HEU benchmark calculations and LEU preliminary calculations for IRR-1
International Nuclear Information System (INIS)
Caner, M.; Shapira, M.; Bettan, M.; Nagler, A.; Gilat, J.
2004-01-01
We performed neutronics calculations for the Soreq Research Reactor, IRR-1. The calculations were done for the purpose of upgrading and benchmarking our codes and methods. The codes used were mainly WIMS-D/4 for cell calculations and the three dimensional diffusion code CITATION for full core calculations. The experimental flux was obtained by gold wire activation methods and compared with our calculated flux profile. The IRR-1 is loaded with highly enriched uranium fuel assemblies, of the plate type. In the framework of preparation for conversion to low enrichment fuel, additional calculations were done assuming the presence of LEU fresh fuel. In these preliminary calculations we investigated the effect on the criticality and flux distributions of the increase of U-238 loading, and the corresponding uranium density.(author)
International Nuclear Information System (INIS)
Sugimura, Naoki; Mori, Masaaki; Hijiya, Masayuki; Ushio, Tadashi; Arakawa, Yasushi
2004-01-01
This paper presents the Hybrid Core Calculation System which is a very rigorous but a practical calculation system applicable to best estimate core design calculations taking advantage of the recent remarkable progress of computers. The basic idea of this system is to generate the correction factors for assembly homogenized cross sections, discontinuity factors, etc. by comparing the CASMO-4 and SIMULATE-3 2-D core calculation results under the consistent calculation condition and then apply them for SIMULATE-3 3-D calculation. The CASMO-4 2-D heterogeneous core calculation is performed for each depletion step with the core conditions previously determined by ordinary SIMULATE-3 core calculation to avoid time consuming iterative calculations searching for the critical boron concentrations while treating the thermal hydraulic feedback. The final SIMULATE-3 3-D calculation using the correction factors is performed with iterative calculations searching for the critical boron concentrations while treating the thermal hydraulic feedback. (author)
National Oceanic and Atmospheric Administration, Department of Commerce — Declination is calculated using the current International Geomagnetic Reference Field (IGRF) model. Declination is calculated using the current World Magnetic Model...
HP-67 calculator programs for thermodynamic data and phase diagram calculations
International Nuclear Information System (INIS)
Brewer, L.
1978-01-01
This report is a supplement to a tabulation of the thermodynamic and phase data for the 100 binary systems of Mo with the elements from H to Lr. The calculations of thermodynamic data and phase equilibria were carried out from 5000 0 K to low temperatures. This report presents the methods of calculation used. The thermodynamics involved is rather straightforward and the reader is referred to any advanced thermodynamic text. The calculations were largely carried out using an HP-65 programmable calculator. In this report, those programs are reformulated for use with the HP-67 calculator; great reduction in the number of programs required to carry out the calculation results
31 CFR 205.27 - How are Interest Calculation Costs calculated?
2010-07-01
... 31 Money and Finance: Treasury 2 2010-07-01 2010-07-01 false How are Interest Calculation Costs calculated? 205.27 Section 205.27 Money and Finance: Treasury Regulations Relating to Money and Finance... this subpart A, other than Interest Calculation Costs, are subject to the procedures and principles of...
Calculating zeros: Non-equilibrium free energy calculations
International Nuclear Information System (INIS)
Oostenbrink, Chris; Gunsteren, Wilfred F. van
2006-01-01
Free energy calculations on three model processes with theoretically known free energy changes have been performed using short simulation times. A comparison between equilibrium (thermodynamic integration) and non-equilibrium (fast growth) methods has been made in order to assess the accuracy and precision of these methods. The three processes have been chosen to represent processes often observed in biomolecular free energy calculations. They involve a redistribution of charges, the creation and annihilation of neutral particles and conformational changes. At very short overall simulation times, the thermodynamic integration approach using discrete steps is most accurate. More importantly, reasonable accuracy can be obtained using this method which seems independent of the overall simulation time. In cases where slow conformational changes play a role, fast growth simulations might have an advantage over discrete thermodynamic integration where sufficient sampling needs to be obtained at every λ-point, but only if the initial conformations do properly represent an equilibrium ensemble. From these three test cases practical lessons can be learned that will be applicable to biomolecular free energy calculations
Energy Technology Data Exchange (ETDEWEB)
Nagao, Yoshiharu [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment
1998-03-01
In material testing reactors like the JMTR (Japan Material Testing Reactor) of 50 MW in Japan Atomic Energy Research Institute, the neutron flux and neutron energy spectra of irradiated samples show complex distributions. It is necessary to assess the neutron flux and neutron energy spectra of an irradiation field by carrying out the nuclear calculation of the core for every operation cycle. In order to advance core calculation, in the JMTR, the application of MCNP to the assessment of core reactivity and neutron flux and spectra has been investigated. In this study, in order to reduce the time for calculation and variance, the comparison of the results of the calculations by the use of K code and fixed source and the use of Weight Window were investigated. As to the calculation method, the modeling of the total JMTR core, the conditions for calculation and the adopted variance reduction technique are explained. The results of calculation are shown. Significant difference was not observed in the results of neutron flux calculations according to the difference of the modeling of fuel region in the calculations by K code and fixed source. The method of assessing the results of neutron flux calculation is described. (K.I.)
International Nuclear Information System (INIS)
Carossi, Jean-Claude
1969-02-01
A CO 2 flowrate calculator has been designed for measuring and recording the gas flow in the loops of Pegase reactor. The analog calculator applies, at every moment, Bernoulli's formula to the values that characterize the carbon dioxide flow through a nozzle. The calculator electronics is described (it includes a sampling calculator and a two-variable function generator), with its amplifiers, triggers, interpolator, multiplier, etc. Calculator operation and setting are presented
National Oceanic and Atmospheric Administration, Department of Commerce — The Magnetic Field Calculator will calculate the total magnetic field, including components (declination, inclination, horizontal intensity, northerly intensity,...
Wisniewski, H.; Gourdain, P.-A.
2017-10-01
APOLLO is an online, Linux based plasma calculator. Users can input variables that correspond to their specific plasma, such as ion and electron densities, temperatures, and external magnetic fields. The system is based on a webserver where a FastCGI protocol computes key plasma parameters including frequencies, lengths, velocities, and dimensionless numbers. FastCGI was chosen to overcome security problems caused by JAVA-based plugins. The FastCGI also speeds up calculations over PHP based systems. APOLLO is built upon the WT library, which turns any web browser into a versatile, fast graphic user interface. All values with units are expressed in SI units except temperature, which is in electron-volts. SI units were chosen over cgs units because of the gradual shift to using SI units within the plasma community. APOLLO is intended to be a fast calculator that also provides the user with the proper equations used to calculate the plasma parameters. This system is intended to be used by undergraduates taking plasma courses as well as graduate students and researchers who need a quick reference calculation.
Global nuclear-structure calculations
International Nuclear Information System (INIS)
Moeller, P.; Nix, J.R.
1990-01-01
The revival of interest in nuclear ground-state octupole deformations that occurred in the 1980's was stimulated by observations in 1980 of particularly large deviations between calculated and experimental masses in the Ra region, in a global calculation of nuclear ground-state masses. By minimizing the total potential energy with respect to octupole shape degrees of freedom in addition to ε 2 and ε 4 used originally, a vastly improved agreement between calculated and experimental masses was obtained. To study the global behavior and interrelationships between other nuclear properties, we calculate nuclear ground-state masses, spins, pairing gaps and Β-decay and half-lives and compare the results to experimental qualities. The calculations are based on the macroscopic-microscopic approach, with the microscopic contributions calculated in a folded-Yukawa single-particle potential
DEFF Research Database (Denmark)
Frederiksen, Morten
2014-01-01
Williamson’s characterisation of calculativeness as inimical to trust contradicts most sociological trust research. However, a similar argument is found within trust phenomenology. This paper re-investigates Williamson’s argument from the perspective of Løgstrup’s phenomenological theory of trust....... Contrary to Williamson, however, Løgstrup’s contention is that trust, not calculativeness, is the default attitude and only when suspicion is awoken does trust falter. The paper argues that while Williamson’s distinction between calculativeness and trust is supported by phenomenology, the analysis needs...... to take actual subjective experience into consideration. It points out that, first, Løgstrup places trust alongside calculativeness as a different mode of engaging in social interaction, rather conceiving of trust as a state or the outcome of a decision-making process. Secondly, the analysis must take...
International Nuclear Information System (INIS)
Scholtyssek, W.
1995-01-01
In the first phase of a benchmark comparison, the CONTAIN code was used to calculate an assumed EPR accident 'medium-sized leak in the cold leg', especially for the first two days after initiation of the accident. The results for global characteristics compare well with those of FIPLOC, MELCOR and WAVCO calculations, if the same materials data are used as input. However, significant differences show up for local quantities such as flows through leakages. (orig.)
Electrical installation calculations basic
Kitcher, Christopher
2013-01-01
All the essential calculations required for basic electrical installation workThe Electrical Installation Calculations series has proved an invaluable reference for over forty years, for both apprentices and professional electrical installation engineers alike. The book provides a step-by-step guide to the successful application of electrical installation calculations required in day-to-day electrical engineering practice. A step-by-step guide to everyday calculations used on the job An essential aid to the City & Guilds certificates at Levels 2 and 3Fo
International Nuclear Information System (INIS)
Richet, Y.; Jacquet, O.; Bay, X.
2005-01-01
The accuracy of an Iterative Monte Carlo calculation requires the convergence of the simulation output process. The present paper deals with a post processing algorithm to suppress the transient due to initialization applied on criticality calculations. It should be noticed that this initial transient suppression aims only at obtaining a stationary output series, then the convergence of the calculation needs to be guaranteed independently. The transient suppression algorithm consists in a repeated truncation of the first observations of the output process. The truncation of the first observations is performed as long as a steadiness test based on Brownian bridge theory is negative. This transient suppression method was previously tuned for a simplified model of criticality calculations, although this paper focuses on the efficiency on real criticality calculations. The efficiency test is based on four benchmarks with strong source convergence problems: 1) a checkerboard storage of fuel assemblies, 2) a pin cell array with irradiated fuel, 3) 3 one-dimensional thick slabs, and 4) an array of interacting fuel spheres. It appears that the transient suppression method needs to be more widely validated on real criticality calculations before any blind using as a post processing in criticality codes
Electrical installation calculations advanced
Kitcher, Christopher
2013-01-01
All the essential calculations required for advanced electrical installation workThe Electrical Installation Calculations series has proved an invaluable reference for over forty years, for both apprentices and professional electrical installation engineers alike. The book provides a step-by-step guide to the successful application of electrical installation calculations required in day-to-day electrical engineering practiceA step-by-step guide to everyday calculations used on the job An essential aid to the City & Guilds certificates at Levels 2 and 3For apprentices and electrical installatio
Dose calculation for electrons
International Nuclear Information System (INIS)
Hirayama, Hideo
1995-01-01
The joint working group of ICRP/ICRU is advancing the works of reviewing the ICRP publication 51 by investigating the data related to radiation protection. In order to introduce the 1990 recommendation, it has been demanded to carry out calculation for neutrons, photons and electrons. As for electrons, EURADOS WG4 (Numerical Dosimetry) rearranged the data to be calculated at the meeting held in PTB Braunschweig in June, 1992, and the question and request were presented by Dr. J.L. Chartier, the responsible person, to the researchers who are likely to undertake electron transport Monte Carlo calculation. The author also has carried out the requested calculation as it was the good chance to do the mutual comparison among various computation codes regarding electron transport calculation. The content that the WG requested to calculate was the absorbed dose at depth d mm when parallel electron beam enters at angle α into flat plate phantoms of PMMA, water and ICRU4-element tissue, which were placed in vacuum. The calculation was carried out by the versatile electron-photon shower computation Monte Carlo code, EGS4. As the results, depth dose curves and the dependence of absorbed dose on electron energy, incident angle and material are reported. The subjects to be investigated are pointed out. (K.I.)
Waste Package Lifting Calculation
International Nuclear Information System (INIS)
H. Marr
2000-01-01
The objective of this calculation is to evaluate the structural response of the waste package during the horizontal and vertical lifting operations in order to support the waste package lifting feature design. The scope of this calculation includes the evaluation of the 21 PWR UCF (pressurized water reactor uncanistered fuel) waste package, naval waste package, 5 DHLW/DOE SNF (defense high-level waste/Department of Energy spent nuclear fuel)--short waste package, and 44 BWR (boiling water reactor) UCF waste package. Procedure AP-3.12Q, Revision 0, ICN 0, calculations, is used to develop and document this calculation
International Nuclear Information System (INIS)
Varga, E.; Visi, Gy.
1982-01-01
Mathematical programmes are given for calculator type PTK 1072 (Hungarian made), to make easier the lengthy calculations applied in examinations in laboratories for control of radioactive materials in food. Basic consideration of making a programme, the method, the mathematical formulae, the variations of calculation and control of program are shown by examples. Making programmes for calculators of other types, too, can be facilitated by adapting the basic consideration. (author)
Invert Effective Thermal Conductivity Calculation
International Nuclear Information System (INIS)
M.J. Anderson; H.M. Wade; T.L. Mitchell
2000-01-01
The objective of this calculation is to evaluate the temperature-dependent effective thermal conductivities of a repository-emplaced invert steel set and surrounding ballast material. The scope of this calculation analyzes a ballast-material thermal conductivity range of 0.10 to 0.70 W/m · K, a transverse beam spacing range of 0.75 to 1.50 meters, and beam compositions of A 516 carbon steel and plain carbon steel. Results from this calculation are intended to support calculations that identify waste package and repository thermal characteristics for Site Recommendation (SR). This calculation was developed by Waste Package Department (WPD) under Office of Civilian Radioactive Waste Management (OCRWM) procedure AP-3.12Q, Revision 1, ICN 0, Calculations
IRIS core criticality calculations
International Nuclear Information System (INIS)
Jecmenica, R.; Trontl, K.; Pevec, D.; Grgic, D.
2003-01-01
Three-dimensional Monte Carlo computer code KENO-VI of CSAS26 sequence of SCALE-4.4 code system was applied for pin-by-pin calculations of the effective multiplication factor for the first cycle IRIS reactor core. The effective multiplication factors obtained by the above mentioned Monte Carlo calculations using 27-group ENDF/B-IV library and 238-group ENDF/B-V library have been compared with the effective multiplication factors achieved by HELIOS/NESTLE, CASMO/SIMULATE, and modified CORD-2 nodal calculations. The results of Monte Carlo calculations are found to be in good agreement with the results obtained by the nodal codes. The discrepancies in effective multiplication factor are typically within 1%. (author)
Parameters calculation of a shielding experiment and evaluation of calculation methodology
International Nuclear Information System (INIS)
Gavazza, S.; Otto, A.C.; Gomes, I.C.; Maiorino, J.R.
1986-01-01
In this text is carried out the evaluation of radiation transport methodology, comparying the calculated reactions and dose rates, for neutrons and gamma-rays, with the experimental measurements obtained on iron shield, irradiated in YAYOI reactor. Were employed the ENDF/B-IV and VITAMIN-C libraries and the AMPX-II modular system for generation of cross sections, collapsed by the ANISN code. The transport calculation were made by using the DOT 3.5 code, adjusting the spectrum of the iron shield boundary source to the reactions and dose rates, measured at the beginning of shield. The distributions calculated for neutrons and gamma-rays, on iron shield, presented coherence with the experimental measurements. (Author) [pt
CALCULATION OF LASER CUTTING COSTS
Bogdan Nedic; Milan Eric; Marijana Aleksijevic
2016-01-01
The paper presents description methods of metal cutting and calculation of treatment costs based on model that is developed on Faculty of mechanical engineering in Kragujevac. Based on systematization and analysis of large number of calculation models of cutting with unconventional methods, mathematical model is derived, which is used for creating a software for calculation costs of metal cutting. Software solution enables resolving the problem of calculating the cost of laser cutting, compar...
Radioactive cloud dose calculations
International Nuclear Information System (INIS)
Healy, J.W.
1984-01-01
Radiological dosage principles, as well as methods for calculating external and internal dose rates, following dispersion and deposition of radioactive materials in the atmosphere are described. Emphasis has been placed on analytical solutions that are appropriate for hand calculations. In addition, the methods for calculating dose rates from ingestion are discussed. A brief description of several computer programs are included for information on radionuclides. There has been no attempt to be comprehensive, and only a sampling of programs has been selected to illustrate the variety available
Nuclear data library in design calculation
International Nuclear Information System (INIS)
Hirano, Go; Kosaka, Shinya
2006-01-01
In core design calculation, nuclear data takes part as multi group cross section library during the assembly calculation, which is the first stage of a core design calculation. This report summarizes the multi group cross section libraries used in assembly calculations and also presents the methods adopted for resonance and assembly calculation. (author)
Closure and Sealing Design Calculation
International Nuclear Information System (INIS)
T. Lahnalampi; J. Case
2005-01-01
The purpose of the ''Closure and Sealing Design Calculation'' is to illustrate closure and sealing methods for sealing shafts, ramps, and identify boreholes that require sealing in order to limit the potential of water infiltration. In addition, this calculation will provide a description of the magma that can reduce the consequences of an igneous event intersecting the repository. This calculation will also include a listing of the project requirements related to closure and sealing. The scope of this calculation is to: summarize applicable project requirements and codes relating to backfilling nonemplacement openings, removal of uncommitted materials from the subsurface, installation of drip shields, and erecting monuments; compile an inventory of boreholes that are found in the area of the subsurface repository; describe the magma bulkhead feature and location; and include figures for the proposed shaft and ramp seals. The objective of this calculation is to: categorize the boreholes for sealing by depth and proximity to the subsurface repository; develop drawing figures which show the location and geometry for the magma bulkhead; include the shaft seal figures and a proposed construction sequence; and include the ramp seal figure and a proposed construction sequence. The intent of this closure and sealing calculation is to support the License Application by providing a description of the closure and sealing methods for the Safety Analysis Report. The closure and sealing calculation will also provide input for Post Closure Activities by describing the location of the magma bulkhead. This calculation is limited to describing the final configuration of the sealing and backfill systems for the underground area. The methods and procedures used to place the backfill and remove uncommitted materials (such as concrete) from the repository and detailed design of the magma bulkhead will be the subject of separate analyses or calculations. Post-closure monitoring will not
International Nuclear Information System (INIS)
Honeck, H.C.
1984-01-01
1 - Description of problem or function: HAMMER performs infinite lattice, one-dimensional cell multigroup calculations, followed (optionally) by one-dimensional, few-group, multi-region reactor calculations with neutron balance edits. 2 - Method of solution: Infinite lattice parameters are calculated by means of multigroup transport theory, composite reactor parameters by few-group diffusion theory. 3 - Restrictions on the complexity of the problem: - Cell calculations - maxima of: 30 thermal groups; 54 epithermal groups; 20 space points; 20 regions; 18 isotopes; 10 mixtures; 3 thermal up-scattering mixtures; 200 resonances per group; no overlap or interference; single level only. - Reactor calculations - maxima of : 40 regions; 40 mixtures; 250 space points; 4 groups
Uncertainty calculations made easier
International Nuclear Information System (INIS)
Hogenbirk, A.
1994-07-01
The results are presented of a neutron cross section sensitivity/uncertainty analysis performed in a complicated 2D model of the NET shielding blanket design inside the ITER torus design, surrounded by the cryostat/biological shield as planned for ITER. The calculations were performed with a code system developed at ECN Petten, with which sensitivity/uncertainty calculations become relatively simple. In order to check the deterministic neutron transport calculations (performed with DORT), calculations were also performed with the Monte Carlo code MCNP. Care was taken to model the 2.0 cm wide gaps between two blanket segments, as the neutron flux behind the vacuum vessel is largely determined by neutrons streaming through these gaps. The resulting neutron flux spectra are in excellent agreement up to the end of the cryostat. It is noted, that at this position the attenuation of the neutron flux is about 1 l orders of magnitude. The uncertainty in the energy integrated flux at the beginning of the vacuum vessel and at the beginning of the cryostat was determined in the calculations. The uncertainty appears to be strongly dependent on the exact geometry: if the gaps are filled with stainless steel, the neutron spectrum changes strongly, which results in an uncertainty of 70% in the energy integrated flux at the beginning of the cryostat in the no-gap-geometry, compared to an uncertainty of only 5% in the gap-geometry. Therefore, it is essential to take into account the exact geometry in sensitivity/uncertainty calculations. Furthermore, this study shows that an improvement of the covariance data is urgently needed in order to obtain reliable estimates of the uncertainties in response parameters in neutron transport calculations. (orig./GL)
MCNP and OMEGA criticality calculations
International Nuclear Information System (INIS)
Seifert, E.
1998-04-01
The reliability of OMEGA criticality calculations is shown by a comparison with calculations by the validated and widely used Monte Carlo code MCNP. The criticality of 16 assemblies with uranium as fissionable is calculated with the codes MCNP (Version 4A, ENDF/B-V cross sections), MCNP (Version 4B, ENDF/B-VI cross sections), and OMEGA. Identical calculation models are used for the three codes. The results are compared mutually and with the experimental criticality of the assemblies. (orig.)
Heterogeneous Calculation of {epsilon}
Energy Technology Data Exchange (ETDEWEB)
Jonsson, Alf
1961-02-15
A heterogeneous method of calculating the fast fission factor given by Naudet has been applied to the Carlvik - Pershagen definition of {epsilon}. An exact calculation of the collision probabilities is included in the programme developed for the Ferranti - Mercury computer.
Heterogeneous Calculation of ε
International Nuclear Information System (INIS)
Jonsson, Alf
1961-02-01
A heterogeneous method of calculating the fast fission factor given by Naudet has been applied to the Carlvik - Pershagen definition of ε. An exact calculation of the collision probabilities is included in the programme developed for the Ferranti - Mercury computer
Uneconomical top calculation method
International Nuclear Information System (INIS)
De Noord, M.; Vanm Sambeek, E.J.W.
2003-08-01
The methodology used to calculate the financial gap of renewable electricity sources and technologies is described. This methodology is used for calculating the production subsidy levels (MEP subsidies) for new renewable electricity projects in 2004 and 2005 in the Netherlands [nl
Calculation of Rydberg interaction potentials
DEFF Research Database (Denmark)
Weber, Sebastian; Tresp, Christoph; Menke, Henri
2017-01-01
for calculating the required electric multipole moments and the inclusion of electromagnetic fields with arbitrary direction. We focus specifically on symmetry arguments and selection rules, which greatly reduce the size of the Hamiltonian matrix, enabling the direct diagonalization of the Hamiltonian up...... to higher multipole orders on a desktop computer. Finally, we present example calculations showing the relevance of the full interaction calculation to current experiments. Our software for calculating Rydberg potentials including all features discussed in this tutorial is available as open source....
Salgado, C A; Salgado, Carlos A.; Wiedemann, Urs Achim
2003-01-01
We calculate the probability (``quenching weight'') that a hard parton radiates an additional energy fraction due to scattering in spatially extended QCD matter. This study is based on an exact treatment of finite in-medium path length, it includes the case of a dynamically expanding medium, and it extends to the angular dependence of the medium-induced gluon radiation pattern. All calculations are done in the multiple soft scattering approximation (Baier-Dokshitzer-Mueller-Peign\\'e-Schiff--Zakharov ``BDMPS-Z''-formalism) and in the single hard scattering approximation (N=1 opacity approximation). By comparison, we establish a simple relation between transport coefficient, Debye screening mass and opacity, for which both approximations lead to comparable results. Together with this paper, a CPU-inexpensive numerical subroutine for calculating quenching weights is provided electronically. To illustrate its applications, we discuss the suppression of hadronic transverse momentum spectra in nucleus-nucleus colli...
Transfer Area Mechanical Handling Calculation
International Nuclear Information System (INIS)
Dianda, B.
2004-01-01
This calculation is intended to support the License Application (LA) submittal of December 2004, in accordance with the directive given by DOE correspondence received on the 27th of January 2004 entitled: ''Authorization for Bechtel SAX Company L.L. C. to Include a Bare Fuel Handling Facility and Increased Aging Capacity in the License Application, Contract Number DE-AC--28-01R W12101'' (Arthur, W.J., I11 2004). This correspondence was appended by further Correspondence received on the 19th of February 2004 entitled: ''Technical Direction to Bechtel SAIC Company L.L. C. for Surface Facility Improvements, Contract Number DE-AC--28-OIRW12101; TDL No. 04-024'' (BSC 2004a). These documents give the authorization for a Fuel Handling Facility to be included in the baseline. The purpose of this calculation is to establish preliminary bounding equipment envelopes and weights for the Fuel Handling Facility (FHF) transfer areas equipment. This calculation provides preliminary information only to support development of facility layouts and preliminary load calculations. The limitations of this preliminary calculation lie within the assumptions of section 5 , as this calculation is part of an evolutionary design process. It is intended that this calculation is superseded as the design advances to reflect information necessary to support License Application. The design choices outlined within this calculation represent a demonstration of feasibility and may or may not be included in the completed design. This calculation provides preliminary weight, dimensional envelope, and equipment position in building for the purposes of defining interface variables. This calculation identifies and sizes major equipment and assemblies that dictate overall equipment dimensions and facility interfaces. Sizing of components is based on the selection of commercially available products, where applicable. This is not a specific recommendation for the future use of these components or their
CALCULATION OF LASER CUTTING COSTS
Directory of Open Access Journals (Sweden)
Bogdan Nedic
2016-09-01
Full Text Available The paper presents description methods of metal cutting and calculation of treatment costs based on model that is developed on Faculty of mechanical engineering in Kragujevac. Based on systematization and analysis of large number of calculation models of cutting with unconventional methods, mathematical model is derived, which is used for creating a software for calculation costs of metal cutting. Software solution enables resolving the problem of calculating the cost of laser cutting, comparison' of costs made by other unconventional methods and provides documentation that consists of reports on estimated costs.
Reactor core performance calculating device
International Nuclear Information System (INIS)
Tominaga, Kenji; Bando, Masaru; Sano, Hiroki; Maruyama, Hiromi.
1995-01-01
The device of the present invention can calculate a power distribution efficiently at high speed by a plurality of calculation means while taking an amount of the reactor state into consideration. Namely, an input device takes data from a measuring device for the amount of the reactor core state such as a large number of neutron detectors disposed in the reactor core for monitoring the reactor state during operation. An input data distribution device comprises a state recognition section and a data distribution section. The state recognition section recognizes the kind and amount of the inputted data and information of the calculation means. The data distribution section analyzes the characteristic of the inputted data, divides them into a several groups, allocates them to each of the calculation means for the purpose of calculating the reactor core performance efficiently at high speed based on the information from the state recognition section. A plurality of the calculation means calculate power distribution of each of regions based on the allocated inputted data, to determine the power distribution of the entire reactor core. As a result, the reactor core can be evaluated at high accuracy and at high speed irrespective of the whole reactor core or partial region. (I.S.)
International Nuclear Information System (INIS)
Thieke, Christian; Nill, Simeon; Oelfke, Uwe; Bortfeld, Thomas
2002-01-01
In inverse planning for intensity-modulated radiotherapy, the dose calculation is a crucial element limiting both the maximum achievable plan quality and the speed of the optimization process. One way to integrate accurate dose calculation algorithms into inverse planning is to precalculate the dose contribution of each beam element to each voxel for unit fluence. These precalculated values are stored in a big dose calculation matrix. Then the dose calculation during the iterative optimization process consists merely of matrix look-up and multiplication with the actual fluence values. However, because the dose calculation matrix can become very large, this ansatz requires a lot of computer memory and is still very time consuming, making it not practical for clinical routine without further modifications. In this work we present a new method to significantly reduce the number of entries in the dose calculation matrix. The method utilizes the fact that a photon pencil beam has a rapid radial dose falloff, and has very small dose values for the most part. In this low-dose part of the pencil beam, the dose contribution to a voxel is only integrated into the dose calculation matrix with a certain probability. Normalization with the reciprocal of this probability preserves the total energy, even though many matrix elements are omitted. Three probability distributions were tested to find the most accurate one for a given memory size. The sampling method is compared with the use of a fully filled matrix and with the well-known method of just cutting off the pencil beam at a certain lateral distance. A clinical example of a head and neck case is presented. It turns out that a sampled dose calculation matrix with only 1/3 of the entries of the fully filled matrix does not sacrifice the quality of the resulting plans, whereby the cutoff method results in a suboptimal treatment plan
Shielding calculational system for plutonium
International Nuclear Information System (INIS)
Zimmerman, M.G.; Thomsen, D.H.
1975-08-01
A computer calculational system has been developed and assembled specifically for calculating dose rates in AEC plutonium fabrication facilities. The system consists of two computer codes and all nuclear data necessary for calculation of neutron and gamma dose rates from plutonium. The codes include the multigroup version of the Battelle Monte Carlo code for solution of general neutron and gamma shielding problems and the PUSHLD code for solution of shielding problems where low energy gamma and x-rays are important. The nuclear data consists of built in neutron and gamma yields and spectra for various plutonium compounds, an automatic calculation of age effects and all cross-sections commonly used. Experimental correlations have been performed to verify portions of the calculational system. (23 tables, 7 figs, 16 refs) (U.S.)
Institute of Scientific and Technical Information of China (English)
无
2002-01-01
The improved form of calculation formula for the activities of the components in binary liquids and solid alloys has been derived based on the free volume theory considering excess entropy and Miedema's model for calculating the formation heat of binary alloys. A calculation method of excess thermodynamic functions for binary alloys, the formulas of integral molar excess properties and partial molar excess properties for solid ordered or disordered binary alloys have been developed. The calculated results are in good agreement with the experimental values.
Energy Technology Data Exchange (ETDEWEB)
Brockway, D.; Soran, P.; Whalen, P.
1985-01-01
A Monte Carlo algorithm to efficiently calculate static alpha eigenvalues, N = ne/sup ..cap alpha..t/, for supercritical systems has been developed and tested. A direct Monte Carlo approach to calculating a static alpha is to simply follow the buildup in time of neutrons in a supercritical system and evaluate the logarithmic derivative of the neutron population with respect to time. This procedure is expensive, and the solution is very noisy and almost useless for a system near critical. The modified approach is to convert the time-dependent problem to a static ..cap alpha../sup -/eigenvalue problem and regress ..cap alpha.. on solutions of a/sup -/ k/sup -/eigenvalue problem. In practice, this procedure is much more efficient than the direct calculation, and produces much more accurate results. Because the Monte Carlo codes are intrinsically three-dimensional and use elaborate continuous-energy cross sections, this technique is now used as a standard for evaluating other calculational techniques in odd geometries or with group cross sections.
International Nuclear Information System (INIS)
Govoni, Marco; Argonne National Lab., Argonne, IL; Galli, Giulia; Argonne National Lab., Argonne, IL
2015-01-01
We present GW calculations of molecules, ordered and disordered solids and interfaces, which employ an efficient contour deformation technique for frequency integration and do not require the explicit evaluation of virtual electronic states nor the inversion of dielectric matrices. We also present a parallel implementation of the algorithm, which takes advantage of separable expressions of both the single particle Green's function and the screened Coulomb interaction. The method can be used starting from density functional theory calculations performed with semilocal or hybrid functionals. The newly developed technique was applied to GW calculations of systems of unprecedented size, including water/semiconductor interfaces with thousands of electrons
Transient anisotropic magnetic field calculation
International Nuclear Information System (INIS)
Jesenik, Marko; Gorican, Viktor; Trlep, Mladen; Hamler, Anton; Stumberger, Bojan
2006-01-01
For anisotropic magnetic material, nonlinear magnetic characteristics of the material are described with magnetization curves for different magnetization directions. The paper presents transient finite element calculation of the magnetic field in the anisotropic magnetic material based on the measured magnetization curves for different magnetization directions. For the verification of the calculation method some results of the calculation are compared with the measurement
International Nuclear Information System (INIS)
Simonkova, J.
1988-01-01
The problems are summed up of the dynamic calculation of cooling towers with forced and natural air draft. The quantities and relations are given characterizing the simultaneous exchange of momentum, heat and mass in evaporative water cooling by atmospheric air in the packings of cooling towers. The method of solution is clarified in the calculation of evaporation criteria and thermal characteristics of countercurrent and cross current cooling systems. The procedure is demonstrated of the calculation of cooling towers, and correction curves and the effect assessed of the operating mode at constant air number or constant outlet air volume flow on their course in ventilator cooling towers. In cooling towers with the natural air draft the flow unevenness is assessed of water and air relative to its effect on the resulting cooling efficiency of the towers. The calculation is demonstrated of thermal and resistance response curves and cooling curves of hydraulically unevenly loaded towers owing to the water flow rate parameter graded radially by 20% along the cross-section of the packing. Flow rate unevenness of air due to wind impact on the outlet air flow from the tower significantly affects the temperatures of cooled water in natural air draft cooling towers of a design with lower demands on aerodynamics, as early as at wind velocity of 2 m.s -1 as was demonstrated on a concrete example. (author). 11 figs., 10 refs
International Nuclear Information System (INIS)
Trkov, A.; Ravnik, M.; Zeleznik, N.
1992-01-01
Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [sl
Exact-exchange-based quasiparticle calculations
International Nuclear Information System (INIS)
Aulbur, Wilfried G.; Staedele, Martin; Goerling, Andreas
2000-01-01
One-particle wave functions and energies from Kohn-Sham calculations with the exact local Kohn-Sham exchange and the local density approximation (LDA) correlation potential [EXX(c)] are used as input for quasiparticle calculations in the GW approximation (GWA) for eight semiconductors. Quasiparticle corrections to EXX(c) band gaps are small when EXX(c) band gaps are close to experiment. In the case of diamond, quasiparticle calculations are essential to remedy a 0.7 eV underestimate of the experimental band gap within EXX(c). The accuracy of EXX(c)-based GWA calculations for the determination of band gaps is as good as the accuracy of LDA-based GWA calculations. For the lowest valence band width a qualitatively different behavior is observed for medium- and wide-gap materials. The valence band width of medium- (wide-) gap materials is reduced (increased) in EXX(c) compared to the LDA. Quasiparticle corrections lead to a further reduction (increase). As a consequence, EXX(c)-based quasiparticle calculations give valence band widths that are generally 1-2 eV smaller (larger) than experiment for medium- (wide-) gap materials. (c) 2000 The American Physical Society
Non-perturbative background field calculations
International Nuclear Information System (INIS)
Stephens, C.R.; Department of Physics, University of Utah, Salt Lake City, Utah 84112)
1988-01-01
New methods are developed for calculating one loop functional determinants in quantum field theory. Instead of relying on a calculation of all the eigenvalues of the small fluctuation equation, these techniques exploit the ability of the proper time formalism to reformulate an infinite dimensional field theoretic problem into a finite dimensional covariant quantum mechanical analog, thereby allowing powerful tools such as the method of Jacobi fields to be used advantageously in a field theory setting. More generally the methods developed herein should be extremely valuable when calculating quantum processes in non-constant background fields, offering a utilitarian alternative to the two standard methods of calculation: perturbation theory in the background field or taking the background field into account exactly. The formalism developed also allows for the approximate calculation of covariances of partial differential equations from a knowledge of the solutions of a homogeneous ordinary differential equation. copyright 1988 Academic Press, Inc
Cleary, David A.
2014-01-01
The usefulness of the JANAF tables is demonstrated with specific equilibrium calculations. An emphasis is placed on the nature of standard chemical potential calculations. Also, the use of the JANAF tables for calculating partition functions is examined. In the partition function calculations, the importance of the zero of energy is highlighted.
Some calculator programs for particle physics
International Nuclear Information System (INIS)
Wohl, C.G.
1982-01-01
Seven calculator programs that do simple chores that arise in elementary particle physics are given. LEGENDRE evaluates the Legendre polynomial series Σa/sub n/P/sub n/(x) at a series of values of x. ASSOCIATED LEGENDRE evaluates the first-associated Legendre polynomial series Σb/sub n/P/sub n/ 1 (x) at a series of values of x. CONFIDENCE calculates confidence levels for chi 2 , Gaussian, or Poisson probability distributions. TWO BODY calculates the c.m. energy, the initial- and final-state c.m. momenta, and the extreme values of t and u for a 2-body reaction. ELLIPSE calculates coordinates of points for drawing an ellipse plot showing the kinematics of a 2-body reaction or decay. DALITZ RECTANGULAR calculates coordinates of points on the boundary of a rectangular Dalitz plot. DALITZ TRIANGULAR calculates coordinates of points on the boundary of a triangular Dalitz plot. There are short versions of CONFIDENCE (EVEN N and POISSON) that calculate confidence levels for the even-degree-of-freedom-chi 2 and the Poisson cases, and there is a short version of TWO BODY (CM) that calculates just the c.m. energy and initial-state momentum. The programs are written for the HP-97 calculator
Two-dimensional sensitivity calculation code: SENSETWO
International Nuclear Information System (INIS)
Yamauchi, Michinori; Nakayama, Mitsuo; Minami, Kazuyoshi; Seki, Yasushi; Iida, Hiromasa.
1979-05-01
A SENSETWO code for the calculation of cross section sensitivities with a two-dimensional model has been developed, on the basis of first order perturbation theory. It uses forward neutron and/or gamma-ray fluxes and adjoint fluxes obtained by two-dimensional discrete ordinates code TWOTRAN-II. The data and informations of cross sections, geometry, nuclide density, response functions, etc. are transmitted to SENSETWO by the dump magnetic tape made in TWOTRAN calculations. The required input for SENSETWO calculations is thus very simple. The SENSETWO yields as printed output the cross section sensitivities for each coarse mesh zone and for each energy group, as well as the plotted output of sensitivity profiles specified by the input. A special feature of the code is that it also calculates the reaction rate with the response function used as the adjoint source in TWOTRAN adjoint calculation and the calculated forward flux from the TWOTRAN forward calculation. (author)
Infinite slab-shield dose calculations
International Nuclear Information System (INIS)
Russell, G.J.
1989-01-01
I calculated neutron and gamma-ray equivalent doses leaking through a variety of infinite (laminate) slab-shields. In the shield computations, I used, as the incident neutron spectrum, the leakage spectrum (<20 MeV) calculated for the LANSCE tungsten production target at 90 degree to the target axis. The shield thickness was fixed at 60 cm. The results of the shield calculations show a minimum in the total leakage equivalent dose if the shield is 40-45 cm of iron followed by 20-15 cm of borated (5% B) polyethylene. High-performance shields can be attained by using multiple laminations. The calculated dose at the shield surface is very dependent on shield material. 4 refs., 4 figs., 1 tab
Calculation of Rydberg interaction potentials
International Nuclear Information System (INIS)
Weber, Sebastian; Büchler, Hans Peter; Tresp, Christoph; Urvoy, Alban; Hofferberth, Sebastian; Menke, Henri; Firstenberg, Ofer
2017-01-01
The strong interaction between individual Rydberg atoms provides a powerful tool exploited in an ever-growing range of applications in quantum information science, quantum simulation and ultracold chemistry. One hallmark of the Rydberg interaction is that both its strength and angular dependence can be fine-tuned with great flexibility by choosing appropriate Rydberg states and applying external electric and magnetic fields. More and more experiments are probing this interaction at short atomic distances or with such high precision that perturbative calculations as well as restrictions to the leading dipole–dipole interaction term are no longer sufficient. In this tutorial, we review all relevant aspects of the full calculation of Rydberg interaction potentials. We discuss the derivation of the interaction Hamiltonian from the electrostatic multipole expansion, numerical and analytical methods for calculating the required electric multipole moments and the inclusion of electromagnetic fields with arbitrary direction. We focus specifically on symmetry arguments and selection rules, which greatly reduce the size of the Hamiltonian matrix, enabling the direct diagonalization of the Hamiltonian up to higher multipole orders on a desktop computer. Finally, we present example calculations showing the relevance of the full interaction calculation to current experiments. Our software for calculating Rydberg potentials including all features discussed in this tutorial is available as open source. (tutorial)
Energy Technology Data Exchange (ETDEWEB)
NONE
1963-07-01
This note constitutes the first edition of a Handbook for the calculation of reactor protections. This handbook makes it possible to calculate simply the different neutron and gamma fluxes and consequently, to fix the minimum quantities of materials necessary under general safety conditions both for the personnel and for the installations. It contains a certain amount of nuclear data, calculation methods, and constants corresponding to the present state of our knowledge. (authors) [French] Cette note constitue la premiere edition du 'Formulaire sur le calcul de la protection des reacteurs'. Ce formulaire permet de calculer de facon simple les difterents flux de neutrons et de gamma et, par suite, de fixer les quantites minima de materiaux a utiliser pour que les conditions generales de securite soient respectees, tant pour le personnel que pour les installations. Il contient un certain nombre de donnees nucleaires, de methodes de calcul et de constantes correspondant a l'etat actuel de nos connaissances. (auteurs)
Radiation damage calculations for compound materials
International Nuclear Information System (INIS)
Greenwood, L.R.
1990-01-01
This paper reports on the SPECOMP computer code, developed to calculate neutron-induced displacement damage cross sections for compound materials such as alloys, insulators, and ceramic tritium breeders for fusion reactors. These new calculations rely on recoil atom energy distributions previously computed with the DISCS computer code, the results of which are stored in SPECTER computer code master libraries. All reaction channels were considered in the DISCS calculations and the neutron cross sections were taken from ENDF/B-V. Compound damage calculations with SPECOMP thus do not need to perform any recoil atom calculations and consequently need no access to ENDF or other neutron data bases. The calculations proceed by determining secondary displacements for each combination of recoil atom and matrix atom using the Lindhard partition of the recoil energy to establish the damage energy
International Nuclear Information System (INIS)
Falero, B.; Bueno, P.; Chaves, M. A.; Ordiales, J. M.; Villafana, O.; Gonzalez, M. J.
2013-01-01
The aim of this study was to develop a software application that performs calculation shields in radiology room depending on the type of equipment. The calculation will be done by selecting the user, the method proposed in the Guide 5.11, the Report 144 and 147 and also for the methodology given by the Portuguese Health Ministry. (Author)
Calculation of sound propagation in fibrous materials
DEFF Research Database (Denmark)
Tarnow, Viggo
1996-01-01
Calculations of attenuation and velocity of audible sound waves in glass wools are presented. The calculations use only the diameters of fibres and the mass density of glass wools as parameters. The calculations are compared with measurements.......Calculations of attenuation and velocity of audible sound waves in glass wools are presented. The calculations use only the diameters of fibres and the mass density of glass wools as parameters. The calculations are compared with measurements....
Mordred: a molecular descriptor calculator.
Moriwaki, Hirotomo; Tian, Yu-Shi; Kawashita, Norihito; Takagi, Tatsuya
2018-02-06
Molecular descriptors are widely employed to present molecular characteristics in cheminformatics. Various molecular-descriptor-calculation software programs have been developed. However, users of those programs must contend with several issues, including software bugs, insufficient update frequencies, and software licensing constraints. To address these issues, we propose Mordred, a developed descriptor-calculation software application that can calculate more than 1800 two- and three-dimensional descriptors. It is freely available via GitHub. Mordred can be easily installed and used in the command line interface, as a web application, or as a high-flexibility Python package on all major platforms (Windows, Linux, and macOS). Performance benchmark results show that Mordred is at least twice as fast as the well-known PaDEL-Descriptor and it can calculate descriptors for large molecules, which cannot be accomplished by other software. Owing to its good performance, convenience, number of descriptors, and a lax licensing constraint, Mordred is a promising choice of molecular descriptor calculation software that can be utilized for cheminformatics studies, such as those on quantitative structure-property relationships.
International Nuclear Information System (INIS)
Andrade Lima, F.R. de
1986-01-01
The sensitivity of non linear responses associated with physical quantities governed by non linear differential systems can be studied using perturbation theory. The equivalence and formal differences between the differential and GPT formalisms are shown and both are used for sensitivity calculations of transient problems in a typical PWR coolant channel. The results obtained are encouraging with respect to the potential of the method for thermalhydraulics calculations normally performed for reactor design and safety analysis. (Author) [pt
Program for TI programmable 59 calculator for calculation of 3H concentration of water samples
International Nuclear Information System (INIS)
Hussain, S.D.; Asghar, G.
1982-09-01
A program has been developed for TI Programmable 59 Calculator of Texas Instruments Inc. to calculate from the observed parameters such as count rate etc. the 3 H (tritium) concentration of water samples processed with/without prior electrolytic enrichment. Procedure to use the program has been described in detail. A brief description of the laboratory treatment of samples and the mathematical equations used in the calculations have been given. (orig./A.B.)
Rotor calculations for neutron spectroscopy; Calculs des rotors de spectrometres a neutrons
Energy Technology Data Exchange (ETDEWEB)
Gobert, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires
1968-07-01
The determination of stress in a rotating disk plane of symmetry normal to the axis of rotation has been studied by a number of investigators. In a recent paper Reich gives an operating process for an analytical solution in an asymmetric rotating disk. In the report we give the calculation of finite difference stress solutions applicable to the two rotating disks. The equations are then programmed for the 360.75 computer by Fortran methods concerning the rotors of choppers. (author) [French] La determination des contraintes dans les disques symetriques, en rotation a ete etudiee par de nombreux auteurs. Dans un recent rapport, Reich donne une solution pour le calcul des disques asymetriques. Ce rapport concerne l'application du calcul des contraintes par differences finies aux deux types de rotors. Les equations ecrites en langage Fortran pour l'ordinateur 360.75 concerne les rotors de choppers. (auteur)
Final disposal room structural response calculations
International Nuclear Information System (INIS)
Stone, C.M.
1997-08-01
Finite element calculations have been performed to determine the structural response of waste-filled disposal rooms at the WIPP for a period of 10,000 years after emplacement of the waste. The calculations were performed to generate the porosity surface data for the final set of compliance calculations. The most recent reference data for the stratigraphy, waste characterization, gas generation potential, and nonlinear material response have been brought together for this final set of calculations
Shielding calculations using FLUKA
International Nuclear Information System (INIS)
Yamaguchi, Chiri; Tesch, K.; Dinter, H.
1988-06-01
The dose equivalent on the surface of concrete shielding has been calculated using the Monte Carlo code FLUKA86 for incident proton energies from 10 to 800 GeV. The results have been compared with some simple equations. The value of the angular dependent parameter in Moyer's equation has been calculated from the locations where the values of the maximum dose equivalent occur. (author)
ENRAF gauge reference level calculations
Energy Technology Data Exchange (ETDEWEB)
Huber, J.H., Fluor Daniel Hanford
1997-02-06
This document describes the method for calculating reference levels for Enraf Series 854 Level Detectors as installed in the tank farms. The reference level calculation for each installed level gauge is contained herein.
RHEIN, Modular System for Reactor Design Calculation
International Nuclear Information System (INIS)
Reiche, Christian; Barz, Hansulrich; Kunzmann, Bernd; Seifert, Eberhard; Wand, Hartmut
1990-01-01
1 - Description of program or function: RHEIN is a modular reactor code system for neutron physics calculations. It consists of a small number of system codes for execution control, data management, and handling support, as well as of the physical calculation routines. The execution is controlled by input data containing mathematical and physical parameters and simple commands for routine calls and data manipulations. The calculation routines are in tune with one another and the system takes care of the data transfer between them. Cross-section libraries with self shielding parameters are added to the system. 2 - Method of solution: The calculation routines can be used for solving the following physics problems: - Calculation of cross-section sets for infinite mediums, taking into account chord length. - Zero-dimensional spectrum calculation in diffusion, P1, or B1 approximation. - One-dimensional calculation in diffusion, P1, or collision probability approximation. - Two-dimensional diffusion calculation. - Cell calculation by THERMOS. - Zone-wise homogenized group collapsing within zero, one, or two-dimensional models. - Normalization, summarizing, etc. - Output of cross-section sets to off systems Sn and Monte-Carlo calculations
Energy Technology Data Exchange (ETDEWEB)
Millman, D. L. [Dept. of Computer Science, Univ. of North Carolina at Chapel Hill (United States); Griesheimer, D. P.; Nease, B. R. [Bechtel Marine Propulsion Corporation, Bertis Atomic Power Laboratory (United States); Snoeyink, J. [Dept. of Computer Science, Univ. of North Carolina at Chapel Hill (United States)
2012-07-01
In this paper we consider a new generalized algorithm for the efficient calculation of component object volumes given their equivalent constructive solid geometry (CSG) definition. The new method relies on domain decomposition to recursively subdivide the original component into smaller pieces with volumes that can be computed analytically or stochastically, if needed. Unlike simpler brute-force approaches, the proposed decomposition scheme is guaranteed to be robust and accurate to within a user-defined tolerance. The new algorithm is also fully general and can handle any valid CSG component definition, without the need for additional input from the user. The new technique has been specifically optimized to calculate volumes of component definitions commonly found in models used for Monte Carlo particle transport simulations for criticality safety and reactor analysis applications. However, the algorithm can be easily extended to any application which uses CSG representations for component objects. The paper provides a complete description of the novel volume calculation algorithm, along with a discussion of the conjectured error bounds on volumes calculated within the method. In addition, numerical results comparing the new algorithm with a standard stochastic volume calculation algorithm are presented for a series of problems spanning a range of representative component sizes and complexities. (authors)
International Nuclear Information System (INIS)
Millman, D. L.; Griesheimer, D. P.; Nease, B. R.; Snoeyink, J.
2012-01-01
In this paper we consider a new generalized algorithm for the efficient calculation of component object volumes given their equivalent constructive solid geometry (CSG) definition. The new method relies on domain decomposition to recursively subdivide the original component into smaller pieces with volumes that can be computed analytically or stochastically, if needed. Unlike simpler brute-force approaches, the proposed decomposition scheme is guaranteed to be robust and accurate to within a user-defined tolerance. The new algorithm is also fully general and can handle any valid CSG component definition, without the need for additional input from the user. The new technique has been specifically optimized to calculate volumes of component definitions commonly found in models used for Monte Carlo particle transport simulations for criticality safety and reactor analysis applications. However, the algorithm can be easily extended to any application which uses CSG representations for component objects. The paper provides a complete description of the novel volume calculation algorithm, along with a discussion of the conjectured error bounds on volumes calculated within the method. In addition, numerical results comparing the new algorithm with a standard stochastic volume calculation algorithm are presented for a series of problems spanning a range of representative component sizes and complexities. (authors)
Buckling feedback of the spectral calculations
International Nuclear Information System (INIS)
Jing Xingqing; Shan Wenzhi; Luo Jingyu
1992-01-01
This paper studies the problems about buckling feedback of spectral calculations in physical calculations of the reactor and presents a useful method by which the buckling feedback of spectral calculations is implemented. The effect of the buckling feedback in spectra and the broad group cross section, convergence of buckling feedback iteration and the effect of the spectral zones dividing are discussed in the calculations. This method has been used for the physical design of HTR-10 MW Test Module
International Nuclear Information System (INIS)
Wu Chuan; Guo Fanqing; Purdy, James A
2006-01-01
This paper reports a step-by-step shielding calculation recipe for a helical tomotherapy unit (TomoTherapy Inc., Madison, WI, USA), recently installed in an existing Varian 600C treatment room. Both primary and secondary radiations (leakage and scatter) are explicitly considered. A typical patient load is assumed. Use factor is calculated based on an analytical formula derived from the tomotherapy rotational beam delivery geometry. Leakage and scatter are included in the calculation based on corresponding measurement data as documented by TomoTherapy Inc. Our calculation result shows that, except for a small area by the therapists' console, most of the existing Varian 600C shielding is sufficient for the new tomotherapy unit. This work cautions other institutions facing the similar situation, where an HT unit is considered for an existing LINAC treatment room, more secondary shielding might be considered at some locations, due to the significantly increased secondary shielding requirement by HT. (note)
47 CFR 1.1623 - Probability calculation.
2010-10-01
... 47 Telecommunication 1 2010-10-01 2010-10-01 false Probability calculation. 1.1623 Section 1.1623 Telecommunication FEDERAL COMMUNICATIONS COMMISSION GENERAL PRACTICE AND PROCEDURE Random Selection Procedures for Mass Media Services General Procedures § 1.1623 Probability calculation. (a) All calculations shall be...
Practical astronomy with your calculator
Duffett-Smith, Peter
1989-01-01
Practical Astronomy with your Calculator, first published in 1979, has enjoyed immense success. The author's clear and easy to follow routines enable you to solve a variety of practical and recreational problems in astronomy using a scientific calculator. Mathematical complexity is kept firmly in the background, leaving just the elements necessary for swiftly making calculations. The major topics are: time, coordinate systems, the Sun, the planetary system, binary stars, the Moon, and eclipses. In the third edition there are entirely new sections on generalised coordinate transformations, nutr
Source and replica calculations
International Nuclear Information System (INIS)
Whalen, P.P.
1994-01-01
The starting point of the Hiroshima-Nagasaki Dose Reevaluation Program is the energy and directional distributions of the prompt neutron and gamma-ray radiation emitted from the exploding bombs. A brief introduction to the neutron source calculations is presented. The development of our current understanding of the source problem is outlined. It is recommended that adjoint calculations be used to modify source spectra to resolve the neutron discrepancy problem
Second reference calculation for the WIPP
International Nuclear Information System (INIS)
Branstetter, L.J.
1985-03-01
Results of the second reference calculation for the Waste Isolation Pilot Plant (WIPP) project using the dynamic relaxation finite element code SANCHO are presented. This reference calculation is intended to predict the response of a typical panel of excavated rooms designed for storage of nonheat-producing nuclear waste. Results are presented that include relevant deformations, relative clay seam displacements, and stress and strain profiles. This calculation is a particular solution obtained by a computer code, which has proven analytic capabilities when compared with other structural finite element codes. It is hoped that the results presented here will be useful in providing scoping values for defining experiments and for developing instrumentation. It is also hoped that the calculation will be useful as part of an exercise in developing a methodology for performing important design calculations by more than one analyst using more than one computer code, and for defining internal Quality Assurance (QA) procedures for such calculations. 27 refs., 15 figs
Three-dimensional electron-beam dose calculations
International Nuclear Information System (INIS)
Shiu, A.S.
1988-01-01
The MDAH pencil-beam algorithm developed by Hogstrom et al (1981) has been widely used in clinics for electron-beam dose calculations for radiotherapy treatment planning. The primary objective of this research was to address several deficiencies of that algorithm and to develop an enhanced version. Two enhancements were incorporated into the pencil-beam algorithm; one models fluence rather than planar fluence, and the other models the bremsstrahlung dose using measured beam data. Comparisons of the resulting calculated dose distributions with measured dose distributions for several test phantoms have been made. From these results it is concluded (1) that the fluence-based algorithm is more accurate to use for the dose calculation in an inhomogeneous slab phantom, and (2) the fluence-based calculation provides only a limited improvement to the accuracy the calculated dose in the region just downstream of the lateral edge of an inhomogeneity. A pencil-beam redefinition model was developed for the calculation of electron-beam dose distributions in three dimensions
Precipitates/Salts Model Sensitivity Calculation
International Nuclear Information System (INIS)
Mariner, P.
2001-01-01
The objective and scope of this calculation is to assist Performance Assessment Operations and the Engineered Barrier System (EBS) Department in modeling the geochemical effects of evaporation on potential seepage waters within a potential repository drift. This work is developed and documented using procedure AP-3.12Q, ''Calculations'', in support of ''Technical Work Plan For Engineered Barrier System Department Modeling and Testing FY 02 Work Activities'' (BSC 2001a). The specific objective of this calculation is to examine the sensitivity and uncertainties of the Precipitates/Salts model. The Precipitates/Salts model is documented in an Analysis/Model Report (AMR), ''In-Drift Precipitates/Salts Analysis'' (BSC 2001b). The calculation in the current document examines the effects of starting water composition, mineral suppressions, and the fugacity of carbon dioxide (CO 2 ) on the chemical evolution of water in the drift
Reaction rate calculations via transmission coefficients
International Nuclear Information System (INIS)
Feit, M.D.; Alder, B.J.
1985-01-01
The transmission coefficient of a wavepacket traversing a potential barrier can be determined by steady state calculations carried out in imaginary time instead of by real time dynamical calculations. The general argument is verified for the Eckart barrier potential by a comparison of transmission coefficients calculated from real and imaginary time solutions of the Schroedinger equation. The correspondence demonstrated here allows a formulation for the reaction rate that avoids difficulties due to both rare events and explicitly time dependent calculations. 5 refs., 2 figs
Nuclear calculation of the thorium reactor
International Nuclear Information System (INIS)
Hirakawa, Naohiro
1998-01-01
Even if for a reactor using thorium (and 233-U), its nuclear design calculation procedure is similar to the case using conventional 235-U, 238-U and plutonium. As nuclear composition varies with time on operation of nuclear reactor, calculation of its mean cross section should be conducted in details. At that time, one-group cross section obtained by integration over a whole of energy range is used for small member group. And, as the nuclear data for a base of its calculation is already prepared by JENDL3.2 and nuclear data library derived from it, the nuclear calculation of a nuclear reactor using thorium has no problem. From such a veiwpoint, IAEA has organized a coordinated research program of 'Potential of Th-based Fuel Cycles to Constrain Pu and to reduce Long-term Waste Toxicities' since 1996. All nations entering this program were regulated so as to institute by selecting a nuclear fuel cycle thinking better by each nation and to examine what cycle is expected by comparing their results. For a promise to conduct such neutral comparison, a comparison of bench mark calculations aiming at PWR was conducted to protect that the obtained results became different because of different calculation method and cross section adopted by each nation. Therefore, it was promoted by entrance of China, Germany, India, Israel, Japan, Korea, Russia and USA. The SWAT system developed by Tohoku University is used for its calculation code, by using which calculated results on the bench mark calculation at the fist and second stages and the nuclear reactor were reported. (G.K.)
CONTAIN calculations; CONTAIN-Rechnungen
Energy Technology Data Exchange (ETDEWEB)
Scholtyssek, W.
1995-08-01
In the first phase of a benchmark comparison, the CONTAIN code was used to calculate an assumed EPR accident `medium-sized leak in the cold leg`, especially for the first two days after initiation of the accident. The results for global characteristics compare well with those of FIPLOC, MELCOR and WAVCO calculations, if the same materials data are used as input. However, significant differences show up for local quantities such as flows through leakages. (orig.)
Feasibility study on embedded transport core calculations
International Nuclear Information System (INIS)
Ivanov, B.; Zikatanov, L.; Ivanov, K.
2007-01-01
The main objective of this study is to develop an advanced core calculation methodology based on embedded diffusion and transport calculations. The scheme proposed in this work is based on embedded diffusion or SP 3 pin-by-pin local fuel assembly calculation within the framework of the Nodal Expansion Method (NEM) diffusion core calculation. The SP 3 method has gained popularity in the last 10 years as an advanced method for neutronics calculation. NEM is a multi-group nodal diffusion code developed, maintained and continuously improved at the Pennsylvania State University. The developed calculation scheme is a non-linear iteration process, which involves cross-section homogenization, on-line discontinuity factors generation, and boundary conditions evaluation by the global solution passed to the local calculation. In order to accomplish the local calculation, a new code has been developed based on the Finite Elements Method (FEM), which is capable of performing both diffusion and SP 3 calculations. The new code will be used in the framework of the NEM code in order to perform embedded pin-by-pin diffusion and SP 3 calculations on fuel assembly basis. The development of the diffusion and SP 3 FEM code is presented first following by its application to several problems. Description of the proposed embedded scheme is provided next as well as the obtained preliminary results of the C3 MOX benchmark. The results from the embedded calculations are compared with direct pin-by-pin whole core calculations in terms of accuracy and efficiency followed by conclusions made about the feasibility of the proposed embedded approach. (authors)
Non-perturbative background field calculations
Stephens, C. R.
1988-01-01
New methods are developed for calculating one loop functional determinants in quantum field theory. Instead of relying on a calculation of all the eigenvalues of the small fluctuation equation, these techniques exploit the ability of the proper time formalism to reformulate an infinite dimensional field theoretic problem into a finite dimensional covariant quantum mechanical analog, thereby allowing powerful tools such as the method of Jacobi fields to be used advantageously in a field theory setting. More generally the methods developed herein should be extremely valuable when calculating quantum processes in non-constant background fields, offering a utilitarian alternative to the two standard methods of calculation—perturbation theory in the background field or taking the background field into account exactly. The formalism developed also allows for the approximate calculation of covariances of partial differential equations from a knowledge of the solutions of a homogeneous ordinary differential equation.
McCarty, George
1982-01-01
How THIS BOOK DIFFERS This book is about the calculus. What distinguishes it, however, from other books is that it uses the pocket calculator to illustrate the theory. A computation that requires hours of labor when done by hand with tables is quite inappropriate as an example or exercise in a beginning calculus course. But that same computation can become a delicate illustration of the theory when the student does it in seconds on his calculator. t Furthermore, the student's own personal involvement and easy accomplishment give hi~ reassurance and en couragement. The machine is like a microscope, and its magnification is a hundred millionfold. We shall be interested in limits, and no stage of numerical approximation proves anything about the limit. However, the derivative of fex) = 67.SgX, for instance, acquires real meaning when a student first appreciates its values as numbers, as limits of 10 100 1000 t A quick example is 1.1 , 1.01 , 1.001 , •••• Another example is t = 0.1, 0.01, in the functio...
International Nuclear Information System (INIS)
Petersen, K.E.
1986-03-01
Risk and reliability analysis is increasingly being used in evaluations of plant safety and plant reliability. The analysis can be performed either during the design process or during the operation time, with the purpose to improve the safety or the reliability. Due to plant complexity and safety and availability requirements, sophisticated tools, which are flexible and efficient, are needed. Such tools have been developed in the last 20 years and they have to be continuously refined to meet the growing requirements. Two different areas of application were analysed. In structural reliability probabilistic approaches have been introduced in some cases for the calculation of the reliability of structures or components. A new computer program has been developed based upon numerical integration in several variables. In systems reliability Monte Carlo simulation programs are used especially in analysis of very complex systems. In order to increase the applicability of the programs variance reduction techniques can be applied to speed up the calculation process. Variance reduction techniques have been studied and procedures for implementation of importance sampling are suggested. (author)
Daylight calculations in practice
DEFF Research Database (Denmark)
Iversen, Anne; Roy, Nicolas; Hvass, Mette
The aim of the project was to obtain a better understanding of what daylight calculations show and also to gain knowledge of how the different daylight simulation programs perform compared with each other. Experience has shown that results for the same room, obtained from two daylight simulation...... programs can give different results. This can be due to restrictions in the program itself and/or be due to the skills of the persons setting up the models. This is crucial as daylight calculations are used to document that the demands and recommendations to daylight levels outlined by building authorities....... The aim of the project was to obtain a better understanding of what daylight calculations show and also to gain knowledge of how the different daylight simulation programs perform compared with each other. Furthermore the aim was to provide knowledge of how to build up the 3D models that were...
PHEBUS-FPTO Benchmark calculations
International Nuclear Information System (INIS)
Shepherd, I.; Ball, A.; Trambauer, K.; Barbero, F.; Olivar Dominguez, F.; Herranz, L.; Biasi, L.; Fermandjian, J.; Hocke, K.
1991-01-01
This report summarizes a set of pre-test predictions made for the first Phebus-FP test, FPT-O. There were many different calculations, performed by various organizations and they represent the first attempt to calculate the whole experimental sequence, from bundle to containment. Quantitative agreement between the various calculations was not good but the particular models in the code responsible for disagreements were mostly identified. A consensus view was formed as to how the test would proceed. It was found that a successful execution of the test will require a different operating procedure than had been assumed here. Critical areas which require close attention are the need to devize a strategy for the power and flow in the bundle that takes account of uncertainties in the modelling and the shroud conductivity and the necessity to develop a reliable method to achieve the desired thermalhydraulic conditions in the containment
Core calculational techniques and procedures
International Nuclear Information System (INIS)
Romano, J.J.
1977-10-01
Described are the procedures and techniques employed by B and W in core design analyses of power peaking, control rod worths, and reactivity coefficients. Major emphasis has been placed on current calculational tools and the most frequently performed calculations over the operating power range
Calculation of Spectra of Solids:
DEFF Research Database (Denmark)
Lindgård, Per-Anker
1975-01-01
The Gilat-Raubenheimer method simplified to tetrahedron division is used to calculate the real and imaginary part of the dynamical response function for electrons. A frequency expansion for the real part is discussed. The Lindhard function is calculated as a test for numerical accuracy...
Economic calculation in socialist countries
Ellman, M.; Durlauf, S.N.; Blume, L.E.
2008-01-01
In the 1930s, when the classical socialist system emerged, economic decisions were based not on detailed and precise economic methods of calculation but on rough and ready political methods. An important method of economic calculation - particularly in the post-Stalin period - was that of
Calculated LET-Spectrum of Antiprotons
DEFF Research Database (Denmark)
Bassler, Niels
-LET components resulting from the annihilation. Though, the calculations of dose-averaged LET in the entry region may suggest that the RBE of antiprotons in the plateau region could significantly differ from unity. Materials and Methods Monte Carlo simulations using FLUKA were performed for calculating...
Radiation-damage calculations with NJOY
International Nuclear Information System (INIS)
MacFarlane, R.E.; Muir, D.W.; Mann, F.W.
1983-01-01
Atomic displacement, gas production, transmutation, and nuclear heating can all be calculated with the NJOY nuclear data processing system using evaluated data in ENDF/B format. Using NJOY helps assure consistency between damage cross sections and those used for transport, and NJOY provides convenient interface formats for linking data to application codes. Unique features of the damage calculation include a simple momentum balance treatment for radiative capture and a new model for (n, particle) reactions based on statistical model calculations. Sample results for iron and nickel are given and compared with the results of other methods
Alaska Village Electric Load Calculator
Energy Technology Data Exchange (ETDEWEB)
Devine, M.; Baring-Gould, E. I.
2004-10-01
As part of designing a village electric power system, the present and future electric loads must be defined, including both seasonal and daily usage patterns. However, in many cases, detailed electric load information is not readily available. NREL developed the Alaska Village Electric Load Calculator to help estimate the electricity requirements in a village given basic information about the types of facilities located within the community. The purpose of this report is to explain how the load calculator was developed and to provide instructions on its use so that organizations can then use this model to calculate expected electrical energy usage.
Reactor calculations and nuclear information
International Nuclear Information System (INIS)
Lang, D.W.
1977-12-01
The relationship of sets of nuclear parameters and the macroscopic reactor quantities that can be calculated from them is examined. The framework of the study is similar to that of Usachev and Bobkov. The analysis is generalised and some properties required by common sense are demonstrated. The form of calculation permits revision of the parameter set. It is argued that any discrepancy between a calculation and measurement of a macroscopic quantity is more useful when applied directly to prediction of other macroscopic quantities than to revision of the parameter set. The mathematical technique outlined is seen to describe common engineering practice. (Author)
Pile Load Capacity – Calculation Methods
Directory of Open Access Journals (Sweden)
Wrana Bogumił
2015-12-01
Full Text Available The article is a review of the current problems of the foundation pile capacity calculations. The article considers the main principles of pile capacity calculations presented in Eurocode 7 and other methods with adequate explanations. Two main methods are presented: α – method used to calculate the short-term load capacity of piles in cohesive soils and β – method used to calculate the long-term load capacity of piles in both cohesive and cohesionless soils. Moreover, methods based on cone CPTu result are presented as well as the pile capacity problem based on static tests.
Range calculations using multigroup transport methods
International Nuclear Information System (INIS)
Hoffman, T.J.; Robinson, M.T.; Dodds, H.L. Jr.
1979-01-01
Several aspects of radiation damage effects in fusion reactor neutron and ion irradiation environments are amenable to treatment by transport theory methods. In this paper, multigroup transport techniques are developed for the calculation of particle range distributions. These techniques are illustrated by analysis of Au-196 atoms recoiling from (n,2n) reactions with gold. The results of these calculations agree very well with range calculations performed with the atomistic code MARLOWE. Although some detail of the atomistic model is lost in the multigroup transport calculations, the improved computational speed should prove useful in the solution of fusion material design problems
Precipitates/Salts Model Sensitivity Calculation
Energy Technology Data Exchange (ETDEWEB)
P. Mariner
2001-12-20
The objective and scope of this calculation is to assist Performance Assessment Operations and the Engineered Barrier System (EBS) Department in modeling the geochemical effects of evaporation on potential seepage waters within a potential repository drift. This work is developed and documented using procedure AP-3.12Q, ''Calculations'', in support of ''Technical Work Plan For Engineered Barrier System Department Modeling and Testing FY 02 Work Activities'' (BSC 2001a). The specific objective of this calculation is to examine the sensitivity and uncertainties of the Precipitates/Salts model. The Precipitates/Salts model is documented in an Analysis/Model Report (AMR), ''In-Drift Precipitates/Salts Analysis'' (BSC 2001b). The calculation in the current document examines the effects of starting water composition, mineral suppressions, and the fugacity of carbon dioxide (CO{sub 2}) on the chemical evolution of water in the drift.
Current interruption transients calculation
Peelo, David F
2014-01-01
Provides an original, detailed and practical description of current interruption transients, origins, and the circuits involved, and how they can be calculated Current Interruption Transients Calculationis a comprehensive resource for the understanding, calculation and analysis of the transient recovery voltages (TRVs) and related re-ignition or re-striking transients associated with fault current interruption and the switching of inductive and capacitive load currents in circuits. This book provides an original, detailed and practical description of current interruption transients, origins,
Argosy 4 - A programme for lattice calculations
International Nuclear Information System (INIS)
MacDougall, J.D.
1965-07-01
This report contains a detailed description of the methods of calculation used in the Argosy 4 computer programme, and of the input requirements and printed results produced by the programme. An outline of the physics of the Argosy method is given. Section 2 describes the lattice calculation, including the burn up calculation, section 3 describes the control rod calculation and section 4 the reflector calculation. In these sections the detailed equations solved by the programme are given. In section 5 input requirements are given, and in section 6 the printed output obtained from an Argosy calculation is described. In section 7 are noted the principal differences between Argosy 4 and earlier versions of the Argosy programme
DEFF Research Database (Denmark)
Nielsen, Claus Werner; Nielsen, Ole-Kenneth
2009-01-01
Many countries are in the process of mapping their national CO2 emissions, but only few have managed to produce an overall report at municipal level yet. Denmark, however, has succeeded in such a project. Using a new national IT-based calculation model, municipalities can calculate the extent...
Calculation of 14 MeV neutron transmission
International Nuclear Information System (INIS)
Vyrskij, M.Yu.; Dubinin, A.A.; Zhuravlev, V.I.; Isaev, N.V.; Klintsov, A.A.; Krivtsov, A.S.; Linge, I.I.; Panfilov, E.I.; Prit'mov, A.P.
1979-01-01
The possibility of using the 28 group constant system (28-GCS) for calculating the transport of neutrons with initial energy of 14 MeV in thermonuclear reactor blankets is studied. A blanket project suggested by the Oak Ridge National Laboratory is used as a test version to estimate applicability of the 28-GCS. Niobium is used in a blanket as a structural material. A mixture of lithium nuclides is used for tritium production. The results of blanket test calculation and the calculational results obtained using the 28-GCS from the UKNDL library are compared. The numerical 28-group calculation of blonket is carried out by means of the ROZ-6 and ROZ-9 codes but not by the Monte-Carlo method as compared with the test calculation. Time of the blanket calculation on the BESM-6 computer by means of the ROZ-9 code in 2P 5 approximation using the 28-GCS amounts to 10 min. It is noted that to create effective codes for the numerical blanket calculation different calculational grids are necessary for different energy grups. The calculations carried out have shown the possibility of using the 28-group library of cross sections for the numerical solution of the neutron transport equation in estimating analysis of blankets
International Nuclear Information System (INIS)
Craig, D.S.; Festarini, G.L.
1986-07-01
The Monte Carlo code, REPC, has been used to calculate resonance reaction rates for the thermal test lattices TRX-1 and MIT-4, and for the CRNL lattices ZEEP-1, 19 UO 2 and 37 UO 2 . These reaction rates were used in the RAHAB cell code to calculate k eff , conversion ratios, and fast fission ratios, for comparison with experimental values. The calculations used the cluster geometry for the 19-, 28-, and 37-element clusters. Calculations were also made using annular representations of the cluster for comparison of the rates with those obtained using the discrete ordinate code OZMA
Subsurface Shielding Source Term Specification Calculation
International Nuclear Information System (INIS)
S.Su
2001-01-01
The purpose of this calculation is to establish appropriate and defensible waste-package radiation source terms for use in repository subsurface shielding design. This calculation supports the shielding design for the waste emplacement and retrieval system, and subsurface facility system. The objective is to identify the limiting waste package and specify its associated source terms including source strengths and energy spectra. Consistent with the Technical Work Plan for Subsurface Design Section FY 01 Work Activities (CRWMS M and O 2001, p. 15), the scope of work includes the following: (1) Review source terms generated by the Waste Package Department (WPD) for various waste forms and waste package types, and compile them for shielding-specific applications. (2) Determine acceptable waste package specific source terms for use in subsurface shielding design, using a reasonable and defensible methodology that is not unduly conservative. This calculation is associated with the engineering and design activity for the waste emplacement and retrieval system, and subsurface facility system. The technical work plan for this calculation is provided in CRWMS M and O 2001. Development and performance of this calculation conforms to the procedure, AP-3.12Q, Calculations
Calculation of neutron kerma in tissues
International Nuclear Information System (INIS)
Vega C, H.R.; Manzanares A, E.
2004-01-01
Neutron kerma of normal and tumor tissues has been calculated using the tissues elemental concentration. A program developed in Math cad contains the kerma factors of C, H, O, N, Na, Mg, P, S, Cl, K, etc. that are in normal and tumor human tissues. Having the elemental composition of any human tissue the neutron kerma can be calculated. The program was tested using the elemental composition of tumor tissues such as sarcoma, melanoma, carcinoma and adenoid cystic, also neutron kerma for adipose and muscle tissue for normal adult was calculated. The results are in agreement with those published in literature. The neutron kerma for water was also calculated because in some dosimetric calculations water is used to describe normal and tumor tissues. From this comparison was found that at larger energies kerma factors are approximately the same, but energies less than 100 eV the differences are large. (Author)
Comparison of RESRAD with hand calculations
International Nuclear Information System (INIS)
Rittmann, P.D.
1995-09-01
This report is a continuation of an earlier comparison done with two other computer programs, GENII and PATHRAE. The dose calculations by the two programs were compared with each other and with hand calculations. These band calculations have now been compared with RESRAD Version 5.41 to examine the use of standard models and parameters in this computer program. The hand calculations disclosed a significant computational error in RESRAD. The Pu-241 ingestion doses are five orders of magnitude too small. In addition, the external doses from some nuclides differ greatly from expected values. Both of these deficiencies have been corrected in later versions of RESRAD
Calculational Tool for Skin Contamination Dose Assessment
Hill, R L
2002-01-01
Spreadsheet calculational tool was developed to automate the calculations preformed for dose assessment of skin contamination. This document reports on the design and testing of the spreadsheet calculational tool.
Calculator. Owning a Small Business.
Parma City School District, OH.
Seven activities are presented in this student workbook designed for an exploration of small business ownership and the use of the calculator in this career. Included are simulated situations in which students must use a calculator to compute property taxes; estimate payroll taxes and franchise taxes; compute pricing, approximate salaries,…
Energy Technology Data Exchange (ETDEWEB)
Kneur, J.L
2006-06-15
This document is divided into 2 parts. The first part describes a particular re-summation technique of perturbative series that can give a non-perturbative results in some cases. We detail some applications in field theory and in condensed matter like the calculation of the effective temperature of Bose-Einstein condensates. The second part deals with the minimal supersymmetric standard model. We present an accurate calculation of the mass spectrum of supersymmetric particles, a calculation of the relic density of supersymmetric black matter, and the constraints that we can infer from models.
Hiatt, Arthur A.
1987-01-01
Ten activities that give learners in grades 5-8 a chance to explore mathematics with calculators are provided. The activity cards involve such topics as odd addends, magic squares, strange projects, and conjecturing rules. (MNS)
The FLUKA atmospheric neutrino flux calculation
Battistoni, G.; Montaruli, T.; Sala, P.R.
2003-01-01
The 3-dimensional (3-D) calculation of the atmospheric neutrino flux by means of the FLUKA Monte Carlo model is here described in all details, starting from the latest data on primary cosmic ray spectra. The importance of a 3-D calculation and of its consequences have been already debated in a previous paper. Here instead the focus is on the absolute flux. We stress the relevant aspects of the hadronic interaction model of FLUKA in the atmospheric neutrino flux calculation. This model is constructed and maintained so to provide a high degree of accuracy in the description of particle production. The accuracy achieved in the comparison with data from accelerators and cross checked with data on particle production in atmosphere certifies the reliability of shower calculation in atmosphere. The results presented here can be already used for analysis by current experiments on atmospheric neutrinos. However they represent an intermediate step towards a final release, since this calculation does not yet include the...
Numerical calculation of the Fresnel transform.
Kelly, Damien P
2014-04-01
In this paper, we address the problem of calculating Fresnel diffraction integrals using a finite number of uniformly spaced samples. General and simple sampling rules of thumb are derived that allow the user to calculate the distribution for any propagation distance. It is shown how these rules can be extended to fast-Fourier-transform-based algorithms to increase calculation efficiency. A comparison with other theoretical approaches is made.
Calculation of the Moments of Polygons.
1987-06-01
2.1) VowUK-1N0+IDIO TUUNTKPlNO.YKNO C Calculate AREA YKXK-YKPIND*IKNO-YKNO*XKP1NO AIKA-hEEA4YKXX C Calculate ACEIT ACENT (1)- ACEIT ( 1) VSUNI4TKIK... ACEIT (2) -ACENT(2) .VSUNYKXK C Calculate SECHON 3ECNON (1) -SCNON( 1) TKXK*(XX~PIdO*VSUNXKKO**2) SECNO(2) -SEn N(2) .yrf* (XKP114*YKP1MO.XKO*YXO+VB1hi
International Nuclear Information System (INIS)
Sekimoto, H.
1987-01-01
The kerma heat production density, tritum production density, and dose in a lithium-fluoride pile with a deuterium-tritum neutron source were calculated with a data processing code, UFO, from the pulse height distribution of a miniature NE213 neutron spectrometer, and compared with the values calculated with a Monte Carlo code, MORSE-CV. Both the UFO and MORSE-CV values agreed with the statistical error (less than 6%) of the MORSE-CV calculations, except for the outer-most point in the pile. The MORSE-CV values were slightly smaller than the UFO values for almost all cases, and this tendency increased with increasing distance from the neutron source
Calculation of the BREN house shielding experiments
International Nuclear Information System (INIS)
Woolson, William A.; Gritzner, Michael L.
1987-01-01
The BREN house transmission experiments provide an excellent set of measurements to validate the calculational procedures that will be used to derive house shielding estimates for the revised dosimetry of the survivors of the Hiroshima and Nagasaki A-bombs. The BREN experiments were performed in realistic full scale models of Japanese residences. Although the radiation spectra and relative intensities of neutrons and gamma rays incident on the houses from the HPRR and the 60 Co source are not appropriate for direct application to the A-bomb survivors, they cover the full energy range of importance. The codes and calculations required to compare with BREN experiments are the same as those needed for the A-bomb dosimetry. They consist of a two-dimensional discrete-ordinates calculation of the free field coupled to an adjoint Monte Carlo calculation in detailed house geometry. The agreement obtained between calculations and the experiments is excellent for neutrons and 60 Co gamma rays. Every house transmission calculation spanning simple to complex configurations and detector locations for the 60 Co and HPRR was within an acceptable margin of error. The gamma-ray TF calculations for the reactor source did not agree well with the experiments. Analysis of this discrepancy, however, strongly indicates that the problem probably does not reside in the calculational procedure but in the measurements themselves. In conclusion, it is believed that the excellent agreement of our calculations with the BREN experiments validates the calculational procedure which is planed to be applied o estimating the house shielding for survivors of the Hiroshima and Nagasaki A-bombs. Certainly, the calculations for Hiroshima and Nagasaki will involve modifications to the code used for the computations reported here, but to the extent that these modifications involve increased calculational complexity to treat more realistic materials and configurations, the benchmark established by these
Fission neutron multiplicity calculations
International Nuclear Information System (INIS)
Maerten, H.; Ruben, A.; Seeliger, D.
1991-01-01
A model for calculating neutron multiplicities in nuclear fission is presented. It is based on the solution of the energy partition problem as function of mass asymmetry within a phenomenological approach including temperature-dependent microscopic energies. Nuclear structure effects on fragment de-excitation, which influence neutron multiplicities, are discussed. Temperature effects on microscopic energy play an important role in induced fission reactions. Calculated results are presented for various fission reactions induced by neutrons. Data cover the incident energy range 0-20 MeV, i.e. multiple chance fission is considered. (author). 28 refs, 13 figs
Lattice cell burnup calculation
International Nuclear Information System (INIS)
Pop-Jordanov, J.
1977-01-01
Accurate burnup prediction is a key item for design and operation of a power reactor. It should supply information on isotopic changes at each point in the reactor core and the consequences of these changes on the reactivity, power distribution, kinetic characters, control rod patterns, fuel cycles and operating strategy. A basic stage in the burnup prediction is the lattice cell burnup calculation. This series of lectures attempts to give a review of the general principles and calculational methods developed and applied in this area of burnup physics
Methodology of shielding calculation for nuclear reactors
International Nuclear Information System (INIS)
Maiorino, J.R.; Mendonca, A.G.; Otto, A.C.; Yamaguchi, Mitsuo
1982-01-01
A methodology of calculation that coupling a serie of computer codes in a net that make the possibility to calculate the radiation, neutron and gamma transport, is described, for deep penetration problems, typical of nuclear reactor shielding. This net of calculation begining with the generation of constant multigroups, for neutrons and gamma, by the AMPX system, coupled to ENDF/B-IV data library, the transport calculation of these radiations by ANISN, DOT 3.5 and Morse computer codes, up to the calculation of absorbed doses and/or equivalents buy SPACETRAN code. As examples of the calculation method, results from benchmark n 0 6 of Shielding Benchmark Problems - ORNL - RSIC - 25, namely Neutron and Secondary Gamma Ray fluence transmitted through a Slab of Borated Polyethylene, are presented. (Author) [pt
Benchmarking criticality safety calculations with subcritical experiments
International Nuclear Information System (INIS)
Mihalczo, J.T.
1984-06-01
Calculation of the neutron multiplication factor at delayed criticality may be necessary for benchmarking calculations but it may not be sufficient. The use of subcritical experiments to benchmark criticality safety calculations could result in substantial savings in fuel material costs for experiments. In some cases subcritical configurations could be used to benchmark calculations where sufficient fuel to achieve delayed criticality is not available. By performing a variety of measurements with subcritical configurations, much detailed information can be obtained which can be compared directly with calculations. This paper discusses several measurements that can be performed with subcritical assemblies and presents examples that include comparisons between calculation and experiment where possible. Where not, examples from critical experiments have been used but the measurement methods could also be used for subcritical experiments
Calculational examination of the Baneberry event
International Nuclear Information System (INIS)
Terhune, R.W.; Glenn, H.D.; Burton, D.E.; McKague, H.L.; Rambo, J.T.
1977-01-01
On December 18, 1970, Baneberry, a 10-kt nuclear device, was detonated at a depth of 278 m in hole U8d at the Nevada Test Site. A shock-induced fissure near ground zero opened and vented radioactive gases and debris into the atmosphere. This report presents the calculational results describing the sequence of dynamic phenomena that produced the vent. The calculations predict the long positive velocity pulse and the surface motion that were observed experimentally. The surface fissure through which the material vented is the same radial distance from ground zero as the maximum horizontal displacement is calculated to be. The calculations predict a final cavity radius that is very close to the measured Baneberry cavity radius. Finally, the calculations indicate that an open fracture path runs from the cavity to the Baneberry Fault, up the fault to the spall region, and then vertically to the surface. This was the vent path predicted by calculations and is consistent with the vent path found from the radioactivity in postshot drill holes. Because of our extensions in computational capabilities, we believe this report advances the state of the art for numerical simulation of the containment problems associated with underground nuclear tests
Core burn-up calculation method of JRR-3
International Nuclear Information System (INIS)
Kato, Tomoaki; Yamashita, Kiyonobu
2007-01-01
SRAC code system is utilized for core burn-up calculation of JRR-3. SRAC code system includes calculation modules such as PIJ, PIJBURN, ANISN and CITATION for making effective cross section and calculation modules such as COREBN and HIST for core burn-up calculation. As for calculation method for JRR-3, PIJBURN (Cell burn-up calculation module) is used for making effective cross section of fuel region at each burn-up step. PIJ, ANISN and CITATION are used for making effective cross section of non-fuel region. COREBN and HIST is used for core burn-up calculation and fuel management. This paper presents details of NRR-3 core burn-up calculation. FNCA Participating countries are expected to carry out core burn-up calculation of domestic research reactor by SRAC code system by utilizing the information of this paper. (author)
Calculation methods for determining dose equivalent
International Nuclear Information System (INIS)
Endres, G.W.R.; Tanner, J.E.; Scherpelz, R.I.; Hadlock, D.E.
1988-01-01
A series of calculations of neutron fluence as a function of energy in an anthropomorphic phantom was performed to develop a system for determining effective dose equivalent for external radiation sources. critical organ dose equivalents are calculated and effective dose equivalents are determined using ICRP-26 methods. Quality factors based on both present definitions and ICRP-40 definitions are used in the analysis. The results of these calculations are presented and discussed
Propagation calculation for reactor cases
Energy Technology Data Exchange (ETDEWEB)
Yang Yanhua [School of Power and Energy Engineering, Shanghai Jiao Tong Univ., Shanghai (China); Moriyama, K.; Maruyama, Y.; Nakamura, H.; Hashimoto, K. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2000-11-01
The propagation of steam explosion for real reactor geometry and conditions are investigated by using the computer code JASMINE-pro. The ex-vessel steam explosion is considered, which is described as follow: during the accident of reactor core meltdown, the molten core melts a hole at the bottom of reactor vessel and causes the higher temperature core fuel being leaked into the water pool below reactor vessel. During the melt-water mixing interaction process, the high temperature melt evaporates the cool water at an extreme high rate and might induce a steam explosion. A steam explosion could experience first the premixing phase and then the propagation explosion phase. For a propagation calculation, we should know the information about the initial fragmentation time, the total melt mass, premixing region size, initial void fraction and distribution of the melt volume fraction, and so on. All the initial conditions used in this calculation are based on analyses from some simple assumptions and the observation from the experiments. The results show that the most important parameter for the initial condition of this phase is the total mass and its initial distribution. This gives the requirement for a premixing calculation. On the other hand, for higher melt volume fraction case, the fragmentation is strong so that the local pressure can exceed over the EOS maximum pressure of the code, which lead to the incorrect calculation or divergence of the calculation. (Suetake, M.)
International Nuclear Information System (INIS)
Mueller, R.G.
1987-06-01
Due to the strong influence of vapour bubbles on the nuclear chain reaction, an exact calculation of neutron physics and thermal hydraulics in light water reactors requires consideration of subcooled boiling. To this purpose, in the present study a dynamic model is derived from the time-dependent conservation equations. It contains new methods for the time-dependent determination of evaporation and condensation heat flow and for the heat transfer coefficient in subcooled boiling. Furthermore, it enables the complete two-phase flow region to be treated in a consistent manner. The calculation model was verified using measured data of experiments covering a wide range of thermodynamic boundary conditions. In all cases very good agreement was reached. The results from the coupling of the new calculation model with a neutron kinetics program proved its suitability for the steady-state and transient calculation of reactor cores. (orig.) [de
PHYSICOCHEMICAL PROPERTY CALCULATIONS
Computer models have been developed to estimate a wide range of physical-chemical properties from molecular structure. The SPARC modeling system approaches calculations as site specific reactions (pKa, hydrolysis, hydration) and `whole molecule' properties (vapor pressure, boilin...
Cardiovascular risk calculation
African Journals Online (AJOL)
James A. Ker
2014-08-20
Aug 20, 2014 ... smoking and elevated blood sugar levels (diabetes mellitus). These risk ... These are risk charts, e.g. FRS, a non-laboratory-based risk calculation, and ... for hard cardiovascular end-points, such as coronary death, myocardial ...
Performance assessment calculational exercises
International Nuclear Information System (INIS)
Barnard, R.W.; Dockery, H.A.
1990-01-01
The Performance Assessment Calculational Exercises (PACE) are an ongoing effort coordinated by Yucca Mountain Project Office. The objectives of fiscal year 1990 work, termed PACE-90, as outlined in the Department of Energy Performance Assessment (PA) Implementation Plan were to develop PA capabilities among Yucca Mountain Project (YMP) participants by calculating performance of a Yucca Mountain (YM) repository under ''expected'' and also ''disturbed'' conditions, to identify critical elements and processes necessary to assess the performance of YM, and to perform sensitivity studies on key parameters. It was expected that the PACE problems would aid in development of conceptual models and eventual evaluation of site data. The PACE-90 participants calculated transport of a selected set of radionuclides through a portion of Yucca Mountain for a period of 100,000 years. Results include analyses of fluid-flow profiles, development of a source term for radionuclide release, and simulations of contaminant transport in the fluid-flow field. Later work included development of a problem definition for perturbations to the originally modeled conditions and for some parametric sensitivity studies. 3 refs
International Nuclear Information System (INIS)
Naka, Takafumi; Yamamoto, Munenari.
1995-01-01
When there are two or more origins of deuterium nuclides in reactor fuels, there are disposed a memory device for an amount of deuterium nuclides for every origin in a noted fuel segment at a certain time point, a device for calculating the amount of nuclides for every origin and current neutron fluxes in the noted fuel segment, and a device for separating and then displaying the amount of deuterium nuclides for every origin. Equations for combustion are dissolved for every origin of the deuterium nuclides based on the amount of the deuterium nuclides for every origin and neutron fluxes, to calculate the current amount of deuterium nuclides for every origin. The amount of deuterium nuclides originated from uranium is calculated ignoring α-decay of curium, while the amount of deuterium nuclides originated from plutonium is calculated ignoring the generation of plutonium formed from neptunium. Deuterium nuclides can be measured and controlled accurately for every origin of the reactor fuels. Even when nuclear fuel materials have two or more nationalities, the measurement and control thereof can be conducted for every country. (N.H.)
Methods of bone marrow dose calculation
International Nuclear Information System (INIS)
Taboaco, R.C.
1982-02-01
Several methods of bone marrow dose calculation for photon irradiation were analised. After a critical analysis, the author proposes the adoption, by the Instituto de Radioprotecao e Dosimetria/CNEN, of Rosenstein's method for dose calculations in Radiodiagnostic examinations and Kramer's method in case of occupational irradiation. It was verified by Eckerman and Simpson that for monoenergetic gamma emitters uniformly distributed within the bone mineral of the skeleton the dose in the bone surface can be several times higher than dose in skeleton. In this way, is also proposed the Calculation of tissue-air ratios for bone surfaces in some irradiation geometries and photon energies to be included in the Rosenstein's method for organ dose calculation in Radiodiagnostic examinations. (Author) [pt
Spreadsheet eases heat balance, payback calculations
International Nuclear Information System (INIS)
Conner, K.P.
1992-01-01
This paper reports that a generalized Lotus type spreadsheet program has been developed to perform the heat balance and simple payback calculations for various turbine-generator (TG) inlet steam pressures. It can be used for potential plant expansions or new cogeneration installations. The program performs the basic heat balance calculations that are associated with turbine-generator, feedwater heating process steam requirements and desuperheating. The printout, shows the basic data and formulation used in the calculations. The turbine efficiency data used are applicable for automatic extraction turbine-generators in the 30-80 MW range. Simple payback calculations are for chemical recovery boilers and power boilers used in the pulp and paper industry. However, the program will also accommodate boilers common to other industries
Calculation models for a nuclear reactor
International Nuclear Information System (INIS)
Tashanii, Ahmed Ali
2010-01-01
Determination of different parameters of nuclear reactors requires neutron transport calculations. Due to complicity of geometry and material composition of the reactor core, neutron calculations were performed for simplified models of the real arrangement. In frame of the present work two models were used for calculations. First, an elementary cell model was used to prepare cross section data set for a homogenized-core reactor model. The homogenized-core reactor model was then used to perform neutron transport calculation. The nuclear reactor is a tank-shaped thermal reactor. The semi-cylindrical core arrangement consists of aluminum made fuel bundles immersed in water which acts as a moderator as well as a coolant. Each fuel bundle consists of aluminum cladded fuel rods arranged in square lattices. (author)
Large scale calculations for hadron spectroscopy
International Nuclear Information System (INIS)
Rebbi, C.
1985-01-01
The talk reviews some recent Monte Carlo calculations for Quantum Chromodynamics, performed on Euclidean lattices of rather large extent. Purpose of the calculations is to provide accurate determinations of quantities, such as interquark potentials or mass eigenvalues, which are relevant for hadronic spectroscopy. Results obtained in quenched QCD on 16 3 x 32 lattices are illustrated, and a discussion of computational resources and techniques required for the calculations is presented. 18 refs.,3 figs., 2 tabs
Effective action calculation in lattice QCD
International Nuclear Information System (INIS)
Hoek, J.
1983-01-01
A method (called the effective action method) devised to make analytic calculations in Quantum Chromodynamics in the region of strong coupling is presented. First, the author deals with developing the calculation of a strong coupling expansion of the generating functional for gauge systems on a lattice with arbitrary sources. An accompanying manual describes the implementation of this calculation on a computer. The next step consists of substituting the expressions for the one-link free energies for a specific gauge group in the result of the previous calculation. This process of substitution, together with the replacement of the sources by a bilinear combination of fermion fields, is described for the group SU(3). More details on the implementation of the substitution scheme on a computer can be found in the accompanying manual. From the effective action thus obtained in terms of meson fields and baryon fields the Green functions of the theory can be derived. As an illustrative application the effective potential determining the vacuum expectation value of the meson field is calculated. (Auth.)
Sandia Strehl Calculator Version 1.0
Energy Technology Data Exchange (ETDEWEB)
2017-12-05
The Sandia Strehl Calculator is designed to calculate the Gibson and Lanni point spread function (PSF), Strehl ratio, and ensquared energy, allowing non-design immersion, coverslip, and sample layers. It also uses Abbe number calculations to determine the refractive index at specific wavelengths when given the refractive index at a different wavelength and the dispersion. The primary application of Sandia Strehl Calculator is to determine the theoretical impacts of using an optical microscope beyond its normal design parameters. Examples of non-design microscope usage include: a) using coverslips of non-design material b) coverslips of different thicknesses c) imaging deep into an aqueous sample with an immersion objective d) imaging a sample at 37 degrees. All of these changes can affect the imaging quality, sometimes profoundly, but are at the same time non-design conditions employed not infrequently. Rather than having to experimentally determine whether the changes will result in unacceptable image quality, Sandia Strehl Calculator uses existing optical theory to determine the approximate effect of the change, saving the need to perform experiments.
Calculation of magnetic hyperfine constants
International Nuclear Information System (INIS)
Bufaical, R.F.; Maffeo, B.; Brandi, H.S.
1975-01-01
The magnetic hyperfine constants of the V sub(K) center in CaF 2 , SrF 2 and BaF 2 have been calculated assuming a phenomenological model, based on the F 2 - 'central molucule', to describe the wavefunction of the defect. Calculations have shown that introduction of a small degree of covalence, between this central molecule and neighboring ions, is necessary to improve the electronic structure description of the defect. It was also shown that the results for the hyperfine constants are strongly dependent on the relaxations of the ions neighboring the central molecule; these relaxations have been determined by fitting the experimental data. The present results are compared with other previous calculations where similar and different theoretical methods have been used
Parameters calculation of shielding experiment
International Nuclear Information System (INIS)
Gavazza, S.
1986-02-01
The radiation transport methodology comparing the calculated reactions and dose rates for neutrons and gama-rays, with experimental measurements obtained on iron shield, irradiated in the YAYOI reactor is evaluated. The ENDF/B-IV and VITAMIN-C libraries and the AMPX-II modular system, for cross sections generation collapsed by the ANISN code were used. The transport calculations were made using the DOT 3.5 code, adjusting the boundary iron shield source spectrum to the reactions and dose rates, measured at the beginning of shield. The neutron and gamma ray distributions calculated on the iron shield presented reasonable agreement with experimental measurements. An experimental arrangement using the IEA-R1 reactor to determine a shielding benchmark is proposed. (Author) [pt
Three dimensions transport calculations for PWR core; Calcul de coeur R.E.P. en transport 3D
Energy Technology Data Exchange (ETDEWEB)
Richebois, E
2000-07-01
The objective of this work is to define improved 3-D core calculation methods based on the transport theory. These methods can be particularly useful and lead to more precise computations in areas of the core where anisotropy and steep flux gradients occur, especially near interface and boundary conditions and in regions of high heterogeneity (bundle with absorbent rods). In order to apply the transport theory a new method for calculating reflector constants has been developed, since traditional methods were only suited for 2-group diffusion core calculations and could not be extrapolated to transport calculations. In this thesis work, the new method for obtaining reflector constants is derived regardless of the number of energy groups and of the operator used. The core calculations results using the reflector constants thereof obtained have been validated on the EDF's power reactor Saint Laurent B1 with MOX loading. The advantages of a 3-D core transport calculation scheme have been highlighted as opposed to diffusion methods; there are a considerable number of significant effects and potential advantages to be gained in rod worth calculations for instance. These preliminary results obtained with on particular cycle will have to be confirmed by more systematic analysis. Accidents like MSLB (main steam line break) and LOCA (loss of coolant accident) should also be investigated and constitute challenging situations where anisotropy is high and/or flux gradients are steep. This method is now being validated for others EDF's PWRs' reactors, as well as for experimental reactors and other types of commercial reactors. (author)
Three dimensions transport calculations for PWR core; Calcul de coeur R.E.P. en transport 3D
Energy Technology Data Exchange (ETDEWEB)
Richebois, E
2000-07-01
The objective of this work is to define improved 3-D core calculation methods based on the transport theory. These methods can be particularly useful and lead to more precise computations in areas of the core where anisotropy and steep flux gradients occur, especially near interface and boundary conditions and in regions of high heterogeneity (bundle with absorbent rods). In order to apply the transport theory a new method for calculating reflector constants has been developed, since traditional methods were only suited for 2-group diffusion core calculations and could not be extrapolated to transport calculations. In this thesis work, the new method for obtaining reflector constants is derived regardless of the number of energy groups and of the operator used. The core calculations results using the reflector constants thereof obtained have been validated on the EDF's power reactor Saint Laurent B1 with MOX loading. The advantages of a 3-D core transport calculation scheme have been highlighted as opposed to diffusion methods; there are a considerable number of significant effects and potential advantages to be gained in rod worth calculations for instance. These preliminary results obtained with on particular cycle will have to be confirmed by more systematic analysis. Accidents like MSLB (main steam line break) and LOCA (loss of coolant accident) should also be investigated and constitute challenging situations where anisotropy is high and/or flux gradients are steep. This method is now being validated for others EDF's PWRs' reactors, as well as for experimental reactors and other types of commercial reactors. (author)
Methods for Melting Temperature Calculation
Hong, Qi-Jun
Melting temperature calculation has important applications in the theoretical study of phase diagrams and computational materials screenings. In this thesis, we present two new methods, i.e., the improved Widom's particle insertion method and the small-cell coexistence method, which we developed in order to capture melting temperatures both accurately and quickly. We propose a scheme that drastically improves the efficiency of Widom's particle insertion method by efficiently sampling cavities while calculating the integrals providing the chemical potentials of a physical system. This idea enables us to calculate chemical potentials of liquids directly from first-principles without the help of any reference system, which is necessary in the commonly used thermodynamic integration method. As an example, we apply our scheme, combined with the density functional formalism, to the calculation of the chemical potential of liquid copper. The calculated chemical potential is further used to locate the melting temperature. The calculated results closely agree with experiments. We propose the small-cell coexistence method based on the statistical analysis of small-size coexistence MD simulations. It eliminates the risk of a metastable superheated solid in the fast-heating method, while also significantly reducing the computer cost relative to the traditional large-scale coexistence method. Using empirical potentials, we validate the method and systematically study the finite-size effect on the calculated melting points. The method converges to the exact result in the limit of a large system size. An accuracy within 100 K in melting temperature is usually achieved when the simulation contains more than 100 atoms. DFT examples of Tantalum, high-pressure Sodium, and ionic material NaCl are shown to demonstrate the accuracy and flexibility of the method in its practical applications. The method serves as a promising approach for large-scale automated material screening in which
Energy Technology Data Exchange (ETDEWEB)
Trkov, A; Ravnik, M; Zeleznik, N [Inst. Jozef Stefan, Ljubljana (Slovenia)
1992-07-01
Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [Slovenian] Opisali smo programski paket CORD-2, ki se uporablja pri projektnih izracunih sredice pri upravljanju tlacnovodnega reaktorja. Prikazana je uporaba paketa in racunskih postopkov za tipicne probleme, ki nastopajo pri projektiranju sredice. Primerjava glavnih rezultatov z eksperimentalnimi vrednostmi je predstavljena kot del preveritvenega procesa. (author)
Ensuring the validity of calculated subcritical limits
International Nuclear Information System (INIS)
Clark, H.K.
1977-01-01
The care taken at the Savannah River Laboratory and Plant to ensure the validity of calculated subcritical limits is described. Close attention is given to ANSI N16.1-1975, ''Validation of Calculational Methods for Nuclear Criticality Safety.'' The computer codes used for criticality safety computations, which are listed and are briefly described, have been placed in the SRL JOSHUA system to facilitate calculation and to reduce input errors. A driver module, KOKO, simplifies and standardizes input and links the codes together in various ways. For any criticality safety evaluation, correlations of the calculational methods are made with experiment to establish bias. Occasionally subcritical experiments are performed expressly to provide benchmarks. Calculated subcritical limits contain an adequate but not excessive margin to allow for uncertainty in the bias. The final step in any criticality safety evaluation is the writing of a report describing the calculations and justifying the margin
Whole core burnup calculations using `MCNP`
Energy Technology Data Exchange (ETDEWEB)
Haran, O; Shaham, Y [Israel Atomic Energy Commission, Beersheba (Israel). Nuclear Research Center-Negev
1996-12-01
Core parameters such as the reactivity, the power distribution and different reactivity coefficients calculated in simulations play an important role in the nuclear reactor handling. Operational safety margins are decided upon, based on the calculated parameters. Thus, the ability to accurately calculate those parameters is of uppermost importance. Such ability exists for fresh cores, using the Monte-Carlo method. The change in the core parameters that results from the core burnup is nowadays calculated within transport codes that simplifies the transport process by using approximations such as the diffusion approximation. The inaccuracy in the burned core parameters arising from the use of such approximations is hard to quantify, leading to an increased gap between the operational routines and the safety limits. A Monte Carlo transport code that caries out accurate static calculations in three dimensional geometries using continuous-energy neutron cross-section data such as the MCNP can be used to generate accurate reaction rates for burnup purposes. Monte Carlo method is statistical by nature, so that the reaction rates calculated will be accurate only to a certain known extent. The purpose of this work was to create a burnup routine that uses the capabilities of the Monte Carlo based MCNP code. It should be noted that burnup using Monte Carlo has been reported in the literatures, but this work is the result of an independent effort (authors).
Whole core burnup calculations using 'MCNP'
International Nuclear Information System (INIS)
Haran, O.; Shaham, Y.
1996-01-01
Core parameters such as the reactivity, the power distribution and different reactivity coefficients calculated in simulations play an important role in the nuclear reactor handling. Operational safety margins are decided upon, based on the calculated parameters. Thus, the ability to accurately calculate those parameters is of uppermost importance. Such ability exists for fresh cores, using the Monte-Carlo method. The change in the core parameters that results from the core burnup is nowadays calculated within transport codes that simplifies the transport process by using approximations such as the diffusion approximation. The inaccuracy in the burned core parameters arising from the use of such approximations is hard to quantify, leading to an increased gap between the operational routines and the safety limits. A Monte Carlo transport code that caries out accurate static calculations in three dimensional geometries using continuous-energy neutron cross-section data such as the MCNP can be used to generate accurate reaction rates for burnup purposes. Monte Carlo method is statistical by nature, so that the reaction rates calculated will be accurate only to a certain known extent. The purpose of this work was to create a burnup routine that uses the capabilities of the Monte Carlo based MCNP code. It should be noted that burnup using Monte Carlo has been reported in the literatures, but this work is the result of an independent effort (authors)
International Nuclear Information System (INIS)
Yan Guanghua; Liu, Chihray; Lu Bo; Palta, Jatinder R; Li, Jonathan G
2008-01-01
The purpose of this study was to choose an appropriate head scatter source model for the fast and accurate independent planar dose calculation for intensity-modulated radiation therapy (IMRT) with MLC. The performance of three different head scatter source models regarding their ability to model head scatter and facilitate planar dose calculation was evaluated. A three-source model, a two-source model and a single-source model were compared in this study. In the planar dose calculation algorithm, in-air fluence distribution was derived from each of the head scatter source models while considering the combination of Jaw and MLC opening. Fluence perturbations due to tongue-and-groove effect, rounded leaf end and leaf transmission were taken into account explicitly. The dose distribution was calculated by convolving the in-air fluence distribution with an experimentally determined pencil-beam kernel. The results were compared with measurements using a diode array and passing rates with 2%/2 mm and 3%/3 mm criteria were reported. It was found that the two-source model achieved the best agreement on head scatter factor calculation. The three-source model and single-source model underestimated head scatter factors for certain symmetric rectangular fields and asymmetric fields, but similar good agreement could be achieved when monitor back scatter effect was incorporated explicitly. All the three source models resulted in comparable average passing rates (>97%) when the 3%/3 mm criterion was selected. The calculation with the single-source model and two-source model was slightly faster than the three-source model due to their simplicity
Energy Technology Data Exchange (ETDEWEB)
Yan Guanghua [Department of Nuclear and Radiological Engineering, University of Florida, Gainesville, FL 32611 (United States); Liu, Chihray; Lu Bo; Palta, Jatinder R; Li, Jonathan G [Department of Radiation Oncology, University of Florida, Gainesville, FL 32610-0385 (United States)
2008-04-21
The purpose of this study was to choose an appropriate head scatter source model for the fast and accurate independent planar dose calculation for intensity-modulated radiation therapy (IMRT) with MLC. The performance of three different head scatter source models regarding their ability to model head scatter and facilitate planar dose calculation was evaluated. A three-source model, a two-source model and a single-source model were compared in this study. In the planar dose calculation algorithm, in-air fluence distribution was derived from each of the head scatter source models while considering the combination of Jaw and MLC opening. Fluence perturbations due to tongue-and-groove effect, rounded leaf end and leaf transmission were taken into account explicitly. The dose distribution was calculated by convolving the in-air fluence distribution with an experimentally determined pencil-beam kernel. The results were compared with measurements using a diode array and passing rates with 2%/2 mm and 3%/3 mm criteria were reported. It was found that the two-source model achieved the best agreement on head scatter factor calculation. The three-source model and single-source model underestimated head scatter factors for certain symmetric rectangular fields and asymmetric fields, but similar good agreement could be achieved when monitor back scatter effect was incorporated explicitly. All the three source models resulted in comparable average passing rates (>97%) when the 3%/3 mm criterion was selected. The calculation with the single-source model and two-source model was slightly faster than the three-source model due to their simplicity.
Equivalent-spherical-shield neutron dose calculations
International Nuclear Information System (INIS)
Russell, G.J.; Robinson, H.
1988-01-01
Neutron doses through 162-cm-thick spherical shields were calculated to be 1090 and 448 mrem/h for regular and magnetite concrete, respectively. These results bracket the measured data, for reinforced regular concrete, of /approximately/600 mrem/h. The calculated fraction of the high-energy (>20 MeV) dose component also bracketed the experimental data. The measured and calculated doses were for a graphite beam stop bombarded with 100 nA of 800-MeV protons. 6 refs., 2 figs., 1 tab
Calculations of optical rotation: Influence of molecular structure
Directory of Open Access Journals (Sweden)
Yu Jia
2012-01-01
Full Text Available Ab initio Hartree-Fock (HF method and Density Functional Theory (DFT were used to calculate the optical rotation of 26 chiral compounds. The effects of theory and basis sets used for calculation, solvents influence on the geometry and values of calculated optical rotation were all discussed. The polarizable continuum model, included in the calculation, did not improve the accuracy effectively, but it was superior to γs. Optical rotation of five or sixmembered of cyclic compound has been calculated and 17 pyrrolidine or piperidine derivatives which were calculated by HF and DFT methods gave acceptable predictions. The nitrogen atom affects the calculation results dramatically, and it is necessary in the molecular structure in order to get an accurate computation result. Namely, when the nitrogen atom was substituted by oxygen atom in the ring, the calculation result deteriorated.
Relativistic Calculations for Be-like Iron
International Nuclear Information System (INIS)
Yang Jianhui; Zhang Jianping; Li Ping; Li Huili
2008-01-01
Relativistic configuration interaction calculations for the states of 1s 2 2s 2 , 1s 2 2s3l (l = s,p,d) and 1s 2 2p3l (l = s,p,d) configurations of iron are carried out using relativistic configuration interaction (RCI) and multi-configuration Dirac-Fock (MCDF) method in the active interaction approach. In the present calculation, a large-scale configuration expansion was used in describing the target states. These results are extensively compared with other available calculative and experimental and observed values, the corresponding present results are in good agreement with experimental and observed values, and some differences are found with other available calculative values. Because more relativistic effects are considered than before, the present results should be more accurate and reliable
Representation and calculation of economic uncertainties
DEFF Research Database (Denmark)
Schjær-Jacobsen, Hans
2002-01-01
Management and decision making when certain information is available may be a matter of rationally choosing the optimal alternative by calculation of the utility function. When only uncertain information is available (which is most often the case) decision-making calls for more complex methods...... of representation and calculation and the basis for choosing the optimal alternative may become obscured by uncertainties of the utility function. In practice, several sources of uncertainties of the required information impede optimal decision making in the classical sense. In order to be able to better handle...... to uncertain economic numbers are discussed. When solving economic models for decision-making purposes calculation of uncertain functions will have to be carried out in addition to the basic arithmetical operations. This is a challenging numerical problem since improper methods of calculation may introduce...
Allswang, John M.
1986-01-01
This article provides two short microcomputer gradebook programs. The programs, written in BASIC for the IBM-PC and Apple II, provide statistical information about class performance and calculate grades either on a normal distribution or based on teacher-defined break points. (JDH)
Calculating Cumulative Binomial-Distribution Probabilities
Scheuer, Ernest M.; Bowerman, Paul N.
1989-01-01
Cumulative-binomial computer program, CUMBIN, one of set of three programs, calculates cumulative binomial probability distributions for arbitrary inputs. CUMBIN, NEWTONP (NPO-17556), and CROSSER (NPO-17557), used independently of one another. Reliabilities and availabilities of k-out-of-n systems analyzed. Used by statisticians and users of statistical procedures, test planners, designers, and numerical analysts. Used for calculations of reliability and availability. Program written in C.
Energy Technology Data Exchange (ETDEWEB)
Ratcliff, Laura E. [Argonne Leadership Computing Facility, Argonne National Laboratory, Lemon IL USA; Mohr, Stephan [Department of Computer Applications in Science and Engineering, Barcelona Supercomputing Center (BSC-CNS), Barcelona Spain; Huhs, Georg [Department of Computer Applications in Science and Engineering, Barcelona Supercomputing Center (BSC-CNS), Barcelona Spain; Deutsch, Thierry [University Grenoble Alpes, INAC-MEM, Grenoble France; CEA, INAC-MEM, Grenoble France; Masella, Michel [Laboratoire de Biologie Structurale et Radiologie, Service de Bioénergétique, Biologie Structurale et Mécanisme, Institut de Biologie et de Technologie de Saclay, CEA Saclay, Gif-sur-Yvette Cedex France; Genovese, Luigi [University Grenoble Alpes, INAC-MEM, Grenoble France; CEA, INAC-MEM, Grenoble France
2016-11-07
During the past decades, quantum mechanical methods have undergone an amazing transition from pioneering investigations of experts into a wide range of practical applications, made by a vast community of researchers. First principles calculations of systems containing up to a few hundred atoms have become a standard in many branches of science. The sizes of the systems which can be simulated have increased even further during recent years, and quantum-mechanical calculations of systems up to many thousands of atoms are nowadays possible. This opens up new appealing possibilities, in particular for interdisciplinary work, bridging together communities of different needs and sensibilities. In this review we will present the current status of this topic, and will also give an outlook on the vast multitude of applications, challenges and opportunities stimulated by electronic structure calculations, making this field an important working tool and bringing together researchers of many different domains.
Calculation of molecular free energies in classical potentials
International Nuclear Information System (INIS)
Farhi, Asaf; Singh, Bipin
2016-01-01
Free energies of molecules can be calculated by quantum chemistry computations or by normal mode classical calculations. However, the first can be computationally impractical for large molecules and the second is based on the assumption of harmonic dynamics. We present a novel, accurate and complete calculation of molecular free energies in standard classical potentials. In this method we transform the molecule by relaxing potential terms which depend on the coordinates of a group of atoms in that molecule and calculate the free energy difference associated with the transformation. Then, since the transformed molecule can be treated as non-interacting systems, the free energy associated with these atoms is analytically or numerically calculated. This two-step calculation can be applied to calculate free energies of molecules or free energy difference between (possibly large) molecules in a general environment. We demonstrate the method in free energy calculations for methanethiol and butane molecules in vacuum and solvent. We suggest the potential application of free energy calculation of chemical reactions in classical molecular simulations. (paper)
International Nuclear Information System (INIS)
Li, D.
1980-01-01
Reviewed is the effect of heat flux of different system parameters on critical density in order to give an initial view on the value of several parameters. A thorough analysis of different equations is carried out to calculate burnout is steam-water flows in uniformly heated tubes, annular, and rectangular channels and rod bundles. Effect of heat flux density distribution and flux twisting on burnout and storage determination according to burnout are commended [ru
Insertion device calculations with mathematica
Energy Technology Data Exchange (ETDEWEB)
Carr, R. [Stanford Synchrotron Radiation Lab., CA (United States); Lidia, S. [Univ. of California, Davis, CA (United States)
1995-02-01
The design of accelerator insertion devices such as wigglers and undulators has usually been aided by numerical modeling on digital computers, using code in high level languages like Fortran. In the present era, there are higher level programming environments like IDL{reg_sign}, MatLab{reg_sign}, and Mathematica{reg_sign} in which these calculations may be performed by writing much less code, and in which standard mathematical techniques are very easily used. The authors present a suite of standard insertion device modeling routines in Mathematica to illustrate the new techniques. These routines include a simple way to generate magnetic fields using blocks of CSEM materials, trajectory solutions from the Lorentz force equations for given magnetic fields, Bessel function calculations of radiation for wigglers and undulators and general radiation calculations for undulators.
Rain Scattering and Co-ordinate Distance Calculation
Directory of Open Access Journals (Sweden)
M. Hajny
1998-12-01
Full Text Available Calculations of scattered field on the rain objects are based on using of Multiple MultiPole (MMP numerical method. Both bi-static scattering function and bi-static scattering cross section are calculated in the plane parallel to Earth surface. The co-ordination area was determined using the simple model of scattering volume [1]. Calculation for frequency 9.595 GHz and antenna elevation of 25Ã‚Â° was done. Obtained results are compared with calculation in accordance to ITU-R recommendation.
40 CFR 1065.675 - CLD quench verification calculations.
2010-07-01
... 40 Protection of Environment 32 2010-07-01 2010-07-01 false CLD quench verification calculations... POLLUTION CONTROLS ENGINE-TESTING PROCEDURES Calculations and Data Requirements § 1065.675 CLD quench verification calculations. Perform CLD quench-check calculations as follows: (a) Perform a CLD analyzer quench...
Soil structure interaction calculations: a comparison of methods
Energy Technology Data Exchange (ETDEWEB)
Wight, L.; Zaslawsky, M.
1976-07-22
Two approaches for calculating soil structure interaction (SSI) are compared: finite element and lumped mass. Results indicate that the calculations with the lumped mass method are generally conservative compared to those obtained by the finite element method. They also suggest that a closer agreement between the two sets of calculations is possible, depending on the use of frequency-dependent soil springs and dashpots in the lumped mass calculations. There is a total lack of suitable guidelines for implementing the lumped mass method of calculating SSI, which leads to the conclusion that the finite element method is generally superior for calculative purposes.
Soil structure interaction calculations: a comparison of methods
International Nuclear Information System (INIS)
Wight, L.; Zaslawsky, M.
1976-01-01
Two approaches for calculating soil structure interaction (SSI) are compared: finite element and lumped mass. Results indicate that the calculations with the lumped mass method are generally conservative compared to those obtained by the finite element method. They also suggest that a closer agreement between the two sets of calculations is possible, depending on the use of frequency-dependent soil springs and dashpots in the lumped mass calculations. There is a total lack of suitable guidelines for implementing the lumped mass method of calculating SSI, which leads to the conclusion that the finite element method is generally superior for calculative purposes
DRY TRANSFER FACILITY CRITICALITY SAFETY CALCULATIONS
International Nuclear Information System (INIS)
C.E. Sanders
2005-01-01
This design calculation updates the previous criticality evaluation for the fuel handling, transfer, and staging operations to be performed in the Dry Transfer Facility (DTF) including the remediation area. The purpose of the calculation is to demonstrate that operations performed in the DTF and RF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Dry Transfer Facility Description Document'' (BSC 2005 [DIRS 173737], p. 3-8). A description of the changes is as follows: (1) Update the supporting calculations for the various Category 1 and 2 event sequences as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2005 [DIRS 171429], Section 7). (2) Update the criticality safety calculations for the DTF staging racks and the remediation pool to reflect the current design. This design calculation focuses on commercial spent nuclear fuel (SNF) assemblies, i.e., pressurized water reactor (PWR) and boiling water reactor (BWR) SNF. U.S. Department of Energy (DOE) Environmental Management (EM) owned SNF is evaluated in depth in the ''Canister Handling Facility Criticality Safety Calculations'' (BSC 2005 [DIRS 173284]) and is also applicable to DTF operations. Further, the design and safety analyses of the naval SNF canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. Also, note that the results for the Monitored Geologic Repository (MGR) Site specific Cask (MSC) calculations are limited to the
Neutron calculation scheme for coupled reactors
International Nuclear Information System (INIS)
Porta, Jacques.
1980-11-01
The CABRI and PHEBUS cores are of the low enrichment rod type in which the fuel is made up of uranium oxide pellets encased in tubular cladding but the SCARABEE core has high enrichment plates, the fuel, an aluminium-uranium alloy (UAl) is metal, rolled into plate form. These three cores in well described rectangular geometry, receive in their centres the very heterogeneous cylindrical test loops (numerous containments of different kinds, large void spaces acting as lagging). After a detailed study of these three reactors, it is found that the search for a calculation scheme (common to the three projects) leads to the elimination of the scattering approximation in our calculations. It is therefore necessary to review the various existing models from a theoretical angle and then to select a particular method, according to the available data processing tools, a choice that will be dictated by the optimization of the parameters: cost in calculation time, difficulties (or ease) of use and accuracy achieved. A problem of experiment interpretation by calculation is dealt with in Chapter 3. The determination of the coupling by calculation is closely linked to the geometrical and energy modelization chosen. But from the experimental angle the determination of the coupling also gives rise to problems with respect to the interpretation of the experimental values obtained by thermal balance determinations, counting of the gamma emission of the fission products of fissile detectors and counting of lanthane 140 in the fuel fission products. The method of calculation is discussed as is the use made of detectors and the counting procedures. In chapter 4, it is not a local modelization that is discussed but an overall one in an original three dimensional calculation [fr
Calculation of resonance integral for fuel cluster
International Nuclear Information System (INIS)
Remsak, S.
1969-01-01
The procedure for calculating the shielding correction, formulated in the previous paper [6], was broadened and applied for a cluster of cylindrical rods. The sam analytical method as in the previous paper was applied. A combination of Gauss method with the method of Almgren and Porn used for solving the same type of integral was used to calculate the geometry functions. CLUSTER code was written for ZUSE-Z-23 computer to calculate the shielding corrections for pairs of fuel rods in the cluster. Computing time for one pair of fuel rods depends on the number of closely placed rod, and for two closely placed rods it is about 3 hours. Calculations were done for clusters containing 7 and 19 UO 2 rods. results show that calculated values of resonance integrals are somewhat higher than the values obtained by Helstrand empirical formula. Taking into account the results for two rods from the previous paper it can be noted that the calculated and empirical values for clusters with 2 and 7 rods are in agreement since the deviations do not exceed the limits of experimental error (±2%). In case of larger cluster with 19 rods deviations are higher than the experimental error. Most probably the calculated values exceed the experimental ones result from the fact that in this paper the shielding correction is calculated only in the region up to 1 keV [sr
Three recent TDHF calculations
International Nuclear Information System (INIS)
Weiss, M.S.
1981-05-01
Three applications of TDHF are discussed. First, vibrational spectra of a post grazing collision 40 Ca nucleus is examined and found to contain many high energy components, qualitatively consistent with recent Orsay experiments. Second, the fusion cross section in energy and angular momentum are calculated for 16 O + 24 Mg to exhibit the parameters of the low l window for this system. A sensitivity of the fusion cross section to the effective two body potential is discussed. Last, a preliminary analysis of 86 Kr + 139 La at E/sub lab/ = 505 MeV calculated in the frozen approximation is displayed, compared to experiment and discussed
Calculation of Critical Temperatures by Empirical Formulae
Directory of Open Access Journals (Sweden)
Trzaska J.
2016-06-01
Full Text Available The paper presents formulas used to calculate critical temperatures of structural steels. Equations that allow calculating temperatures Ac1, Ac3, Ms and Bs were elaborated based on the chemical composition of steel. To elaborate the equations the multiple regression method was used. Particular attention was paid to the collection of experimental data which was required to calculate regression coefficients, including preparation of data for calculation. The empirical data set included more than 500 chemical compositions of structural steel and has been prepared based on information available in literature on the subject.
Relativistic multiple scattering X-alpha calculations
International Nuclear Information System (INIS)
Chermette, H.; Goursot, A.
1986-01-01
The necessity to include self-consistent relativistic corrections in molecular calculations has been pointed out for all compounds involving heavy atoms. Most of the changes in the electronic properties are due to the mass-velocity and the so-called Darwin terms so that the use of Wood and Boring's Hamiltonian is very convenient for this purpose as it can be easily included in MSXalpha programs. Although the spin orbit operator effects are only obtained by perturbation theory, the results compare fairly well with experiment and with other relativistic calculations, namely Hartree-Fock-Slater calculations
IAEA sodium void reactivity benchmark calculations
International Nuclear Information System (INIS)
Hill, R.N.; Finck, P.J.
1992-01-01
In this paper, the IAEA-1 992 ''Benchmark Calculation of Sodium Void Reactivity Effect in Fast Reactor Core'' problem is evaluated. The proposed design is a large axially heterogeneous oxide-fueled fast reactor as described in Section 2; the core utilizes a sodium plenum above the core to enhance leakage effects. The calculation methods used in this benchmark evaluation are described in Section 3. In Section 4, the calculated core performance results for the benchmark reactor model are presented; and in Section 5, the influence of steel and interstitial sodium heterogeneity effects is estimated
BN-600 Phase III benchmark calculations
International Nuclear Information System (INIS)
Hill, R.N.; Grimm, K.N.
2002-01-01
Calculations for a Hexagonal-Z model of the BN-600 reactor with a partial mixed oxide loading, based on a joint IPPE/OBMK loading configuration that contained three uranium enrichment zones and one plutonium enrichment zone in the core, have been performed at ANL. Control-rod worths and reactivity feedback coefficients were calculated using both homogeneous and heterogeneous models. These values were calculated with either first-order perturbation theory methods (Triangle-Z geometry), nodal eigenvalue differences (Hexagonal-Z geometry), or Monte Carlo eigenvalue differences. Both spatially-dependent and region integrated values are shown
Hamming generalized corrector for reactivity calculation
International Nuclear Information System (INIS)
Suescun-Diaz, Daniel; Ibarguen-Gonzalez, Maria C.; Figueroa-Jimenez, Jorge H.
2014-01-01
This work presents the Hamming method generalized corrector for numerically resolving the differential equation of delayed neutron precursor concentration from the point kinetics equations for reactivity calculation, without using the nuclear power history or the Laplace transform. A study was carried out of several correctors with their respective modifiers with different time step calculations, to offer stability and greater precision. Better results are obtained for some correctors than with other existing methods. Reactivity can be calculated with precision of the order h 5 , where h is the time step. (orig.)
Improvements for Monte Carlo burnup calculation
Energy Technology Data Exchange (ETDEWEB)
Shenglong, Q.; Dong, Y.; Danrong, S.; Wei, L., E-mail: qiangshenglong@tsinghua.org.cn, E-mail: d.yao@npic.ac.cn, E-mail: songdr@npic.ac.cn, E-mail: luwei@npic.ac.cn [Nuclear Power Inst. of China, Cheng Du, Si Chuan (China)
2015-07-01
Monte Carlo burnup calculation is development trend of reactor physics, there would be a lot of work to be done for engineering applications. Based on Monte Carlo burnup code MOI, non-fuel burnup calculation methods and critical search suggestions will be mentioned in this paper. For non-fuel burnup, mixed burnup mode will improve the accuracy of burnup calculation and efficiency. For critical search of control rod position, a new method called ABN based on ABA which used by MC21 will be proposed for the first time in this paper. (author)
Shielding calculations for the SNO detector
International Nuclear Information System (INIS)
Earle, E.D.; Wong, P.Y.
1987-05-01
The gamma-ray background into the central D 2 O vessel of the SNO detector due to Th and U in the rock, concrete, and photomultipliers is calculated. A cylindrical geometry and concrete thicknesses of 0.5 and 1 m are assumed. The effect of adding boron to the concrete is also considered. It is concluded that backgrounds from (α,n) reactions can be reduced to the required level. These calculations will assist in finalizing the detector design but additional calculations will be required as new design details become known
Using Financial Calculators in a Business Mathematics Course.
Heller, William H.; Taylor, Monty B.
2000-01-01
Discusses the authors' experiences with integrating financial calculators into a business mathematics course. Presents a brief overview of the operation of financial calculators, reviews some of the more common models, discusses how to use the equation solver utility on other calculators to emulate a financial calculator, and explores the…
42 CFR 102.80 - Calculation of medical benefits.
2010-10-01
... 42 Public Health 1 2010-10-01 2010-10-01 false Calculation of medical benefits. 102.80 Section 102... COMPENSATION PROGRAM Calculation and Payment of Benefits § 102.80 Calculation of medical benefits. In calculating medical benefits, the Secretary will take into consideration all reasonable costs for those...
Isochronous cyclotron closed equilibrium orbit calculation program description
International Nuclear Information System (INIS)
Kiyan, I.N.; Vorozhtsov, S.B.; Tarashkevich, R.
2003-01-01
The Equilibrium Orbit Research Program - EORP, written in C++ with the use of Visual C++ is described. The program is intended for the calculation of the particle rotation frequency and particle kinetic energy in the closed equilibrium orbits of an isochronous cyclotron, where the closed equilibrium orbits are described through the radius and particle momentum angle: r eo (θ) and φ p (θ). The program algorithm was developed on the basis of articles, lecture notes and original analytic calculations. The results of calculations by the EORP were checked and confirmed by using the results of calculations by the numerical methods. The discrepancies between the EORP results and the numerical method results for the calculations of the particle rotation frequency and particle kinetic energy are within the limits of ±1·10 -4 . The EORP results and the numerical method results for the calculations of r eo (θ) and φ p (θ) practically coincide. All this proves the accuracy of calculations by the EORP for the isochronous cyclotrons with the azimuthally varied fields. As is evident from the results of calculations, the program can be used for the calculations of both straight - sector and spiral-sector isochronous cyclotrons. (author)
The Calculation of Flooding Level using CFX Code
International Nuclear Information System (INIS)
Oh, Seo Bin; Kim, Keon Yeop; Lee, Hyung Ho
2015-01-01
The plant design should consider internal flooding by postulated pipe ruptures, component failures, actuation of spray systems, and improper system alignment. The flooding causes failure of safety-related equipment and affects the integrity of the structure. The safety-related equipment should be installed above the flood level for protection against flooding effects. Conservative estimates of the flood level are important when a DBA occurs. The flooding level can be calculated simply applying Bernoulli's equation. However, in this study, a realistic calculation is performed with ANSYS CFX code. In calculation with CFX, air-core vortex phenomena, and turbulent flow can be simulated, which cannot be calculated analytically. The flooding level is evaluated by analytical calculation and CFX analysis for an assumed condition. The flood level is calculated as 0.71m and 1.1m analytically and with CFX simulation, respectively. Comparing the analytical calculation and simulation, they are similar, but the analytical calculation is not conservative. There are many factors reducing the drainage capacity such as air-core vortex, intake of air, and turbulent flow. Therefore, in case of flood level evaluation by analytical calculation, a sufficient safety margin should be considered
Text book of dose calculation for operators
International Nuclear Information System (INIS)
Aoyagi, Haruki; Gonda, Kozo
1979-07-01
This is a text book of dose calculation for the operators of the reprocessing factory of Power Reactor and Nuclear Fuel Development Corporation. The radiations considered are beta-ray and gamma-ray. The method used is a point attenuation nuclear integral method. Radiation sources are considered as the assemblies of point sources. Dose from each point source is calculated, then, total dose is obtained by the integration for all sources. Attenuation is calculated by considering the attenuation owing to distance and the absorption by absorbers. The build-up factor is introduced for the correction for scattered gamma-ray. The build-up factor is given in a table for various scatterers. The operators are able to calculate dose by themselves. The results of integral calculation expressed with formulas are given in graphs. (Kato, T.)
Prenatal radiation exposure. Dose calculation
International Nuclear Information System (INIS)
Scharwaechter, C.; Schwartz, C.A.; Haage, P.; Roeser, A.
2015-01-01
The unborn child requires special protection. In this context, the indication for an X-ray examination is to be checked critically. If thereupon radiation of the lower abdomen including the uterus cannot be avoided, the examination should be postponed until the end of pregnancy or alternative examination techniques should be considered. Under certain circumstances, either accidental or in unavoidable cases after a thorough risk assessment, radiation exposure of the unborn may take place. In some of these cases an expert radiation hygiene consultation may be required. This consultation should comprise the expected risks for the unborn while not perturbing the mother or the involved medical staff. For the risk assessment in case of an in-utero X-ray exposition deterministic damages with a defined threshold dose are distinguished from stochastic damages without a definable threshold dose. The occurrence of deterministic damages depends on the dose and the developmental stage of the unborn at the time of radiation. To calculate the risks of an in-utero radiation exposure a three-stage concept is commonly applied. Depending on the amount of radiation, the radiation dose is either estimated, roughly calculated using standard tables or, in critical cases, accurately calculated based on the individual event. The complexity of the calculation thereby increases from stage to stage. An estimation based on stage one is easily feasible whereas calculations based on stages two and especially three are more complex and often necessitate execution by specialists. This article demonstrates in detail the risks for the unborn child pertaining to its developmental phase and explains the three-stage concept as an evaluation scheme. It should be noted, that all risk estimations are subject to considerable uncertainties.
International Nuclear Information System (INIS)
Pukhov, A.
2003-01-01
CalcHEP is a clone of the CompHEP project which is developed by the author outside of the CompHEP group. CompHEP/CalcHEP are packages for automatic calculations of elementary particle decay and collision properties in the lowest order of perturbation theory. The main idea prescribed into the packages is to make available passing on from the Lagrangian to the final distributions effectively with a high level of automation. According to this, the packages were created as a menu driven user friendly programs for calculations in the interactive mode. From the other side, long-time calculations should be done in the non-interactive regime. Thus, from the beginning CompHEP has a problem of batch calculations. In CompHEP 33.23 the batch session was realized by mean of interactive menu which allows to the user to formulate the task for batch. After that the not-interactive session was launched. This way is too restricted, not flexible, and leads to doubling in programming. In this article I discuss another approach how one can force an interactive program to work in non-interactive mode. This approach was realized in CalcHEP 2.1 disposed on http://theory.sinp.msu.ru/~pukhov/calchep.html
Photoproduction data for heating calculations
International Nuclear Information System (INIS)
Van der Marck, Steven C.; Koning, Arjan J.; Rochman, Dimitri
2008-01-01
For irradiations in a materials test reactor, the prediction of the amount of gamma heating in the reactor is important. Only a good predictive calculation will lead to an irradiation in which the specified temperatures are reached. The photons produced by fission product decay are often missing in spectrum calculations for a reactor, but the contribution of the photons can be computed effectively using engineering correlations for the amount of fission product decay and the ensuing photon spectrum. The prompt photons are usually calculated by a spectrum code based on the underlying nuclear data libraries. For most of the important nuclides, the nuclear data libraries contain data for the photon productions rates. However, there are still many nuclides for which the photon production data are missing, and some of these nuclides contribute to gamma heating. In this paper it is estimated what the contributions to heating are from photon production on nuclides such as 236 U, 238 Pu, 135 I, 135 Xe, 147 Pm, 148 Pm, 148m Pm, and 149 Sm. Also, simple arguments are given to judge the effect from photon production on all other (lumped) fission products, and from 28 Al decay. For all these calculations the High Flux Reactor is used as an example. (authors)
IN-DRIFT MICROBIAL COMMUNITIES MODEL VALIDATION CALCULATIONS
Energy Technology Data Exchange (ETDEWEB)
D.M. Jolley
2001-12-18
The objective and scope of this calculation is to create the appropriate parameter input for MING 1.0 (CSCI 30018 V1.0, CRWMS M&O 1998b) that will allow the testing of the results from the MING software code with both scientific measurements of microbial populations at the site and laboratory and with natural analogs to the site. This set of calculations provides results that will be used in model validation for the ''In-Drift Microbial Communities'' model (CRWMS M&O 2000) which is part of the Engineered Barrier System Department (EBS) process modeling effort that eventually will feed future Total System Performance Assessment (TSPA) models. This calculation is being produced to replace MING model validation output that is effected by the supersession of DTN M09909SPAMINGl.003 using its replacement DTN M00106SPAIDMO 1.034 so that the calculations currently found in the ''In-Drift Microbial Communities'' AMR (CRWMS M&O 2000) will be brought up to date. This set of calculations replaces the calculations contained in sections 6.7.2, 6.7.3 and Attachment I of CRWMS M&O (2000) As all of these calculations are created explicitly for model validation, the data qualification status of all inputs can be considered corroborative in accordance with AP-3.15Q. This work activity has been evaluated in accordance with the AP-2.21 procedure, ''Quality Determinations and Planning for Scientific, Engineering, and Regulatory Compliance Activities'', and is subject to QA controls (BSC 2001). The calculation is developed in accordance with the AP-3.12 procedure, Calculations, and prepared in accordance with the ''Technical Work Plan For EBS Department Modeling FY 01 Work Activities'' (BSC 200 1) which includes controls for the management of electronic data.
In-Drift Microbial Communities Model Validation Calculations
Energy Technology Data Exchange (ETDEWEB)
D. M. Jolley
2001-09-24
The objective and scope of this calculation is to create the appropriate parameter input for MING 1.0 (CSCI 30018 V1.0, CRWMS M&O 1998b) that will allow the testing of the results from the MING software code with both scientific measurements of microbial populations at the site and laboratory and with natural analogs to the site. This set of calculations provides results that will be used in model validation for the ''In-Drift Microbial Communities'' model (CRWMS M&O 2000) which is part of the Engineered Barrier System Department (EBS) process modeling effort that eventually will feed future Total System Performance Assessment (TSPA) models. This calculation is being produced to replace MING model validation output that is effected by the supersession of DTN MO9909SPAMING1.003 using its replacement DTN MO0106SPAIDM01.034 so that the calculations currently found in the ''In-Drift Microbial Communities'' AMR (CRWMS M&O 2000) will be brought up to date. This set of calculations replaces the calculations contained in sections 6.7.2, 6.7.3 and Attachment I of CRWMS M&O (2000) As all of these calculations are created explicitly for model validation, the data qualification status of all inputs can be considered corroborative in accordance with AP-3.15Q. This work activity has been evaluated in accordance with the AP-2.21 procedure, ''Quality Determinations and Planning for Scientific, Engineering, and Regulatory Compliance Activities'', and is subject to QA controls (BSC 2001). The calculation is developed in accordance with the AP-3.12 procedure, Calculations, and prepared in accordance with the ''Technical Work Plan For EBS Department Modeling FY 01 Work Activities'' (BSC 2001) which includes controls for the management of electronic data.
In-Drift Microbial Communities Model Validation Calculation
Energy Technology Data Exchange (ETDEWEB)
D. M. Jolley
2001-10-31
The objective and scope of this calculation is to create the appropriate parameter input for MING 1.0 (CSCI 30018 V1.0, CRWMS M&O 1998b) that will allow the testing of the results from the MING software code with both scientific measurements of microbial populations at the site and laboratory and with natural analogs to the site. This set of calculations provides results that will be used in model validation for the ''In-Drift Microbial Communities'' model (CRWMS M&O 2000) which is part of the Engineered Barrier System Department (EBS) process modeling effort that eventually will feed future Total System Performance Assessment (TSPA) models. This calculation is being produced to replace MING model validation output that is effected by the supersession of DTN MO9909SPAMING1.003 using its replacement DTN MO0106SPAIDM01.034 so that the calculations currently found in the ''In-Drift Microbial Communities'' AMR (CRWMS M&O 2000) will be brought up to date. This set of calculations replaces the calculations contained in sections 6.7.2, 6.7.3 and Attachment I of CRWMS M&O (2000) As all of these calculations are created explicitly for model validation, the data qualification status of all inputs can be considered corroborative in accordance with AP-3.15Q. This work activity has been evaluated in accordance with the AP-2.21 procedure, ''Quality Determinations and Planning for Scientific, Engineering, and Regulatory Compliance Activities'', and is subject to QA controls (BSC 2001). The calculation is developed in accordance with the AP-3.12 procedure, Calculations, and prepared in accordance with the ''Technical Work Plan For EBS Department Modeling FY 01 Work Activities'' (BSC 2001) which includes controls for the management of electronic data.
In-Drift Microbial Communities Model Validation Calculations
International Nuclear Information System (INIS)
Jolley, D.M.
2001-01-01
The objective and scope of this calculation is to create the appropriate parameter input for MING 1.0 (CSCI 30018 V1.0, CRWMS MandO 1998b) that will allow the testing of the results from the MING software code with both scientific measurements of microbial populations at the site and laboratory and with natural analogs to the site. This set of calculations provides results that will be used in model validation for the ''In-Drift Microbial Communities'' model (CRWMS MandO 2000) which is part of the Engineered Barrier System Department (EBS) process modeling effort that eventually will feed future Total System Performance Assessment (TSPA) models. This calculation is being produced to replace MING model validation output that is effected by the supersession of DTN MO9909SPAMING1.003 using its replacement DTN MO0106SPAIDM01.034 so that the calculations currently found in the ''In-Drift Microbial Communities'' AMR (CRWMS MandO 2000) will be brought up to date. This set of calculations replaces the calculations contained in sections 6.7.2, 6.7.3 and Attachment I of CRWMS MandO (2000) As all of these calculations are created explicitly for model validation, the data qualification status of all inputs can be considered corroborative in accordance with AP-3.15Q. This work activity has been evaluated in accordance with the AP-2.21 procedure, ''Quality Determinations and Planning for Scientific, Engineering, and Regulatory Compliance Activities'', and is subject to QA controls (BSC 2001). The calculation is developed in accordance with the AP-3.12 procedure, Calculations, and prepared in accordance with the ''Technical Work Plan For EBS Department Modeling FY 01 Work Activities'' (BSC 2001) which includes controls for the management of electronic data
IN-DRIFT MICROBIAL COMMUNITIES MODEL VALIDATION CALCULATIONS
International Nuclear Information System (INIS)
D.M. Jolley
2001-01-01
The objective and scope of this calculation is to create the appropriate parameter input for MING 1.0 (CSCI 30018 V1.0, CRWMS M andO 1998b) that will allow the testing of the results from the MING software code with both scientific measurements of microbial populations at the site and laboratory and with natural analogs to the site. This set of calculations provides results that will be used in model validation for the ''In-Drift Microbial Communities'' model (CRWMS M andO 2000) which is part of the Engineered Barrier System Department (EBS) process modeling effort that eventually will feed future Total System Performance Assessment (TSPA) models. This calculation is being produced to replace MING model validation output that is effected by the supersession of DTN M09909SPAMINGl.003 using its replacement DTN M00106SPAIDMO 1.034 so that the calculations currently found in the ''In-Drift Microbial Communities'' AMR (CRWMS M andO 2000) will be brought up to date. This set of calculations replaces the calculations contained in sections 6.7.2, 6.7.3 and Attachment I of CRWMS M andO (2000) As all of these calculations are created explicitly for model validation, the data qualification status of all inputs can be considered corroborative in accordance with AP-3.15Q. This work activity has been evaluated in accordance with the AP-2.21 procedure, ''Quality Determinations and Planning for Scientific, Engineering, and Regulatory Compliance Activities'', and is subject to QA controls (BSC 2001). The calculation is developed in accordance with the AP-3.12 procedure, Calculations, and prepared in accordance with the ''Technical Work Plan For EBS Department Modeling FY 01 Work Activities'' (BSC 200 1) which includes controls for the management of electronic data
International Nuclear Information System (INIS)
M. Gross
2004-01-01
The purpose of this scientific analysis is to define the sampled values of stochastic (random) input parameters for (1) rockfall calculations in the lithophysal and nonlithophysal zones under vibratory ground motions, and (2) structural response calculations for the drip shield and waste package under vibratory ground motions. This analysis supplies: (1) Sampled values of ground motion time history and synthetic fracture pattern for analysis of rockfall in emplacement drifts in nonlithophysal rock (Section 6.3 of ''Drift Degradation Analysis'', BSC 2004 [DIRS 166107]); (2) Sampled values of ground motion time history and rock mechanical properties category for analysis of rockfall in emplacement drifts in lithophysal rock (Section 6.4 of ''Drift Degradation Analysis'', BSC 2004 [DIRS 166107]); (3) Sampled values of ground motion time history and metal to metal and metal to rock friction coefficient for analysis of waste package and drip shield damage to vibratory motion in ''Structural Calculations of Waste Package Exposed to Vibratory Ground Motion'' (BSC 2004 [DIRS 167083]) and in ''Structural Calculations of Drip Shield Exposed to Vibratory Ground Motion'' (BSC 2003 [DIRS 163425]). The sampled values are indices representing the number of ground motion time histories, number of fracture patterns and rock mass properties categories. These indices are translated into actual values within the respective analysis and model reports or calculations. This report identifies the uncertain parameters and documents the sampled values for these parameters. The sampled values are determined by GoldSim V6.04.007 [DIRS 151202] calculations using appropriate distribution types and parameter ranges. No software development or model development was required for these calculations. The calculation of the sampled values allows parameter uncertainty to be incorporated into the rockfall and structural response calculations that support development of the seismic scenario for the
Calculating radiation exposure and dose
International Nuclear Information System (INIS)
Hondros, J.
1987-01-01
This paper discusses the methods and procedures used to calculate the radiation exposures and radiation doses to designated employees of the Olympic Dam Project. Each of the three major exposure pathways are examined. These are: gamma irradiation, radon daughter inhalation and radioactive dust inhalation. A further section presents ICRP methodology for combining individual pathway exposures to give a total dose figure. Computer programs used for calculations and data storage are also presented briefly
Numerical calculations near spatial infinity
International Nuclear Information System (INIS)
Zenginoglu, Anil
2007-01-01
After describing in short some problems and methods regarding the smoothness of null infinity for isolated systems, I present numerical calculations in which both spatial and null infinity can be studied. The reduced conformal field equations based on the conformal Gauss gauge allow us in spherical symmetry to calculate numerically the entire Schwarzschild-Kruskal spacetime in a smooth way including spacelike, null and timelike infinity and the domain close to the singularity
Calculation of a toroidal labyrinth shields
International Nuclear Information System (INIS)
Sul'kin, A.G.
1979-01-01
Calculation of protective case with a toroidal labyrinth channel, being one of the main design elements of hose gamma-devices, is presented. The case provides relative isotropic distribution of radiation outside protection limits. The main geometric parameters of the channel are determined: r-radius of the channel hole, rho-bend radius of the channel axis, β-angle of the channel bend. General exposure dose rate of γ-radiation in the detection point at l distance (usually l=100 m during calculations), is also calculated. Differential current dose albedo values have been found for certain combinations of parameters of the labyrinth channel. It is considered for simplification of labyrinth channel calculations, that backward radiation scattering passes, without energy change and isotropically, due to which differential current albedo values of γ-radiation for any incidence angle may be determined from integral albedo current values by the empirie formula
Nuclear structure calculations for astrophysical applications
International Nuclear Information System (INIS)
Moeller, P.; Kratz, K.L.
1992-01-01
Here we present calculated results on such diverse properties as nuclear energy levels, ground-state masses and shapes, β-decay properties and fission-barrier heights. Our approach to these calculations is to use a unified theoretical framework within which the above properties can all be studied. The results are obtained in the macroscopic-microscopic approach in which a microscopic nuclear-structure single-particle model with extensions is combined with a macroscopic model, such as the liquid drop model. In this model the total potential energy of the nucleus may be calculated as a function of shape. The maxima and minima in this function correspond to such features as the ground state, fission saddle points and shape-isomeric states. Various transition rate matrix elements are determined from wave-functions calculated in the single-particle model with pairing and other relevant residual interactions taken into account
Temperature calculation in fire safety engineering
Wickström, Ulf
2016-01-01
This book provides a consistent scientific background to engineering calculation methods applicable to analyses of materials reaction-to-fire, as well as fire resistance of structures. Several new and unique formulas and diagrams which facilitate calculations are presented. It focuses on problems involving high temperature conditions and, in particular, defines boundary conditions in a suitable way for calculations. A large portion of the book is devoted to boundary conditions and measurements of thermal exposure by radiation and convection. The concepts and theories of adiabatic surface temperature and measurements of temperature with plate thermometers are thoroughly explained. Also presented is a renewed method for modeling compartment fires, with the resulting simple and accurate prediction tools for both pre- and post-flashover fires. The final chapters deal with temperature calculations in steel, concrete and timber structures exposed to standard time-temperature fire curves. Useful temperature calculat...
Radiation damage calculations for compound materials
International Nuclear Information System (INIS)
Greenwood, L.R.
1989-01-01
Displacement damage calculations can be performed for 40 elements in the energy range up to 20 MeV with the SPECTER computer code. A recent addition to the code, called SPECOMP, can intermix atomic recoil energy distributions for any four elements to calculate the proper displacement damage for compound materials. The calculations take advantage of the atomic recoil data in the SPECTER libraries, which were determined by the DISCS computer code, using evaluated neutron cross section and angular distribution data in ENDF/B-V. Resultant damage cross sections for any compound can be added to the SPECTER libraries for the routine calculation of displacements in any given neutron field. Users do not require access to neutron cross section files. Results are presented for a variety of fusion materials and a new ceramic superconductor material. Future plans and nuclear data needs are discussed. 11 refs., 6 figs., 1 tab
Direct calculation of wind turbine tip loss
DEFF Research Database (Denmark)
Wood, D.H.; Okulov, Valery; Bhattacharjee, D.
2016-01-01
. We develop three methods for the direct calculation of the tip loss. The first is the computationally expensive calculation of the velocities induced by the helicoidal wake which requires the evaluation of infinite sums of products of Bessel functions. The second uses the asymptotic evaluation......The usual method to account for a finite number of blades in blade element calculations of wind turbine performance is through a tip loss factor. Most analyses use the tip loss approximation due to Prandtl which is easily and cheaply calculated but is known to be inaccurate at low tip speed ratio...
Quantum mechanical methods for calculation of force constants
International Nuclear Information System (INIS)
Mullally, D.J.
1985-01-01
The focus of this thesis is upon the calculation of force constants; i.e., the second derivatives of the potential energy with respect to nuclear displacements. This information is useful for the calculation of molecular vibrational modes and frequencies. In addition, it may be used for the location and characterization of equilibrium and transition state geometries. The methods presented may also be applied to the calculation of electric polarizabilities and infrared and Raman vibrational intensities. Two approaches to this problem are studied and evaluated: finite difference methods and analytical techniques. The most suitable method depends on the type and level of theory used to calculate the electronic wave function. Double point displacement finite differencing is often required for accurate calculation of the force constant matrix. These calculations require energy and gradient calculations on both sides of the geometry of interest. In order to speed up these calculations, a novel method is presented that uses geometry dependent information about the wavefunction. A detailed derivation for the analytical evaluation of force constants with a complete active space multiconfiguration self consistent field wave function is presented
There Is Time for Calculation in Speed Chess, and Calculation Accuracy Increases With Expertise.
Chang, Yu-Hsuan A; Lane, David M
2016-01-01
The recognition-action theory of chess skill holds that expertise in chess is due primarily to the ability to recognize familiar patterns of pieces. Despite its widespread acclaim, empirical evidence for this theory is indirect. One source of indirect evidence is that there is a high correlation between speed chess and standard chess. Assuming that there is little or no time for calculation in speed chess, this high correlation implies that calculation is not the primary factor in standard chess. Two studies were conducted analyzing 100 games of speed chess. In Study 1, we examined the distributions of move times, and the key finding was that players often spent considerable time on a few moves. Moreover, stronger players were more likely than weaker players to do so. Study 2 examined skill differences in calculation by examining poor moves. The stronger players made proportionally fewer blunders (moves that a 2-ply search would have revealed to be errors). Overall, the poor moves made by the weaker players would have required a less extensive search to be revealed as poor moves than the poor moves made by the stronger players. Apparently, the stronger players are searching deeper and more accurately. These results are difficult to reconcile with the view that speed chess does not allow players time to calculate extensively and call into question the assertion that the high correlation between speed chess and standard chess supports recognition-action theory.
“DRYPACK” - a calculation and analysis tool
DEFF Research Database (Denmark)
Andreasen, M.B.; Toftegaard, R.; Schneider, P.
2013-01-01
drying processes. Moreover, it is possible to change the configuration of the dryer by including/changing energy saving components to illustrate the potential of the new configuration. The calculation tool is demonstrated in four different case studies, where the actual energy consumption and possible......“DryPack” is a calculation tool that visualises the energy consumption of airbased and superheated steam drying processes. With “DryPack”, it is possible to add different components to a simple drying process, and thereby increase the flexibility, which makes it possible to analyse the most common...... energy consumption reductions by using “DryPack” are calculated. With the “DryPack” calculation tool, it is possible to calculate four different unit operations with moist air (dehumidification of air, humidification of air, mixing of two air streams, and heating of air). In addition, a Mollier diagram...
McMullan, Miriam; Jones, Ray; Lea, Susan
2011-06-01
Nurses need to be competent and confident in performing drug calculations to ensure patient safety. The purpose of this study is to compare an interactive e-drug calculations package, developed using Cognitive Load Theory as its theoretical framework, with traditional handout learning support on nursing students' drug calculation ability, self-efficacy and support material satisfaction. A cluster randomised controlled trial comparing the e-package with traditional handout learning support was conducted with a September cohort (n=137) and a February cohort (n=92) of second year diploma nursing students. Students from each cohort were geographically dispersed over 3 or 4 independent sites. Students from each cohort were invited to participate, halfway through their second year, before and after a 12 week clinical practice placement. During their placement the intervention group received the e-drug calculations package while the control group received traditional 'handout' support material. Drug calculation ability and self-efficacy tests were given to the participants pre- and post-intervention. Participants were given the support material satisfaction scale post-intervention. Students in both cohorts randomised to e-learning were more able to perform drug calculations than those receiving the handout (September: mean 48.4% versus 34.7%, p=0.027; February: mean 47.6% versus 38.3%, p=0.024). February cohort students using the e-package were more confident in performing drug calculations than those students using handouts (self-efficacy mean 56.7% versus 45.8%, p=0.022). There was no difference in improved self-efficacy between intervention and control for students in the September cohort. Students who used the package were more satisfied with its use than the students who used the handout (mean 29.6 versus 26.5, p=0.001), particularly with regard to the package enhancing their learning (p=0.023), being an effective way to learn (p=0.005), providing practice and
Electronics Environmental Benefits Calculator
U.S. Environmental Protection Agency — The Electronics Environmental Benefits Calculator (EEBC) was developed to assist organizations in estimating the environmental benefits of greening their purchase,...
Assessment of seismic margin calculation methods
International Nuclear Information System (INIS)
Kennedy, R.P.; Murray, R.C.; Ravindra, M.K.; Reed, J.W.; Stevenson, J.D.
1989-03-01
Seismic margin review of nuclear power plants requires that the High Confidence of Low Probability of Failure (HCLPF) capacity be calculated for certain components. The candidate methods for calculating the HCLPF capacity as recommended by the Expert Panel on Quantification of Seismic Margins are the Conservative Deterministic Failure Margin (CDFM) method and the Fragility Analysis (FA) method. The present study evaluated these two methods using some representative components in order to provide further guidance in conducting seismic margin reviews. It is concluded that either of the two methods could be used for calculating HCLPF capacities. 21 refs., 9 figs., 6 tabs
Recent skyshine calculations at Jefferson Lab
International Nuclear Information System (INIS)
Degtyarenko, P.
1997-01-01
New calculations of the skyshine dose distribution of neutrons and secondary photons have been performed at Jefferson Lab using the Monte Carlo method. The dose dependence on neutron energy, distance to the neutron source, polar angle of a source neutron, and azimuthal angle between the observation point and the momentum direction of a source neutron have been studied. The azimuthally asymmetric term in the skyshine dose distribution is shown to be important in the dose calculations around high-energy accelerator facilities. A parameterization formula and corresponding computer code have been developed which can be used for detailed calculations of the skyshine dose maps
RA-0 reactor. New neutronic calculations
International Nuclear Information System (INIS)
Rumis, D.; Leszczynski, F.
1990-01-01
An updating of the neutronic calculations performed at the RA-0 reactor, located at the Natural, Physical and Exact Sciences Faculty of Cordoba National University, are herein described. The techniques used for the calculation of a reactor like the RA-0 allows prediction in detail of the flux behaviour in the core's interior and in the reflector, which will be helpful for experiments design. In particular, the use of WIMSD4 code to make calculations on the reactor implies a novelty in the possible applications of this code to solve the problems that arise in practice. (Author) [es
International Nuclear Information System (INIS)
Mochamad Nasrullah; Nurlaila
2014-01-01
The government plan to build Experimental Power Reactor (EPR) requires measurable economic assessment. The purpose of the study was to recalculate Gas Turbine High Temperature Reactor of 300 MWe (GTHTR 300) and compare the results with reference data. Then calculate generation cost of GTHTR 3, 5 and 10 MWe using the scale factor calculation. The methodology used is covered the generation cost calculation using the Mini G4Econs spread sheet models published by IAEA. Result of the verification calculation showed that a relatively similar, which means that the calculation model could be used to calculate for same other cases. Afterward, using scale factor, smaller scale reactor could be calculated. The calculation result show that electricity generation cost of SMR-HTR type with load factor 90% and discount rate 10% for power capacity 3, 5 and 10 MWe are 29.5, 22.68 and 16.17 cents$/kWh respectively. However, because the EPR is planning to be built as a non-commercial power reactors, then 5 % and 3 % of discount rate could be chosen, each of those discount rate will result electricity generation cost of 10.37 cents$/kWh and 8.56 cents$/kWh respectively. These result could be considered by the government for developing SMR type of HTR. (author)
Evaluation of students' knowledge about paediatric dosage calculations.
Özyazıcıoğlu, Nurcan; Aydın, Ayla İrem; Sürenler, Semra; Çinar, Hava Gökdere; Yılmaz, Dilek; Arkan, Burcu; Tunç, Gülseren Çıtak
2018-01-01
Medication errors are common and may jeopardize the patient safety. As paediatric dosages are calculated based on the child's age and weight, risk of error in dosage calculations is increasing. In paediatric patients, overdose drug prescribed regardless of the child's weight, age and clinical picture may lead to excessive toxicity and mortalities while low doses may delay the treatment. This study was carried out to evaluate the knowledge of nursing students about paediatric dosage calculations. This research, which is of retrospective type, covers a population consisting of all the 3rd grade students at the bachelor's degree in May, 2015 (148 students). Drug dose calculation questions in exam papers including 3 open ended questions on dosage calculation problems, addressing 5 variables were distributed to the students and their responses were evaluated by the researchers. In the evaluation of the data, figures and percentage distribution were calculated and Spearman correlation analysis was applied. Exam question on the dosage calculation based on child's age, which is the most common method in paediatrics, and which ensures right dosages and drug dilution was answered correctly by 87.1% of the students while 9.5% answered it wrong and 3.4% left it blank. 69.6% of the students was successful in finding the safe dose range, and 79.1% in finding the right ratio/proportion. 65.5% of the answers with regard to Ml/dzy calculation were correct. Moreover, student's four operation skills were assessed and 68.2% of the students were determined to have found the correct answer. When the relation among the questions on medication was examined, a significant relation (correlation) was determined between them. It is seen that in dosage calculations, the students failed mostly in calculating ml/dzy (decimal). This result means that as dosage calculations are based on decimal values, calculations may be ten times erroneous when the decimal point is placed wrongly. Moreover, it
Computer automation for protection factor calculations of buildings
International Nuclear Information System (INIS)
Farafat, M.A.Z.; Madian, A.H.
2011-01-01
The protection factors of buildings are different according to the constructional and architectural specifications. Uk and USA performed a calculation using manual method to calculate the protection factor for any building which may protect the people in it from gamma rays and fall-out.The manual calculation method is very complex which is very difficult to use, for that reason the researchers simplify this method in proposed form which will be easy to understand and use. Also the researchers have designed a computer program ,in visual basic, to calculate the different protection factors for buildings. The program aims to provide the missing time in the calculation processes to calculate the protection in some spaces for any building through entering specifications data for any building .The program will modify the protection factor in very short time which will save the effort and time in comparison with the manual calculation.
Probability calculations for three-part mineral resource assessments
Ellefsen, Karl J.
2017-06-27
Three-part mineral resource assessment is a methodology for predicting, in a specified geographic region, both the number of undiscovered mineral deposits and the amount of mineral resources in those deposits. These predictions are based on probability calculations that are performed with computer software that is newly implemented. Compared to the previous implementation, the new implementation includes new features for the probability calculations themselves and for checks of those calculations. The development of the new implementation lead to a new understanding of the probability calculations, namely the assumptions inherent in the probability calculations. Several assumptions strongly affect the mineral resource predictions, so it is crucial that they are checked during an assessment. The evaluation of the new implementation leads to new findings about the probability calculations,namely findings regarding the precision of the computations,the computation time, and the sensitivity of the calculation results to the input.
Numeric calculation of celestial bodies with spreadsheet analysis
Koch, Alexander
2016-04-01
The motion of the planets and moons in our solar system can easily be calculated for any time by the Kepler laws of planetary motion. The Kepler laws are a special case of the gravitational law of Newton, especially if you consider more than two celestial bodies. Therefore it is more basic to calculate the motion by using the gravitational law. But the problem is, that by gravitational law it is not possible to calculate the state of motion with only one step of calculation. The motion has to be numerical calculated for many time intervalls. For this reason, spreadsheet analysis is helpful for students. Skills in programmes like Excel, Calc or Gnumeric are important in professional life and can easily be learnt by students. These programmes can help to calculate the complex motions with many intervalls. The more intervalls are used, the more exact are the calculated orbits. The sutdents will first get a quick course in Excel. After that they calculate with instructions the 2-D-coordinates of the orbits of Moon and Mars. Step by step the students are coding the formulae for calculating physical parameters like coordinates, force, acceleration and velocity. The project is limited to 4 weeks or 8 lessons. So the calcualtion will only include the calculation of one body around the central mass like Earth or Sun. The three-body problem can only be shortly discussed at the end of the project.
Temperature Calculations in the Coastal Modeling System
2017-04-01
ERDC/CHL CHETN-IV-110 April 2017 Approved for public release; distribution is unlimited . Temperature Calculations in the Coastal Modeling...tide) and river discharge at model boundaries, wave radiation stress, and wind forcing over a model computational domain. Physical processes calculated...calculated in the CMS using the following meteorological parameters: solar radiation, cloud cover, air temperature, wind speed, and surface water temperature
Calculation methods for determining dose equivalent
International Nuclear Information System (INIS)
Endres, G.W.R.; Tanner, J.E.; Scherpelz, R.I.; Hadlock, D.E.
1987-11-01
A series of calculations of neutron fluence as a function of energy in an anthropomorphic phantom was performed to develop a system for determining effective dose equivalent for external radiation sources. Critical organ dose equivalents are calculated and effective dose equivalents are determined using ICRP-26 [1] methods. Quality factors based on both present definitions and ICRP-40 definitions are used in the analysis. The results of these calculations are presented and discussed. The effective dose equivalent determined using ICRP-26 methods is significantly smaller than the dose equivalent determined by traditional methods. No existing personnel dosimeter or health physics instrument can determine effective dose equivalent. At the present time, the conversion of dosimeter response to dose equivalent is based on calculations for maximal or ''cap'' values using homogeneous spherical or cylindrical phantoms. The evaluated dose equivalent is, therefore, a poor approximation of the effective dose equivalent as defined by ICRP Publication 26. 3 refs., 2 figs., 1 tab
Magnetic Field Grid Calculator
National Oceanic and Atmospheric Administration, Department of Commerce — The Magnetic Field Properties Calculator will computes the estimated values of Earth's magnetic field(declination, inclination, vertical component, northerly...
Radar Signature Calculation Facility
Federal Laboratory Consortium — FUNCTION: The calculation, analysis, and visualization of the spatially extended radar signatures of complex objects such as ships in a sea multipath environment and...
Dose calculation system for remotely supporting radiotherapy
International Nuclear Information System (INIS)
Saito, K.; Kunieda, E.; Narita, Y.; Kimura, H.; Hirai, M.; Deloar, H. M.; Kaneko, K.; Ozaki, M.; Fujisaki, T.; Myojoyama, A.; Saitoh, H.
2005-01-01
The dose calculation system IMAGINE is being developed keeping in mind remotely supporting external radiation therapy using photon beams. The system is expected to provide an accurate picture of the dose distribution in a patient body, using a Monte Carlo calculation that employs precise models of the patient body and irradiation head. The dose calculation will be performed utilising super-parallel computing at the dose calculation centre, which is equipped with the ITBL computer, and the calculated results will be transferred through a network. The system is intended to support the quality assurance of current, widely carried out radiotherapy and, further, to promote the prevalence of advanced radiotherapy. Prototypes of the modules constituting the system have already been constructed and used to obtain basic data that are necessary in order to decide on the concrete design of the system. The final system will be completed in 2007. (authors)
Core physics calculation and analysis for SNRE
International Nuclear Information System (INIS)
Xie Jiachun; Zhao Shouzhi; Jia Baoshan
2010-01-01
Five different precise calculation models have been set up for Small Nuclear Rocket Engine (SNRE) core based on MCNP code, and then the effective multiplication constant, drum control worth and power distribution were calculated. The results from different models indicate that the model in which elements are homogeneous could be used in the reactivity calculation, but a detailed description of elements have to be used in the element internal power distribution calculation. The results of physics parameters show that the basic characteristics of SNRE are reasonable. The drum control worth is sufficient. The power distribution is symmetrical and reasonable. All of the parameters can satisfy the design requirement. (authors)
Coil protection calculator for TFTR
International Nuclear Information System (INIS)
Marsala, R.J.; Lawson, J.E.; Persing, R.G.; Senko, T.R.; Woolley, R.D.
1989-01-01
A new coil protection system (CPS) is being developed to replace the existing TFTR magnetic coil fault detector. The existing fault detector sacrifices TFTR operating capability for simplicity. The new CPS, when installed in October of 1988, will permit operation up to the actual coil stress limits parameters in real-time. The computation will be done in a microprocessor based Coil Protection Calculator (CPC) currently under construction at PPL. THe new CPC will allow TFTR to operate with higher plasma currents and will permit the optimization of pulse repetition rates. The CPC will provide real-time estimates of critical coil and bus temperatures and stresses based on real-time redundant measurements of coil currents, coil cooling water inlet temperature, and plasma current. The critical parameter calculations are compared to prespecified limits. If these limits are reached or exceeded, protection action will be initiated to a hard wired control system (HCS), which will shut down the power supplies. The CPC consists of a redundant VME based microprocessor system which will sample all input data and compute all stress quantities every ten milliseconds. Thermal calculations will be approximated every 10ms with an exact solution occurring every second. The CPC features continuous cross-checking of redundant input signal, automatic detection of internal failure modes, monitoring and recording of calculated results, and a quick, functional verification of performance via an internal test system. (author)
Impact of neutron resonance treatments on reactor calculation
International Nuclear Information System (INIS)
Leszczynski, F.
1988-01-01
The neutron resonance treatment on reactor calculation is one of the not completely resolved problems of reactor theory. The calculation required on design, fuel management and accident analysis of nuclear reactors contains adjust coefficients and semi-empirical values introduced on the computer codes; these values are obtained comparing calculation results with experimental values and more exact calculation results. This is made when the characteristics of the analyzed system are such that this type of comparisons are possible. The impact that one fixed resonance treatment method have on the final evaluation of physics reactor parameters, reactivity, power distribution, etc., is useful to know. In this work, the differences between calculated parameters with two different methods of resonance treatment in cell calculations are shown. It is concluded that improvements on resonance treatment are necessary for growing the reliability on core calculations results. Finally, possible improvements, easy to implement in current computer codes, are presented. (Author) [es
CONSUL code package application for LMFR core calculations
Energy Technology Data Exchange (ETDEWEB)
Chibinyaev, A.V.; Teplov, P.S.; Frolova, M.V. [RNC ' Kurchatovskiy institute' , Kurchatov sq.1, Moscow (Russian Federation)
2008-07-01
CONSUL code package designed for the calculation of reactor core characteristics has been developed at the beginning of 90's. The calculation of nuclear reactor core characteristics is carried out on the basis of correlated neutron, isotope and temperature distributions. The code package has been generally used for LWR core characteristics calculations. At present CONSUL code package was adapted to calculate liquid metal fast reactors (LMFR). The comparisons with IAEA computational test 'Evaluation of benchmark calculations on a fast power reactor core with near zero sodium void effect' and BN-1800 testing calculations are presented in the paper. The IAEA benchmark core is based on the innovative core concept with sodium plenum above the core BN-800. BN-1800 core is the next development step which is foreseen for the Russian fast reactor concept. The comparison of the operational parameters has shown good agreement and confirms the possibility of CONSUL code package application for LMFR core calculation. (authors)
International Nuclear Information System (INIS)
Devooght, J.; Lefvert, T.; Stankiewiez, J.
1981-01-01
This chapter deals with the work done in reactor dynamics within the Coordinated Research Program on Transport Theory and Advanced Reactor Calculations by three groups in Belgium, Poland, Sweden and Italy. Discretization methods in diffusion theory, collision probability methods in time-dependent neutron transport and singular perturbation method are represented in this paper
Sussman, Allen; Taylor, Elizabeth J; Patel, Mona; Ward, Jeanne; Alva, Shridhara; Lawrence, Andrew; Ng, Ronald
2012-03-01
Patients consider multiple parameters in adjusting prandial insulin doses for optimal glycemic control. Difficulties in calculations can lead to incorrect doses or induce patients to administer fixed doses, rely on empirical estimates, or skip boluses. A multicenter study was conducted with 205 diabetes subjects who were on multiple daily injections of rapid/ short-acting insulin. Using the formula provided, the subjects manually calculated two prandial insulin doses based on one high and one normal glucose test result, respectively. They also determined the two doses using the FreeStyle InsuLinx Blood Glucose Monitoring System, which has a built-in, automated bolus calculator. After dose determinations, the subjects completed opinion surveys. Of the 409 insulin doses manually calculated by the subjects, 256 (63%) were incorrect. Only 23 (6%) of the same 409 dose determinations were incorrect using the meter, and these errors were due to either confirmed or potential deviations from the study instructions by the subjects when determining dose with meter. In the survey, 83% of the subjects expressed more confidence in the meter-calculated doses than the manually calculated doses. Furthermore, 87% of the subjects preferred to use the meter than manual calculation to determine prandial insulin doses. Insulin-using patients made errors in more than half of the manually calculated insulin doses. Use of the automated bolus calculator in the FreeStyle InsuLinx meter minimized errors in dose determination. The patients also expressed confidence and preference for using the meter. This may increase adherence and help optimize the use of mealtime insulin. © 2012 Diabetes Technology Society.
Compilation report of VHTRC temperature coefficient benchmark calculations
Energy Technology Data Exchange (ETDEWEB)
Yasuda, Hideshi; Yamane, Tsuyoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1995-11-01
A calculational benchmark problem has been proposed by JAERI to an IAEA Coordinated Research Program, `Verification of Safety Related Neutronic Calculation for Low-enriched Gas-cooled Reactors` to investigate the accuracy of calculation results obtained by using codes of the participating countries. This benchmark is made on the basis of assembly heating experiments at a pin-in block type critical assembly, VHTRC. Requested calculation items are the cell parameters, effective multiplication factor, temperature coefficient of reactivity, reaction rates, fission rate distribution, etc. Seven institutions from five countries have joined the benchmark works. Calculation results are summarized in this report with some remarks by the authors. Each institute analyzed the problem by applying the calculation code system which was prepared for the HTGR development of individual country. The values of the most important parameter, k{sub eff}, by all institutes showed good agreement with each other and with the experimental ones within 1%. The temperature coefficient agreed within 13%. The values of several cell parameters calculated by several institutes did not agree with the other`s ones. It will be necessary to check the calculation conditions again for getting better agreement. (J.P.N.).
Parallel processing of neutron transport in fuel assembly calculation
International Nuclear Information System (INIS)
Song, Jae Seung
1992-02-01
Group constants, which are used for reactor analyses by nodal method, are generated by fuel assembly calculations based on the neutron transport theory, since one or a quarter of the fuel assembly corresponds to a unit mesh in the current nodal calculation. The group constant calculation for a fuel assembly is performed through spectrum calculations, a two-dimensional fuel assembly calculation, and depletion calculations. The purpose of this study is to develop a parallel algorithm to be used in a parallel processor for the fuel assembly calculation and the depletion calculations of the group constant generation. A serial program, which solves the neutron integral transport equation using the transmission probability method and the linear depletion equation, was prepared and verified by a benchmark calculation. Small changes from the serial program was enough to parallelize the depletion calculation which has inherent parallel characteristics. In the fuel assembly calculation, however, efficient parallelization is not simple and easy because of the many coupling parameters in the calculation and data communications among CPU's. In this study, the group distribution method is introduced for the parallel processing of the fuel assembly calculation to minimize the data communications. The parallel processing was performed on Quadputer with 4 CPU's operating in NURAD Lab. at KAIST. Efficiencies of 54.3 % and 78.0 % were obtained in the fuel assembly calculation and depletion calculation, respectively, which lead to the overall speedup of about 2.5. As a result, it is concluded that the computing time consumed for the group constant generation can be easily reduced by parallel processing on the parallel computer with small size CPU's
Description of the CAREM Reactor Neutronic Calculation Codes
International Nuclear Information System (INIS)
Villarino, Eduardo; Hergenreder, Daniel
2000-01-01
In this work is described the neutronic calculation line used to design the CAREM reactor.A description of the codes used and the interfaces between the different programs are presented.Both, the normal calculation line and the alternative or verification calculation line are included.The calculation line used to obtain the kinetics parameters (effective delayed-neutron fraction and prompt-neutron lifetime) is also included
Two dimensional burn-up calculation of TRIGA core
International Nuclear Information System (INIS)
Persic, A.; Ravnik, M.; Slavic, S.
1996-01-01
TRIGLAV is a new computer program for burn-up calculation of mixed core of research reactors. The code is based on diffusion model in two dimensions and iterative procedure is applied for its solution. The material data used in the model are calculated with the transport program WIMS. In regard to fission density distribution and energy produced by the reactor the burn-up increment of fuel elements is determined. In this paper the calculation model of diffusion constants and burn-up calculation are described and some results of calculations for TRIGA MARK II reactor are presented. (author)
Local expressions for one-loop calculations
International Nuclear Information System (INIS)
Wasson, D.A.; Koonin, S.E.
1991-01-01
We develop local expressions for the contributions of the short-wavelength vacuum modes to the one-loop vacuum energy. These expressions significantly improve the convergence properties of various ''brute-force'' calculational methods. They also provide a continuous series of approximations that interpolate between the brute-force calculations and the derivative expansion
Calculated apparent yields of rare gas fission products
International Nuclear Information System (INIS)
Delucchi, A.A.
1975-01-01
The apparent fission yield of the rare gas fission products from four mass chains is calculated as a function of separation time for six different fissioning systems. A plot of the calculated fission yield along with a one standard deviation error band is given for each rare gas fission product and for each fissioning system. Those parameters in the calculation that were major contributors to the calculated standard deviation at each separation time were identified and the results presented on a separate plot. To extend the usefulness of these calculations as new and better values for the input parameters become available, a third plot was generated for each system which shows how sensitive the derived fission yield is to a change in any given parameter used in the calculation. (U.S.)
Adaptation of GRS calculation codes for Soviet reactors
International Nuclear Information System (INIS)
Langenbuch, S.; Petri, A.; Steinborn, J.; Stenbok, I.A.; Suslow, A.I.
1994-01-01
The use of ATHLET for incident calculation of WWER has been tested and verified in numerous calculations. Further adaptation may be needed for the WWER 1000 plants. Coupling ATHLET with the 3D nuclear model BIPR-8 for WWER cores clearly improves studies of the influence of neutron kinetics. In the case of FBMK reactors ATHLET calculations show that typical incidents in the complex RMBK reactors can be calculated even though verification still has to be worked on. Results of the 3D-core model QUABOX/CUBBOX-HYCA show good correlation of calculated and measured values in reactor plants. Calculations carried out to date were used to check essential parameters influencing RBMK core behaviour especially dependence of effective voidre activity on the number of control rods. (orig./HP) [de
Agriculture-related radiation dose calculations
International Nuclear Information System (INIS)
Furr, J.M.; Mayberry, J.J.; Waite, D.A.
1987-10-01
Estimates of radiation dose to the public must be made at each stage in the identification and qualification process leading to siting a high-level nuclear waste repository. Specifically considering the ingestion pathway, this paper examines questions of reliability and adequacy of dose calculations in relation to five stages of data availability (geologic province, region, area, location, and mass balance) and three methods of calculation (population, population/food production, and food production driven). Calculations were done using the model PABLM with data for the Permian and Palo Duro Basins and the Deaf Smith County area. Extra effort expended in gathering agricultural data at succeeding environmental characterization levels does not appear justified, since dose estimates do not differ greatly; that effort would be better spent determining usage of food types that contribute most to the total dose; and that consumption rate and the air dispersion factor are critical to assessment of radiation dose via the ingestion pathway. 17 refs., 9 figs., 32 tabs
Good Practices in Free-energy Calculations
Pohorille, Andrew; Jarzynski, Christopher; Chipot, Christopher
2013-01-01
As access to computational resources continues to increase, free-energy calculations have emerged as a powerful tool that can play a predictive role in drug design. Yet, in a number of instances, the reliability of these calculations can be improved significantly if a number of precepts, or good practices are followed. For the most part, the theory upon which these good practices rely has been known for many years, but often overlooked, or simply ignored. In other cases, the theoretical developments are too recent for their potential to be fully grasped and merged into popular platforms for the computation of free-energy differences. The current best practices for carrying out free-energy calculations will be reviewed demonstrating that, at little to no additional cost, free-energy estimates could be markedly improved and bounded by meaningful error estimates. In energy perturbation and nonequilibrium work methods, monitoring the probability distributions that underlie the transformation between the states of interest, performing the calculation bidirectionally, stratifying the reaction pathway and choosing the most appropriate paradigms and algorithms for transforming between states offer significant gains in both accuracy and precision. In thermodynamic integration and probability distribution (histogramming) methods, properly designed adaptive techniques yield nearly uniform sampling of the relevant degrees of freedom and, by doing so, could markedly improve efficiency and accuracy of free energy calculations without incurring any additional computational expense.
General-purpose software for science technology calculation
International Nuclear Information System (INIS)
Aikawa, Hiroshi
1999-01-01
We have developed many general-purpose softwares for parallel processing of science technology calculation. This paper reported six softwares such as STA (Seamless Thinking Aid) basic soft, parallel numerical computation library, grid formation software for parallel computer, real-time visualizing system, parallel benchmark test system and object-oriented parallel programing method. STA is a user interface software to perform a total environment for parallel programing, a network computing environment for various parallel computers and a desktop computing environment via Web. Some examples using the above softwares are explained. One of them is a simultaneous parallel calculation of both analysis of flow and structure of supersonic transport to design of them. The other is various kinds of computer parallel calculations for nuclear fusion reaction such as a molecular dynamic calculation and a calculation of reactor structure and fluid. These softs are opened to the public by the home page {http://guide.tokai.jaeri.go.jp/ccse/}. (S.Y.)
International Nuclear Information System (INIS)
Weill, J.; Tellier; Bonnemay; Craigne; Chareton; Di Falco
1969-02-01
After a definition of hybrid calculation (combination of analogue and digital calculation) with a distinction between series and parallel hybrid computing, and a description of a hybrid computer structure and of task sharing between computers, this course proposes a description of hybrid hardware used in Saclay and Cadarache computing centres, and of operations performed by these systems. The next part addresses issues related to programming languages and software. The fourth part describes how a problem is organised for its processing on these computers. Methods of hybrid analysis are then addressed: resolution of optimisation problems, of partial differential equations, and of integral equations by means of different methods (gradient, maximum principle, characteristics, functional approximation, time slicing, Monte Carlo, Neumann iteration, Fischer iteration)
Nuclear data preparation and discrete ordinates calculation
International Nuclear Information System (INIS)
Carmignani, B.
1980-01-01
These lectures deal with the use of the GAM-GATHER and GAM-THERMOS chains for the calculation of lattice cross sections and within use of the discrete ordinates one dimensional ANISN code for the calculation of criticality and flux distribution of the cell and of the whole reactor. As an example the codes are applied to the calculation of a PWR. Results of different approximations are compared. (author)
40 CFR 1065.650 - Emission calculations.
2010-07-01
... field testing, you may calculate the ratio of total mass to total work, where these individual values... negative work rate values in the integration to calculate total work from that work path. Some work paths may result in a negative total work. Include negative total work values from any work path in the...
Parallel computational in nuclear group constant calculation
International Nuclear Information System (INIS)
Su'ud, Zaki; Rustandi, Yaddi K.; Kurniadi, Rizal
2002-01-01
In this paper parallel computational method in nuclear group constant calculation using collision probability method will be discuss. The main focus is on the calculation of collision matrix which need large amount of computational time. The geometry treated here is concentric cylinder. The calculation of collision probability matrix is carried out using semi analytic method using Beckley Naylor Function. To accelerate computation speed some computer parallel used to solve the problem. We used LINUX based parallelization using PVM software with C or fortran language. While in windows based we used socket programming using DELPHI or C builder. The calculation results shows the important of optimal weight for each processor in case there area many type of processor speed
Achieving High Accuracy in Calculations of NMR Parameters
DEFF Research Database (Denmark)
Faber, Rasmus
quantum chemical methods have been developed, the calculation of NMR parameters with quantitative accuracy is far from trivial. In this thesis I address some of the issues that makes accurate calculation of NMR parameters so challenging, with the main focus on SSCCs. High accuracy quantum chemical......, but no programs were available to perform such calculations. As part of this thesis the CFOUR program has therefore been extended to allow the calculation of SSCCs using the CC3 method. CC3 calculations of SSCCs have then been performed for several molecules, including some difficult cases. These results show...... vibrations must be included. The calculation of vibrational corrections to NMR parameters has been reviewed as part of this thesis. A study of the basis set convergence of vibrational corrections to nuclear shielding constants has also been performed. The basis set error in vibrational correction...
Benchmark calculation of subchannel analysis codes
International Nuclear Information System (INIS)
1996-02-01
In order to evaluate the analysis capabilities of various subchannel codes used in thermal-hydraulic design of light water reactors, benchmark calculations were performed. The selected benchmark problems and major findings obtained by the calculations were as follows: (1)As for single-phase flow mixing experiments between two channels, the calculated results of water temperature distribution along the flow direction were agreed with experimental results by tuning turbulent mixing coefficients properly. However, the effect of gap width observed in the experiments could not be predicted by the subchannel codes. (2)As for two-phase flow mixing experiments between two channels, in high water flow rate cases, the calculated distributions of air and water flows in each channel were well agreed with the experimental results. In low water flow cases, on the other hand, the air mixing rates were underestimated. (3)As for two-phase flow mixing experiments among multi-channels, the calculated mass velocities at channel exit under steady-state condition were agreed with experimental values within about 10%. However, the predictive errors of exit qualities were as high as 30%. (4)As for critical heat flux(CHF) experiments, two different results were obtained. A code indicated that the calculated CHF's using KfK or EPRI correlations were well agreed with the experimental results, while another code suggested that the CHF's were well predicted by using WSC-2 correlation or Weisman-Pei mechanistic model. (5)As for droplets entrainment and deposition experiments, it was indicated that the predictive capability was significantly increased by improving correlations. On the other hand, a remarkable discrepancy between codes was observed. That is, a code underestimated the droplet flow rate and overestimated the liquid film flow rate in high quality cases, while another code overestimated the droplet flow rate and underestimated the liquid film flow rate in low quality cases. (J.P.N.)
Thermohydraulic calculations of PWR primary circuits
International Nuclear Information System (INIS)
Botelho, D.A.
1984-01-01
Some mathematical and numerical models from Retran computer codes aiming to simulate reactor transients, are presented. The equations used for calculating one-dimensional flow are integrated using mathematical methods from Flash code, with steam code to correlate the variables from thermodynamic state. The algorithm obtained was used for calculating a PWR reactor. (E.G.) [pt
Fast reactor calculational route for Pu burning core design
Energy Technology Data Exchange (ETDEWEB)
Hunter, S. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center
1998-01-01
This document provides a description of a calculational route, used in the Reactor Physics Research Section for sensitivity studies and initial design optimization calculations for fast reactor cores. The main purpose in producing this document was to provide a description of and user guides to the calculational methods, in English, as an aid to any future user of the calculational route who is (like the author) handicapped by a lack of literacy in Japanese. The document also provides for all users a compilation of information on the various parts of the calculational route, all in a single reference. In using the calculational route (to model Pu burning reactors) the author identified a number of areas where an improvement in the modelling of the standard calculational route was warranted. The document includes comments on and explanations of the modelling assumptions in the various calculations. Practical information on the use of the calculational route and the computer systems is also given. (J.P.N.)
Gaseous Nitrogen Orifice Mass Flow Calculator
Ritrivi, Charles
2013-01-01
The Gaseous Nitrogen (GN2) Orifice Mass Flow Calculator was used to determine Space Shuttle Orbiter Water Spray Boiler (WSB) GN2 high-pressure tank source depletion rates for various leak scenarios, and the ability of the GN2 consumables to support cooling of Auxiliary Power Unit (APU) lubrication during entry. The data was used to support flight rationale concerning loss of an orbiter APU/hydraulic system and mission work-arounds. The GN2 mass flow-rate calculator standardizes a method for rapid assessment of GN2 mass flow through various orifice sizes for various discharge coefficients, delta pressures, and temperatures. The calculator utilizes a 0.9-lb (0.4 kg) GN2 source regulated to 40 psia (.276 kPa). These parameters correspond to the Space Shuttle WSB GN2 Source and Water Tank Bellows, but can be changed in the spreadsheet to accommodate any system parameters. The calculator can be used to analyze a leak source, leak rate, gas consumables depletion time, and puncture diameter that simulates the measured GN2 system pressure drop.
Calculation of CSF spaces in CT
Energy Technology Data Exchange (ETDEWEB)
Hacker, H; Artmann, H [Frankfurt Univ. (Germany, F.R.). Abt. fuer Neuroradiologie
1978-01-01
Objective digital determination of CSF spaces is discussed, with ventricular and subarachnoid spaces handled separately. This method avoids the difficulty of visual definition of ventricular borders in planimetric measurements. The principle is to count automatically all pixels corresponding to CSF in a given region with a Hounsfield unit and to multiply this number by the pixel size. This will give the total surface area of CSF spaces in square millimeters. The calculation of pixel values for CSF spaces and brain tissue is experimentally formulated taking the intersection of the Gaussian curves for ventricular content and brain tissue. In practice, the determination of CSF spaces is done by first calculating a histogram of the total brain in a given slice defining all CSF spaces. Next a histogram of a region including ventricles with adjoining tissue is calculated and the ventricular size is calculated. By subtraction of the ventricle value from the total CSF space value, the subarachnoid space size is obtained. The advantages of this mehtod will be discussed.
Selection of skin dose calculation methodologies
International Nuclear Information System (INIS)
Farrell, W.E.
1987-01-01
This paper reports that good health physics practice dictates that a dose assessment be performed for any significant skin contamination incident. There are, however, several methodologies that could be used, and while there is probably o single methodology that is proper for all cases of skin contamination, some are clearly more appropriate than others. This can be demonstrated by examining two of the more distinctly different options available for estimating skin dose the calculational methods. The methods compiled by Healy require separate beta and gamma calculations. The beta calculational method is the derived by Loevinger, while the gamma dose is calculated from the equation for dose rate from an infinite plane source with an absorber between the source and the detector. Healy has provided these formulas in graphical form to facilitate rapid dose rate determinations at density thicknesses of 7 and 20 mg/cm 2 . These density thicknesses equate to the regulatory definition of the sensitive layer of the skin and a more arbitrary value to account of beta absorption in contaminated clothing
Three dimensions transport calculations for PWR core
International Nuclear Information System (INIS)
Richebois, E.
2000-01-01
The objective of this work is to define improved 3-D core calculation methods based on the transport theory. These methods can be particularly useful and lead to more precise computations in areas of the core where anisotropy and steep flux gradients occur, especially near interface and boundary conditions and in regions of high heterogeneity (bundle with absorbent rods). In order to apply the transport theory a new method for calculating reflector constants has been developed, since traditional methods were only suited for 2-group diffusion core calculations and could not be extrapolated to transport calculations. In this thesis work, the new method for obtaining reflector constants is derived regardless of the number of energy groups and of the operator used. The core calculations results using the reflector constants thereof obtained have been validated on the EDF's power reactor Saint Laurent B1 with MOX loading. The advantages of a 3-D core transport calculation scheme have been highlighted as opposed to diffusion methods; there are a considerable number of significant effects and potential advantages to be gained in rod worth calculations for instance. These preliminary results obtained with on particular cycle will have to be confirmed by more systematic analysis. Accidents like MSLB (main steam line break) and LOCA (loss of coolant accident) should also be investigated and constitute challenging situations where anisotropy is high and/or flux gradients are steep. This method is now being validated for others EDF's PWRs' reactors, as well as for experimental reactors and other types of commercial reactors. (author)
KENO-IV code benchmark calculation, (6)
International Nuclear Information System (INIS)
Nomura, Yasushi; Naito, Yoshitaka; Yamakawa, Yasuhiro.
1980-11-01
A series of benchmark tests has been undertaken in JAERI in order to examine the capability of JAERI's criticality safety evaluation system consisting of the Monte Carlo calculation code KENO-IV and the newly developed multigroup constants library MGCL. The present report describes the results of a benchmark test using criticality experiments about Plutonium fuel in various shape. In all, 33 cases of experiments have been calculated for Pu(NO 3 ) 4 aqueous solution, Pu metal or PuO 2 -polystyrene compact in various shape (sphere, cylinder, rectangular parallelepiped). The effective multiplication factors calculated for the 33 cases distribute widely between 0.955 and 1.045 due to wide range of system variables. (author)
Calculating lattice thermal conductivity: a synopsis
Fugallo, Giorgia; Colombo, Luciano
2018-04-01
We provide a tutorial introduction to the modern theoretical and computational schemes available to calculate the lattice thermal conductivity in a crystalline dielectric material. While some important topics in thermal transport will not be covered (including thermal boundary resistance, electronic thermal conduction, and thermal rectification), we aim at: (i) framing the calculation of thermal conductivity within the general non-equilibrium thermodynamics theory of transport coefficients, (ii) presenting the microscopic theory of thermal conduction based on the phonon picture and the Boltzmann transport equation, and (iii) outlining the molecular dynamics schemes to calculate heat transport. A comparative and critical addressing of the merits and drawbacks of each approach will be discussed as well.
Automated one-loop calculations with GOSAM
International Nuclear Information System (INIS)
Cullen, Gavin; Greiner, Nicolas; Heinrich, Gudrun; Reiter, Thomas; Luisoni, Gionata
2011-11-01
We present the program package GoSam which is designed for the automated calculation of one-loop amplitudes for multi-particle processes in renormalisable quantum field theories. The amplitudes, which are generated in terms of Feynman diagrams, can be reduced using either D-dimensional integrand-level decomposition or tensor reduction. GoSam can be used to calculate one-loop QCD and/or electroweak corrections to Standard Model processes and offers the flexibility to link model files for theories Beyond the Standard Model. A standard interface to programs calculating real radiation is also implemented. We demonstrate the flexibility of the program by presenting examples of processes with up to six external legs attached to the loop. (orig.)
Automated one-loop calculations with GOSAM
Energy Technology Data Exchange (ETDEWEB)
Cullen, Gavin [Edinburgh Univ. (United Kingdom). School of Physics and Astronomy; Deutsches Elektronen-Synchrotron, Zeuthen [DESY; Germany; Greiner, Nicolas [Illinois Univ., Urbana-Champaign, IL (United States). Dept. of Physics; Max-Planck-Institut fuer Physik, Muenchen (Germany); Heinrich, Gudrun; Reiter, Thomas [Max-Planck-Institut fuer Physik, Muenchen (Germany); Luisoni, Gionata [Durham Univ. (United Kingdom). Inst. for Particle Physics Phenomenology; Mastrolia, Pierpaolo [Max-Planck-Institut fuer Physik, Muenchen (Germany); Padua Univ. (Italy). Dipt. di Fisica; Ossola, Giovanni [New York City Univ., NY (United States). New York City College of Technology; New York City Univ., NY (United States). The Graduate School and University Center; Tramontano, Francesco [European Organization for Nuclear Research (CERN), Geneva (Switzerland)
2011-11-15
We present the program package GoSam which is designed for the automated calculation of one-loop amplitudes for multi-particle processes in renormalisable quantum field theories. The amplitudes, which are generated in terms of Feynman diagrams, can be reduced using either D-dimensional integrand-level decomposition or tensor reduction. GoSam can be used to calculate one-loop QCD and/or electroweak corrections to Standard Model processes and offers the flexibility to link model files for theories Beyond the Standard Model. A standard interface to programs calculating real radiation is also implemented. We demonstrate the flexibility of the program by presenting examples of processes with up to six external legs attached to the loop. (orig.)
Poloidal field equilibrium calculations for JET
International Nuclear Information System (INIS)
Khalafallah, A.K.
1976-01-01
The structure of the JET 2D Poloidal Field Analysis Package is discussed. The ability to cope with different plasma current density distributions (skin, flat or peaked), each with a range of Beta poloidal values and varying plasma shapes is a new feature of these calculations. It is possible to construct instant-by-instant pictures of equilibrium configurations for various plasma build up scenarios taking into account the level of flux in the iron core and return limbs. The equilibrium configurations are calculated for two possible sequences of plasma build up. Examples of the magnetic field calculations being carried out under contract to JET at the Rutherford Laboratory, using a 3D code, are also given
Development of My Footprint Calculator
Mummidisetti, Karthik
The Environmental footprint is a very powerful tool that helps an individual to understand how their everyday activities are impacting environmental surroundings. Data shows that global climate change, which is a growing concern for nations all over the world, is already affecting humankind, plants and animals through raising ocean levels, droughts & desertification and changing weather patterns. In addition to a wide range of policy measures implemented by national and state governments, it is necessary for individuals to understand the impact that their lifestyle may have on their personal environmental footprint, and thus over the global climate change. "My Footprint Calculator" (myfootprintcalculator.com) has been designed to be one the simplest, yet comprehensive, web tools to help individuals calculate and understand their personal environmental impact. "My Footprint Calculator" is a website that queries users about their everyday habits and activities and calculates their personal impact on the environment. This website was re-designed to help users determine their environmental impact in various aspects of their lives ranging from transportation and recycling habits to water and energy usage with the addition of new features that will allow users to share their experiences and their best practices with other users interested in reducing their personal Environmental footprint. The collected data is stored in the database and a future goal of this work plans to analyze the collected data from all users (anonymously) for developing relevant trends and statistics.
First-principles calculations of mobility
Krishnaswamy, Karthik
First-principles calculations can be a powerful predictive tool for studying, modeling and understanding the fundamental scattering mechanisms impacting carrier transport in materials. In the past, calculations have provided important qualitative insights, but numerical accuracy has been limited due to computational challenges. In this talk, we will discuss some of the challenges involved in calculating electron-phonon scattering and carrier mobility, and outline approaches to overcome them. Topics will include the limitations of models for electron-phonon interaction, the importance of grid sampling, and the use of Gaussian smearing to replace energy-conserving delta functions. Using prototypical examples of oxides that are of technological importance-SrTiO3, BaSnO3, Ga2O3, and WO3-we will demonstrate computational approaches to overcome these challenges and improve the accuracy. One approach that leads to a distinct improvement in the accuracy is the use of analytic functions for the band dispersion, which allows for an exact solution of the energy-conserving delta function. For select cases, we also discuss direct quantitative comparisons with experimental results. The computational approaches and methodologies discussed in the talk are general and applicable to other materials, and greatly improve the numerical accuracy of the calculated transport properties, such as carrier mobility, conductivity and Seebeck coefficient. This work was performed in collaboration with B. Himmetoglu, Y. Kang, W. Wang, A. Janotti and C. G. Van de Walle, and supported by the LEAST Center, the ONR EXEDE MURI, and NSF.
Unit Cost Compendium Calculations
U.S. Environmental Protection Agency — The Unit Cost Compendium (UCC) Calculations raw data set was designed to provide for greater accuracy and consistency in the use of unit costs across the USEPA...
Friction and wear calculation methods
Kragelsky, I V; Kombalov, V S
1981-01-01
Friction and Wear: Calculation Methods provides an introduction to the main theories of a new branch of mechanics known as """"contact interaction of solids in relative motion."""" This branch is closely bound up with other sciences, especially physics and chemistry. The book analyzes the nature of friction and wear, and some theoretical relationships that link the characteristics of the processes and the properties of the contacting bodies essential for practical application of the theories in calculating friction forces and wear values. The effect of the environment on friction and wear is a
Small portable speed calculator
Burch, J. L.; Billions, J. C.
1973-01-01
Calculator is adapted stopwatch calibrated for fast accurate measurement of speeds. Single assembled unit is rugged, self-contained, and relatively inexpensive to manufacture. Potential market includes automobile-speed enforcement, railroads, and field-test facilities.
International Nuclear Information System (INIS)
Goto, Minoru; Takamatsu, Kuniyoshi
2007-03-01
The HTTR temperature coefficients required for the core dynamics calculations had been calculated from the HTTR core calculation results by the diffusion code with which the corrections had been performed using the core calculation results by the Monte-Carlo code MVP. This calculation method for the temperature coefficients was considered to have some issues to be improved. Then, the calculation method was improved to obtain the temperature coefficients in which the corrections by the Monte-Carlo code were not required. Specifically, from the point of view of neutron spectrum calculated by lattice calculations, the lattice model was revised which had been used for the calculations of the temperature coefficients. The HTTR core calculations were performed by the diffusion code with the group constants which were generated by the lattice calculations with the improved lattice model. The core calculations and the lattice calculations were performed by the SRAC code system. The HTTR core dynamics calculation was performed with the temperature coefficient obtained from the core calculation results. In consequence, the core dynamics calculation result showed good agreement with the experimental data and the valid temperature coefficient could be calculated only by the diffusion code without the corrections by Monte-Carlo code. (author)
Microbial Communities Model Parameter Calculation for TSPA/SR
International Nuclear Information System (INIS)
D. Jolley
2001-01-01
This calculation has several purposes. First the calculation reduces the information contained in ''Committed Materials in Repository Drifts'' (BSC 2001a) to useable parameters required as input to MING V1.O (CRWMS M and O 1998, CSCI 30018 V1.O) for calculation of the effects of potential in-drift microbial communities as part of the microbial communities model. The calculation is intended to replace the parameters found in Attachment II of the current In-Drift Microbial Communities Model revision (CRWMS M and O 2000c) with the exception of Section 11-5.3. Second, this calculation provides the information necessary to supercede the following DTN: M09909SPAMING1.003 and replace it with a new qualified dataset (see Table 6.2-1). The purpose of this calculation is to create the revised qualified parameter input for MING that will allow ΔG (Gibbs Free Energy) to be corrected for long-term changes to the temperature of the near-field environment. Calculated herein are the quadratic or second order regression relationships that are used in the energy limiting calculations to potential growth of microbial communities in the in-drift geochemical environment. Third, the calculation performs an impact review of a new DTN: M00012MAJIONIS.000 that is intended to replace the currently cited DTN: GS9809083 12322.008 for water chemistry data used in the current ''In-Drift Microbial Communities Model'' revision (CRWMS M and O 2000c). Finally, the calculation updates the material lifetimes reported on Table 32 in section 6.5.2.3 of the ''In-Drift Microbial Communities'' AMR (CRWMS M and O 2000c) based on the inputs reported in BSC (2001a). Changes include adding new specified materials and updating old materials information that has changed
Reactor calculation benchmark PCA blind test results
International Nuclear Information System (INIS)
Kam, F.B.K.; Stallmann, F.W.
1980-01-01
Further improvement in calculational procedures or a combination of calculations and measurements is necessary to attain 10 to 15% (1 sigma) accuracy for neutron exposure parameters (flux greater than 0.1 MeV, flux greater than 1.0 MeV, and dpa). The calculational modeling of power reactors should be benchmarked in an actual LWR plant to provide final uncertainty estimates for end-of-life predictions and limitations for plant operations. 26 references, 14 figures, 6 tables
Reactor calculation benchmark PCA blind test results
Energy Technology Data Exchange (ETDEWEB)
Kam, F.B.K.; Stallmann, F.W.
1980-01-01
Further improvement in calculational procedures or a combination of calculations and measurements is necessary to attain 10 to 15% (1 sigma) accuracy for neutron exposure parameters (flux greater than 0.1 MeV, flux greater than 1.0 MeV, and dpa). The calculational modeling of power reactors should be benchmarked in an actual LWR plant to provide final uncertainty estimates for end-of-life predictions and limitations for plant operations. 26 references, 14 figures, 6 tables.
Procedures for Calculating Residential Dehumidification Loads
Energy Technology Data Exchange (ETDEWEB)
Winkler, Jon [National Renewable Energy Lab. (NREL), Golden, CO (United States); Booten, Chuck [National Renewable Energy Lab. (NREL), Golden, CO (United States)
2016-06-01
Residential building codes and voluntary labeling programs are continually increasing the energy efficiency requirements of residential buildings. Improving a building's thermal enclosure and installing energy-efficient appliances and lighting can result in significant reductions in sensible cooling loads leading to smaller air conditioners and shorter cooling seasons. However due to fresh air ventilation requirements and internal gains, latent cooling loads are not reduced by the same proportion. Thus, it's becoming more challenging for conventional cooling equipment to control indoor humidity at part-load cooling conditions and using conventional cooling equipment in a non-conventional building poses the potential risk of high indoor humidity. The objective of this project was to investigate the impact the chosen design condition has on the calculated part-load cooling moisture load, and compare calculated moisture loads and the required dehumidification capacity to whole-building simulations. Procedures for sizing whole-house supplemental dehumidification equipment have yet to be formalized; however minor modifications to current Air-Conditioner Contractors of America (ACCA) Manual J load calculation procedures are appropriate for calculating residential part-load cooling moisture loads. Though ASHRAE 1% DP design conditions are commonly used to determine the dehumidification requirements for commercial buildings, an appropriate DP design condition for residential buildings has not been investigated. Two methods for sizing supplemental dehumidification equipment were developed and tested. The first method closely followed Manual J cooling load calculations; whereas the second method made more conservative assumptions impacting both sensible and latent loads.
The cellular approach to band structure calculations
International Nuclear Information System (INIS)
Verwoerd, W.S.
1982-01-01
A short introduction to the cellular approach in band structure calculations is given. The linear cellular approach and its potantial applicability in surface structure calculations is given some consideration in particular
Reactor perturbation calculations by Monte Carlo methods
International Nuclear Information System (INIS)
Gubbins, M.E.
1965-09-01
Whilst Monte Carlo methods are useful for reactor calculations involving complicated geometry, it is difficult to apply them to the calculation of perturbation worths because of the large amount of computing time needed to obtain good accuracy. Various ways of overcoming these difficulties are investigated in this report, with the problem of estimating absorbing control rod worths particularly in mind. As a basis for discussion a method of carrying out multigroup reactor calculations by Monte Carlo methods is described. Two methods of estimating a perturbation worth directly, without differencing two quantities of like magnitude, are examined closely but are passed over in favour of a third method based on a correlation technique. This correlation method is described, and demonstrated by a limited range of calculations for absorbing control rods in a fast reactor. In these calculations control rod worths of between 1% and 7% in reactivity are estimated to an accuracy better than 10% (3 standard errors) in about one hour's computing time on the English Electric KDF.9 digital computer. (author)
Adjoint spectrum calculation in fuel heterogeneous cells
International Nuclear Information System (INIS)
Suster, Luis Carlos
1998-01-01
In most codes for cells calculation, the multigroup cross sections are generated taking into consideration the conservation of the reaction rates in the forward spectrum. However, for certain uses of the perturbation theory it's necessary to use the average of the parameters for energy macrogroups over the forward and the adjoint spectra. In this thesis the adjoint spectrum was calculated from the adjoint neutron balance equations, that were obtained for a heterogeneous unit cell. The collision probabilities method was used to obtain these equations. In order optimize the computational run-time, the Gaussian quadrature method was used in the calculation of the neutron balance equations, forward and adjoint. This method of integration was also used for the Doppler broadening functions calculation, necessary for obtaining the energy dependent cross sections. In order to calculate the reaction rates and the average cross sections, using both the forward and the adjoint neutron spectra, the most important resonances of the U 238 were considered. The results obtained with the method show significant differences for the different cross sections weighting schemes. (author)
Handout on shielding calculation
International Nuclear Information System (INIS)
Heilbron Filho, P.F.L.
1991-01-01
In order to avoid the difficulties of the radioprotection supervisors in the tasks related to shielding calculations, is presented in this paper the basic concepts of shielding theory. It also includes exercises and examples. (author)
Casio Graphical Calculator Project.
Stott, Nick
2001-01-01
Shares experiences of a project aimed at developing and refining programs written on a Casio FX9750G graphing calculator. Describes in detail some programs used to develop mental strategies and problem solving skills. (MM)
Can medical students calculate drug doses? | Harries | Southern ...
African Journals Online (AJOL)
... with calculations when the drug concentration was expressed either as a ratio or a percentage. Conclusion: Our findings support calls for the standardised labelling of drugs in solution and for dosage calculation training in the medical curriculum. Keywords: drug dosage calculations, clinical competence, medication errors
Shielding calculations for ships carrying irradiated nuclear fuel
International Nuclear Information System (INIS)
Burstall, R.F.; Dean, M.H.
1983-01-01
A number of ships have been constructed to carry irradiated fuel from Japan to the UK and France, for reprocessing. About twenty transport flasks may be carried on each voyage. Permanent shielding must be provided on the ships to ensure that no member of the crew receives an annual dose rate greater than a specified limit. As the fuel is of varying type and radiation history, and as flasks of differing designs are used, many calculations are needed. There are a number of difficulties in making shielding calculations for the ships. The geometry is complex, dimensions are large, and considerable air spaces are involved. The paper considers possible methods of calculation. The line-of-sight method is chosen for most of the calculations, for both gamma radiation and neutrons. The basic data which is used in the calculations is described. As the methods of calculation are somewhat approximate, it is necessary to provide confirmation that they are sufficiently accurate. Validation has been provided in two ways. First, measurements have been made on board the ships, and these have been checked against calculation. Second, a simplified model of the flasks and ship has been set up, and calculations checked against more sophisticated methods. Results of the validation checks are presented, and it is shown that adequate accuracy is achieved. (author)
Shielding calculations for ships carrying irradiated nuclear fuel
International Nuclear Information System (INIS)
Dean, M.H.
1985-01-01
A number of ships have been constructed to carry irradiated fuel from Japan to the U.K. and France, for reprocessing. About 20 transport flasks may be carried on each voyage. Permanent shielding must be provided on the ships to ensure that no member of the crew receives an annual dose greater than a specified limit. As the fuel is of varying type and radiation history, and as flasks of differing designs are used, many shielding calculations are needed. There are a number of difficulties in making shielding calculations for the ships. The geometry is complex, dimensions are large and considerable air spaces are involved. The paper considers possible methods of calculation. The line-of-sight method is chosen for most of the calculations, for both γ-radiation and neutrons. The basic data which is used in the calculations is described. As the methods of calculation are somewhat approximate, it is necessary to provide confirmation that they are sufficiently accurate. Validation has been provided in two ways. First, measurements have been made on board one of the ships, Pacific Crane, and these have been checked against calculation. Second, a simplified model of the flasks and ship has been set up, and calculations checked against more sophisticated methods. Results of the validation checks are presented, and it is shown that adequate accuracy is achieved. (author)
Program for the surface muon spectra calculation
International Nuclear Information System (INIS)
Arkatov, Yu.M.; Voloshchuk, V.I.; Zolenko, V.A.; Prokhorets, I.M.; Soldatov, S.A.
1987-01-01
Program for the ''surface'' muon spectrum calculation is described. The algorithm is based on simulation of coordinates of π-meson birth point and direction of its escape from meson-forming target (MFT) according to angular distribution with the use of Monte Carlo method. Ionization losses of π-(μ)-mesons in the target are taken into account in the program. Calculation of ''surface'' muon spectrum is performed in the range of electron energies from 150 MeV up to 1000 MeV. Spectra of π-mesons are calculated with account of ionization losses in the target and without it. Distributions over lengths of π-meson paths in MFT and contribution of separate sections of the target to pion flux at the outlet of meson channel are calculated as well. Meson-forming target for calculation can be made of any material. The program provides for the use of the MFT itself in the form of photon converter or photon converter is located in front of the target. The program is composed of 13 subprograms; 2 of them represent generators of pseudorandom numbers, distributed uniformly in the range from 0 up to 1, and numbers with Gauss distribution. Example of calculation for copper target of 3 cm length, electron beam current-1 μA, energy-300 MeV is presented
Fluence-convolution broad-beam (FCBB) dose calculation
Energy Technology Data Exchange (ETDEWEB)
Lu Weiguo; Chen Mingli, E-mail: wlu@tomotherapy.co [TomoTherapy Inc., 1240 Deming Way, Madison, WI 53717 (United States)
2010-12-07
IMRT optimization requires a fast yet relatively accurate algorithm to calculate the iteration dose with small memory demand. In this paper, we present a dose calculation algorithm that approaches these goals. By decomposing the infinitesimal pencil beam (IPB) kernel into the central axis (CAX) component and lateral spread function (LSF) and taking the beam's eye view (BEV), we established a non-voxel and non-beamlet-based dose calculation formula. Both LSF and CAX are determined by a commissioning procedure using the collapsed-cone convolution/superposition (CCCS) method as the standard dose engine. The proposed dose calculation involves a 2D convolution of a fluence map with LSF followed by ray tracing based on the CAX lookup table with radiological distance and divergence correction, resulting in complexity of O(N{sup 3}) both spatially and temporally. This simple algorithm is orders of magnitude faster than the CCCS method. Without pre-calculation of beamlets, its implementation is also orders of magnitude smaller than the conventional voxel-based beamlet-superposition (VBS) approach. We compared the presented algorithm with the CCCS method using simulated and clinical cases. The agreement was generally within 3% for a homogeneous phantom and 5% for heterogeneous and clinical cases. Combined with the 'adaptive full dose correction', the algorithm is well suitable for calculating the iteration dose during IMRT optimization.
Development of Calculation Algorithm for ECCS Kinematic Shock
Energy Technology Data Exchange (ETDEWEB)
Lee, Seung-Chan; Yoon, Duk-Joo; Ha, Sang-Jun [KHNP-CRI, Daejeon (Korea, Republic of)
2014-10-15
The void fraction of inverted U-pipes in front of SI(Safety Injection) pumps impact on the pipe system of ECCS(Emergency Core Cooling Systems). This phenomena is called as 'Kinematic Shock'. The purpose of this paper is to achieve the more exactly calculation when the kinematic shock is calculated by simplified equation. The behavior of the void packet of the ECCS pipes is illustrated by the simplified (other name is kinematic shock equation).. The kinematic shock is defined as the depth of total length of void clusters in the pipes of ECCS when the void cluster is continually reached along the part of pipes in vertical direction. In this paper, the simplified equation is evaluated by comparing calculation error each other.]. The more exact methods of calculating the depth of the kinematic shock in ECCS is achieved. The error of kinematic shock calculation is strongly depended on the calculation search gap and the order of Taylor's expansion. From this study, to select the suitable search gap and the suitable calculation order, differential root method, secant method, and Taylor's expansion form are compared one another.
An adaptive sampling scheme for deep-penetration calculation
International Nuclear Information System (INIS)
Wang, Ruihong; Ji, Zhicheng; Pei, Lucheng
2013-01-01
As we know, the deep-penetration problem has been one of the important and difficult problems in shielding calculation with Monte Carlo Method for several decades. In this paper, an adaptive Monte Carlo method under the emission point as a sampling station for shielding calculation is investigated. The numerical results show that the adaptive method may improve the efficiency of the calculation of shielding and might overcome the under-estimation problem easy to happen in deep-penetration calculation in some degree
On the calculation of soft phase space integral
International Nuclear Information System (INIS)
Zhu, Hua Xing
2015-01-01
The recent discovery of the Higgs boson at the LHC attracts much attention to the precise calculation of its production cross section in quantum chromodynamics. In this work, we discuss the calculation of soft triple-emission phase space integral, which is an essential ingredient in the recently calculated soft-virtual corrections to Higgs boson production at next-to-next-to-next-to-leading order. The main techniques used this calculation are method of differential equation for Feynman integral, and integration of harmonic polylogarithms.
Calculation of the resonance cross section functions
International Nuclear Information System (INIS)
Slipicevic, K.F.
1967-11-01
This paper includes the procedure for calculating the Doppler broadened line shape functions ψ and χ which are needed for calculation of resonance cross section functions. The obtained values are given in tables
Calculation of the resonance cross section functions
Energy Technology Data Exchange (ETDEWEB)
Slipicevic, K F [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)
1967-11-15
This paper includes the procedure for calculating the Doppler broadened line shape functions {psi} and {chi} which are needed for calculation of resonance cross section functions. The obtained values are given in tables.
Calculation and definition of safety indicators
International Nuclear Information System (INIS)
Cristian, I.; Branzeu, N.; Vidican, D.; Vladescu, G.
1997-01-01
This paper presents, based on Cernavoda safety indicators proposal, the purpose definition and calculation formulas for each of the selected safety indicators. Five categories of safety indicators for Cernavoda Unit 1 were identified, namely: overall plant safety performance; initiating events; safety system availability, physical barrier integrity; indirect indicators. Definition, calculation and use of some safety indicators are shown in a tabular form. (authors)
Efficient Finite Element Calculation of Nγ
DEFF Research Database (Denmark)
Clausen, Johan; Damkilde, Lars; Krabbenhøft, K.
2007-01-01
This paper deals with the computational aspects of the Mohr-Coulomb material model, in particular the calculation of the bearing capacity factor Nγfor a strip and a circular footing.......This paper deals with the computational aspects of the Mohr-Coulomb material model, in particular the calculation of the bearing capacity factor Nγfor a strip and a circular footing....
Breit–Pauli atomic structure calculations for Fe XI
International Nuclear Information System (INIS)
Aggarwal, Sunny; Singh, Jagjit; Mohan, Man
2013-01-01
Energy levels, oscillator strengths, and transition probabilities are calculated for the lowest-lying 165 energy levels of Fe XI using configuration-interaction wavefunctions. The calculations include all the major correlation effects. Relativistic effects are included in the Breit–Pauli approximation by adding mass-correction, Darwin, and spin–orbit interaction terms to the non-relativistic Hamiltonian. For comparison with the calculated ab initio energy levels, we have also calculated the energy levels by using the fully relativistic multiconfiguration Dirac–Fock method. The calculated results are in close agreement with the National Institute of Standards and Technology compilation and other available results. New results are predicted for many of the levels belonging to the 3s3p 4 3d and 3s3p 3 3d 2 configurations, which are very important in astrophysics, relevant, for example, to the recent observations by the Hinode spacecraft. We expect that our extensive calculations will be useful to experimentalists in identifying the fine structure levels in their future work
Isochronous Cyclotron Closed Equilibrium Orbit Calculation Program Description
Kian, I N; Tarashkevich, R
2003-01-01
The Equilibrium Orbit Research Program - EORP, written in C++ with the use of Visual C++ is described. The program is intended for the calculation of the particle rotation frequency and particle kinetic energy in the closed equilibrium orbits of an isochronous cyclotron, where the closed equilibrium orbits are described through the radius and particle momentum angle: r_{eo}(\\theta) and \\varphi_{p}(\\theta). The program algorithm was developed on the basis of articles, lecture notes and original analytic calculations. The results of calculations by the EORP were checked and confirmed by using the results of calculations by the numerical methods. The discrepancies between the EORP results and the numerical method results for the calculations of the particle rotation frequency and particle kinetic energy are within the limits of \\pm1\\cdot10^{-4}. The EORP results and the numerical method results for the calculations of r_{eo}(\\theta) and \\varphi_{p}(\\theta) practically coincide. All this proves the accuracy of ca...
Some experience of shielding calculations by combinatorial method
International Nuclear Information System (INIS)
Korobejnikov, V.V.; Oussanov, V.I.
1996-01-01
Some aspects of the compound systems shielding calculations by a combinatorial approach are discussed. The effectiveness of such an approach is based on the fundamental characteristic of a compound system: if some element of the system have in itself mathematical or physical properties favorable for calculation, these properties may be used in a combinatorial approach and are lost when the system is being calculated in the whole by a direct approach. The combinatorial technique applied is well known. A compound system are being splitting for two or more auxiliary subsystems (so that calculation each of them is a more simple problem than calculation of the original problem (or at last is a soluble problem if original one is not). Calculation of every subsystem are carried out by suitable method and code, the coupling being made through boundary conditions or boundary source. The special consideration in the paper is given to a fast reactor shielding combinatorial analysis and to the testing of the results received. (author)
Am/Cm Vitrification Process: Vitrification Material Balance Calculations
International Nuclear Information System (INIS)
Smith, F.G.
2000-01-01
This report documents material balance calculations for the Americium/Curium vitrification process and describes the basis used to make the calculations. The material balance calculations reported here start with the solution produced by the Am/Cm pretreatment process as described in ``Material Balance Calculations for Am/Cm Pretreatment Process (U)'', SRT-AMC-99-0178 [1]. Following pretreatment, small batches of the product will be further treated with an additional oxalic acid precipitation and washing. The precipitate from each batch will then be charged to the Am/Cm melter with glass cullet and vitrified to produce the final product. The material balance calculations in this report are designed to provide projected compositions of the melter glass and off-gas streams. Except for decanted supernate collected from precipitation and precipitate washing, the flowsheet neglects side streams such as acid washes of empty tanks that would go directly to waste. Complete listings of the results of the material balance calculations are provided in the Appendices to this report
Microcomputer generated pipe support calculations
International Nuclear Information System (INIS)
Hankinson, R.F.; Czarnowski, P.; Roemer, R.E.
1991-01-01
The cost and complexity of pipe support design has been a continuing challenge to the construction and modification of commercial nuclear facilities. Typically, pipe support design or qualification projects have required large numbers of engineers centrally located with access to mainframe computer facilities. Much engineering time has been spent repetitively performing a sequence of tasks to address complex design criteria and consolidating the results of calculations into documentation packages in accordance with strict quality requirements. The continuing challenges of cost and quality, the need for support engineering services at operating plant sites, and the substantial recent advances in microcomputer systems suggested that a stand-alone microcomputer pipe support calculation generator was feasible and had become a necessity for providing cost-effective and high quality pipe support engineering services to the industry. This paper outlines the preparation for, and the development of, an integrated pipe support design/evaluation software system which maintains all computer programs in the same environment, minimizes manual performance of standard or repetitive tasks, and generates a high quality calculation which is consistent and easily followed
Calculational methods for lattice cells
International Nuclear Information System (INIS)
Askew, J.R.
1980-01-01
At the current stage of development, direct simulation of all the processes involved in the reactor to the degree of accuracy required is not an economic proposition, and this is achieved by progressive synthesis of models for parts of the full space/angle/energy neutron behaviour. The split between reactor and lattice calculations is one such simplification. Most reactors are constructed of repetitions of similar geometric units, the fuel elements, having broadly similar properties. Thus the provision of detailed predictions of their behaviour is an important step towards overall modelling. We shall be dealing with these lattice methods in this series of lectures, but will refer back from time to time to their relationship with overall reactor calculation The lattice cell is itself composed of somewhat similar sub-units, the fuel pins, and will itself often rely upon a further break down of modelling. Construction of a good model depends upon the identification, on physical and mathematical grounds, of the most helpful division of the calculation at this level
Calculation of saturated hydraulic conductivity of bentonite
International Nuclear Information System (INIS)
He Jun
2006-01-01
Hydraulic conductivity test has some defects such as weak repeatability, time-consuming. Taking bentonite as dual porous media, the calculation formula of the distance, d 2 , between montmorillonite in intraparticle pores is deduced. Improved calculated method of hydraulic conductivity is obtained using d 2 and Poiseuille law. The method is valid through the comparison with results of test and other methods. The method is very convenient to calculate hydraulic conductivity of bentonite of certain montmorillonite content and void ratio. (authors)
Evaluation of fusion-evaporation cross-section calculations
Blank, B.; Canchel, G.; Seis, F.; Delahaye, P.
2018-02-01
Calculated fusion-evaporation cross sections from five different codes are compared to experimental data. The present comparison extents over a large range of nuclei and isotopic chains to investigate the evolution of experimental and calculated cross sections. All models more or less overestimate the experimental cross sections. We found reasonable agreement by using the geometrical average of the five model calculations and dividing the average by a factor of 11.2. More refined analyses are made for example for the 100Sn region.
Computation cluster for Monte Carlo calculations
Energy Technology Data Exchange (ETDEWEB)
Petriska, M.; Vitazek, K.; Farkas, G.; Stacho, M.; Michalek, S. [Dep. Of Nuclear Physics and Technology, Faculty of Electrical Engineering and Information, Technology, Slovak Technical University, Ilkovicova 3, 81219 Bratislava (Slovakia)
2010-07-01
Two computation clusters based on Rocks Clusters 5.1 Linux distribution with Intel Core Duo and Intel Core Quad based computers were made at the Department of the Nuclear Physics and Technology. Clusters were used for Monte Carlo calculations, specifically for MCNP calculations applied in Nuclear reactor core simulations. Optimization for computation speed was made on hardware and software basis. Hardware cluster parameters, such as size of the memory, network speed, CPU speed, number of processors per computation, number of processors in one computer were tested for shortening the calculation time. For software optimization, different Fortran compilers, MPI implementations and CPU multi-core libraries were tested. Finally computer cluster was used in finding the weighting functions of neutron ex-core detectors of VVER-440. (authors)
Daylight calculations using constant luminance curves
Energy Technology Data Exchange (ETDEWEB)
Betman, E. [CRICYT, Mendoza (Argentina). Laboratorio de Ambiente Humano y Vivienda
2005-02-01
This paper presents a simple method to manually estimate daylight availability and to make daylight calculations using constant luminance curves calculated with local illuminance and irradiance data and the all-weather model for sky luminance distribution developed in the Atmospheric Science Research Center of the University of New York (ARSC) by Richard Perez et al. Work with constant luminance curves has the advantage that daylight calculations include the problem's directionality and preserve the information of the luminous climate of the place. This permits accurate knowledge of the resource and a strong basis to establish conclusions concerning topics related to the energy efficiency and comfort in buildings. The characteristics of the proposed method are compared with the method that uses the daylight factor. (author)
Statistical calculation of hot channel factors
International Nuclear Information System (INIS)
Farhadi, K.
2007-01-01
It is a conventional practice in the design of nuclear reactors to introduce hot channel factors to allow for spatial variations of power generation and flow distribution. Consequently, it is not enough to be able to calculate the nominal temperature distributions of fuel element, cladding, coolant, and central fuel. Indeed, one must be able to calculate the probability that the imposed temperature or heat flux limits in the entire core is not exceeded. In this paper, statistical methods are used to calculate hot channel factors for a particular case of a heterogeneous, Material Testing Reactor (MTR) and compare the results obtained from different statistical methods. It is shown that among the statistical methods available, the semi-statistical method is the most reliable one
Computation cluster for Monte Carlo calculations
International Nuclear Information System (INIS)
Petriska, M.; Vitazek, K.; Farkas, G.; Stacho, M.; Michalek, S.
2010-01-01
Two computation clusters based on Rocks Clusters 5.1 Linux distribution with Intel Core Duo and Intel Core Quad based computers were made at the Department of the Nuclear Physics and Technology. Clusters were used for Monte Carlo calculations, specifically for MCNP calculations applied in Nuclear reactor core simulations. Optimization for computation speed was made on hardware and software basis. Hardware cluster parameters, such as size of the memory, network speed, CPU speed, number of processors per computation, number of processors in one computer were tested for shortening the calculation time. For software optimization, different Fortran compilers, MPI implementations and CPU multi-core libraries were tested. Finally computer cluster was used in finding the weighting functions of neutron ex-core detectors of VVER-440. (authors)
Methods for thermal reactor lattice calculations
International Nuclear Information System (INIS)
Schneider, A.
1976-12-01
The American code HAMMER and the British code WIMS, for the analysis of thermal reactor lattices, have been investigated. The primary objective of this investigation was to identify the causes for the discrepancies that exist between the calculated and the experimentally determined reactivity of clean critical experiments. Three phases have been undertaken in the research: (a) Detailed comparison between the group cross-sections used by the codes; (b) Definition of the various approximations incorporated into the codes; (c) Comparison between the values of a variety of reaction rates calculated by the two codes. It was concluded that the main cause of discrepancy between calculations and experiments is due to data inaccuracies, while approximations introduced in solving the transport equation are of smaller importance
Fluorescein isothiocyanate: Molecular characterization by theoretical calculations
Energy Technology Data Exchange (ETDEWEB)
Casanovas, Jordi [Departament de Quimica, Escola Politecnica Superior, Universitat de Lleida, c/Jaume II No 69, Lleida E-25001 (Spain); Jacquemin, Denis [Laboratoire de Chimie Theorique Appliquee, Facultes Universitaires Notre-Dame de la Paix, rue de Bruxelles, 61, B-5000 Namur (Belgium)], E-mail: denis.jacquemin@fundp.ac.be; Perpete, Eric A. [Laboratoire de Chimie Theorique Appliquee, Facultes Universitaires Notre-Dame de la Paix, rue de Bruxelles, 61, B-5000 Namur (Belgium); Aleman, Carlos [Departament d' Enginyeria Quimica, E. T. S. d' Enginyeria Industrial de Barcelona, Universitat Politecnica de Catalunya, Diagonal 647, Barcelona E-08028 (Spain)], E-mail: carlos.aleman@upc.edu
2008-12-10
Quantum mechanical calculations have been used to investigate the conformation, molecular geometry, basicity and spectroscopic properties of fluorescein isothiocyanate in both the gas-phase and aqueous solution. Specifically, calculations have been performed considering the neutral, monoanionic and dianionic forms of this important fluorescent compound. Results reveal that for the neutral form multiple conformational states are possible, all them with significant contributions, and the stability of the different conformers is similar in the gas-phase and aqueous solution. Calculation of the excitation energies revealed that spectroscopic properties are very sensitive to the relaxation effect in solution. A good agreement has been reached obtained between the experimental and theoretical values derived from time-dependent density functional theory methods for the neutral form, whereas for charged species the calculations fail to accurately reproduce the measured trends.
Programmable calculator: alternative to minicomputer-based analyzer
International Nuclear Information System (INIS)
Hochel, R.C.
1979-01-01
Described are a number of typical field and laboratory counting systems that use standard stand-alone multichannel analyzers (MCA) interfaced to a Hewlett-Packard Company (HP 9830) programmable calculator. Such systems can offer significant advantages in cost and flexibility over a minicomputyr-based system. Because most laboratories tend to accumulate MCA's over the years, the programmable calculator also offers an easy way to upgrade the laboratory while making optimum use of existing systems. Software programs are easily tailored to fit a variety of general or specific applications. The only disadvantage of the calculator vs a computer-based system is in speed of analyses; however, for most applications this handicap is minimal. Applications discussed give a brief overview of the power and flexibility of the MCA-calculator approach to automated counting and data reduction
Calculation of tritium release from reactor's stack
International Nuclear Information System (INIS)
Akhadi, M.
1996-01-01
Method for calculation of tritium release from nuclear to environment has been discussed. Part of gas effluent contain tritium in form of HTO vapor released from reactor's stack was sampled using silica-gel. The silica-gel was put in the water to withdraw HTO vapor absorbed by silica-gel. Tritium concentration in the water was measured by liquid scintillation counter of Aloka LSC-703. Tritium concentration in the gas effluent and total release of tritium from reactor's stack during certain interval time were calculated using simple mathematic formula. This method has examined for calculation of tritium release from JRR-3M's stack of JAERI, Japan. From the calculation it was obtained the value of tritium release as much as 4.63 x 10 11 Bq during one month. (author)
Intercavitary implants dosage calculation
International Nuclear Information System (INIS)
Rehder, B.P.
The use of spacial geometry peculiar to each treatment for the attainment of intercavitary and intersticial implants dosage calculation is presented. The study is made in patients with intercavitary implants by applying a modified Manchester technique [pt
40 CFR 1033.530 - Duty cycles and calculations.
2010-07-01
... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Duty cycles and calculations. 1033.530... CONTROLS CONTROL OF EMISSIONS FROM LOCOMOTIVES Test Procedures § 1033.530 Duty cycles and calculations. This section describes how to apply the duty cycle to measured emission rates to calculate cycle...
Continuum variational and diffusion quantum Monte Carlo calculations
International Nuclear Information System (INIS)
Needs, R J; Towler, M D; Drummond, N D; Lopez RIos, P
2010-01-01
This topical review describes the methodology of continuum variational and diffusion quantum Monte Carlo calculations. These stochastic methods are based on many-body wavefunctions and are capable of achieving very high accuracy. The algorithms are intrinsically parallel and well suited to implementation on petascale computers, and the computational cost scales as a polynomial in the number of particles. A guide to the systems and topics which have been investigated using these methods is given. The bulk of the article is devoted to an overview of the basic quantum Monte Carlo methods, the forms and optimization of wavefunctions, performing calculations under periodic boundary conditions, using pseudopotentials, excited-state calculations, sources of calculational inaccuracy, and calculating energy differences and forces. (topical review)
First-principles calculations of novel materials
Sun, Jifeng
Computational material simulation is becoming more and more important as a branch of material science. Depending on the scale of the systems, there are many simulation methods, i.e. first-principles calculation (or ab-initio), molecular dynamics, mesoscale methods and continuum methods. Among them, first-principles calculation, which involves density functional theory (DFT) and based on quantum mechanics, has become to be a reliable tool in condensed matter physics. DFT is a single-electron approximation in solving the many-body problems. Intrinsically speaking, both DFT and ab-initio belong to the first-principles calculation since the theoretical background of ab-initio is Hartree-Fock (HF) approximation and both are aimed at solving the Schrodinger equation of the many-body system using the self-consistent field (SCF) method and calculating the ground state properties. The difference is that DFT introduces parameters either from experiments or from other molecular dynamic (MD) calculations to approximate the expressions of the exchange-correlation terms. The exchange term is accurately calculated but the correlation term is neglected in HF. In this dissertation, DFT based first-principles calculations were performed for all the novel materials and interesting materials introduced. Specifically, the DFT theory together with the rationale behind related properties (e.g. electronic, optical, defect, thermoelectric, magnetic) are introduced in Chapter 2. Starting from Chapter 3 to Chapter 5, several representative materials were studied. In particular, a new semiconducting oxytelluride, Ba2TeO is studied in Chapter 3. Our calculations indicate a direct semiconducting character with a band gap value of 2.43 eV, which agrees well with the optical experiment (˜ 2.93 eV). Moreover, the optical and defects properties of Ba2TeO are also systematically investigated with a view to understanding its potential as an optoelectronic or transparent conducting material. We find
COSANI-2, Gamma Doses from SABINE Calculation, Activity from ANISN Flux Calculation
International Nuclear Information System (INIS)
Dupont, C.
1975-01-01
1 - Nature of physical problem solved: Retrieval of SABINE and/or ANISN results. Calculates in case of SABINE results the individual contributions of capture gamma rays in each region to the total gamma dose and to the total gamma heating may calculate in case of ANISN new activity rates starting from ANISN flux saved on tape and activity cross sections taken on an ANISN binary library tape. The program can draw on a BENSON plotter any of the following quantities: - group flux; - activity rates; - dose rates; - neutron spectra for SABINE; - neutron or gamma direct or adjoint spectra for ANISN; - gamma heating and dose rate for SABINE including individual contributions from each region. Several ANISN and/or SABINE cases can be drawn on the same graph for comparison purposes. 2 - Restrictions on the complexity of the problem: Maximum number of: - tapes containing ANISN and/or SABINE results: 5; - curves per graph: 3; - regions: 40; - points per curve: 500; - energy groups: 200
Criticality calculation method for mixer-settlers
International Nuclear Information System (INIS)
Gonda, Kozo; Aoyagi, Haruki; Nakano, Ko; Kamikawa, Hiroshi.
1980-01-01
A new criticality calculation code MACPEX has been developed to evaluate and manage the criticality of the process in the extractor of mixer-settler type. MACPEX can perform the combined calculation with the PUREX process calculation code MIXSET, to get the neutron flux and the effective multiplication constant in the mixer-settlers. MACPEX solves one-dimensional diffusion equation by the explicit difference method and the standard source-iteration technique. The characteristics of MACPEX are as follows. 1) Group constants of 4 energy groups for the 239 Pu-H 2 O solution, water, polyethylene and SUS 28 are provided. 2) The group constants of the 239 Pu-H 2 O solution are given by the functional formulae of the plutonium concentration, which is less than 50 g/l. 3) Two boundary conditions of the vacuum condition and the reflective condition are available in this code. 4) The geometrical bucklings can be calculated for a certain energy group and/or region by using the three dimentional neutron flux profiles obtained by CITATION. 5) The buckling correction search can be carried out in order to get a desired k sub(eff). (author)
Source term calculations - Ringhals 2 PWR
International Nuclear Information System (INIS)
Johansson, L.L.
1998-02-01
This project was performed within the fifth and final phase of sub-project RAK-2.1 of the Nordic Co-operative Reactor Safety Program, NKS.RAK-2.1 has also included studies of reflooding of degraded core, recriticality and late phase melt progression. Earlier source term calculations for Swedish nuclear power plants are based on the integral code MAAP. A need was recognised to compare these calculations with calculations done with mechanistic codes. In the present work SCDAP/RELAP5 and CONTAIN were used. Only limited results could be obtained within the frame of RAK-2.1, since many problems were encountered using the SCDAP/RELAP5 code. The main obstacle was the extremely long execution times of the MOD3.1 version, but also some dubious fission product calculations. However, some interesting results were obtained for the studied sequence, a total loss of AC power. The report describes the modelling approach for SCDAP/RELAP5 and CONTAIN, and discusses results for the transient including the event of a surge line creep rupture. The study will probably be completed later, providing that an improved SCDAP/RELAP5 code version becomes available. (au) becomes available. (au)
42 CFR 102.82 - Calculation of death benefits.
2010-10-01
... 42 Public Health 1 2010-10-01 2010-10-01 false Calculation of death benefits. 102.82 Section 102... COMPENSATION PROGRAM Calculation and Payment of Benefits § 102.82 Calculation of death benefits. (a... paragraph (d) of this section for the death benefit available to dependents. (2) Deceased person means an...
PLUTONIUM/HIGH-LEVEL VITRIFIED WASTE BDBE DOSE CALCULATION
Energy Technology Data Exchange (ETDEWEB)
J.A. Ziegler
2000-11-20
The purpose of this calculation is to provide a dose consequence analysis of high-level waste (HLW) consisting of plutonium immobilized in vitrified HLW to be handled at the proposed Monitored Geologic Repository at Yucca Mountain for a beyond design basis event (BDBE) under expected conditions using best estimate values for each calculation parameter. In addition to the dose calculation, a plutonium respirable particle size for dose calculation use is derived. The current concept for this waste form is plutonium disks enclosed in cans immobilized in canisters of vitrified HLW (i.e., glass). The plutonium inventory at risk used for this calculation is selected from Plutonium Immobilization Project Input for Yucca Mountain Total Systems Performance Assessment (Shaw 1999). The BDBE examined in this calculation is a nonmechanistic initiating event and the sequence of events that follow to cause a radiological release. This analysis will provide the radiological releases and dose consequences for a postulated BDBE. Results may be considered in other analyses to determine or modify the safety classification and quality assurance level of repository structures, systems, and components. This calculation uses best available technical information because the BDBE frequency is very low (i.e., less than 1.0E-6 events/year) and is not required for License Application for the Monitored Geologic Repository. The results of this calculation will not be used as part of a licensing or design basis.
Calculation of projected ranges
International Nuclear Information System (INIS)
Biersack, J.P.
1980-09-01
The concept of multiple scattering is reconsidered for obtaining the directional spreading of ion motion as a function of energy loss. From this the mean projection of each pathlength element of the ion trajectory is derived which - upon summation or integration - leads to the desired mean projected range. In special cases, the calculation can be carried out analytically, otherwise a simple general algorithm is derived which is suitable even for the smallest programmable calculators. Necessary input for the present treatment consists only of generally accessable stopping power and straggling formulas. The procedure does not rely on scattering cross sections, e.g. power potential or f(t 1 sup(/) 2 ) approximations. The present approach lends itself easily to include electronic straggling or to treat composed target materials, or even to account for the so-called time integral. (orig.)
Selfconsistent calculations at finite temperatures
International Nuclear Information System (INIS)
Brack, M.; Quentin, P.
1975-01-01
Calculations have been done for the spherical nuclei 40 Ca, 208 Pb and the hypothetical superheavy nucleus with Z=114, A=298, as well as for the deformed nucleus 168 Yb. The temperature T was varied from zero up to 5 MeV. For T>3 MeV, some numerical problems arise in connection with the optimization of the basis when calculating deformed nuclei. However, at these high temperatures the occupation numbers in the continuum are sufficiently large so that the nucleus starts evaporating particles and no equilibrium state can be described. Results are obtained for excitation energies and entropies. (Auth.)
Calculation methods in program CCRMN
Energy Technology Data Exchange (ETDEWEB)
Chonghai, Cai [Nankai Univ., Tianjin (China). Dept. of Physics; Qingbiao, Shen [Chinese Nuclear Data Center, Beijing, BJ (China)
1996-06-01
CCRMN is a program for calculating complex reactions of a medium-heavy nucleus with six light particles. In CCRMN, the incoming particles can be neutrons, protons, {sup 4}He, deuterons, tritons and {sup 3}He. the CCRMN code is constructed within the framework of the optical model, pre-equilibrium statistical theory based on the exciton model and the evaporation model. CCRMN is valid in 1{approx} MeV energy region, it can give correct results for optical model quantities and all kinds of reaction cross sections. This program has been applied in practical calculations and got reasonable results.
Molecular calculations with B functions
International Nuclear Information System (INIS)
Steinborn, E.O.; Homeier, H.H.H.; Ema, I.; Lopez, R.; Ramirez, G.
2000-01-01
A program for molecular calculations with B functions is reported and its performance is analyzed. All the one- and two-center integrals and the three-center nuclear attraction integrals are computed by direct procedures, using previously developed algorithms. The three- and four-center electron repulsion integrals are computed by means of Gaussian expansions of the B functions. A new procedure for obtaining these expansions is also reported. Some results on full molecular calculations are included to show the capabilities of the program and the quality of the B functions to represent the electronic functions in molecules
Results of recent calculations using realistic potentials
International Nuclear Information System (INIS)
Friar, J.L.
1987-01-01
Results of recent calculations for the triton using realistic potentials with strong tensor forces are reviewed, with an emphasis on progress made using the many different calculational schemes. Several test problems are suggested. 49 refs., 5 figs
Characteristic parameters of drift chambers calculation
International Nuclear Information System (INIS)
Duran, I.; Martinez-Laso, L.
1989-01-01
We present here the methods we used to analyse the characteristic parameters of drift chambers. The algorithms to calculate the electric potential in any point for any drift chamber geometry are presented. We include the description of the programs used to calculate the electric field, the drift paths, the drift velocity and the drift time. The results and the errors are discussed. (Author) 7 refs
Exact and approximate multiple diffraction calculations
International Nuclear Information System (INIS)
Alexander, Y.; Wallace, S.J.; Sparrow, D.A.
1976-08-01
A three-body potential scattering problem is solved in the fixed scatterer model exactly and approximately to test the validity of commonly used assumptions of multiple scattering calculations. The model problem involves two-body amplitudes that show diffraction-like differential scattering similar to high energy hadron-nucleon amplitudes. The exact fixed scatterer calculations are compared to Glauber approximation, eikonal-expansion results and a noneikonal approximation
Calculation of the counting efficiency for extended sources
International Nuclear Information System (INIS)
Korun, M.; Vidmar, T.
2002-01-01
A computer program for calculation of efficiency calibration curves for extended samples counted on gamma- and X ray spectrometers is described. The program calculates efficiency calibration curves for homogeneous cylindrical samples placed coaxially with the symmetry axis of the detector. The method of calculation is based on integration over the sample volume of the efficiencies for point sources measured in free space on an equidistant grid of points. The attenuation of photons within the sample is taken into account using the self-attenuation function calculated with a two-dimensional detector model. (author)
Methane on Mars: Thermodynamic Equilibrium and Photochemical Calculations
Levine, J. S.; Summers, M. E.; Ewell, M.
2010-01-01
The detection of methane (CH4) in the atmosphere of Mars by Mars Express and Earth-based spectroscopy is very surprising, very puzzling, and very intriguing. On Earth, about 90% of atmospheric ozone is produced by living systems. A major question concerning methane on Mars is its origin - biological or geological. Thermodynamic equilibrium calculations indicated that methane cannot be produced by atmospheric chemical/photochemical reactions. Thermodynamic equilibrium calculations for three gases, methane, ammonia (NH3) and nitrous oxide (N2O) in the Earth s atmosphere are summarized in Table 1. The calculations indicate that these three gases should not exist in the Earth s atmosphere. Yet they do, with methane, ammonia and nitrous oxide enhanced 139, 50 and 12 orders of magnitude above their calculated thermodynamic equilibrium concentration due to the impact of life! Thermodynamic equilibrium calculations have been performed for the same three gases in the atmosphere of Mars based on the assumed composition of the Mars atmosphere shown in Table 2. The calculated thermodynamic equilibrium concentrations of the same three gases in the atmosphere of Mars is shown in Table 3. Clearly, based on thermodynamic equilibrium calculations, methane should not be present in the atmosphere of Mars, but it is in concentrations approaching 30 ppbv from three distinct regions on Mars.
Three-dimensional equilibria and Mercier stability calculations
International Nuclear Information System (INIS)
Lynch, V.E.; Dominguez, N.; Carreras, B.A.; Varias, A.; Alejaldre, C.; Fraguas, A.L.
1989-01-01
It is well known that an equilibrium to be used for stability calculations must be extremely accurate. These high accuracy requirements, in a fixed boundary calculation, are translated into high accuracy in the representation of the boundary. These requirements are even stricter for stellarator configurations, for which all the information about the magnetic configuration is given externally through the boundary. Many Fourier components are required to accurately represent the boundary input from a realistic coil system. For torsatron-type configurations, as many as 50 components can be needed to describe the last closed magnetic surface for the vacuum field. For a heliac configuration, the number of components can go up to 200. For 3-D calculations, there is another question of accuracy that does not apply to stability calculations for axisymmetric systems. This is the role of resonant components in the calculation of the geodesic curvature or the Pfirsch-Schlueter current. As Boozer argues, local flattening of the pressure profile eliminates the singularities generated by the resonant components. However, to implement it in a numerical calculation and to eliminate the resonant components, it is necessary to work in a coordinate system with straight magnetic field lines. This creates another problem, since the equilibrium representation in a straight magnetic field lines coordinate system requires many more components than the optimal equilibrium representation developed by Hirshman and co-workers over the last decade and implemented in the VMEC equilibrium code. In this paper, we use the VMEC equilibrium code and tranform the results to the straight magnetic field line coordinate system to calculate the input for the stability analysis. The accuracy of the transformation and the convergence of the equilibrium in the new coordinate system are the major points discussed in this paper. 6 refs., 1 fig
Calculation system for physical analysis of boiling water reactors
International Nuclear Information System (INIS)
Bouveret, F.
2001-01-01
Although Boiling Water Reactors generate a quarter of worldwide nuclear electricity, they have been only little studied in France. A certain interest now shows up for these reactors. So, the aim of the work presented here is to contribute to determine a core calculation methodology with CEA (Commissariat a l'Energie Atomique) codes. Vapour production in the reactor core involves great differences in technological options from pressurised water reactor. We analyse main physical phenomena for BWR and offer solutions taking them into account. BWR fuel assembly heterogeneity causes steep thermal flux gradients. The two dimensional collision probability method with exact boundary conditions makes possible to calculate accurately the flux in BWR fuel assemblies using the APOLLO-2 lattice code but induces a very long calculation time. So, we determine a new methodology based on a two-level flux calculation. Void fraction variations in assemblies involve big spectrum changes that we have to consider in core calculation. We suggest to use a void history parameter to generate cross-sections libraries for core calculation. The core calculation code has also to calculate the depletion of main isotopes concentrations. A core calculation associating neutronics and thermal-hydraulic codes lays stress on points we still have to study out. The most important of them is to take into account the control blade in the different calculation stages. (author)
Shielding calculation for bremsstrahlung from β-emitters
International Nuclear Information System (INIS)
Ichimiya, Tsutomu
1990-01-01
Accompanying the revision of radiation injury prevention law, the shielding calculation method for photon corresponding to the dose equivalent was shown. However, regarding the electron from β decay nuclide and bremsstrahlung caused by shielding material, the shielding calculation method corresponding to the 1 cm dose equivalent has not been reported, hence, in this report, the spectrum of β-ray is calculated and the 1 cm dose equivalent transmission rate of the bremsstrahlung was calculated for three kinds of shielding materials (iron, lead, concrete). As the result of consideration, it is sufficient to think about the bremsstrahlung due to negative electron emission accompanying β-decay. In β-decay, electrons which constitute the continuous spectrum with maximum energy are emitted. The shape of the spectrum differs with nuclides. The maximum energy of β-ray of generally used nuclides is mostly below 3MeV and, besides, the electron ray itself is easily shielded, while the strength of bremsstrahlung depends on the atomic number of shielding materials and its generating mechanism is complicated. In this report, the actual shielding calculation method for bremsstrahlung is shown with regard to the most frequently used β-decay nuclides. (M.T.)
Practical applications of internal dose calculations
International Nuclear Information System (INIS)
Carbaugh, E.H.
1994-06-01
Accurate estimates of intake magnitude and internal dose are the goal for any assessment of an actual intake of radioactivity. When only one datum is available on which to base estimates, the choices for internal dose assessment become straight-forward: apply the appropriate retention or excretion function, calculate the intake, and calculate the dose. The difficulty comes when multiple data and different types of data become available. Then practical decisions must be made on how to interpret conflicting data, or how to adjust the assumptions and techniques underlying internal dose assessments to give results consistent with the data. This article describes nine types of adjustments which can be incorporated into calculations of intake and internal dose, and then offers several practical insights to dealing with some real-world internal dose puzzles
A Novel Hybrid Similarity Calculation Model
Directory of Open Access Journals (Sweden)
Xiaoping Fan
2017-01-01
Full Text Available This paper addresses the problems of similarity calculation in the traditional recommendation algorithms of nearest neighbor collaborative filtering, especially the failure in describing dynamic user preference. Proceeding from the perspective of solving the problem of user interest drift, a new hybrid similarity calculation model is proposed in this paper. This model consists of two parts, on the one hand the model uses the function fitting to describe users’ rating behaviors and their rating preferences, and on the other hand it employs the Random Forest algorithm to take user attribute features into account. Furthermore, the paper combines the two parts to build a new hybrid similarity calculation model for user recommendation. Experimental results show that, for data sets of different size, the model’s prediction precision is higher than the traditional recommendation algorithms.
Parallel scalability of Hartree-Fock calculations
Chow, Edmond; Liu, Xing; Smelyanskiy, Mikhail; Hammond, Jeff R.
2015-03-01
Quantum chemistry is increasingly performed using large cluster computers consisting of multiple interconnected nodes. For a fixed molecular problem, the efficiency of a calculation usually decreases as more nodes are used, due to the cost of communication between the nodes. This paper empirically investigates the parallel scalability of Hartree-Fock calculations. The construction of the Fock matrix and the density matrix calculation are analyzed separately. For the former, we use a parallelization of Fock matrix construction based on a static partitioning of work followed by a work stealing phase. For the latter, we use density matrix purification from the linear scaling methods literature, but without using sparsity. When using large numbers of nodes for moderately sized problems, density matrix computations are network-bandwidth bound, making purification methods potentially faster than eigendecomposition methods.
LCEs for Naval Reactor Benchmark Calculations
International Nuclear Information System (INIS)
W.J. Anderson
1999-01-01
The purpose of this engineering calculation is to document the MCNP4B2LV evaluations of Laboratory Critical Experiments (LCEs) performed as part of the Disposal Criticality Analysis Methodology program. LCE evaluations documented in this report were performed for 22 different cases with varied design parameters. Some of these LCEs (10) are documented in existing references (Ref. 7.1 and 7.2), but were re-run for this calculation file using more neutron histories. The objective of this analysis is to quantify the MCNP4B2LV code system's ability to accurately calculate the effective neutron multiplication factor (k eff ) for various critical configurations. These LCE evaluations support the development and validation of the neutronics methodology used for criticality analyses involving Naval reactor spent nuclear fuel in a geologic repository
Calculation of Gilbert damping in ferromagnetic ﬁlms
Directory of Open Access Journals (Sweden)
Edwards D. M.
2013-01-01
Full Text Available The Gilbert damping constant in the phenomenological Landau-Lifshitz-Gilbert equation which describes the dynamics of magnetization, is calculated for Fe, Co and Ni bulk ferromagnets, Co ﬁlms and Co/Pd bilayers within a nine-band tight-binding model with spin-orbit coupling included. The calculational effciency is remarkably improved by introducing ﬁnite temperature into the electronic occupation factors and subsequent summation over the Matsubara frequencies. The calculated dependence of Gilbert damping constant on scattering rate for bulk Fe, Co and Ni is in good agreement with the results of previous ab initio calculations. Calculations are reported for ferromagnetic Co metallic ﬁlms and Co/Pd bilayers. The dependence of the Gilbert damping constant on Co ﬁlm thickness, for various scattering rates, is studied and compared with recent experiments.
Analysis of error in Monte Carlo transport calculations
International Nuclear Information System (INIS)
Booth, T.E.
1979-01-01
The Monte Carlo method for neutron transport calculations suffers, in part, because of the inherent statistical errors associated with the method. Without an estimate of these errors in advance of the calculation, it is difficult to decide what estimator and biasing scheme to use. Recently, integral equations have been derived that, when solved, predicted errors in Monte Carlo calculations in nonmultiplying media. The present work allows error prediction in nonanalog Monte Carlo calculations of multiplying systems, even when supercritical. Nonanalog techniques such as biased kernels, particle splitting, and Russian Roulette are incorporated. Equations derived here allow prediction of how much a specific variance reduction technique reduces the number of histories required, to be weighed against the change in time required for calculation of each history. 1 figure, 1 table
Energy Technology Data Exchange (ETDEWEB)
Lefevre, R.; D' Harcourt, A.; Dupuy, G. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires
1968-07-01
The aim of this work is to allow the practical execution of the estimated calculation of the electronic devices reliability and to standardize the source and the approach of the calculations while giving a systematic character to their execution mode. The systematic character of the calculations allows a comparison of the reliability of different materials and a rapid control of the calculations validity; at last, it brings out the use conditions of all the components of a set. A reliability calculation made according to the method described here reveals: -components badly used -the relative influence on the reliability of the set, of a component or a components group taking into account of the number, of the characteristics and of the use of these ones. At last, the results of the calculation allows to organize the exploitation (availability) and the maintenance (staff, stock of components) of the materials. The failure rates given in this book are only relating to the components service-life and do not correspond to precocious failures. (authors) [French] L'objet du present recueil est d'une part de permettre l'execution pratique du calcul previsionnel de la fiabilite des equipements electroniques et d'autre part d'uniformiser l'origine et la presentation de calculs en donnant un caractere systematique a leur mode d'execution. Le caractere systematique des calculs permet une comparaison de la fiabilite de differents materiels et un controle rapide de la validite des calculs; enfin il met en relief les conditions d'utilisation de tous les composants d'un ensemble. Un calcul de fiabilite realise selon la methode preconise par ce document permet la mise en evidence: -des composants mal utilises -de l'influence relative sur la fiabilite de l'ensemble, d'un composant ou d'un groupe de composants compte tenu du nombre, des caracteristiques et de l'utilisation de ceux-ci. Enfin, les resultats du
Study on the concurrent calculation with a multitransputer network
International Nuclear Information System (INIS)
Inomata, Shinji; Suzuki, Katsuo
1992-06-01
A relation between the calculation processing time and the overhead of multitransputer network is studied in performing several concurrent calculations. Hardware configuration we used was a ring-type linked nine transputer network on an INMOS B008 mother board which was inserted into a slot of a personal computer compatible with IBM PC/AT. Further we used the occam2 toolset of MS-DOS version to write the calculation programs in the occam2 language. On arithmetic operations, some trigonometric and transcendental functions and numerical integrations, the calculation time and overhead were measured. The results show that the relative calculation speed is proportional to the number of transputers when the ratio of calculation time to overhead is much greater than 1.0. And the overhead is consumed in inter-communicating among transputers for the parallel calculation set-up. (author)
International Nuclear Information System (INIS)
Etemad, M.A.
1981-04-01
The one dimensional discrete ordinates code ANISN-F was used to calculate the thermal neutron flux distribution in water from a Ra-Be neutron source. The calculations were performed in order to investigate the different possibilities of the code as well as to verify the results of the calculations in terms of comparisons to corresponding experimental data. Two different group cross section libraries were used in the calculations and conclusions were drawn on the adequacy of these libraries for a fixed source type calculation. Furthermore, critically calculations were performed for an infinite homogeneous slab of multiplying material using different angular and spatial approximations. The results of these calculations were then compared to the corresponding results previously obtained at this department by a different method and a different code. (author)
Subcritical calculation of the nuclear material warehouse
International Nuclear Information System (INIS)
Garcia M, T.; Mazon R, R.
2009-01-01
In this work the subcritical calculation of the nuclear material warehouse of the Reactor TRIGA Mark III labyrinth in the Mexico Nuclear Center is presented. During the adaptation of the nuclear warehouse (vault I), the fuel was temporarily changed to the warehouse (vault II) and it was also carried out the subcritical calculation for this temporary arrangement. The code used for the calculation of the effective multiplication factor, it was the Monte Carlo N-Particle Extended code known as MCNPX, developed by the National Laboratory of Los Alamos, for the particles transport. (Author)
Neutronic parameters calculations of a CANDU reactor
International Nuclear Information System (INIS)
Zamonsky, G.
1991-01-01
Neutronic calculations that reproduce in a simplified way some aspects of a CANDU reactor design were performed. Starting from some prefixed reactor parameters, cylindrical and uniform iron adjuster rods were designed. An appropriate refueling scheme was established, defininig in a 2 zones model their dimensions and exit burnups. The calculations have been done using the codes WIMS-D4 (cell), SNOD (reactivity device simulations) and PUMA (reactor). Comparing with similar calculations done with codes and models usually employed for CANDU design, it is concluded that the models and methods used are appropriate. (Author) [es
Electron and bremsstrahlung penetration and dose calculation
Watts, J. W., Jr.; Burrell, M. O.
1972-01-01
Various techniques for the calculation of electron and bremsstrahlung dose deposition are described. Energy deposition, transmission, and reflection coefficients for electrons incident on plane slabs are presented, and methods for their use in electron dose calculations were developed. A method using the straight-ahead approximation was also developed, and the various methods were compared and found to be in good agreement. Both accurate and approximate methods of calculating bremsstrahlung dose were derived and compared. Approximation is found to give a good estimate of dose where the electron spectrum falls off exponentially with energy.
Realistic shell-model calculations for Sn isotopes
International Nuclear Information System (INIS)
Covello, A.; Andreozzi, F.; Coraggio, L.; Gargano, A.; Porrino, A.
1997-01-01
We report on a shell-model study of the Sn isotopes in which a realistic effective interaction derived from the Paris free nucleon-nucleon potential is employed. The calculations are performed within the framework of the seniority scheme by making use of the chain-calculation method. This provides practically exact solutions while cutting down the amount of computational work required by a standard seniority-truncated calculation. The behavior of the energy of several low-lying states in the isotopes with A ranging from 122 to 130 is presented and compared with the experimental one. (orig.)
Fluidization calculation on nuclear fuel kernel coating
International Nuclear Information System (INIS)
Sukarsono; Wardaya; Indra-Suryawan
1996-01-01
The fluidization of nuclear fuel kernel coating was calculated. The bottom of the reactor was in the from of cone on top of the cone there was a cylinder, the diameter of the cylinder for fluidization was 2 cm and at the upper part of the cylinder was 3 cm. Fluidization took place in the cone and the first cylinder. The maximum and the minimum velocity of the gas of varied kernel diameter, the porosity and bed height of varied stream gas velocity were calculated. The calculation was done by basic program
The MiAge Calculator: a DNA methylation-based mitotic age calculator of human tissue types.
Youn, Ahrim; Wang, Shuang
2018-01-01
Cell division is important in human aging and cancer. The estimation of the number of cell divisions (mitotic age) of a given tissue type in individuals is of great interest as it allows not only the study of biological aging (using a new molecular aging target) but also the stratification of prospective cancer risk. Here, we introduce the MiAge Calculator, a mitotic age calculator based on a novel statistical framework, the MiAge model. MiAge is designed to quantitatively estimate mitotic age (total number of lifetime cell divisions) of a tissue using the stochastic replication errors accumulated in the epigenetic inheritance process during cell divisions. With the MiAge model, the MiAge Calculator was built using the training data of DNA methylation measures of 4,020 tumor and adjacent normal tissue samples from eight TCGA cancer types and was tested using the testing data of DNA methylation measures of 2,221 tumor and adjacent normal tissue samples of five other TCGA cancer types. We showed that within each of the thirteen cancer types studied, the estimated mitotic age is universally accelerated in tumor tissues compared to adjacent normal tissues. Across the thirteen cancer types, we showed that worse cancer survivals are associated with more accelerated mitotic age in tumor tissues. Importantly, we demonstrated the utility of mitotic age by showing that the integration of mitotic age and clinical information leads to improved survival prediction in six out of the thirteen cancer types studied. The MiAge Calculator is available at http://www.columbia.edu/∼sw2206/softwares.htm .
Calculating ε'/ε in the standard model
International Nuclear Information System (INIS)
Sharpe, S.R.
1988-01-01
The ingredients needed in order to calculate ε' and ε are described. Particular emphasis is given to the non-perturbative calculations of matrix elements by lattice methods. The status of the electromagnetic contribution to ε' is reviewed. 15 refs
Impact of image noise on gamma index calculation
International Nuclear Information System (INIS)
Chen, M; Mo, X; Parnell, D; Olivera, G; Galmarini, D; Lu, W
2014-01-01
Purpose: The Gamma Index defines an asymmetric metric between the evaluated image and the reference image. It provides a quantitative comparison that can be used to indicate sample-wised pass/fail on the agreement of the two images. The Gamma passing/failing rate has become an important clinical evaluation tool. However, the presence of noise in the evaluated and/or reference images may change the Gamma Index, hence the passing/failing rate, and further, clinical decisions. In this work, we systematically studied the impact of the image noise on the Gamma Index calculation. Methods: We used both analytic formulation and numerical calculations in our study. The numerical calculations included simulations and clinical images. Three different noise scenarios were studied in simulations: noise in reference images only, in evaluated images only, and in both. Both white and spatially correlated noises of various magnitudes were simulated. For clinical images of various noise levels, the Gamma Index of measurement against calculation, calculation against measurement, and measurement against measurement, were evaluated. Results: Numerical calculations for both the simulation and clinical data agreed with the analytic formulations, and the clinical data agreed with the simulations. For the Gamma Index of measurement against calculation, its distribution has an increased mean and an increased standard deviation as the noise increases. On the contrary, for the Gamma index of calculation against measurement, its distribution has a decreased mean and stabilized standard deviation as the noise increases. White noise has greater impact on the Gamma Index than spatially correlated noise. Conclusions: The noise has significant impact on the Gamma Index calculation and the impact is asymmetric. The Gamma Index should be reported along with the noise levels in both reference and evaluated images. Reporting of the Gamma Index with switched roles of the images as reference and
Impact of Image Noise on Gamma Index Calculation
Chen, M.; Mo, X.; Parnell, D.; Olivera, G.; Galmarini, D.; Lu, W.
2014-03-01
Purpose: The Gamma Index defines an asymmetric metric between the evaluated image and the reference image. It provides a quantitative comparison that can be used to indicate sample-wised pass/fail on the agreement of the two images. The Gamma passing/failing rate has become an important clinical evaluation tool. However, the presence of noise in the evaluated and/or reference images may change the Gamma Index, hence the passing/failing rate, and further, clinical decisions. In this work, we systematically studied the impact of the image noise on the Gamma Index calculation. Methods: We used both analytic formulation and numerical calculations in our study. The numerical calculations included simulations and clinical images. Three different noise scenarios were studied in simulations: noise in reference images only, in evaluated images only, and in both. Both white and spatially correlated noises of various magnitudes were simulated. For clinical images of various noise levels, the Gamma Index of measurement against calculation, calculation against measurement, and measurement against measurement, were evaluated. Results: Numerical calculations for both the simulation and clinical data agreed with the analytic formulations, and the clinical data agreed with the simulations. For the Gamma Index of measurement against calculation, its distribution has an increased mean and an increased standard deviation as the noise increases. On the contrary, for the Gamma index of calculation against measurement, its distribution has a decreased mean and stabilized standard deviation as the noise increases. White noise has greater impact on the Gamma Index than spatially correlated noise. Conclusions: The noise has significant impact on the Gamma Index calculation and the impact is asymmetric. The Gamma Index should be reported along with the noise levels in both reference and evaluated images. Reporting of the Gamma Index with switched roles of the images as reference and
Reference Monte Carlo calculations of Maria reactor core
International Nuclear Information System (INIS)
Andrzejewski, K.; Kulikowska, T.
2002-01-01
The reference Monte Carlo calculations of MARIA reactor core have been carried to evaluate accuracy of the calculations at each stage of its neutron-physics analysis using deterministic codes. The elementary cell has been calculated with two main goals; evaluation of effects of simplifications introduced in deterministic lattice spectrum calculations by the WIMS code and evaluation of library data in recently developed WIMS libraries. In particular the beryllium data of those libraries needed evaluation. The whole core calculations mainly the first MARIA critical experiment and the first critical core after the 8-year break in operation. Both cores contained only fresh fuel elements but only in the first critical core the beryllium blocks were not poisoned by Li-6 and He-3. Thus the MCNP k-eff results could be compared with the experiment. The MCNP calculations for the cores with beryllium poisoned suffered the deficiency of uncertainty in the poison concentration, but a comparison of power distribution shows that realistic poison levels have been carried out for the operating reactor MARIA configurations. (author)
Electromagnetic design calculation of the control rod drive mechanism
International Nuclear Information System (INIS)
Zhu Qirong; Zhu Jingchang
1991-01-01
Electromagnetic design calculation of the step-by-step magnetic jacking control rod drive mechanism includes magnetic field force calculation and design calculation of magnetomotive force for three electromagnetic iron and their coilds. The basic principle and method of electromagnetic design calculation had been expounded to take the lift magnet and lift coil for example
Molecular mechanics calculations on cobalt phthalocyanine dimers
Heuts, J.P.A.; Schipper, E.T.W.M.; Piet, P.; German, A.L.
1995-01-01
In order to obtain insight into the structure of cobalt phthalocyanine dimers, molecular mechanics calculations were performed on dimeric cobalt phthalocyanine species. Molecular mechanics calculations are first presented on monomeric cobalt(II) phthalocyanine. Using the Tripos force field for the
First principles calculations for analysis martensitic transformations
International Nuclear Information System (INIS)
Harmon, B.N.; Zhao, G.L.; Ho, K.M.; Chan, C.T.; Ye, Y.Y.; Ding, Y.; Zhang, B.L.
1993-01-01
The change in crystal energy is calculated for atomic displacements corresponding to phonons, elastic shears, and lattice transformations. Anomalies in the phonon dispersion curves of NiAl and NiTi are analyzed and recent calculations for TiPd alloys are presented
Void coefficient of reactivity calculation for AP-600 core
International Nuclear Information System (INIS)
Suparlina, L.; Budiono, T.A.; Mardha, A.; Tukiran
1998-01-01
Void coefficient of reactivity as one of reactor kinetics parameters has been carried out. The calculation was done into two steps which is cell calculation using WIMSD/4 and core calculation using Batan-2DIFF code programs with the condition of beginning of cycle with all fresh fuels elements and all control rods withdrawn. The one dimension transport program in four neutron energy groups is used to calculate the cell generation of various core materials cell has been calculated in 1/4 fuel element with cluster model and square pitch arrange. Moderator density have been reduced until 20% for the void coefficient of reactivity calculation. Macroscopic cross-section as the out put of WIMSD/4 is being used as the input at the diffusion neutron program for core calculation. The void coefficient of reactivity of the AP-600 core can be determined with regular neutron flux and adjoint in four energy groups and X-Y geometry. The results is shown that the K eff calculation value is different 5.2% from the design data
Validation of Calculations in a Digital Thermometer Firmware
Batagelj, V.; Miklavec, A.; Bojkovski, J.
2014-04-01
State-of-the-art digital thermometers are arguably remarkable measurement instruments, measuring outputs from resistance thermometers and/or thermocouples. Not only that they can readily achieve measuring accuracies in the parts-per-million range, but they also incorporate sophisticated algorithms for the transformation calculation of the measured resistance or voltage to temperature. These algorithms often include high-order polynomials, exponentials and logarithms, and must be performed using both standard coefficients and particular calibration coefficients. The numerical accuracy of these calculations and the associated uncertainty component must be much better than the accuracy of the raw measurement in order to be negligible in the total measurement uncertainty. In order for the end-user to gain confidence in these calculations as well as to conform to formal requirements of ISO/IEC 17025 and other standards, a way of validation of these numerical procedures performed in the firmware of the instrument is required. A software architecture which allows a simple validation of internal measuring instrument calculations is suggested. The digital thermometer should be able to expose all its internal calculation functions to the communication interface, so the end-user can compare the results of the internal measuring instrument calculation with reference results. The method can be regarded as a variation of the black-box software validation. Validation results on a thermometer prototype with implemented validation ability show that the calculation error of basic arithmetic operations is within the expected rounding error. For conversion functions, the calculation error is at least ten times smaller than the thermometer effective resolution for the particular probe type.
Assessing the reliability of calculated catalytic ammonia synthesis rates
DEFF Research Database (Denmark)
Medford, Andrew James; Wellendorff, Jess; Vojvodic, Aleksandra
2014-01-01
We introduce a general method for estimating the uncertainty in calculated materials properties based on density functional theory calculations. We illustrate the approach for a calculation of the catalytic rate of ammonia synthesis over a range of transition-metal catalysts. The correlation...... between errors in density functional theory calculations is shown to play an important role in reducing the predicted error on calculated rates. Uncertainties depend strongly on reaction conditions and catalyst material, and the relative rates between different catalysts are considerably better described...
Medication calculation skills of graduating nursing students in Finland.
Grandell-Niemi, H; Hupli, M; Leino-Kilpi, H
2001-01-01
The aim of this study was to describe the basic mathematical proficiency and the medication calculation skills of graduating nursing students in Finland. A further concern was with how students experienced the teaching of medication calculation. We wanted to find out whether these experiences were associated with various background factors and the students' medication calculation skills. In spring 1997 the population of graduating nursing students in Finland numbered around 1280; the figure for the whole year was 2640. A convenience sample of 204 students completed a questionnaire specially developed for this study. The instrument included structured questions, statements and a medication calculation test. The response rate was 88%. Data analysis was based on descriptive statistics. The students found it hard to learn mathematics and medication calculation skills. Those who evaluated their mathematical and medication calculation skills as sufficient successfully solved the problems included in the questionnaire. It was felt that the introductory course on medication calculation was uninteresting and poorly organised. Overall the students' mathematical skills were inadequate. One-fifth of the students failed to pass the medication calculation test. A positive correlation was shown between the student's grade in mathematics (Sixth Form College) and her skills in medication calculation.
Languages for structural calculations
International Nuclear Information System (INIS)
Thomas, J.B.; Chambon, M.R.
1988-01-01
The differences between human and computing languages are recalled. It is argued that they are to some extent structured in antagonistic ways. Languages in structural calculation, in the past, present, and future, are considered. The contribution of artificial intelligence is stressed [fr
Radiative flux calculations at UV and visible wavelengths
International Nuclear Information System (INIS)
Grossman, A.S.; Grant, K.E.; Wuebbles, D.J.
1993-10-01
A radiative transfer model to calculate the short wavelength fluxes at altitudes between 0 and 80 km has been developed at LLNL. The wavelength range extends from 175--735 nm. This spectral range covers the UV-B wavelength region, 250--350 nm, with sufficient resolution to allow comparison of UV-B measurements with theoretical predictions. Validation studies for the model have been made for both UV-B ground radiation calculations and tropospheric solar radiative forcing calculations for various ozone distributions. These studies indicate that the model produces results which agree well with respect to existing UV calculations from other published models
Quantifying environmental performance using an environmental footprint calculator
Energy Technology Data Exchange (ETDEWEB)
Smith, D.B.; Loney, A.C.; Chan, V. [Conestoga-Rovers & Associates, Waterloo, Ontario (Canada)
2009-07-01
This paper provides a case study using relevant key performance indicators (KPIs) to evaluate the environmental performance of a business. Using recognized calculation and reporting frameworks, Conestoga-Rovers & Associates (CRA) designed the Environmental Footprint Calculator to quantify the environmental performance of a Canadian construction materials company. CRA designed the Environmental Footprint calculator for our client to track and report their environmental performance in accordance with their targets, based on requirements of relevant guidance documents. The objective was to design a tool that effectively manages, calculates, and reports environmental performance to various stakeholders in a user-friendly format. (author)
Radionuclide inventory calculation in VVER and BWR reactor
International Nuclear Information System (INIS)
Bouhaddane, A.; Farkas, F.; Slugen, V.; Ackermann, L.; Schienbein, M.
2014-01-01
The paper shows different aspects in the radionuclide inventory determination. Precise determination of the neutron flux distribution, presented for a BRW reactor, is vital for the activation calculations. The precision can be improved utilizing variance reduction methods as importance treatment, weight windows etc. Direct calculation of the radionuclide inventory via Monte Carlo code is presented for a VVER reactor. Burn-up option utilized in this calculation appears to be proper for reactor internal components. However, it will not be probably effective outside the reactor core. Further calculations in this area are required to support the forth-set findings. (authors)
Quantifying environmental performance using an environmental footprint calculator
International Nuclear Information System (INIS)
Smith, D.B.; Loney, A.C.; Chan, V.
2009-01-01
This paper provides a case study using relevant key performance indicators (KPIs) to evaluate the environmental performance of a business. Using recognized calculation and reporting frameworks, Conestoga-Rovers & Associates (CRA) designed the Environmental Footprint Calculator to quantify the environmental performance of a Canadian construction materials company. CRA designed the Environmental Footprint calculator for our client to track and report their environmental performance in accordance with their targets, based on requirements of relevant guidance documents. The objective was to design a tool that effectively manages, calculates, and reports environmental performance to various stakeholders in a user-friendly format. (author)
40 CFR 98.413 - Calculating GHG emissions.
2010-07-01
... 40 Protection of Environment 20 2010-07-01 2010-07-01 false Calculating GHG emissions. 98.413 Section 98.413 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) MANDATORY GREENHOUSE GAS REPORTING Suppliers of Industrial Greenhouse Gases § 98.413 Calculating...
Manual method for dose calculation in gynecologic brachytherapy
International Nuclear Information System (INIS)
Vianello, Elizabeth A.; Almeida, Carlos E. de; Biaggio, Maria F. de
1998-01-01
This paper describes a manual method for dose calculation in brachytherapy of gynecological tumors, which allows the calculation of the doses at any plane or point of clinical interest. This method uses basic principles of vectorial algebra and the simulating orthogonal films taken from the patient with the applicators and dummy sources in place. The results obtained with method were compared with the values calculated with the values calculated with the treatment planning system model Theraplan and the agreement was better than 5% in most cases. The critical points associated with the final accuracy of the proposed method is related to the quality of the image and the appropriate selection of the magnification factors. This method is strongly recommended to the radiation oncology centers where are no treatment planning systems available and the dose calculations are manually done. (author)
Theoretical calculations of positron lifetimes for metal oxides
International Nuclear Information System (INIS)
Mizuno, Masataka; Araki, Hideki; Shirai, Yasuharu
2004-01-01
Our recent positron lifetime measurements for metal oxides suggest that positron lifetimes of bulk state in metal oxides are shorter than previously reported values. We have performed theoretical calculations of positron lifetimes for bulk and vacancy states in MgO and ZnO using first-principles electronic structure calculations and discuss the validity of positron lifetime calculations for insulators. By comparing the calculated positron lifetimes to the experimental values, it wa found that the semiconductor model well reproduces the experimental positron lifetime. The longer positron lifetime previously reported can be considered to arise from not only the bulk but also from the vacancy induced by impurities. In the case of cation vacancy, the calculated positron lifetime based on semiconductor model is shorter than the experimental value, which suggests that the inward relaxation occurs around the cation vacancy trapping the positron. (author)
Am/Cm Vitrification Process: Pretreatment Material Balance Calculations
International Nuclear Information System (INIS)
Smith, F.G.
2001-01-01
This report documents material balance calculations for the pretreatment steps required to prepare the Americium/Curium solution currently stored in Tank 17.1 in the F-Canyon for vitrification. The material balance uses the latest analysis of the tank contents to provide a best estimate calculation of the expected plant operations during the pretreatment process. The material balance calculations primarily follow the material that directly leads to melter feed. Except for vapor products of the denitration reactions and treatment of supernate from precipitation and precipitate washing, the flowsheet does not include side streams such as acid washes of the empty tanks that would go directly to waste. The calculation also neglects tank heels. This report consolidates previously reported results, corrects some errors found in the spreadsheet and provides a more detailed discussion of the calculation basis
Estimate of cryoscopic calculations accuracy from fusibility diagrams
International Nuclear Information System (INIS)
Viting, L.M.; Gorbovskaya, G.P.
1975-01-01
The melting points of some lead and zinc salts, that can be used as solvents for ferrites in systems: PbMoO 4 -MgFe 2 O 4 , Zn 2 V 2 O 7 -NiFe 2 O 4 , Pb 3 (VO 4 ) 2 -MgFe 2 O 4 , have been calculated in accordance with the hypotetical mechanism of the solvent dissociation. The accuracy of cryoscopic calculations based on melting point curves is evaluated. Cryoscopic calculations permit to determin the solvent activity with the accuracy of +-0.3% and the heat of its fusion, with the accuracy of +-3%. The comparison of the calculated and experimental values of the entropy of melting, as well as calculated and experimental values of the cryoscopic constant elucidates the mechanism of dissociation of both the dissolved compound and the solvent
Calculation of groundwater travel time
International Nuclear Information System (INIS)
Arnett, R.C.; Sagar, B.; Baca, R.G.
1984-12-01
Pre-waste-emplacement groundwater travel time is one indicator of the isolation capability of the geologic system surrounding a repository. Two distinct modeling approaches exist for prediction of groundwater flow paths and travel times from the repository location to the designated accessible environment boundary. These two approaches are: (1) the deterministic approach which calculates a single value prediction of groundwater travel time based on average values for input parameters and (2) the stochastic approach which yields a distribution of possible groundwater travel times as a function of the nature and magnitude of uncertainties in the model inputs. The purposes of this report are to (1) document the theoretical (i.e., mathematical) basis used to calculate groundwater pathlines and travel times in a basalt system, (2) outline limitations and ranges of applicability of the deterministic modeling approach, and (3) explain the motivation for the use of the stochastic modeling approach currently being used to predict groundwater pathlines and travel times for the Hanford Site. Example calculations of groundwater travel times are presented to highlight and compare the differences between the deterministic and stochastic modeling approaches. 28 refs
Reactor physics calculations in the Nordic countries
International Nuclear Information System (INIS)
Hoeglund, R.
1995-01-01
The seventh biennial meeting on reactor physics calculations in the Nordic countries was arranged by VTT Energy on May 8-9, 1995. 26 papers on different subjects in the field of reactor physics were presented by 45 participants representing research establishments, technical universities, utilities, consultants and suppliers. Resent development and verification of the program systems of ABB Atom, Risoe, Scandpower, Studsvik and VTT Energy were the main topic of the meeting. Benchmarking of the two assembly codes CASMO-4 and HELIOS is proceeding. Cross section data calculated with CASMO-HEX have been validated for the Loviisa reactors. On core analysis ABB atom gives a description on its latest core simulator version POLCA7 with the calculation Core Master 2 and the BWR core supervision system Core Watch. Transient calculations with HEXTRAN, HEXTRAN- PLIM, TRAB, RAMONA, SIMULATE-3K and a code based on PRESTO II/POLCA7 were also presented
AMBER and Ecolego Intercomparisons using Calculations from SR 97
International Nuclear Information System (INIS)
Maul, Philip; Robinson, Peter; Avila, Rodolfo; Broed, Robert; Pereira, Antonio
2003-08-01
The AMBER compartmental modelling software has been used by Quintessa in undertaking Performance Assessment (PA) calculations for SKI for the last four years. In particular, it was used in helping to identify key issues in the safety case for SFR 1. SSI and the Norwegian Radiation Protection Authority (NRPA) are supporting the development of Ecolego which is based on the Matlab/Simulink general purpose modelling software and provides a user-friendly Graphical User Interface which is particularly suitable for undertaking PA calculations. At SKI/SSI expert group meetings in September 2002, Quintessa presented some radionuclide transport calculations for a deep repository for spent fuel using AMBER. Comparisons were presented with calculations undertaken by SKB in their SR 97 study. It was agreed that it would be useful to undertake some intercomparison calculations using AMBER and Ecolego based on calculation cases considered in SR 97. The reasons for undertaking such intercomparisons include: 1. to give confidence in the use of AMBER and Ecolego for whole system PA calculations for a deep repository; and 2. to provide SKI and SSI with an understanding of some of the technical issues raised by seeking to repeat the SR 97 calculations with different models and software. This report documents the intercomparison calculations that have been undertaken. The main conclusions that have been drawn are: 1. Both modelling packages could successfully reproduce SKB's SR 97 calculations for the near field, geosphere and biosphere. 2. The results of the deterministic calculations for the near field and both deterministic and probabilistic calculations for the biosphere, where AMBER and Ecolego used compartment models, were typically in agreement to within one significant figure. The identical results for the calculated equilibrium inventories in Bay sediments illustrate that the calculations by AMBER and Ecolego are essentially the same when the compartment models are identical
Making sense of cancer risk calculators on the web.
Levy, Andrea Gurmankin; Sonnad, Seema S; Kurichi, Jibby E; Sherman, Melani; Armstrong, Katrina
2008-03-01
Cancer risk calculators on the internet have the potential to provide users with valuable information about their individual cancer risk. However, the lack of oversight of these sites raises concerns about low quality and inconsistent information. These concerns led us to evaluate internet cancer risk calculators. After a systematic search to find all cancer risk calculators on the internet, we reviewed the content of each site for information that users should seek to evaluate the quality of a website. We then examined the consistency of the breast cancer risk calculators by having 27 women complete 10 of the breast cancer risk calculators for themselves. We also completed the breast cancer risk calculators for a hypothetical high- and low-risk woman, and compared the output to Surveillance Epidemiology and End Results estimates for the average same-age and same-race woman. Nineteen sites were found, 13 of which calculate breast cancer risk. Most sites do not provide the information users need to evaluate the legitimacy of a website. The breast cancer calculator sites vary in the risk factors they assess to calculate breast cancer risk, how they operationalize each risk factor and in the risk estimate they provide for the same individual. Internet cancer risk calculators have the potential to provide a public health benefit by educating individuals about their risks and potentially encouraging preventive health behaviors. However, our evaluation of internet calculators revealed several problems that call into question the accuracy of the information that they provide. This may lead the users of these sites to make inappropriate medical decisions on the basis of misinformation.
CRACKEL: a computer code for CFR fuel management calculations
International Nuclear Information System (INIS)
Burstall, R.F.; Ball, M.A.; Thornton, D.E.J.
1975-12-01
The CRACKLE computer code is designed to perform rapid fuel management surveys of CFR systems. The code calculates overall features such as reactivity, power distributions and breeding gain, and also calculates for each sub-assembly plutonium content and power output. A number of alternative options are built into the code, in order to permit different fuel management strategies to be calculated, and to perform more detailed calculations when necessary. A brief description is given of the methods of calculation, and the input facilities of CRACKLE, with examples. (author)
Automatic calculations of electroweak processes
International Nuclear Information System (INIS)
Ishikawa, T.; Kawabata, S.; Kurihara, Y.; Shimizu, Y.; Kaneko, T.; Kato, K.; Tanaka, H.
1996-01-01
GRACE system is an excellent tool for calculating the cross section and for generating event of the elementary process automatically. However it is not always easy for beginners to use. An interactive version of GRACE is being developed so as to be a user friendly system. Since it works exactly in the same environment as PAW, all functions of PAW are available for handling any histogram information produced by GRACE. As its application the cross sections of all elementary processes with up to 5-body final states induced by e + e - interaction are going to be calculated and to be summarized as a catalogue. (author)
Benchmark neutron porosity log calculations
International Nuclear Information System (INIS)
Little, R.C.; Michael, M.; Verghese, K.; Gardner, R.P.
1989-01-01
Calculations have been made for a benchmark neutron porosity log problem with the general purpose Monte Carlo code MCNP and the specific purpose Monte Carlo code McDNL. For accuracy and timing comparison purposes the CRAY XMP and MicroVax II computers have been used with these codes. The CRAY has been used for an analog version of the MCNP code while the MicroVax II has been used for the optimized variance reduction versions of both codes. Results indicate that the two codes give the same results within calculated standard deviations. Comparisons are given and discussed for accuracy (precision) and computation times for the two codes
Lattice calculations in gauge theory
International Nuclear Information System (INIS)
Rebbi, C.
1985-01-01
The lattice formulation of quantum gauge theories is discussed as a viable technique for quantitative studies of nonperturbative effects in QCD. Evidence is presented to ascertain that whole classes of lattice actions produce a universal continuum limit. Discrepancies between numerical results from Monto Carlo simulations for the pure gauge system and for the system with gauge and quark fields are discussed. Numerical calculations for QCD require very substantial computational resources. The use of powerful vector processors of special purpose machines, in extending the scope and magnitude or the calculations is considered, and one may reasonably expect that in the near future good quantitative predictions will be obtained for QCD
Calculation of electron-helium scattering
International Nuclear Information System (INIS)
Fursa, D.V.; Bray, I.
1994-11-01
We present the Convergent Close-Coupling (CCC) theory for the calculation of electron-helium scattering. We demonstrate its applicability at a range of projectile energies of 1.5 to 500 eV to scattering from the ground state to n ≤3 states. Excellent agreement with experiment is obtained with the available differential, integrated, ionization, and total cross sections, as well as with the electron-impact coherence parameters up to and including the 3 3 D state excitation. Comparison with other theories demonstrates that the CCC theory is the only general reliable method for the calculation of electron helium scattering. (authors). 66 refs., 2 tabs., 24 figs
Accurate quantum chemical calculations
Bauschlicher, Charles W., Jr.; Langhoff, Stephen R.; Taylor, Peter R.
1989-01-01
An important goal of quantum chemical calculations is to provide an understanding of chemical bonding and molecular electronic structure. A second goal, the prediction of energy differences to chemical accuracy, has been much harder to attain. First, the computational resources required to achieve such accuracy are very large, and second, it is not straightforward to demonstrate that an apparently accurate result, in terms of agreement with experiment, does not result from a cancellation of errors. Recent advances in electronic structure methodology, coupled with the power of vector supercomputers, have made it possible to solve a number of electronic structure problems exactly using the full configuration interaction (FCI) method within a subspace of the complete Hilbert space. These exact results can be used to benchmark approximate techniques that are applicable to a wider range of chemical and physical problems. The methodology of many-electron quantum chemistry is reviewed. Methods are considered in detail for performing FCI calculations. The application of FCI methods to several three-electron problems in molecular physics are discussed. A number of benchmark applications of FCI wave functions are described. Atomic basis sets and the development of improved methods for handling very large basis sets are discussed: these are then applied to a number of chemical and spectroscopic problems; to transition metals; and to problems involving potential energy surfaces. Although the experiences described give considerable grounds for optimism about the general ability to perform accurate calculations, there are several problems that have proved less tractable, at least with current computer resources, and these and possible solutions are discussed.
Undergraduate paramedic students cannot do drug calculations
Eastwood, Kathryn; Boyle, Malcolm J; Williams, Brett
2012-01-01
BACKGROUND: Previous investigation of drug calculation skills of qualified paramedics has highlighted poor mathematical ability with no published studies having been undertaken on undergraduate paramedics. There are three major error classifications. Conceptual errors involve an inability to formulate an equation from information given, arithmetical errors involve an inability to operate a given equation, and finally computation errors are simple errors of addition, subtraction, division and multiplication. The objective of this study was to determine if undergraduate paramedics at a large Australia university could accurately perform common drug calculations and basic mathematical equations normally required in the workplace. METHODS: A cross-sectional study methodology using a paper-based questionnaire was administered to undergraduate paramedic students to collect demographical data, student attitudes regarding their drug calculation performance, and answers to a series of basic mathematical and drug calculation questions. Ethics approval was granted. RESULTS: The mean score of correct answers was 39.5% with one student scoring 100%, 3.3% of students (n=3) scoring greater than 90%, and 63% (n=58) scoring 50% or less, despite 62% (n=57) of the students stating they ‘did not have any drug calculations issues’. On average those who completed a minimum of year 12 Specialist Maths achieved scores over 50%. Conceptual errors made up 48.5%, arithmetical 31.1% and computational 17.4%. CONCLUSIONS: This study suggests undergraduate paramedics have deficiencies in performing accurate calculations, with conceptual errors indicating a fundamental lack of mathematical understanding. The results suggest an unacceptable level of mathematical competence to practice safely in the unpredictable prehospital environment. PMID:25215067
Undergraduate paramedic students cannot do drug calculations.
Eastwood, Kathryn; Boyle, Malcolm J; Williams, Brett
2012-01-01
Previous investigation of drug calculation skills of qualified paramedics has highlighted poor mathematical ability with no published studies having been undertaken on undergraduate paramedics. There are three major error classifications. Conceptual errors involve an inability to formulate an equation from information given, arithmetical errors involve an inability to operate a given equation, and finally computation errors are simple errors of addition, subtraction, division and multiplication. The objective of this study was to determine if undergraduate paramedics at a large Australia university could accurately perform common drug calculations and basic mathematical equations normally required in the workplace. A cross-sectional study methodology using a paper-based questionnaire was administered to undergraduate paramedic students to collect demographical data, student attitudes regarding their drug calculation performance, and answers to a series of basic mathematical and drug calculation questions. Ethics approval was granted. The mean score of correct answers was 39.5% with one student scoring 100%, 3.3% of students (n=3) scoring greater than 90%, and 63% (n=58) scoring 50% or less, despite 62% (n=57) of the students stating they 'did not have any drug calculations issues'. On average those who completed a minimum of year 12 Specialist Maths achieved scores over 50%. Conceptual errors made up 48.5%, arithmetical 31.1% and computational 17.4%. This study suggests undergraduate paramedics have deficiencies in performing accurate calculations, with conceptual errors indicating a fundamental lack of mathematical understanding. The results suggest an unacceptable level of mathematical competence to practice safely in the unpredictable prehospital environment.
Calculation of positron annihilation in high Tc copper oxides
International Nuclear Information System (INIS)
Barbiellini, B.; Jarlborg, T.; Gauthier, M.; Shukla, A.
1992-01-01
LMTO calculations have been done in order to obtain electronic and positronic states in YBaCuO compounds. The annihilation rates have been calculated including the enhancement effects. Folded momentum density spectra are calculated and used for prediction of Fermi surface signals. The influence of lattice defects on the spectra is studied by the comparison between experimental and calculated positron lifetimes. (orig.)
7 CFR 760.209 - Livestock payment calculations.
2010-01-01
... 7 Agriculture 7 2010-01-01 2010-01-01 false Livestock payment calculations. 760.209 Section 760..., DEPARTMENT OF AGRICULTURE SPECIAL PROGRAMS INDEMNITY PAYMENT PROGRAMS Emergency Assistance for Livestock, Honeybees, and Farm-Raised Fish Program § 760.209 Livestock payment calculations. (a) Payments for an...
Reactor physics calculations on HTR type configurations
Energy Technology Data Exchange (ETDEWEB)
Klippel, H.T.; Hogenbirk, A.; Stad, R.C.L. van der; Janssen, A.J.; Kuijper, J.C.; Levin, P.
1995-04-01
In this paper a short description of the ECN nuclear analysis code system is given with respect to application in HTR reactor physics calculations. First results of calculations performed on the PROTEUS benchmark are shown. Also first results of a HTGR benchmark are given. (orig.).
Reactor physics calculations on HTR type configurations
International Nuclear Information System (INIS)
Klippel, H.T.; Hogenbirk, A.; Stad, R.C.L. van der; Janssen, A.J.; Kuijper, J.C.; Levin, P.
1995-04-01
In this paper a short description of the ECN nuclear analysis code system is given with respect to application in HTR reactor physics calculations. First results of calculations performed on the PROTEUS benchmark are shown. Also first results of a HTGR benchmark are given. (orig.)
Hartree-Fock calculations of nuclear masses
International Nuclear Information System (INIS)
Quentin, P.
1976-01-01
Hartree-Fock calculations pertaining to the determination of nuclear binding energies throughout the whole chart of nuclides are reviewed. Such an approach is compared with other methods. Main techniques in use are shortly presented. Advantages and drawbacks of these calculations are also discussed with a special emphasis on the extrapolation towards nuclei far from the stability valley. Finally, a discussion of some selected results from light to superheavy nuclei, is given [fr
A revised calculational model for fission
Energy Technology Data Exchange (ETDEWEB)
Atchison, F
1998-09-01
A semi-empirical parametrization has been developed to calculate the fission contribution to evaporative de-excitation of nuclei with a very wide range of charge, mass and excitation-energy and also the nuclear states of the scission products. The calculational model reproduces measured values (cross-sections, mass distributions, etc.) for a wide range of fissioning systems: Nuclei from Ta to Cf, interactions involving nucleons up to medium energy and light ions. (author)
Ti-84 Plus graphing calculator for dummies
McCalla
2013-01-01
Get up-to-speed on the functionality of your TI-84 Plus calculator Completely revised to cover the latest updates to the TI-84 Plus calculators, this bestselling guide will help you become the most savvy TI-84 Plus user in the classroom! Exploring the standard device, the updated device with USB plug and upgraded memory (the TI-84 Plus Silver Edition), and the upcoming color screen device, this book provides you with clear, understandable coverage of the TI-84's updated operating system. Details the new apps that are available for download to the calculator via the USB cabl
Importance iteration in MORSE Monte Carlo calculations
International Nuclear Information System (INIS)
Kloosterman, J.L.; Hoogenboom, J.E.
1994-01-01
An expression to calculate point values (the expected detector response of a particle emerging from a collision or the source) is derived and implemented in the MORSE-SGC/S Monte Carlo code. It is outlined how these point values can be smoothed as a function of energy and as a function of the optical thickness between the detector and the source. The smoothed point values are subsequently used to calculate the biasing parameters of the Monte Carlo runs to follow. The method is illustrated by an example that shows that the obtained biasing parameters lead to a more efficient Monte Carlo calculation
Importance iteration in MORSE Monte Carlo calculations
International Nuclear Information System (INIS)
Kloosterman, J.L.; Hoogenboom, J.E.
1994-02-01
An expression to calculate point values (the expected detector response of a particle emerging from a collision or the source) is derived and implemented in the MORSE-SGC/S Monte Carlo code. It is outlined how these point values can be smoothed as a function of energy and as a function of the optical thickness between the detector and the source. The smoothed point values are subsequently used to calculate the biasing parameters of the Monte Carlo runs to follow. The method is illustrated by an example, which shows that the obtained biasing parameters lead to a more efficient Monte Carlo calculation. (orig.)
Fully self-consistent GW calculations for molecules
DEFF Research Database (Denmark)
Rostgaard, Carsten; Jacobsen, Karsten Wedel; Thygesen, Kristian Sommer
2010-01-01
We calculate single-particle excitation energies for a series of 34 molecules using fully self-consistent GW, one-shot G0W0, Hartree-Fock (HF), and hybrid density-functional theory (DFT). All calculations are performed within the projector-augmented wave method using a basis set of Wannier...... functions augmented by numerical atomic orbitals. The GW self-energy is calculated on the real frequency axis including its full frequency dependence and off-diagonal matrix elements. The mean absolute error of the ionization potential (IP) with respect to experiment is found to be 4.4, 2.6, 0.8, 0.4, and 0...
Determination of orbitals for use in configuration interaction calculations
International Nuclear Information System (INIS)
Dunning, T.H. Jr.; Davidson, E.R.; Ruedenberg, K.; Hinze, J.
1978-01-01
For a full configuration interaction (CI) calculation the choice of orbitals is completely irrelevant, i.e., the calculated wavefunction is unaffected by an arbitrary unitary transformation of the orbitals; it depends only on the space spanned by the original basis set. For most chemical systems it is not possible to realistically carry out a full CI calculation, so that specification of the orbital set is important. Even for less-than-full CI calculations, it can be shown, however, that for certain types of calculations the wavefunction is unaffected by restricted transformations among the orbital set. For example, for CI calculations based on a single configuration plus a complete set of excitations of a given type (single, double, etc.), the calculated wavefunction is independent of transformations among the set of occupied orbitals and among the set of virtual orbitals. The wavefunction does, however, depend on transformations which mix the occupied and virtual orbitals
Advances in supercell calculation methods and comparison with measurements
Energy Technology Data Exchange (ETDEWEB)
Arsenault, B [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Baril, R; Hotte, G [Hydro-Quebec, Central Nucleaire Gentilly, Montreal, Quebec (Canada)
1996-07-01
In the last few years, modelling techniques have been developed in new supercell computer codes. These techniques have been used to model the CANDU reactivity devices. One technique is based on one- and two-dimensional transport calculations with the WIMS-AECL lattice code followed by super homogenization and three-dimensional flux calculations in a modified version of the MULTICELL code. The second technique is based on two- and three-dimensional transport calculations in DRAGON. The code calculates the lattice properties by solving the transport equation in a two-dimensional geometry followed by supercell calculations in three dimensions. These two calculation schemes have been used to calculate the incremental macroscopic properties of CANDU reactivity devices. The supercell size has also been modified to define incremental properties over a larger region. The results show improved agreement between the reactivity worth of zone controllers and adjusters. However, at the same time the agreement between measured and simulated flux distributions deteriorated somewhat. (author)
JNC results of BN-600 benchmark calculation (phase 4)
International Nuclear Information System (INIS)
Ishikawa, Makoto
2003-01-01
The present work is the results of JNC, Japan, for the Phase 4 of the BN-600 core benchmark problem (Hex-Z fully MOX fuelled core model) organized by IAEA. The benchmark specification is based on 1) the RCM report of IAEA CRP on 'Updated Codes and Methods to Reduce the Calculational Uncertainties of LMFR Reactivity Effects, Action 3.12' (Calculations for BN-600 fully fuelled MOX core for subsequent transient analyses). JENDL-3.2 nuclear data library was used for calculating 70 group ABBN-type group constants. Cell models for fuel assembly and control rod calculations were applied: homogeneous and heterogeneous (cylindrical supercell) model. Basic diffusion calculation was three-dimensional Hex-Z model, 18 group (Citation code). Transport calculations were 18 group, three-dimensional (NSHEC code) based on Sn-transport nodal method developed at JNC. The generated thermal power per fission was based on Sher's data corrected on the basis of ENDF/B-IV data library. Calculation results are presented in Tables for intercomparison
Ab initio theory and calculations of X-ray spectra
International Nuclear Information System (INIS)
Rehr, J.J.; Kas, J.J.; Prange, M.P.; Sorini, A.P.; Takimoto, Y.; Vila, F.
2009-01-01
There has been dramatic progress in recent years both in the calculation and interpretation of various x-ray spectroscopies. However, current theoretical calculations often use a number of simplified models to account for many-body effects, in lieu of first principles calculations. In an effort to overcome these limitations we describe in this article a number of recent advances in theory and in theoretical codes which offer the prospect of parameter free calculations that include the dominant many-body effects. These advances are based on ab initio calculations of the dielectric and vibrational response of a system. Calculations of the dielectric function over a broad spectrum yield system dependent self-energies and mean-free paths, as well as intrinsic losses due to multielectron excitations. Calculations of the dynamical matrix yield vibrational damping in terms of multiple-scattering Debye-Waller factors. Our ab initio methods for determining these many-body effects have led to new, improved, and broadly applicable x-ray and electron spectroscopy codes. (authors)
Hybrid reduced order modeling for assembly calculations
Energy Technology Data Exchange (ETDEWEB)
Bang, Youngsuk, E-mail: ysbang00@fnctech.com [FNC Technology, Co. Ltd., Yongin-si (Korea, Republic of); Abdel-Khalik, Hany S., E-mail: abdelkhalik@purdue.edu [Purdue University, West Lafayette, IN (United States); Jessee, Matthew A., E-mail: jesseema@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mertyurek, Ugur, E-mail: mertyurek@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN (United States)
2015-12-15
Highlights: • Reducing computational cost in engineering calculations. • Reduced order modeling algorithm for multi-physics problem like assembly calculation. • Non-intrusive algorithm with random sampling. • Pattern recognition in the components with high sensitive and large variation. - Abstract: While the accuracy of assembly calculations has considerably improved due to the increase in computer power enabling more refined description of the phase space and use of more sophisticated numerical algorithms, the computational cost continues to increase which limits the full utilization of their effectiveness for routine engineering analysis. Reduced order modeling is a mathematical vehicle that scales down the dimensionality of large-scale numerical problems to enable their repeated executions on small computing environment, often available to end users. This is done by capturing the most dominant underlying relationships between the model's inputs and outputs. Previous works demonstrated the use of the reduced order modeling for a single physics code, such as a radiation transport calculation. This manuscript extends those works to coupled code systems as currently employed in assembly calculations. Numerical tests are conducted using realistic SCALE assembly models with resonance self-shielding, neutron transport, and nuclides transmutation/depletion models representing the components of the coupled code system.
Hybrid reduced order modeling for assembly calculations
International Nuclear Information System (INIS)
Bang, Youngsuk; Abdel-Khalik, Hany S.; Jessee, Matthew A.; Mertyurek, Ugur
2015-01-01
Highlights: • Reducing computational cost in engineering calculations. • Reduced order modeling algorithm for multi-physics problem like assembly calculation. • Non-intrusive algorithm with random sampling. • Pattern recognition in the components with high sensitive and large variation. - Abstract: While the accuracy of assembly calculations has considerably improved due to the increase in computer power enabling more refined description of the phase space and use of more sophisticated numerical algorithms, the computational cost continues to increase which limits the full utilization of their effectiveness for routine engineering analysis. Reduced order modeling is a mathematical vehicle that scales down the dimensionality of large-scale numerical problems to enable their repeated executions on small computing environment, often available to end users. This is done by capturing the most dominant underlying relationships between the model's inputs and outputs. Previous works demonstrated the use of the reduced order modeling for a single physics code, such as a radiation transport calculation. This manuscript extends those works to coupled code systems as currently employed in assembly calculations. Numerical tests are conducted using realistic SCALE assembly models with resonance self-shielding, neutron transport, and nuclides transmutation/depletion models representing the components of the coupled code system.
Ab-initio calculations for dilute magnetic semiconductors
Energy Technology Data Exchange (ETDEWEB)
Belhadji, Brahim
2008-03-03
This thesis focusses on ab-initio calculations for the electronic structure and the magnetic properties of dilute magnetic semiconductors (DMS). In particular we aim at the understanding of the complex exchange interactions in these systems. Our calculations are based on density functional theory, being ideally suited for a description of the material specific properties of the considered DMS. Moreover we use the KKR Green function method in connection with the coherent potential approximation (CPA), which allows to include the random substitutional disorder in a mean field-like approximation for the electronic structure. Finally we calculate the exchange coupling constants J{sub ij} between two impurities in a CPA medium by using the Lichtenstein formula and from this calculate the Curie temperature by a numerically exact Monte Carlo method. Based on this analysis we found and investigated four different exchange mechanisms being of importance in DMS systems: Double exchange, p-d exchange, antiferromagnetic superexchanges, and ferromagnetic superexchange. A second topic we have investigated in this thesis is the pressure dependence of the exchange interactions and the Curie temperatures in (Ga,Mn)As and (In,Mn)As, using the LDA and the LDA+U approximations. Exact calculations of T{sub C} by Monte Carlo simulations show a somehow different behavior. (orig.)
Application of backtracking algorithm to depletion calculations
International Nuclear Information System (INIS)
Wu Mingyu; Wang Shixi; Yang Yong; Zhang Qiang; Yang Jiayin
2013-01-01
Based on the theory of linear chain method for analytical depletion calculations, the burnup matrix is decoupled by the divide and conquer strategy and the linear chain with Markov characteristic is formed. The density, activity and decay heat of every nuclide in the chain then can be calculated by analytical solutions. Every possible reaction path of the nuclide must be considered during the linear chain establishment process. To confirm the calculation precision and efficiency, the algorithm which can cover all the reaction paths and search the paths automatically according to the problem description and precision restrictions should be found. Through analysis and comparison of several kinds of searching algorithms, the backtracking algorithm was selected to establish and calculate the linear chains in searching process using depth first search (DFS) method, forming an algorithm which can solve the depletion problem adaptively and with high fidelity. The complexity of the solution space and time was analyzed by taking into account depletion process and the characteristics of the backtracking algorithm. The newly developed depletion program was coupled with Monte Carlo program MCMG-Ⅱ to calculate the benchmark burnup problem of the first core of China Experimental Fast Reactor (CEFR) and the preliminary verification and validation of the program were performed. (authors)
Detection limit calculations for different total reflection techniques
International Nuclear Information System (INIS)
Sanchez, H.J.
2000-01-01
In this work, theoretical calculations of detection limits for different total-reflection techniques are presented.. Calculations include grazing incidence (TXRF) and gracing exit (GEXRF) conditions. These calculations are compared with detection limits obtained for conventional x-ray fluorescence (XRF). In order to compute detection limits the Shiraiwa and Fujino's model to calculate x-ray fluorescence intensities was used. This model made certain assumptions and approximations to achieve the calculations, specially in the case of the geometrical conditions of the sample, and the incident and takeoff beams. Nevertheless the calculated data of detection limits for conventional XRF and total-reflection XRF show a good agreement with previous results. The model proposed here allows to analyze the different sources of background and the influence of the excitation geometry, which contribute to the understanding of the physical processes involved in the XRF analysis by total reflection. Finally, a comparison between detection limits in total-reflection analysis at grazing incidence and at grazing exit is carried out. Here a good agreement with the theoretical predictions of the reversibility principle is found, showing that detection limits are similar for both techniques. (author)
3-D calculations for comparison with the experiments
Energy Technology Data Exchange (ETDEWEB)
Alrsen, A M; Bosser, R
1973-09-27
In order to analyse the axial power profile measurements an attempt has been made to do full 3-D calculations for the Dragon reactor. The calculations are still at a very early stage, but the methods used will be outlined here together with the plans for investigations to be carried out in the near future. Some preliminary-results are reported as no final results have yet been obtained. 3-D calculations are rather expensive because of the computer time consumption. It is therefore essential, before too many big computer jobs are spent, to find approximations which can save calculation time. On the other hand some savings, for instance in the number of mesh points, may cause totally wrong results. The ''proper'' calculations have therefore to be proceeded by a number of preliminary investigations, to ensure optimum accuracy and computer expenses. This report contains some of these preliminary studies.
Variational calculation for the ground state of 12C
International Nuclear Information System (INIS)
Consoni, L.H.A.; Coelho, H.T.; Das, T.K.
1983-01-01
A variational calculation is done for the ground state of a 3α-particle system. Two simple trial wavefunctions are used and results are compared with an exact calculation done by the Hyperspherical Harmonic method. A modifed Ali-Bodmer potential for the α-α interaction is considered for all calculations. It is found that these simple wave functions can be very useful for phenomenological calculations. (Author) [pt
Calculation of the superconducting transition temperature in niobium
International Nuclear Information System (INIS)
Perlov, C.M.
1982-01-01
The author presents calculations of the superconducting transition temperature, T/sub c/, the electron-phonon coupling constant, lambda, and the spectral function, α 2 f(ω), for niobium. The author's calculations are based on an empirical pseudopotential method (EPM) band structure. Phonon linewidths are also given for longitudinal and transverse branches along different directions. The necessary electron-phonon matrix elements are evaluated using only the rigid-ion approximation by applying Green's theorem. The calculated value of T/sub c/ is 8.4 K which differs from the measured value by only 9%; the calculated lambda is 1.02. The spectral function and linewidths are compared to experimental and previous theoretical results
Data base to compare calculations and observations
International Nuclear Information System (INIS)
Tichler, J.L.
1985-01-01
Meteorological and climatological data bases were compared with known tritium release points and diffusion calculations to determine if calculated concentrations could replace measure concentrations at the monitoring stations. Daily tritium concentrations were monitored at 8 stations and 16 possible receptors. Automated data retrieval strategies are listed
Benchmark calculations of power distribution within assemblies
International Nuclear Information System (INIS)
Cavarec, C.; Perron, J.F.; Verwaerde, D.; West, J.P.
1994-09-01
The main objective of this Benchmark is to compare different techniques for fine flux prediction based upon coarse mesh diffusion or transport calculations. We proposed 5 ''core'' configurations including different assembly types (17 x 17 pins, ''uranium'', ''absorber'' or ''MOX'' assemblies), with different boundary conditions. The specification required results in terms of reactivity, pin by pin fluxes and production rate distributions. The proposal for these Benchmark calculations was made by J.C. LEFEBVRE, J. MONDOT, J.P. WEST and the specification (with nuclear data, assembly types, core configurations for 2D geometry and results presentation) was distributed to correspondents of the OECD Nuclear Energy Agency. 11 countries and 19 companies answered the exercise proposed by this Benchmark. Heterogeneous calculations and homogeneous calculations were made. Various methods were used to produce the results: diffusion (finite differences, nodal...), transport (P ij , S n , Monte Carlo). This report presents an analysis and intercomparisons of all the results received
Dose rate calculations for a reconnaissance vehicle
International Nuclear Information System (INIS)
Grindrod, L.; Mackey, J.; Salmon, M.; Smith, C.; Wall, S.
2005-01-01
A Chemical Nuclear Reconnaissance System (CNRS) has been developed by the British Ministry of Defence to make chemical and radiation measurements on contaminated terrain using appropriate sensors and recording equipment installed in a land rover. A research programme is under way to develop and validate a predictive capability to calculate the build-up of contamination on the vehicle, radiation detector performance and dose rates to the occupants of the vehicle. This paper describes the geometric model of the vehicle and the methodology used for calculations of detector response. Calculated dose rates obtained using the MCBEND Monte Carlo radiation transport computer code in adjoint mode are presented. These address the transient response of the detectors as the vehicle passes through a contaminated area. Calculated dose rates were found to agree with the measured data to be within the experimental uncertainties, thus giving confidence in the shielding model of the vehicle and its application to other scenarios. (authors)
Microscopic calculation of the 4He system
International Nuclear Information System (INIS)
Hofmann, H.M.
1996-01-01
We report on a consistent, microscopic calculation of the bound and scattering states in the 4 He system employing a realistic nucleon-nucleon potential in the framework of the resonating group model (RGM). We present for comparison with these microscopic RGM calculations the results from a charge-independent, Coulomb-corrected R-matrix analysis of all types of data for reactions in the A=4 system. Comparisons are made between the phase shifts, and with a selection of measurements from each reaction, as well as between the resonance spectra obtained from both calculations. In general, the comparisons are favorable, but distinct differences are observed between the RGM calculations and some of the polarisation data. The partial-wave decomposition of the experimental data produced by the R-matrix analysis shows that these differences can be attributed to just a few S-matrix elements, for which inadequate tensor-force strength in the N-N interaction used appears to be responsible. (orig.)
Calculated fission properties of the heaviest elements
International Nuclear Information System (INIS)
Moeller, P.; Nix, J.R.; Swiatecki, W.J.
1986-09-01
A quantitative calculation is presented that shows where high-kinetic-energy symmetric fission occurs and why it is associated with a sudden and large decrease in fission half-lives. The study is based on calculations of potential-energy surfaces in the macroscopic-microscopic model and a semi-empirical model for the nuclear inertia. For the macroscopic part a Yukawa-plus-exponential model is used and for the microscopic part a folded-Yukawa single-particle potential is used. The three-quadratic-surface parameterization generates shapes for which the potential-energy surfaces are calculated. The use of this parameterization and the use of the finite-range macroscopic model allows for the study of two touching spheres and similar shapes. The results of the calculations in terms of potential-energy surfaces and fission half-lives are presented for heavy even nuclei. The surfaces are displayed in the form of contour diagrams as functions of two moments of the shape. 53 refs., 15 figs., 1 tab
Calculation analysis on steady state natural circulation characteristics
International Nuclear Information System (INIS)
Wang Fei; Nie Changhua; Huang Yanping
2005-01-01
The calculation results of single-phase steady state natural circulation characteristics by using Retran02 code have been presented, good agreement is achieved between the verified calculation result and the experimental data which were conducted at a test facility. Based on the calculation model, some sensibility analyses were made and much deeper understanding for single-phase steady state natural circulation characteristics was obtained. (author)
Comparison of calculational methods for EBT reactor nucleonics
International Nuclear Information System (INIS)
Henninger, R.J.; Seed, T.J.; Soran, P.D.; Dudziak, D.J.
1980-01-01
Nucleonic calculations for a preliminary conceptual design of the first wall/blanket/shield/coil assembly for an EBT reactor are described. Two-dimensional Monte Carlo, and one- and two-dimensional discrete-ordinates calculations are compared. Good agreement for the calculated values of tritium breeding and nuclear heating is seen. We find that the three methods are all useful and complementary as a design of this type evolves
New developments in multireference and complete configuration interaction calculations
International Nuclear Information System (INIS)
Knowles, P.J.; Werner, H.J.
1987-01-01
Some recently developed techniques for the calculation of Hamiltonian matrix elements in molecular electronic structure calculations are described. These techniques allow the very rapid calculation, in any desired order, of one particle coupling coefficients between spin symmetry adapted basis functions of arbitrary structure. The matrix elements that are required, for either internally contracted multireference CI calculations, or full CI calculations, are then obtainable from suitable summations over resolutions of the identity, which has been shown previously to be rather efficient; this is especially true on vector computers, since all arithmetic can be formulated as matrix multiplications. These ideas have culminated in the preparation of a new multireference CI program which is capable of handling very large numbers of reference configurations. Application of the new techniques to full CI calculations are also presented
Validation of dose calculation programmes for recycling
International Nuclear Information System (INIS)
Menon, Shankar; Brun-Yaba, Christine; Yu, Charley; Cheng, Jing-Jy; Williams, Alexander
2002-12-01
This report contains the results from an international project initiated by the SSI in 1999. The primary purpose of the project was to validate some of the computer codes that are used to estimate radiation doses due to the recycling of scrap metal. The secondary purpose of the validation project was to give a quantification of the level of conservatism in clearance levels based on these codes. Specifically, the computer codes RESRAD-RECYCLE and CERISE were used to calculate radiation doses to individuals during the processing of slightly contaminated material, mainly in Studsvik, Sweden. Calculated external doses were compared with measured data from different steps of the process. The comparison of calculations and measurements shows that the computer code calculations resulted in both overestimations and underestimations of the external doses for different recycling activities. The SSI draws the conclusion that the accuracy is within one order of magnitude when experienced modellers use their programmes to calculate external radiation doses for a recycling process involving material that is mainly contaminated with cobalt-60. No errors in the codes themselves were found. Instead, the inaccuracy seems to depend mainly on the choice of some modelling parameters related to the receptor (e.g., distance, time, etc.) and simplifications made to facilitate modelling with the codes (e.g., object geometry). Clearance levels are often based on studies on enveloping scenarios that are designed to cover all realistic exposure pathways. It is obvious that for most practical cases, this gives a margin to the individual dose constraint (in the order of 10 micro sievert per year within the EC). This may be accentuated by the use of conservative assumptions when modelling the enveloping scenarios. Since there can obviously be a fairly large inaccuracy in the calculations, it seems reasonable to consider some degree of conservatism when establishing clearance levels based on
Validation of dose calculation programmes for recycling
Energy Technology Data Exchange (ETDEWEB)
Menon, Shankar [Menon Consulting, Nykoeping (Sweden); Brun-Yaba, Christine [Inst. de Radioprotection et Securite Nucleaire (France); Yu, Charley; Cheng, Jing-Jy [Argonne National Laboratory, IL (United States). Environmental Assessment Div.; Bjerler, Jan [Studsvik Stensand, Nykoeping (Sweden); Williams, Alexander [Dept. of Energy (United States). Office of Environmental Management
2002-12-01
This report contains the results from an international project initiated by the SSI in 1999. The primary purpose of the project was to validate some of the computer codes that are used to estimate radiation doses due to the recycling of scrap metal. The secondary purpose of the validation project was to give a quantification of the level of conservatism in clearance levels based on these codes. Specifically, the computer codes RESRAD-RECYCLE and CERISE were used to calculate radiation doses to individuals during the processing of slightly contaminated material, mainly in Studsvik, Sweden. Calculated external doses were compared with measured data from different steps of the process. The comparison of calculations and measurements shows that the computer code calculations resulted in both overestimations and underestimations of the external doses for different recycling activities. The SSI draws the conclusion that the accuracy is within one order of magnitude when experienced modellers use their programmes to calculate external radiation doses for a recycling process involving material that is mainly contaminated with cobalt-60. No errors in the codes themselves were found. Instead, the inaccuracy seems to depend mainly on the choice of some modelling parameters related to the receptor (e.g., distance, time, etc.) and simplifications made to facilitate modelling with the codes (e.g., object geometry). Clearance levels are often based on studies on enveloping scenarios that are designed to cover all realistic exposure pathways. It is obvious that for most practical cases, this gives a margin to the individual dose constraint (in the order of 10 micro sievert per year within the EC). This may be accentuated by the use of conservative assumptions when modelling the enveloping scenarios. Since there can obviously be a fairly large inaccuracy in the calculations, it seems reasonable to consider some degree of conservatism when establishing clearance levels based on
Faas, S.; Snijders, Jaap; van Lenthe, J.H.; HernandezLaguna, A; Maruani, J; McWeeny, R; Wilson, S
2000-01-01
In this paper we present the first application of the ZORA (Zeroth Order Regular Approximation of the Dirac Fock equation) formalism in Ab Initio electronic structure calculations. The ZORA method, which has been tested previously in the context of Density Functional Theory, has been implemented in
Weldon Spring dose calculations
International Nuclear Information System (INIS)
Dickson, H.W.; Hill, G.S.; Perdue, P.T.
1978-09-01
In response to a request by the Oak Ridge Operations (ORO) Office of the Department of Energy (DOE) for assistance to the Department of the Army (DA) on the decommissioning of the Weldon Spring Chemical Plant, the Health and Safety Research Division of the Oak Ridge National Laboratory (ORNL) performed limited dose assessment calculations for that site. Based upon radiological measurements from a number of soil samples analyzed by ORNL and from previously acquired radiological data for the Weldon Spring site, source terms were derived to calculate radiation doses for three specific site scenarios. These three hypothetical scenarios are: a wildlife refuge for hunting, fishing, and general outdoor recreation; a school with 40 hr per week occupancy by students and a custodian; and a truck farm producing fruits, vegetables, meat, and dairy products which may be consumed on site. Radiation doses are reported for each of these scenarios both for measured uranium daughter equilibrium ratios and for assumed secular equilibrium. Doses are lower for the nonequilibrium case
Guidelines for the analysis of free energy calculations.
Klimovich, Pavel V; Shirts, Michael R; Mobley, David L
2015-05-01
Free energy calculations based on molecular dynamics simulations show considerable promise for applications ranging from drug discovery to prediction of physical properties and structure-function studies. But these calculations are still difficult and tedious to analyze, and best practices for analysis are not well defined or propagated. Essentially, each group analyzing these calculations needs to decide how to conduct the analysis and, usually, develop its own analysis tools. Here, we review and recommend best practices for analysis yielding reliable free energies from molecular simulations. Additionally, we provide a Python tool, alchemical-analysis.py, freely available on GitHub as part of the pymbar package (located at http://github.com/choderalab/pymbar), that implements the analysis practices reviewed here for several reference simulation packages, which can be adapted to handle data from other packages. Both this review and the tool covers analysis of alchemical calculations generally, including free energy estimates via both thermodynamic integration and free energy perturbation-based estimators. Our Python tool also handles output from multiple types of free energy calculations, including expanded ensemble and Hamiltonian replica exchange, as well as standard fixed ensemble calculations. We also survey a range of statistical and graphical ways of assessing the quality of the data and free energy estimates, and provide prototypes of these in our tool. We hope this tool and discussion will serve as a foundation for more standardization of and agreement on best practices for analysis of free energy calculations.
International Nuclear Information System (INIS)
Sharma, Subhash; Ott, Joseph; Williams, Jamone; Dickow, Danny
2011-01-01
Monte Carlo dose calculation algorithms have the potential for greater accuracy than traditional model-based algorithms. This enhanced accuracy is particularly evident in regions of lateral scatter disequilibrium, which can develop during treatments incorporating small field sizes and low-density tissue. A heterogeneous slab phantom was used to evaluate the accuracy of several commercially available dose calculation algorithms, including Monte Carlo dose calculation for CyberKnife, Analytical Anisotropic Algorithm and Pencil Beam convolution for the Eclipse planning system, and convolution-superposition for the Xio planning system. The phantom accommodated slabs of varying density; comparisons between planned and measured dose distributions were accomplished with radiochromic film. The Monte Carlo algorithm provided the most accurate comparison between planned and measured dose distributions. In each phantom irradiation, the Monte Carlo predictions resulted in gamma analysis comparisons >97%, using acceptance criteria of 3% dose and 3-mm distance to agreement. In general, the gamma analysis comparisons for the other algorithms were <95%. The Monte Carlo dose calculation algorithm for CyberKnife provides more accurate dose distribution calculations in regions of lateral electron disequilibrium than commercially available model-based algorithms. This is primarily because of the ability of Monte Carlo algorithms to implicitly account for tissue heterogeneities, density scaling functions; and/or effective depth correction factors are not required.
Broyden's method in nuclear structure calculations
International Nuclear Information System (INIS)
Baran, Andrzej; Bulgac, Aurel; Forbes, Michael McNeil; Hagen, Gaute; Nazarewicz, Witold; Schunck, Nicolas; Stoitsov, Mario V.
2008-01-01
Broyden's method, widely used in quantum chemistry electronic-structure calculations for the numerical solution of nonlinear equations in many variables, is applied in the context of the nuclear many-body problem. Examples include the unitary gas problem, the nuclear density functional theory with Skyrme functionals, and the nuclear coupled-cluster theory. The stability of the method, its ease of use, and its rapid convergence rates make Broyden's method a tool of choice for large-scale nuclear structure calculations
eQuilibrator--the biochemical thermodynamics calculator.
Flamholz, Avi; Noor, Elad; Bar-Even, Arren; Milo, Ron
2012-01-01
The laws of thermodynamics constrain the action of biochemical systems. However, thermodynamic data on biochemical compounds can be difficult to find and is cumbersome to perform calculations with manually. Even simple thermodynamic questions like 'how much Gibbs energy is released by ATP hydrolysis at pH 5?' are complicated excessively by the search for accurate data. To address this problem, eQuilibrator couples a comprehensive and accurate database of thermodynamic properties of biochemical compounds and reactions with a simple and powerful online search and calculation interface. The web interface to eQuilibrator (http://equilibrator.weizmann.ac.il) enables easy calculation of Gibbs energies of compounds and reactions given arbitrary pH, ionic strength and metabolite concentrations. The eQuilibrator code is open-source and all thermodynamic source data are freely downloadable in standard formats. Here we describe the database characteristics and implementation and demonstrate its use.
eQuilibrator—the biochemical thermodynamics calculator
Flamholz, Avi; Noor, Elad; Bar-Even, Arren; Milo, Ron
2012-01-01
The laws of thermodynamics constrain the action of biochemical systems. However, thermodynamic data on biochemical compounds can be difficult to find and is cumbersome to perform calculations with manually. Even simple thermodynamic questions like ‘how much Gibbs energy is released by ATP hydrolysis at pH 5?’ are complicated excessively by the search for accurate data. To address this problem, eQuilibrator couples a comprehensive and accurate database of thermodynamic properties of biochemical compounds and reactions with a simple and powerful online search and calculation interface. The web interface to eQuilibrator (http://equilibrator.weizmann.ac.il) enables easy calculation of Gibbs energies of compounds and reactions given arbitrary pH, ionic strength and metabolite concentrations. The eQuilibrator code is open-source and all thermodynamic source data are freely downloadable in standard formats. Here we describe the database characteristics and implementation and demonstrate its use. PMID:22064852
Radionuclide release calculations for SAR-08
International Nuclear Information System (INIS)
Thomson, Gavin; Miller, Alex; Smith, Graham; Jackson, Duncan
2008-04-01
Following a review by the Swedish regulatory authorities of the post-closure safety assessment of the SFR 1 disposal facility for low and intermediate waste (L/ILW), SAFE, the SKB has prepared an updated assessment called SAR-08. This report describes the radionuclide release calculations that have been undertaken as part of SAR-08. The information, assumptions and data used in the calculations are reported and the results are presented. The calculations address issues raised in the regulatory review, but also take account of new information including revised inventory data. The scenarios considered include the main case of expected behaviour of the system, with variants; low probability releases, and so-called residual scenarios. Apart from these scenario uncertainties, data uncertainties have been examined using a probabilistic approach. Calculations have been made using the AMBER software. This allows all the component features of the assessment model to be included in one place. AMBER has been previously used to reproduce results the corresponding calculations in the SAFE assessment. It is also used in demonstration of the IAEA's near surface disposal assessment methodology ISAM and has been subject to very substantial verification tests and has been used in verifying other assessment codes. Results are presented as a function of time for the release of radionuclides from the near field, and then from the far field into the biosphere. Radiological impacts of the releases are reported elsewhere. Consideration is given to each radionuclide and to each component part of the repository. The releases from the entire repository are also presented. The peak releases rates are, for most scenarios, due to organic C-14. Other radionuclides which contribute to peak release rates include inorganic C-14, Ni-59 and Ni-63. (author)
Radionuclide release calculations for SAR-08
Energy Technology Data Exchange (ETDEWEB)
Thomson, Gavin; Miller, Alex; Smith, Graham; Jackson, Duncan (Enviros Consulting Ltd, Wolverhampton (United Kingdom))
2008-04-15
Following a review by the Swedish regulatory authorities of the post-closure safety assessment of the SFR 1 disposal facility for low and intermediate waste (L/ILW), SAFE, the SKB has prepared an updated assessment called SAR-08. This report describes the radionuclide release calculations that have been undertaken as part of SAR-08. The information, assumptions and data used in the calculations are reported and the results are presented. The calculations address issues raised in the regulatory review, but also take account of new information including revised inventory data. The scenarios considered include the main case of expected behaviour of the system, with variants; low probability releases, and so-called residual scenarios. Apart from these scenario uncertainties, data uncertainties have been examined using a probabilistic approach. Calculations have been made using the AMBER software. This allows all the component features of the assessment model to be included in one place. AMBER has been previously used to reproduce results the corresponding calculations in the SAFE assessment. It is also used in demonstration of the IAEA's near surface disposal assessment methodology ISAM and has been subject to very substantial verification tests and has been used in verifying other assessment codes. Results are presented as a function of time for the release of radionuclides from the near field, and then from the far field into the biosphere. Radiological impacts of the releases are reported elsewhere. Consideration is given to each radionuclide and to each component part of the repository. The releases from the entire repository are also presented. The peak releases rates are, for most scenarios, due to organic C-14. Other radionuclides which contribute to peak release rates include inorganic C-14, Ni-59 and Ni-63. (author)
Willow growing - Methods of calculation and profitability
International Nuclear Information System (INIS)
Rosenqvist, H.
1997-01-01
The calculation method presented here makes it possible to conduct profitability comparisons between annual and perennial crops and in addition take the planning situation into account. The method applied is a modified total step calculation. The difference between a traditional total step calculation and the modified version is the way in which payments and disbursements are taken into account over a period of several years. This is achieved by combining the present value method and the annuity method. The choice of interest rate has great bearing on the result in perennial calculations. The various components influencing the interest rate are analysed and factors relating to the establishment of the interest rate in different situations are described. The risk factor can be an important variable component of the interest rate calculation. Risk is also addressed from an approach in accordance with portfolio theory. The application of the methods sheds light on the profitability of Salix cultivation from the viewpoint of business economics, and also how different factors influence the profitability of Salix cultivation. Aspects studied are harvesting intervals, the importance of yield level, the competitiveness of Salix versus grain cultivation, the influence of income taxes on profitability etc. Methods for evaluation of activities concerning cultivation of a perennial crop are described and also involve the application of nitrogen fertilization to Salix cultivation. Studies have been performed using these methods to look into nitrogen fertilizer profitability in Salix cultivation during the first rotation period. Nitrogen fertilizer profitability has been investigated involving both production functions and cost calculations, taking the year fertilization into consideration. 72 refs., 2 figs., 52 tabs
Numerical challenges of short range wake field calculations
Energy Technology Data Exchange (ETDEWEB)
Lau, Thomas; Gjonaj, Erion; Weiland, Thomas [Technische Universitaet Darmstadt (Germany). Institut fuer Theorie Elektromagnetischer Felder (TEMF)
2011-07-01
For present and future accelerator projects with ultra short bunches the accurate and reliable calculation of short range wake fields is an important issue. However, the numerical calculation of short range wake fields is a numerical challenging task. The presentation gives an overview over the numerical challenges and techniques for short range wake field calculations. Finally, some simulation results obtained by the program PBCI developed at the TU Darmstadt are presented.
Development of Monte Carlo decay gamma-ray transport calculation system
Energy Technology Data Exchange (ETDEWEB)
Sato, Satoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Kawasaki, Nobuo [Fujitsu Ltd., Tokyo (Japan); Kume, Etsuo [Japan Atomic Energy Research Inst., Center for Promotion of Computational Science and Engineering, Tokai, Ibaraki (Japan)
2001-06-01
In the DT fusion reactor, it is critical concern to evaluate the decay gamma-ray biological dose rates after the reactor shutdown exactly. In order to evaluate the decay gamma-ray biological dose rates exactly, three dimensional Monte Carlo decay gamma-ray transport calculation system have been developed by connecting the three dimensional Monte Carlo particle transport calculation code and the induced activity calculation code. The developed calculation system consists of the following four functions. (1) The operational neutron flux distribution is calculated by the three dimensional Monte Carlo particle transport calculation code. (2) The induced activities are calculated by the induced activity calculation code. (3) The decay gamma-ray source distribution is obtained from the induced activities. (4) The decay gamma-rays are generated by using the decay gamma-ray source distribution, and the decay gamma-ray transport calculation is conducted by the three dimensional Monte Carlo particle transport calculation code. In order to reduce the calculation time drastically, a biasing system for the decay gamma-ray source distribution has been developed, and the function is also included in the present system. In this paper, the outline and the detail of the system, and the execution example are reported. The evaluation for the effect of the biasing system is also reported. (author)
The development and validation of control rod calculation methods
International Nuclear Information System (INIS)
Rowlands, J.L.; Sweet, D.W.; Franklin, B.M.
1979-01-01
Fission rate distributions have been measured in the zero power critical facility, ZEBRA, for a series of eight different arrays of boron carbide control rods. Diffusion theory calculations have been compared with these measurements. The normalised fission rates differ by up to about 30% in some regions, between the different arrays, and these differences are well predicted by the calculations. A development has been made to a method used to produce homogenised cross sections for lattice regions containing control rods. Calculations show that the method also reproduces the reaction rate within the rod and the fission rate dip at the surface of the rod in satisfactory agreement with the more accurate calculations which represent the fine structure of the rod. A comparison between diffusion theory and transport theory calculations of control rod reactivity worths in the CDFR shows that for the standard design method the finite mesh approximation and the difference between diffusion theory and transport theory (the transport correction) tend to cancel and result in corrections to be applied to the standard mesh diffusion theory calculations of about +- 2% or less. This result applies for mesh centred finite difference diffusion theory codes and for the arrays of natural boron carbide control rods for which the calculations were made. Improvements have also been made to the effective diffusion coefficients used in diffusion theory calculations for control rod followers and these give satisfactory agreement with transport theory calculations. (U.K.)
Phase-shift calculation using continuum-discretized states
International Nuclear Information System (INIS)
Suzuki, Y.; Horiuchi, W.; Arai, K.
2009-01-01
We present a method for calculating scattering phase shifts which utilizes continuum-discretized states obtained in a bound-state type calculation. The wrong asymptotic behavior of the discretized state is remedied by means of the Green's function formalism. Test examples confirm the accuracy of the method. The α+n scattering is described using realistic nucleon-nucleon potentials. The 3/2 - and 1/2 - phase shifts obtained in a single-channel calculation are too small in comparison with experiment. The 1/2 + phase shifts are in reasonable agreement with experiment, and gain contributions both from the tensor and central components of the nucleon-nucleon potential.
Calculation and measurement of fog droplet size
International Nuclear Information System (INIS)
Laali, A.R.; Courant, J.J.; Kleitz, A.
1991-01-01
This paper outlines the elements involved in calculation and measurement of fog droplet size in steam turbines. The condensation calculations are performed for a 600 MW LP fossil fired, and for a 900 MW LP nuclear turbine. A simplified method based on classical condensation theory is used for these calculations. The fog droplet size measurement are carried out downstream of the last moving blades of these turbines in order to validate the program. The comparison between the results could lead to a better understanding of the condensation process in steam turbines. Some large droplet (re-entrained droplet) measurements are also taken using a microvideo probe
Hypervelocity impact cratering calculations
Maxwell, D. E.; Moises, H.
1971-01-01
A summary is presented of prediction calculations on the mechanisms involved in hypervelocity impact cratering and response of earth media. Considered are: (1) a one-gram lithium-magnesium alloys impacting basalt normally at 6.4 km/sec, and (2) a large terrestrial impact corresponding to that of Sierra Madera.
Preliminary isodose calculation for gynecological curietherapy
International Nuclear Information System (INIS)
Bridier, A.; Dutreix, A.; Gerbaulet, A.; Chassagne, D.
1981-01-01
We present a preliminary method of calculating the dimensions of the reference isodose, based upon the geometrical distribution and length of the sources used, their linear activity and the length of treatment, that does not require use of a computer. Inversely, this method can be used to determine the factors necessary to produce a given shape of isodose, and also to predict the change in shape of the isodose that will be produced by altering the various factors. This method was derived from a systematic computer study of dose distribution in which each factor was varied independently of all others. The dimensions of the isodoses, calculated by this method, were found to be in agreement with those derived from computer calculation to within an error of about 2 mm. The method is only applicable for a limited range of positions of the vaginal sources. The influence of the positions of these sources along the line of the axis of uterine catheter and of their inclination to this line, are currently being studied. The results are presented as mathematical formulae relating each dimension of the isodose curves to the features of the application, but could equally well be expressed in tabular form that would be more convenient for everyday use. An example of the calculation used is given to facilitate understanding of the method [fr
42 CFR 403.253 - Calculation of benefits.
2010-10-01
... the gross premiums are computed to provide coverage. (iv) Reserve for future contingent benefits means... the loss ratio calculation period. (iii) Net premium means the level portion of the gross premium used... period, to— (B) The total policy reserve at the last day of the loss ratio calculation period: and (ii...
GPU based acceleration of first principles calculation
International Nuclear Information System (INIS)
Tomono, H; Tsumuraya, K; Aoki, M; Iitaka, T
2010-01-01
We present a Graphics Processing Unit (GPU) accelerated simulations of first principles electronic structure calculations. The FFT, which is the most time-consuming part, is about 10 times accelerated. As the result, the total computation time of a first principles calculation is reduced to 15 percent of that of the CPU.
Molecular dynamics calculation of shear viscosity for molten salt
International Nuclear Information System (INIS)
Okamoto, Yoshihiro; Yokokawa, Mitsuo; Ogawa, Toru
1993-12-01
A computer program of molecular dynamics simulation has been made to calculate shear viscosity of molten salt. Correlation function for an off-diagonal component of stress tensor can be obtained as the results of calculation. Shear viscosity is calculated by integration of the correlation function based on the Kubo-type formula. Shear viscosities for a molten KCl ranging in temperature from 1047K to 1273K were calculated using the program. Calculation of 10 5 steps (1 step corresponds to 5 x 10 -15 s) was performed for each temperature in the 216 ions system. The obtained results were in good agreement with the reported experimental values. The program has been vectorized to achieve a faster computation in supercomputer. It makes possible to calculate the viscosity using a large number of statistics amounting to several million MD steps. (author)
TRIGA fuel element burnup determination by measurement and calculation
International Nuclear Information System (INIS)
Zagar, T.; Ravnik, M.; Persic, A.; Jeraj, R.
2000-01-01
To estimate the accuracy of the fuel element burnup calculation different factors influencing the calculation were studied. To cover different aspects of burnup calculations, two in-house developed computer codes were used in calculations. The first (TRIGAP) is based on a one-dimensional two-group diffusion approximation, and the second (TRIGLAV) is based on a two-dimensional four-group diffusion equation. Both codes use WIMSD program with different libraries forunit-cell cross section data calculation. The burnup accumulated during the operating history of the TRIGA reactor at Josef Stefan Institute was calculated for all fuel elements. Elements used in the core during this period were standard SS 8.5% fuel elements, standard SS 12% fuel elements and highly enriched FLIP fuel elements. During the considerable period of operational history, FLIP and standard fuel elements were used simultaneously in mixed cores. (authors)
Monte Carlo perturbation theory in neutron transport calculations
International Nuclear Information System (INIS)
Hall, M.C.G.
1980-01-01
The need to obtain sensitivities in complicated geometrical configurations has resulted in the development of Monte Carlo sensitivity estimation. A new method has been developed to calculate energy-dependent sensitivities of any number of responses in a single Monte Carlo calculation with a very small time penalty. This estimation typically increases the tracking time per source particle by about 30%. The method of estimation is explained. Sensitivities obtained are compared with those calculated by discrete ordinates methods. Further theoretical developments, such as second-order perturbation theory and application to k/sub eff/ calculations, are discussed. The application of the method to uncertainty analysis and to the analysis of benchmark experiments is illustrated. 5 figures
Linear filtering applied to Monte Carlo criticality calculations
International Nuclear Information System (INIS)
Morrison, G.W.; Pike, D.H.; Petrie, L.M.
1975-01-01
A significant improvement in the acceleration of the convergence of the eigenvalue computed by Monte Carlo techniques has been developed by applying linear filtering theory to Monte Carlo calculations for multiplying systems. A Kalman filter was applied to a KENO Monte Carlo calculation of an experimental critical system consisting of eight interacting units of fissile material. A comparison of the filter estimate and the Monte Carlo realization was made. The Kalman filter converged in five iterations to 0.9977. After 95 iterations, the average k-eff from the Monte Carlo calculation was 0.9981. This demonstrates that the Kalman filter has the potential of reducing the calculational effort of multiplying systems. Other examples and results are discussed
Cohesion energy calculations for ternary ionic novel crystals
International Nuclear Information System (INIS)
Vazquez P, G.; Cabrera, E.; Mijangos, R.R.; Valdez, E.; Duarte, C.
2001-01-01
The present work calculates the value of the link energy of a crystalline ternary structure newly formed by alkali halides. The ternary structure prepared with different concentrations of KCl x KBrRbCl 2 maintains a very good miscibility and stability. The calculation is based on the use of a generalization of the Vegard law (which generally is valid for binary compounds) for calculating the values of the lattice constant and the repulsive m exponent. The value of the lattice parameter given by X-ray diffractometry agrees with the close approximation of the calculated value of the method used. It also compares the value of energy cohesion obtained by the Born expression with more complex approximations. (Author)
THE ACCOUNTING POSTEMPLOYMENT BENEFITS BASED ON ACTUARIAL CALCULATIONS
Directory of Open Access Journals (Sweden)
Anna CEBOTARI
2017-11-01
Full Text Available The accounting post-employment benefits, based on actuarial calculations, at present remains a subject studied in Moldova only theoretically. Applying actuarial calculations of accounting in fact denotes its character of evolving. Because national accounting standards have been adapted to international, which, in turn, require the valuation of assets and debts at fair value, there is a need to draw up exact calculations on which stands the theory of probability and mathematical statistics. One of the main objectives of accounting information is reflected in its financial situations and providing internal and external users of the entity. Hence, arises the need to reflect highly reliable information that can be provided by applying actuarial calculations.
Parallel plasma fluid turbulence calculations
International Nuclear Information System (INIS)
Leboeuf, J.N.; Carreras, B.A.; Charlton, L.A.; Drake, J.B.; Lynch, V.E.; Newman, D.E.; Sidikman, K.L.; Spong, D.A.
1994-01-01
The study of plasma turbulence and transport is a complex problem of critical importance for fusion-relevant plasmas. To this day, the fluid treatment of plasma dynamics is the best approach to realistic physics at the high resolution required for certain experimentally relevant calculations. Core and edge turbulence in a magnetic fusion device have been modeled using state-of-the-art, nonlinear, three-dimensional, initial-value fluid and gyrofluid codes. Parallel implementation of these models on diverse platforms--vector parallel (National Energy Research Supercomputer Center's CRAY Y-MP C90), massively parallel (Intel Paragon XP/S 35), and serial parallel (clusters of high-performance workstations using the Parallel Virtual Machine protocol)--offers a variety of paths to high resolution and significant improvements in real-time efficiency, each with its own advantages. The largest and most efficient calculations have been performed at the 200 Mword memory limit on the C90 in dedicated mode, where an overlap of 12 to 13 out of a maximum of 16 processors has been achieved with a gyrofluid model of core fluctuations. The richness of the physics captured by these calculations is commensurate with the increased resolution and efficiency and is limited only by the ingenuity brought to the analysis of the massive amounts of data generated
CONDOR: neutronic code for fuel elements calculation with rods
International Nuclear Information System (INIS)
Villarino, E.A.
1990-01-01
CONDOR neutronic code is used for the calculation of fuel elements formed by fuel rods. The method employed to obtain the neutronic flux is that of collision probabilities in a multigroup scheme on two-dimensional geometry. This code utilizes new calculation algorithms and normalization of such collision probabilities. Burn-up calculations can be made before the alternative of applying variational methods for response flux calculations or those corresponding to collision normalization. (Author) [es
Spallation reactions: calculations
International Nuclear Information System (INIS)
Bertini, H.W.
1975-01-01
Current methods for calculating spallation reactions over various energy ranges are described and evaluated. Recent semiempirical fits to existing data will probably yield the most accurate predictions for these reactions in general. However, if the products in question have binding energies appreciably different from their isotropic neighbors and if the cross section is approximately 30 mb or larger, then the intranuclear-cascade-evaporation approach is probably better suited. (6 tables, 12 figures, 34 references) (U.S.)
Graphical comparison of calculated internal conversion coefficients
International Nuclear Information System (INIS)
Ewbank, W.B.
1980-11-01
Calculated values of the coefficients of internal conversion of gamma rays in the K shell and L 1 , L 2 , L 3 subshells from published tabulations by Band and Trzhaskovskaya and by Roesel et al. at Data Nucl. Data Tables, 21, 92-514(1978) are compared with values obtained by computer interpolation among tabulated values of Hager and Seltzer Nucl. Data, A4, 1-235(1968). In some cases, agreement among the three calculations is remarkably good, and differences are generally less than 5%. In a few cases, there are differences as large as 20 to 50%, corresponding to the threshold effect described by Roesel et al. The Z-dependent resonance minimum described by Roesel et al. is also observed in the comparison of E1-E4 conversion in the L 1 subshell. In several cases (notably M1-M4 conversion in the K shell and L 1 subshell), the Band and Roesel calculations show dramatically different dependence on gamma energy and atomic number. For Z = 100, the Band calculation for E4 conversion in the L 3 subshell shows irregular behavior at energies below the K-shell binding energy. A few high-quality measurements of internal conversion coefficients (+-5%) would help greatly to establish a basis for choice among the theoretical calculations. 32 figures
Georgia fishery study: implications for dose calculations
International Nuclear Information System (INIS)
Turcotte, M.D.S.
1983-01-01
Fish consumption will contribute a major portion of the estimated individual and population doses from L-Reactor liquid releases and Cs-137 remobilization in Steel Creek. It is therefore important that the values for fish consumption used in dose calculations be as realistic as possible. Since publication of the L-Reactor Environmental Information Document (EID), data have become available on sport fishing in the Savannah River. These data provide SRP with site-specific sport fish harvest and consumption values for use in dose calculations. The Georgia fishery data support the total population fish consumption and calculated dose reported in the EID. The data indicate, however, that both the EID average and maximum individual fish consumption have been underestimated, although each to a different degree. The average fish consumption value used in the EID is approximately 3% below the lower limit of the fish consumption range calculated using the Georgia data. A fish consumption value of 11.3 kg/yr should be used to recalculate dose to the average individual from L-Reactor restart. Maximum fish consumption in the EID has been underestimated by approximately 60%, and doses to the maximum individual should also be recalculated. Future dose calculations should utilize an average fish consumption value of 11.3 kg/yr, and a maximum fish consumption value of 34 kg/yr
A quantitative calculation for software reliability evaluation
Energy Technology Data Exchange (ETDEWEB)
Lee, Young-Jun; Lee, Jang-Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2016-10-15
To meet these regulatory requirements, the software used in the nuclear safety field has been ensured through the development, validation, safety analysis, and quality assurance activities throughout the entire process life cycle from the planning phase to the installation phase. A variety of activities, such as the quality assurance activities are also required to improve the quality of a software. However, there are limitations to ensure that the quality is improved enough. Therefore, the effort to calculate the reliability of the software continues for a quantitative evaluation instead of a qualitative evaluation. In this paper, we propose a quantitative calculation method for the software to be used for a specific operation of the digital controller in an NPP. After injecting random faults in the internal space of a developed controller and calculating the ability to detect the injected faults using diagnostic software, we can evaluate the software reliability of a digital controller in an NPP. We tried to calculate the software reliability of the controller in an NPP using a new method that differs from a traditional method. It calculates the fault detection coverage after injecting the faults into the software memory space rather than the activity through the life cycle process. We attempt differentiation by creating a new definition of the fault, imitating the software fault using the hardware, and giving a consideration and weights for injection faults.
Computer Package for Graphite Total Cross-Section Calculations
International Nuclear Information System (INIS)
Adib, M.; Fathalla, M.
2008-01-01
An additive formula is given which allows calculating the contribution of the total neut.>neutron transmission through crystalline graphite. The formula takes into account the graphite form of poly or pyrolytic crystals and its parameters. Computer package Graphite has been designed in order to provide the required calculations in the neutron energy range from 0.1 MeV to 10 eV. The package includes three codes: PCG (Polycrystalline Graphite), PG (Pyrolytic Graphite) and HOPG (Highly Oriented Pyrolytic Graphite) for calculating neutron transmission through fine graphite powder (polycrystalline), neutron transmission and removal coefficient of PG crystal in terms of its mosaic spread for neutrons incident along its c-axis and the transmission of neutrons incident on HOPG crystal at different angles, respectively. For comparison of the experimental neutron transmission data with the calculated values, the program takes into consideration the effect of both wavelength and neutron beam divergence in either 2 constant wavelength spread mode (δλ=constant) or constant wavelength resolution mode (δλ/λ=constant). In order to check the validity for application of computer package Graphite in cross-section calculations, a comparison between calculated values with the available experimental data were carried out. An overall agreement is indicated with an accuracy sufficient for determine the neutron transmission characteristics
Calculations of the ground state of 16O
International Nuclear Information System (INIS)
Pieper, S.C.
1989-01-01
One of the central problems in nuclear physics is the description of nuclei as systems of nucleons interacting via realistic potentials. There are two main aspects of this problem: specification of the Hamiltonian, and calculation of the ground states of nuclei with the given interaction. Realistic interactions must contain both two- and three-nucleon potentials and these potentials have a complicated non-central operator structure consisting, for example, of spin, isospin and tensor dependences. This structure results in formidable many-body problems in the computation of the ground states of nuclei. At present, reliable solutions of the Faddeev equations for the A = 3 nuclei with such interactions are routine. Recently, Carlson has made an essentially exact GFMC calculation of the He ground state using just a two-nucleon interaction, and there are reliable variational calculations for more complete potential models. Nuclear matter calculations can also be made with reasonable reliability. However, there have been very few calculations of nuclei with A > 5 using realistic interactions, and none with a modern three-nucleon interaction. In the present paper I present a new technique for variational calculations for such nuclei and apply it to the ground state of 16 O. 15 refs., 2 figs., 3 tabs
Three-dimensional space-charge calculation method
International Nuclear Information System (INIS)
Lysenko, W.P.; Wadlinger, E.A.
1980-09-01
A method is presented for calculating space-charge forces on individual particles in a particle tracing simulation code. Poisson's equation is solved in three dimensions with boundary conditions specified on an arbitrary surface. When the boundary condition is defined by an impressed radio-frequency field, the external electric fields as well as the space-charge fields are determined. A least squares fitting procedure is used to calculate the coefficients of expansion functions, which need not be orthogonal nor individually satisfy the boundary condition
Nuclear friction calculated from nucleon currents
International Nuclear Information System (INIS)
Pi, M.; Vinas, X.; Barranco, M.; La Rana, G.; Leray, S.; Lucas, R.; Ngo, C.; Tomasi, E.
1984-01-01
Nuclear friction can be connected to the number of nucleons exchanged between two interacting nuclei. The proximity scaling allows to reduce this problem to a calculation of the nucleon current between two semi infinite slabs of nuclear matter facing each other. In this paper we review the approximations and the results concerning this problem with a special emphasis on the physical ideas. Applications of nucleons currents to Fermi jets and to the calculation of a part of the imaginary potential are also discussed
Lutnaes, Ola B; Teale, Andrew M; Helgaker, Trygve; Tozer, David J; Ruud, Kenneth; Gauss, Jürgen
2009-10-14
An accurate set of benchmark rotational g tensors and magnetizabilities are calculated using coupled-cluster singles-doubles (CCSD) theory and coupled-cluster single-doubles-perturbative-triples [CCSD(T)] theory, in a variety of basis sets consisting of (rotational) London atomic orbitals. The accuracy of the results obtained is established for the rotational g tensors by careful comparison with experimental data, taking into account zero-point vibrational corrections. After an analysis of the basis sets employed, extrapolation techniques are used to provide estimates of the basis-set-limit quantities, thereby establishing an accurate benchmark data set. The utility of the data set is demonstrated by examining a wide variety of density functionals for the calculation of these properties. None of the density-functional methods are competitive with the CCSD or CCSD(T) methods. The need for a careful consideration of vibrational effects is clearly illustrated. Finally, the pure coupled-cluster results are compared with the results of density-functional calculations constrained to give the same electronic density. The importance of current dependence in exchange-correlation functionals is discussed in light of this comparison.
Danish Sector Guide for Calculation of the Actual Energy Consumption
DEFF Research Database (Denmark)
Mortensen, Lone Hedegaard
2016-01-01
, the innovation network for sustainable construction, InnoBYG started work on a Danish sector guide for the calculation of actual energy consumption in relation to upgrading of buildings. The focus was to make a common guide for energy calculations that can be used by consultants performing calculations...... consumption compared with the estimated energy demand by calculation. The paper concludes that the result of an energy calculation should not be given as a single figure but rather as a spread between the best and worst case for the assumed conditions. Finally, a brief update on current actions is given...... related to the sector guide for calculation of actual energy consumption. Keywords – Energy calculations, actual energy consumption, energy perfomance...
Zero Temperature Hope Calculations
International Nuclear Information System (INIS)
Rozsnyai, B. F.
2002-01-01
The primary purpose of the HOPE code is to calculate opacities over a wide temperature and density range. It can also produce equation of state (EOS) data. Since the experimental data at the high temperature region are scarce, comparisons of predictions with the ample zero temperature data provide a valuable physics check of the code. In this report we show a selected few examples across the periodic table. Below we give a brief general information about the physics of the HOPE code. The HOPE code is an ''average atom'' (AA) Dirac-Slater self-consistent code. The AA label in the case of finite temperature means that the one-electron levels are populated according to the Fermi statistics, at zero temperature it means that the ''aufbau'' principle works, i.e. no a priory electronic configuration is set, although it can be done. As such, it is a one-particle model (any Hartree-Fock model is a one particle model). The code is an ''ion-sphere'' model, meaning that the atom under investigation is neutral within the ion-sphere radius. Furthermore, the boundary conditions for the bound states are also set at the ion-sphere radius, which distinguishes the code from the INFERNO, OPAL and STA codes. Once the self-consistent AA state is obtained, the code proceeds to generate many-electron configurations and proceeds to calculate photoabsorption in the ''detailed configuration accounting'' (DCA) scheme. However, this last feature is meaningless at zero temperature. There is one important feature in the HOPE code which should be noted; any self-consistent model is self-consistent in the space of the occupied orbitals. The unoccupied orbitals, where electrons are lifted via photoexcitation, are unphysical. The rigorous way to deal with that problem is to carry out complete self-consistent calculations both in the initial and final states connecting photoexcitations, an enormous computational task. The Amaldi correction is an attempt to address this problem by distorting the
40 CFR 98.363 - Calculating GHG emissions.
2010-07-01
... combustion device (metric tons CH4/yr). V = Average annual volumetric flow rate, calculated in Equation JJ-7... CH4 flow to digester combustion device, as calculated in Equation JJ-6 of this section (metric tons CH4). DE = CH4 destruction efficiency from flaring or burning in engine (lesser of manufacturer's...
Radiation protection calculations for diagnostic medical equipment
International Nuclear Information System (INIS)
Klueter, R.
1992-01-01
The standards DIN 6812 and DIN 6844 define the radiation protection requirements to be met by biomedical radiography equipment or systems for nuclear medicine. The paper explains the use of a specific computer program for radiation protection calculations. The program offers menu-controlled calculation, with free choice of the relevant nuclides. (DG) [de
Fast, large-scale hologram calculation in wavelet domain
Shimobaba, Tomoyoshi; Matsushima, Kyoji; Takahashi, Takayuki; Nagahama, Yuki; Hasegawa, Satoki; Sano, Marie; Hirayama, Ryuji; Kakue, Takashi; Ito, Tomoyoshi
2018-04-01
We propose a large-scale hologram calculation using WAvelet ShrinkAge-Based superpositIon (WASABI), a wavelet transform-based algorithm. An image-type hologram calculated using the WASABI method is printed on a glass substrate with the resolution of 65 , 536 × 65 , 536 pixels and a pixel pitch of 1 μm. The hologram calculation time amounts to approximately 354 s on a commercial CPU, which is approximately 30 times faster than conventional methods.
Equilibrium fission model calculations
International Nuclear Information System (INIS)
Beckerman, M.; Blann, M.
1976-01-01
In order to aid in understanding the systematics of heavy ion fission and fission-like reactions in terms of the target-projectile system, bombarding energy and angular momentum, fission widths are calculated using an angular momentum dependent extension of the Bohr-Wheeler theory and particle emission widths using angular momentum coupling
21 CFR 868.1890 - Predictive pulmonary-function value calculator.
2010-04-01
... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Predictive pulmonary-function value calculator... SERVICES (CONTINUED) MEDICAL DEVICES ANESTHESIOLOGY DEVICES Diagnostic Devices § 868.1890 Predictive pulmonary-function value calculator. (a) Identification. A predictive pulmonary-function value calculator is...
Environment-based pin-power reconstruction method for homogeneous core calculations
International Nuclear Information System (INIS)
Leroyer, H.; Brosselard, C.; Girardi, E.
2012-01-01
Core calculation schemes are usually based on a classical two-step approach associated with assembly and core calculations. During the first step, infinite lattice assemblies calculations relying on a fundamental mode approach are used to generate cross-sections libraries for PWRs core calculations. This fundamental mode hypothesis may be questioned when dealing with loading patterns involving several types of assemblies (UOX, MOX), burnable poisons, control rods and burn-up gradients. This paper proposes a calculation method able to take into account the heterogeneous environment of the assemblies when using homogeneous core calculations and an appropriate pin-power reconstruction. This methodology is applied to MOX assemblies, computed within an environment of UOX assemblies. The new environment-based pin-power reconstruction is then used on various clusters of 3x3 assemblies showing burn-up gradients and UOX/MOX interfaces, and compared to reference calculations performed with APOLLO-2. The results show that UOX/MOX interfaces are much better calculated with the environment-based calculation scheme when compared to the usual pin-power reconstruction method. The power peak is always better located and calculated with the environment-based pin-power reconstruction method on every cluster configuration studied. This study shows that taking into account the environment in transport calculations can significantly improve the pin-power reconstruction so far as it is consistent with the core loading pattern. (authors)
Calculation method for gamma dose rates from Gaussian puffs
Energy Technology Data Exchange (ETDEWEB)
Thykier-Nielsen, S; Deme, S; Lang, E
1995-06-01
The Lagrangian puff models are widely used for calculation of the dispersion of releases to the atmosphere. Basic output from such models is concentration of material in the air and on the ground. The most simple method for calculation of the gamma dose from the concentration of airborne activity is based on the semi-infinite cloud model. This method is however only applicable for puffs with large dispersion parameters, i.e. for receptors far away from the release point. The exact calculation of the cloud dose using volume integral requires large computer time usually exceeding what is available for real time calculations. The volume integral for gamma doses could be approximated by using the semi-infinite cloud model combined with correction factors. This type of calculation procedure is very fast, but usually the accuracy is poor because only a few of the relevant parameters are considered. A multi-parameter method for calculation of gamma doses is described here. This method uses precalculated values of the gamma dose rates as a function of E{sub {gamma}}, {sigma}{sub y}, the asymmetry factor - {sigma}{sub y}/{sigma}{sub z}, the height of puff center - H and the distance from puff center R{sub xy}. To accelerate the calculations the release energy, for each significant radionuclide in each energy group, has been calculated and tabulated. Based on the precalculated values and suitable interpolation procedure the calculation of gamma doses needs only short computing time and it is almost independent of the number of radionuclides considered. (au) 2 tabs., 15 ills., 12 refs.
Calculation method for gamma dose rates from Gaussian puffs
International Nuclear Information System (INIS)
Thykier-Nielsen, S.; Deme, S.; Lang, E.
1995-06-01
The Lagrangian puff models are widely used for calculation of the dispersion of releases to the atmosphere. Basic output from such models is concentration of material in the air and on the ground. The most simple method for calculation of the gamma dose from the concentration of airborne activity is based on the semi-infinite cloud model. This method is however only applicable for puffs with large dispersion parameters, i.e. for receptors far away from the release point. The exact calculation of the cloud dose using volume integral requires large computer time usually exceeding what is available for real time calculations. The volume integral for gamma doses could be approximated by using the semi-infinite cloud model combined with correction factors. This type of calculation procedure is very fast, but usually the accuracy is poor because only a few of the relevant parameters are considered. A multi-parameter method for calculation of gamma doses is described here. This method uses precalculated values of the gamma dose rates as a function of E γ , σ y , the asymmetry factor - σ y /σ z , the height of puff center - H and the distance from puff center R xy . To accelerate the calculations the release energy, for each significant radionuclide in each energy group, has been calculated and tabulated. Based on the precalculated values and suitable interpolation procedure the calculation of gamma doses needs only short computing time and it is almost independent of the number of radionuclides considered. (au) 2 tabs., 15 ills., 12 refs
Monte Carlo neutron and gamma-ray calculations
International Nuclear Information System (INIS)
Mendelsohn, Edgar
1987-01-01
Kerma in tissue and the activation produced in sulfur and cobalt due to prompt neutrons from the Hiroshima and Nagasaki bombs were calculated out to 2000 m from the hypocenter in 100 m increments. As neutron sources weapon output spectra calculated by investigators from the Los Alamos National Laboratory (LANL) were used. Other parameters, such as burst height and air and ground densities and compositions, were obtained from recent sources. The LLNL Monte Carlo transport code TART was used for these calculations. TART accesses the well-established 1985 ENDL cross-section library, which has built-in reaction cross sections. The zoning for this problem was a full two-dimensional geometry with a ceiling height of 1100 m and a ground thickness of 30 cm. For the Hiroshima calculations (including sulfur activation) and untilted source was used. However, a special sulfur activation problem using a source tilted 15 deg was run for which the ratios to the untilted case are reported. The TART code uses a technique for solving the transport equation that is different from that of the ORNL DOT code; it also draws on a specially evaluated cross-section library (ENDL) and uses a larger group structure than DOT. One of the purposes of this work was to instill confidence in the DOT calculations that will be used directly in the dose reassessment of A-bomb survivors. The TART results were compared with values calculated with the DOT code by investigators from ORNL and found to be in good agreement for the most part. However, the sulfur activation comparison is disappointing. Because the sulfur activation is caused by higher energy neutrons (which should have experienced fewer collisions than those causing cobalt activation, for example), better agreement than what is reported here would be expected
The Monte Carlo applied for calculation dose
International Nuclear Information System (INIS)
Peixoto, J.E.
1988-01-01
The Monte Carlo method is showed for the calculation of absorbed dose. The trajectory of the photon is traced simulating sucessive interaction between the photon and the substance that consist the human body simulator. The energy deposition in each interaction of the simulator organ or tissue per photon is also calculated. (C.G.C.) [pt
Calculated Atomic Volumes of the Actinide Metals
DEFF Research Database (Denmark)
Skriver, H.; Andersen, O. K.; Johansson, B.
1979-01-01
The equilibrium atomic volume is calculated for the actinide metals. It is possible to account for the localization of the 5f electrons taking place in americium.......The equilibrium atomic volume is calculated for the actinide metals. It is possible to account for the localization of the 5f electrons taking place in americium....
Simplified shielding calculation system for high-intensity proton accelerators
Energy Technology Data Exchange (ETDEWEB)
Masumura, Tomomi; Nakashima, Hiroshi; Nakane, Yoshihiro; Sasamoto, Nobuo [Center for Neutron Science, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan)
2000-06-01
A simplified shielding calculation system is developed for applying conceptual shielding design of facilities in the joint project for high-intensity proton accelerators. The system is composed of neutron transmission calculation part for bulk shielding using simplified formulas: Moyer model and Tesch's formula, and neutron skyshine calculation part using an empirical formula: Stapleton's formula. The system is made with the Microsoft Excel software for user's convenience. This report provides a manual for the system as well as calculation conditions used in the calculation such as Moyer model's parameters. In this report preliminary results based on data at December 8, 1999, are also shown as an example. (author)
Standard Guide for Benchmark Testing of Light Water Reactor Calculations
American Society for Testing and Materials. Philadelphia
2010-01-01
1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactor fluences with ...
The calculation of CSF spaces in CT
International Nuclear Information System (INIS)
Hacker, H.; Artmann, H.
1978-01-01
Objective digital determination of CSF spaces is discussed, with ventricular and subarachnoid spaces handled separately. This method avoids the difficulty of visual definition of ventricular borders in planimetric measurements. The principle is to count automatically all pixels corresponding to CSF in a given region with a Hounsfield unit and to multiply this number by the pixel size. This will give the total surface area of CSF spaces in square millimeters. The calculation of pixel values for CSF spaces and brain tissue is experimentally formulated taking the intersection of the Gaussian curves for ventricular content and brain tissue. In practice, the determination of CSF spaces is done by first calculating a histogram of the total brain in a given slice defining all CSF spaces. Next a histogram of a region including ventricles with adjoining tissue is calculated and the ventricular size is calculated. By subtraction of the ventricle value from the total CSF space value, the subarachnoid space size is obtained. The advantages of this mehtod will be discussed. (orig.) [de
CALCULATION: PRECIPITATION CHARACTERISITICS FOR STORM WATER MANAGEMENT
International Nuclear Information System (INIS)
D. Ambos
2000-01-01
This Calculation is intended to satisfy engineering requirements for maximum 60-minute precipitation amounts for 50 and 100-year return periods at and near Yucca Mountain. This data requirement is documented in the ''Interface Control Document for Support Operations to Surface Facilities Operations Functional and Organizational Interfaces'' (CRWMS M and O 1998a). These developed data will supplement the information on 0.1 hour to 6-hour (in 0.1-hour increments) probable maximum precipitation (PMP) presented in the report, ''Precipitation Design Criteria for Storm Water Management'' (CRWMS M and O 1998b). The Reference Information Base (RIB) item, Precipitation ''Characteristics for Storm Water Management'' (M09902RIB00045 .OOO), was developed based on CRWMS M and O (1998b) and will be supplemented (via revision) with the information developed in this Calculation. The ''Development Plan for the Calculation: Precipitation Characteristics for Storm Water Management'' (CRWMS M and O 2000) was prepared in accordance with AP-2.l3Q, ''Technical Product Development Planning''. This calculation was developed in accordance with AP-3.12Q, Rev. O/ICN 2
76 FR 39242 - Federal Acquisition Regulation; TINA Interest Calculations
2011-07-05
... pricing data. This rule replaces the term ``simple interest'' as the requirement for calculating interest...-AL73 Federal Acquisition Regulation; TINA Interest Calculations AGENCIES: Department of Defense (DoD... interest calculations be applied to Government overpayments as a result of defective cost or pricing data...
ORIGEN2 calculations supporting TRIGA irradiated fuel data package
Energy Technology Data Exchange (ETDEWEB)
Schmittroth, F.A.
1996-09-20
ORIGEN2 calculations were performed for TRIGA spent fuel elements from the Hanford Neutron Radiography Facility. The calculations support storage and disposal and results include mass, activity,and decay heat. Comparisons with underwater dose-rate measurements were used to confirm and adjust the calculations.
10 CFR 474.3 - Petroleum-equivalent fuel economy calculation.
2010-01-01
... 10 Energy 3 2010-01-01 2010-01-01 false Petroleum-equivalent fuel economy calculation. 474.3..., DEVELOPMENT, AND DEMONSTRATION PROGRAM; PETROLEUM-EQUIVALENT FUEL ECONOMY CALCULATION § 474.3 Petroleum-equivalent fuel economy calculation. (a) The petroleum-equivalent fuel economy for an electric vehicle is...
Nuclear data and multigroup methods in fast reactor calculations
International Nuclear Information System (INIS)
Gur, Y.
1975-03-01
The work deals with fast reactor multigroup calculations, and the efficient treatment of basic nuclear data, which serves as raw material for the calculations. Its purpose is twofold: to build a computer code system that handles a large, detailed library of basic neutron cross section data, (such as ENDF/B-III) and yields a compact set of multigroup cross sections for reactor calculations; to use the code system for comparative analysis of different libraries, in order to discover basic uncertainties that still exist in the measurement of neutron cross sections, and to determine their influence upon uncertainties in nuclear calculations. A program named NANICK which was written in two versions is presented. The first handles the American basic data library, ENDF/B-III, while the second handles the German basic data library, KEDAK. The mathematical algorithm is identical in both versions, and only the file management is different. This program calculates infinitely diluted multigroup cross sections and scattering matrices. It is complemented by the program NASIF that calculates shielding factors from resonance parameters. Different versions of NASIF were written to handle ENDF/B-III or KEDAK. New methods for evaluating in reactor calculations the long term behavior of the neutron flux as well as its fine structure are described and an efficient calculation of the shielding factors from resonance parameters is offered. (B.G.)
RP-10: commissioning. Reproduction by physical experiences calculation
International Nuclear Information System (INIS)
Higa, Manabu; Madariaga, M.R.
1990-01-01
This work presents the neutronic calculation results, most of which were carried out after such experiences, to verify the calculation methodology developed at the Analysis and Calculation Department of the National Atomic Energy Commission (CNEA). The results obtained were satisfactory, proving that the calculation methodology used is adequate for the design of this type of reactors. The only important disagreement is to evaluate the reactivity excess and cut reactivity, but this responds to a criterion difference and/or that of definition for these parameters. The positions of criticality with errors lower than 100 pcm were predicted. The differential and integral reactivities for the calibration of bars, as well as the flux distribution, are reproduced in a reasonable degree in relation to differences inferior to 10%. (Author) [es
Landfill Gas Energy Benefits Calculator
This page contains the LFG Energy Benefits Calculator to estimate direct, avoided, and total greenhouse gas reductions, as well as environmental and energy benefits, for a landfill gas energy project.
Efficient sound barrier calculations with the BEM
DEFF Research Database (Denmark)
Juhl, Peter Møller; Cutanda Henriquez, Vicente
2018-01-01
The Boundary Element Method has been used for calculating the effect of introducing sound barriers for some decades. The method has also been used for optimizing the shape of the barrier and in some cases the effects of introducing sound absorption. However, numerical calculations are still quite...... time consuming and inconvenient to use, which is limiting their use for many practical problems. Moreover, measurements are mostly taken in one-third or full octave bands opposed to the numerical computations at specific frequencies, which then has to be conducted using a fine density in frequencies....... This paper addresses some of the challenges and possible solutions for developing BEM into a more efficient tool for sound barrier calculations....
Method for consequence calculations for severe accidents
International Nuclear Information System (INIS)
Nielsen, F.
1988-07-01
This report was commissioned by the Swedish State Power Board. The report contains a calculation of radiation doses in the surroundings caused by a theoretical core meltdown accident at Forsmark reactor No 3. The accident sequence chosen for the calculating was a release caused by total power failure. The calculations were made by means of the PLUCON4 code. Meteorological data for two years from the Forsmark meteorological tower were analysed to find representative weather situations. As typical weather, Pasquill D was chosen with a wind speed of 5 m/s, and as extreme weather, Pasquill F with a wind speed of 2 m/s. 23 tabs., 37 ills., 20 refs. (author)
Automated one-loop calculations with GoSam
International Nuclear Information System (INIS)
Cullen, Gavin; Greiner, Nicolas; Heinrich, Gudrun; Reiter, Thomas; Luisoni, Gionata; Mastrolia, Pierpaolo; Ossola, Giovanni; Tramontano, Francesco
2012-01-01
We present the program package GoSam which is designed for the automated calculation of one-loop amplitudes for multi-particle processes in renormalisable quantum field theories. The amplitudes, which are generated in terms of Feynman diagrams, can be reduced using either D-dimensional integrand-level decomposition or tensor reduction. GoSam can be used to calculate one-loop QCD and/or electroweak corrections to Standard Model processes and offers the flexibility to link model files for theories Beyond the Standard Model. A standard interface to programs calculating real radiation is also implemented. We demonstrate the flexibility of the program by presenting examples of processes with up to six external legs attached to the loop. (orig.)
Automated One-Loop Calculations with GoSam
Cullen, Gavin; Heinrich, Gudrun; Luisoni, Gionata; Mastrolia, Pierpaolo; Ossola, Giovanni; Reiter, Thomas; Tramontano, Francesco
2012-01-01
We present the program package GoSam which is designed for the automated calculation of one-loop amplitudes for multi-particle processes in renormalisable quantum field theories. The amplitudes, which are generated in terms of Feynman diagrams, can be reduced using either D-dimensional integrand-level decomposition or tensor reduction. GoSam can be used to calculate one-loop QCD and/or electroweak corrections to Standard Model processes and offers the flexibility to link model files for theories Beyond the Standard Model. A standard interface to programs calculating real radiation is also implemented. We demonstrate the flexibility of the program by presenting examples of processes with up to six external legs attached to the loop.
TDHF calculations for heavy-ion collisions
International Nuclear Information System (INIS)
Dhar, A.K.
1981-01-01
In considering the TDHF theory for heavy-ion reaction calculations it is shown that this parameter-free approach spans a wide range of nuclear phenomena ranging from elastic scattering to fusion, including dissipative and collective processes, in a unified manner. The subject is considered under the headings: (1) TDHF equations, conservation laws, effective hamiltonian and initial conditions. (2) Symmetries and filling approximation. (3) Qualitative features of TDHF dynamics. (4) Comparison with experiment (fusion results, deep-inelastic reaction studies, particle emission from TDHF calculations). (U.K.)
Transmission pipeline calculations and simulations manual
Menon, E Shashi
2014-01-01
Transmission Pipeline Calculations and Simulations Manual is a valuable time- and money-saving tool to quickly pinpoint the essential formulae, equations, and calculations needed for transmission pipeline routing and construction decisions. The manual's three-part treatment starts with gas and petroleum data tables, followed by self-contained chapters concerning applications. Case studies at the end of each chapter provide practical experience for problem solving. Topics in this book include pressure and temperature profile of natural gas pipelines, how to size pipelines for specified f
Newnes circuit calculations pocket book with computer programs
Davies, Thomas J
2013-01-01
Newnes Circuit Calculations Pocket Book: With Computer Programs presents equations, examples, and problems in circuit calculations. The text includes 300 computer programs that help solve the problems presented. The book is comprised of 20 chapters that tackle different aspects of circuit calculation. The coverage of the text includes dc voltage, dc circuits, and network theorems. The book also covers oscillators, phasors, and transformers. The text will be useful to electrical engineers and other professionals whose work involves electronic circuitry.
ECP evaluation by water radiolysis and ECP model calculations
Energy Technology Data Exchange (ETDEWEB)
Hanawa, S.; Nakamura, T.; Uchida, S. [Japan Atomic Energy Agency, Tokai-mura, Ibaraki (Japan); Kus, P.; Vsolak, R.; Kysela, J. [Nuclear Research Inst. Rez plc, Rez (Czech Republic)
2010-07-01
In-pile ECP measurements data was evaluated by water radiolysis calculations. The data was obtained by using an in-pile loop in an experimental reactor, LVR-15, at the Nuclear Research Institute (NRI) in Czech Republic. Three types of ECP sensors, a Pt electrode, an Ag/AgCl sensor and a zirconia membrane sensor containing Ag/Ag{sub 2}O were used at several levels of the irradiation rig at various neutron flux and gamma rates. For water radiolysis calculation, the in-pile loop was modeled to several nodes following their design specifications, operating conditions such as flow rates, dose rate distributions of neutron and gamma-ray and so on. Concentration of chemical species along the water flow was calculated by a radiolysis code, WRAC-J. The radiolysis calculation results were transferred to an ECP model. In the model, anodic and cathodic current densities were calculated with combination of an electrochemistry model and an oxide film growth model. The measured ECP data were compared with the radiolysis/ECP calculation results, and applicability the of radiolysis model was confirmed. In addition, anomalous phenomenon appears in the in-pile loop was also investigated by radiolysis calculations. (author)
CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS
International Nuclear Information System (INIS)
C.E. Sanders
2005-01-01
This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for the current design of the CHF and may not reflect the ongoing design evolution of the facility
Nozzle flow calculation for real gases
International Nuclear Information System (INIS)
Bier, K.; Ehrler, F.; Hartz, U.; Kissau, G.
1977-01-01
The flow of CHF 2 Cl vapor (refrigerant R 22) through a Laval nozzle of annular geometry has been investigated in the region near the saturation line with stagnation pressures up to 85 per cent of the critical pressure. Static pressure profiles measured along the nozzle axis were found in good agreement with profiles calculated for one-dimensional isentropic flow of the real gas the thermal properties of which were derived from an equation of state proposed previously by Rombusch. Minor deviations between measured and calculated static pressure curves occur in the supersonic part of the mozzle, especially when supersaturated states of the vapour are passed. These deviations can be attributed to uncertainties in the calculation of the enthalpy and to a small influence of the static pressure probe. An additional investigation was concerned with an approximate calculation of the nozzle flow of real gases. In this approximation the well known relations of ideal gas dynamics are applied, the ratio of specific heats for the ideal gas being replaced, however, by a suitably adapted isentropic exponent, which can be determined e.g. from measured values of the Laval pressure or of the mass flow. For pressure ratios p/po between 1 and approximately 0.1, corresponding to Mach numbers up to approximately 2.2, all the interesting properties of the investigated flow of CHF 2 Cl vapour are approximated within a few per cent. (orig.) [de
Calculating Free Energies Using Average Force
Darve, Eric; Pohorille, Andrew; DeVincenzi, Donald L. (Technical Monitor)
2001-01-01
A new, general formula that connects the derivatives of the free energy along the selected, generalized coordinates of the system with the instantaneous force acting on these coordinates is derived. The instantaneous force is defined as the force acting on the coordinate of interest so that when it is subtracted from the equations of motion the acceleration along this coordinate is zero. The formula applies to simulations in which the selected coordinates are either unconstrained or constrained to fixed values. It is shown that in the latter case the formula reduces to the expression previously derived by den Otter and Briels. If simulations are carried out without constraining the coordinates of interest, the formula leads to a new method for calculating the free energy changes along these coordinates. This method is tested in two examples - rotation around the C-C bond of 1,2-dichloroethane immersed in water and transfer of fluoromethane across the water-hexane interface. The calculated free energies are compared with those obtained by two commonly used methods. One of them relies on determining the probability density function of finding the system at different values of the selected coordinate and the other requires calculating the average force at discrete locations along this coordinate in a series of constrained simulations. The free energies calculated by these three methods are in excellent agreement. The relative advantages of each method are discussed.
SCALE Sensitivity Calculations Using Contributon Theory
International Nuclear Information System (INIS)
Rearden, Bradley T.; Perfetti, Chris; Williams, Mark L.; Petrie, Lester M. Jr.
2010-01-01
The SCALE TSUNAMI-3D sensitivity and uncertainty analysis sequence computes the sensitivity of k-eff to each constituent multigroup cross section using adjoint techniques with the KENO Monte Carlo codes. A new technique to simultaneously obtain the product of the forward and adjoint angular flux moments within a single Monte Carlo calculation has been developed and implemented in the SCALE TSUNAMI-3D analysis sequence. A new concept in Monte Carlo theory has been developed for this work, an eigenvalue contributon estimator, which is an extension of previously developed fixed-source contributon estimators. A contributon is a particle for which the forward solution is accumulated, and its importance to the response, which is equivalent to the adjoint solution, is simultaneously accumulated. Thus, the contributon is a particle coupled with its contribution to the response, in this case k-eff. As implemented in SCALE, the contributon provides the importance of a particle exiting at any energy or direction for each location, energy and direction at which the forward flux solution is sampled. Although currently implemented for eigenvalue calculations in multigroup mode in KENO, this technique is directly applicable to continuous-energy calculations for many other responses such as fixed-source sensitivity analysis and quantification of reactor kinetics parameters. This paper provides the physical bases of eigenvalue contributon theory, provides details of implementation into TSUNAMI-3D, and provides results of sample calculations.
On the relativistic calculation of spontaneous emission
International Nuclear Information System (INIS)
Boudet, R.
1993-01-01
In a recent work, Barut and Salamin (1988) have derived a method for calculating the relativistic decay rates in atoms, in a formulation of quantum electrodynamics based upon the electron's self-energy. The decay rate appears as the imaginary part of a formula giving a complex energy shift, the real part of the formula being the Lamb shift. The presence of the the decay rate in the imaginary part of a formula, giving an energy in its real part, may appear a bit strange. A confirmation of the Barut and Alamin calculation, by means of a quite different point of view, would be useful. Therefore in this work the Einstein A coefficients are calculated, in all cases of degeneracies of the Dirac transition currents, by means of the energy balance method. This point of view is based on the balance between the energy released during the transitions of electrons from a higher state to a lower one, and the flux of the Poynting vector of the classical electromagnetic field, created by the electrons, through a sphere a large radius. The particularity of the present work lies in the direct calculation of the relativistic Dirac transition currents and the fact that the dipole and Pauli approximations are avoided. The quantum part of the relativistic calculation is based on the determination of the transition charge currents in the Darwin solutions of the Dirac equation. 13 refs
International Nuclear Information System (INIS)
Kim, Jung-Ha; Hill, Robin; Kuncic, Zdenka
2012-01-01
The Monte Carlo (MC) method has proven invaluable for radiation transport simulations to accurately determine radiation doses and is widely considered a reliable computational measure that can substitute a physical experiment where direct measurements are not possible or feasible. In the EGSnrc/BEAMnrc MC codes, there are several user-specified parameters and customized transport algorithms, which may affect the calculation results. In order to fully utilize the MC methods available in these codes, it is essential to understand all these options and to use them appropriately. In this study, the effects of the electron transport algorithms in EGSnrc/BEAMnrc, which are often a trade-off between calculation accuracy and efficiency, were investigated in the buildup region of a homogeneous water phantom and also in a heterogeneous phantom using the DOSRZnrc user code. The algorithms and parameters investigated include: boundary crossing algorithm (BCA), skin depth, electron step algorithm (ESA), global electron cutoff energy (ECUT) and electron production cutoff energy (AE). The variations in calculated buildup doses were found to be larger than 10% for different user-specified transport parameters. We found that using BCA = EXACT gave the best results in terms of accuracy and efficiency in calculating buildup doses using DOSRZnrc. In addition, using the ESA = PRESTA-I option was found to be the best way of reducing the total calculation time without losing accuracy in the results at high energies (few keV ∼ MeV). We also found that although choosing a higher ECUT/AE value in the beam modelling can dramatically improve computation efficiency, there is a significant trade-off in surface dose uncertainty. Our study demonstrates that a careful choice of user-specified transport parameters is required when conducting similar MC calculations. (note)
Parameters calculation of fuel assembly with complex geometry
International Nuclear Information System (INIS)
Wu Hongchun; Ju Haitao; Yao Dong
2006-01-01
The code DRAGON was developed for CANDU reactor by Ecole Polytechnique de Montreal of Canada. In order to validate the DRAGON code's applicability for complex geometry fuel assembly calculation, the rod shape fuel assembly of PWR benchmark problem and the plate shape fuel assembly of MTR benchmark problem were analyzed by DRAGON code. Some other shape fuel assemblies were also discussed simply. Calculation results show that the DRAGON code can be used to calculate variform fuel assembly and the precision is high. (authors)
Ab initio valence calculations in chemistry
Cook, D B
1974-01-01
Ab Initio Valence Calculations in Chemistry describes the theory and practice of ab initio valence calculations in chemistry and applies the ideas to a specific example, linear BeH2. Topics covered include the Schrödinger equation and the orbital approximation to atomic orbitals; molecular orbital and valence bond methods; practical molecular wave functions; and molecular integrals. Open shell systems, molecular symmetry, and localized descriptions of electronic structure are also discussed. This book is comprised of 13 chapters and begins by introducing the reader to the use of the Schrödinge
Effective operators in nuclear-structure calculations
International Nuclear Information System (INIS)
Barrett, Bruce R
2005-01-01
A brief review of the history of the use of many-body perturbation theory to determine effective operators for shell-model calculations, i.e., for calculations in truncated model spaces, is given, starting with the ground-breaking work of Arima and Horie for electromagnetic moments. The problems encountered in utilizing this approach are discussed. New methods based on unitary-transformation approaches are introduced and analyzed. The old problems persist, but the new methods allow us to obtain a better insight into the nature of the physics involved in these processes
Criticality calculation of non-ordinary systems
Energy Technology Data Exchange (ETDEWEB)
Kalugin, A. V., E-mail: Kalugin-AV@nrcki.ru; Tebin, V. V. [National Research Centre Kurchatov Institute (Russian Federation)
2016-12-15
The specific features of calculation of the effective multiplication factor using the Monte Carlo method for weakly coupled and non-asymptotic multiplying systems are discussed. Particular examples are considered and practical recommendations on detection and Monte Carlo calculation of systems typical in numerical substantiation of nuclear safety for VVER fuel management problems are given. In particular, the problems of the choice of parameters for the batch mode and the method for normalization of the neutron batch, as well as finding and interpretation of the eigenvalue spectrum for the integral fission matrix, are discussed.
Many body calculations in atomic physics
International Nuclear Information System (INIS)
Kelly, H.P.
1985-01-01
The use of the many-body perturbation theory for atomic calculations are reviewed. The major emphasis is on the use of the linked-cluster many-body perturbation theory derived by Brueckner and Goldstone. Applications of many-body theory to calculations of hyperfine structure are examined. Auger rates and parity violation are discussed. Photoionization is reviewed, and the authors show how many-body perturbation theory can be applied to problems ranging from structural properties such as correlation energies and hyperfine structure to dynamical properties such as transitions induced by weak neutral currents and photoionization cross sections
Heat Exchanger Support Bracket Design Calculations
International Nuclear Information System (INIS)
Rucinski, Russ
1995-01-01
This engineering note documents the design of the heat exchanger support brackets. The heat exchanger is roughly 40 feet long, 22 inches in diameter and weighs 6750 pounds. It will be mounted on two identical support brackets that are anchored to a concrete wall. The design calculations were done for one bracket supporting the full weight of the heat exchanger, rounded up to 6800 pounds. The design follows the American Institute of Steel Construction (AISC) Manual of steel construction, Eighth edition. All calculated stresses and loads on welds were below allowables.
Ammonia synthesis from first principles calculations
DEFF Research Database (Denmark)
Honkala, Johanna Karoliina; Hellman, Anders; Remediakis, Ioannis
2005-01-01
. When the size distribution of ruthenium particles measured by transmission electron microscopy was used as the [ink between the catalyst material and the theoretical treatment, the calculated rate was within a factor of 3 to 20 of the experimental rate. This offers hope for computer-based methods......The rate of ammonia synthesis over a nanoparticle ruthenium catalyst can be calculated directly on the basis of a quantum chemical treatment of the problem using density functional theory. We compared the results to measured rates over a ruthenium catalyst supported on magnesium aluminum spinet...
Examples of 3-D field calculations using GFUN
International Nuclear Information System (INIS)
Lari, R.J.
1981-01-01
Magnets are described that were calculated using GFUN. A four-step procedure is used to calculate magnets using GFUN. First, the interactive TSO system is used to draw the geometry of the magnet on a Tektronix 4012 graphic display unit. When the geometry is correct, it is stored on a disc file that is shared by the batch computers 3033 and 195. A file JCL and data can be created on TSO and submitted to the batch computers to calculate the magnetization of the steel tetrahedron elements. The results of this approx. 1 hour batch job are stored on disc. In the same job, or in a separate one, the fields can be calculated at desired points and stored on a shared disc. The fourth step is to plot these fields interactively in TSO
Calculation of Si(Li) x-ray detector efficiencies
International Nuclear Information System (INIS)
Zaluzec, N.; Holton, R.
1984-01-01
The calculation of detector efficiency functions is an important step in the quantitative analysis of x-ray spectra when approached by a standardless technique. In this regard, it becomes essential that the analyst not only model the physical aspects of the absorption and transmission of the various windows present, but also use the most accurate data available for the mass absorption coefficients required in these calculations. The topic of modeling the size and shape of the windows present is beyond the scope of this paper and the authors instead concentrate on the mass absorption coefficients used in the calculations and their implications to efficiency calculations. For the purposes of this paper, the authors consider that the relative detector efficiency function of a conventional Si(Li) detector can be modeled by a simple expression
International Nuclear Information System (INIS)
Thompson, W.L.; Deutsch, O.L.; Booth, T.E.
1980-04-01
Several Monte Carlo techniques are compared in the transport of neutrons of different source energies through two different deep-penetration problems each with two parts. The first problem involves transmission through a 200-cm concrete slab. The second problem is a 90 0 bent pipe jacketed by concrete. In one case the pipe is void, and in the other it is filled with liquid sodium. Calculations are made with two different Los Alamos Monte Carlo codes: the continuous-energy code MCNP and the multigroup code MCMG
Method for consequence calculations for severe accidents
International Nuclear Information System (INIS)
Nielsen, F.; Thykier-Nielsn, S.
1987-03-01
This report was commissioned by the Swedish State Power Board. The report contains a calculation of radiation doses in the surroundings caused by a theoretical core meltdown accident at Forsmark reactor No 3. The assumption used for the calculations were a 0.06% release of iodine and cesium corresponding to a 0.1% release through the FILTRA plant at Barsebaeck. The calculations were made by means of the PLUCON4 code. Meteorological data for two years from the Forsmark meteorological tower were analysed to find representative weather situations. As typical weather pasquill D was chosen with wind speed 5 m/s, and as extreme weather, Pasquill F with wind speed 2 m/s. 23 tabs., 36 ills., 21 refs. (author)
Calculating seismic of slabs ITA NNP Garona
International Nuclear Information System (INIS)
Ezeberry, J. I.; Guerrero, A.; Gamarra, J.; Beltran, F.
2014-01-01
This article describes the methodology that Idom has employed to perform the seismic evaluation of slabs within the ITA project of the NPP Santa Maria de Garona. Seismic calculations that have been conducted include consideration of the effects of the interaction of soil structure as well as the possible take-off containers with respect to slab during the earthquake. Therefore, the main contribution of the work is the study of the coupling of rolling containers with the flexibility of the whole ground-slab For calculations has been used ABAQUS/Explicit program, allowing to solve effectively the nonlinearities listed above using explicit integration algorithms over time. The results of the calculations reflect the importance of jointly analyse the seismic responses of slab and containers. (Author)
Accelerating GW calculations with optimal polarizability basis
Energy Technology Data Exchange (ETDEWEB)
Umari, P.; Stenuit, G. [CNR-IOM DEMOCRITOS Theory Elettra Group, Basovizza (Trieste) (Italy); Qian, X.; Marzari, N. [Department of Materials Science and Engineering, MIT, Cambridge, MA (United States); Giacomazzi, L.; Baroni, S. [CNR-IOM DEMOCRITOS Theory Elettra Group, Basovizza (Trieste) (Italy); SISSA - Scuola Internazionale Superiore di Studi Avanzati, Trieste (Italy)
2011-03-15
We present a method for accelerating GW quasi-particle (QP) calculations. This is achieved through the introduction of optimal basis sets for representing polarizability matrices. First the real-space products of Wannier like orbitals are constructed and then optimal basis sets are obtained through singular value decomposition. Our method is validated by calculating the vertical ionization energies of the benzene molecule and the band structure of crystalline silicon. Its potentialities are illustrated by calculating the QP spectrum of a model structure of vitreous silica. Finally, we apply our method for studying the electronic structure properties of a model of quasi-stoichiometric amorphous silicon nitride and of its point defects. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)
The validation of neutron kinetic calculations of CEGB reactors
International Nuclear Information System (INIS)
Emmett, J.C.A.; Hutt, P.K.; Nunn, D.L.; Waterson, R.H.
1982-01-01
Reactor kinetic calculations are required by the CEGB to predict space and time varying neutron fluxes through the course of various hypothesized core transients. These transients arise through flow or reactivity perturbations occurring in a part of the core. A description is given of the results of dual programmes of work undertaken at BNL to validate such calculations. Firstly, analyses have been carried out to establish how data for these calculations should best be derived. Secondly, experimental measurements have been compared against the predictions of such calculations with data derived in the recommended way. (author)
Method and program for complex calculation of heterogeneous reactor
International Nuclear Information System (INIS)
Kalashnikov, A.G.; Glebov, A.P.; Elovskaya, L.F.; Kuznetsova, L.I.
1988-01-01
An algorithm and the GITA program for complex one-dimensional calculation of a heterogeneous reactor which permits to conduct calculations for the reactor and its cell simultaneously using the same algorithm are described. Multigroup macrocross sections for reactor zones in the thermal energy range are determined according to the technique for calculating a cell with complicate structure and then the continuous multi group calculation of the reactor in the thermal energy range and in the range of neutron thermalization is made. The kinetic equation is solved using the Pi- and DSn- approximations [fr
Direct calculation of current drive efficiency in FISIC code
International Nuclear Information System (INIS)
Wright, J.C.; Phillips, C.K.; Bonoli, P.T.
1996-01-01
Two-dimensional RF modeling codes use a parameterization (1) of current drive efficiencies to calculate fast wave driven currents. This parameterization assumes a uniform quasi-linear diffusion coefficient and requires a priori knowledge of the wave polarizations. These difficulties may be avoided by a direct calculation of the quasilinear diffusion coefficient from the Kennel-Englemann form with the field polarizations calculated by the full wave code, FISIC (2). Current profiles are calculated using the adjoint formulation (3). Comparisons between the two formulations are presented. copyright 1996 American Institute of Physics
Calculations of radiation levels during reactor operations for safeguard inspections
International Nuclear Information System (INIS)
Sobhy, M.
1996-01-01
When an internal core spent fuel storage is used in the shield tank to accommodate a large number of spent fuel baskets, physical calculations are performed to determine the number of these spent fuel elements which can be accommodated and still maintain the gamma activity outside under the permissible limit. The corresponding reactor power level is determined. The radioactivity calculations are performed for this internal storage at different axial levels to avoid the criticality of the reactor core. Transport theory is used in calculations based on collision probability for multi group cell calculations. Diffusion theory is used in three dimensions in the core calculations. The nuclear fuel history is traced and radioactive decay is calculated, since reactor fission products are very sensitive to power level. The radioactivity is calculated with a developed formula based on fuel basket loading integrity. (author)
International Nuclear Information System (INIS)
Ferraro, V.; Igawa, N.; Marinelli, V.
2010-01-01
A calculation code, named INLUX-DBR, is presented, which is a modified version of INLUX code, able to predict the illuminance distribution on the inside surfaces of a room with six walls and a window, and on the work plane. At each desired instant the code solves the system of the illuminance equations of each surface element, characterized by the latter's reflection coefficient and its view factors toward the other elements. In the model implemented in the code, the sky-diffuse luminance distribution, the sun beam light and the light reflected from the ground toward the room are considered. The code was validated by comparing the calculated values of illuminance with the experimental values measured inside a scale model (1:5) of a building room, in various sky conditions of overcast, clear and intermediate days. The validation is performed using the sky luminance data measured by a sky scanner and the measured beam illuminance of the sun as input data. A comparative analysis of some of the well-known calculation models of sky luminance, namely Perez, Igawa and CIE models was also carried out, comparing the code predictions and the measured values of inside illuminance in the scale model.
theory and calculation of the design of nuclear reactor
International Nuclear Information System (INIS)
Refaat, R.A.
1994-01-01
For the sake of formation of a complete general code for nuclear power reactor design, this thesis deals with a great part of this code. the code links the solution of the neutron integral transport equation by the multigroup treatment (76 energy groups) for the calculation of the reactor cell parameters by the fuel management program that solves the neutron diffusion equation inside a large number of nuclear fuel assemblies. the lattice cell code is modified to accommodate the calculation of lattice cell parameters for more than one enrichment ( one after the other). it is also modified to calculate the burn up parameters using unequal time steps. these two modifications are complicated but necessary for the link between the cell program and fuel management program. the comparison between the results of the fitted cross sections and that given by the cell calculations shows the necessity of using the cell code cross sections. this is also necessary for the sake of generality for any type of reactors. the comparison for the fuel management calculation depending on fitted data and that depending on cell calculation data insures the necessity for using the cell data i.e. insures the necessity of linking the cell calculation program by the fuel management program
Body Mass Index: Calculator for Child and Teen
... Healthy Weight Sample Link BMI Percentile Calculator for Child and Teen English Version Language: English Español (Spanish) ... and Weight Accurately At Home BMI Calculator for Child and Teen ( English | Metric ) 1. Birth Date : Month: ...
Diffusion Monte Carlo calculation of three-body systems
International Nuclear Information System (INIS)
Lu Mengjiao; Lin Qihu; Ren Zhongzhou
2012-01-01
The application of the diffusion Monte Carlo algorithm in three-body systems is studied. We develop a program and use it to calculate the property of various three-body systems. Regular Coulomb systems such as atoms, molecules, and ions are investigated. The calculation is then extended to exotic systems where electrons are replaced by muons. Some nuclei with neutron halos are also calculated as three-body systems consisting of a core and two external nucleons. Our results agree well with experiments and others' work. (authors)
Fe IX CALCULATIONS FOR THE SOLAR DYNAMICS OBSERVATORY
International Nuclear Information System (INIS)
Foster, Adam R.; Testa, Paola
2011-01-01
New calculations of the energy levels, radiative transition rates, and collisional excitation rates of Fe IX have been carried out using the Flexible Atomic Code, paying close attention to experimentally identified levels and extending existing calculations to higher energy levels. For lower levels, R-matrix collisional excitation rates from earlier work have been used. Significant emission is predicted by these calculations in the 5f-3d transitions, which will impact analysis of Solar Dynamics Observatory Atmospheric Imaging Assembly observations using the 94 A filter.
Drop Calculations of HLW Canister and Pu Can-in-Canister
International Nuclear Information System (INIS)
Sreten Mastilovic
2001-01-01
The objective of this calculation is to determine the structural response of the standard high-level waste (HLW) canister and the canister containing the cans of immobilized plutonium (Pu) (''can-in-canister'' [CIC] throughout this document) subjected to drop DBEs (design basis events) during the handling operation. The evaluated DBE in the former case is 7-m (23-ft) vertical (flat-bottom) drop. In the latter case, two 2-ft (0.61-m) corner (oblique) drops are evaluated in addition to the 7-m vertical drop. These Pu CIC calculations are performed at three different temperatures: room temperature (RT) (20 C), T = 200 F = 93.3 C , and T = 400 F = 204 C ; in addition to these the calculation characterized by the highest maximum stress intensity is performed at T = 750 F = 399 C as well. The scope of the HLW canister calculation is limited to reporting the calculation results in terms of: stress intensity and effective plastic strain in the canister, directional residual strains at the canister outer surface, and change of canister dimensions. The scope of Pu CIC calculation is limited to reporting the calculation results in terms of stress intensity, and effective plastic strain in the canister. The information provided by the sketches from Reference 26 (Attachments 5.3,5.5,5.8, and 5.9) is that of the potential CIC design considered in this calculation, and all obtained results are valid for this design only. This calculation is associated with the Plutonium Immobilization Project and is performed by the Waste Package Design Section in accordance with Reference 24. It should be noted that the 9-m vertical drop DBE, included in Reference 24, is not included in the objective of this calculation since it did not become a waste acceptance requirement. AP-3.124, ''Calculations'', is used to perform the calculation and develop the document
The Collective Practice of Calculation
DEFF Research Database (Denmark)
Schrøder, Ida
The calculation of costs plays an increasingly large role in the decision-making processes of public sector human service organizations. This has brought scholars of management accounting to investigate the relationship between caring professions and demands to make economic entities of the service...... productions as either a process of hybridization or conflict. With these approaches, though, they fail to interrogate the possibility that professional action might not be either the one or the other, but entail a broad variety of relationships between calculations and judgements. This paper elaborates...... and judgement to reach decisions to invest in social services. The line is not drawn between the two, but between the material arrangements that make decisions possible. This implies that the insisting on qualitatively based decisions gives the professionals agency to collectively engage in practical...
SRAC-95, Cell Calculation with Burnup, Fuel Management for Thermal Reactors
International Nuclear Information System (INIS)
Tsuchihashi, K.; Ishiguro, Y.; Kaneko, K.; Ido, M.
2004-01-01
1 - Description of program or function: General neutronics calculation including cell calculation with burn-up, core calculation for any type of thermal reactor. Core burn-up calculation and fuel management by an auxiliary code. 2 - Method of solution: Collision probability method, 1D and 2D Sn for cell calculation; 1D, 2D and 3D diffusion for core calculation. 3 - Restrictions on the complexity of the problem: 20 regions for a continuous energy resonance absorption calculation and 16 steps for cell burn-up
Study of the accuracy of radiation field calculations in media
International Nuclear Information System (INIS)
Bolyatko, V.V.; Vyrskij, M.Yu.; Ilyushkin, A.I.; Mashkovich, V.P.; Sakharov, V.K.; Stroganov, A.A.
1981-01-01
The sensitivity p of the radiation transport calculations to variations of input parameters Xsub(i) is theoretically analyzed, and the calculational errors induced by uncertainties of initial data are evaluated. Two calculational methods are considered: the direct substitution method using the ROZ-5 code and method using the linear perturbation theory. In order to calculate p(Xsub(i)) and bilinear convolutions of the conjugated transport equations the ZAKAT code has been developed. The calculations use the ZAKAT, ROZ-11 and APAMAKO-2F codes. As an example of practical use of the method proposed a shielding composition characteristic for fast reactors was analyzed. A plane monodirectional neutron beam of the BR-10 reactor falls onto a 5-layer stainless steel (1Kh18N10T)-carbon barrier. The sensitivily of the neutron dose absorbed in tissue to the cross sections of all the shielding constituents and to the source and detector representation functions has been calculated. A comparison of the calculations with experimental data proves the validity of the calculational method [ru
Calculational approach to ionization spectrometer design
International Nuclear Information System (INIS)
Gabriel, T.A.
1974-01-01
Many factors contribute to the design and overall performance of an ionization spectrometer. These factors include the conditions under which the spectrometer is to be used, the required performance, the development of the hadronic and electromagnetic cascades, leakage and binding energies, saturation effects of densely ionizing particles, nonuniform light collection, sampling fluctuations, etc. The calculational procedures developed at Oak Ridge National Laboratory that have been applied to many spectrometer designs and that include many of the influencing factors in spectrometer design are discussed. The incident-particle types which can be considered with some generality are protons, neutrons, pions, muons, electrons, positrons, and gamma rays. Charged kaons can also be considered but with less generality. The incident-particle energy range can extend into the hundreds of GeV range. The calculations have been verified by comparison with experimental data but only up to approximately 30 GeV. Some comparisons with experimental data are also discussed and presented so that the flexibility of the calculational methods can be demonstrated. (U.S.)
TTS-Polttopuu - cost calculation model for fuelwood
International Nuclear Information System (INIS)
Naett, H.; Ryynaenen, S.
1999-01-01
The TTS-Institutes's Forestry Department has developed a computer based cost-calculation model, 'TTS-Polttopuu', for the calculation of unit costs and resource needs in the harvesting systems for wood chips and split firewood. The model enables to determine the productivity and device cost per operating hour by each working stage of the harvesting system. The calculation model also enables the user to find out how changes in the productivity and cost bases of different harvesting chains influence the unit cost of the whole system. The harvesting chain includes the cutting of delimbed and non-delimbed fuelwood, forest haulage, road transportation, chipping and chopping of longwood at storage. This individually operating software was originally developed to serve research needs, but it also serves the needs of the forestry and agricultural education, training and extension as well as individual firewood producers. The system requirements for this cost calculation model are at least 486- level processor with the Windows 95/98 -operating system, 16 MB of memory (RAM) and 5 MB of available hard-disk. This development work was carried out in conjunction with the nation-wide BIOENERGY-research programme. (orig.)
NEPTUNE: a modular system for light-water reactor calculation
International Nuclear Information System (INIS)
Bouchard, J.; Kanevoky, A.; Reuss, P.
1975-01-01
A complete modular system of light water reactor calculations has been designed. It includes basic nuclear data processing, the APOLLO phase: transport calculations for cells, multicells, fuel assemblies or reactors, the NEPTUNE phase: reactor calculations. A fuel management module, devoted to the automatic determination of the best shuffling strategy is included in NEPTUNE [fr
International Nuclear Information System (INIS)
Penner, J.E.; Haselman, L.C.; Edwards, L.L.
1985-01-01
Smoke from raging fires produced in the aftermath of a major nuclear exchange has been predicted to cause large decreases in surface temperatures. However, the extent of the decrease and even the sign of the temperature change, depend on how the smoke is distributed with altitude. We present a model capable of evaluating the initial distribution of lofted smoke above a massive fire. Calculations are shown for a two-dimensional slab version of the model and a full three-dimensional version. The model has been evaluated by simulating smoke heights for the Hamburg firestorm of 1943 and a smaller scale oil fire which occurred in Long Beach in 1958. Our plume heights for these fires are compared to those predicted by the classical Morton-Taylor-Turner theory for weakly buoyant plumes. We consider the effect of the added buoyancy caused by condensation of water-laden ground level air being carried to high altitude with the convection column as well as the effects of background wind on the calculated smoke plume heights for several fire intensities. We find that the rise height of the plume depends on the assumed background atmospheric conditions as well as the fire intensity. Little smoke is injected into the stratosphere unless the fire is unusually intense, or atmospheric conditions are more unstable than we have assumed. For intense fires significant amounts of water vapor are condensed raising the possibility of early scavenging of smoke particles by precipitation. 26 references, 11 figures
Dynamical calculations for RHEED intensity oscillations
Daniluk, Andrzej
2005-03-01
A practical computing algorithm working in real time has been developed for calculating the reflection high-energy electron diffraction from the molecular beam epitaxy growing surface. The calculations are based on the use of a dynamical diffraction theory in which the electrons are taken to be diffracted by a potential, which is periodic in the dimension perpendicular to the surface. The results of the calculations are presented in the form of rocking curves to illustrate how the diffracted beam intensities depend on the glancing angle of the incident beam. Program summaryTitle of program: RHEED Catalogue identifier:ADUY Program summary URL:http://cpc.cs.qub.ac.uk/summaries/ADUY Program obtainable from:CPC Program Library, Queen's University of Belfast, N. Ireland Computer for which the program is designed and others on which it has been tested: Pentium-based PC Operating systems or monitors under which the program has been tested: Windows 9x, XP, NT, Linux Programming language used: Borland C++ Memory required to execute with typical data: more than 1 MB Number of bits in a word: 64 bits Number of processors used: 1 Distribution format:tar.gz Number of lines in distributed program, including test data, etc.:982 Number of bytes in distributed program, including test data, etc.: 126 051 Nature of physical problem: Reflection high-energy electron diffraction (RHEED) is a very useful technique for studying growth and surface analysis of thin epitaxial structures prepared by the molecular beam epitaxy (MBE). Nowadays, RHEED is used in many laboratories all over the world where researchers deal with the growth of materials by MBE. The RHEED technique can reveal, almost instantaneously, changes either in the coverage of the sample surface by adsorbates or in the surface structure of a thin film. In most cases the interpretation of experimental results is based on the use of dynamical diffraction approaches. Such approaches are said to be quite useful in qualitative and
Updated Collisional Ionization Equilibrium Calculated for Optically Thin Plasmas
Savin, Daniel Wolf; Bryans, P.; Badnell, N. R.; Gorczyca, T. W.; Laming, J. M.; Mitthumsiri, W.
2010-03-01
Reliably interpreting spectra from electron-ionized cosmic plasmas requires accurate ionization balance calculations for the plasma in question. However, much of the atomic data needed for these calculations have not been generated using modern theoretical methods and their reliability are often highly suspect. We have carried out state-of-the-art calculations of dielectronic recombination (DR) rate coefficients for the hydrogenic through Na-like ions of all elements from He to Zn as well as for Al-like to Ar-like ions of Fe. We have also carried out state-of-the-art radiative recombination (RR) rate coefficient calculations for the bare through Na-like ions of all elements from H to Zn. Using our data and the recommended electron impact ionization data of Dere (2007), we present improved collisional ionization equilibrium calculations (Bryans et al. 2006, 2009). We compare our calculated fractional ionic abundances using these data with those presented by Mazzotta et al. (1998) for all elements from H to Ni. This work is supported in part by the NASA APRA and SHP SR&T programs.
Impurities in semiconductors: total energy and infrared absorption calculations
International Nuclear Information System (INIS)
Yndurain, F.
1987-01-01
A new method to calculate the electronic structure of infinite nonperiodic system is discussed. The calculations are performed using atomic pseudopotentials and a basis of atomic Gaussiam wave functions. The Hartree-Fock self consistent equations are solved in the cluster-Bethe lattice system. Electron correlation is partially included in second order pertubation approximation. The formalism is applied to hydrogenated amorphous silicon. Total energy calculations of finite clusters of silicon atom in the presence of impurities, are also presented. The results show how atomic oxygen breaks the covalent silicon silicon bond forming a local configuration similar to that of SiO 2 . Calculations of the infrared absorption due to the presence of atomic oxygen in cristalline silicon are presented. The Born Hamiltonian to calculate the vibrational modes of the system and a simplied model to describe the infrared absorption mechanism are used. The interstitial and the the substitutional cases are considered and analysed. The position of the main infrared absorption peak, their intensities and their isotope shifts are calculated. The results are satisfactory agreement with the available data. (author) [pt
Cluster-cell calculation using the method of generalized homogenization
International Nuclear Information System (INIS)
Laletin, N.I.; Boyarinov, V.F.
1988-01-01
The generalized-homogenization method (GHM), used for solving the neutron transfer equation, was applied to calculating the neutron distribution in the cluster cell with a series of cylindrical cells with cylindrically coaxial zones. Single-group calculations of the technological channel of the cell of an RBMK reactor were performed using GHM. The technological channel was understood to be the reactor channel, comprised of the zirconium rod, the water or steam-water mixture, the uranium dioxide fuel element, and the zirconium tube, together with the adjacent graphite layer. Calculations were performed for channels with no internal sources and with unit incoming current at the external boundary as well as for channels with internal sources and zero current at the external boundary. The PRAKTINETs program was used to calculate the symmetric neutron distributions in the microcell and in channels with homogenized annular zones. The ORAR-TsM program was used to calculate the antisymmetric distribution in the microcell. The accuracy of the calculations were compared for the two channel versions
Power calculator for instrumental variable analysis in pharmacoepidemiology.
Walker, Venexia M; Davies, Neil M; Windmeijer, Frank; Burgess, Stephen; Martin, Richard M
2017-10-01
Instrumental variable analysis, for example with physicians' prescribing preferences as an instrument for medications issued in primary care, is an increasingly popular method in the field of pharmacoepidemiology. Existing power calculators for studies using instrumental variable analysis, such as Mendelian randomization power calculators, do not allow for the structure of research questions in this field. This is because the analysis in pharmacoepidemiology will typically have stronger instruments and detect larger causal effects than in other fields. Consequently, there is a need for dedicated power calculators for pharmacoepidemiological research. The formula for calculating the power of a study using instrumental variable analysis in the context of pharmacoepidemiology is derived before being validated by a simulation study. The formula is applicable for studies using a single binary instrument to analyse the causal effect of a binary exposure on a continuous outcome. An online calculator, as well as packages in both R and Stata, are provided for the implementation of the formula by others. The statistical power of instrumental variable analysis in pharmacoepidemiological studies to detect a clinically meaningful treatment effect is an important consideration. Research questions in this field have distinct structures that must be accounted for when calculating power. The formula presented differs from existing instrumental variable power formulae due to its parametrization, which is designed specifically for ease of use by pharmacoepidemiologists. © The Author 2017. Published by Oxford University Press on behalf of the International Epidemiological Association
Energy mesh optimization for multi-level calculation schemes
International Nuclear Information System (INIS)
Mosca, P.; Taofiki, A.; Bellier, P.; Prevost, A.
2011-01-01
The industrial calculations of third generation nuclear reactors are based on sophisticated strategies of homogenization and collapsing at different spatial and energetic levels. An important issue to ensure the quality of these calculation models is the choice of the collapsing energy mesh. In this work, we show a new approach to generate optimized energy meshes starting from the SHEM 281-group library. The optimization model is applied on 1D cylindrical cells and consists of finding an energy mesh which minimizes the errors between two successive collision probability calculations. The former is realized over the fine SHEM mesh with Livolant-Jeanpierre self-shielded cross sections and the latter is performed with collapsed cross sections over the energy mesh being optimized. The optimization is done by the particle swarm algorithm implemented in the code AEMC and multigroup flux solutions are obtained from standard APOLLO2 solvers. By this new approach, a set of new optimized meshes which encompass from 10 to 50 groups has been defined for PWR and BWR calculations. This set will allow users to adapt the energy detail of the solution to the complexity of the calculation (assembly, multi-assembly, two-dimensional whole core). Some preliminary verifications, in which the accuracy of the new meshes is measured compared to a direct 281-group calculation, show that the 30-group optimized mesh offers a good compromise between simulation time and accuracy for a standard 17 x 17 UO 2 assembly with and without control rods. (author)
Practical use of a microcomputer for ventilation calculations
International Nuclear Information System (INIS)
D'Albrand, N.; Froger, C.; Josien, J.P.
1983-01-01
Calculations of ventilation networks are necessary to elaborate the projects and to handle the gradual development of a ventilation network. The means which have been used up to the present to tackle these problems are the simulator and the computer, each with its own advantages and disadvantages. These means can be improved by considering the following needs of the user: short response time in the calculation of the state of a network; easy data input and presentation of the results (diagram, visual display); keeping records of the results. Cerchar has developed a program for the calculation of ventilation networks for use with a microcomputer psi 80 (Kontron) which has been constructed around a microprocessor Z 80. The data of the network to be calculated are entered by a keyboard and stored on a small disc; the data and the results can be displayed on a screen or produced as a print-out. This program is suitable for the calculation of networks composed of 400 branches and 300 nodes and comprising up to 9 different curves of characteristics as far as the mine fans are concerned. Cerchar investigates at present the visual display of a network on an interactive graphic terminal which, when perfected, will put at the disposal of the mine operators a tool for use by people other than data processing experts; this enables them to have permanently on the site their network for immediate access to consult it or make new calculations [fr
The number processing and calculation system: evidence from cognitive neuropsychology.
Salguero-Alcañiz, M P; Alameda-Bailén, J R
2015-04-01
Cognitive neuropsychology focuses on the concepts of dissociation and double dissociation. The performance of number processing and calculation tasks by patients with acquired brain injury can be used to characterise the way in which the healthy cognitive system manipulates number symbols and quantities. The objective of this study is to determine the components of the numerical processing and calculation system. Participants consisted of 6 patients with acquired brain injuries in different cerebral localisations. We used Batería de evaluación del procesamiento numérico y el cálculo, a battery assessing number processing and calculation. Data was analysed using the difference in proportions test. Quantitative numerical knowledge is independent from number transcoding, qualitative numerical knowledge, and calculation. Recodification is independent from qualitative numerical knowledge and calculation. Quantitative numerical knowledge and calculation are also independent functions. The number processing and calculation system comprises at least 4 components that operate independently: quantitative numerical knowledge, number transcoding, qualitative numerical knowledge, and calculation. Therefore, each one may be damaged selectively without affecting the functioning of another. According to the main models of number processing and calculation, each component has different characteristics and cerebral localisations. Copyright © 2013 Sociedad Española de Neurología. Published by Elsevier Espana. All rights reserved.
Thermal Margin Calculation of the CAREM-25 Core
International Nuclear Information System (INIS)
Mazufri, C.M
2000-01-01
During the operation in steady state and anticipated operational transient of a nuclear reactor it is necessary to avoid the damage in the fuel elements induced by thermal or hydraulic effects.To satisfy that design bases safety limits are established and calculation methodologies are defined to verify them.In the particular case of the reactor CAREM-25 reactor where the core is cooled by natural circulation, it is not adequate to use directly the same calculation methodologies from typical PWR and BWR.The low cooling flow rate and not having channels in the fuel elements (open-channel fuels) produce that most of the models and computer programs typically used must be carefully validated.As result of that process, an adequate calculation procedure for this reactor type was developed.In the present work, the thermal-hydraulic design criteria of the core and the design bases, the uncertainties factors, and the thermal margin results of the core are described.Despite that the methodology of DNBR calculation is under a validation process and considering the available calculation tools, it is possible to assure that the core fulfills the safety regulations in steady state conditions
Calculation of transport coefficients in an axisymmetric plasma
International Nuclear Information System (INIS)
Shumaker, D.E.
1977-01-01
A method of calculating the transport coefficient in an axisymmetric toroidal plasma is presented. This method is useful in calculating the transport coefficients in a Tokamak plasma confinement device. The particle density and temperature are shown to be a constant on a magnetic flux surface. Transport equations are given for the total particle flux and total energy flux crossing a closed toroidal surface. Also transport equations are given for the toroidal magnetic flux. A computer code was written to calculate the transport coefficients for a three species plasma, electrons and two species of ions. This is useful for calculating the transport coefficients of a plasma which contains impurities. It was found that the particle and energy transport coefficients are increased by a large amount, and the transport coefficients for the toroidal magnetic field are reduced by a small amount
Audit calculation for the LOCA methodology for KSNP
Energy Technology Data Exchange (ETDEWEB)
Lee, Un Chul; Park, Chang Hwan; Choi, Yong Won; Yoo, Jun Soo [Seoul National Univ., Seoul (Korea, Republic of)
2006-11-15
The objective of this research is to perform the audit regulatory calculation for the LOCA methodology for KSNP. For LBLOCA calculation, several uncertainty variables and new ranges of those are added to those of previous KINS-REM to improve the applicability of KINS-REM for KSNP LOCA. And those results are applied to LBLOCA audit calculation by statistical method. For SBLOCA calculation, after selecting BATHSY9.1.b, which is not used by KHNP, the results of RELAP5/Mod3.3 and RELAP5/MOD3.3ef-sEM for KSNP SBLOCA are compared to evaluate the conservativeness or applicability of RELAP5/MOD3.3ef-sEM code for KSNP SBLOCA. The result of this research can be used to support the activities of KINS for reviewing the LOCA methodology for KSNP proposed by KHNP.
Time step length versus efficiency of Monte Carlo burnup calculations
International Nuclear Information System (INIS)
Dufek, Jan; Valtavirta, Ville
2014-01-01
Highlights: • Time step length largely affects efficiency of MC burnup calculations. • Efficiency of MC burnup calculations improves with decreasing time step length. • Results were obtained from SIE-based Monte Carlo burnup calculations. - Abstract: We demonstrate that efficiency of Monte Carlo burnup calculations can be largely affected by the selected time step length. This study employs the stochastic implicit Euler based coupling scheme for Monte Carlo burnup calculations that performs a number of inner iteration steps within each time step. In a series of calculations, we vary the time step length and the number of inner iteration steps; the results suggest that Monte Carlo burnup calculations get more efficient as the time step length is reduced. More time steps must be simulated as they get shorter; however, this is more than compensated by the decrease in computing cost per time step needed for achieving a certain accuracy
Development of the code package KASKAD for calculations of WWERs
International Nuclear Information System (INIS)
Bolobov, P.A.; Lazarenko, A.P.; Tomilov, M.Ju.
2008-01-01
The new version of software package for neutron calculation of WWER cores KASKAD 2007 consists of some calculating and service modules, which are integrated in the common framework. The package is based on the old version, which was expanded with some new functions and the new calculating modules, such as: -the BIPR-2007 code is the new one which performs calculation of power distribution in three-dimensional geometry for 2-group neutron diffusion calculation. This code is based on the BIPR-8KN model, provides all possibilities of BIPR-7A code and uses the same input data; -the PERMAK-2007 code is pin-by-pin few-group multilayer and 3-D code for neutron diffusion calculation; -graphical user interface for input data preparation of the TVS-M code. The report also includes some calculation results obtained with modified version of the KASKAD 2007 package. (Authors)
Evaluation of RSG-GAS Core Management Based on Burnup Calculation
International Nuclear Information System (INIS)
Lily Suparlina; Jati Susilo
2009-01-01
Evaluation of RSG-GAS Core Management Based on Burnup Calculation. Presently, U 3 Si 2 -Al dispersion fuel is used in RSG-GAS core and had passed the 60 th core. At the beginning of each cycle the 5/1 fuel reshuffling pattern is used. Since 52 nd core, operators did not use the core fuel management computer code provided by vendor for this activity. They use the manually calculation using excel software as the solving. To know the accuracy of the calculation, core calculation was carried out using two kinds of 2 dimension diffusion codes Batan-2DIFF and SRAC. The beginning of cycle burn-up fraction data were calculated start from 51 st to 60 th using Batan-EQUIL and SRAC COREBN. The analysis results showed that there is a disparity in reactivity values of the two calculation method. The 60 th core critical position resulted from Batan-2DIFF calculation provide the reduction of positive reactivity 1.84 % Δk/k, while the manually calculation results give the increase of positive reactivity 2.19 % Δk/k. The minimum shutdown margin for stuck rod condition for manual and Batan-3DIFF calculation are -3.35 % Δk/k dan -1.13 % Δk/k respectively, it means that both values met the safety criteria, i.e <-0.5 % Δk/k. Excel program can be used for burn-up calculation, but it is needed to provide core management code to reach higher accuracy. (author)
Three dimensional diffusion calculations of nuclear reactors
International Nuclear Information System (INIS)
Caspo, N.
1981-07-01
This work deals with the three dimensional calculation of nuclear reactors using the code TRITON. The purposes of the work were to perform three-dimensional computations of the core of the Soreq nuclear reactor and of the power reactor ZION and to validate the TRITON code. Possible applications of the TRITON code in Soreq reactor calculations and in power reactor research are suggested. (H.K.)
Development of the code for filter calculation
International Nuclear Information System (INIS)
Gritzay, O.O.; Vakulenko, M.M.
2012-01-01
This paper describes a calculation method, which commonly used in the Neutron Physics Department to develop a new neutron filter or to improve the existing neutron filter. This calculation is the first step of the traditional filter development procedure. It allows easy selection of the qualitative and quantitative contents of a composite filter in order to receive the filtered neutron beam with given parameters