Benchmark calculations of sodium fast critical experiments
International Nuclear Information System (INIS)
The high expectations from fast critical experiments impose the additional requirements on reliability of final reconstructed values, obtained in experiments at critical facility. Benchmark calculations of critical experiments are characterized by impossibility of complete experiment reconstruction, the large amounts of input data (dependent and independent) with very different reliability. It should also take into account different sensitivity of the measured and appropriate calculated characteristics to the identical changes of geometry parameters, temperature, and isotopic composition of individual materials. The calculations of critical facility experiments are produced for the benchmark models, generated by the specific reconstructing codes with its features when adjusting model parameters, and using the nuclear data library. The generated benchmark model, providing the agreed calculated and experimental values for one or more neutronic characteristics can lead to considerable differences for other key characteristics. The sensitivity of key neutronic characteristics to the extra steel allocation in the core, and ENDF/B nuclear data sources is performed using a few calculated models of BFS-62-3A and BFS1-97 critical assemblies. The comparative analysis of the calculated effective multiplication factor, spectral indices, sodium void reactivity, and radial fission-rate distributions leads to quite different models, providing the best agreement the calculated and experimental neutronic characteristics. This fact should be considered during the refinement of computational models and code-verification purpose. (author)
Karma1.1 benchmark calculations for the numerical benchmark problems and the critical experiments
International Nuclear Information System (INIS)
The transport lattice code KARMA 1.1 has been developed at KAERI for the reactor physics analysis of the pressurized water reactor. This program includes the multi-group library processed from ENDF/B-VI R8 and also utilizes the macroscopic cross sections for the benchmark problems. Benchmark calculations were performed for the C5G7 and the KAERI benchmark problems given with seven group cross sections, for various fuels loaded in the operating pressurized water reactors in South Korea, and for the critical experiments including CE, B and W and KRITZ. Benchmark results show that KARMA 1.1 is working reasonably. (author)
MCNP calculations for Russian criticality-safety benchmarks
International Nuclear Information System (INIS)
The current edition of the International Handbook of Evaluated Criticality Safety Benchmark Experiments contains evaluations of 20 critical experiments performed and evaluated by the Institute for Experimental Physics of the Russian Federal Nuclear Center (VNIIEF) at Arzamas-16 and 16 critical experiments performed and evaluated by the Institute for Technical Physics of the Russian Federal Nuclear Center (VNIITF) at Chelyabinsk-70. These fast-spectrum experiments are of particular interest for data testing of ENDF/B-VI because they contain uranium metal systems of intermediate enrichment as well as uranium and plutonium metal systems with reflectors such as graphite, stainless steel, polyethylene, beryllium, and beryllium oxide. This paper presents the first published results for such systems using cross-section libraries based on ENDF/B-VI
International Nuclear Information System (INIS)
Criticality calculation codes/code systems MCNP, MVP, SCALE and JACS, which are currently typically used in Japan for nuclear criticality safety evaluation, were benchmarked for so called dissolver-typed systems, i.e., fuel rod arrays immersed in fuel solution. The benchmark analyses were made for the evaluated critical experiments published in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook: one evaluation representing five critical configurations from heterogeneous core of low-enriched uranium dioxides at the Japan Atomic Energy Research Institute and two evaluations representing 16 critical configurations from heterogeneous core of mixed uranium and plutonium dioxides (MOXs) at the Battelle Pacific Northwest Laboratories of the U.S.A. The results of the analyses showed that the minimum values of the neutron multiplication factor obtained with MCNP, MVP, SCALE and JACS were 0.993, 0.990, 0.993, 0.972, respectively, which values are from 2% to 4% larger than the maximum permissible multiplication factor of 0.95. (author)
Benchmark test of JEF-1 evaluation by calculating fast criticalities
International Nuclear Information System (INIS)
JEF-1 basic evaluation was tested by calculating fast critical experiments using the cross section discrete-ordinates transport code ONEDANT with P/sub 3/S/sub 16/ approximation. In each computation a spherical one dimensional model was used, together with a 174 neutron group VITAMIN-E structured JEF-1 based nuclear data library, generated at EIR with NJOY and TRANSX-CTR. It is found that the JEF-1 evaluation gives accurate results comparable with ENDF/B-V and that eigenvalues agree well within 10 mk whereas reaction rates deviate by up to 10% from the experiment. U-233 total and fission cross sections seem to be underestimated in the JEF-1 evaluation in the fast energy range between 0.1 and 1 MeV. This confirms previous analysis based on diffusion theory with 71 neutron groups, performed by H. Takano and E. Sartori at NEA Data Bank. (author)
Benchmark calculations by the nuclear criticality safety analysis code system JACS(MGCL, KENO-IV)
International Nuclear Information System (INIS)
Since 1980, as many as 1394 cases of benchmark calculations on criticality problems have been performed by the KENO-IV Monte Carlo calculation code with the MGCL cross section data library. The code system is a part of the criticality safety evaluation code system JACS developed at JAERI. The code validation results have been published in a series of JAERI-M reports and others. This report summarizes these results and the reliability of the code system systematically. The number of the calculated cases briefly described in this report together with their experimental systems and data are 502 for 17 kinds of homogeneous single-unit systems, 331 for 8 kinds of homogeneous multi-unit systems and 561 for 16 kinds of heterogeneous systems. Discussions and interpretations are made on the calculated keff's (neutron multiplication factors) with their bias errors. The factors related to the bias errors are confirmed together with their causes and trends. (author)
OECD/NEA burnup credit calculational criticality benchmark Phase I-B results
International Nuclear Information System (INIS)
In most countries, criticality analysis of LWR fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. This assumption has led to the design of widely spaced and/or highly poisoned storage and transport arrays. If credit is assumed for fuel burnup, initial enrichment limitations can be raised in existing systems, and more compact and economical arrays can be designed. Such reliance on the reduced reactivity of spent fuel for criticality control is referred to as burnup credit. The Burnup Credit Working Group, formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development, has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods agree to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods agree within 11% about the average for all fission products studied. Most deviations are less than 10%, and many are less than 5%. The exceptions are Sm 149, Sm 151, and Gd 155
International Nuclear Information System (INIS)
The report describes the final results of Phase IIIA Benchmarks conducted by the Burnup Credit Criticality Calculation Working Group under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA). The benchmarks are intended to confirm the predictive capability of the current computer code and data library combinations for the neutron multiplication factor (keff) of a layer of irradiated BWR fuel assembly array model. In total 22 benchmark problems are proposed for calculations of keff. The effects of following parameters are investigated: cooling time, inclusion/exclusion of FP nuclides and axial burnup profile, and inclusion of axial profile of void fraction or constant void fractions during burnup. Axial profiles of fractional fission rates are further requested for five cases out of the 22 problems. Twenty-one sets of results are presented, contributed by 17 institutes from 9 countries. The relative dispersion of keff values calculated by the participants from the mean value is almost within the band of ±1%Δk/k. The deviations from the averaged calculated fission rate profiles are found to be within ±5% for most cases. (author)
Benchmark data for validating irradiated fuel compositions used in criticality calculations
International Nuclear Information System (INIS)
To establish criticality safety margins utilizing burnup credit in the storage and transport of spent reactor fuels requires a knowledge of the uncertainty in the calculated fuel composition used in making the reactivity assessment. To provide data for validating such calculated burnup fuel compositions, radiochemical assays have been obtained as part of the United States Department of Energy From-Reactor Cask Development Program. Assay results and associated operating histories on the initial three samples analyzed in this effort are presented. The three samples were taken from different axial regions of a Pressurized Water Reactor fuel rod and represent radiation exposures of about 37, 27, and 44 GWd/MTU. The data are presented in a benchmark type format to facilitate identification/referencing and computer code input
OECD/NEA Burnup Credit Calculational Criticality Benchmark Phase I-B Results
International Nuclear Information System (INIS)
Burnup credit is an ongoing technical concern for many countries that operate commercial nuclear power reactors. In a multinational cooperative effort to resolve burnup credit issues, a Burnup Credit Working Group has been formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development. This working group has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide, and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods are in agreement to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods are within 11% agreement about the average for all fission products studied. Furthermore, most deviations are less than 10%, and many are less than 5%. The exceptions are 149Sm, 151Sm, and 155Gd
International Nuclear Information System (INIS)
A method for classifying benchmark results of criticality calculations according to similarity was proposed in this paper. After formulation of the method utilizing correlation coefficients, it was applied to burnup credit criticality benchmarks Phase III-A and II-A, which were conducted by the Expert Group on Burnup Credit Criticality Safety under auspices of the Nuclear Energy Agency of the Organisation for Economic Cooperation and Development (OECD/NEA). Phase III-A benchmark was a series of criticality calculations for irradiated Boiling Water Reactor (BWR) fuel assemblies, whereas Phase II-A benchmark was a suite of criticality calculations for irradiated Pressurized Water Reactor (PWR) fuel pins. These benchmark problems and their results were summarized. The correlation coefficients were calculated and sets of benchmark calculation results were classified according to the criterion that the values of the correlation coefficients were no less than 0.15 for Phase III-A and 0.10 for Phase II-A benchmarks. When a couple of benchmark calculation results belonged to the same group, one calculation result was found predictable from the other. An example was shown for each of the Benchmarks. While the evaluated nuclear data seemed the main factor for the classification, further investigations were required for finding other factors. (author)
International Nuclear Information System (INIS)
The MUSE project, carried out within the European fifth Framework Program, focuses on the coupling of a sub-critical reactor core with an external neutron source. In the first stage of the project a benchmark has been defined in order to define a reference calculational route, which is able to accurately predict the neutronics behavior in an accelerator driven system. Benchmark calculations will be carried out by several members of the project and the results will be compared, also with experimental results. The contribution of NRG to the project consists of the benchmark calculations and additional work that focuses on the calculation of 3D distributions of reaction yields. This paper discusses the non-conventional methods used to perform the benchmark calculations, including the 3D reaction yield distributions. The 3D distributions calculated for the sub-critical core will be Shown and discussed. With the ORANGE-extension to MCNP it is possible to tally 3D distributions, without adding extra cells and surfaces to the geometry and without a significant slowing down of the calculation. These are major advantages when compared to the conventional way of tallying in the MCNP-code. The distributions show details that can be understood in terms of the expected neutron behavior in the different parts of the geometry. For instance, the results show that: 1) a large number of fast neutrons is found in the fuel regions, 2) the reflector region shows an increased number of slower neutrons and 3) the reaction yield in the shielding region declines steeply. The extension therefore seems a useful tool in generating a better understanding of the behavior of neutrons throughout large and complex geometries like accelerator driven systems, but we also expect to use the extension in a variety of different fields. (authors)
International Nuclear Information System (INIS)
In this report we investigate the adequacy of the available 233U cross-section data for calculation of experimental critical systems. The 233U evaluations provided in two evaluated nuclear data libraries, the U.S. Data Bank [ENDF/B (Evaluated Nuclear Data Files)] and the Japanese Data Bank [JENDL (Japanese Evaluated Nuclear Data Library)] are examined. Calculations were performed for six thermal and ten fast experimental critical systems using the Sn transport XSDRNPM code. To verify the performance of the 233U cross-section data for nuclear criticality safety application in which the neutron energy spectrum is predominantly in the epithermal energy range, calculations of four numerical benchmark systems with energy spectra in the intermediate energy range were done. These calculations serve only as an indication of the difference in calculated results that may be expected when the two 233U cross-section evaluations are used for problems with neutron spectra in the intermediate energy range. Additionally, comparisons of experimental and calculated central fission rate ratios were also made. The study has suggested that an ad hoc 233U evaluation based on the JENDL library provides better overall results for both fast and thermal experimental critical systems
Energy Technology Data Exchange (ETDEWEB)
Leal, L.C.
1993-01-01
In this report we investigate the adequacy of the available {sup 233}U cross-section data for calculation of experimental critical systems. The {sup 233}U evaluations provided in two evaluated nuclear data libraries, the U. S. Data Bank [ENDF/B (Evaluated Nuclear Data Files)] and the Japanese Data Bank [JENDL (Japanese Evaluated Nuclear Data Library)] are examined. Calculations were performed for six thermal and ten fast experimental critical systems using the Sn transport XSDRNPM code. To verify the performance of the {sup 233}U cross-section data for nuclear criticality safety application in which the neutron energy spectrum is predominantly in the epithermal energy range, calculations of four numerical benchmark systems with energy spectra in the intermediate energy range were done. These calculations serve only as an indication of the difference in calculated results that may be expected when the two {sup 233}U cross-section evaluations are used for problems with neutron spectra in the intermediate energy range. Additionally, comparisons of experimental and calculated central fission rate ratios were also made. The study has suggested that an ad hoc {sup 233}U evaluation based on the JENDL library provides better overall results for both fast and thermal experimental critical systems.
Evaluation of CRISTO II storage arrays benchmark with TRIPOLI-4.2 criticality calculations
International Nuclear Information System (INIS)
The new lattice feature of TRIPOLI-4.2 geometry package was applied to model the CRISTO II storage arrays of PWR fuels with various kinds of neutron absorber plates. The new 'Kcoll' collision estimator of TRIPOLI-4.2 code was utilized to evaluate the infinite multiplication factors, Kinf. Comparing with the published ICS-BEP benchmark results of CRISTO II experiments and of three different continuous-energy Monte Carlo codes - TRIPOLI-4.1 (JEF2.2), MCNP4B2 (ENDF/B-V) and MCNP4XS (ENDF/B-VI.r4), the present study using cost-effective modeling, JEF2.2 and ENDF/B-VI.r4 libraries obtained satisfactory results. (orig.)
International Nuclear Information System (INIS)
The report describes the final results of the Phase IIIB Benchmark conducted by the Expert Group on Burnup Credit Criticality Safety under the auspices of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD). The Benchmark was intended to compare the predictability of current computer code and data library combinations for the atomic number densities of an irradiated PWR fuel assembly model. The fuel assembly was irradiated under specific power of 25.6 MW/tHM up to 40 GWd/tHM and cooled for five years. The void fraction was assumed to be uniform throughout the channel box and constant, at 0, 40 and 70%, during burnup. In total, 16 results were submitted from 13 institutes of 7 countries. The calculated atomic number densities of 12 actinides and 20 fission product nuclides were found to be for the most part within a range of ±10% relative to the average, although some results, esp. 155Eu and gadolinium isotopes, exceeded the band, which will require further investigation. Pin-wise burnup results agreed well among the participants. The results in the infinite neutron multiplication factor k∞ also accorded well with each other for void fractions of 0 and 40%; however some results deviated from the averaged value noticeably for the void fraction of 70%. (author)
Energy Technology Data Exchange (ETDEWEB)
Okuno, Hiroshi; Naito, Yoshitaka; Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2002-02-01
The report describes the final results of the Phase IIIB Benchmark conducted by the Expert Group on Burnup Credit Criticality Safety under the auspices of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD). The Benchmark was intended to compare the predictability of current computer code and data library combinations for the atomic number densities of an irradiated PWR fuel assembly model. The fuel assembly was irradiated under specific power of 25.6 MW/tHM up to 40 GWd/tHM and cooled for five years. The void fraction was assumed to be uniform throughout the channel box and constant, at 0, 40 and 70%, during burnup. In total, 16 results were submitted from 13 institutes of 7 countries. The calculated atomic number densities of 12 actinides and 20 fission product nuclides were found to be for the most part within a range of {+-}10% relative to the average, although some results, esp. {sup 155}Eu and gadolinium isotopes, exceeded the band, which will require further investigation. Pin-wise burnup results agreed well among the participants. The results in the infinite neutron multiplication factor k{sub {infinity}} also accorded well with each other for void fractions of 0 and 40%; however some results deviated from the averaged value noticeably for the void fraction of 70%. (author)
Selecting benchmarks for reactor calculations
Alhassan, Erwin; Sjöstrand, Henrik; Duan, Junfeng; Helgesson, Petter; Pomp, Stephan; Österlund, Michael; Rochman, Dimitri; Koning, Arjan J.
2014-01-01
Criticality, reactor physics, fusion and shielding benchmarks are expected to play important roles in GENIV design, safety analysis and in the validation of analytical tools used to design these reactors. For existing reactor technology, benchmarks are used to validate computer codes and test nuclear data libraries. However the selection of these benchmarks are usually done by visual inspection which is dependent on the expertise and the experience of the user and there by resulting in a user...
Selecting benchmarks for reactor calculations
International Nuclear Information System (INIS)
Criticality, reactor physics, fusion and shielding benchmarks are expected to play important roles in GENIV design, safety analysis and in the validation of analytical tools used to design these reactors. For existing reactor technology, benchmarks are used to validate computer codes and test nuclear data libraries. However the selection of these benchmarks are usually done by visual inspection which is dependent on the expertise and the experience of the user and thereby resulting in a user bias in the process. In this paper we present a method for the selection of these benchmarks for reactor applications and uncertainty reduction based on Total Monte Carlo (TMC) method. Similarities between an application case and one or several benchmarks are quantified using the correlation coefficient. Based on the method, we also propose two approaches for reducing nuclear data uncertainty using integral benchmark experiments as an additional constrain in the TMC method: a binary accept/reject method and a method of uncertainty reduction using weights. Finally, the methods were applied to a full Lead Fast Reactor core and a set of criticality benchmarks. (author)
Benchmark calculations on simple reactor systems
International Nuclear Information System (INIS)
The development of some calculation methods is described. Tests of these and other methods on benchmark problems are reported. The following items are treated: 1) Criticality of spheres and slabs for monoenergetic neutrons with Carlviks method. 2) High precision S sub (n) calculations on critical slabs. 3) Comparison of angular quadrature methods in S sub (n) calculations. 4) Tests of a standard ANISN program. 5) Presence of complex time eigenvalues in a fundamental problem. (Author)
International Nuclear Information System (INIS)
The DECD/NEA Expert Group on Burn-up Credit was established in 1991 to address scientific and technical issues connected with the use of burn-up credit in nuclear fuel cycle operations. Following the completion of six benchmark exercises with uranium oxide (UOX) fuels irradiated in pressurised water reactors (PWRs) and boiling water reactors (BWRs), the present report concerns mixed uranium and plutonium oxide (MOX) fuels irradiated in PWRs. The exercises consisted of inventory calculations of MOX fuels for two initial plutonium compositions. The depletion calculations were carried out using three representations of the MOX assemblies and their interface with UOX assemblies. This enabled the investigation of the spatial and spectral effects during the irradiation of the MOX fuels. (author)
EPRI depletion benchmark calculations using PARAGON
International Nuclear Information System (INIS)
Highlights: • PARAGON depletion calculations are benchmarked against the EPRI reactivity decrement experiments. • Benchmarks cover a wide range of enrichments, burnups, cooling times, and burnable absorbers, and different depletion and storage conditions. • Results from PARAGON-SCALE scheme are more conservative relative to the benchmark data. • ENDF/B-VII based data reduces the excess conservatism and brings the predictions closer to benchmark reactivity decrement values. - Abstract: In order to conservatively apply burnup credit in spent fuel pool criticality analyses, code validation for both fresh and used fuel is required. Fresh fuel validation is typically done by modeling experiments from the “International Handbook.” A depletion validation can determine a bias and bias uncertainty for the worth of the isotopes not found in the fresh fuel critical experiments. Westinghouse’s burnup credit methodology uses PARAGON™ (Westinghouse 2-D lattice physics code) and its 70-group cross-section library, which have been benchmarked, qualified, and licensed both as a standalone transport code and as a nuclear data source for core design simulations. A bias and bias uncertainty for the worth of depletion isotopes, however, are not available for PARAGON. Instead, the 5% decrement approach for depletion uncertainty is used, as set forth in the Kopp memo. Recently, EPRI developed a set of benchmarks based on a large set of power distribution measurements to ascertain reactivity biases. The depletion reactivity has been used to create 11 benchmark cases for 10, 20, 30, 40, 50, and 60 GWd/MTU and 3 cooling times 100 h, 5 years, and 15 years. These benchmark cases are analyzed with PARAGON and the SCALE package and sensitivity studies are performed using different cross-section libraries based on ENDF/B-VI.3 and ENDF/B-VII data to assess that the 5% decrement approach is conservative for determining depletion uncertainty
Benchmark calculations for EGS5
International Nuclear Information System (INIS)
In the past few years, EGS4 has undergone an extensive upgrade to EGS5, in particularly in the areas of low-energy electron physics, low-energy photon physics, PEGS cross section generation, and the coding from Mortran to Fortran programming. Benchmark calculations have been made to assure the accuracy, reliability and high quality of the EGS5 code system. This study reports three benchmark examples that show the successful upgrade from EGS4 to EGS5 based on the excellent agreements among EGS4, EGS5 and measurements. The first benchmark example is the 1969 Crannell Experiment to measure the three-dimensional distribution of energy deposition for 1-GeV electrons shower in water and aluminum tanks. The second example is the 1995 Compton-scattered spectra measurements for 20-40 keV, linearly polarized photon by Namito et. al., in KEK, which was a main part of the low-energy photon expansion work for both EGS4 and EGS5. The third example is the 1986 heterogeneity benchmark experiment by Shortt et. al., who used a monoenergetic 20-MeV electron beam to hit the front face of a water tank containing both air and aluminum cylinders and measured spatial depth dose distribution using a small solid-state detector. (author)
KENO-IV code benchmark calculation, (6)
International Nuclear Information System (INIS)
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(NO3)4 aqueous solution, Pu metal or PuO2-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)
Introduction to 'International Handbook of Criticality Safety Benchmark Experiments'
International Nuclear Information System (INIS)
The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in 1992 by the United States Department of Energy. The project quickly became an international effort as scientists from other interested countries became involved. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) is now an official activity of the Organization for Economic Cooperation and Development-Nuclear Energy Agency (OECD-NEA). 'International Handbook of Criticality Safety Benchmark Experiments' was prepared and is updated year by year by the working group of the project. This handbook contains criticality safety benchmark specifications that have been derived from experiments that were performed at various nuclear critical facilities around the world. The benchmark specifications are intended for use by criticality safety engineers to validate calculation techniques used. The author briefly introduces the informative handbook and would like to encourage Japanese engineers who are in charge of nuclear criticality safety to use the handbook. (author)
HEU benchmark calculations and LEU preliminary calculations for IRR-1
International Nuclear Information System (INIS)
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)
PHEBUS-FPTO Benchmark calculations
International Nuclear Information System (INIS)
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
KENO-IV code benchmark calculation, (4)
International Nuclear Information System (INIS)
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 multi-group constants library MGCL. The present paper describes the results of a test using criticality experiments about slab-cylinder system of uranium nitrate solution. In all, 128 cases of experiments have been calculated for the slab-cylinder configuration with and without plexiglass reflector, having the various critical parameters such as the number of cylinders and height of the uranium nitrate solution. It is shown among several important results that the code and library gives a fairly good multiplication factor, that is, k sub(eff) -- 1.0 for heavily reflected cases, whereas k sub(eff) -- 0.91 for the unreflected ones. This suggests the necessity of more advanced treatment of the criticality calculation for the system where neutrons can easily leak out during slowing down process. (author)
One dimensional benchmark calculations using diffusion theory
International Nuclear Information System (INIS)
This is a comparative study by using different one dimensional diffusion codes which are available at our Nuclear Engineering Department. Some modifications have been made in the used codes to fit the problems. One of the codes, DIFFUSE, solves the neutron diffusion equation in slab, cylindrical and spherical geometries by using 'Forward elimination- Backward substitution' technique. DIFFUSE code calculates criticality, critical dimensions and critical material concentrations and adjoint fluxes as well. It is used for the space and energy dependent neutron flux distribution. The whole scattering matrix can be used if desired. Normalisation of the relative flux distributions to the reactor power, plotting of the flux distributions and leakage terms for the other two dimensions have been added. Some modifications also have been made for the code output. Two Benchmark problems have been calculated with the modified version and the results are compared with BBD code which is available at our department and uses same techniques of calculation. Agreements are quite good in results such as k-eff and the flux distributions for the two cases studies. (author)
OECD/NEA Burnup Credit Criticality Benchmark
International Nuclear Information System (INIS)
The report describes the final result of the phase-1A of the Burnup Credit Criticality Benchmark conducted by OECD/NEA. The phase-1A benchmark problem is an infinite array of a simple PWR spent fuel rod. The analysis has been performed for the PWR spent fuels of 30 and 40 GWd/t after 1 and 5 years of cooling time. In total, 25 results from 19 institutes of 11 countries have been submitted. For the nuclides in spent fuel, 7 major actinides and 15 major fission products (FP) are selected for the benchmark calculation. In the case of 30 GWd/t burnup, it is found that the major actinides and the major FPs contribute more than 50% and 30% of the total reactivity loss due to burnup, respectively. Therefore, more than 80% of the reactivity loss can be covered by 22 nuclides. However, the larger deviation among the reactivity losses by participants has been found for cases including EPs than the cases with only actinides, indicating the existence of relatively large uncertainties in FP cross sections. The large deviation seen also in the case of the fresh fuel has been found to reduce sufficiently by replacing the cross section library from ENDF-B/IV with that from ENDF-B/V and taking the known bias of MONK6 into account. (author)
''FULL-CORE'' VVER-440 calculation benchmark
International Nuclear Information System (INIS)
Because of the difficulties with experimental validation of power distribution predicted by macro-code on the pin by pin level we decided to prepare a calculation benchmark named ''FULL-CORE'' VVER-440. This benchmark is a two-dimensional (2D) calculation benchmark based on the VVER-440 reactor core cold state geometry with taking into account the geometry of explicit radial reflector. The main task of this benchmark is to test the pin by pin power distribution in fuel assemblies predicted by macro-codes that are used for neutron-physics calculations especially for VVER-440 reactors. The proposal of this benchmark was presented at the 21st Symposium of AER in 2011. The reference solution has been calculated by MCNP code using Monte Carlo method and the results have been published in the AER community. The results of reference calculation were presented at the 22nd Symposium of AER in 2012. In this paper we will compare the available macro-codes results of this calculation benchmark.
International Criticality Safety Benchmark Evaluation Project (ICSBEP) - ICSBEP 2015 Handbook
International Nuclear Information System (INIS)
The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in October of 1992 by the United States Department of Energy (DOE). The project quickly became an international effort as scientists from other interested countries became involved. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) became an official activity of the Nuclear Energy Agency (NEA) in 1995. This handbook contains criticality safety benchmark specifications that have been derived from experiments performed at various critical facilities around the world. The benchmark specifications are intended for use by criticality safety engineers to validate calculation techniques used to establish minimum subcritical margins for operations with fissile material and to determine criticality alarm requirements and placement. Many of the specifications are also useful for nuclear data testing. Example calculations are presented; however, these calculations do not constitute a validation of the codes or cross-section data. The evaluated criticality safety benchmark data are given in nine volumes. These volumes span approximately 69000 pages and contain 567 evaluations with benchmark specifications for 4874 critical, near-critical or subcritical configurations, 31 criticality alarm placement/shielding configurations with multiple dose points for each, and 207 configurations that have been categorised as fundamental physics measurements that are relevant to criticality safety applications. New to the handbook are benchmark specifications for neutron activation foil and thermoluminescent dosimeter measurements performed at the SILENE critical assembly in Valduc, France as part of a joint venture in 2010 between the US DOE and the French Alternative Energies and Atomic Energy Commission (CEA). A photograph of this experiment is shown on the front cover. Experiments that are found unacceptable for use as criticality safety benchmark experiments are discussed in these
The International Criticality Safety Benchmark Evaluation Project (ICSBEP)
International Nuclear Information System (INIS)
The International Criticality Safety Benchmark Evaluation Project (ICSBEP) was initiated in 1992 by the United States Department of Energy. The ICSBEP became an official activity of the Organisation for Economic Cooperation and Development (OECD) - Nuclear Energy Agency (NEA) in 1995. Representatives from the United States, United Kingdom, France, Japan, the Russian Federation, Hungary, Republic of Korea, Slovenia, Yugoslavia, Kazakhstan, Israel, Spain, and Brazil are now participating. The purpose of the ICSBEP is to identify, evaluate, verify, and formally document a comprehensive and internationally peer-reviewed set of criticality safety benchmark data. The work of the ICSBEP is published as an OECD handbook entitled 'International Handbook of Evaluated Criticality Safety Benchmark Experiments.' The 2003 Edition of the Handbook contains benchmark model specifications for 3070 critical or subcritical configurations that are intended for validating computer codes that calculate effective neutron multiplication and for testing basic nuclear data. (author)
Benchmark calculations of power distribution within assemblies
International Nuclear Information System (INIS)
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 (Pij, Sn, Monte Carlo). This report presents an analysis and intercomparisons of all the results received
Reactor calculation benchmark PCA blind test results
International Nuclear Information System (INIS)
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
Two benchmarks for qualification of pressure vessel fluence calculational methodology
International Nuclear Information System (INIS)
Two benchmarks for the qualification of the pressure vessel fluence calculational methodology were formulated and are briefly described. The Pool Critical Assembly (PCA) benchmark is based on the experiments performed at the PCA in Oak Ridge. The measured quantities to be compared against the calculated values are the equivalent fission fluxes at several locations in front, behind, and inside the pressure-vessel wall simulator. This benchmark is particularly suitable to test the capabilities of the calculational methodology and cross-section libraries to predict in-vessel gradients because only a few approximations are necessary in the analysis. The HBR-2 benchmark is based on the data for the H.B. Robinson-2 plant, which is a 2,300 MW (thermal) pressurized light-water reactor. The benchmark provides the reactor geometry, the material compositions, the core power distributions, and the power historical data. The quantities to be calculated are the specific activities of the radiometric monitors that were irradiated in the surveillance capsule and in the cavity location during one fuel cycle. The HBR-2 benchmark requires modeling approximations, power-to-neutron source conversion, and treatment of time dependant variations. It can therefore be used to test the overall performance and adequacy of the calculational methodology for power-reactor pressure-vessel flux calculations. Both benchmarks were analyzed with the DORT code and the BUGLE-96 cross-section library that is based on ENDF/B-VI evaluations. The calculations agreed with the measurements within 10%, and the calculations underpredicted the measurements in all the cases. This indicates that the ENDF/B-VI cross sections resolve most of the discrepancies between the measurements and calculations. The decrease of the CIM ratios with increased thickness of iron, which was typical for pre-ENDF/B-VI libraries, is almost completely removed
The MCNP6 Analytic Criticality Benchmark Suite
Energy Technology Data Exchange (ETDEWEB)
Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Codes Group
2016-06-16
Analytical benchmarks provide an invaluable tool for verifying computer codes used to simulate neutron transport. Several collections of analytical benchmark problems [1-4] are used routinely in the verification of production Monte Carlo codes such as MCNP® [5,6]. Verification of a computer code is a necessary prerequisite to the more complex validation process. The verification process confirms that a code performs its intended functions correctly. The validation process involves determining the absolute accuracy of code results vs. nature. In typical validations, results are computed for a set of benchmark experiments using a particular methodology (code, cross-section data with uncertainties, and modeling) and compared to the measured results from the set of benchmark experiments. The validation process determines bias, bias uncertainty, and possibly additional margins. Verification is generally performed by the code developers, while validation is generally performed by code users for a particular application space. The VERIFICATION_KEFF suite of criticality problems [1,2] was originally a set of 75 criticality problems found in the literature for which exact analytical solutions are available. Even though the spatial and energy detail is necessarily limited in analytical benchmarks, typically to a few regions or energy groups, the exact solutions obtained can be used to verify that the basic algorithms, mathematics, and methods used in complex production codes perform correctly. The present work has focused on revisiting this benchmark suite. A thorough review of the problems resulted in discarding some of them as not suitable for MCNP benchmarking. For the remaining problems, many of them were reformulated to permit execution in either multigroup mode or in the normal continuous-energy mode for MCNP. Execution of the benchmarks in continuous-energy mode provides a significant advance to MCNP verification methods.
International handbook of evaluated criticality safety benchmark experiments
International Nuclear Information System (INIS)
The primary purpose of the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Working Group is to compile critical and subcritical benchmark experiment data into a standardised format that allows criticality safety analysts to easily use the data to validate calculation tools and cross-section libraries. ICSBEP work includes: - Identifying a comprehensive set of critical benchmark data and, to the extent possible, verify the data by reviewing original and subsequently revised documentation, and by talking with the experimenters or individuals who are familiar with the experimenters or the experimental facility; - Evaluating the data and quantify overall uncertainties through various types of sensitivity analysis; - Compiling the data into a standardised format; - Performing calculations of each experiment with standard criticality safety codes; - Formally documenting the work into a single source of verified benchmark critical data. The work of the ICSBEP is documented as an International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook). Currently, the handbook spans nearly 67,000 pages and contains 561 evaluations representing 4839 critical, near-critical, or subcritical configurations, 24 criticality alarm placement/shielding configurations with multiple dose points for each, and 207 configurations that have been categorised as fundamental physics measurements that are relevant to criticality safety applications. The handbook is intended for use by criticality safety analysts to perform necessary validations of their calculational techniques and is expected to be a valuable tool for decades to come. The ICSBEP Handbook is produced in electronic format (pdf files) where the experiments are grouped into evaluations and categorised by: fissile media (plutonium, highly enriched uranium, intermediate and mixed enrichment uranium, low enriched uranium, uranium-233, mixed plutonium-uranium and special isotope systems
Benchmark calculations for fusion blanket development
International Nuclear Information System (INIS)
Benchmark problems representing the leading fusion blanket concepts are presented. Benchmark calculations for self-cooled Li17Pb83 and helium-cooled blankets were performed. Multigroup data libraries generated from ENDF/B-IV and V files using the NJOY and AMPX processing codes with different weighting functions were used. The sensitivity of the tritium breeding ratio to group structure and weighting spectrum increases as the thickness and Li enrichment decrease with up to 20% discrepancies for thin natural Li17Pb83 blankets. (author)
International handbook of evaluated criticality safety benchmark experiments
International Nuclear Information System (INIS)
The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in October of 1992 by the United States Department of Energy. The project quickly became an international effort as scientists from other interested countries became involved. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) became an official activity of the Organization for Economic Cooperation and Development - Nuclear Energy Agency (OECD-NEA) in 1995. This handbook contains criticality safety benchmark specifications that have been derived from experiments performed at various nuclear critical facilities around the world. The benchmark specifications are intended for use by criticality safety engineers to validate calculational techniques used to establish minimum subcritical margins for operations with fissile material and to determine criticality alarm requirement and placement. Many of the specifications are also useful for nuclear data testing. Example calculations are presented; however, these calculations do not constitute a validation of the codes or cross section data. The evaluated criticality safety benchmark data are given in nine volumes. These volumes span over 55,000 pages and contain 516 evaluations with benchmark specifications for 4,405 critical, near critical, or subcritical configurations, 24 criticality alarm placement / shielding configurations with multiple dose points for each, and 200 configurations that have been categorized as fundamental physics measurements that are relevant to criticality safety applications. Experiments that are found unacceptable for use as criticality safety benchmark experiments are discussed in these evaluations; however, benchmark specifications are not derived for such experiments (in some cases models are provided in an appendix). Approximately 770 experimental configurations are categorized as unacceptable for use as criticality safety benchmark experiments. Additional evaluations are in progress and will be
Testing of 233U evaluations with criticality benchmarks
International Nuclear Information System (INIS)
To validate and improve the quality of the complete set of evaluated nuclear reaction data for 233U, criticality benchmarks with fast, epithermal and thermal spectra from ICSBEP handbook were selected to test 233U evaluations from CENDL-3.1, ENDF/B-Ⅶ.0, JENDL-3.3 and JENDL-4.0. The effective multiplication factors keff of selected benchmarks were calculated with the Monte Carlo code MCNP5 and compared with the benchmark values. The results were analyzed with trend against energy spectrum index and sensitivity analysis. In present validation, the underestimation of keff for benchmarks with thermal, epithermal or some of fast spectra is the main problem existed in the tested evaluations. From the view of thermal reactors design, the 233U evaluation from ENDF/B-Ⅶ.0 shows better performance than other file tested, but still overestimates the contribution of capture reaction in resonance region. (authors)
FENDL-2 and associated benchmark calculations
International Nuclear Information System (INIS)
The present Report contains the Summary of the IAEA Advisory Group Meeting on ''The FENDL-2 and Associated Benchmark Calculations'' convened on 18-22 November 1991, at the IAEA Headquarters in Vienna, Austria, by the IAEA Nuclear Data Section. The Advisory Group Meeting Conclusions and Recommendations and the Report on the Strategy for the Future Development of the FENDL and on Future Work towards establishing FENDL-2 are also included in this Summary Report. (author). 1 ref., 4 tabs
Benchmark calculations for MTR type cores
International Nuclear Information System (INIS)
The benchmark neutronies design study of MTR cores has been performed for various fuel enrichments. The reactivities and fluxes for fresh core have been evaluated. The reference calculations have been performed for a 10MW(th) reactor but the method is applicable to other power levels. As the results are in good agreement with those obtained at other establishments, the method of analysis used in this report for a fresh core can be relied upon with a fair amount of confidence. (authors)
Benchmark testing calculations for 232Th
International Nuclear Information System (INIS)
The cross sections of 232Th from CNDC and JENDL-3.3 were processed with NJOY97.45 code in the ACE format for the continuous-energy Monte Carlo Code MCNP4C. The Keff values and central reaction rates based on CENDL-3.0, JENDL-3.3 and ENDF/B-6.2 were calculated using MCNP4C code for benchmark assembly, and the comparisons with experimental results are given. (author)
International Nuclear Information System (INIS)
The OECD/NEA Expert Group on Burn-up Credit was established in 1991 to address scientific and technical issues connected with the use of burn-up credit in nuclear fuel cycle operations. Following the completion of six benchmark exercises with uranium oxide fuels irradiated in pressurised water reactors (PWRs) and boiling water reactors (BWRs), the present report concerns mixed uranium and plutonium oxide (MOX) fuels irradiated in PWRs. The report summarises and analyses the solutions to the specified exercises provided by 37 contributors from 10 countries. The exercises were based upon the calculation of infinite PWR fuel pin cell reactivity for fresh and irradiated MOX fuels with various MOX compositions, burn-ups and cooling times. In addition, several representations of the MOX fuel assembly were tested in order to check various levels of approximations commonly used in reactor physics calculations. (authors)
International Nuclear Information System (INIS)
Continuous-energy Monte Carlo eigenvalue calculations have been performed for a selection of HEU-MET-FAST, IEU-MET-FAST, HEU-SOL-THERM, LEU-COMP-THERM, and LEU-SOL-THERM benchmarks using ENDF/B (primarily VI.8), JEFF-3.0, and JENDL-3.3 cross sections. These benchmarks allow for testing the cross-section data for both common reactor nuclides such as 1H, 16O, and 235,238U and structural and shielding elements such as Al, Ti, Fe, Ni, and Pb. The latest cross-section libraries yield near-unity eigenvalues for unreflected or water-reflected HEU-SOL-THERM and LEU-SOL-THERM systems. Near-unity eigenvalues are also obtained for bare HEU-MET-FAST and IEU-MET-FAST systems, but small deviations from unity are observed in both FAST and THERM benchmarks as a function of nonhydrogenous reflector material and thickness. The long-standing problem of lower eigenvalues in water-reflected low-enriched-uranium fuel lattice systems remains, regardless of cross-section library
42 CFR 422.258 - Calculation of benchmarks.
2010-10-01
... 42 Public Health 3 2010-10-01 2010-10-01 false Calculation of benchmarks. 422.258 Section 422.258... and Plan Approval § 422.258 Calculation of benchmarks. (a) The term “MA area-specific non-drug monthly... the plan bids. (c) Calculation of MA regional non-drug benchmark amount. CMS calculates the...
BEGAFIP. Programming service, development and benchmark calculations
International Nuclear Information System (INIS)
This report summarizes improvements to BEGAFIP (the Swedish equivalent to the Oak Ridge computer code ORIGEN). The improvements are: addition of a subroutine making it possible to calculate neutron sources, exchange of fission yields and branching ratios in the data library to those published by Meek and Rider in 1978. In addition, BENCHMARK-calculations have been made with BEGAFIP as well as with ORIGEN regarding the build-up of actinides for a fuel burnup of 33 MWd/kg U. The results were compared to those arrived upon from the more sophisticated code CASMO. (author)
COVE 2A Benchmarking calculations using NORIA
International Nuclear Information System (INIS)
Six steady-state and six transient benchmarking calculations have been performed, using the finite element code NORIA, to simulate one-dimensional infiltration into Yucca Mountain. These calculations were made to support the code verification (COVE 2A) activity for the Yucca Mountain Site Characterization Project. COVE 2A evaluates the usefulness of numerical codes for analyzing the hydrology of the potential Yucca Mountain site. Numerical solutions for all cases were found to be stable. As expected, the difficulties and computer-time requirements associated with obtaining solutions increased with infiltration rate. 10 refs., 128 figs., 5 tabs
Critical benchmark results for a modified 16O evaluation
International Nuclear Information System (INIS)
The effect of a uniform reduction in the elastic scattering cross-section for 16O on critical benchmarks is quantified and discussed. It is hypothesised that current evaluations for 16O systematically overestimate elastic scattering by about 3% due to a normalisation error in various experimental data. Selected critical benchmarks from the HEU-SOL-THERM (HST) series of the International Handbook of Evaluated Criticality Safety Benchmark Experiments were simulated using the MC21 Monte Carlo code. The benchmark results show that a decrease in the elastic scattering cross-section to agree with high-precision experimental measurements leads to higher leakage and lower benchmark eigenvalues. Additionally, a trend with the above-thermal leakage fraction was observed. The sensitivity of this trend to the first Legendre polynomial coefficient of the elastic scattering angular distribution was calculated. Based on the observed sensitivity, a 35% decrease in the first-order Legendre polynomial coefficient would be required to eliminate the trend with above-thermal leakage fraction. (authors)
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.).
International Nuclear Information System (INIS)
The reactivity effect of the asymmetry of axial burnup profile in burnup credit criticality safety is studied for a realistic PWR spent fuel transport cask proposed in the current OECD/NEA Phase II-C benchmark problem. The axial burnup profiles are simulated in 21 material zones based on in-core flux measurements varying from strong asymmetry to more or less no asymmetry. Criticality calculations in a 3-D model have been performed using the continuous energy Monte Carlo code MCNP-4B2 and the nuclear data library JENDL-3.2. Calculation conditions are determined with consideration of the axial fission source convergence. Calculations are carried out not only for cases proposed in the benchmark but also for additional cases assuming symmetric burnup profile. The actinide-only approach supposed for first domestic introduction of burnup credit into criticality evaluation is also considered in addition to the actinide plus fission product approach adopted in the benchmark. The calculated results show that keff and the end effect increase almost linearly with increasing burnup axial offset that is defined as one of typical parameters showing the intensity of axial burnup asymmetry. The end effect is more sensitive to the asymmetry of burnup profile for the higher burnup. For an axially distributed burnup, the axial fission source distribution becomes strongly asymmetric as its peak shifts toward the top end of the fuel's active zone where the local burnup is less than that of the bottom end. The peak of fission source distribution becomes higher with the increase of either the asymmetry of burnup profile or the assembly-averaged burnup. The conservatism of the assumption of uniform axial burnup based on the actinide-only approach is estimated quantitatively in comparison with the keff result calculated with experiment-based strongest asymmetric axial burnup profile with the actinide plus fission product approach. (author)
Benchmark analysis of KRITZ-2 critical experiments
International Nuclear Information System (INIS)
In the KRITZ-2 critical experiments, criticality and pin power distributions were measured at room temperature and high temperature (about 245 degC) for three different cores (KRITZ-2:1, KRITZ-2:13, KRITZ-2:19) loading slightly enriched UO2 or MOX fuels. Recently, international benchmark problems were provided by ORNL and OECD/NEA based on the KRITZ-2 experimental data. The published experimental data for the system with slightly enriched fuels at high temperature are rare in the world and they are valuable for nuclear data testing. Thus, the benchmark analysis was carried out with a continuous-energy Monte Carlo code MVP and its four nuclear data libraries based on JENDL-3.2, JENDL-3.3, JEF-2.2 and ENDF/B-VI.8. As a result, fairly good agreements with the experimental data were obtained with any libraries for the pin power distributions. However, the JENDL-3.3 and ENDF/B-VI.8 give under-prediction of criticality and too negative isothermal temperature coefficients for slightly enriched UO2 cores, although the older nuclear data JENDL-3.2 and JEF-2.2 give rather good agreements with the experimental data. From the detailed study with an infinite unit cell model, it was found that the differences among the results with different libraries are mainly due to the different fission cross section of U-235 in the energy range below 1.0 eV. (author)
LCEs for Naval Reactor Benchmark Calculations
International Nuclear Information System (INIS)
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 (keff) 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
LCEs for Naval Reactor Benchmark Calculations
Energy Technology Data Exchange (ETDEWEB)
W.J. Anderson
1999-07-19
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{sub 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.
International Nuclear Information System (INIS)
Different evaluated (n,d) energy-angle elastic scattering distributions produce k-effective differences in MCNP5 simulations of critical experiments involving heavy water (D2O) of sufficient magnitude to suggest a need for new (n,d) scattering measurements and/or distributions derived from modern theoretical nuclear models, especially at neutron energies below a few MeV. The present work focuses on the small reactivity change of 2O coolant-void-reactivity calculation bias for simulations of two pairs of critical experiments performed in the ZED-2 reactor at the Chalk River Laboratories when different nuclear data libraries are used for deuterium. The deuterium data libraries tested include Endf/B-VII.0, Endf/B-VI.4, JENDL-3.3 and a new evaluation, labelled Bonn-B, which is based on recent theoretical nuclear-model calculations. Comparison calculations were also performed for a simplified, two-region, spherical model having an inner, 250-cm radius, homogeneous sphere of UO2, without and with deuterium, and an outer 20-cm-thick deuterium reflector. A notable observation from this work is the reduction of about 0.4 mk in the MCNP5 ZED-2 CVR calculation bias that is obtained when the O-in-UO2 thermal scattering data comes from Endf-B-VII.0. (author)
Status of the international criticality safety benchmark evaluation project (ICSBEP)
International Nuclear Information System (INIS)
Since ICNC'99, four new editions of the International Handbook of Evaluated Criticality Safety Benchmark Experiments have been published. The number of benchmark specifications in the Handbook has grown from 2157 in 1999 to 3073 in 2003, an increase of nearly 1000 specifications. These benchmarks are used to validate neutronics codes and nuclear cross-section data. Twenty evaluations representing 192 benchmark specifications were added to the Handbook in 2003. The status of the International Criticality Safety Benchmark Evaluation Project (ICSBEP) is provided in this paper along with a summary of the newly added benchmark specifications that appear in the 2003 Edition of the Handbook. (author)
Benchmarking Calculations of Excitonic Couplings between Bacteriochlorophylls.
Kenny, Elise P; Kassal, Ivan
2016-01-14
Excitonic couplings between (bacterio)chlorophyll molecules are necessary for simulating energy transport in photosynthetic complexes. Many techniques for calculating the couplings are in use, from the simple (but inaccurate) point-dipole approximation to fully quantum-chemical methods. We compared several approximations to determine their range of applicability, noting that the propagation of experimental uncertainties poses a fundamental limit on the achievable accuracy. In particular, the uncertainty in crystallographic coordinates yields an uncertainty of about 20% in the calculated couplings. Because quantum-chemical corrections are smaller than 20% in most biologically relevant cases, their considerable computational cost is rarely justified. We therefore recommend the electrostatic TrEsp method across the entire range of molecular separations and orientations because its cost is minimal and it generally agrees with quantum-chemical calculations to better than the geometric uncertainty. Understanding these uncertainties can guard against striving for unrealistic precision; at the same time, detailed benchmarks can allow important qualitative questions-which do not depend on the precise values of the simulation parameters-to be addressed with greater confidence about the conclusions. PMID:26651217
OECD/Nea benchmark calculations for accelerator driven systems
International Nuclear Information System (INIS)
In order to evaluate the performances of the codes and the nuclear data, the Nuclear Science Committee of the OECD/NEA organised in July 1999 a benchmark exercise on a lead-bismuth cooled sub-critical system driven by a beam of 1 GeV protons. The benchmark model is based on the ALMR reference design and is optimised to burn minor actinides using a 'double strata' fuel cycle strategy. Seven organisations (ANL, CIEMAT, KAERI, JAERI, PSI/CEA, RIT and SCK-CEN) have contributed to this exercise using different basic data libraries (ENDF/B-VI, JEF-2.2 and JENDL-3.2) and various reactor calculation methods. Significant discrepancies are observed in important neutronic parameters, such as keff, reactivity swing with burn-up and neutron flux distributions. (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 ...
International Nuclear Information System (INIS)
In year 2008 the Atomic Energy National Commission (CNEA) of Argentina, and the Brazilian Institute of Energetic and Nuclear Research (IPEN), under the frame of Nuclear Energy Argentine Brazilian Agreement (COBEN), among many others, included the project “Validation and Verification of Calculation Methods used for Research and Experimental Reactors. At this time, it was established that the validation was to be performed with models implemented in the deterministic codes HUEMUL and PUMA (cell and reactor codes) developed by CNEA and those ones implemented in MCNP by CNEA and IPEN. The necessary data for these validations would correspond to theoretical-experimental reference cases in the research reactor IPEN/MB-01 located in São Paulo, Brazil. The staff of the group Reactor and Nuclear Power Studies (SERC) of CNEA, from the argentine side, performed calculations with deterministic models (HUEMUL-PUMA) and probabilistic methods (MCNP) modeling a great number of physical situations of de reactor, which previously have been studied and modeled by members of the Center of Nuclear Engineering of the IPEN, whose results were extensively provided to CNEA. In this paper results for critical configurations are shown. (author)
Validation of IRBURN calculation code system through burnup benchmark analysis
International Nuclear Information System (INIS)
Assessment of the reactor fuel composition during the irradiation time, fuel management and criticality safety analysis require the utilization of a validated burnup calculation code system. In this work a newly developed burnup calculation code system, IRBURN, is introduced for the estimation and analysis of the fuel burnup in LWR reactors. IRBURN provides the full capabilities of the Monte Carlo neutron and photon transport code MCNP4C as well as the versatile code for calculating the buildup and decay of nuclides in nuclear materials, ORIGEN2.1, along with other data processing and linking subroutines. This code has the capability of using different depletion calculation schemes. The accuracy and precision of the implemented algorithms to estimate the eigenvalue and spent fuel isotope concentrations are demonstrated by validation against reliable benchmark problem analyses. A comparison of IRBURN results with experimental data demonstrates that the code predicts the spent fuel concentrations within 10% accuracy. Furthermore, standard deviations of the average values for isotopic concentrations including IRBURN data decreases considerably in comparison with the same parameter excluding IRBURN results, except for a few sets of isotopes. The eigenvalue comparison between our results and the benchmark problems shows a good prediction of the k-inf values during the entire burnup history with the maximum difference of 1% at 100 MWd/kgU.
Benchmark calculation of nuclear design code for HCLWR
International Nuclear Information System (INIS)
In the calculation of the lattice cell for High Conversion Light Water Reactors, big differences of nuclear design parameters appear between the results obtained by various methods and nuclear data libraries. The validity of the calculation can be verified by the critical experiment. The benchmark calculation is also efficient for the estimation of the validity in wide range of lattice parameters and burnup. As we do not have many measured data. The benchmark calculations were done by JAERI and MAPI, using SRAC and WIMS-E respectively. The problem covered the wide range of lattice parameters, i.e., from tight lattice to the current PWR lattice. The comparison was made on the effective multiplication factor, conversion ratio, and reaction rate of each nuclide, including burnup and void effects. The difference of the result is largest at the tightest lattice. But even at that lattice, the difference of the effective multiplication factor is only 1.4 %. The main cause of the difference is the neutron absorption rate U-238 in resonance energy region. The difference of other nuclear design parameters and their cause were also grasped. (author)
Benchmark assemblies of the Los Alamos Critical Assemblies Facility
International Nuclear Information System (INIS)
Several critical assemblies of precisely known materials composition and easily calculated and reproducible geometries have been constructed at the Los Alamos National Laboratory. Some of these machines, notably Jezebel, Flattop, Big Ten, and Godiva, have been used as benchmark assemblies for the comparison of the results of experimental measurements and computation of certain nuclear reaction parameters. These experiments are used to validate both the input nuclear data and the computational methods. The machines and the applications of these machines for integral nuclear data checks are described
Danish calculations of the NEACRP pin-power benchmark
International Nuclear Information System (INIS)
This report describes calculations performed for the NEACRP pin-power benchmark. The calculations are made with the code NEM2D, a diffusion theory code based on the nodal expansion method. (au) (15 tabs., 15 ills., 5 refs.)
TRIGA Mark II Criticality Benchmark Experiment with Burned Fuel
International Nuclear Information System (INIS)
The experimental results of criticality benchmark experiments performed at the Jozef Stefan Institute TRIGA Mark II reactor are presented. The experiments were performed with partly burned fuel in two compact and uniform core configurations in the same arrangements as were used in the fresh fuel criticality benchmark experiment performed in 1991. In the experiments, both core configurations contained only 12 wt% U-ZrH fuel with 20% enriched uranium. The first experimental core contained 43 fuel elements with average burnup of 1.22 MWd or 2.8% 235U burned. The last experimental core configuration was composed of 48 fuel elements with average burnup of 1.15 MWd or 2.6% 235U burned. The experimental determination of keff for both core configurations, one subcritical and one critical, are presented. Burnup for all fuel elements was calculated in two-dimensional four-group diffusion approximation using the TRIGLAV code. The burnup of several fuel elements was measured also by the reactivity method
Benchmark problems and results for verifying resonance calculation methodologies
International Nuclear Information System (INIS)
Resonance calculation is one of the most important procedures for the multi-group neutron transport calculation. With the development of nuclear reactor concepts, many new types of fuel assembly are raised. Compared to the traditional designs, most of the new fuel assemblies have different fuel types either with complex isotopes or with complicated geometry. This makes the traditional resonance calculation method invalid. Recently, many advanced resonance calculation methods are proposed. However, there are few benchmark problems for evaluating those methods with a comprehensive comparison. In this paper, we design 5 groups of benchmark problems including 21 typical cases of different geometries and fuel contents. The reference results of the benchmark problems are generated based on the sub-group method, ultra-fine group method, function expanding method and Monte Carlo method. It is shown that those benchmark problems and their results could be helpful to evaluate the validity of the newly developed resonance calculation method in the future work. (authors)
International Nuclear Information System (INIS)
The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in October of 1992 by the United States Department of Energy. The project quickly became an international effort as scientists from other interested countries became involved. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) became an official activity of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) in 1995. This handbook contains criticality safety benchmark specifications that have been derived from experiments performed at various nuclear critical experiment facilities around the world. The benchmark specifications are intended for use by criticality safety engineers to validate calculational techniques used to establish minimum subcritical margins for operations with fissile material and to determine criticality alarm requirement and placement. Many of the specifications are also useful for nuclear data testing. Example calculations are presented; however, these calculations do not constitute a validation of the codes or cross section data. The evaluated criticality safety benchmark data are given in nine volumes. These volumes span nearly 66,000 pages and contain 558 evaluations with benchmark specifications for 4,798 critical, near critical or subcritical configurations, 24 criticality alarm placement/shielding configurations with multiple dose points for each and 200 configurations that have been categorised as fundamental physics measurements that are relevant to criticality safety applications. New to the Handbook are benchmark specifications for Critical, Bare, HEU(93.2)- Metal Sphere experiments referred to as ORSphere that were performed by a team of experimenters at Oak Ridge National Laboratory in the early 1970's. A photograph of this assembly is shown on the front cover
Criticality benchmarking of ANET Monte Carlo code
International Nuclear Information System (INIS)
In this work the new Monte Carlo code ANET is tested on criticality calculations. ANET is developed based on the high energy physics code GEANT of CERN and aims at progressively satisfying several requirements regarding both simulations of GEN II/III reactors, as well as of innovative nuclear reactor designs such as the Accelerator Driven Systems (ADSs). Here ANET is applied on three different nuclear configurations, including a subcritical assembly, a Material Testing Reactor and the conceptual configuration of an ADS. In the first case, calculation of the effective multiplication factor (keff) are performed for the Training Nuclear Reactor of the Aristotle University of Thessaloniki, while in the second case keff is computed for the fresh fueled core of the Portuguese research reactor (RPJ) just after its conversion to Low Enriched Uranium, considering the control rods at the position that renders the reactor critical. In both cases ANET computations are compared with corresponding results obtained by three different well established codes, including both deterministic (XSDRNPM/CITATION) and Monte Carlo (TRIPOLI, MCNP). In the RPI case, keff computations are also compared with observations during the reactor core commissioning since the control rods are considered at criticality position. The above verification studies show ANET to produce reasonable results since they are satisfactorily compared with other models as well as with observations. For the third case (ADS), preliminary ANET computations of keff for various intensities of the proton beam are presented, showing also a reasonable code performance concerning both the order of magnitude and the relative variation of the computed parameter. (author)
Benchmark on the Kritz-2 Leu and MOX critical experiments
International Nuclear Information System (INIS)
In the framework of the joint activities of the OECD/NEA Working Party on the Physics of Plutonium Fuels and Innovative Fuel Cycles (WPPR)1 and the Task Force on Reactor-based Plutonium Disposition (TFRPD), an international benchmark exercise based on KRITZ UO2 and MOX critical configurations was launched in October 2000. The aim of this exercise was to investigate the capabilities of the current production codes and nuclear data libraries to analyse MOX-fuelled systems, and to compare the accuracy of the predictions for the MOX- and UO2-fuelled configurations. Institutions from 7 countries participated in this exercise, providing 13 solutions. The report provides comparative analyses of calculated and measured results, as well as intercomparisons of some of the results obtained by participants by calculation only. (author)
IRIS core criticality calculations
International Nuclear Information System (INIS)
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)
Criticality calculations on BARC parallel processor- ANUPAM
International Nuclear Information System (INIS)
Parallel processing offers an increase in computational speed beyond the technological limitations of single processor systems. BARC has recently developed a parallel processing system (ANUPAM) based Multiple Instruction Multiple Data (MIMD) distributed memory architecture. In the work reported here, the sequential version of Monte Carlo code MONALI is modified to work on the ANUPAM for criticality calculations. The problem of random number generation in a parallel environment is handled using leapfrog technique. The code is modified to use variable number of slave processors. The parallel version of MONALI is used to calculate multiplication factor, fluxes and absorptions in one of the 8x8 fuel assemblies of IAEA BWR benchmark in 69 groups. To compare gain in execution time, the benchmark is also solved on LANDMARK and ND-570 systems (both serial) using the sequential version of the code. Speedup and efficiencies achieved on varying the number of slave processors are encouraging. (author). 5 refs., 1 tab
The ORSphere Benchmark Evaluation and Its Potential Impact on Nuclear Criticality Safety
Energy Technology Data Exchange (ETDEWEB)
John D. Bess; Margaret A. Marshall; J. Blair Briggs
2013-10-01
In the early 1970’s, critical experiments using an unreflected metal sphere of highly enriched uranium (HEU) were performed with the focus to provide a “very accurate description…as an ideal benchmark for calculational methods and cross-section data files.” Two near-critical configurations of the Oak Ridge Sphere (ORSphere) were evaluated as acceptable benchmark experiments for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook). The results from those benchmark experiments were then compared with additional unmoderated and unreflected HEU metal benchmark experiment configurations currently found in the ICSBEP Handbook. For basic geometries (spheres, cylinders, and slabs) the eigenvalues calculated using MCNP5 and ENDF/B-VII.0 were within 3 of their respective benchmark values. There appears to be generally a good agreement between calculated and benchmark values for spherical and slab geometry systems. Cylindrical geometry configurations tended to calculate low, including more complex bare HEU metal systems containing cylinders. The ORSphere experiments do not calculate within their 1s uncertainty and there is a possibility that the effect of the measured uncertainties for the GODIVA I benchmark may need reevaluated. There is significant scatter in the calculations for the highly-correlated ORCEF cylinder experiments, which are constructed from close-fitting HEU discs and annuli. Selection of a nuclear data library can have a larger impact on calculated eigenvalue results than the variation found within calculations of a given experimental series, such as the ORCEF cylinders, using a single nuclear data set.
MOx benchmark calculations by deterministic and Monte Carlo codes
International Nuclear Information System (INIS)
Highlights: ► MOx based depletion calculation. ► Methodology to create continuous energy pseudo cross section for lump of minor fission products. ► Mass inventory comparison between deterministic and Monte Carlo codes. ► Higher deviation was found for several isotopes. - Abstract: A depletion calculation benchmark devoted to MOx fuel is an ongoing objective of the OECD/NEA WPRS following the study of depletion calculation concerning UOx fuels. The objective of the proposed benchmark is to compare existing depletion calculations obtained with various codes and data libraries applied to fuel and back-end cycle configurations. In the present work the deterministic code NEWT/ORIGEN-S of the SCALE6 codes package and the Monte Carlo based code MONTEBURNS2.0 were used to calculate the masses of inventory isotopes. The methodology to apply the MONTEBURNS2.0 to this benchmark is also presented. Then the results from both code were compared.
Critical power prediction by CATHARE2 of the OECD/NRC BFBT benchmark
International Nuclear Information System (INIS)
Highlights: • We used CATHARE code to calculate the critical power exercises of the OECD/NRC BFBT benchmark. • We considered both steady-state and transient critical power tests of the benchmark. • We used both the 1D and 3D features of the CATHARE code to simulate the experiments. • Acceptable prediction of the critical power and its location in the bundle is obtained using appropriate modelling. - Abstract: This paper presents an application of the French best estimate thermal-hydraulic code CATHARE 2 to calculate the critical power and departure from nucleate boiling (DNB) exercises of the International OECD/NRC BWR Fuel Bundle Test (BFBT) benchmark. The assessment activity is performed comparing the code calculation results with available in the framework of the benchmark experimental data from Japanese Nuclear Power Engineering Corporation (NUPEC). Two-phase flow calculations on prediction of the critical power have been carried out both in steady state and transient cases, using one-dimensional and three-dimensional modelling. Results of the steady-state critical power tests calculation have shown the ability of CATHARE code to predict reasonably the critical power and its location, using appropriate modelling
Critical power prediction by CATHARE2 of the OECD/NRC BFBT benchmark
Energy Technology Data Exchange (ETDEWEB)
Lutsanych, Sergii, E-mail: s.lutsanych@ing.unipi.it [San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa, Via Livornese 1291, 56122, San Piero a Grado, Pisa (Italy); Sabotinov, Luben, E-mail: luben.sabotinov@irsn.fr [Institut for Radiological Protection and Nuclear Safety (IRSN), 31 avenue de la Division Leclerc, 92262 Fontenay-aux-Roses (France); D’Auria, Francesco, E-mail: francesco.dauria@dimnp.unipi.it [San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa, Via Livornese 1291, 56122, San Piero a Grado, Pisa (Italy)
2015-03-15
Highlights: • We used CATHARE code to calculate the critical power exercises of the OECD/NRC BFBT benchmark. • We considered both steady-state and transient critical power tests of the benchmark. • We used both the 1D and 3D features of the CATHARE code to simulate the experiments. • Acceptable prediction of the critical power and its location in the bundle is obtained using appropriate modelling. - Abstract: This paper presents an application of the French best estimate thermal-hydraulic code CATHARE 2 to calculate the critical power and departure from nucleate boiling (DNB) exercises of the International OECD/NRC BWR Fuel Bundle Test (BFBT) benchmark. The assessment activity is performed comparing the code calculation results with available in the framework of the benchmark experimental data from Japanese Nuclear Power Engineering Corporation (NUPEC). Two-phase flow calculations on prediction of the critical power have been carried out both in steady state and transient cases, using one-dimensional and three-dimensional modelling. Results of the steady-state critical power tests calculation have shown the ability of CATHARE code to predict reasonably the critical power and its location, using appropriate modelling.
WIPP Benchmark calculations with the large strain SPECTROM codes
International Nuclear Information System (INIS)
This report provides calculational results from the updated Lagrangian structural finite-element programs SPECTROM-32 and SPECTROM-333 for the purpose of qualifying these codes to perform analyses of structural situations in the Waste Isolation Pilot Plant (WIPP). Results are presented for the Second WIPP Benchmark (Benchmark II) Problems and for a simplified heated room problem used in a parallel design calculation study. The Benchmark II problems consist of an isothermal room problem and a heated room problem. The stratigraphy involves 27 distinct geologic layers including ten clay seams of which four are modeled as frictionless sliding interfaces. The analyses of the Benchmark II problems consider a 10-year simulation period. The evaluation of nine structural codes used in the Benchmark II problems shows that inclusion of finite-strain effects is not as significant as observed for the simplified heated room problem, and a variety of finite-strain and small-strain formulations produced similar results. The simplified heated room problem provides stratigraphic complexity equivalent to the Benchmark II problems but neglects sliding along the clay seams. The simplified heated problem does, however, provide a calculational check case where the small strain-formulation produced room closures about 20 percent greater than those obtained using finite-strain formulations. A discussion is given of each of the solved problems, and the computational results are compared with available published results. In general, the results of the two SPECTROM large strain codes compare favorably with results from other codes used to solve the problems
Full CI benchmark calculations on CH3
Bauschlicher, Charles W., Jr.; Taylor, Peter R.
1987-01-01
Full CI calculations have been performed on the CH3 radical. The full CI results are compared to those obtained using CASSCF/multireference CI and coupled-pair functional methods, both at the equilibrium CH distance and at geometries with the three CH bonds extended. In general, the performance of the approximate methods is similar to that observed in calculations on other molecules in which one or two bonds were stretched.
Energy Technology Data Exchange (ETDEWEB)
J. Blair Briggs; Lori Scott; Enrico Sartori; Yolanda Rugama
2008-09-01
Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) continue to expand their efforts and broaden their scope to identify, evaluate, and provide integral benchmark data for method and data validation. Benchmark model specifications provided by these two projects are used heavily by the international reactor physics, nuclear data, and criticality safety communities. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. The status of the IRPhEP and ICSBEP is discussed in this paper, and the future of the two projects is outlined and discussed. Selected benchmarks that have been added to the IRPhEP and ICSBEP handbooks since PHYSOR’06 are highlighted, and the future of the two projects is discussed.
Photon shielding calculations for a radiation waste facility benchmark
Energy Technology Data Exchange (ETDEWEB)
Estes, G.P.; Urban, W.T.; Heath, A.R.
1985-11-01
Photon transport calculations have been performed for the ANS 6.2.1 radiation waste facility shielding benchmark using the continuous energy Monte Carlo code MCNP, and ONEDANT and TWODANT discrete ordinates codes. Comparisons are made of integral dose rates and flux spectra calculated with the three codes for various geometries, cross-section sets, and source and output energy group structures.
Benchmarking calculations of excitonic couplings between bacteriochlorophylls
Kenny, Elise P
2015-01-01
Excitonic couplings between (bacterio)chlorophyll molecules are necessary for simulating energy transport in photosynthetic complexes. Many techniques for calculating the couplings are in use, from the simple (but inaccurate) point-dipole approximation to fully quantum-chemical methods. We compared several approximations to determine their range of applicability, noting that the propagation of experimental uncertainties poses a fundamental limit on the achievable accuracy. In particular, the uncertainty in crystallographic coordinates yields an uncertainty of about 20% in the calculated couplings. Because quantum-chemical corrections are smaller than 20% in most biologically relevant cases, their considerable computational cost is rarely justified. We therefore recommend the electrostatic TrEsp method across the entire range of molecular separations and orientations because its cost is minimal and it generally agrees with quantum-chemical calculations to better than the geometric uncertainty. We also caution ...
Benchmark calculation of CANDU end shielding system
Energy Technology Data Exchange (ETDEWEB)
Roh, Gyuhong; Choi, Hangbok [KAERI, Taejon (Korea, Republic of)
1998-05-01
A shielding analysis was performed for the end shield of CANDU 6 reactor. The one-dimensional discrete ordinate code ANISN with a 38-group neutron-gamma library, extracted from DLC-37D library, was used to estimate the dose rate for the natural uranium CANDU reactor. For comparison, MCNP-4B calculation was performed for the same system using continuous, discrete and multi-group libraries. The comparison has shown that the total dose rate of the ANISN calculation agrees well with that of the MCNP calculation. However, the individual dose rate (neutron and gamma) has shown opposite trends between ANISN and MCNP estimates, which may require a consistent library generation for both codes.
Benchmarking calculations of excitonic couplings between bacteriochlorophylls
Kenny, Elise P.; Kassal, Ivan
2015-01-01
Excitonic couplings between (bacterio)chlorophyll molecules are necessary for simulating energy transport in photosynthetic complexes. Many techniques for calculating the couplings are in use, from the simple (but inaccurate) point-dipole approximation to fully quantum-chemical methods. We compared several approximations to determine their range of applicability, noting that the propagation of experimental uncertainties poses a fundamental limit on the achievable accuracy. In particular, the ...
Quantum critical benchmark for density functional theory
Grabowski, Paul E.; Burke, Kieron
2014-01-01
Two electrons at the threshold of ionization represent a severe test case for electronic structure theory. A pseudospectral method yields a very accurate density of the two-electron ion with nuclear charge close to the critical value. Highly accurate energy components and potentials of Kohn-Sham density functional theory are given, as well as a useful parametrization of the critical density. The challenges for density functional approximations and the strength of correlation are also discussed.
Providing Nuclear Criticality Safety Analysis Education through Benchmark Experiment Evaluation
International Nuclear Information System (INIS)
One of the challenges that today's new workforce of nuclear criticality safety engineers face is the opportunity to provide assessment of nuclear systems and establish safety guidelines without having received significant experience or hands-on training prior to graduation. Participation in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and/or the International Reactor Physics Experiment Evaluation Project (IRPhEP) provides students and young professionals the opportunity to gain experience and enhance critical engineering skills.
Benchmark density functional theory calculations for nanoscale conductance
Strange, Mikkel; Bækgaard, Iben Sig Buur; Thygesen, Kristian Sommer; Jacobsen, Karsten Wedel
2008-01-01
We present a set of benchmark calculations for the Kohn-Sham elastic transmission function of five representative single-molecule junctions. The transmission functions are calculated using two different density functional theory methods, namely an ultrasoft pseudopotential plane-wave code in combination with maximally localized Wannier functions and the norm-conserving pseudopotential code SIESTA which applies an atomic orbital basis set. All calculations have been converged with respect to t...
Benchmark Calculations For A VVER-1000 Assembly Using SRAC
International Nuclear Information System (INIS)
This work presents the neutronic calculation results of a VVER-1000 assembly using SRAC with 107 energy groups in comparison with the benchmark values in the OECD/NEA report. The main neutronic characteristics which were calculated in this comparison include infinite multiplication factors (k-inf), nuclide densities as the function of burnup and pin-wise power distribution. Calculations were conducted with various conditions of fuel, coolant and boron content in coolant. (author)
Benchmark Calculations of Noncovalent Interactions of Halogenated Molecules
Czech Academy of Sciences Publication Activity Database
Řezáč, Jan; Riley, Kevin Eugene; Hobza, Pavel
2012-01-01
Roč. 8, č. 11 (2012), s. 4285-4292. ISSN 1549-9618 R&D Projects: GA ČR GBP208/12/G016 Institutional support: RVO:61388963 Keywords : halogenated molecules * noncovalent interactions * benchmark calculations Subject RIV: CF - Physical ; Theoretical Chemistry Impact factor: 5.389, year: 2012
Benchmark density functional theory calculations for nanoscale conductance
DEFF Research Database (Denmark)
Strange, Mikkel; Bækgaard, Iben Sig Buur; Thygesen, Kristian Sommer;
2008-01-01
We present a set of benchmark calculations for the Kohn-Sham elastic transmission function of five representative single-molecule junctions. The transmission functions are calculated using two different density functional theory methods, namely an ultrasoft pseudopotential plane-wave code in...... observe a systematic downshift of the SIESTA transmission functions relative to the plane-wave results. The effect diminishes as the atomic orbital basis is enlarged; however, the convergence can be rather slow....
Benchmark calculations of thermal reaction rates. I - Quantal scattering theory
Chatfield, David C.; Truhlar, Donald G.; Schwenke, David W.
1991-01-01
The thermal rate coefficient for the prototype reaction H + H2 yields H2 + H with zero total angular momentum is calculated by summing, averaging, and numerically integrating state-to-state reaction probabilities calculated by time-independent quantum-mechanical scattering theory. The results are very carefully converged with respect to all numerical parameters in order to provide high-precision benchmark results for confirming the accuracy of new methods and testing their efficiency.
Criticality safety benchmark evaluation project: Recovering the past
Energy Technology Data Exchange (ETDEWEB)
Trumble, E.F.
1997-06-01
A very brief summary of the Criticality Safety Benchmark Evaluation Project of the Westinghouse Savannah River Company is provided in this paper. The purpose of the project is to provide a source of evaluated criticality safety experiments in an easily usable format. Another project goal is to search for any experiments that may have been lost or contain discrepancies, and to determine if they can be used. Results of evaluated experiments are being published as US DOE handbooks.
MCNP (trademark) ENDF/B-VI iron benchmark calculations
Court, J. D.; Hendricks, J. S.
Four iron shielding benchmarks have been calculated for, we believe the first time, with MCNP4A and its new ENDF/B-VI library. These calculations are part of the Hiroshima/Nagasaki dose re-evaluation for the National Academy of Sciences and the Defense Nuclear Agency. We believe these calculations are significant because they validate MCNP and the new ENDF/B-VI libraries. These calculations are compared to ENDF/B-V, experiment, and in some cases the recommended MCNP data library (a T-2 evaluation) and ENDF/IV.
TRX and UO2 criticality benchmarks with SAM-CE
International Nuclear Information System (INIS)
A set of thermal reactor benchmark calculations with SAM-CE which have been conducted at both MAGI and at BNL are described. Their purpose was both validation of the SAM-CE reactor eigenvalue capability developed by MAGI and a substantial contribution to the data testing of both ENDF/B-IV and ENDF/B-V libraries. This experience also resulted in increased calculational efficiency of the code and an example is given. The benchmark analysis included the TRX-1 infinite cell using both ENDF/B-IV and ENDF/B-V cross section sets and calculations using ENDF/B-IV of the TRX-1 full core and TRX-2 cell. BAPL-UO2-1 calculations were conducted for the cell using both ENDF/B-IV and ENDF/B-V and for the full core with ENDF/B-V
Energy Technology Data Exchange (ETDEWEB)
Miyoshi, Yoshinori; Yamamoto, Toshihiro; Nakamura, Takemi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2001-08-01
In order to validate the availability of criticality calculation codes and related nuclear data library, a series of fundamental benchmark experiments on low enriched uranyl nitrate solution have been performed with a Static Experiment Criticality Facility, STACY in JAERI. The basic core composed of a single tank with water reflector was used for accumulating the systematic data with well-known experimental uncertainties. This paper presents the outline of the core configurations of STACY, the standard calculation model, and calculation results with a Monte Carlo code and JENDL 3.2 nuclear data library. (author)
Benchmark calculations on nuclear characteristics of JRR-4 HEU core by SRAC code system
International Nuclear Information System (INIS)
The reduced enrichment program for the JRR-4 has been progressing based on JAERI's RERTR (Reduced Enrichment Research and Test Reactor) program. The SRAC (JAERI Thermal Reactor Standard Code System for Reactor Design and Analysis) is used for the neutronic design of the JRR-4 LEU Core. This report describes the benchmark calculations on the neutronic characteristics of the JRR-4 HEU Core in order to validate the calculation method. The benchmark calculations were performed on the various kind of neutronic characteristics such as excess reactivity, criticality, control rod worth, thermal neutron flux distribution, void coefficient, temperature coefficient, mass coefficient, kinetic parameters and poisoning effect by Xe-135 build up. As the result, it was confirmed that these calculated values are in satisfactory agreement with the measured values. Therefore, the calculational method by the SRAC was validated. (author)
AGING FACILITY CRITICALITY SAFETY CALCULATIONS
International Nuclear Information System (INIS)
The purpose of this design calculation is to revise and update the previous criticality calculation for the Aging Facility (documented in BSC 2004a). This design calculation will also demonstrate and ensure that the storage and aging operations to be performed in the Aging Facility meet the criticality safety design criteria in the ''Project Design Criteria Document'' (Doraswamy 2004, Section 4.9.2.2), and the functional nuclear criticality safety requirement described in the ''SNF Aging System Description Document'' (BSC [Bechtel SAIC Company] 2004f, p. 3-12). The scope of this design calculation covers the systems and processes for aging commercial spent nuclear fuel (SNF) and staging Department of Energy (DOE) SNF/High-Level Waste (HLW) prior to its placement in the final waste package (WP) (BSC 2004f, p. 1-1). Aging commercial SNF is a thermal management strategy, while staging DOE SNF/HLW will make loading of WPs more efficient (note that aging DOE SNF/HLW is not needed since these wastes are not expected to exceed the thermal limits form emplacement) (BSC 2004f, p. 1-2). The description of the changes in this revised document is as follows: (1) Include DOE SNF/HLW in addition to commercial SNF per the current ''SNF Aging System Description Document'' (BSC 2004f). (2) Update the evaluation of Category 1 and 2 event sequences for the Aging Facility as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004c, Section 7). (3) Further evaluate the design and criticality controls required for a storage/aging cask, referred to as MGR Site-specific Cask (MSC), to accommodate commercial fuel outside the content specification in the Certificate of Compliance for the existing NRC-certified storage casks. In addition, evaluate the design required for the MSC that will accommodate DOE SNF/HLW. This design calculation will achieve the objective of providing the criticality safety results to support the preliminary design of the Aging
JNC results of BN-600 benchmark calculation (phase 3)
International Nuclear Information System (INIS)
The present work is the result of phase 3 BN-600 core benchmark problem, meaning burnup and heterogeneity. Analytical method applied consisted of: JENDL-3.2 nuclear data library, group constants (70 group, ABBN type self shielding transport factors), heterogeneous cell model for fuel and control rod, basic diffusion calculation (CITATION code), transport theory and mesh size correction (NSHEX code based on SN transport nodal method developed by JNC). Burnup and heterogeneity calculation results are presented obtained by applying both diffusion and transport approach for beginning and end of cycle
Canister Transfer Facility Criticality Calculations
Energy Technology Data Exchange (ETDEWEB)
J.E. Monroe-Rammsy
2000-10-13
The objective of this calculation is to evaluate the criticality risk in the surface facility for design basis events (DBE) involving Department of Energy (DOE) Spent Nuclear Fuel (SNF) standardized canisters (Civilian Radioactive Waste Management System [CRWMS] Management and Operating Contractor [M&O] 2000a). Since some of the canisters will be stored in the surface facility before they are loaded in the waste package (WP), this calculation supports the demonstration of concept viability related to the Surface Facility environment. The scope of this calculation is limited to the consideration of three DOE SNF fuels, specifically Enrico Fermi SNF, Training Research Isotope General Atomic (TRIGA) SNF, and Mixed Oxide (MOX) Fast Flux Test Facility (FFTF) SNF.
Criticality Benchmark Analysis of the HTTR Annular Startup Core Configurations
Energy Technology Data Exchange (ETDEWEB)
John D. Bess
2009-11-01
One of the high priority benchmarking activities for corroborating the Next Generation Nuclear Plant (NGNP) Project and Very High Temperature Reactor (VHTR) Program is evaluation of Japan's existing High Temperature Engineering Test Reactor (HTTR). The HTTR is a 30 MWt engineering test reactor utilizing graphite moderation, helium coolant, and prismatic TRISO fuel. A large amount of critical reactor physics data is available for validation efforts of High Temperature Gas-cooled Reactors (HTGRs). Previous international reactor physics benchmarking activities provided a collation of mixed results that inaccurately predicted actual experimental performance.1 Reevaluations were performed by the Japanese to reduce the discrepancy between actual and computationally-determined critical configurations.2-3 Current efforts at the Idaho National Laboratory (INL) involve development of reactor physics benchmark models in conjunction with the International Reactor Physics Experiment Evaluation Project (IRPhEP) for use with verification and validation methods in the VHTR Program. Annular cores demonstrate inherent safety characteristics that are of interest in developing future HTGRs.
Criticality Benchmark Analysis of the HTTR Annular Startup Core Configurations
International Nuclear Information System (INIS)
One of the high priority benchmarking activities for corroborating the Next Generation Nuclear Plant (NGNP) Project and Very High Temperature Reactor (VHTR) Program is evaluation of Japan's existing High Temperature Engineering Test Reactor (HTTR). The HTTR is a 30 MWt engineering test reactor utilizing graphite moderation, helium coolant, and prismatic TRISO fuel. A large amount of critical reactor physics data is available for validation efforts of High Temperature Gas-cooled Reactors (HTGRs). Previous international reactor physics benchmarking activities provided a collation of mixed results that inaccurately predicted actual experimental performance.1 Reevaluations were performed by the Japanese to reduce the discrepancy between actual and computationally-determined critical configurations.2-3 Current efforts at the Idaho National Laboratory (INL) involve development of reactor physics benchmark models in conjunction with the International Reactor Physics Experiment Evaluation Project (IRPhEP) for use with verification and validation methods in the VHTR Program. Annular cores demonstrate inherent safety characteristics that are of interest in developing future HTGRs.
AGING FACILITY CRITICALITY SAFETY CALCULATIONS
Energy Technology Data Exchange (ETDEWEB)
C.E. Sanders
2004-09-10
The purpose of this design calculation is to revise and update the previous criticality calculation for the Aging Facility (documented in BSC 2004a). This design calculation will also demonstrate and ensure that the storage and aging operations to be performed in the Aging Facility meet the criticality safety design criteria in the ''Project Design Criteria Document'' (Doraswamy 2004, Section 4.9.2.2), and the functional nuclear criticality safety requirement described in the ''SNF Aging System Description Document'' (BSC [Bechtel SAIC Company] 2004f, p. 3-12). The scope of this design calculation covers the systems and processes for aging commercial spent nuclear fuel (SNF) and staging Department of Energy (DOE) SNF/High-Level Waste (HLW) prior to its placement in the final waste package (WP) (BSC 2004f, p. 1-1). Aging commercial SNF is a thermal management strategy, while staging DOE SNF/HLW will make loading of WPs more efficient (note that aging DOE SNF/HLW is not needed since these wastes are not expected to exceed the thermal limits form emplacement) (BSC 2004f, p. 1-2). The description of the changes in this revised document is as follows: (1) Include DOE SNF/HLW in addition to commercial SNF per the current ''SNF Aging System Description Document'' (BSC 2004f). (2) Update the evaluation of Category 1 and 2 event sequences for the Aging Facility as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004c, Section 7). (3) Further evaluate the design and criticality controls required for a storage/aging cask, referred to as MGR Site-specific Cask (MSC), to accommodate commercial fuel outside the content specification in the Certificate of Compliance for the existing NRC-certified storage casks. In addition, evaluate the design required for the MSC that will accommodate DOE SNF/HLW. This design calculation will achieve the objective of providing the
Benchmarking Outcomes in the Critically Injured Burn Patient
Klein, Matthew B.; Goverman, Jeremy; Hayden, Douglas L.; Fagan, Shawn P.; McDonald-Smith, Grace P.; Alexander, Andrew K.; Gamelli, Richard L.; Gibran, Nicole S.; Finnerty, Celeste C.; Jeschke, Marc G.; Arnoldo, Brett; Wispelwey, Bram; Mindrinos, Michael N.; Xiao, Wenzhong; Honari, Shari E.; Mason, Philip H.; Schoenfeld, David A.; Herndon, David N.; Tompkins, Ronald G.
2014-01-01
Objective To determine and compare outcomes with accepted benchmarks in burn care at six academic burn centers. Background Since the 1960s, U.S. morbidity and mortality rates have declined tremendously for burn patients, likely related to improvements in surgical and critical care treatment. We describe the baseline patient characteristics and well-defined outcomes for major burn injuries. Methods We followed 300 adults and 241 children from 2003–2009 through hospitalization using standard operating procedures developed at study onset. We created an extensive database on patient and injury characteristics, anatomic and physiological derangement, clinical treatment, and outcomes. These data were compared with existing benchmarks in burn care. Results Study patients were critically injured as demonstrated by mean %TBSA (41.2±18.3 for adults and 57.8±18.2 for children) and presence of inhalation injury in 38% of the adults and 54.8% of the children. Mortality in adults was 14.1% for those less than 55 years old and 38.5% for those age ≥55 years. Mortality in patients less than 17 years old was 7.9%. Overall, the multiple organ failure rate was 27%. When controlling for age and %TBSA, presence of inhalation injury was not significant. Conclusions This study provides the current benchmark for major burn patients. Mortality rates, notwithstanding significant % TBSA and presence of inhalation injury, have significantly declined compared to previous benchmarks. Modern day surgical and medically intensive management has markedly improved to the point where we can expect patients less than 55 years old with severe burn injuries and inhalation injury to survive these devastating conditions. PMID:24722222
Calculation of potassium critical temperature
International Nuclear Information System (INIS)
The paper describes the algorithm of the functional prediction which is based on the selforganization of nonlinear algebraic models. The calculation procedure includes the module for the recognition of the dependence type hitch allows to restrict the number of choice of the prediction functions at the each step of the model building. The characteristic property of this algorithm is bootstrap method application as the external criteria of the selforganization. The calculation module is built using APL*PLUS and the user-friendly interface is implemented using Clipper 5.01 under Windows control. When using the algorithm and the programs, the critical point of potassium has been predicted on the base of the solubility curves of liquid and steam. 9 refs.; 1 fig.; 1 tab
Muse-4 benchmark calculations using MCNP-4C and different nuclear data libraries
International Nuclear Information System (INIS)
Current calculation methods and nuclear data are well validated for conventional nuclear reactor systems. However there is a further need for validating the computational tools and the nuclear data for ADS applications. The OECD/NEA, in co-operation with CIEMAT (Spain) and CEA (France), therefore launched a benchmark based on the MUSE-4 experiments being carried out at Cadarache, France, to simulate the neutronics of a source-driven sub-critical system. This paper summarises the calculated results of the MUSE-4 benchmark obtained from the Monte Carlo code MCNP (Version 4Ca) using different nuclear data evaluations, and shows the sensitivity of the requested results with regard to the nuclear data used. All the calculated results will be compared against measured data after the completion of the experiments foreseen for the end of 2003. (author)
Experimental Criticality Benchmarks for SNAP 10A/2 Reactor Cores
Energy Technology Data Exchange (ETDEWEB)
Krass, A.W.
2005-12-19
This report describes computational benchmark models for nuclear criticality derived from descriptions of the Systems for Nuclear Auxiliary Power (SNAP) Critical Assembly (SCA)-4B experimental criticality program conducted by Atomics International during the early 1960's. The selected experimental configurations consist of fueled SNAP 10A/2-type reactor cores subject to varied conditions of water immersion and reflection under experimental control to measure neutron multiplication. SNAP 10A/2-type reactor cores are compact volumes fueled and moderated with the hydride of highly enriched uranium-zirconium alloy. Specifications for the materials and geometry needed to describe a given experimental configuration for a model using MCNP5 are provided. The material and geometry specifications are adequate to permit user development of input for alternative nuclear safety codes, such as KENO. A total of 73 distinct experimental configurations are described.
Criticality benchmark guide for light-water-reactor fuel in transportation and storage packages
International Nuclear Information System (INIS)
This report is designed as a guide for performing criticality benchmark calculations for light-water-reactor (LWR) fuel applications. The guide provides documentation of 180 criticality experiments with geometries, materials, and neutron interaction characteristics representative of transportation packages containing LWR fuel or uranium oxide pellets or powder. These experiments should benefit the U.S. Nuclear Regulatory Commission (NRC) staff and licensees in validation of computational methods used in LWR fuel storage and transportation concerns. The experiments are classified by key parameters such as enrichment, water/fuel volume, hydrogen-to-fissile ratio (H/X), and lattice pitch. Groups of experiments with common features such as separator plates, shielding walls, and soluble boron are also identified. In addition, a sample validation using these experiments and a statistical analysis of the results are provided. Recommendations for selecting suitable experiments and determination of calculational bias and uncertainty are presented as part of this benchmark guide
Benchmark calculation with MOSRA-SRAC for burnup of a BWR fuel assembly
International Nuclear Information System (INIS)
The Japan Atomic Energy Agency has developed the Modular Reactor Analysis Code System MOSRA to improve the applicability of neutronic characteristics modeling. The cell calculation module MOSRA-SRAC is based on the collision probability method and is one of the core modules of the MOSRA system. To test the module on a real-world problem, it was combined with the benchmark program 'Burnup Credit Criticality Benchmark Phase IIIC.' In this program participants are requested to submit the neutronic characteristics of burnup calculations for a BWR fuel assembly containing fuel rods poisoned with gadolinium (Gd2O3), which is similar to the fuel assembly at TEPCO's Fukushima Daiichi Nuclear Power Station. Because of certain restrictions of the MOSRA-SRAC burnup calculations part of the geometry model was homogenized. In order to verify the validity of MOSRA-SRAC, including the effects of the homogenization, the calculated burnup dependent infinite multiplication factor and the nuclide compositions were compared with those obtained with the burnup calculation code MVP-BURN which had already been validated for many benchmark problems. As a result of the comparisons, the applicability of MOSRA-SRAC module for the BWR assembly has been verified. Furthermore, it can be shown that the effects of the homogenization are smaller than the effects due to the calculation method for both multiplication factor and compositions. (author)
International Nuclear Information System (INIS)
Full text: New releases of nuclear data files made available during the few recent years. The reference MCNP5 code (1) for Monte Carlo calculations is usually distributed with only one standard nuclear data library for neutron interactions based on ENDF/B-VI. The main goal of this work is to process new neutron cross sections libraries in ACE continuous format for MCNP code based on the most recent data files recently made available for the scientific community : ENDF/B-VII.b2, ENDF/B-VI (release 8), JEFF3.0, JEFF-3.1, JENDL-3.3 and JEF2.2. In our data treatment, we used the modular NJOY system (release 99.9) (2) in conjunction with its most recent upadates. Assessment of the processed point wise cross sections libraries performances was made by means of some criticality prediction and analysis of other integral parameters for a set of reactor benchmarks. Almost all the analyzed benchmarks were taken from the international handbook of Evaluated criticality safety benchmarks experiments from OECD (3). Some revised benchmarks were taken from references (4,5). These benchmarks use Pu-239 or U-235 as the main fissionable materiel in different forms, different enrichments and cover various geometries. Monte Carlo calculations were performed in 3D with maximum details of benchmark description and the S(α,β) cross section treatment was adopted in all thermal cases. The resulting one standard deviation confidence interval for the eigenvalue is typically +/-13% to +/-20 pcm
Burn up calculation applied to the NEACRP fast breeder benchmark
International Nuclear Information System (INIS)
The burn up calculations have been performed for the NEACRP fast breeder benchmark. The calculated core parameters are based on the proposal at the NEACRP meeting due to Hammer (CEA). The present calculations have been perfomed basing on JENDL-2 (Japanese Evaluated Nuclear Data Library) instead of JENDL-1 which was used for the previous international comparison calculation of a large LMFBR. The core parameters of the fresh core have been recalculated using JENDL-2 in order to enable direct comparison with those of the end-of-cycle core. The effective microscopic cross sections for fresh core elements have been obtained with use of the ESELEM5 code in 25 groups by weighting with a fundamental mode fine spectrum. Those of F.P. and Actinide nuclides have been generated by using the PROF-GROUCH-G2 code by weighting with 1/E and fission spectrum. The calculations based on the seventy group constants set (JENDL-2B-70) have been performed for a comparison. The burn up calculations have been performed in R-Z geometry by the diffusion theory code PHENIX. The irradiated fuel composition have been obtained at the end of-cycle of the inner core zone 1 by using the zero dimensional burn-up code, FPG S-3. The final report has been submitted to Hammer and intercomparison of solution will be made at NEACRP. Tables of group cross sections for Actinides and F.P. are shown in Appendixes. (author)
International Nuclear Information System (INIS)
Full text: The International Criticality Safety Benchmark Evaluation Project (ICSBEP) was initiated in October of 1992 by the Department of Energy Defence Programs, now NNSA. The U.S. effort to support and provide leadership for the ICSBEP has been funded by DOE-DP since that time. The project is managed through the Idaho National Engineering and Environmental Laboratory (INEEL), but involves nationally known criticality safety experts from Los Alamos National Laboratory, Lawrence Livermore National Laboratory, Savannah River Technology Center, Oak Ridge National Laboratory and the Y-12 Plant, Hanford, Argonne National Laboratory, and the Rocky Flat Plant. An International Criticality Safety Data Exchange component was added to the project during 1994. Representatives from the United Kingdom, France, Japan, the Russian Federation, Hungary, Republic of Korea, Slovenia, Yugoslavia, Kazakhstan, Spain, Israel, Brazil, and Poland are now participating on the project and China, South Africa, and the Czech Republic have indicated that they plan to contribute to the project. The ICSBEP is an official activity of the OECD-NEA. The United States is the lead country, providing most of the administrative support. The purpose of the ICSBEP is to: 1. Identify and evaluate a comprehensive set of criticality related benchmark data. 2. Verify the data, to the extent possible, by reviewing original and subsequently revised documentation, logbook data when possible, and by talking with the experimenters or individuals who are familiar with the experimenters or the experimental facility. 3. Compile the data into a standardized format. 4. Perform calculations of each experiment with standard criticality safety codes. 5. Formally document the work into a single source of verified and internationally peer reviewed benchmark critical data. Each experiment evaluation undergoes a thorough internal review by someone within the evaluator's organization. The internal reviewers verifies: 1. The
Gas-cooled fast breeder reactor shielding benchmark calculation
Energy Technology Data Exchange (ETDEWEB)
Rouse, C.A.; Mathews, D.R.; Koch, P.K.
1977-01-01
This report summarizes the results of a shielding benchmark calculation performed by General Atomic (GA) and Oak Ridge National Laboratory (ORNL). The problem analyzed was a neutron-coupled gamma ray transport calculation of the core blanket shield of the 300-MW(e) gas-cooled fast breeder reactor (GCFR). Comparison of the initial GA and ORNL results indicated good agreement for fast fluxes (E greater than 0.9 MeV and E greater than 0.086 MeV) but poor agreement for epithermal and thermal neutron fluxes. Examination of the results revealed that a deficiency in the GA fine-group cross section preparation code was responsible for the differences in the GA and ORNL iron cross sections. Modification of the GA cross sections to include self-shielding was accomplished, and the updated GA benchmark calculation performed with the self-shielded iron cross sections was in excellent agreement with the ORNL results for fast neutron fluxes with E greater than 0.9 MeV and E greater than 0.086 MeV and in good agreement for epithermal and thermal fluxes. The agreement of the gamma heating rates also improved significantly. Thus, it was concluded that the good agreement of the GA and ORNL neutron-coupled gamma ray transport calculation indicates that (1) the methods and cross sections used by both laboratories were compatible and consistent and (2) the use of 24 neutron energy groups and 15 gamma energy groups by GA was adequate compared with the use of 51 neutron energy groups and 25 gamma energy groups by ORNL.
Testing of cross section libraries for TRIGA criticality benchmark
International Nuclear Information System (INIS)
Influence of various up-to-date cross section libraries on the multiplication factor of TRIGA benchmark as well as the influence of fuel composition on the multiplication factor of the system composed of various types of TRIGA fuel elements was investigated. It was observed that keff calculated by using the ENDF/B VII cross section library is systematically higher than using the ENDF/B-VI cross section library. The main contributions (∼220 pcm) are from 235U and Zr. (author)
Calculations of EURACOS iron benchmark experiment using the HYBRID method
International Nuclear Information System (INIS)
In this paper, the HYBRID method is used in the calculations of the iron benchmark experiment at the EURACOS-II device. The saturation activities of the 32S(n,p)32P reaction at different depths in an iron block are computed with ENDF/B-IV data to compare with the measurements. At the outer layers of the iron block, the HYBRID calculation gives increasingly higher results than the VITAMIN-C multigroup calculation. With the adjustment of the two- to one-dimensional ratios, the HYBRID results agree with the measurements to within 10% at most penetration depths, a considerable improvement over the VITAMIN-C multigroup results. The development of a collapsing method for the HYBRID cross sections provides a more direct and practical way of using the HYBRID method in the two-dimensional calculations. It is observed that half of the window effect is smeared in the collapsing treatment, but it still provides a better cross-section set than the VITAMIN-C cross sections for the deep-penetration calculations
DeCART benchmark calculation for LWR next generation fuels
International Nuclear Information System (INIS)
DeCART (Deterministic Core Analysis based on Ray Tracing) is a three-dimensional whole-core transport code capable of a direct core calculation at power generating conditions. Recently, a depletion capability has been implemented into the DeCART code to predict the depleted composition in the fuel. The representative depletion methods include the exponential matrix method and the linearization method. While most of the transport lattice codes adopt the linearization method for a better efficiency in the computing time, the Monte Carlo depletion codes adopt the exponential matrix method. The drawback of the linearization method is in its fixed formulation which causes difficulties in the modification of the depletion chains and the programming itself. The drawback of the exponential matrix method is the relatively expensive computing time. However, the computing time for a depletion calculation is quite small when compared with that for the main transport calculation. Therefore, the DeCART code adopts the exponential matrix method of ORIGEN-2 for the depletion calculation. In this paper, some features of the depletion method implemented in DeCART are described first, and next the depletion capability is examined by solving a LWR next generation fuel benchmark problem
Calculations of different transmutation concepts. An international benchmark exercise
International Nuclear Information System (INIS)
In April 1996, the NEA Nuclear Science Committee (NSC) Expert Group on Physics Aspects of Different Transmutation Concepts launched a benchmark exercise to compare different transmutation concepts based on pressurised water reactors (PWRs), fast reactors, and an accelerator-driven system. The aim was to investigate the physics of complex fuel cycles involving reprocessing of spent PWR reactor fuel and its subsequent reuse in different reactor types. The objective was also to compare the calculated activities for individual isotopes as a function of time for different plutonium and minor actinide transmutation scenarios in different reactor systems. This report gives the analysis of results of the 15 solutions provided by the participants: six for the PWRs, six for the fast reactor and three for the accelerator case. Various computer codes and nuclear data libraries were applied. (author)
Galileo probe forebody thermal protection - Benchmark heating environment calculations
Balakrishnan, A.; Nicolet, W. E.
1981-01-01
Solutions are presented for the aerothermal heating environment for the forebody heatshield of candidate Galileo probe. Entry into both the nominal and cool-heavy model atmospheres were considered. Solutions were obtained for the candidate heavy probe with a weight of 310 kg and a lighter probe with a weight of 290 kg. In the flowfield analysis, a finite difference procedure was employed to obtain benchmark predictions of pressure, radiative and convective heating rates, and the steady-state wall blowing rates. Calculated heating rates for entry into the cool-heavy model atmosphere were about 60 percent higher than those predicted for the entry into the nominal atmosphere. The total mass lost for entry into the cool-heavy model atmosphere was about 146 kg and the mass lost for entry into the nominal model atmosphere was about 101 kg.
International Nuclear Information System (INIS)
In this paper we present the results of our calculations of the OECD NEA benchmark on generation-IV advanced sodium-cooled fast reactor (SFR) concepts. The aim of this benchmark is to study the core design features, moreover the feedback and transient behaviour of four SFR concepts. At the present state, static global neutronic parameters, e.g. keff, effective delayed neutron fraction, Doppler constant, sodium void worth, control rod worth, power distribution; and burnup were calculated for both the beginning and the end of cycle. In the benchmark definition, the following core descriptions were specified: two large cores (3600 MW thermal power) with carbide and oxide fuel, and two medium cores (1000 MW thermal power) with metal and oxide fuel. The calculations were performed by using the ECCO module of the ERANOS code system at the subassembly level, and with the KIKO3DMG code at the core level. The former code produced the assembly homogenized cross sections applying 1968 group collision probability calculations; the latter one determined the core multiplication factor, the radial power distribution using a 3D nodal diffusion method in 9 energy groups. We examined the effects of increasing the energy groups to 17 in the core calculation. The reflector and shield assembly homogenization methodology was also tested: a “homogeneous region model” was compared with a “concentric cylindrical core” calculation. The breeding ratio was also determined for the beginning of cycle. (author)
Benchmark on deterministic time-dependent transport calculations without spatial homogenisation
International Nuclear Information System (INIS)
The space-time neutron kinetics benchmark on deterministic transport calculations without spatial homogenization C5G7-TD has been developed and proposed for verification of codes solving the time-dependent neutron transport equation. The well-known C5G7 benchmark has been chosen as the base for new benchmark. The proposed benchmark has been calculated by SUHAM-TD code, which realizes the surface harmonic method (SHM). Authors hope to attract the attention of other researchers in order to involve them to participate in calculations of the proposed benchmark. (author)
Benchmark Evaluation of the Medium-Power Reactor Experiment Program Critical Configurations
Energy Technology Data Exchange (ETDEWEB)
Margaret A. Marshall; John D. Bess
2013-02-01
A series of small, compact critical assembly (SCCA) experiments were performed in 1962-1965 at the Oak Ridge National Laboratory Critical Experiments Facility (ORCEF) for the Medium-Power Reactor Experiment (MPRE) program. The MPRE was a stainless-steel clad, highly enriched uranium (HEU)-O2 fuelled, BeO reflected reactor design to provide electrical power to space vehicles. Cooling and heat transfer were to be achieved by boiling potassium in the reactor core and passing vapor directly through a turbine. Graphite- and beryllium-reflected assemblies were constructed at ORCEF to verify the critical mass, power distribution, and other reactor physics measurements needed to validate reactor calculations and reactor physics methods. The experimental series was broken into three parts, with the third portion of the experiments representing the beryllium-reflected measurements. The latter experiments are of interest for validating current reactor design efforts for a fission surface power reactor. The entire series has been evaluated as acceptable benchmark experiments and submitted for publication in the International Handbook of Evaluated Criticality Safety Benchmark Experiments and in the International Handbook of Evaluated Reactor Physics Benchmark Experiments.
Benchmark calculations of target heat deposition and bulk shielding
Energy Technology Data Exchange (ETDEWEB)
Takada, Hiroshi; Sakamoto, Yukio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1996-09-01
As a first step of a design study of the neutron science research center using an intense proton accelerator of 1.5 GeV with a current of 1 mA, a benchmark calculation was carried out with the NMTC/JAERI-MCNP-4A code system for the heat deposition in thick targets of Cu, Pb and U bombarded with 1.2 GeV protons. The thickness of bulk shielding around a spallation target was also estimated with the Moyer model and Sn calculation. It was found from these calculations that the code system reproduced well the experimental heat distribution around the beam axis. However, the code gave rather lower heat deposition at peripheral region of the target. As for the bulk shielding, it was estimated that the shielding made of iron having the thickness of 4 m surrounded by ordinary concrete with the thickness of 1 m was required for the 1.5 GeV proton incidence on a stopping-length Ta target with the diameter of 15 cm. (author)
ICSBEP criticality benchmarking for nuclear data validations, KAMINI, PURNIMA-II and PURNIMA-I
International Nuclear Information System (INIS)
India has contributed three experimental benchmarks to the International handbook of the International Criticality safety Benchmark Evaluation Project (ICSBEP) of the US-DOE/NEA-DB. This presentation describes the interesting experience in creating these three Indian experimental benchmarks for nuclear data and code validation studies. The concept of definition of benchmark is also reviewed for convenience. Series of sensitivity studies are performed to assess the various uncertainties that arise in knowledge of the description of the actual system
Analysis and evaluation of critical experiments for validation of neutron transport calculations
International Nuclear Information System (INIS)
The calculation schemes, computational codes and nuclear data used in neutronic design require validation to obtain reliable results. In the nuclear criticality safety field this reliability also translates into a higher level of safety in procedures involving fissile material. The International Criticality Safety Benchmark Evaluation Project is an OECD/NEA activity led by the United States, in which participants from over 20 countries evaluate and publish criticality safety benchmarks. The product of this project is a set of benchmark experiment evaluations that are published annually in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. With the recent participation of Argentina, this information is now available for use by the neutron calculation and criticality safety groups in Argentina. This work presents the methodology used for the evaluation of experimental data, some results obtained by the application of these methods, and some examples of the data available in the Handbook.
Benchmark calculations of 150-group cross section library for LMR's
International Nuclear Information System (INIS)
For the purpose of diversification of selection of cross section library for neutron calculation of LMR, the 150 multi-group cross section library was generated from ENDF-VI release. The set was then examined by analyzing measured reactivity quantities such as control rod worth, Doppler effect and sodium void effect for BFS critical assemblies that we obtained through the critical experiment plan for developing the KALIMER core design. The calculated results based on 9 group structure using the new set were also compared with those of JEF set based on the same group structure and compared with those of the same set based on 25 group structure to find the proper group structure. ENDF-VI-based set shows a small deviation in predicting measured integral quantities in comparison with the previous set and a small group effect
Shielding benchmark calculations of selected spent fuel storage cask experiments
Energy Technology Data Exchange (ETDEWEB)
Broadhead, B.L.; Tang, J.S.; Parks, C.V. (Oak Ridge National Lab., TN (United States)); Taniuchi, H. (Kobe Steel Ltd. (Japan))
1993-01-01
This paper describes the application of the three-dimensional Monte Carlo code MORSE-SGC, as implemented in the SCALE system calculational sequence SAS4, to the analysis of a series of benchmark spent fuel storage cask measurements performed at the Idaho National Engineering Laboratory. A total of five storage cask problems were analyzed to determine the expected accuracies of computational analyses using well-established Monte Carlo codes. The results presented herein represent the current status of the work. Predicted neutron dose results generally compare very favorably (within 30%) with the measurements for the cask lid, bottom, and along the cask side. Gamma-ray dose rates exhibit differing trends, depending on the measurement location. For lid and bottom doses, as well as side doses near the endfittings, agreement is again within 30%, although several exceptions are seen. However, for gamma doses along the cask side and adjacent to the active fuel, a factor of 2 overprediction is noted. Investigations into the cause of these discrepancies are currently in progress.
Shielding benchmark calculations of selected spent fuel storage cask experiments
Energy Technology Data Exchange (ETDEWEB)
Broadhead, B.L.; Tang, J.S.; Parks, C.V. [Oak Ridge National Lab., TN (United States); Taniuchi, H. [Kobe Steel Ltd. (Japan)
1993-03-01
This paper describes the application of the three-dimensional Monte Carlo code MORSE-SGC, as implemented in the SCALE system calculational sequence SAS4, to the analysis of a series of benchmark spent fuel storage cask measurements performed at the Idaho National Engineering Laboratory. A total of five storage cask problems were analyzed to determine the expected accuracies of computational analyses using well-established Monte Carlo codes. The results presented herein represent the current status of the work. Predicted neutron dose results generally compare very favorably (within 30%) with the measurements for the cask lid, bottom, and along the cask side. Gamma-ray dose rates exhibit differing trends, depending on the measurement location. For lid and bottom doses, as well as side doses near the endfittings, agreement is again within 30%, although several exceptions are seen. However, for gamma doses along the cask side and adjacent to the active fuel, a factor of 2 overprediction is noted. Investigations into the cause of these discrepancies are currently in progress.
Solution of the BEAVRS benchmark using the nTRACER direct whole core calculation code
International Nuclear Information System (INIS)
The BEAVRS (Benchmark for Evaluation and Validation of Reactor Simulation) benchmark is solved by the nTRACER direct whole core calculation code to assess its accuracy and to examine the solution dependence on modeling parameters. A sophisticated nTRACER core model representing the BEAVRS core is prepared after a series of sensitivity study to ensure solution accuracy. The resulting solutions for several hot-zero-power (HZP) states are compared first with the corresponding Monte Carlo solutions, which consist of the McCARD solutions for the assembly problems and the OpenMC solutions for the core problems, and then with the measured data which include the control rod worths (CRWs) and incore detector signals as well as the critical boron concentrations (CBC). The core depletion calculation is performed for the initial and second cycles with a set of approximated power histories and the calculated CBCs are compared with the measured data. The comparison results show that the criticality, control rod bank worths at HZP and the boron let-down curves of two cycles agree well with the measurements within 180 pcm and 25 ppm, respectively. (author)
International Nuclear Information System (INIS)
The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in 1992 by the United States Department of Energy. The project quickly became an international effort as scientists from other interested countries became involved. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) is now an official activity of the Organization for Economic Cooperation and Development-Nuclear Energy Agency (OECD-NEA). 'International Handbook of Criticality Safety Benchmark Experiments' was prepared and is updated yearly by the working group of the project. This handbook contains criticality safety benchmark specifications that have been derived from experiments that were performed at various nuclear criticality facilities around the world. However, the handbook lacks criticality data of 20 wt%-enriched uranium fuel. The author proposes to make benchmark specifications derived from modern research reactors in Asia. Future evaluations of these reactors will facilitate to fill the 'enrichment gap'. (author)
International Nuclear Information System (INIS)
In order to evaluate criticality accident analysis codes, a criticality accident benchmark problem was made based on the TRACY experiment. It is evaluated by the contributors of the expert group on criticality excursion analysis, a group of criticality safety WP of OECD/NEA/NSC. This paper reports the detail of TRACY Benchmark I and II, and preliminary results of its analysis using AGNES code. (author)
Benchmark calculations for hexagonal lattices with different methods
International Nuclear Information System (INIS)
Necessity to increase the safety conditions of exploitation of recently designed core of modern nuclear reactors causes stronger requirements to the precision of neutron-physical analysis. To get more precise characteristics of nuclear reactor cells and assembly one can increase the accuracy of neutron-physical calculation analysis by taking account the spectral effects. This paper deals with the analysis of the ZR-6 series of experiments using some components of the KARATE code system. The goal of our investigations is the comparison of measured and calculated parameters of perturbed hexagonal lattices containing Gd2O3 in Al2O3 matrix or water holes/ The quoted results include: the critical y parameters Hcr, dρ/dh and the absorber rod efficiency: Δρ. The experiments are based on doubling time measurements. The calculations have been compared not only to the measured data but to the Monte Carlo code results, too (Authors)
IPEN/MB-01 heavy reflector benchmark calculations using Serpent code
International Nuclear Information System (INIS)
A series of critical experiments with water-moderated square-pitched lattices with low-enriched uranium fuel rods was conducted at the IPEN/MB-01 research reactor facility, in 2005. Later, this data become some benchmarks. In one of these experiments the west face of the reactor core was covered with a set of thin SS-304 plates to simulate a heavy reflector as used in the EPR reactor (LEU-COMP-HERM-043). The plates are 3 mm thick and their width and axial length were large enough to cover one whole side of the active core of the reactor. The critical configurations were found as a function of the number of plates. Fuel rods containing UO2 with uranium enriched to 4.3% 235U were arranged in specific geometric configurations to be as close as possible to the critical state. In this work, these benchmark configurations with heavy reflectors were modeled using the Serpent Monte Carlo Code. Serpent uses a universe-based geometry model, which allows the description of practically any three-dimensional fuel or reactor configuration. Neutron transport is based on a combination of surface-to-surface ray-tracing and the Woodcock delta-tracking method. Woodcock method is many times faster than ray-tracing, so compared to MCNP code, Serpent code can bring huge gains in processing time of reactor calculations and reaction rate calculations. The results of these calculations were compared with experimental data and calculations with codes MCNP5 and SCALE6 (KENO-VI) using ENDF/B-VII.0 as cross-section input data. The codes performances are compared in terms of CPU calculation time and agreement with experimental data. Additional y, sensitivity on keff of Serpent woodcock threshold parameter was analyzed. (author)
Benchmark Calculations of OECD/NEA Reactivity-Initiated Accidents
International Nuclear Information System (INIS)
The benchmark- Phase I was done from 2011 to 2013 with a consistent set of four experiments on very similar highly irradiated fuel rods tested under different experimental conditions: low temperature, low pressure, stagnant water coolant, very short power pulse (NSRR VA-1), high temperature, medium pressure, stagnant water coolant, very short power pulse (NSRR VA-3), high temperature, low pressure, flowing sodium coolant, larger power pulse (CABRI CIP0-1), high temperature, high pressure, flowing water coolant, medium width power pulse (CABRI CIP3-1). Based on the importance of the thermal-hydraulics aspects revealed during the Phase I, the specifications of the benchmark-Phase II was elaborated in 2014. The benchmark-Phase II focused on the deeper understanding of the differences in modeling of the different codes. The work on the benchmark- Phase II program will last the end of 2015. The benchmark cases for RIA are simulated with the code of FRAPTRAN 1.5, in order to understand the phenomena during RIA and to check the capacity of the code itself. The results of enthalpy, cladding strain and outside temperature among 21 parameters asked by the benchmark program are summarized, and they seem to reasonably reflect the actual phenomena, except for them of case 6
Analysis of the international criticality benchmark no 19 of a realistic fuel dissolver
International Nuclear Information System (INIS)
The dispersion of the order of 12000 pcm in the results of the international criticality fuel dissolver benchmark calculation, exercise OECD/19, showed the necessity of analysing the calculational methods used in this case. The APOLLO/PIC method developed to treat this type of problem permits us to propose international reference values. The problem studied here, led us to investigate two supplementary parameters in addition to the double heterogeneity of the fuel: the reactivity variation as a function of moderation and the effects of the size of the fuel pellets during dissolution. The following conclusions were obtained: The fast cross-section sets used by the international SCALE package introduces a bias of - 3000 pcm in undermoderated lattices. More generally, the fast and resonance nuclear data in criticality codes are not sufficiently reliable. Geometries with micro-pellets led to an underestimation of reactivity at the end of dissolution of 3000 pcm in certain 1988 Sn calculations; this bias was avoided in the up-dated 1990 computation because of a correct use of calculation tools. The reactivity introduced by the dissolved fuel is underestimated by 3000 pcm in contributions based on the standard NITAWL module in the SCALE code. More generally, the neutron balance analysis pointed out that standard ND self shielding formalism cannot account for 238U resonance mutual self-shielding in the pellet-fissile liquor interaction. The combination of these three types of bias explain the underestimation of all of the international contributions of the reactivity of dissolver lattices by -2000 to -6000 pcm. The improved 1990 calculations confirm the need to use rigorous methods in the calculation of systems which involve the fuel double heterogeneity. This study points out the importance of periodic benchmarking exercises for probing the efficacity of criticality codes, data libraries and the users
Benchmark calculations by the thermal reactor standard nuclear design code system SRAC
International Nuclear Information System (INIS)
This report summarizes the present status of the thermal reactor standard nuclear design code system SRAC developed by the nuclear design working group of the JAERI thermal reactor standard code committee which was started on July 1978. Descriptions are given at first on the brief introduction and the process of development of the code system SRAC, and then, the several benchmark tests performed to evaluate the performance of the code system. The results show the good predictions of the experimental keff values of the critical facilities; TCA for LWR, JMTRC for JAERI MTR, DCA for the Japanese Advanced Thermal Reactor and SHE for VHTR. A trial to the IAEA benchmark calculations on the Reduction of uranium Enrichment of Research and Test Reactors yields satisfactory agreements with the results of ANL. Another test to evaluate the fast group constants was also attempted by tracing the fast reactor benchmark problems which have been used to evaluate nuclear data file in the FBR reactor physics field. (author)
International Nuclear Information System (INIS)
Highlights: • Performance estimation of nuclear-data benchmark was investigated. • Point detector contribution played a benchmark role not only to the neutron producing the detector contribution but also equally to all the upstream transport neutrons. • New functions were defined to give how well the contribution could be interpreted for benchmarking. • Benchmark performance could be evaluated only by a forward Monte Carlo calculation. -- Abstract: The author's group has been investigating how the performance estimation of nuclear-data benchmark using experiment and its analysis by Monte Carlo code should be carried out especially at 14 MeV. We have recently found that a detector contribution played a benchmark role not only to the neutron producing the detector contribution but also equally to all the upstream neutrons during the neutron history. This result would propose that the benchmark performance could be evaluated only by a forward Monte Carlo calculation. In this study, we thus defined new functions to give how well the contribution could be utilized for benchmarking using the point detector, and described that it was deeply related to the newly introduced “partial adjoint contribution”. By preparing these functions before benchmark experiments, one could know beforehand how well and for which nuclear data the experiment results could do benchmarking in forward Monte Carlo calculations
International Nuclear Information System (INIS)
Some high-quality reactor physics benchmark experiments are being re-evaluated with today's state-of-the-art methods, particularly using that of detailed 3-dimensional models. One experiment analysed in the framework of the International Reactor Physics Benchmark Experiments (IRPhE) project is SNEAK-7A. This assembly is characterised by a Pu-fuelled fast critical assembly in the Karlsruhe Fast Critical Facility for the purpose of testing cross section data and calculational methods. As the detailed information on the SNEAK-7A benchmark experiment becomes available, the purpose of this paper is to model this experiment as closely as possible to the configuration as it existed in the critical facility. The experimental keff was determined to be 1.0010, which is 29.6 cents supercritical. The realistic modelling of the SNEAK-7A assembly was performed using the DANTSYS code capability for X-Y-Z geometry. The calculated core eigenvalue from THREEDANT is 1.00975. With corrections applied for core plate cell heterogeneity and mesh sizes, the best-estimate core criticality with JEF-2.2-based cross-sections turns out to be 1.01137. While the plate heterogeneity effect from flux redistribution was at first estimated to be as large as 387 pcm from plate cell calculations, it proves to be 142 pcm when the core-wide heterogeneity effects are accounted for. In order to figure out the over-prediction of core eigenvalue, spectral indices are investigated, which led to projecting that the 238U capture cross-sections are being underestimated. This fact is confirmed in the comparison of the central material worth of 238U with the measured value. When the sensitivity of core eigenvalue to the cross section is used, the newly estimated core eigenvalue is 1.00175, which is very close to the measured core eigenvalue, when the 238U capture cross-section was assumed to increase by 5% implied from the comparison of spectral indices. Once the details in the old critical experiments are
Energy Technology Data Exchange (ETDEWEB)
Kim, S.J. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kodeli, I.; Sartori, E. [OECD NEA DataBank, 92 - Issy les Moulineaux (France)
2003-07-01
Some high-quality reactor physics benchmark experiments are being re-evaluated with today's state-of-the-art methods, particularly using that of detailed 3-dimensional models. One experiment analysed in the framework of the International Reactor Physics Benchmark Experiments (IRPhE) project is SNEAK-7A. This assembly is characterised by a Pu-fuelled fast critical assembly in the Karlsruhe Fast Critical Facility for the purpose of testing cross section data and calculational methods. As the detailed information on the SNEAK-7A benchmark experiment becomes available, the purpose of this paper is to model this experiment as closely as possible to the configuration as it existed in the critical facility. The experimental keff was determined to be 1.0010, which is 29.6 cents supercritical. The realistic modelling of the SNEAK-7A assembly was performed using the DANTSYS code capability for X-Y-Z geometry. The calculated core eigenvalue from THREEDANT is 1.00975. With corrections applied for core plate cell heterogeneity and mesh sizes, the best-estimate core criticality with JEF-2.2-based cross-sections turns out to be 1.01137. While the plate heterogeneity effect from flux redistribution was at first estimated to be as large as 387 pcm from plate cell calculations, it proves to be 142 pcm when the core-wide heterogeneity effects are accounted for. In order to figure out the over-prediction of core eigenvalue, spectral indices are investigated, which led to projecting that the {sup 238}U capture cross-sections are being underestimated. This fact is confirmed in the comparison of the central material worth of {sup 238}U with the measured value. When the sensitivity of core eigenvalue to the cross section is used, the newly estimated core eigenvalue is 1.00175, which is very close to the measured core eigenvalue, when the {sup 238}U capture cross-section was assumed to increase by 5% implied from the comparison of spectral indices. Once the details in the old critical
International Nuclear Information System (INIS)
At the nineteenth AER symposium a benchmark on core burnup calculations for WWER-1000 reactors was proposed for further validation and verification of the reactor physics code systems. The work was continued in the framework of a project supported by the German BMU3). During the preparation of the calculations results corrections, refinement and additions the benchmark specification were done. The benchmark includes two stages: the first step comprises the data library preparation for all fuel assembly types used in the core loadings. The second step consists of the 3D core burnup calculation together with calculations of critical states for hot zero power conditions. The benchmark specification contains the description of the fuel assemblies (FA) for the few group data preparation, the core loading patterns and the load follow as well as a set of reference data such as boron acid concentration in the coolant, cycle length, measured reactivity coefficients and power density distributions for successive cycles of a WWER-1000 reactor core. Different reactor physics codes were used to produce solutions. FA burnup codes such as NESSEL, CASMO or HELIOS were used for data preparation. The core calculations were performed using codes such as DYN3D, TRAPEZ as well as several data libraries. The results of the calculations made by different organisations (IBBS, FZD, SSTC) are presented and discussed. The data needed to produce solutions as well as most of the calculated data are attached in the appendices of the paper presented. (Authors)
Preparation of a criticality benchmark based on experiments performed at the RA-6 reactor
International Nuclear Information System (INIS)
The operation and fuel management of a reactor uses neutronic modeling to predict its behavior in operational and accidental conditions. This modeling uses computational tools and nuclear data that must be contrasted against benchmark experiments to ensure its accuracy. These benchmarks have to be simple enough to be possible to model with the desired computer code and have quantified and bound uncertainties. The start-up of the RA-6 reactor, final stage of the conversion and renewal project, allowed us to obtain experimental results with fresh fuel. In this condition the material composition of the fuel elements is precisely known, which contributes to a more precise modeling of the critical condition. These experimental results are useful to evaluate the precision of the models used to design the core, based on U3Si2 and cadmium wires as burnable poisons, for which no data was previously available. The analysis of this information can be used to validate models for the analysis of similar configurations, which is necessary to follow the operational history of the reactor and perform fuel management. The analysis of the results and the generation of the model were done following the methodology established by International Criticality Safety Benchmark Evaluation Project, which gathers and analyzes experimental data for critical systems. The results were very satisfactory resulting on a value for the multiplication factor of the model of 1.0000 ± 0.0044, and a calculated value of 0.9980 ± 0.0001 using MCNP 5 and ENDF/B-VI. The utilization of as-built dimensions and compositions, and the sensitivity analysis allowed us to review the design calculations and analyze their precision, accuracy and error compensation.
Benchmark calculations in multigroup and multidimensional time-dependent transport
International Nuclear Information System (INIS)
It is widely recognized that reliable benchmarks are essential in many technical fields in order to assess the response of any approximation to the physics of the problem to be treated and to verify the performance of the numerical methods used. The best possible benchmarks are analytical solutions to paradigmatic problems where no approximations are actually introduced and the only error encountered is connected to the limitations of computational algorithms. Another major advantage of analytical solutions is that they allow a deeper understanding of the physical features of the model, which is essential for the intelligent use of complicated codes. In neutron transport theory, the need for benchmarks is particularly great. In this paper, the authors propose to establish accurate numerical solutions to some problems concerning the migration of neutron pulses. Use will be made of the space asymptotic theory, coupled with a Laplace transformation inverted by a numerical technique directly evaluating the inversion integral
OECD/NEA burnup credit criticality benchmark. Result of phase IIA
International Nuclear Information System (INIS)
The report describes the final result of the Phase IIA of the Burnup Credit Criticality Benchmark conducted by OECD/NEA. In the Phase IIA benchmark problems, the effect of an axial burnup profile of PWR spent fuels on criticality (end effect) has been studied. The axial profiles at 10, 30 and 50 GWd/t burnup have been considered. In total, 22 results from 18 institutes of 10 countries have been submitted. The calculated multiplication factors from the participants have lain within the band of ± 1% Δk. For the irradiation up to 30 GWd/t, the end effect has been found to be less than 1.0% Δk. But, for the 50 GWd/t case, the effect is more than 4.0% Δk when both actinides and FPs are taken into account, whereas it remains less than 1.0% Δk when only actinides are considered. The fission density data have indicated the importance end regions have in the criticality safety analysis of spent fuel systems. (author)
OECD/NEA burnup credit criticality benchmark. Result of phase IIA
Energy Technology Data Exchange (ETDEWEB)
Takano, Makoto; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1996-02-01
The report describes the final result of the Phase IIA of the Burnup Credit Criticality Benchmark conducted by OECD/NEA. In the Phase IIA benchmark problems, the effect of an axial burnup profile of PWR spent fuels on criticality (end effect) has been studied. The axial profiles at 10, 30 and 50 GWd/t burnup have been considered. In total, 22 results from 18 institutes of 10 countries have been submitted. The calculated multiplication factors from the participants have lain within the band of {+-} 1% {Delta}k. For the irradiation up to 30 GWd/t, the end effect has been found to be less than 1.0% {Delta}k. But, for the 50 GWd/t case, the effect is more than 4.0% {Delta}k when both actinides and FPs are taken into account, whereas it remains less than 1.0% {Delta}k when only actinides are considered. The fission density data have indicated the importance end regions have in the criticality safety analysis of spent fuel systems. (author).
International Nuclear Information System (INIS)
One of the important issues regarding deterministic transport methods for whole core calculations is that homogenized techniques can introduce errors into results. On the other hand, with modern computation abilities, direct whole core heterogeneous calculations are becoming increasingly feasible. This report provides an analysis of the results obtained from a challenging benchmark on deterministic MOX fuel assembly transport calculations without spatial homogenization. A majority of the participants obtained solutions that were more than acceptable for typical reactor calculations. The report will be of particular interest to reactor physicists and transport code developers. (author)
Calculation of SPERT Reactor benchmarks using 3D diffusion code DIREN
International Nuclear Information System (INIS)
The three dimensional diffusion code DIREN was developed at Institute for Nuclear Research (INR) Pitesti for reactor physics calculations for natural uranium and advanced CANDU reactors. Cell codes used are WIMS (from NEA library) and DRAGON (available in open source system). The latter is used also for super cell modeling of reactor control devices. These codes and the auxiliary programs were linked together in a calculation system. In order to apply WIMS-DRAGON-DIREN system to LWR, first the reactor SPERT benchmarks problems were calculated. The core including the control rods was modeled in three dimensional geometry. Following the calculations of the critical height (Hcrit), three dimensional power and flux distributions were obtained. The standard procedure used for CANDU reactor calculations (incremental cross sections for reactivity devices) underestimated the worth of control rods. A simple procedure to obtain the internal boundary conditions was developed using the super cell code DRAGON. Also the DIREN 3D diffusion code was modified to apply inner boundary conditions at control rods assigned volumes. Applying the inner boundary conditions yielded results closer to the measured values (e.g. the measured Hcrit was 49.53 cm as compared to 53.15 cm, the calculated one on 7 groups for nominal temperature). The reactivity coefficients for temperature and density required in transient's simulations were also calculated. The sample test problem T83 (hot stand-by, fast transient) was simulated using the RELAP code. (authors)
Calculational Benchmark Problems for VVER-1000 Mixed Oxide Fuel Cycle
Energy Technology Data Exchange (ETDEWEB)
Emmett, M.B.
2000-03-17
Standard problems were created to test the ability of American and Russian computational methods and data regarding the analysis of the storage and handling of Russian pressurized water reactor (VVER) mixed oxide fuel. Criticality safety and radiation shielding problems were analyzed. Analysis of American and Russian multiplication factors for fresh fuel storage for low-enriched uranium (UOX), weapons- (MOX-W) and reactor-grade (MOX-R) MOX differ by less than 2% for all variations of water density. For shielding calculations for fresh fuel, the ORNL results for the neutron source differ from the Russian results by less than 1% for UOX and MOX-R and by approximately 3% for MOX-W. For shielding calculations for fresh fuel assemblies, neutron dose rates at the surface of the assemblies differ from the Russian results by 5% to 9%; the level of agreement for gamma dose varies depending on the type of fuel, with UOX differing by the largest amount. The use of different gamma group structures and instantaneous versus asymptotic decay assumptions also complicate the comparison. For the calculation of dose rates from spent fuel in a shipping cask, the neutron source for UOX after 3-year cooling is within 1% and for MOX-W within 5% of one of the Russian results while the MOX-R difference is the largest at over 10%. These studies are a portion of the documentation required by the Russian nuclear regulatory authority, GAN, in order to certify Russian programs and data as being acceptably accurate for the analysis of mixed oxide fuels.
Results of the isotopic concentrations of VVER calculational burnup credit benchmark No. 2(CB2)
International Nuclear Information System (INIS)
Results of the nuclide concentrations are presented of VVER Burnup Credit Benchmark No. 2(CB2) that were performed in The Nuclear Technology Center of Cuba with available codes and libraries. The CB2 benchmark specification as the second phase of the VVER burnup credit benchmark is summarized. The CB2 benchmark focused on VVER burnup credit study proposed on the 97' AER Symposium. The obtained results are isotopic concentrations of spent fuel as a function of the burnup and cooling time. The depletion point 'ORIGEN2' code and other codes were used for the calculation of the spent fuel concentration. (author)
PWR assembly transport calculation: A validation benchmark using DRAGON, PENTRAN, and MCNP
International Nuclear Information System (INIS)
This paper presents a 2D PWR fuel assembly benchmark performed with 3 transport codes: DRAGON which uses the collision probability method, PENTRAN, an Sn transport code, and MCNP, a Monte Carlo code. First, DRAGON was used to produce a 2-group pin-by-pin cross-section library associated with 45 materials that describe the fuel assembly. Using the same library, it was then possible to perform comparisons between DRAGON and MCNP, and between PENTRAN and MCNP. Here, MCNP was considered as the reference multigroup Monte Carlo tool used to validate the deterministic codes. This type of 2-group benchmark can be utilized to evaluate the performance of different solvers using the very same cross-sections. The transport solutions provided here May be used as references for further comparisons with industrial reactor core codes using a diffusion or a SPn solver, and generally relying on 2-group cross-sections. Results show an excellent overall agreement between the 3 codes, with discrepancies that are less than 0.5% on the pin-by-pin flux, and less than 20 pcm on the keff. Therefore, it May be concluded that these deterministic codes are reliable tools to perform criticality transport calculations for PWR lattices. Moreover, the use of multigroup Monte Carlo appears as an efficient independent technique to perform detailed code to code comparisons relying on the same cross-section library. The present work May be considered as the first step of a 3D PWR core benchmark using DRAGON generated cross-sections and comparing PENTRAN and MCNP multigroup calculations. (authors)
Benchmark calculations for reduced density-matrix functional theory
Lathiotakis, N.N.; Marques, Miguel A. L.
2008-01-01
Reduced density-matrix functional theory (RDMFT) is a promising alternative approach to the problem of electron correlation. Like standard density functional theory, it contains an unknown exchange-correlation functional, for which several approximations have been proposed in the last years. In this article, we benchmark some of these functionals in an extended set of molecules with respect to total and atomization energies. Our results show that the most recent RDMFT functionals give very sa...
The Activities of the International Criticality Safety Benchmark Evaluation Project (ICSBEP)
Energy Technology Data Exchange (ETDEWEB)
Briggs, Joseph Blair
2001-10-01
The International Criticality Safety Benchmark Evaluation Project (ICSBEP) was initiated in 1992 by the United States Department of Energy. The ICSBEP became an official activity of the Organization for Economic Cooperation and Development (OECD) – Nuclear Energy Agency (NEA) in 1995. Representatives from the United States, United Kingdom, France, Japan, the Russian Federation, Hungary, Republic of Korea, Slovenia, Yugoslavia, Kazakhstan, Spain, and Israel are now participating. The purpose of the ICSBEP is to identify, evaluate, verify, and formally document a comprehensive and internationally peer-reviewed set of criticality safety benchmark data. The work of the ICSBEP is published as an OECD handbook entitled “International Handbook of Evaluated Criticality Safety Benchmark Experiments”. The 2001 Edition of the Handbook contains benchmark specifications for 2642 critical or subcritical configurations that are intended for use in validation efforts and for testing basic nuclear data.
International Nuclear Information System (INIS)
In the reactor physics calculation, solutions for the neutron transport equation are obtained mostly by the discrete ordinates method, referred as an SN method. A number of computer codes that use SN method require regular mesh (such as rectangular, cylindrical or spherical) to model the problems geometry. Using such a specific regular mesh leads to the simplest difference equations but may require an excessive number of mesh points to describe complicated geometries adequately. The MUST (Multi-group Unstructured geometry SN Transport) code uses unstructured tetrahedral elements so that it can be applied to solve complicated geometry. However, even the simple criticality benchmark problems (i.e., Godiva and VERA1B) can be difficult ones due to a curved surface. When a curved surface is meshed with tetrahedral elements, original volume may not be conserved because curved surface is modeled with several faces of tetrahedral elements. Instead of conserving volume, in this paper, an equivalent mass technique is applied to the criticality benchmark problems and the effects of it are shown
Computer simulation of Masurca critical and subcritical experiments. Muse-4 benchmark. Final report
International Nuclear Information System (INIS)
The efficient and safe management of spent fuel produced during the operation of commercial nuclear power plants is an important issue. In this context, partitioning and transmutation (P and T) of minor actinides and long-lived fission products can play an important role, significantly reducing the burden on geological repositories of nuclear waste and allowing their more effective use. Various systems, including existing reactors, fast reactors and advanced systems have been considered to optimise the transmutation scheme. Recently, many countries have shown interest in accelerator-driven systems (ADS) due to their potential for transmutation of minor actinides. Much R and D work is still required in order to demonstrate their desired capability as a whole system, and the current analysis methods and nuclear data for minor actinide burners are not as well established as those for conventionally-fuelled systems. Recognizing a need for code and data validation in this area, the Nuclear Science Committee of the OECD/NEA has organised various theoretical benchmarks on ADS burners. Many improvements and clarifications concerning nuclear data and calculation methods have been achieved. However, some significant discrepancies for important parameters are not fully understood and still require clarification. Therefore, this international benchmark based on MASURCA experiments, which were carried out under the auspices of the EC 5. Framework Programme, was launched in December 2001 in co-operation with the CEA (France) and CIEMAT (Spain). The benchmark model was oriented to compare simulation predictions based on available codes and nuclear data libraries with experimental data related to TRU transmutation, criticality constants and time evolution of the neutronic flux following source variation, within liquid metal fast subcritical systems. A total of 16 different institutions participated in this first experiment based benchmark, providing 34 solutions. The large number
Calculation of the CB1 burnup credit benchmark reaction rates with MCNP4B
International Nuclear Information System (INIS)
The first calculational VVER-440 burnup credit benchmark CB1 in 1996. VTT Energy participated in the calculation of the CB1 benchmark with three different codes: CASMO-4, KENO-VI and MCNP4B. However, the reaction rates and the fission ν were calculated only with CASMO-4. Now, the neutron absorption and production reaction rates and the fission ν values have been calculated at VTT Energy with the MCNP4B Monte Carlo code using the ENDF60 neutron data library. (author)
A proposal of a benchmark for calculation of the power distribution next to the absorber
International Nuclear Information System (INIS)
A proposal of a new benchmark problem was formulated to consider the characteristics of the VVER-440 fuel assembly with enrichment zoning, i.e. to study the space dependence of the power distribution near to a control assembly. A quite detailed geometry and the material composition of the fuel and the control assemblies were modeled by the help of MCNP calculations in AEKI. The results of the MCNP calculations were built in the KARATE code system as the new albedo matrices. The comparison of the KARATE calculation results and the MCNP calculations for this benchmark is presented. (author)
ZPPR-21 critical benchmark analyses with ENDF/B-V and -VII data
International Nuclear Information System (INIS)
The six benchmark problems for the ZPPR-21 critical experiments phases A through F were analyzed using the ENDF/B-V2 and ENDF/B-VII.0 data. For reference calculations, Monte Carlo simulations were performed using the VIM code with the continuous energy cross sections. For deterministic calculations, composition and region dependent multi-group cross sections were generated using the ETOE2-2/MC2-2/SDX code system and core calculations were performed with the TWODANT discrete ordinate transport code. Based on sensitivity studies, deterministic core calculations were carried out with 230 energy groups, ∼1 cm spatial mesh size, S24 angular approximation, and P5 anisotropic scattering order Comparisons showed that the core multiplication factor determined with TWODANT agreed well with the VIM Monte Carlo solution within 0.20 %Δk for the ENDF/B-V2 data and 0.27 %Δk for the ENDF/B-VII.0 data. Detailed comparisons of reaction rates were also made between VIM and TWODANT solutions. The results showed that the multigroup cross sections generated with MC2-2 accurately reproduced the isotopic reaction rates of VIM calculations. (authors)
International Nuclear Information System (INIS)
Characteristics and contents of the CSRL-V (Criticality Safety Reference Library based on ENDF/B-V data) 227-neutron-group AMPX master and pointwise cross-section libraries are described. Results obtained in using CSRL-V to calculate performance parameters of selected thermal reactor and criticality safety benchmarks are discussed
Development of neutral transport lattice code DENT-2D and benchmark calculation
International Nuclear Information System (INIS)
We developed new transport lattice code called DENT-2D (Deterministic Neutral Particle Transport Code in 2-D imensional Space)primarily to generate few- group constants for the reactor physics analysis diffusion codes. This code is designed to be coupled with KAERI reactor analysis nodal code, MASTER [1] ,to complete the design system package. CASMO-3 and HELIOS have been used in generating the few- group constant for MASTER. Currently DENT-2D includes only neutron particle transport calculation in 2-dimensional Cartesian geometry. The characteristics method is adopted for the spatial discretization, which is advantageous for the treatment of the complicated geometry structure and the highly anisotropic scattering. The subgroup method is used for the resonance treatment. B1 approximation has been used to obtain the criticality spectrum considering the leakage effect in the real core situation. The exponential matrix method has been used for the depletion calculation. The results of benchmark calculations show that the prediction capability of DENT-2D is comparable to the other lattice codes such as HELIOS and CASMO-3
Monte Carlo Calculations of Pebble Bed Benchmark Configurations of the PROTEUS Facility
International Nuclear Information System (INIS)
Under the auspices of the International Atomic Energy Agency, a series of well-documented benchmark experiments were performed at the Proteus facility of the Swiss Paul Scherrer Institute. Thirteen critical pebble bed reactor configurations were assembled, with ten of them deterministic with a precise location of the low-enriched fuel and moderator pebbles. Seven of these configurations were modeled with a very high spatial resolution with the Monte Carlo code MCNP with details that go from the fuel kernel (0.5 mm in diameter) to the walls surrounding the facility. The calculations of the k's of the configurations agree quite well with the experiments (within a fraction of a dollar). A sensitivity analysis is included to discuss the possibility of a small bias; also biases introduced by customary approximations of production codes were calculated. The experiments and the analysis of this paper might be very useful tools to check the calculational accuracy of procedures used in the emerging work related to pebble bed modular gas-cooled reactors
Benchmark analysis of the DeCART MOC code with the VENUS-2 critical experiment
International Nuclear Information System (INIS)
Computational benchmarks based on well-defined problems with a complete set of input and a unique solution are often used as a means of verifying the reliability of numerical solutions. VENUS is a widely used MOX benchmark problem for the validation of numerical methods and nuclear data set. In this paper, the results of benchmarking the DeCART (Deterministic Core Analysis based on Ray Tracing) integral transport code is reported using the OECD/NEA VENUS-2 MOX benchmark problem. Both 2-D and 3-D DeCART calculations were performed and comparisons are reported with measured data, as well as with the results of other benchmark participants. In general the DeCART results agree well with both the experimental data as well as those of other participants. (authors)
Benchmarking Monte Carlo codes for criticality safety using subcritical measurements
International Nuclear Information System (INIS)
Monte Carlo codes that are used for criticality safety evaluations are typically validated using critical experiments in which the neutron multiplication factor is unity. However, the conditions for most fissile material operations do not coincide to those of the critical experiments. This paper demonstrates that Monte Carlo methods and nuclear data can be validated using subcritical measurements whose conditions may coincide more closely to actual configurations of fissile material. (orig.)
Energy Technology Data Exchange (ETDEWEB)
NONE
1998-06-01
This volume of the progress report provides documentation of reactor physics and criticality safety studies conducted in the Russian Federation during fiscal year 1997 and sponsored by the Fissile Materials Disposition Program of the US Department of Energy. Descriptions of computational and experimental benchmarks for the verification and validation of computer programs for neutron physics analyses are included. All benchmarks include either plutonium, uranium, or mixed uranium and plutonium fuels. Calculated physics parameters are reported for all of the contaminated benchmarks that the United States and Russia mutually agreed in November 1996 were applicable to mixed-oxide fuel cycles for light-water reactors.
Results of the isotopic concentrations of VVER calculational burnup credit benchmark no. 2(cb2
International Nuclear Information System (INIS)
The characterization of the irradiated fuel materials is becoming more important with the Increasing use of nuclear energy in the world. The purpose of this document is to present the results of the nuclide concentrations calculated Using Calculation VVER Burnup Credit Benchmark No. 2(CB2). The calculations were Performed in The Nuclear Technology Center of Cuba. The CB2 benchmark specification as the second phase of the VVER burnup credit benchmark is Summarized in [1]. The CB2 benchmark focused on VVER burnup credit study proposed on the 97' AER Symposium [2]. It should provide a comparison of the ability of various code systems And data libraries to predict VVER-440 spent fuel isotopes (isotopic concentrations) using Depletion analysis. This phase of the benchmark calculations is still in progress. CB2 should be finished by summer 1999 and evaluated results could be presented on the next AER Symposium. The obtained results are isotopic concentrations of spent fuel as a function of the burnup and Cooling time. The depletion point ORIGEN2[3] code was used for the calculation of the spent Fuel concentration. The depletion analysis was performed using the VVER-440 irradiated fuel assemblies with in-core Irradiation time of 3 years, burnup of the 30000 mwd/TU, and an after discharge cooling Time of 0 and 1 year. This work also comprises the results obtained by other codes[4].
Benchmark calculations for elastic fermion-dimer scattering
Bour, Shahin; Lee, Dean; Meißner, Ulf-G
2012-01-01
We present continuum and lattice calculations for elastic scattering between a fermion and a bound dimer in the shallow binding limit. For the continuum calculation we use the Skorniakov-Ter-Martirosian (STM) integral equation to determine the scattering length and effective range parameter to high precision. For the lattice calculation we use the finite-volume method of L\\"uscher. We take into account topological finite-volume corrections to the dimer binding energy which depend on the momentum of the dimer. After subtracting these effects, we find from the lattice calculation kappa a_fd = 1.174(9) and kappa r_fd = -0.029(13). These results agree well with the continuum values kappa a_fd = 1.17907(1) and kappa r_fd = -0.0383(3) obtained from the STM equation. We discuss applications to cold atomic Fermi gases, deuteron-neutron scattering in the spin-quartet channel, and lattice calculations of scattering for nuclei and hadronic molecules at finite volume.
Benchmark calculations on residue production within the EURISOL DS project; Part I: thin targets
David, J.C; Boudard, A; Doré, D; Leray, S; Rapp, B; Ridikas, D; Thiollière, N
Report on benchmark calculations on residue production in thin targets. Calculations were performed using MCNPX 2.5.0 coupled to a selection of reaction models. The results were compared to nuclide production cross-sections measured in GSI in inverse kinematics
Benchmark calculations on residue production within the EURISOL DS project; Part II: thick targets
David, J.-C; Boudard, A; Doré, D; Leray, S; Rapp, B; Ridikas, D; Thiollière, N
Benchmark calculations on residue production using MCNPX 2.5.0. Calculations were compared to mass-distribution data for 5 different elements measured at ISOLDE, and to specific activities of 28 radionuclides in different places along the thick target measured in Dubna.
Benchmark calculations of the shielding constants in the water dimer
Pecul, Magdalena; Lewandowski, Józef; Sadlej, Joanna
2001-01-01
The NMR shielding constants in (H 2O) 2 have been calculated using GIAO-SCF, MP2, MP4 and CCSD methods and for a range of basis sets. According to the obtained results the 6-311++G ** or aug-cc-pVDZ basis sets are recommended for SCF calculations, and the aug-cc-pVXZ series is suggested for correlated calculations of the interaction-induced changes in the shielding constants. The counterpoise correction improves the results towards the basis set limit and is essential in the case of 17O shielding. Correlation effects are substantial for the changes in 17O shielding, less so for 1H shielding. They are overestimated by the MP2 method.
JNC results of BFS-62-3A benchmark calculation (CRP: Phase 5)
International Nuclear Information System (INIS)
The present work is the results of JNC, Japan, for the Phase 5 of IAEA CRP benchmark problem (BFS-62-3A critical experiment). Analytical Method of JNC is based on Nuclear Data Library JENDL-3.2; Group Constant Set JFS-3-J3.2R: 70-group, ABBN-type self-shielding factor table based on JENDL-3.2; Effective Cross-section - Current-weighted multigroup transport cross-section. Cell model for the BFS as-built tube and pellets was (Case 1) Homogeneous Model based on IPPE definition; (Case 2) Homogeneous atomic density equivalent to JNC's heterogeneous calculation only to cross-check the adjusted correction factors; (Case 3) Heterogeneous model based on JNC's evaluation, One-dimensional plate-stretch model with Tone's background cross-section method (CASUP code). Basic diffusion Calculation was done in 18-groups and three-dimensional Hex-Z model (by the CITATION code), with Isotropic diffusion coefficients (Case 1 and 2), and Benoist's anisotropic diffusion coefficients (Case 3). For sodium void reactivity, the exact perturbation theory was applied both to basic calculation and correction calculations, ultra-fine energy group correction - approx. 100,000 group constants below 50 keV, and ABBN-type 175 group constants with shielding factors above 50 keV. Transport theory and mesh size correction 18-group, was used for three-dimensional Hex-Z model (the MINIHEX code based on the S4-P0 transport method, which was developed by JNC. Effective delayed Neutron fraction in the reactivity scale was fixed at 0.00623 by IPPE evaluation. Analytical Results of criticality values and sodium void reactivity coefficient obtained by JNC are presented. JNC made a cross-check of the homogeneous model and the adjusted correction factors submitted by IPPE, and confirmed they are consistent. JNC standard system showed quite satisfactory analytical results for the criticality and the sodium void reactivity of BFS-62-3A experiment. JNC calculated the cross-section sensitivity coefficients of BFS
Calculations with ANSYS/FLOTRAN to a core catcher benchmark
International Nuclear Information System (INIS)
There are numerous experiments for the exploration of the corium spreading behaviour, but comparable data have not been available up to now in the field of the long-term behaviour of a corium expanded in a core catcher. For the calculations a pure liquid oxidic melt with a homogeneous internal heat source was assumed. The melt was distributed uniformly over the spreading area of the EPR core catcher. All codes applied the well known k-ε-turbulence-model to simulate the turbulent flow regime of this melt configuration. While the FVM-code calculations were performed with three dimensional models using a simple symmetry, the problem was modelled two-dimensionally with ANSYS due to limited CPU performance. In addition, the 2D results of ANSYS should allow a comparison for the planned second stage of the calculations. In this second stage, the behaviour of a segregated metal oxide melt should be examined. However, first estimates and pre-calculations showed that a 3D simulation of the problem is not possible with any of the codes due to lacking computer performance. (orig.)
International Nuclear Information System (INIS)
A criticality benchmark experiment performed at the Jozef Stefan Institute TRIGA Mark II research reactor is described. This experiment and its evaluation are given as examples of benchmark experiments at research reactors. For this reason the differences and possible problems compared to other benchmark experiments are particularly emphasized. General guidelines for performing criticality benchmarks in research reactors are given. The criticality benchmark experiment was performed in a normal operating reactor core using commercially available fresh 20% enriched fuel elements containing 12 wt% uranium in uranium-zirconium hydride fuel material. Experimental conditions to minimize experimental errors and to enhance computer modeling accuracy are described. Uncertainties in multiplication factor due to fuel composition and geometry data are analyzed by sensitivity analysis. The simplifications in the benchmark model compared to the actual geometry are evaluated. Sample benchmark calculations with the MCNP and KENO Monte Carlo codes are given
Calculating the Fuzzy Project Network Critical Path
Directory of Open Access Journals (Sweden)
Nasser Shahsavari Pour
2012-04-01
Full Text Available A project network consists of various activities. To determine the length of project time and the amount of the needed sources, the time of project completion must correctly and exactly be calculated, so the critical path is calculated. The activities on this path have no floating. It means that there is no delay on these activities. As a result the calculation of the critical path in a project network has a special importance. In this paper a simple method for calculation the critical path is proposed. Assignment an exact time on any activity in real world is not correct; So the fuzzy and uncertainty theories are used to assigned a length of time on any activities. In the present study the trapezoidal fuzzy numbers are assigned to the length of activity time, and the total time of the project is also a fuzzy number. In addition, to compare the fuzzy numbers, ranking of fuzzy numbers are used. Finally a practical example will show the efficiency of the method.
International Nuclear Information System (INIS)
An accurate determination of damage fluence accumulated by reactor pressure vessels (RPV) as a function of time is essential in order to evaluate the vessel integrity for both pressurized thermal shock (PTS) transients and end-of-life considerations. The desired accuracy for neutron exposure parameters such as displacements per atom or fluence (E > 1 MeV) is of the order of 20 to 30%. However, these types of accuracies can only be obtained realistically by validation of nuclear data and calculational methods in benchmark facilities. The purposes of this paper are to review the needs and requirements for benchmark experiments, to discuss the status of current benchmark experiments, to summarize results and conclusions obtained so far, and to suggest areas where further benchmarking is needed
Results of the isotopic concentrations of WWER calculation Burnup Credit Benchmark NO.2 (CB2)
International Nuclear Information System (INIS)
The purpose of this document is to present the results of the nuclide concentrations of the WWER Burnup Credit Benchmark NO.2 (CB2) that were performed in The Nuclear Technology Center of Cuba with available codes and libraries. The CB2 benchmark specification as the second phase of the WWER burnup credit benchmark is summarized in [1]. The CB2 benchmark focused on WWER burnup credit study proposed on the 97' Atomic Energy Research symposium [2]. The obtained results are isotopic concentrations of spent fuel as a function of the burnup and cooling time. The depletion point 'ORIGEN2'[3] code was used for the calculation of the spent fuel concentration. This work also comprises the results obtained by other codes [4]. (Author)
International Nuclear Information System (INIS)
It was determined to perform the second NEACRP benchmark calculation on High Conversion Light Water Reactor (HCLWR) lattices at the 31st NEACRP meeting on October, 1988. The object was to clarify the physics problems induced in the data and method on HCLWR lattice analyses and also to obtain the reference solutions for deterministic codes by using continuous energy Monte Carlo codes. In the new problems, the analysis for the PROTEUS-LWHCR experiments were added. JAERI participated in this benchmark comparison by use of the VIM code (Monte Carlo method) and the SRAC code (collision probability method) with the libraries based on the JENDL-2 file. In this report, all of the calculated results are summarized. Some additional investigation will be also shown on resonance treatment and geometrical modelling relevant to the benchmark calculation. (author)
Computational benchmark for calculation of silane and siloxane thermochemistry.
Cypryk, Marek; Gostyński, Bartłomiej
2016-01-01
Geometries of model chlorosilanes, R3SiCl, silanols, R3SiOH, and disiloxanes, (R3Si)2O, R = H, Me, as well as the thermochemistry of the reactions involving these species were modeled using 11 common density functionals in combination with five basis sets to examine the accuracy and applicability of various theoretical methods in organosilicon chemistry. As the model reactions, the proton affinities of silanols and siloxanes, hydrolysis of chlorosilanes and condensation of silanols to siloxanes were considered. As the reference values, experimental bonding parameters and reaction enthalpies were used wherever available. Where there are no experimental data, W1 and CBS-QB3 values were used instead. For the gas phase conditions, excellent agreement between theoretical CBS-QB3 and W1 and experimental thermochemical values was observed. All DFT methods also give acceptable values and the precision of various functionals used was comparable. No significant advantage of newer more advanced functionals over 'classical' B3LYP and PBEPBE ones was noted. The accuracy of the results was improved significantly when triple-zeta basis sets were used for energy calculations, instead of double-zeta ones. The accuracy of calculations for the reactions in water solution within the SCRF model was inferior compared to the gas phase. However, by careful estimation of corrections to the ΔHsolv and ΔGsolv of H(+) and HCl, reasonable values of thermodynamic quantities for the discussed reactions can be obtained. PMID:26781663
A primer for criticality calculations with DANTSYS
Energy Technology Data Exchange (ETDEWEB)
Busch, R.D. [Univ. of New Mexico, Albuquerque, NM (United States). Nuclear Criticality Safety Group
1997-08-01
With the closure of many experimental facilities, the nuclear safety analyst has to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. Although deterministic methods often do not provide exact models of a system, a substantial amount of reliable information on nuclear systems can be obtained using these methods if the user understands their limitations. To guide criticality specialists in this area, the Nuclear Criticality Safety Group at the University of New Mexico (UNM) in cooperation with the Radiation Transport Group at Los Alamos National Laboratory (LANL) has designed a primer to help the analyst understand and use the DANTSYS deterministic transport code for nuclear criticality safety analyses. DANTSYS is the new name of the group of codes formerly known as: ONEDANT, TWODANT, TWOHEX, TWOGQ, and THREEDANT. The primer is designed to teach bu example, with each example illustrating two or three DANTSYS features useful in criticality analyses. Starting with a Quickstart chapter, the primer gives an overview of the basic requirements for DANTSYS input and allows the user to quickly run a simple criticality problem with DANTSYS. Each chapter has a list of basic objectives at the beginning identifying the goal of the chapter and the individual DANTSYS features covered in detail in the chapter example problems. On completion of the primer, it is expected that the user will be comfortable doing criticality calculations with DANTSYS and can handle 60--80% of the situations that normally arise in a facility. The primary provides a set of input files that can be selective modified by the user to fit each particular problem.
A primer for criticality calculations with DANTSYS
International Nuclear Information System (INIS)
With the closure of many experimental facilities, the nuclear safety analyst has to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. Although deterministic methods often do not provide exact models of a system, a substantial amount of reliable information on nuclear systems can be obtained using these methods if the user understands their limitations. To guide criticality specialists in this area, the Nuclear Criticality Safety Group at the University of New Mexico (UNM) in cooperation with the Radiation Transport Group at Los Alamos National Laboratory (LANL) has designed a primer to help the analyst understand and use the DANTSYS deterministic transport code for nuclear criticality safety analyses. DANTSYS is the new name of the group of codes formerly known as: ONEDANT, TWODANT, TWOHEX, TWOGQ, and THREEDANT. The primer is designed to teach bu example, with each example illustrating two or three DANTSYS features useful in criticality analyses. Starting with a Quickstart chapter, the primer gives an overview of the basic requirements for DANTSYS input and allows the user to quickly run a simple criticality problem with DANTSYS. Each chapter has a list of basic objectives at the beginning identifying the goal of the chapter and the individual DANTSYS features covered in detail in the chapter example problems. On completion of the primer, it is expected that the user will be comfortable doing criticality calculations with DANTSYS and can handle 60--80% of the situations that normally arise in a facility. The primary provides a set of input files that can be selective modified by the user to fit each particular problem
Isopiestic density law of actinide nitrates applied to criticality calculations
International Nuclear Information System (INIS)
Up to now, criticality safety experts used density laws fitted on experimental data and applied them in and outside the measurement range. Depending on the case, such an approach could be wrong for nitrate solutions. Seven components are concerned: UO2(NO3)2, U(NO3)4, Pu(NO3)4, Pu(NO3)3, Th(NO3)4, Am(NO3)3 and HNO3. To get rid of this problem, a new methodology based on the thermodynamic concept of binary electrolytes solutions mixtures at constant water activity, so called 'isopiestic' solutions, has been developed by IRSN to calculate the nitrate solutions density. This article shortly presents the theoretical aspects of the method, its qualification using benchmarks and its implementation in IRSN graphical user interface. (author)
Monte Carlo criticality calculation for Pebble-type HTR-PROTEUS core
International Nuclear Information System (INIS)
These days, pebble-bed and other High-Temperature Gas-cooled Reactor (HTGR) designs are once again in vogue in connection with hydrogen production. In this study, as a part of establishing Monte Carlo computation system for HTGR core analysis, some criticality calculations for pebble-type HTGR were carried out using MCNP code. Firstly, the pebble-bed cores of HTR-PROTEUS critical facility in Swiss were selected for the benchmark model, and, after the detailed MCNP modeling of the whole facility, criticality calculations were performed. It was also investigated the homogenization effect of TRISO fuel on criticality
Benchmarking of the ZR-6 critical assemblies using WIMS
International Nuclear Information System (INIS)
During the 1970 and early 1980 a wide ranging series of experiments was performed in the ZR-6 facility in Budapest. The cores consisted of arrays of UO2 fuel rods on a hexagonal pitch with light water moderator. Criticality was achieved by varying the moderator height.(Authors)
Benchmark calculations for evaluation methods of gas volumetric leakage rate
International Nuclear Information System (INIS)
A containment function of radioactive materials transport casks is essential for safe transportation to prevent the radioactive materials from being released into environment. Regulations such as IAEA standard determined the limit of radioactivity to be released. Since is not practical for the leakage tests to measure directly the radioactivity release from a package, as gas volumetric leakages rates are proposed in ANSI N14.5 and ISO standards. In our previous works, gas volumetric leakage rates for several kinds of gas from various leaks were measured and two evaluation methods, 'a simple evaluation method' and 'a strict evaluation method', were proposed based on the results. The simple evaluation method considers the friction loss of laminar flow with expansion effect. The strict evaluating method considers an exit loss in addition to the friction loss. In this study, four worked examples were completed for on assumed large spent fuel transport cask (Type B Package) with wet or dry capacity and at three transport conditions; normal transport with intact fuels or failed fuels, and an accident in transport. The standard leakage rates and criteria for two kinds of leak test were calculated for each example by each evaluation method. The following observations are made based upon the calculations and evaluations: the choked flow model of ANSI method greatly overestimates the criteria for tests ; the laminar flow models of both ANSI and ISO methods slightly overestimate the criteria for tests; the above two results are within the design margin for ordinary transport condition and all methods are useful for the evaluation; for severe condition such as failed fuel transportation, it should pay attention to apply a choked flow model of ANSI method. (authors)
D.C. Blitz (David)
2011-01-01
textabstractBenchmarking benchmarks is a bundle of six studies that are inspired by the prevalence of benchmarking in academic finance research as well as in investment practice. Three studies examine if current benchmark asset pricing models adequately describe the cross-section of stock returns. W
International Nuclear Information System (INIS)
This paper provides benchmark comparisons of the MCNPX Monte Carlo code to a series of integral critical experiments performed at the Toshiba Nuclear Critical Assembly (NCA) facility from 1994 to 2001 [1;2]. The beta-1 release version of ENDF/B-VII is used for all nuclides process with NJOY99 (update 96) executed with the beta test version of MCNPX 2.6.A [3]. A total of fifty-two (52) low enriched, UO2 pin-lattice in water experiments were analyzed with experimental W/F ratios from 0.791 to 1.756. The lattices were designed to simulate that of 8 x 8 and 9 x 9 Boiling Water Reactor (BWR) lattices with hollow aluminum tubes inserted between the fuel rods to simulate voiding conditions in approximately half of the experiments. In addition to measured critical lattice configurations, a series of individual pin-power fission density estimates were made via gross gamma scans of individual fuel pins after irradiation. This data is also used to benchmark the Monte Carlo fission density calculations to confirm the code and cross-section applicability for use as a benchmarking tool for the LANCER02 lattice physics code [4]. (authors)
CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS
International Nuclear Information System (INIS)
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
International Nuclear Information System (INIS)
The ENDF/B-VII.1 library is the latest revision to the United States' Evaluated Nuclear Data File (ENDF). The ENDF library is currently in its seventh generation, with ENDF/B-VII.0 being released in 2006. This revision expands upon that library, including the addition of new evaluated files (was 393 neutron files previously, now 418 including replacement of elemental vanadium and zinc evaluations with isotopic evaluations) and extension or updating of many existing neutron data files. Complete details are provided in the companion paper [1]. This paper focuses on how accurately application libraries may be expected to perform in criticality calculations with these data. Continuous energy cross section libraries, suitable for use with the MCNP Monte Carlo transport code, have been generated and applied to a suite of nearly one thousand critical benchmark assemblies defined in the International Criticality Safety Benchmark Evaluation Project's International Handbook of Evaluated Criticality Safety Benchmark Experiments. This suite covers uranium and plutonium fuel systems in a variety of forms such as metallic, oxide or solution, and under a variety of spectral conditions, including unmoderated (i.e., bare), metal reflected and water or other light element reflected. Assembly eigenvalues that were accurately predicted with ENDF/B-VII.0 cross sections such as unmoderated and uranium reflected 235U and 239Pu assemblies, HEU solution systems and LEU oxide lattice systems that mimic commercial PWR configurations continue to be accurately calculated with ENDF/B-VII.1 cross sections, and deficiencies in predicted eigenvalues for assemblies containing selected materials, including titanium, manganese, cadmium and tungsten are greatly reduced. Improvements are also confirmed for selected actinide reaction rates such as 236U capture. Other deficiencies, such as the overprediction of Pu solution system critical eigenvalues and a decreasing trend in calculated eigenvalue for
BENCHMARKING UPGRADED HOTSPOT DOSE CALCULATIONS AGAINST MACCS2 RESULTS
Energy Technology Data Exchange (ETDEWEB)
Brotherton, Kevin
2009-04-30
The radiological consequence of interest for a documented safety analysis (DSA) is the centerline Total Effective Dose Equivalent (TEDE) incurred by the Maximally Exposed Offsite Individual (MOI) evaluated at the 95th percentile consequence level. An upgraded version of HotSpot (Version 2.07) has been developed with the capabilities to read site meteorological data and perform the necessary statistical calculations to determine the 95th percentile consequence result. These capabilities should allow HotSpot to join MACCS2 (Version 1.13.1) and GENII (Version 1.485) as radiological consequence toolbox codes in the Department of Energy (DOE) Safety Software Central Registry. Using the same meteorological data file, scenarios involving a one curie release of {sup 239}Pu were modeled in both HotSpot and MACCS2. Several sets of release conditions were modeled, and the results compared. In each case, input parameter specifications for each code were chosen to match one another as much as the codes would allow. The results from the two codes are in excellent agreement. Slight differences observed in results are explained by algorithm differences.
Criticality calculations with MCNP trademark: A primer
International Nuclear Information System (INIS)
With the closure of many experimental facilities, the nuclear criticality safety analyst increasingly is required to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. However, in many cases, the analyst has little experience with the specific codes available at his/her facility. This primer will help you, the analyst, understand and use the MCNP Monte Carlo code for nuclear criticality safety analyses. It assumes that you have a college education in a technical field. There is no assumption of familiarity with Monte Carlo codes in general or with MCNP in particular. Appendix A gives an introduction to Monte Carlo techniques. The primer is designed to teach by example, with each example illustrating two or three features of MCNP that are useful in criticality analyses. Beginning with a Quickstart chapter, the primer gives an overview of the basic requirements for MCNP input and allows you to run a simple criticality problem with MCNP. This chapter is not designed to explain either the input or the MCNP options in detail; but rather it introduces basic concepts that are further explained in following chapters. Each chapter begins with a list of basic objectives that identify the goal of the chapter, and a list of the individual MCNP features that are covered in detail in the unique chapter example problems. It is expected that on completion of the primer you will be comfortable using MCNP in criticality calculations and will be capable of handling 80 to 90 percent of the situations that normally arise in a facility. The primer provides a set of basic input files that you can selectively modify to fit the particular problem at hand
SCALE 5.1 - criticality and inventory calculation for WWER-440 fuel
International Nuclear Information System (INIS)
The latest version of SCALE system (SCALE 5.1) was tested for criticality and inventory calculation for WWER-440 fuel. The criticality calculations (the KENO VI module) were tested on experimental critical cores (393 experiments from ICSBEP) and numerical benchmarks CB1, CB3 and CB4. The cross sections are prepared either by the NITAWL module (original way, used in SCALE 4.x) or by the CENTRM module (in SCALE 5.1 default). In the article are compared results by using both ways. The 44-group and 238-group libraries were used. The inventory calculations (the ORIGEN-S and TRITON modules) were tested on experiments in Russia (measurement in 80-ies, ISTC 2670) and numerical benchmark CB2. The ORIGEN-S module uses the WWER(3.6) library, the TRITON module uses the 44-group library (Authors)
International Nuclear Information System (INIS)
Defining precisely the burnup of the nuclear fuel is important from the point of view of core design calculations, safety analyses, criticality calculations (e.g. burnup credit calculations), etc. This paper deals with the uncertainties of MULTICELL calculations obtained by the solution of the OECD NEA UAM PWR pin cell burnup benchmark. In this assessment Monte-Carlo type statistical analyses are applied and the energy dependent covariance matrices of the cross-sections are taken into account. Additionally, the impact of the uncertainties of the fission yields is also considered. The target quantities are the burnup dependent uncertainties of the infinite multiplication factor, the two-group cross-sections, the reaction rates and the number densities of some isotopes up to the burnup of 60 MWd/kgU. In the paper the burnup dependent tendencies of the corresponding uncertainties and their sources are analyzed.
Energy Technology Data Exchange (ETDEWEB)
Kereszturi, Andras [Hungarian Academy of Sciences, Budapest (Hungary). Centre for Energy Research; Panka, Istvan
2015-09-15
Defining precisely the burnup of the nuclear fuel is important from the point of view of core design calculations, safety analyses, criticality calculations (e.g. burnup credit calculations), etc. This paper deals with the uncertainties of MULTICELL calculations obtained by the solution of the OECD NEA UAM PWR pin cell burnup benchmark. In this assessment Monte-Carlo type statistical analyses are applied and the energy dependent covariance matrices of the cross-sections are taken into account. Additionally, the impact of the uncertainties of the fission yields is also considered. The target quantities are the burnup dependent uncertainties of the infinite multiplication factor, the two-group cross-sections, the reaction rates and the number densities of some isotopes up to the burnup of 60 MWd/kgU. In the paper the burnup dependent tendencies of the corresponding uncertainties and their sources are analyzed.
International Nuclear Information System (INIS)
An accurate knowledge of irradiated fuel composition is of utmost importance regarding properties such as criticality, activity or residual heat generation. These magnitudes are in turn essential to fuel transport and storage and depend on many parameters, from which of course fuel type is essential. In the frame of activities devoted to fuel cycle issues, the NEA WPRS proposed a Depletion Calculation Benchmark to compare results and trends with different codes and libraries. While Phase 1 dealt with UOX fuel, Phase 2 is devoted to MOX fuel. The present paper aims at comparing isotopic compositions for MOX fuel obtained by GRS and AREVA with different codes and libraries. (orig.)
A full CI treatment of Ne atom - A benchmark calculation performed on the NAS CRAY 2
Bauschlicher, C. W., Jr.; Langhoff, S. R.; Partridge, H.; Taylor, P. R.
1986-01-01
Full CI calculations are performed for Ne atom using Gaussian basis sets of up to triple-zeta plus double polarization quality. The total valence correlation energy through double, triple, quadruple and octuple excitations is compared for eight different basis sets. These results are expected to be an important benchmark for calibrating methods for estimating the importance of higher excitations.
DEFF Research Database (Denmark)
Grandjean, Philippe; Budtz-Joergensen, Esben
2013-01-01
follow-up of a Faroese birth cohort were used. Serum-PFC concentrations were measured at age 5 years, and serum antibody concentrations against tetanus and diphtheria toxoids were obtained at ages 7 years. Benchmark dose results were calculated in terms of serum concentrations for 431 children with...
The solution of the LEU and MOX WWER-1000 calculation benchmark with the CARATE - multicell code
International Nuclear Information System (INIS)
Preparations for disposition of weapons grade plutonium in WWER-1000 reactors are in progress. Benchmark: Defined by the Kurchatov Institute (S. Bychkov, M. Kalugin, A. Lazarenko) to assess the applicability of computer codes for weapons grade MOX assembly calculations. Framework: 'Task force on reactor-based plutonium disposition' of OECD Nuclear Energy Agency. (Authors)
Benchmark Testing of a New ^{56}Fe Evaluation for Criticality Safety Applications
Energy Technology Data Exchange (ETDEWEB)
Leal, Luiz C [ORNL; Ivanov, E. [Institut de Radioprotection et de Surete Nucleaire
2015-01-01
The SAMMY code was used to evaluate resonance parameters of the ^{56}Fe cross section in the resolved resonance energy range of 0–2 MeV using transmission data, capture, elastic, inelastic, and double differential elastic cross sections. The resonance analysis was performed with the code SAMMY that fits R-matrix resonance parameters using the generalized least-squares technique (Bayes’ theory). The evaluation yielded a set of resonance parameters that reproduced the experimental data very well, along with a resonance parameter covariance matrix for data uncertainty calculations. Benchmark tests were conducted to assess the evaluation performance in benchmark calculations.
FUEL HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS
Energy Technology Data Exchange (ETDEWEB)
C.E. Sanders
2005-06-30
The purpose of this design calculation is to perform a criticality evaluation of the Fuel Handling Facility (FHF) and the operations and processes performed therein. The current intent of the FHF is to receive transportation casks whose contents will be unloaded and transferred to waste packages (WP) or MGR Specific Casks (MSC) in the fuel transfer bays. Further, the WPs will also be prepared in the FHF for transfer to the sub-surface facility (for disposal). The MSCs will be transferred to the Aging Facility for storage. The criticality evaluation of the FHF features the following: (I) Consider the types of waste to be received in the FHF as specified below: (1) Uncanistered commercial spent nuclear fuel (CSNF); (2) Canistered CSNF (with the exception of horizontal dual-purpose canister (DPC) and/or multi-purpose canisters (MPCs)); (3) Navy canistered SNF (long and short); (4) Department of Energy (DOE) canistered high-level waste (HLW); and (5) DOE canistered SNF (with the exception of MCOs). (II) Evaluate the criticality analyses previously performed for the existing Nuclear Regulatory Commission (NRC)-certified transportation casks (under 10 CFR 71) to be received in the FHF to ensure that these analyses address all FHF conditions including normal operations, and Category 1 and 2 event sequences. (III) Evaluate FHF criticality conditions resulting from various Category 1 and 2 event sequences. Note that there are currently no Category 1 and 2 event sequences identified for FHF. Consequently, potential hazards from a criticality point of view will be considered as identified in the ''Internal Hazards Analysis for License Application'' document (BSC 2004c, Section 6.6.4). (IV) Assess effects of potential moderator intrusion into the fuel transfer bay for defense in depth. The SNF/HLW waste transfer activity (i.e., assembly and canister transfer) that is being carried out in the FHF has been classified as safety category in the &apos
The fifth Atomic Energy Research dynamic benchmark calculation with HEXTRAN-SMABRE
International Nuclear Information System (INIS)
The fifth Atomic Energy Research dynamic benchmark is the first Atomic Energy Research benchmark for coupling of the thermohydraulic codes and three-dimensional reactor dynamic core models. In VTT HEXTRAN 2.7 is used for the core dynamics and SMABRE 4.6 as a thermohydraulic model for the primary and secondary loops. The plant model for SMABRE is based mainly on two input models. the Loviisa model and standard WWER-440/213 plant model. The primary circuit includes six separate loops, totally 505 nodes and 652 junctions. The reactor pressure vessel is divided into six parallel channels. In HEXTRAN calculation 176 symmetry is used in the core. In the sequence of main steam header break at the hot standby state, the liquid temperature is decreased symmetrically in the core inlet which leads to return to power. In the benchmark, no isolations of the steam generators are assumed and the maximum core power is about 38 % of the nominal power at four minutes after the break opening in the HEXTRAN-SMABRE calculation. Due to boric acid in the high pressure safety injection water, the power finally starts to decrease. The break flow is pure steam in the HEXTRAN-SMABRE calculation during the whole transient even in the swell levels in the steam generators are very high due to flashing. Because of sudden peaks in the preliminary results of the steam generator heat transfer, the SMABRE drift-flux model was modified. The new model is a simplified version of the EPRI correlation based on test data. The modified correlation behaves smoothly. In the calculations nuclear data is based on the ENDF/B-IV library and it has been evaluated with the CASMO-HEX code. The importance of the nuclear data was illustrated by repeating the benchmark calculation with using three different data sets. Optimal extensive data valid from hot to cold conditions were not available for all types of fuel enrichments needed in this benchmark.(Author)
Energy Technology Data Exchange (ETDEWEB)
Lara, Rafael G.; Maiorino, Jose R., E-mail: rafael.lara@aluno.ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br [Universidade Federal do ABC (UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais Aplicadas
2013-07-01
This work aimed at the implementation and qualification of MCNP code in a supercomputer of the Universidade Federal do ABC, so that may be available a next-generation simulation tool for precise calculations of nuclear reactors and systems subject to radiation. The implementation of this tool will have multidisciplinary applications, covering various areas of engineering (nuclear, aerospace, biomedical), radiation physics and others.
Primer for criticality calculations with DANTSYS
International Nuclear Information System (INIS)
With the closure of many experimental facilities, the nuclear criticality safety analyst is increasingly required to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. However, in many cases, the analyst has little experience with the specific codes available at his or her facility. Typically, two types of codes are available: deterministic codes such as ANISN or DANTSYS that solve an approximate model exactly and Monte Carlo Codes such as KENO or MCNP that solve an exact model approximately. Often, the analyst feels that the deterministic codes are too simple and will not provide the necessary information, so most modeling uses Monte Carlo methods. This sometimes means that hours of effort are expended to produce results available in minutes from deterministic codes. A substantial amount of reliable information on nuclear systems can be obtained using deterministic methods if the user understands their limitations. To guide criticality specialists in this area, the Nuclear Criticality Safety Group at the University of New Mexico in cooperation with the Radiation Transport Group at Los Alamos National Laboratory has designed a primer to help the analyst understand and use the DANTSYS deterministic transport code for nuclear criticality safety analyses. (DANTSYS is the name of a suite of codes that users more commonly know as ONEDANT, TWODANT, TWOHEX, and THREEDANT.) It assumes a college education in a technical field, but there is no assumption of familiarity with neutronics codes in general or with DANTSYS in particular. The primer is designed to teach by example, with each example illustrating two or three DANTSYS features useful in criticality analyses
Real Variance Estimation of BEAVRS whole core benchmark in Monte Carlo Eigenvalue Calculations
International Nuclear Information System (INIS)
For whole core analysis by the MC eigenvalue mode calculations, some severe problems are encountered because these systems have higher dominance ratios (DRs) than a fuel assembly (FA) or critical facilities. It is well known that the apparent variance of a local tally like pin power is differ from the real variance considerably. In McCARD code, four approaches for the real variance estimation were implemented. This benchmark provides a detailed description of fuel assemblies, burnable absorbers, in-core fission detectors, core loading patterns, and numerous in-vessel components with three-dimensional (3D) scale. In this study, we perform a real variance estimation of MC tally for the design parameter such as keff, pin fission power, FA-wise fission power for BEAVRS fresh core using McCARD. In addition, this paper presents a new method to estimate the real variance called history-based sampling method, briefly. In this study, the real variance estimations for the BEAVRS whole core benchmark were performed using Gelbard's batch method, Ueki's inter-cycle correction method, and Shim's HB method, which were implemented in McCARD. As expected, it was observed that the apparent variance of local MC tally estimate such as pin or FA-wise fission power tends to be smaller than its real variance while that of the global MC tally such as keff is comparable to the reference. To investigate the difference of the real to apparent variance ratio between global and local MC tally, the correlation coefficients between each pin or FA fission power are calculated using McCARD. Because the correlation coefficients between neighbor pins is near 1.0, the error by FSD inter-cycle correlation would be propagated. In addition, this paper presented a new variance estimation method called the HS method. The HS method has several advantages over the HB method. The HS method is very easy to implement into a existing MC code and it does not require additional parameters such as
Two-dimension calculation of proposed benchmark core analysis for the BN-600 hybrid reactor
International Nuclear Information System (INIS)
This paper presents primary calculation results of the proposed benchmark for a hybrid UOX/MOX fuelled core of the BN-600 reactor. The analysis in this paper uses a R-Z homogeneous model of the BN-600 reactor. Calculation results include effective multiplication factors obtained by both diffusion and Monte Carlo methods; fuel Doppler constants; steel Doppler constants; sodium density coefficient; steel density coefficients; fuel density coefficient; absorber density coefficient; axial and radial expansion coefficients; dynamic parameters; power distribution
Three dimension calculation of proposed benchmark core analysis for the BN-600 hybrid reactor
International Nuclear Information System (INIS)
This paper presents primary calculation results of the proposed benchmark for a hybrid UOX/MOX fuelled core of the BN-600 reactor. The analysis in this paper uses a HEX-Z homogeneous model of the BN-600 reactor. Calculation results include effective multiplication factors obtained by both diffusion and Monte Carlo methods; fuel Doppler constants; steel Doppler constants; sodium density coefficient; steel density coefficients; fuel density coefficient; absorber density coefficient; axial and radial expansion coefficients; dynamic parameters; power distribution
The potential use of criticality benchmark experiments in nuclear data evaluation
International Nuclear Information System (INIS)
The presence of significant systematic errors even in the latest nuclear data compilations can be shown by making Monte Carlo calculations for critical systems. Calculations have been made for forty-four critical systems. Modelling errors, which used to plague such calculations, have been eliminated, and discrepancies between calculated and experimental eigenvalues of critical systems can now be confidently ascribed to errors in the nuclear data. The Monte Carlo code MONK is particularly suitable for these calculations. (author)
Peneliau, Y.; Litaize, O.; Archier, P.; De Saint Jean, C.
2014-04-01
A large set of nuclear data are investigated to improve the calculation predictions of the new neutron transport simulation codes. With the next generation of nuclear power plants (GEN IV projects), one expects to reduce the calculated uncertainties which are mainly coming from nuclear data and are still very important, before taking into account integral information in the adjustment process. In France, future nuclear power plant concepts will probably use MOX fuel, either in Sodium Fast Reactors or in Gas Cooled Fast Reactors. Consequently, the knowledge of 239Pu cross sections and other nuclear data is crucial issue in order to reduce these sources of uncertainty. The Prompt Fission Neutron Spectra (PFNS) for 239Pu are part of these relevant data (an IAEA working group is even dedicated to PFNS) and the work presented here deals with this particular topic. The main international data files (i.e. JEFF-3.1.1, ENDF/B-VII.0, JENDL-4.0, BRC-2009) have been considered and compared with two different spectra, coming from the works of Maslov and Kornilov respectively. The spectra are first compared by calculating their mathematical moments in order to characterize them. Then, a reference calculation using the whole JEFF-3.1.1 evaluation file is performed and compared with another calculation performed with a new evaluation file, in which the data block containing the fission spectra (MF=5, MT=18) is replaced by the investigated spectra (one for each evaluation). A set of benchmarks is used to analyze the effects of PFNS, covering criticality cases and mock-up cases in various neutron flux spectra (thermal, intermediate, and fast flux spectra). Data coming from many ICSBEP experiments are used (PU-SOL-THERM, PU-MET-FAST, PU-MET-INTER and PU-MET-MIXED) and French mock-up experiments are also investigated (EOLE for thermal neutron flux spectrum and MASURCA for fast neutron flux spectrum). This study shows that many experiments and neutron parameters are very sensitive to
Institute of Scientific and Technical Information of China (English)
XIAO Hai; LI Jun
2008-01-01
Benchmark calculations on the molar atomization enthalpy, geometry, and vibrational frequencies of uranium hexafluoride (UF6) have been performed by using relativistic density functional theory (DFT) with various levels of relativistic effects, different types of basis sets, and exchange-correlation functionals. Scalar relativistic effects are shown to be critical for the structural properties. The spin-orbit coupling effects are important for the calculated energies, but are much less important for other calculated ground-state properties of closed-shell UF6. We conclude through systematic investigations that ZORA- and RECP-based relativistic DFT methods are both appropriate for incorporating relativistic effects. Comparisons of different types of basis sets (Slater, Gaussian, and plane-wave types) and various levels of theoretical approximation of the exchange-correlation functionals were also made.
Lutnæs, O.B.; Teale, A.M.; Helgaker, T.; Tozer, D J; Ruud, K.; Gauss, J.
2009-01-01
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 th...
International Nuclear Information System (INIS)
The validation of a code for criticality safety analysis requires the recalculation of benchmark experiments. The selected benchmark experiments are chosen such that they have properties similar to the application case that has to be assessed. A common source of benchmark experiments is the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments' (ICSBEP Handbook) compiled by the 'International Criticality Safety Benchmark Evaluation Project' (ICSBEP). In order to take full advantage of the information provided by the individual benchmark descriptions for the application case, the recommended procedure is to perform an uncertainty analysis. The latter is based on the uncertainties of experimental results included in most of the benchmark descriptions. They can be performed by means of the Monte Carlo sampling technique. The consideration of uncertainties is also being introduced in the supplementary sheet of DIN 25478 'Application of computer codes in the assessment of criticality safety'. However, for a correct treatment of uncertainties taking into account the individual uncertainties of the benchmark experiments is insufficient. In addition, correlations between benchmark experiments have to be handled correctly. For example, these correlations can arise due to different cases of a benchmark experiment sharing the same components like fuel pins or fissile solutions. Thus, manufacturing tolerances of these components (e.g. diameter of the fuel pellets) have to be considered in a consistent manner in all cases of the benchmark experiment. At the 2012 meeting of the Expert Group on 'Uncertainty Analysis for Criticality Safety Assessment' (UACSA) of the OECD/NEA a benchmark proposal was outlined that aimed for the determination of the impact on benchmark correlations on the estimation of the computational bias of the neutron multiplication factor (keff). The analysis presented here is based on this proposal. (orig.)
Some comments on cold hydrogenous moderators, simple synthetic kernels and benchmark calculations
International Nuclear Information System (INIS)
The author comments on three general subjects which are not directly related, but which in his opinion are very relevant to the objectives of the workshop. The first of these is parahydrogen moderators, about which recurring questions have been raised during the Workshop. The second topic is related to the use of simple synthetic scattering kernels in conjunction with the neutron transport equation to carry out elementary mathematical analyses and simple computational analyses in order to understand the gross physics of time-dependent neutron transport initiated by pulsed sources in cold moderators. The third subject is that of 'simple' benchmark calculations by which is meant calculations that are simple compared to the very large scale combined spallation, slowing-down, thermalization calculations using MCNP and other large Monte Carlo codes. Such benchmark problems can be created so that they are closely related to both the geometric configuration and material composition of cold moderators of interest and still can be solved using steady-state deterministic transport codes to calculate the asymptotic time-decay constant, and the time-asymptotic energy spectrum of neutrons in the cold moderator and the spectrum of the cold neutrons leaking from it (neither of which should be expected to be Maxwellian in these small leakage-dominated systems). These would provide rather precise benchmark solutions against which the results of the large scale calculations carried out for the whole spallation, slowing-down, thermalization system -- for the same decoupled cold moderator -- could be compared.
Neutron transport calculations of some fast critical assemblies
International Nuclear Information System (INIS)
To analyse the influence of the input variables of the transport codes upon the neutronic results (eigenvalues, generation times, . . . ) four Benchmark calculations have been performed. Sensitivity analysis have been applied to express these dependences in a useful way, and also to get an unavoidable experience to carry out calculations achieving the required accuracy and doing them in practical computing times. (Author) 29 refs
International Nuclear Information System (INIS)
Monte Carlo simulations are regarded as the most accurate method of solving complex problems of radiation transport. Therefore, they have great potential to realize more exact dose calculations for treatment planning in radiation therapy. However, there is a lack of information on how correct the results of Monte Carlo calculations are on an absolute basis. A practical verification of the calculations can be performed by direct comparison with a benchmark experiment. Thereby, the uncertainties of the experimental result and of the simulation also have to be considered to make a meaningful comparison between the experiment and the simulation possible. This dissertation presents a benchmark experiment and its results, including the uncertainty, which can be used to test the accuracy of Monte Carlo calculations in the field of radiation therapy. The experiment was planned to have parallels to clinical radiation therapy, among other things, with respect to the radiation applied, the materials used and the manner of dose detection. The benchmark experiment aimed at an absolute comparison with a simulation result and because of this it was necessary to use a special research accelerator as a radiation source in the experiment. The accurate characterization of the accelerator beam was a precondition to define a realistic radiation source for the Monte Carlo simulation. Therefore, this work also deals with the characterization of the source and investigations regarding the X-ray target used. Additionally, the dissertation contains the verification of the widely used Monte Carlo program EGSnrc by the benchmark experiment. The simulation of the experiment by EGSnrc, the results and the estimation of the uncertainty related to the simulation are documented in this work.The results and findings of this dissertation end in a comparison between the results of the benchmark experiment and the corresponding calculations with EGSnrc. The benchmark experiment and the simulations
Doppler coefficient of reactivity - Benchmark calculations for different enrichments of UO2
International Nuclear Information System (INIS)
Continuous-energy Monte Carlo code, MCNP along with its cross section data library in ACE format based on ENDF/B-V and VI has been used to analyze a new computational benchmark circulated by LANL (LA-UR-06-2968) on Doppler coefficient for different types UO2. Doppler coefficient has been computed by calculating the Eigen values of some selected idealized PWR fuel pin cell configurations with seven different fuel enrichments of UO2. Even though the benchmark contained configurations for different kinds of mixed oxide fuel configurations, the same could not be analyzed for evaluating the Doppler coefficient due to lack of nuclear data with us for some of the isotopes. The pin cell configuration is modeled in 3-D geometry by assuming an infinite dimension instead of reflecting boundary conditions in the axial direction and reflective boundary conditions are assumed on all other four sides of the pin cell. With this geometry model of the pin cell, first an initial criticality run is made with 1.0 million active histories (i.e. 1000 active and 50 skipped cycles with 1000 histories per cycle). The fission source file (SRCTP) from the last cycle of this run is then used as converged input source for the final run with 14 million histories (3500 active cycles and 500 skipped with 4000 histories per cycle). The intermediate MCNP run confirmed the sampling of fission sites in the entire fuel cell region modeled. Thermal treatment (Sαβ) option is used to take care of binding effect of hydrogen in water. The corresponding light water Sαβ cross- section treatments for temperature 600 K (HZP) is used. Doppler coefficients for all the UO2 cases are estimated using two different cross-section sets: Case-I based on ENDF/B-VI and Case II based on ENDF/B-V. In Case-I, fuel temperature changes from 600 K (Hot Zero Power) to 900 K (Hot Full Power) but in Case-II due to lack of data in the MCNP data library available, fuel temperature change assumed is from 587 K (Hot Zero Power) to
Calculational benchmark comparisons for a low sodium void worth actinide burner core design
International Nuclear Information System (INIS)
Recently, a number of low void worth core designs with non-conventional core geometries have been proposed. Since these designs lack a good experimental and computational data base, benchmark calculations are useful for the identification of possible biases in performance characteristics predictions. In this paper, a simplified benchmark model of a metal fueled, low void worth actinide burner design is detailed: and two independent neutronic performance evaluations are compared. Calculated performance characteristics are evaluated for three spatially uniform compositions (fresh uranium/plutonium, batch-averaged uranium/transuranic, and batch-averaged uranium/transuranic with fission products) and a regional depleted distribution obtained from a benchmark depletion calculation. For each core composition, the flooded and voided multiplication factor, power peaking factor, sodium void worth (and its components), flooded Doppler coefficient and control rod worth predictions are compared. In addition, the burnup swing, average discharge burnup, peak linear power, and fresh fuel enrichment are calculated for the depletion case. In general, remarkably good agreement is observed between the evaluations. The mot significant difference in predicted performance characteristics is a 0.3-05% Δk/(kk') bias in the sodium void worth. Significant differences in the transmutation rate of higher actinides are also observed; however, these differences do not cause discrepancies in the performance predictions
EA-MC Neutronic Calculations on IAEA ADS Benchmark 3.2
International Nuclear Information System (INIS)
The neutronics and the transmutation properties of the IAEA ADS benchmark 3.2 setup, the 'Yalina' experiment or ISTC project B-70, have been studied through an extensive amount of 3-D Monte Carlo calculations at CERN. The simulations were performed with the state-of-the-art computer code package EA-MC, developed at CERN. The calculational approach is outlined and the results are presented in accordance with the guidelines given in the benchmark description. A variety of experimental conditions and parameters are examined; three different fuel rod configurations and three types of neutron sources are applied to the system. Reactivity change effects introduced by removal of fuel rods in both central and peripheral positions are also computed. Irradiation samples located in a total of 8 geometrical positions are examined. Calculations of capture reaction rates in 129I, 237Np and 243Am samples and of fission reaction rates in 235U, 237Np and 243Am samples are presented. Simulated neutron flux densities and energy spectra as well as spectral indices inside experimental channels are also given according to benchmark specifications. Two different nuclear data libraries, JAR-95 and JENDL-3.2, are applied for the calculations
Proposal of a benchmark for core burnup calculations for a VVER-1000 reactor core
International Nuclear Information System (INIS)
In the framework of a project supported by the German BMU the code DYN3D should be further validated and verified. During the work a lack of a benchmark on core burnup calculations for VVER-1000 reactors was noticed. Such a benchmark is useful for validating and verifying the whole package of codes and data libraries for reactor physics calculations including fuel assembly modelling, fuel assembly data preparation, few group data parametrisation and reactor core modelling. The benchmark proposed specifies the core loading patterns of burnup cycles for a VVER-1000 reactor core as well as a set of operational data such as load follow, boron concentration in the coolant, cycle length, measured reactivity coefficients and power density distributions. The reactor core characteristics chosen for comparison and the first results obtained during the work with the reactor physics code DYN3D are presented. This work presents the continuation of efforts of the projects mentioned to estimate the accuracy of calculated characteristics of VVER-1000 reactor cores. In addition, the codes used for reactor physics calculations of safety related reactor core characteristics should be validated and verified for the cases in which they are to be used. This is significant for safety related evaluations and assessments carried out in the framework of licensing and supervision procedures in the field of reactor physics. (authors)
Calculational benchmark comparisons for a low sodium void worth actinide burner core design
International Nuclear Information System (INIS)
Recently, a number of low void worth core designs with non-conventional core geometries have been proposed. Since these designs lack a good experimental and computational database, benchmark calculations are useful for the identification of possible biases in performance characteristics predictions. In this paper, a simplified benchmark model of a metal fueled, low void worth actinide burner design is detailed; and two independent neutronic performance evaluations are compared. Calculated performance characteristics are evaluated for three spatially uniform compositions (fresh uranium/plutonium, batch-averaged uranium/transuranic, and batch-averaged uranium/transuranic with fission products) and a regional depleted distribution obtained from a benchmark depletion calculation. For each core composition, the flooded and voided multiplication factor, power peaking factor, sodium void worth (and its components), flooded Doppler coefficient and control rod worth predictions are compared. In addition, the burnup swing, average discharge burnup, peak linear power, and fresh fuel enrichment are calculated for the depletion case. In general, remarkably good agreement is observed between the evaluations. The most significant difference is predicted performance characteristics is a 0.3--0.5% Δk/(kk) bias in the sodium void worth. Significant differences in the transmutation rate of higher actinides are also observed; however, these differences do not cause discrepancies in the performing predictions
Calculation of the IAEA ADS neutronics benchmark (stage-1) (2D discrete coordinate method)
International Nuclear Information System (INIS)
To study the neutronics for the ADS system, a set of computation software based on discrete ordinate method is selected and established. The set is tested through an IAEA benchmark. In the test process, the understanding and using of this software set are improved. The benchmark is analyzed. The calculations include the effective multiplication factor keff , the required strength of the spallation neutron source for 1.5 GW thermal power, the distribution of power density and the spectrum index, and the void effect at the beginning of life, BOL; the spatial and time-dependent density distribution of various nuclides (actinides and fission products) for burn-up process. The results are given in figures and tables and are consistent with calculations made abroad. The conclusion is that this software set can be applied to the optimization of design study for the ADS system
International Nuclear Information System (INIS)
The facility for incineration of long-lived minor actinides and some dangerous fission products should be an important feature of the future nuclear power (NP). For many reasons the liquid-fuel reactor driven by accelerator can be considered as the perspective reactor- burner for radioactive waste. The fuel of such reactor is the fluoride molten salt composition with minor actinides (Np, Cm, Am) and some fission products (99Tc, 129I, etc.). Preliminary analysis shows that the values of keff, calculated with different codes and nuclear data differ up to several percents for such fuel compositions. Reliable critical and subcritical benchmark experiments with molten salt fuel compositions with significant quantities of minor actinides are absent. One of the main tasks for the numerical study of this problem is the estimation of nuclear data for such fuel compositions and verification of the different numerical codes used for the calculation of keff, neutron spectra and reaction rates. It is especially important for the resonance region where experimental data are poor or absent. The calculation benchmark of the cascade subcritical molten salt reactor is developed. For the chosen nuclear fuel composition the comparison of the results obtained by three different Monte-Carlo codes (MCNP4A, MCU, and C95) using three different nuclear data libraries are presented. This report concerns the investigation of subcritical molten salt reactor unit main peculiarities carried out at the beginning of ISTC project 1486. (author)
INTRACOIN level 1 benchmark calculations with EIR codes CONZRA, RANCH and RANCHN
International Nuclear Information System (INIS)
The authors present the results from calculations of INTRACOIN level 1, case 1 and 2 (one-dimensional advection-dispersion) benchmarks. The codes used are CONZRA and RANCH, corresponding to a semi-analytical solution of the transport equation, and RANCHN based on a fully numerical solution in the framework of the pseudo-spectral method. The influence of various boundary conditions is investigated. Excellent agreement between results from the different solution approaches is obtained. (Auth.)
Fast neutron fluence calculation benchmark analysis based on 3D MC-SN bidirectional coupling method
International Nuclear Information System (INIS)
The Monte Carlo (MC)-discrete ordinates (SN) bidirectional coupling method is an efficient approach to solve shielding calculation of the large complex nuclear facility. The test calculation was taken by the application of the MC-SN bidirectional coupling method on the shielding calculation of the large PWR nuclear facility. Based on the characteristics of NUREG/CR-6115 PWR benchmark model issued by the NRC, 3D Monte Carlo code was employed to accurately simulate the structure from the core to the thermal shield and the dedicated model of the calculation parts locating in the pressure vessel, while the TORT was used for the calculation from the thermal shield to the second down-comer region. The transform between particle probability distribution of MC and angular flux density of SN was realized by the interface program to achieve the coupling calculation. The calculation results were compared with MCNP and DORT solutions of benchmark report and satisfactory agreements were obtained. The preliminary validity of feasibility by using the method to solve shielding problem of a large complex nuclear device was proved. (authors)
Energy Technology Data Exchange (ETDEWEB)
Primm III, RT
2002-05-29
This volume of the progress report provides documentation of reactor physics and criticality safety studies conducted in the US during fiscal year 1997 and sponsored by the Fissile Materials Disposition Program of the US Department of Energy. Descriptions of computational and experimental benchmarks for the verification and validation of computer programs for neutron physics analyses are included. All benchmarks include either plutonium, uranium, or mixed uranium and plutonium fuels. Calculated physics parameters are reported for all of the computational benchmarks and for those experimental benchmarks that the US and Russia mutually agreed in November 1996 were applicable to mixed-oxide fuel cycles for light-water reactors.
RB reactor as the U-D2O benchmark criticality system
International Nuclear Information System (INIS)
From a rich and valuable database fro 580 different reactor cores formed up to now in the RB nuclear reactor, a selected and well recorded set is carefully chosen and preliminarily proposed as a new uranium-heavy water benchmark criticality system for validation od reactor design computer codes and data libraries. The first results of validation of the MCNP code and adjoining neutron cross section libraries are resented in this paper. (author)
International Nuclear Information System (INIS)
This paper describes details of the IAEA/CRP benchmark calculation by JAEA on the control rod withdrawal test in the Phenix End-of-Life Experiments. The power distribution deviation by the control rod insertion/withdrawal, which is the major target of the benchmark, is well simulated by calculation. In addition to the CRP activities, neutron and photon heat transport effect is evaluated in the nuclear heating calculation of the benchmark analysis. It is confirmed that the neutron and photon heat transport effect contributes to the improvement of the absolute power calculation results in the breeder blanket region. (author)
Uncertainties in Monte Carlo-based absorbed dose calculations for an experimental benchmark
International Nuclear Information System (INIS)
There is a need to verify the accuracy of general purpose Monte Carlo codes like EGSnrc, which are commonly employed for investigations of dosimetric problems in radiation therapy. A number of experimental benchmarks have been published to compare calculated values of absorbed dose to experimentally determined values. However, there is a lack of absolute benchmarks, i.e. benchmarks without involved normalization which may cause some quantities to be cancelled. Therefore, at the Physikalisch-Technische Bundesanstalt a benchmark experiment was performed, which aimed at the absolute verification of radiation transport calculations for dosimetry in radiation therapy. A thimble-type ionization chamber in a solid phantom was irradiated by high-energy bremsstrahlung and the mean absorbed dose in the sensitive volume was measured per incident electron of the target. The characteristics of the accelerator and experimental setup were precisely determined and the results of a corresponding Monte Carlo simulation with EGSnrc are presented within this study. For a meaningful comparison, an analysis of the uncertainty of the Monte Carlo simulation is necessary. In this study uncertainties with regard to the simulation geometry, the radiation source, transport options of the Monte Carlo code and specific interaction cross sections are investigated, applying the general methodology of the Guide to the expression of uncertainty in measurement. Besides studying the general influence of changes in transport options of the EGSnrc code, uncertainties are analyzed by estimating the sensitivity coefficients of various input quantities in a first step. Secondly, standard uncertainties are assigned to each quantity which are known from the experiment, e.g. uncertainties for geometric dimensions. Data for more fundamental quantities such as photon cross sections and the I-value of electron stopping powers are taken from literature. The significant uncertainty contributions are identified as
Uncertainties in Monte Carlo-based absorbed dose calculations for an experimental benchmark
Renner, F.; Wulff, J.; Kapsch, R.-P.; Zink, K.
2015-10-01
There is a need to verify the accuracy of general purpose Monte Carlo codes like EGSnrc, which are commonly employed for investigations of dosimetric problems in radiation therapy. A number of experimental benchmarks have been published to compare calculated values of absorbed dose to experimentally determined values. However, there is a lack of absolute benchmarks, i.e. benchmarks without involved normalization which may cause some quantities to be cancelled. Therefore, at the Physikalisch-Technische Bundesanstalt a benchmark experiment was performed, which aimed at the absolute verification of radiation transport calculations for dosimetry in radiation therapy. A thimble-type ionization chamber in a solid phantom was irradiated by high-energy bremsstrahlung and the mean absorbed dose in the sensitive volume was measured per incident electron of the target. The characteristics of the accelerator and experimental setup were precisely determined and the results of a corresponding Monte Carlo simulation with EGSnrc are presented within this study. For a meaningful comparison, an analysis of the uncertainty of the Monte Carlo simulation is necessary. In this study uncertainties with regard to the simulation geometry, the radiation source, transport options of the Monte Carlo code and specific interaction cross sections are investigated, applying the general methodology of the Guide to the expression of uncertainty in measurement. Besides studying the general influence of changes in transport options of the EGSnrc code, uncertainties are analyzed by estimating the sensitivity coefficients of various input quantities in a first step. Secondly, standard uncertainties are assigned to each quantity which are known from the experiment, e.g. uncertainties for geometric dimensions. Data for more fundamental quantities such as photon cross sections and the I-value of electron stopping powers are taken from literature. The significant uncertainty contributions are identified as
International Nuclear Information System (INIS)
Four calculational benchmarks have been selected to compare various nuclear data libraries based on both ENDF/B-IV and V, and to compare results from various transport codes. Discrepancies up to 20% in tritium production from 7Li were found and have been attributed mainly to differences in current ENDF/B-IV and V evaluations, while approx.4% is attributed to differences in the group structure of the libraries used. Results from MCNP and VIP Monte Carlo codes are in good agreement, but MORSE calculations show good agreement only for high threshold reactions
Benchmarking of calculation schemes in Apollo2 and COBAYA3 for VVER lattices
Zheleva, Nonka; Ivanov, Plamen; Todorova, Galina; Kolev, Nikola; Herrero Carrascosa, José Javier
2013-01-01
This paper presents solutions of the NURISP VVER lattice benchmark using APOLLO2, TRIPOLI4 and COBAYA3 pin-by-pin. The main objective is to validate MOC based calculation schemes for pin-by-pin cross-section generation with APOLLO2 against TRIPOLI4 reference results. A specific objective is to test the APOLLO2 generated cross-sections and interface discontinuity factors in COBAYA3 pin-by-pin calculations with unstructured mesh. The VVER-1000 core consists of large hexagonal assemblies with 2m...
International Nuclear Information System (INIS)
A benchmark problem was proposed to reproduce an experiment for target membrane structure cooling of Accelerator Driven System at the 10th meeting of IWGAR (International Working Group of Advanced Nuclear Reactors Thermal Hydraulic) by the Fluid Phenomena in Energy Exchanges Section of IAHR (International Association of Hydraulic Engineering and Research). The benchmark calculation has been carried out with AQUA and FLUENT codes to estimate the code validity for liquid metal thermal-hydraulics application. As a result of comparison between numerical analyses and experiment, it is concluded as follows: Inlet flow rate at the distributing grid much affects a coolant temperature and temperature pulsation near the membrane. The coolant temperature decreases and the pulsation decays rapidly as the flow rate toward the membrane center increases. On downstream of the distributing grid, numerical results agree with experimental data except that numerical analysis tends to overestimate the coolant temperature pulsation. Numerical results show that the decrease of coolant temperature and the dissipation of pulsation tend to be underestimated when the flow rate toward the membrane center increases. In FLUENT code, the dissipation of coolant temperature is underestimated more than in AQUA code because FLUENT code tends to overestimate the flow rate toward the membrane center. But the same tendency of the dissipation behavior is shown in AQUA code. A turbulent model is less influenced on the coolant behavior in this benchmark analysis. Because Prandtl (Pr) number of liquid metal is low and the turbulent flow is not developed sufficiently in the conditions of the experiment. (author)
Benchmark results for the critical slab and sphere problem in one-speed neutron transport theory
International Nuclear Information System (INIS)
Research highlights: → The critical slab and sphere problem in neutron transport under Case eigenfunction formalism is considered. → These equations reduce to integral expressions involving X functions. → Gauss quadrature is not ideal but DE quadrature is well-suited. → Several fold decrease in computational effort with improved accuracy is realisable. - Abstract: In this paper benchmark numerical results for the one-speed criticality problem with isotropic scattering for the slab and sphere are reported. The Fredholm integral equations of the second kind based on the Case eigenfunction formalism are numerically solved by Neumann iterations with the Double Exponential quadrature.
Benchmark calculations on neutrons streaming through mazes at proton accelerator facilities
International Nuclear Information System (INIS)
In accelerator shielding designs one of the important issues is to estimate radiation streaming through mazes and ducts. In order to validate the accuracy of the calculation methods concerning such neutron streaming, benchmark analyses were carried out using two kinds of benchmark problems based on past experiments. The analyses showed that the design methods were applicable to neutron streaming calculations of proton accelerator facilities with an uncertainty within a factor of two. In the analyses, relative comparisons were conducted using a radiation source generated by GeV energy protons, and absolute comparisons were conducted using a low-energy neutron source of a few tens of MeV. A radiation streaming experiment was planned and carried out at KEK using a radiation source produced by a thin copper target irradiated by 12 GeV protons. The preliminary experimental analysis is presented below. In addition, the authors propose to compile benchmark problems on radiation streaming for accelerator facilities and to search for possible new streaming experiments at other facilities. (authors)
International Nuclear Information System (INIS)
This paper presents a benchmark framework established as a basis for investigation of the validity of multi-group approximation with respect to the continuous energy approach, of the level of spatial homogenization with respect to heterogeneous solution, and of the level of angular approximation to the linear Boltzmann transport equation in respect to the Monte Carlo reference solution. Several steady-state solutions of this benchmark have been generated using three different computer codes focusing on the two-dimensional (2-D) geometry model. MCNP5 has been used to generate the reference solution using the continuous energy library. HELIOS is then used for both to solve the problem using a 45 group cross-section library and to generate new sets of few-group cross-sections for the core simulator NEM. The results from the diffusion option of the NEM code on pin-by-pin and Fuel Assembly (FA) basis are presented and discussed in the paper. The benchmark is being designed for evaluation of number of energy groups (number of energy groups and energy cut off points) and spatial (homogenized assembly level vs. homogenized pin cell level) representation needed for high-fidelity reactor core calculation schemes developed at the Pennsylvania State Univ. such as NEM SP3, hybrid NEM-BEM and some recent developments of embedded three-dimensional pin-by-pin diffusion / SP3 finite element calculation schemes. (authors)
Update of KASHIL-E6 library for shielding analysis and benchmark calculations
Energy Technology Data Exchange (ETDEWEB)
Kim, D. H.; Kil, C. S.; Jang, J. H. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
2004-07-01
For various shielding and reactor pressure vessel dosimetry applications, a pseudo-problem-independent neutron-photon coupled MATXS-format library based on the last release of ENDF/B-VI has been generated as a part of the update program for KASHIL-E6, which was based on ENDF/B-VI.5. It has VITAMIN-B6 neutron and photon energy group structures, i.e., 199 groups for neutron and 42 groups for photon. The neutron and photon weighting functions and the Legendre order of scattering are same as KASHIL-E6. The library has been validated through some benchmarks: the PCA-REPLICA and NESDIP-2 experiments for LWR pressure vessel facility benchmark, the Winfrith Iron88 experiment for validation of iron data, and the Winfrith Graphite experiment for validation of graphite data. These calculations were performed by the TRANSXlDANTSYS code system. In addition, the substitutions of the JENDL-3.3 and JEFF-3.0 data for Fe, Cr, Cu and Ni, which are very important nuclides for shielding analyses, were investigated to estimate the effects on the benchmark calculation results.
International Nuclear Information System (INIS)
The paper gives a brief survey of the fifth three-dimensional dynamic Atomic Energy Research benchmark calculation results received with the code DYN3D/ATHLET at NRI Rez. This benchmark was defined at the seventh Atomic Energy Research Symposium (Hoernitz near Zittau, 1997). Its initiating event is a symmetrical break of the main steam header at the end of the first fuel cycle and hot shutdown conditions with one stuck out control rod group. The calculations were performed with the externally coupled codes ATHLET Mod.1.1 Cycle C and DYN3DH1.1/M3. The standard WWER-440/213 input deck of ATHLET code was adopted for benchmark purposes and for coupling with the code DYN3D. The first part of paper contains a brief characteristics of NPP input deck and reactor core model. The second part shows the time dependencies of important global and local parameters. In comparison with the results published at the eighth Atomic Energy Research Symposium (Bystrice nad Pernstejnem, 1998), the results published in this paper are based on improved ATHLET descriptions of control and safety systems. (Author)
Results of the fifth three-dimensional dynamic atomic energy research benchmark problem calculation
International Nuclear Information System (INIS)
The pare gives a brief survey of the fifth three-dimensional dynamic atomic energy research benchmark calculation results received with the code DYN3D/ATHLET at NRI Rez. This benchmark was defined at the seventh AER Symposium. Its initiating event is a symmetrical break of the main steam header at the end of the first fuel cycle and hot shutdown conditions with one stuck out control rot group. The calculations were performed with the externally coupled codes ATHLET Mod.1.1 Cycle C and DYN3DH1.1/M3. The Kasseta library was used for the generation of reactor core neutronic parameters. The standard WWER-440/213 input deck of ATHLET code was adopted for benchmark purposes and for coupling with the code DYN3D. The first part of paper contains a brief characteristics of NPP input deck and reactor core model. The second part shows the time dependencies of important global, fuel assembly and loops parameters.(Author)
International Nuclear Information System (INIS)
The JIPNR-Sosny of the NAS of Belarus created and explored a number of uranium-containing critical assemblies of the BTS series in designing fast BRIG-300 reactor with N2O4 ↔ 2NO2 ↔ 2NO + O2 coolant and the PVER fast-resonance neutron spectrum reactor with a steam-water coolant. Research in the physics of these reactors was performed on fast-thermal critical assemblies at the critical facility Roza. Structurally, these critical assemblies consisted of fast and thermal reactor cores and the buffer zones located between them, intended for leakage spectrum neutron conversion from a thermal zone to a spectrum of neutrons of the modelled fast reactor. Fast zones are a non-uniform hexagonal lattice of cylindrical fuel rods with fuel composition based on metal U (90% U-235), UO2 (36% U-235), depleted U (0.4% U-235), rods with SiO2; a buffer zone is a non-uniform hexagonal lattice of cylindrical fuel rods based on UO2 (36% U-235), natural U and depleted U (0.4% U-235), rods with B4C and made from stainless steel; a thermal zone is a uniform rectangular uranium-polyethylene lattice of cylindrical fuel rods based on the fuel composition UO2+Mg (10% U-235). For obtaining benchmark data on the criticality, computational models have been developed and the analysis of experiments has been carried out to estimate the experimental results as criticality benchmark data. The JIPNR-Sosny of the NAS of Belarus also prepared experiments on the criticality of multiplying systems simulating some physical features of the core of fast low power small-size gas-cooled reactors with UZrCN nuclear fuel. For these purposes, the critical assemblies P-20 were developed at the critical facility “Giacint”. These assemblies represent a uniform hexagonal lattice of fuel cassette: the central area is based on cylindrical fuel rods with UZrCN (19.75% U-235), the peripheral area is based on the cylindrical fuel rods with metallic U (90% U-235), UO2 (36% U-235) and natural U; and the reflector on
Benchmarking of MCNP against B ampersand W LRC Core XI critical experiments
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The MCNP Monte Carlo code and its ENDF/B-V continuous-energy cross- section library previously has been benchmarked against a variety of critical experiments, and that benchmarking recently has been extended to include its ENDF/B-VI continuous-energy cross-section library and additional critical experiments. This study further extends the benchmarking of MCNP and its two continuous-energy libraries to 17 large-scale mockup experiments that closely resemble the core of a pressurized water reactor (PWR). The experiments were performed at Babcock ampersand Wilcox's Lynchburg Research Center in 1970 and 1971. The series was designated as Core XI, and the individual experiments were characterized as different ''loadings.'' The experiments were performed inside a large aluminum tank that contained borated water. The water height for each loading was exactly 145 cm, and the soluble boron concentration in the water was adjusted until the configuration was slightly supercritical, with a value of 1.0007 for keff. Pin-by-pin power distributions were measured for several of the loadings
OECD/NEA Benchmark Calculations for an Accelerator-Driven Minor Actinide Burner
International Nuclear Information System (INIS)
Noticing the current interest in accelerator-driven systems as actinide waste burners, the OECD/NEA has organised an international benchmark exercise for evaluating the performance of computational tools and nuclear data for this type of system. The benchmark model simulates a lead-bismuth cooled sub-critical system driven by a beam of 1 GeV protons. The core design is similar to that of an ALMR, and the fuel composition is typical for a minor actinide burner in a 'double strata' fuel cycle. Lead-bismuth was chosen as target material. Since the intention was to validate data and codes in the energy region below 20 MeV, a predefined spallation neutron source was provided to the benchmark participants. The solutions from seven organisations (ANL, CIEMAT, KAERI, JAERI, PSI/CEA, RIT and SCK-CEN) are based on three different basic data libraries (ENDF/B-VI, JEF-2.2 and JENDL-3.2) and both deterministic and Monte Carlo reactor codes. Significant discrepancies are observed for important neutronic parameters such as initial keff, burn-up reactivity swing and flux distribution. Additional investigations of the basic nuclear data, the data processing methods and the approximations for the reactor simulation will be necessary to understand the origin of all observed discrepancies. (authors)
Intercomparison of Monte Carlo and SN sensitivity calculations for a 14 MeV neutron benchmark
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An inter-comparison has been performed of probabilistic and deterministic sensitivity calculations with the objective to check and validate the Monte Carlo technique for calculating point detector sensitivities as being implemented in MCSEN, a local version of the MCNP4A code. A suitable 14 MeV neutron benchmark problem on an iron assembly has been considered to this end. Good agreement has been achieved for the calculated individual sensitivity profiles, the uncertainties and the neutron flux spectra as well. It is concluded that the Monte Carlo technique for calculating point detector sensitivities and related uncertainties as being implemented in MCSEN is well qualified for sensitivity and uncertainty analyses of fusion neutronics integral experiments. (orig.)
DeCART code verifications by numerical benchmark calculations of HTTR
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DeCART code verifications have been performed through the numerical benchmark calculations of HTTR. The reference calculations have been carried out using the Monte Carlo McCARD code in which a double heterogeneity model was used. Verification results show that the DeCART code gives less negative MTC and RTC than the McCARD code does and thus the DeCART code underestimates the multiplication factors at states with high moderator and reflector temperatures. However, the DeCART code predicts more negative FTC than McCARD code does. In the depletion calculation for the HTTR single cell and single block, the error of the DeCART code increases with burnup. While the DeCART code error in a 2-dimensional core depletion calculation decreases with burnup up to around 500 FPD. (author)
JNC results of BN-600 hybrid core benchmark calculations (3-D)
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This paper presents the phase 2 calculation results of the benchmark core of the BN-600 reactor (3-d modelling). The analytical method applied included the following: JENDL-3.2 nuclear data library; 70 group ABBN type self shielding factor table for group constants; reference delayed neutron yield and spectrum adopted; effective cross section obtained by SLAROM code; basic calculation done by using 18 group two dimensional RZ model (CITATION code) with region dependent fission spectra; transport theory and mesh size correction (TWOTRAN code); perturbation calculation done by diffusion, first order perturbation reactivity mapping method (PERKY code). Calculation results include effective multiplication factors; fuel Doppler constants; steel Doppler constants; sodium density coefficient; steel density coefficients; fuel density coefficient; absorber density coefficient; axial and radial expansion coefficients; dynamic parameters; power distribution; beta and neutron life time; reaction rate distribution
Validation of the Monteburns code for criticality calculation of TRIGA reactors
Energy Technology Data Exchange (ETDEWEB)
Dalle, Hugo Moura [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil); Jeraj, Robert [Jozef Stafan Institute, Ljubljana (Slovenia)
2002-07-01
Use of Monte Carlo methods in burnup calculations of nuclear fuel has become practical due to increased speed of computers. Monteburns is an automated computational tool that links the Monte Carlo code MCNP with the burnup and decay code ORIGEN2.1. This code system was used to simulate a criticality benchmark experiment with burned fuel on a TRIGA Mark II research reactor. Two core configurations were simulated and k{sub eff} values calculated. The comparison between the calculated and experimental values shows good agreement, which indicates that the MCNP/Monteburns/ORIGEN2.1 system gives reliable results for neutronic simulations of TRIGA reactors. (author)
International Nuclear Information System (INIS)
The International Criticality Safety Benchmark Evaluation Project (ICSBEP) was initiated in 1992 and has become a major internationally recognized program. The purpose of the ICSBEP is to identify, evaluate, verify, and formally document a comprehensive and internationally peer-reviewed set of criticality safety benchmark data. The work of the ICSBEP is published as an Organization for Economic Cooperation and Development (OECD) handbook entitled 'International Handbook of Criticality Safety Benchmark Experiments'. More than 150 scientists from around the world have combined their efforts to produce this handbook, which currently spans more than 19 000 pages and contains benchmark specifications for more than 2352 critical configurations. The handbook is intended for use by criticality safety analysts to perform necessary validations of their calculational techniques. The 2001 edition of the 'International Handbook of Criticality Safety Benchmark Experiments' is scheduled for publication in September of 2001 and could contain as many as 30 new evaluations of experimental data. Included in the list of 'in-progress' evaluations are: 1. the ZPPR-21 experiments entitled 'Criticality Studies for Integral Fast Reactors'; 2. RAPSODIE mixed plutonium/uranium fuel rods in water (IPSN, France); 3. mixed uranium/plutonium (29.87%) nitrate solutions poisoned with gadolinium (IPSN, France); 4. PuO2-UO2-polystyrene cubes with poison plates (Westinghouse SMS, United States); 5. highly enriched uranyl nitrate solution in steel containers with 'pipe' intersections (Westinghouse SMS, United States); 6. HEU metal in oil (Westinghouse SMS, United States); 7. an evaluation of experiments with liquid mixtures of HEU hexafluoride and hydrofluoric acid (IPSN, France); 8. critical experiments of stainless steel clad SPERT fuel in water (INEEL, United States); 9. HEU foils reflected by SiO2 and polyethylene (LANL, United States); 10. Un-reflected highly enriched uranyl nitrate subcritical
Benchmark calculation with improved VVER-440/213 RPV CFD model
International Nuclear Information System (INIS)
A detailed RPV model of WWER-440/213 type reactors was developed in BME NTI in the last years. This model contains the main structural elements as inlet and outlet nozzles, guide baffles of hydro-accumulators, alignment drifts, perforated plates, brake- and guide tube chamber and simplified core. For the meshing and simulations ANSYS software's (ICEM 12.0 and CFX 12.0) were used. With the new vessel model a series of parameter studies were performed considering turbulence models, discretization schemes, and modeling methods. In steady state the main results were presented on last AER Symposium in Varna. The model is suitable for different transient calculations as well. The purpose of the suggested new benchmark (seventh Dynamic AER Benchmark) is to investigate the reactor dynamic effects of coolant mixing in the WWER-440/213 reactor vessel and to compare the different codes. The task of this benchmark is to investigate the start up of the sixth main coolant pump. The computation was carried out with the help of ATHLET/BIPRVVER code in Kurchatov Institute for this transient and was repeated with ANSYS CFX 12.0 at our Institute. (Authors)
International Nuclear Information System (INIS)
A benchmark analysis of the transient BFBT data [1], measured in an 8x8 fuel assembly design under typical BWR transient conditions, was performed using the VIPRE-W/MEFISTO-T code package. This is a continuation of the BFBT steady-state benchmark activities documented in [2] and [3]. All available transient void and pressure drop experimental data were considered and the measurements were compared with the predictions of the VIPRE-W sub-channel analysis code using various modeling approaches, including the EPRI drift flux void correlation. Detailed analyses of the code results were performed and it was demonstrated that the VIPRE-W transient predictions are generally reliable over the tested conditions. Available transient dryout data were also considered and the measurements were compared with the predictions of the VIPRE-W/ MEFISTO-T film flow calculations. The code calculates the transient multi-film flowrate distributions in the BFBT bundle, including the effect of spacer grids on drop deposition enhancement, and the dryout criterion corresponds to the total liquid film disappearance. After calibration of the grid enhancement effect with a very small subset of the steady-state critical power database, the code could predict the time and location of transient dryout with very good accuracy. (author)
MCNP benchmark calculation: GCFR grid-plate shield design, configuration II.A
International Nuclear Information System (INIS)
This report describes the Monte Carlo MCNP analysis of one of the GCFR Shield Design experimental configurations which has been constructed and analyzed at the Test Shielding Facility in ORNL. It is a part of the benchmarking program for MCNP, which has been agreed upon with HRB, Mannheim. The calculated response results for the selected detectors agree within 10 % with the measured ones, what can be considered as a very good agreement. The code appears to be a reliable tool for the analysis of similar systems. (author)
Calculations to an IAHR-benchmark test using the CFD-code CFX-4
Energy Technology Data Exchange (ETDEWEB)
Krepper, E.
1998-10-01
The calculation concerns a test, which was defined as a benchmark for 3-D codes by the working group of advanced nuclear reactor types of IAHR (International Association of Hydraulic Research). The test is well documented and detailed measuring results are available. The test aims at the investigation of phenomena, which are important for heat removal at natural circulation conditions in a nuclear reactor. The task for the calculation was the modelling of the forced flow field of a single phase incompressible fluid with consideration of heat transfer and influence of gravity. These phenomena are typical also for other industrial processes. The importance of correct modelling of these phenomena also for other applications is a motivation for performing these calculations. (orig.)
Benchmark calculations of neutron dose rates at transport and storage casks
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The application of numerical calculations methods for demonstration of sufficient radiation shielding of radioactive waste transport and storage casks requires a validation based on appropriate measurements of gamma and neutron sources. The results of the comparison of measured data and calculations using the Monte Carlo program MCNP show deviations dependent on the loading of the cask within the standard deviation which is dominated by the measuring method. Considering the neutrons scattered at the salt MCNP (in case of disposal in the salt) tends to underestimate the nominal values, but still within the double standard deviation. This accuracy is not reached with MAVRIC. Based on AHE (active handling experiments) data benchmark calculations were performed that can be used as reference value. The total accuracy results from the accuracy of the source term and the measurement of the neutron dose rate with a deviation of 15%.
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The Static Experiment Critical Facility, STACY was constructed in the Nuclear Fuel Cycle Safety Engineering Research Facility, NUCEF of the Japan Atomic Energy Research Institute in order to produce the fundamental critical data of uranyl nitrate solution, plutonium nitrate solution and their mixture. A series of experiments using single core tank have been performed using 10% enriched uranyl nitrate solution since the first criticality in 1995. Benchmark data of STACY are now used for verification of Japanese criticality safety code system and nuclear data libraries. Kinetic parameters, temperature coefficients and reflector effects of structural material are also measured using single homogeneous core. It is on schedule to make experiments for neutron interaction effect and for simulating the dissolving process with a heterogeneous core using low enriched uranyl nitrate solution. After these experiments, systematic critical and subcritical experiments on plutonium nitrate solution will start in five years. This paper reviews the main results of STACY since the initial criticality and describes the criticality properties of the experimental cores in the future program. (author)
Calculation of Upper Subcritical Limits for Nuclear Criticality in a Repository
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The purpose of this document is to present the methodology to be used for development of the Subcritical Limit (SL) for post closure conditions for the Yucca Mountain repository. The SL is a value based on a set of benchmark criticality multiplier, keff results that are outputs of the MCNP calculation method. This SL accounts for calculational biases and associated uncertainties resulting from the use of MCNP as the method of assessing keff. The context for an SL estimate include the range of applicability (based on the set of MCNP results) and the type of SL required for the application at hand. This document will include illustrative calculations for each of three approaches. The data sets used for the example calculations are identified in Section 5.1. These represent three waste categories, and SLs for each of these sets of experiments will be computed in this document. Future MCNP data sets will be analyzed using the methods discussed here. The treatment of the biases evaluated on sets of keff results via MCNP is statistical in nature. This document does not address additional non-statistical contributions to the bias margin, acknowledging that regulatory requirements may impose additional administrative penalties. Potentially, there are other biases or margins that should be accounted for when assessing criticality (keff). Only aspects of the bias as determined using the stated assumptions and benchmark critical data sets will be included in the methods and sample calculations in this document. The set of benchmark experiments used in the validation of the computational system should be representative of the composition, configuration, and nuclear characteristics for the application at hand. In this work, a range of critical experiments will be the basis of establishing the SL for three categories of waste types that will be in the repository. The ultimate purpose of this document is to present methods that will effectively characterize the MCNP computations
Comparisons of the MCNP criticality benchmark suite with ENDF/B-VI.8, JENDL-3.3, and JEFF-3.0
International Nuclear Information System (INIS)
A comparative study has been performed with the latest evaluated nuclear data libraries ENDF/B-VI.8, JENDL-3.3, and JEFF-3.0. The study has been conducted through the benchmark calculations for 91 criticality problems with the libraries processed for MCNP4C. The calculation results have been compared with those of the ENDF60 library. The self-shielding effects of the unresolved-resonance (UR) probability tables have also been estimated for each library. The χ2 differences between the MCNP results and experimental data were calculated for the libraries. (author)
Criticality safety calculations of storage canisters
International Nuclear Information System (INIS)
In the planned Swedish repository for deep disposal of spent nuclear fuel the fuel assemblies will be stored in storage canisters made of cast iron and copper. To assure safe storage of the fuel the requirement is that the normal criticality safety criteria have to be met. The effective neutron multiplication factor must not exceed 0.95 in the most reactive conditions including different kinds of uncertainties. In this report it is shown that the criteria could be met if credit for the reactivity decrease due to the burn up of the fuel is taken into account. The criticality safety criteria are based on the US NRC regulatory requirements for transportation and storage of spent fuel
Energy Technology Data Exchange (ETDEWEB)
Renner, Franziska
2014-10-02
Monte Carlo simulations are regarded as the most accurate method of solving complex problems of radiation transport. Therefore, they have great potential to realize more exact dose calculations for treatment planning in radiation therapy. However, there is a lack of information on how correct the results of Monte Carlo calculations are on an absolute basis. A practical verification of the calculations can be performed by direct comparison with a benchmark experiment. Thereby, the uncertainties of the experimental result and of the simulation also have to be considered to make a meaningful comparison between the experiment and the simulation possible. This dissertation presents a benchmark experiment and its results, including the uncertainty, which can be used to test the accuracy of Monte Carlo calculations in the field of radiation therapy. The experiment was planned to have parallels to clinical radiation therapy, among other things, with respect to the radiation applied, the materials used and the manner of dose detection. The benchmark experiment aimed at an absolute comparison with a simulation result and because of this it was necessary to use a special research accelerator as a radiation source in the experiment. The accurate characterization of the accelerator beam was a precondition to define a realistic radiation source for the Monte Carlo simulation. Therefore, this work also deals with the characterization of the source and investigations regarding the X-ray target used. Additionally, the dissertation contains the verification of the widely used Monte Carlo program EGSnrc by the benchmark experiment. The simulation of the experiment by EGSnrc, the results and the estimation of the uncertainty related to the simulation are documented in this work.The results and findings of this dissertation end in a comparison between the results of the benchmark experiment and the corresponding calculations with EGSnrc. The benchmark experiment and the simulations
International Nuclear Information System (INIS)
The reliability of calculation tools to evaluate and calculate dose rates appearing behind multi-layered shields is important with regard to the certification of transport and storage casks. Actual benchmark databases like SINBAD do not offer such configurations because they were developed for reactor and accelerator purposes. Due to this, a bench-mark-suite based on own experiments that contain dose rates measured in different distances and levels from a transport and storage cask and on a public benchmark to validate Monte-Carlo-transport-codes has been developed. The analysed and summarised experiments include a 60Co point-source located in a cylindrical cask, a 252Cf line-source shielded by iron and polyethylene (PE) and a bare 252Cf source moderated by PE in a concrete-labyrinth with different inserted shielding materials to quantify neutron streaming effects on measured dose rates. In detail not only MCNPTM (version 5.1.6) but also MAVRIC, included in the SCALE 6.1 package, have been compared for photon and neutron transport. Aiming at low deviations between calculation and measurement requires precise source term specification and exact measurements of the dose rates which have been evaluated carefully including known uncertainties. In MAVRIC different source-descriptions with respect to the group-structure of the nuclear data library are analysed for the calculation of gamma dose rates because the energy lines of 60Co can only be modelled in groups. In total the comparison shows that MCNPTM fits very wall to the measurements within up to two standard deviations and that MAVRIC behaves similarly under the prerequisite that the source-model can be optimized. (author)
International Nuclear Information System (INIS)
Highlights: • We analize the performance of neutron scattering libraries for D and O in D2O for nuclear criticality calculations. • We calculated 65 ICSBEP benchmark cases from 8 heavy water moderated thermal systems using MCNP5. • A significant improvement is found when our library is combined with the ROSFOND-2010 evaluation for deuterium. • In 48 of the 65 benchmark cases we obtained a C/E ratio closer to 1.0. • The percentage of benchmark cases calculated within 1-sigma increases from 42% to 82%, compared to ENDF/B-VII calculations. - Abstract: To improve the evaluations in thermal sublibraries, we developed a set of thermal neutron scattering cross sections (scattering kernels) for the deuterium and oxygen bound in heavy water in the ENDF-6 format. These new libraries are based on molecular dynamics simulations and recent experimental data, and result in an improvement of the calculation of single neutron scattering quantities. In this work, we show how the use of this new set of cross sections also improves the calculation of thermal critical systems moderated and/or reflected with heavy water. The use of the new thermal scattering library for heavy water, combined with the ROSFOND-2010 evaluation of the deuterium cross sections, results in an improvement of the C/E ratio in 48 out of 65 benchmark cases calculated with the Monte Carlo code MCNP5, in comparison with the existing library based on the ENDF/B-VII evaluation
Benchmark calculation of APOLLO2 and SLAROM-UF in a fast reactor lattice
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A lattice cell benchmark calculation is carried out for APOLLO2 and SLAROM-UF on the infinite lattice of a simple pin cell featuring a fast reactor. The accuracy in k-infinity and reaction rates is investigated in their reference and standard level calculations. In the 1st reference level calculation, APOLLO2 and SLAROM-UF agree with the reference value of k-infinity obtained by a continuous energy Monte Carlo calculation within 50 pcm. However, larger errors are observed in a particular reaction rate and energy range. A major problem common to both codes is in the cross section library of 239Pu in the unresolved energy range. In the 2nd reference level calculation, which is based on the ECCO 1968 group structure, both results of k-infinity agree with the reference value within 100 pcm. The resonance overlap effect is observed by several percents in cross sections of heavy nuclides. In the standard level calculation based on the APOLLO2 library creation methodology, a discrepancy appears by more than 300 pcm. A restriction is revealed in APOLLO2. Its standard cross section library does not have a sufficiently small background cross section to evaluate the self-shielding effect of 56Fe cross sections. The restriction can be removed by introducing the mixture self-shielding treatment recently introduced to APOLLO2. SLAROM-UF original standard level calculation based on the JFS-3 library creation methodology is the best among the standard level calculations. Improvement from the SLAROM-UF standard level calculation is achieved mainly by use of a proper weight function for light or intermediate nuclides. (author)
Benchmark calculation of APOLLO-2 and SLAROM-UF in a fast reactor lattice
International Nuclear Information System (INIS)
A lattice cell benchmark calculation is carried out for APOLLO2 and SLAROM-UF on the infinite lattice of a simple pin cell featuring a fast reactor. The accuracy in k-infinity and reaction rates is investigated in their reference and standard level calculations. In the 1. reference level calculation, APOLLO2 and SLAROM-UF agree with the reference value of k-infinity obtained by a continuous energy Monte Carlo calculation within 50 pcm. However, larger errors are observed in a particular reaction rate and energy range. The major problem common to both codes is in the cross section library of 239Pu in the unresolved energy range. In the 2. reference level calculation, which is based on the ECCO 1968 group structure, both results of k-infinity agree with the reference value within 100 pcm. The resonance overlap effect is observed by several percents in cross sections of heavy nuclides. In the standard level calculation based on the APOLLO2 library creation methodology, a discrepancy appears by more than 300 pcm. A restriction is revealed in APOLLO2. Its standard cross section library does not have a sufficiently small background cross section to evaluate the self shielding effect on 56Fe cross sections. The restriction can be removed by introducing the mixture self-shielding treatment recently introduced to APOLLO2. SLAROM-UF original standard level calculation based on the JFS-3 library creation methodology is the best among the standard level calculations. Improvement from the SLAROM-UF standard level calculation is achieved mainly by use of a proper weight function for light or intermediate nuclides. (author)
Concrete Spent Fuel Cask Criticality Calculation
International Nuclear Information System (INIS)
A preliminary analysis of the concrete cask for the intermediate dry storage of the spent fuel of NPP Krsko should include an estimation of the effective multiplication factor. Assuming 16x16 fuel elements, 4.3% initial enrichment, 45 GWd/tU burnup and 10 years cooling time, a concrete spent fuel capacity of 10 spent fuel assemblies is proposed. Fuel assemblies are placed inside inner cavity in a 'basket' - a boron (1%) doped steel structure. Heavy concrete (25% Fe), 45 cm thick, is enclosed in a carbon steel shell. There is also a stainless steel (SS304) lining of the storage cavity. Isotope inventory of the spent fuel after a 10 years cooling time is calculated using ORIGEN-S functional module of the SCALE-4.2 code package. The effective multiplication factor keff of dry (helium filled) and wet (water filled) cask for fresh and used fuel is calculated using CSAS4 Monte Carlo method based control module of the same SCALE-4.2 code package. The obtained results of keff of the dry cask for fresh and spent fuel are well below the required 0.95 value, but those for the water filled cask are above this value. Therefore, several additional calculations of the keff varying the thickness of a boral basket structure which had replaced the stainless steel one were done. It turned out that at least a 1.5 cm thick boral wall was needed to meet the required 0.95 value for keff. (author)
International Nuclear Information System (INIS)
The criticality analysis of the TRIGA-II benchmark experiment at the Musashi Institute of Technology Research Reactor (MuITR, 100kW) was performed by the three-dimensional continuous-energy Monte Carlo code (MCNP4A). To minimize errors due to an inexact geometry model, all fresh fuels and control rods as well as vicinity of the core were precisely modeled. Effective multiplication factors (keff) in the initial core critical experiment and in the excess reactivity adjustment for the several fuel-loading patterns as well as the fuel element reactivity worth distributions were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated keff overestimated the experimental data by about 1.0%Δk/k for both the initial core and the several fuel-loading arrangements (fuels or graphite elements were added only to the outer-ring), but the discrepancy increased to 1.8%Δk/k for the some fuel-loading patterns (graphite elements were inserted into the inner-ring). The comparison result of the fuel element worth distribution showed above tendency. All in all, the agreement between the MCNP predictions and the experimentally determined values is good, which indicates that the Monte Carlo model is enough to simulate criticality of the TRIGA-II reactor. (author)
An Analytical Benchmark for the Calculation of Current Distribution in Superconducting Cables
Bottura, L; Fabbri, M G
2002-01-01
The validation of numerical codes for the calculation of current distribution and AC loss in superconducting cables versus experimental results is essential, but could be affected by approximations in the electromagnetic model or incertitude in the evaluation of the model parameters. A preliminary validation of the codes by means of a comparison with analytical results can therefore be very useful, in order to distinguish among different error sources. We provide here a benchmark analytical solution for current distribution that applies to the case of a cable described using a distributed parameters electrical circuit model. The analytical solution of current distribution is valid for cables made of a generic number of strands, subjected to well defined symmetry and uniformity conditions in the electrical parameters. The closed form solution for the general case is rather complex to implement, and in this paper we give the analytical solutions for different simplified situations. In particular we examine the ...
Calculation of four thermal reactor benchmark problems in X-Y geometry
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Four LWR benchmark problems in X-Y geometry are calculated using the ''Surface Flux'' transport code SURCU based on the solution of the integral transport equation. Quadruple spherical harmonics (QPN), orthogonal in angular quadrants, are used to approximate the incoming and outgoing surface fluxes in angular quadrants at the interface between space intervals, while the spatial distribution of neutron flux and sources is represented by a Legendre polynomial expansion. Results are compared with those obtained from the following codes: (1) TWOTRAN-II and DOT 3.5, based on Ssub(N) theory and employing flat and linear flux approximations in space, (2) MARSYAS and TRITON, based on first collision probability theory with flat or linear flux approximations in space, and (3) QP1, using spherical harmonics up to first order to approximate incoming and outgoing currents in angular quadrants at the mesh interfaces and up to first order Legendre polynomial flux and source approximations in space. (Auth.)
The University of Pisa calculations for the Phase I of the OECD/NEA UAM Benchmark
International Nuclear Information System (INIS)
In this paper we present the Univ. of Pisa preliminary results for the first exercise of the Phase I of the OECD/NEA Benchmark on the Uncertainty in Analysis and Modeling. The scope of exercise one is to address the uncertainties due to the basic nuclear data as well as the impact of processing the nuclear and covariance data, selection of multi-group structure and self-shielding treatment. DRAGON code and TSUNAMI code were employed, using the available covariance data matrix. The execution of DRAGON calculations required the use of ANGELO and LAMBDA codes for the extension of the covariance matrix from the original SCALE 44 group structure to DRAGON 69 group structure. The uncertainties for the main cross sections were evaluated and are presented here. (authors)
International Nuclear Information System (INIS)
The kyoto university reactor physics experiments on the university critical assembly is used to benchmark validate the NCNSRC calculations methodology. This methodology has two lines, diffusion and Monte Carlo. The diffusion line includes the codes WIMSD4 for cell calculations and the two dimensional diffusion code DIXY2 for core calculations. The transport line uses the MULTIKENO-Code vax Version. Analysis is performed for the criticality, and the temperature coefficients of reactivity (TCR) for the light water moderated and reflected cores, of the different cores utilized in the experiments. The results of both Eigen value and TCR approximately reproduced the experimental and theoretical Kyoto results. However, some conclusions are drawn about the adequacy of the standard wimsd4 library. This paper is an extension of the NCNSRC efforts to assess and validate computer tools and methods for both Et-R R-1 and Et-MMpr-2 research reactors. 7 figs., 1 tab
Study on the conservative factors for burnup credit criticality calculation
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When applies the burnup credit technology to perform criticality safety analysis for spent fuel storage or transportation problems, it is important for one to confirm that all the conditions adopted are adequate to cover the severest conditions that may encounter in the engineering applications. Taking the OECD/NEA burnup credit criticality benchmarks as sample problems, we study the effect of some important factors that may affect the conservatism of' the results for spent fuel system criticality safety analysis. Effects caused by different nuclides credit strategy, different cooling time and axial burnup profile are studied by use of the STARBUCS module of SCALE5. 1 software package, and related conclusions about the conservatism of these factors are drawn. (authors)
IAEA GT-MHR Benchmark Calculations Using the HELIOS/MASTER Two-Step Procedure
Energy Technology Data Exchange (ETDEWEB)
Lee, Kyung Hoon; Kim, Kang Seog; Cho, Jin Young; Song, Jae Seung; Noh, Jae Man; Lee, Chung Chan; Zee, Sung Quun
2007-05-15
A new two-step procedure based on the HELISO/MASTER code system has been developed for the prismatic VHTR physics analysis. This procedure employs the HELIOS code for the transport lattice calculation to generate a few group constants, and the MASTER code for the 3-dimensional core calculation to perform the reactor physics analysis. Double heterogeneity effect due to the random distribution of the particulate fuel could be dealt with the recently developed reactivity-equivalent physical transformation (RPT) method. The strong spectral effects of the graphite moderated reactor core could be solved both by optimizing the number of energy groups and group boundaries, and by employing a partial core model instead of a single block one to generate a few group cross sections. Burnable poisons in the inner reflector and asymmetrically located large control rod can be treated by adopting the equivalence theory applied for the multi-block models to generate surface dependent discontinuity factors. Effective reflector cross sections were generated by using a simple mini-core model and an equivalence theory. In this study the IAEA GT-MHR benchmark problems with a plutonium fuel were analyzed by using the HELIOS/MASTER code package and the Monte Carlo code MCNP. Benchmark problems include pin, block and core models. The computational results of the HELIOS/MASTER code system were compared with those of MCNP and other participants. The results show that the 2-step procedure using HELIOS/MASTER can be applied to the reactor physics analysis for the prismatic VHTR with a good accuracy.
Dose Rate Experiment at JET for Benchmarking the Calculation Direct One Step Method
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Neutrons produced by D-D and D-T plasmas induce the activation of tokamak materials and of components. The development of reliable methods to assess dose rates is a key issue for maintenance and operating nuclear machines, in normal and off-normal conditions. In the frame of the EFDA Fusion Technology work programme, a computational tool based upon MCNP Monte Carlo code has been developed to predict the dose rate after shutdown: it is called Direct One Step Method (D1S). The D1S is an innovative approach in which the decay gammas are coupled to the neutrons as in the prompt case and they are transported in one single step in the same run. Benchmarking of this new tool with experimental data taken in a complex geometry like that of a tokamak is a fundamental step to test the reliability of the D1S method. A dedicated benchmark experiment was proposed for the 2005-2006 experimental campaign of JET. Two irradiation positions have been selected for the benchmark: one inner position inside the vessel, not far from the plasma, called the 2 upper irradiation end (IE2), where neutron fluence is relatively high. The second position is just outside a vertical port in an external position (EX). Here the neutron flux is lower and the dose rate to be measured is not very far from the residual background. Passive detectors are used for in-vessel measurements: the high sensitivity Thermo Luminescent Dosimeters (TLDs) GR-200A (natural LiF), which ensure measurements down to environmental dose level. An active detector of Geiger-Muller (GM) type is used for out of vessel dose rate measurement. Before their use the detectors were calibrated in a secondary gamma-ray standard (Cs-137 and Co-60) facility in term of air-kerma. The background measurement was carried-out in the period July -September 2005 in the outside position EX using the GM tube and in September 2005 inside the vacuum vessel using TLD detectors located in the 2 Upper irradiation end IE2. In the present work
Verification of HELIOS-MASTER system through benchmark of critical experiments
Energy Technology Data Exchange (ETDEWEB)
Kim, Ha Yong; Kim, Kyo Yun; Cho, Byung Oh; Lee, Chung Chan; Zee, Sung Quun
1999-03-01
HELIOS-MASTER code system is verified through the benchmark of the critical experiments that were performed by RRC Kurchatov Institute with water moderated hexagonally pitched lattices of highly enriched uranium fuel rods (80w/o). We also used the same input by using MCNP code that was described in evaluation report, and compare our results with those of evaluation report. HELIOS developed by Scandpower A/S is a two-dimensional transport program for generation of group cross sections and MASTER developed by KAERI is a three-dimensional nuclear design and analysis code based on the two group diffusion theory. It solves neutronics model with AFEN (Analytic Function Expansion Nodal) method for hexagonal geometry. The results show that HELIOS-MASTER code system is fast and accurate enough so that this code system can be used as nuclear core analysis tool for hexagonal geometry. (author). 4 refs., 4 tabs., 10 figs.
Verification of HELIOS-MASTER system through benchmark of critical experiments
International Nuclear Information System (INIS)
HELIOS-MASTER code system is verified through the benchmark of the critical experiments that were performed by RRC Kurchatov Institute with water moderated hexagonally pitched lattices of highly enriched uranium fuel rods (80w/o). We also used the same input by using MCNP code that was described in evaluation report, and compare our results with those of evaluation report. HELIOS developed by Scandpower A/S is a two-dimensional transport program for generation of group cross sections and MASTER developed by KAERI is a three-dimensional nuclear design and analysis code based on the two group diffusion theory. It solves neutronics model with AFEN (Analytic Function Expansion Nodal) method for hexagonal geometry. The results show that HELIOS-MASTER code system is fast and accurate enough so that this code system can be used as nuclear core analysis tool for hexagonal geometry. (author). 4 refs., 4 tabs., 10 figs
International Nuclear Information System (INIS)
This report provides the specification for the uncertainty exercises of the international OECD/NEA, NRC and NUPEC BFBT benchmark problem including the elemental task. The specification was prepared jointly by Pennsylvania State University (PSU), USA and the Japan Nuclear Energy Safety (JNES) Organisation, in cooperation with the OECD/NEA and the Commissariat a l'energie atomique (CEA Saclay, France). The work is sponsored by the US NRC, METI-Japan, the OECD/NEA and the Nuclear Engineering Program (NEP) of Pennsylvania State University. This uncertainty specification covers the fourth exercise of Phase I (Exercise-I-4), and the third exercise of Phase II (Exercise II-3) as well as the elemental task. The OECD/NRC BFBT benchmark provides a very good opportunity to apply uncertainty analysis (UA) and sensitivity analysis (SA) techniques and to assess the accuracy of thermal-hydraulic models for two-phase flows in rod bundles. During the previous OECD benchmarks, participants usually carried out sensitivity analysis on their models for the specification (initial conditions, boundary conditions, etc.) to identify the most sensitive models or/and to improve the computed results. The comprehensive BFBT experimental database (NEA, 2006) leads us one step further in investigating modelling capabilities by taking into account the uncertainty analysis in the benchmark. The uncertainties in input data (boundary conditions) and geometry (provided in the benchmark specification) as well as the uncertainties in code models can be accounted for to produce results with calculational uncertainties and compare them with the measurement uncertainties. Therefore, uncertainty analysis exercises were defined for the void distribution and critical power phases of the BFBT benchmark. This specification is intended to provide definitions related to UA/SA methods, sensitivity/ uncertainty parameters, suggested probability distribution functions (PDF) of sensitivity parameters, and selected
Design of an efficient calculation model of BWR cold critical experiments for validation
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purposes. This means that the computational effort can be reduced by more than an order of magnitude without significant effect on the final results. For the same reason the work needed to prepare benchmarks from the experiments is reduced significantly. The procedure can be looked upon as being a conservative calculation model for criticality code validation or a way to create simplified benchmarks from cold critical experiments. (author)
Energy Technology Data Exchange (ETDEWEB)
Ellis, R.J.
2001-01-11
A series of unit pin-cell benchmark problems have been analyzed related to irradiation of mixed oxide fuel in VVER-1000s (water-water energetic reactors). One-dimensional, discrete-ordinates eigenvalue calculations of these benchmarks were performed at ORNL using the SAS2H control sequence module of the SCALE-4.3 computational code system, as part of the Fissile Materials Disposition Program (FMDP) of the US DOE. Calculations were also performed using the SCALE module CSAS to confirm the results. The 238 neutron energy group SCALE nuclear data library 238GROUPNDF5 (based on ENDF/B-V) was used for all calculations. The VVER-1000 pin-cell benchmark cases modeled with SAS2H included zero-burnup calculations for eight fuel material variants (from LEU UO{sub 2} to weapons-grade MOX) at five different reactor states, and three fuel depletion cases up to high burnup. Results of the SAS2H analyses of the VVER-1000 neutronics benchmarks are presented in this report. Good general agreement was obtained between the SAS2H results, the ORNL results using HELIOS-1.4 with ENDF/B-VI nuclear data, and the results from several Russian benchmark studies using the codes TVS-M, MCU-RFFI/A, and WIMS-ABBN. This SAS2H benchmark study is useful for the verification of HELIOS calculations, the HELIOS code being the principal computational tool at ORNL for physics studies of assembly design for weapons-grade plutonium disposition in Russian reactors.
Benchmark calculations on residue production within the EURISOL DS project. Part 1: thin targets
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We have begun this benchmark study using mass distribution data of reaction products obtained at GSI in inverse kinematics. This step has allowed us to make a first selection among 10 spallation models; in this way the first assessment of the quality of the models was obtained. Then, in a second part, experimental mass distributions for some elements, which either are interesting as radioactive ion beams or important due to the safety and radioprotection issues (alpha or gamma emitters), will be also compared to model calculations. These data have been obtained for an equivalent 0.8 or 1.0 GeV proton beam, which is approximately the proposed projectile energy. We note that in realistic thick targets the proton beam will be slowed down and some secondary particles will be produced. Therefore, the residual nuclei production at lower energies is also important. For this reason, we also performed in the third part of this work some excitation function calculations and the associated data obtained with gamma-spectroscopy to test the models in a wide projectile energy range. We conclude that INCL4/Abla and Isabel/Abla are the best model combinations which we recommend. We also note that the agreement between model and data are better with 1 GeV protons than with 100-200 MeV protons
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The Spallation Neutron Source (SNS) will provide an intense source of low-energy neutrons for experimental use. The low-energy neutrons are produced by the interaction of a high-energy (1.0 GeV) proton beam on a mercury (Hg) target and slowed down in liquid hydrogen or light water moderators. Computer codes and computational techniques are being benchmarked against relevant experimental data to validate and verify the tools being used to predict the performance of the SNS. The LAHET Code System (LCS), which includes LAHET, HTAPE ad HMCNP (a modified version of MCNP version 3b), have been applied to the analysis of experiments that were conducted in the Alternating Gradient Synchrotron (AGS) facility at Brookhaven National Laboratory (BNL). In the AGS experiments, foils of various materials were placed around a mercury-filled stainless steel cylinder, which was bombarded with protons at 1.6 GeV. Neutrons created in the mercury target, activated the foils. Activities of the relevant isotopes were accurately measured and compared with calculated predictions. Measurements at BNL were provided in part by collaborating scientists from JAERI as part of the AGS Spallation Target Experiment (ASTE) collaboration. To date, calculations have shown good agreement with measurements
Diffusion benchmark calculations of a VVER-440 core with 180 deg symmetry
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A diffusion benchmark of the VVER-440 core with 180 deg symmetry and fixed cross sections is proposed. The new benchmark is the modification of Seidel's 3-dimensional 30 degree benchmark, which plays an important role in the verification and validation of nodal neutronic codes. In the new benchmark the 180 deg symmetry is assured by a stuck eccentric control assembly. The recommended reference solution is derived from diverse solutions of the DIF3D finite difference code. The results of the HEXAN module of the KARATE code system are also presented. (author)
International Nuclear Information System (INIS)
The need to refine models for best-estimate calculations, based on good-quality experimental data, has been expressed in many recent meetings in the field of nuclear applications. The modeling needs arising in this respect should not be limited to the currently available macroscopic methods but should be extended to next-generation analysis techniques that focus on more microscopic processes. One of the most valuable databases identified for the thermalhydraulics modeling was developed by the Nuclear Power Engineering Corporation (NUPEC), Japan. From 1987 to 1995, NUPEC performed steady-state and transient critical power and departure from nucleate boiling (DNB) test series based on the equivalent full-size mock-ups. Considering the reliability not only of the measured data, but also other relevant parameters such as the system pressure, inlet sub-cooling and rod surface temperature, these test series supplied the first substantial database for the development of truly mechanistic and consistent models for boiling transition and critical heat flux. Over the last few years the Pennsylvania State University (PSU) under the sponsorship of the U.S. Nuclear Regulatory Commission (NRC) has prepared, organized, conducted and summarized the OECD/NRC Full-size Fine-mesh Bundle Tests (BFBT) Benchmark. The international benchmark activities have been conducted in cooperation with the Nuclear Energy Agency/Organization for Economic Co-operation and Development (NEA/OECD) and Japan Nuclear Energy Safety (JNES) organization, Japan. Consequently, the JNES has made available the Boiling Water Reactor (BWR) NUPEC database for the purposes of the benchmark. Based on the success of the OECD/NRC BFBT benchmark the JNES has decided to release also the data based on the NUPEC Pressurized Water Reactor (PWR) subchannel and bundle tests for another follow-up international benchmark entitled OECD/NRC PWR Subchannel and Bundle Tests (PSBT) benchmark. This paper presents an application of
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During the past two years, a Working Group established by the Organization for Economic Co-Operation and Development's Nuclear Energy Agency (OECD-NEA) has been developing a set of criticality benchmark problems which could be used to help establish the validity of criticality safety computer programs and their associated nuclear data for calculation of ksub(eff) for spent light water reactor (LWR) fuel transport containers. The basic goal of this effort was to identify a set of actual critical experiments which would contain the various material and geometric properties present in spent LWR transport contrainers. These data, when used by the various computational methods, are intended to demonstrate the ability of each method to accurately reproduce the experimentally measured ksub(eff) for the parameters under consideration
Self organized criticality - analytical calculations and open problems
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Some analytical calculations and results concerning self organized critical state in the sand pile-like cellular automate defined on the Bethe and square lattices are showed. The possibility of achieving a self organized critical state in nonconservative model system is discussed. (author)
Technique for technological calculation of critical flow of boiling water
International Nuclear Information System (INIS)
Average values of friction factor and mach number for a critical flow of boiling water are determined on the basis of computerized processing of experimental data. Empirical formula, relating these values, which can be used for technological calculations of critical conditions of boiling water flow through transport pipelines, is derived
Energy Technology Data Exchange (ETDEWEB)
Kiedrowski, Brian C. [Los Alamos National Laboratory
2012-06-19
Within the last decade, there has been increasing interest in the calculation of cross section sensitivity coefficients of k{sub eff} for integral experiment design and uncertainty analysis. The OECD/NEA has an Expert Group devoted to Sensitivity and Uncertainty Analysis within the Working Party for Nuclear Criticality Safety. This expert group has developed benchmarks to assess code capabilities and performance for doing sensitivity and uncertainty analysis. Phase III of a set of sensitivity benchmarks evaluates capabilities for computing sensitivity coefficients. MCNP6 has the capability to compute cross section sensitivities for k{sub eff} using continuous-energy physics. To help verify this capability, results for the Phase III benchmark cases are generated and submitted to the Expert Group for comparison. The Phase III benchmark has three cases: III.1, an array of MOX fuel pins, III.2, a series of infinite lattices of MOX fuel pins with varying pitches, and III.3 two spheres with homogeneous mixtures of UF{sub 4} and polyethylene with different enrichments.
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The theoretical and adjusted Watt spectrum representations for 235U are used as weighting functions to calculate Keff and θf28/θf25 for the benchmark Godiva. The results obtained show that the values of Keff and θf28/θf25 are not affected by spectrum form change. (author)
Additional nuclear criticality safety calculations for small-diameter containers
International Nuclear Information System (INIS)
This report documents additional criticality safety analysis calculations for small diameter containers, which were originally documented in Reference 1. The results in Reference 1 indicated that some of the small diameter containers did not meet the criteria established for criticality safety at the Portsmouth facility (Keff +2σ<.95) when modeled under various contingency assumptions of reflection and moderation. The calculations performed in this report reexamine those cases which did not meet the criticality safety criteria. In some cases, unnecessary conservatism is removed, and in other cases mass or assay limits are established for use with the respective containers
MCNP Perturbation Capability for Monte Carlo Criticality Calculations
International Nuclear Information System (INIS)
The differential operator perturbation capability in MCNP4B has been extended to automatically calculate perturbation estimates for the track length estimate of keff in MCNP4B. The additional corrections required in certain cases for MCNP4B are no longer needed. Calculating the effect of small design changes on the criticality of nuclear systems with MCNP is now straightforward
MCNP Perturbation Capability for Monte Carlo Criticality Calculations
Energy Technology Data Exchange (ETDEWEB)
Hendricks, J.S.; Carter, L.L.; McKinney, G.W.
1999-09-20
The differential operator perturbation capability in MCNP4B has been extended to automatically calculate perturbation estimates for the track length estimate of k{sub eff} in MCNP4B. The additional corrections required in certain cases for MCNP4B are no longer needed. Calculating the effect of small design changes on the criticality of nuclear systems with MCNP is now straightforward.
Analysis of BFS-62-3A critical experiment benchmark model - IGCAR results
International Nuclear Information System (INIS)
The BFS 62-3A assembly is a full scale model of BN-600 hybrid core. The MOX zone is represented as a ring between medium enriched (MEZ) and high enriched zones (HEZ). The hybrid core with steel reflector is represented in a 120 deg sector of BFS. For a homogenised 3-D core of BFS, equivalent experimental data of keff and SVRE values were derived by including the following corrections to the actually obtained experimental results: (a) heterogeneity effect and (b) 3-D model simplification effect. The nuclear data used was XSET-98. It is a 26 group set with ABBN type self-shielding factor table. The benchmark models were analysed by diffusion theory. 3-D calculations were done by TREDFR code in 26 groups with 6 triangular meshes per fuel assembly. The number of triangles was 24414. Axial mesh size corrections were estimated for some cases. The convergence criteria for were 0.000001 for keff and 0.0001 for point wise fission source. The multiplication factor of the reference core of the benchmark is compared with measured. The multiplication factor is predicted with in the uncertainty margin. The SVRE values were computed as Δk/k1k2 and compared to measured values. It is found that the predictions are with in the uncertainty margin except in the MOX region. Reason for this needs to be investigated. As a first step, axial mesh size effect was estimated for MOX SVRE (sodium void reactivity effect) case with use finer meshes in the reference core as well the MOX voided core. By increasing the axial meshe from 35 to 54 both the keff reduced by the same amount leaving the MOX SVRE worth unchanged
Diffusion benchmark calculations of a WWER-440 core with 180 deg symmetry
International Nuclear Information System (INIS)
A diffusion benchmark of the VVER-440 core with 180 degree symmetry and fixed cross sections is proposed. The new benchmark is the modification of Seidel's 3 dimensional 30 degree benchmark, which plays an important role in the verification and validation of nodal neutronic codes. In the 180 degree symmetry is assured by a stuck eccentric control assembly. The recommended reference solution is derived from diverse solution of the DIF3D finite difference code. The results of the HEXAN module of the KARATE code system are also presented.(Authors)
JNC's review and proposal for BN-600 hybrid core benchmark calculation
International Nuclear Information System (INIS)
This contribution includes questions on benchmark description (geometry, composition, data evaluation) and proposals for the BN-600 benchmark project. Proposals are related to benchmark of cell heterogeneity evaluation (fuel assembly, control rod); additional burnup properties (burnup reactivity loss, fuel composition change); analysis by using the cross section sensitivity method (application of perturbation theory, influence of cross section difference, estimation of analytical method difference); evaluation of BN-600 design value and its errors (best estimated design value of hybrid core, error estimation of the design value)
International Nuclear Information System (INIS)
To justify the use of a commercial Computational Fluid Dynamics (CFD) code for a CANDU fuel channel analysis, especially for the radiation heat transfer dominant conditions, the CFX-10 code is tested against three benchmark problems which were used for the validation of a radiation heat transfer in the CANDU analysis code, a CATHENA. These three benchmark problems are representative of the CANDU fuel channel configurations from a simple geometry to a whole fuel channel geometry. For the solutions of the benchmark problems, the temperature or the net radiation heat flux boundary conditions are prescribed for each radiating surface to determine the radiation heat transfer rate or the surface temperature, respectively by using the network method. The Discrete Transfer Model (DTM) is used for the CFX-10 radiation model and its calculation results are compared with the solutions of the benchmark problems. The CFX-10 results for the three benchmark problems are in close agreement with those solutions, so it is concluded that the CFX-10 with a DTM radiation model can be applied to the CANDU fuel channel analysis where a surface radiation heat transfer is a dominant mode of the heat transfer. (author)
Criticality and reactor physics benchmark experiments. Influence of nuclear data uncertainties
International Nuclear Information System (INIS)
A number of LWR-type criticality and reactor physics experiments, mainly from the ICSBEP and IRPhEP Handbooks, many of which were already used in the past for international benchmarks in the framework of OECD/NEA working groups, are being evaluated with respect to uncertainties in the basic nuclear data. For this, the sampling based uncertainty and sensitivity analysis tool XSUSA along with the Monte Carlo code KE-NO-V.a as transport solver is employed. Particular emphasis is put on experiments where differential quantities, mainly reaction rate distributions, were measured; the uncertainties of such quantities are not directly accessible to tools based on first order perturbation theory. With respect to multiplication factors and reactivity differences, all results are compared with corresponding results obtained with TSUNAMI-3D from the SCALE 6.1 system; the agreement is very good for all assemblies under consideration. With respect to fission rate distributions, the uncertainty analyses yield only moderate uncertainties from nuclear data; therefore, in general the total uncertainty is dominated by measurement uncertainties, which include the uncertainties of technological parameters. The work is continuously being extended; in the future, also non-LWR specific assemblies, mainly relevant for GEN-IV reactors, will be investigated. (author)
A thermo mechanical benchmark calculation of a hexagonal can in the BTI accident with INCA code
International Nuclear Information System (INIS)
The thermomechanical behaviour of an hexagonal can in a benchmark problem (simulating the conditions of a BTI accident in a fuel assembly) is examined by means of the INCA code and the results systematically compared with those of ADINA
International Nuclear Information System (INIS)
This paper provides a description of work on the joint analysis of the entire set of the experiments with the solutions of highly enriched uranium in light water from the ''International Handbook of Evaluated Criticality Safety Benchmark Experiments''. The purpose of the work was to analyze the experiments for interconsistency, discover and evaluate possible correlations between them, discover and eliminate systematic errors and disagreements, and get a consistent set of evaluated experiments for future use in validation of calculations of critical mass of solutions of highly enriched uranium of different concentrations in light water and evaluation of uncertainty of these calculations. The paper describes in details how the correlations between the experimental uncertainties were determined as well as how systematic errors were discovered. (author)
Validation of criticality calculation for systems with MOX powders
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In 2005-2006, a series of experiments referred to as BFS/MOX was conducted at the BFS-1 experimental facility in IPPE, Russia. The program was designed to provide a basis for validation of criticality calculations for MOX fuel manufacturing processes and particularly with low-moderated MOX fissile media. The extensive experimental program was performed on those configurations, including criticality and reactor-type parameters measurements. The experiments were evaluated, peer reviewed, and analyzed with various codes and cross-section data. Criticality validation study was performed employing sensitivity/uncertainty technique based on the generalized linear least square method. The paper illustrates different tools' performance when calculating criticality for the BFS/MOX configurations and focused upon the validation process and results for generic application systems with weapons-grade plutonium. (authors)
International Nuclear Information System (INIS)
For many design and ageing considerations fracture mechanics is needed to evaluate cracked components. The major parameters used are K and J. For that, the different codes (RSE-M appendix 5, RCC-MRx appendix A16, R6 rule, ASME B and PV Code Section XI, API, VERLIFE, Russian Code..) propose compendia of stress intensity factors, and for some of them compendia of limit loads for usual situations, in terms of component geometry, type of defect and loading conditions. The benchmark bench-KJ, proposed in the frame of the OECD/IAGE Group, aims to compare these different estimation schemes by comparison to reference analyses done by Finite Element Method, for representative cases (pipes and elbows, mechanical or/and thermal loadings, different type and size of cracks). The objective is to have a global comparison of the procedures but also of all independent elements as stress intensity factor or reference stress. The benchmark will cover simple cases with basic mechanical loads like pressure and bending up to complex load combinations and complex geometries (cylinders and elbows) including cladding or welds: these cases are classified into 6 tasks. Twenty-nine partners are involved in this benchmark. This paper gives a short overview of the different tasks of the benchmark and presents the analysis of the results for the four first tasks, devoted on the elastic stress intensity factor calculation (task 1) and J calculation in cracked pipes (tasks 2 and 3). (authors)
International Nuclear Information System (INIS)
In this article are compared theoretical results by new version of the SCALE5 code with experiments or other theoretical calculation for: 1. criticality: -measurement on ZR-6 and LR-0; - numerical benchmark No. 1.3 and 4 (CB1, CB3, CB4); 2. nuclide compositions: - measurement in Kurchatov institute for 3,6 %; - measurement in JAERI(PWR 17x17); - numerical benchmark No. 2-Source (CB2); 3. sources and decay heat:- numerical benchmark No.2-Source (CB2-S); The focus is on modules KENO, TRITON and ORIGEN-S (Authors)
Criticality safety calculations of the Soreq research reactor storage pool
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The IRR-l spent fuel is to be relocated in a storage pool. The present paper describes the actual facility and summarizes the Monte Carlo criticality safety calculations. The fuel elements are to be placed inside cadmium boxes to reduce their reactivity. The fuel element is 7.6 cm by 8.0 cm in the horizontal plane. The cadmium box is effectively 9.7 cm by 9.7 cm, providing significant water between the cadmium and the fuel element. The present calculations show that the spent fuel storage pool is criticality safe even for fresh fuel elements. (author)
Camps, Peter; Misselt, Karl; Bianchi, Simone; Lunttila, Tuomas; Pinte, Christophe; Natale, Giovanni; Juvela, Mika; Fischera, Joerg; Fitzgerald, Michael P.; Gordon, Karl; Baes, Maarten; Steinacker, Jürgen
2015-08-01
Context. Thermal emission by stochastically heated dust grains (SHGs) plays an important role in the radiative transfer (RT) problem for a dusty medium. It is therefore essential to verify that RT codes properly calculate the dust emission before studying the effects of spatial distribution and other model parameters on the simulated observables. Aims: We define an appropriate problem for benchmarking dust emissivity calculations in the context of RT simulations, specifically including the emission from SHGs. Our aim is to provide a self-contained guide for implementors of such functionality and to offer insight into the effects of the various approximations and heuristics implemented by the participating codes to accelerate the calculations. Methods: The benchmark problem definition includes the optical and calorimetric material properties and the grain size distributions for a typical astronomical dust mixture with silicate, graphite, and PAH components. It also includes a series of analytically defined radiation fields to which the dust population is to be exposed and instructions for the desired output. We processed this problem using six RT codes participating in this benchmark effort and compared the results to a reference solution computed with the publicly available dust emission code DustEM. Results: The participating codes implement different heuristics to keep the calculation time at an acceptable level. We study the effects of these mechanisms on the calculated solutions and report on the level of (dis)agreement between the participating codes. For all but the most extreme input fields, we find agreement within 10% across the important wavelength range 3 μm ≤ λ ≤ 1000 μm. Conclusions: We conclude that the relevant modules in RT codes can and do produce fairly consistent results for the emissivity spectra of SHGs. This work can serve as a reference for implementors of dust RT codes, and it will pave the way for a more extensive benchmark effort
The calculational VVER burnup Credit Benchmark No.3 results with the ENDF/B-VI rev.5 (1999)
Energy Technology Data Exchange (ETDEWEB)
Rodriguez Gual, Maritza [Centro de Tecnologia Nuclear, La Habana (Cuba). E-mail: mrgual@ctn.isctn.edu.cu
2000-07-01
The purpose of this papers to present the results of CB3 phase of the VVER calculational benchmark with the recent evaluated nuclear data library ENDF/B-VI Rev.5 (1999). This results are compared with the obtained from the other participants in the calculations (Czech Republic, Finland, Hungary, Slovaquia, Spain and the United Kingdom). The phase (CB3) of the VVER calculation benchmark is similar to the Phase II-A of the OECD/NEA/INSC BUC Working Group benchmark for PWR. The cases without burnup profile (BP) were performed with the WIMS/D-4 code. The rest of the cases have been carried with DOTIII discrete ordinates code. The neutron library used was the ENDF/B-VI rev. 5 (1999). The WIMS/D-4 (69 groups) is used to collapse cross sections from the ENDF/B-VI Rev. 5 (1999) to 36 groups working library for 2-D calculations. This work also comprises the results of CB1 (obtained with ENDF/B-VI rev. 5 (1999), too) and CB3 for cases with Burnup of 30 MWd/TU and cooling time of 1 and 5 years and for case with Burnup of 40 MWd/TU and cooling time of 1 year. (author)
Iterative acceleration methods for Monte Carlo and deterministic criticality calculations
International Nuclear Information System (INIS)
If you have ever given up on a nuclear criticality calculation and terminated it because it took so long to converge, you might find this thesis of interest. The author develops three methods for improving the fission source convergence in nuclear criticality calculations for physical systems with high dominance ratios for which convergence is slow. The Fission Matrix Acceleration Method and the Fission Diffusion Synthetic Acceleration (FDSA) Method are acceleration methods that speed fission source convergence for both Monte Carlo and deterministic methods. The third method is a hybrid Monte Carlo method that also converges for difficult problems where the unaccelerated Monte Carlo method fails. The author tested the feasibility of all three methods in a test bed consisting of idealized problems. He has successfully accelerated fission source convergence in both deterministic and Monte Carlo criticality calculations. By filtering statistical noise, he has incorporated deterministic attributes into the Monte Carlo calculations in order to speed their source convergence. He has used both the fission matrix and a diffusion approximation to perform unbiased accelerations. The Fission Matrix Acceleration method has been implemented in the production code MCNP and successfully applied to a real problem. When the unaccelerated calculations are unable to converge to the correct solution, they cannot be accelerated in an unbiased fashion. A Hybrid Monte Carlo method weds Monte Carlo and a modified diffusion calculation to overcome these deficiencies. The Hybrid method additionally possesses reduced statistical errors
Iterative acceleration methods for Monte Carlo and deterministic criticality calculations
Energy Technology Data Exchange (ETDEWEB)
Urbatsch, T.J.
1995-11-01
If you have ever given up on a nuclear criticality calculation and terminated it because it took so long to converge, you might find this thesis of interest. The author develops three methods for improving the fission source convergence in nuclear criticality calculations for physical systems with high dominance ratios for which convergence is slow. The Fission Matrix Acceleration Method and the Fission Diffusion Synthetic Acceleration (FDSA) Method are acceleration methods that speed fission source convergence for both Monte Carlo and deterministic methods. The third method is a hybrid Monte Carlo method that also converges for difficult problems where the unaccelerated Monte Carlo method fails. The author tested the feasibility of all three methods in a test bed consisting of idealized problems. He has successfully accelerated fission source convergence in both deterministic and Monte Carlo criticality calculations. By filtering statistical noise, he has incorporated deterministic attributes into the Monte Carlo calculations in order to speed their source convergence. He has used both the fission matrix and a diffusion approximation to perform unbiased accelerations. The Fission Matrix Acceleration method has been implemented in the production code MCNP and successfully applied to a real problem. When the unaccelerated calculations are unable to converge to the correct solution, they cannot be accelerated in an unbiased fashion. A Hybrid Monte Carlo method weds Monte Carlo and a modified diffusion calculation to overcome these deficiencies. The Hybrid method additionally possesses reduced statistical errors.
Benchmark Calculations of Interaction Energies in Noncovalent Complexes and Their Applications.
Řezáč, Jan; Hobza, Pavel
2016-05-11
Data sets of benchmark interaction energies in noncovalent complexes are an important tool for quantifying the accuracy of computational methods used in this field, as well as for the development of new computational approaches. This review is intended as a guide to conscious use of these data sets. We discuss their construction and accuracy, list the data sets available in the literature, and demonstrate their application to validation and parametrization of quantum-mechanical computational methods. In practical model systems, the benchmark interaction energies are usually obtained using composite CCSD(T)/CBS schemes. To use these results as a benchmark, their accuracy should be estimated first. We analyze the errors of this methodology with respect to both the approximations involved and the basis set size. We list the most prominent data sets covering various aspects of the field, from general ones to sets focusing on specific types of interactions or systems. The benchmark data are then used to validate more efficient computational approaches, including those based on explicitly correlated methods. Special attention is paid to the transition to large systems, where accurate benchmarking is difficult or impossible, and to the importance of nonequilibrium geometries in parametrization of more approximate methods. PMID:26943241
Acceleration and increased control of convergence in criticality calculations
International Nuclear Information System (INIS)
IRSN is developing a numerical simulation code called Moret to assess the nuclear criticality risk. This tool is designed to perform 3D simulations of neutron transport in a given system. It achieves this by adopting a probabilistic approach known as Monte Carlo, in which the transport of several successive generations of neutrons is calculated from an initial neutron distribution in the system under study. These generations are simulated until it is considered that convergence of the effective neutron multiplication coefficient (or Keff) - which characterizes the gap before reaching the critical state - has been reached. Insufficient convergence can lead to underestimation of both Keff and the criticality risk. During this thesis work, A. Jinaphanh sought to improve the reliability of values by developing a new method for initializing calculations, together with a criterion used to reliably determine whether or not convergence has been reached. (author)
TRIGA criticality experiment for testing burn-up calculations
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Persic, Andreja; Ravnik, Matjaz; Zagar, Tomaz [Jozef Stefan Institute, Reactor Physics Division, Ljubljana (Slovenia)
1999-07-01
A criticality experiment with partly burned TRIGA fuel is described. 20 wt % enriched standard TRIGA fuel elements initially containing 12 wt % U are used. Their average burn-up is 1.4 MWd. Fuel element burn-up is calculated in 2-D four group diffusion approximation using TRIGLAV code. The burn-up of several fuel elements is also measured by reactivity method. The excess reactivity of several critical and subcritical core configurations is measured. Two core configurations contain the same fuel elements in the same arrangement as were used in the fresh TRIGA fuel criticality experiment performed in 1991. The results of the experiment may be applied for testing the computer codes used for fuel burn-up calculations. (author)
International Nuclear Information System (INIS)
The quantum Monte Carlo (QMC) technique is used to generate accurate energy benchmarks for methane-water clusters containing a single methane monomer and up to 20 water monomers. The benchmarks for each type of cluster are computed for a set of geometries drawn from molecular dynamics simulations. The accuracy of QMC is expected to be comparable with that of coupled-cluster calculations, and this is confirmed by comparisons for the CH4-H2O dimer. The benchmarks are used to assess the accuracy of the second-order Møller-Plesset (MP2) approximation close to the complete basis-set limit. A recently developed embedded many-body technique is shown to give an efficient procedure for computing basis-set converged MP2 energies for the large clusters. It is found that MP2 values for the methane binding energies and the cohesive energies of the water clusters without methane are in close agreement with the QMC benchmarks, but the agreement is aided by partial cancelation between 2-body and beyond-2-body errors of MP2. The embedding approach allows MP2 to be applied without loss of accuracy to the methane hydrate crystal, and it is shown that the resulting methane binding energy and the cohesive energy of the water lattice agree almost exactly with recently reported QMC values
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Gillan, M. J., E-mail: m.gillan@ucl.ac.uk [London Centre for Nanotechnology, University College London, Gordon St., London WC1H 0AH (United Kingdom); Department of Physics and Astronomy, University College London, Gower St., London WC1E 6BT (United Kingdom); Thomas Young Centre, University College London, Gordon St., London WC1H 0AH (United Kingdom); Alfè, D. [London Centre for Nanotechnology, University College London, Gordon St., London WC1H 0AH (United Kingdom); Department of Physics and Astronomy, University College London, Gower St., London WC1E 6BT (United Kingdom); Thomas Young Centre, University College London, Gordon St., London WC1H 0AH (United Kingdom); Department of Earth Sciences, University College London, Gower St., London WC1E 6BT (United Kingdom); Manby, F. R. [Centre for Computational Chemistry, School of Chemistry, University of Bristol, Bristol BS8 1TS (United Kingdom)
2015-09-14
The quantum Monte Carlo (QMC) technique is used to generate accurate energy benchmarks for methane-water clusters containing a single methane monomer and up to 20 water monomers. The benchmarks for each type of cluster are computed for a set of geometries drawn from molecular dynamics simulations. The accuracy of QMC is expected to be comparable with that of coupled-cluster calculations, and this is confirmed by comparisons for the CH{sub 4}-H{sub 2}O dimer. The benchmarks are used to assess the accuracy of the second-order Møller-Plesset (MP2) approximation close to the complete basis-set limit. A recently developed embedded many-body technique is shown to give an efficient procedure for computing basis-set converged MP2 energies for the large clusters. It is found that MP2 values for the methane binding energies and the cohesive energies of the water clusters without methane are in close agreement with the QMC benchmarks, but the agreement is aided by partial cancelation between 2-body and beyond-2-body errors of MP2. The embedding approach allows MP2 to be applied without loss of accuracy to the methane hydrate crystal, and it is shown that the resulting methane binding energy and the cohesive energy of the water lattice agree almost exactly with recently reported QMC values.
Gillan, M J; Alfè, D; Manby, F R
2015-09-14
The quantum Monte Carlo (QMC) technique is used to generate accurate energy benchmarks for methane-water clusters containing a single methane monomer and up to 20 water monomers. The benchmarks for each type of cluster are computed for a set of geometries drawn from molecular dynamics simulations. The accuracy of QMC is expected to be comparable with that of coupled-cluster calculations, and this is confirmed by comparisons for the CH4-H2O dimer. The benchmarks are used to assess the accuracy of the second-order Møller-Plesset (MP2) approximation close to the complete basis-set limit. A recently developed embedded many-body technique is shown to give an efficient procedure for computing basis-set converged MP2 energies for the large clusters. It is found that MP2 values for the methane binding energies and the cohesive energies of the water clusters without methane are in close agreement with the QMC benchmarks, but the agreement is aided by partial cancelation between 2-body and beyond-2-body errors of MP2. The embedding approach allows MP2 to be applied without loss of accuracy to the methane hydrate crystal, and it is shown that the resulting methane binding energy and the cohesive energy of the water lattice agree almost exactly with recently reported QMC values. PMID:26374005
TRIGA FUEL PHASE I AND II CRITICALITY CALCULATION
Energy Technology Data Exchange (ETDEWEB)
L. Angers
1999-11-23
The purpose of this calculation is to characterize the criticality aspect of the codisposal of TRIGA (Training, Research, Isotopes, General Atomic) reactor spent nuclear fuel (SNF) with Savannah River Site (SRS) high-level waste (HLW). The TRIGA SNF is loaded into a Department of Energy (DOE) standardized SNF canister which is centrally positioned inside a five-canister defense SRS HLW waste package (WP). The objective of the calculation is to investigate the criticality issues for the WP containing the five SRS HLW and DOE SNF canisters in various stages of degradation. This calculation will support the analysis that will be performed to demonstrate the viability of the codisposal concept for the Monitored Geologic Repository (MGR).
Criticality calculations with MCNP{sup TM}: A primer
Energy Technology Data Exchange (ETDEWEB)
Mendius, P.W. [ed.; Harmon, C.D. II; Busch, R.D.; Briesmeister, J.F.; Forster, R.A.
1994-08-01
The purpose of this Primer is to assist the nuclear criticality safety analyst to perform computer calculations using the Monte Carlo code MCNP. Because of the closure of many experimental facilities, reliance on computer simulation is increasing. Often the analyst has little experience with specific codes available at his/her facility. This Primer helps the analyst understand and use the MCNP Monte Carlo code for nuclear criticality analyses. It assumes no knowledge of or particular experience with Monte Carlo codes in general or with MCNP in particular. The document begins with a Quickstart chapter that introduces the basic concepts of using MCNP. The following chapters expand on those ideas, presenting a range of problems from simple cylinders to 3-dimensional lattices for calculating keff confidence intervals. Input files and results for all problems are included. The Primer can be used alone, but its best use is in conjunction with the MCNP4A manual. After completing the Primer, a criticality analyst should be capable of performing and understanding a majority of the calculations that will arise in the field of nuclear criticality safety.
Calculation of Critical Values for Somerville's FDR Procedures
Directory of Open Access Journals (Sweden)
Paul N. Somerville
2007-04-01
Full Text Available A Fortran 95 program has been written to calculate critical values for the step-up and step-down FDR procedures developed by Somerville (2004. The program allows for arbitrary selection of number of hypotheses, FDR rate, one- or two-sided hypotheses, common correlation coefficient of the test statistics and degrees of freedom. An MCV (minimum critical value may be specified, or the program will calculate a specified number of critical values or steps in an FDR procedure. The program can also be used to efficiently ascertain an upper bound to the number of hypotheses which the procedure will reject, given either the values of the test statistics, or their p values. Limiting the number of steps in an FDR procedure can be used to control the number or proportion of false discoveries (Somerville and Hemmelmann 2007. Using the program to calculate the largest critical values makes possible efficient use of the FDR procedures for very large numbers of hypotheses
International Nuclear Information System (INIS)
The neutron generation time Λ plays an important role in the reactor kinetics. However, it is not straightforward nor standard in most continuous energy Monte Carlo codes which are able to calculate the prompt neutron lifetime lp directly. The difference between Λ and lp are sometimes very apparent. As very few delayed neutrons are produced in the reactor, they have little influence on Λ. Thus on the assumption that no delayed neutrons are produced in the system, the prompt kinetics equations for critical system and subcritical system with an external source are proposed. And then the equations are applied to calculating Λ with pulsed neutron technique using Monte Carlo. Only one fission neutron source is simulated with Monte Carlo in critical system while two neutron sources, including a fission source and an external source, are simulated for subcritical system. Calculations are performed on both critical benchmarks and subcritical system with an external source and the results are consistent with the reference values. (author)
Visualization and analyses of MCNP criticality calculation results
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Careful assessment of the results of a calculation by the code itself can detect mistakes in the problem setup and execution. MCNP has over four hundred error messages that inform the user of FATAL or WARNING errors that have been discovered during processing of just the input file. MCNP4A performs a self assessment of the calculated results to aid the user in determining the quality of the Monte Carlo results. MCNP4A contains new built-in sensitivity analyses of the Monte Carlo calculation that provide the user with simple WARNING messages for both criticality and fixed source calculations. The goal of the new analyses described in this paper is to provide the MCNP criticality practitioner with enough information in the output to assess the validity of the keff calculation and any associated tallies. The results of these checks are presented in the keff results summary, several keff tables and graphs, and tally tables and graphs. Plots of keff at the workstation are also available as the problem is running or in a postprocessing mode to assess problem performance and results. Plots of the fission source by cycle supply valuable visual information, although they are not yet available in the production version of MCNP
EXTERNAL CRITICALITY CALCULATION FOR DOE SNF CODISPOSAL WASTE PACKAGES
International Nuclear Information System (INIS)
The purpose of this document is to evaluate the potential for criticality for the fissile material that could accumulate in the near-field (invert) and in the far-field (host rock) beneath the U.S. Department of Energy (DOE) spent nuclear fuel (SNF) codisposal waste packages (WPs) as they degrade in the proposed monitored geologic repository at Yucca Mountain. The scope of this calculation is limited to the following DOE SNF types: Shippingport Pressurized Water Reactor (PWR), Enrico Fermi, Fast Flux Test Facility (FFTF), Fort St. Vrain, Melt and Dilute, Shippingport Light Water Breeder Reactor (LWBR), N-Reactor, and Training, Research, Isotope, General Atomics reactor (TRIGA). The results of this calculation are intended to be used for estimating the probability of criticality in the near-field and in the far-field. There are no limitations on use of the results of this calculation. The calculation is associated with the waste package design and was developed in accordance with the technical work plan, ''Technical Work Plan for: Department of Energy Spent Nuclear Fuel and Plutonium Disposition Work Packages'' (Bechtel SAIC Company, LLC [BSC], 2002a). This calculation is subject to the Quality Assurance Requirements and Description (QARD) per the activity evaluation under work package number P6212310Ml in the technical work plan TWP-MGR-MD-0000 101 (BSC 2002a)
Energy Technology Data Exchange (ETDEWEB)
Joo, Hyung Kook; Noh, Jae Man; Lee, Hyung Chul; Yoo, Jae Woon
2006-01-15
In this report, we verified the NUREC code transient calculation capability using OECD NEA/US NRC PWR MOX/UO2 Core Transient Benchmark Problem. The benchmark problem consists of Part 1, a 2-D problem with given T/H conditions, Part 2, a 3-D problem at HFP condition, Part 3, a 3-D problem at HZP condition, and Part 4, a transient state initiated by a control rod ejection at HZP condition in Part 3. In Part 1, the results of NUREC code agreed well with the reference solution obtained from DeCART calculation except for the pin power distributions at the rodded assemblies. In Part 2, the results of NUREC code agreed well with the reference DeCART solutions. In Part 3, some results of NUREC code such as critical boron concentration and core averaged delayed neutron fraction agreed well with the reference PARCS 2G solutions. But the error of the assembly power at the core center was quite large. The pin power errors of NUREC code at the rodded assemblies was much smaller the those of PARCS code. The axial power distribution also agreed well with the reference solution. In Part 4, the results of NUREC code agreed well with those of PARCS 2G code which was taken as the reference solution. From the above results we can conclude that the results of NUREC code for steady states and transient states of the MOX loaded LWR core agree well with those of the other codes.
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In this report, we verified the NUREC code transient calculation capability using OECD NEA/US NRC PWR MOX/UO2 Core Transient Benchmark Problem. The benchmark problem consists of Part 1, a 2-D problem with given T/H conditions, Part 2, a 3-D problem at HFP condition, Part 3, a 3-D problem at HZP condition, and Part 4, a transient state initiated by a control rod ejection at HZP condition in Part 3. In Part 1, the results of NUREC code agreed well with the reference solution obtained from DeCART calculation except for the pin power distributions at the rodded assemblies. In Part 2, the results of NUREC code agreed well with the reference DeCART solutions. In Part 3, some results of NUREC code such as critical boron concentration and core averaged delayed neutron fraction agreed well with the reference PARCS 2G solutions. But the error of the assembly power at the core center was quite large. The pin power errors of NUREC code at the rodded assemblies was much smaller the those of PARCS code. The axial power distribution also agreed well with the reference solution. In Part 4, the results of NUREC code agreed well with those of PARCS 2G code which was taken as the reference solution. From the above results we can conclude that the results of NUREC code for steady states and transient states of the MOX loaded LWR core agree well with those of the other codes
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DeCART is a 3-dimensional whole-core code based on the synthesis of 2-D radial MOC (Method Of Characteristics) transport and 1-D axial nodal diffusion methods. This code has been applied to the PWR physics analysis. Recently its' geometry treatment capability has been extended to deal with the hexagonal meshes for the VHTR (Very High Temperature gas-cooled Reactor) physics analysis, which requires a verification of the applicability to the VHTR fuels. The Argonne national laboratory has developed the numerical benchmark problems based on the Compact Nuclear Power Source (CNPS) experiments conducted at the Los Alamos National Laboratory (LANL) in the late 1980s in order to support the validation and verification work for the VHTR physics codes. Development of the numerical benchmarks was required from a lack of experimental information on the design data uncertainties and the inconsistency in the design data from different sources. Two- and three-dimensional numerical benchmarks based on the CNPS experiment are specified for the verification of the VHTR physics. In this study the DeCART code was assessed by performing the CNPS benchmark calculations and comparing the results with the MCNP ones
Energy Technology Data Exchange (ETDEWEB)
Lee, Kyung Hoon; Cho, Jin Young; Kim, Kang Seog; Joo, Hyung Kook; Lee, Chung Chan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
2006-07-01
DeCART is a 3-dimensional whole-core code based on the synthesis of 2-D radial MOC (Method Of Characteristics) transport and 1-D axial nodal diffusion methods. This code has been applied to the PWR physics analysis. Recently its' geometry treatment capability has been extended to deal with the hexagonal meshes for the VHTR (Very High Temperature gas-cooled Reactor) physics analysis, which requires a verification of the applicability to the VHTR fuels. The Argonne national laboratory has developed the numerical benchmark problems based on the Compact Nuclear Power Source (CNPS) experiments conducted at the Los Alamos National Laboratory (LANL) in the late 1980s in order to support the validation and verification work for the VHTR physics codes. Development of the numerical benchmarks was required from a lack of experimental information on the design data uncertainties and the inconsistency in the design data from different sources. Two- and three-dimensional numerical benchmarks based on the CNPS experiment are specified for the verification of the VHTR physics. In this study the DeCART code was assessed by performing the CNPS benchmark calculations and comparing the results with the MCNP ones.
Recent R and D around the Monte-Carlo code Tripoli-4 for criticality calculation
Energy Technology Data Exchange (ETDEWEB)
Hugot, F.X.; Lee, Y.K.; Malvagi, F. [CEA - DEN/DANS/DM2S/SERMA/LTSD, Saclay (France)
2008-07-01
TRIPOLI-4 [1] is the fourth generation of the TRIPOLI family of Monte Carlo codes developed from the 60's by CEA. It simulates the 3D transport of neutrons, photons, electrons and positrons as well as coupled neutron-photon propagation and electron-photons cascade showers. The code addresses radiation protection and shielding problems, as well as criticality and reactor physics problems through both critical and subcritical neutronics calculations. It uses full pointwise as well as multigroup cross-sections. The code has been validated through several hundred benchmarks as well as measurement campaigns. It is used as a reference tool by CEA as well as its industrial and institutional partners, and in the NURESIM [2] European project. Section 2 reviews its main features, with emphasis on the latest developments. Section 3 presents some recent R and D for criticality calculations. Fission matrix, Eigen-values and eigenvectors computations will be exposed. Corrections on the standard deviation estimator in the case of correlations between generation steps will be detailed. Section 4 presents some preliminary results obtained by the new mesh tally feature. The last section presents the interest of using XML format output files. (authors)
Teale, Andrew M.; Lutnæs, Ola B.; Helgaker, Trygve; Tozer, David J.; Gauss, Jürgen
2013-01-01
Accurate sets of benchmark nuclear-magnetic-resonance shielding constants and spin-rotation constants are calculated using coupled-cluster singles-doubles (CCSD) theory and coupled-cluster singles-doubles-perturbative-triples [CCSD(T)] theory, in a variety of basis sets consisting of (rotational) London atomic orbitals. The accuracy of the calculated coupled-cluster constants is established by a careful comparison with experimental data, taking into account zero-point vibrational corrections. Coupled-cluster basis-set convergence is analyzed and extrapolation techniques are employed to estimate basis-set-limit quantities, thereby establishing an accurate benchmark data set. Together with the set provided for rotational g-tensors and magnetizabilities in our previous work [O. B. Lutnæs, A. M. Teale, T. Helgaker, D. J. Tozer, K. Ruud, and J. Gauss, J. Chem. Phys. 131, 144104 (2009)], 10.1063/1.3242081, it provides a substantial source of consistently calculated high-accuracy data on second-order magnetic response properties. The utility of this benchmark data set is demonstrated by examining a wide variety of Kohn-Sham exchange-correlation functionals for the calculation of these properties. None of the existing approximate functionals provide an accuracy competitive with that provided by CCSD or CCSD(T) theory. The need for a careful consideration of vibrational effects is clearly illustrated. Finally, the pure coupled-cluster results are compared with the results of Kohn-Sham calculations constrained to give the same electronic density. Routes to future improvements are discussed in light of this comparison.
Criticality calculations for a critical assembly, graphite moderate, using 20% enriched uranium
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The construction of a Zero Power Reactor (ZPR) at the Instituto de Energia Atomica in order to measure the neutron characteristics (parameters) of HTGR reactors is proposed. The necessary quantity fissile uranium for these measurements has been calculed. Criticality studies of graphite moderated critical assemblies containing thorium have been made and the critical mass of each of several typical commercial HTGR compositions has been calculated using computer codes HAMMER and CITATION. Assemblies investigated contained a central cylindrical core region, simulating a typical commercial HTGR composition, a uranium-graphite driver region and a outer pure graphite reflector region. It is concluded that a 10Kg inventory of fissile uranium will be required for a program of measurements utilizing each of the several calculated assemblies
Second Order Perturbations of Monte Carlo Criticality Calculations
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Perturbation techniques are powerful tools for determining the effects of small changes, or perturbations, to a problem. Perturbations have long been problematic in Monte Carlo calculations because the effects of small changes to the problem are usually masked by the inherent statistical uncertainties. The recently released MCNP4B Monte Carlo computer code uses the differential operator technique, to calculate changes in tallies caused by perturbations in density and composition over given energy ranges and reaction types. This technique will allow for precise calculation of the changes in tallies even if the standard deviation of the unperturbed tally is larger than the change. The differential operator is approximated by a second order Taylor series. The implementation of the Taylor series expansion assumes that the coefficients are independent of any perturbed cross-sections. However, if the tally is multiplied by cross-section data this assumption is invalid and incorrect results will be generated. Of significant interest is the use of perturbations in criticality calculations. Although the criticality source feature for MCNP cannot directly calculate perturbed eigenvalues, a track-length estimate for Keff can be tallied and the perturbation feature can be applied to this tally. However, since the tally multiplies the flux by the macroscopic fission cross-section, this tally is dependent on perturbed cross-section data and incorrect results will be calculated by the perturbation feature. In order to compute the correct tally, a correction term is needed that will account for the dependence of the Taylor series coefficients on the perturbed cross-section data
Neutron batch size optimisation methodology for Monte Carlo criticality calculations
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Highlights: • A method is suggested for improving efficiency of MC criticality calculations. • The method optimises the number of neutrons simulated per cycle. • The optimal number of neutrons per cycle depends on allocated computing time. - Abstract: We present a methodology that improves the efficiency of conventional power iteration based Monte Carlo criticality calculations by optimising the number of neutron histories simulated per criticality cycle (the so-called neutron batch size). The chosen neutron batch size affects both the rate of convergence (in computing time) and magnitude of bias in the fission source. Setting a small neutron batch size ensures a rapid simulation of criticality cycles, allowing the fission source to converge fast to its stationary state; however, at the same time, the small neutron batch size introduces a large systematic bias in the fission source. It follows that for a given allocated computing time, there is an optimal neutron batch size that balances these two effects. We approach this problem by studying the error in the cumulative fission source, i.e. the fission source combined over all simulated cycles, as all results are commonly combined over the simulated cycles. We have deduced a simplified formula for the error in the cumulative fission source, taking into account the neutron batch size, the dominance ratio of the system, the error in the initial fission source and the allocated computing time (in the form of the total number of simulated neutron histories). Knowing how the neutron batch size affects the error in the cumulative fission source allows us to find its optimal value. We demonstrate the benefits of the method on a number of numerical test calculations
Criticality calculations with MCNP{trademark}: A primer
Energy Technology Data Exchange (ETDEWEB)
Harmon, C.D. II; Busch, R.D.; Briesmeister, J.F.; Forster, R.A. [New Mexico Univ., Albuquerque, NM (United States)
1994-06-06
With the closure of many experimental facilities, the nuclear criticality safety analyst increasingly is required to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. However, in many cases, the analyst has little experience with the specific codes available at his/her facility. This primer will help you, the analyst, understand and use the MCNP Monte Carlo code for nuclear criticality safety analyses. It assumes that you have a college education in a technical field. There is no assumption of familiarity with Monte Carlo codes in general or with MCNP in particular. Appendix A gives an introduction to Monte Carlo techniques. The primer is designed to teach by example, with each example illustrating two or three features of MCNP that are useful in criticality analyses. Beginning with a Quickstart chapter, the primer gives an overview of the basic requirements for MCNP input and allows you to run a simple criticality problem with MCNP. This chapter is not designed to explain either the input or the MCNP options in detail; but rather it introduces basic concepts that are further explained in following chapters. Each chapter begins with a list of basic objectives that identify the goal of the chapter, and a list of the individual MCNP features that are covered in detail in the unique chapter example problems. It is expected that on completion of the primer you will be comfortable using MCNP in criticality calculations and will be capable of handling 80 to 90 percent of the situations that normally arise in a facility. The primer provides a set of basic input files that you can selectively modify to fit the particular problem at hand.
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A general purpose user's version of the EGS4 code system has been developed to make EGS4 easily applicable to the safety analysis of nuclear fuel cycle facilities. One such application involves the determination of skyshine dose for a variety of photon sources. To verify the accuracy of the code, it was benchmarked with Kansas State University (KSU) photon skyshine experiment of 1977. The results of the simulation showed that this version of EGS4 would be appicable to the skyshine calculation. (author)
Feldman, U.; Landi, E.; Doschek, G. A.
2006-10-01
The accuracy of available spectral codes is dependent on the quality of the atomic data and transition rates that they include, and can only be tested by benchmarking predicted line emissivities with observations from plasmas whose physical properties are known with precision. In the present work we describe a few high-resolution spectra emitted by solar flare plasmas under condition of ionization equilibrium, and one quiet Sun off-disk region spectrum, and we propose these datasets as benchmarks for the assessment of the accuracy of existing spectral codes in the 1.84-1.90 Å and 3.17-3.22 Å X-ray ranges and in the 500-1600 Å far ultraviolet range.
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Highlights: • Benchmark study performed for the neutronic calculations of TRIGA research reactors. • WIMSD-5B/CITATION is the utilized code system along with the WIMSD-IAEA-69 library. • The studied condensed spectra are five and seven energy groups spectra. • Analyzed: lattice parameters, reactivities, CR worth, flux and power distribution. • The lattice and neutronic parameters showed the accuracy of both condensed spectra. - Abstract: The objective of this paper is to assess the suitability and accuracy of the deterministic diffusion method for the neutronic calculations of the TRIGA Mark-III research reactors using the WIMSD/CITATION code system in proposed condensed energy spectra of five and seven energy groups with one and three thermal groups respectively. The utilized cell transport calculations code and core diffusion calculations code are the WIMSD-5B and the CITVAP v3.1 codes respectively, along with the WIMSD-IAEA-69 nuclear data library. Firstly, the assessment goes through analyzing the integral parameters – keff, ρ238, δ235, δ238, and C* – of the TRX and BAPL benchmark lattices and comparison with experimental and previous reference results using other ENDLs at the full energy spectra which show good agreement with the references at both spectra. Secondly, evaluation of the 3D nuclear characteristics of three different cores of the TRR-1/M1 TRIGA Mark-III Thai research reactor at the condensed energy spectra. The results include the excess reactivities of the cores and the worth of selected control rods which were compared with reference Monte Carlo results and experimental values. The results show good agreement with the references at both energy spectra and the better accuracy are attainable in the five energy groups spectrum. The results also include neutron flux distributions which are evaluated for future comparisons with other calculational techniques even they are comparable to reactors and fuels of the same type. The
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More than 50 neutron benchmark calculations have recently been completed as part of an ongoing program to validate the MCNP Monte Carlo radiation transport code. The benchmark calculations reported here are part of an ongoing multiyear, multiperson effort to benchmark version 4 of the MCNP code. The MCNP is a Monte Carlo three-dimensional general-purpose, continuous-energy neutron, photon, and electron transport code. It is used around the world for many applications including aerospace, oil-well logging, physics experiments, criticality safety, reactor analysis, medical imaging, defense applications, accelerator design, radiation hardening, radiation shielding, health physics, fusion research, and education. The first phase of the benchmark project consisted of analytic and photon problems. The second phase consists of the ENDF/B-V neutron problems reported in this paper and in more detail in the comprehensive report. A cooperative program being carried out a General Electric, San Jose, consists of light water reactor benchmark problems. A subsequent phase focusing on electron problems is planned
MUPO, Critical 43 Group Spectra Calculation for Homogeneous Reactor
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1 - Nature of physical problem solved: MUPO calculates the critical spectrum of a bare homogeneous reactor in 43 groups. This spectrum is used to evaluate condensed microscopic cross-sections. An option for this programme is to read in the library data from cards and write the binary library tape -DRAGON LIBRARY 3-. 2 - Method of solution: 3 options. Introduction of an additional absorber to account for a control poison, source iteration technique, and a buckling iteration. 3 - Restrictions on the complexity of the problem: 110 materials
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Shutdown dose rate (SDDR) inside and around the diagnostics ports of ITER is performed at PPPL/UCLA using the 3-D, FEM, Discrete Ordinates code, ATTILA, along with its updated FORNAX transmutation/decay gamma library. Other ITER partners assess SDDR using codes based on the Monte Carlo (MC) approach (e.g. MCNP code) for transport calculation and the radioactivity inventory code FISPACT or other equivalent decay data libraries for dose rate assessment. To reveal the range of discrepancies in the results obtained by various analysts, an extensive experimental and calculation benchmarking effort has been undertaken to validate the capability of ATTILA for dose rate assessment. On the experimental validation front, the comparison was performed using the measured data from two SDDR experiments performed at the FNG facility, Italy. Comparison was made to the experimental data and to MC results obtained by other analysts. On the calculation validation front, the ATTILA's predictions were compared to other results at key locations inside a calculation benchmark whose configuration duplicates an upper diagnostics port plug (UPP) in ITER. Both serial and parallel version of ATTILA-7.1.0 are used in the PPPL/UCLA analysis performed with FENDL-2.1/FORNAX databases. In the FNG 1st experimental, it was shown that ATTILA's dose rates are largely over estimated (by ∼30–60%) with the ANSI/ANS-6.1.1 flux-to-dose factors whereas the ICRP-74 factors give better agreement (10–20%) with the experimental data and with the MC results at all cooling times. In the 2nd experiment, there is an under estimation in SDDR calculated by both MCNP and ATTILA based on ANSI/ANS-6.1.1 for cooling times up to ∼4 days after irradiation. Thereafter, an over estimation is observed (∼5–10% with MCNP and ∼10–15% with ATTILA). As for the calculation benchmark, the agreement is much better based on ICRP-74 1996 data. The divergence among all dose rate results at ∼11 days cooling time is no
TMI-2 criticality studies: lower-vessel rubble and analytical benchmarking
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A bounding strategy has been adopted for assuring subcriticality during all TMI-2 defueling operations. The strategy is based upon establishing a safe soluble boron level for the entire reactor core in an optimum reactivity configuration. This paper presents the determination of a fuel rubble model which yields a maximum infinite lattice multiplication factor and the subsequent application of cell-averaged constants in finite system analyses. Included in the analyses are the effects of fuel burnup determined from a simplified power history of the reactor. A discussion of the analytical methods employed and the determination of an analytical bias with benchmark crictical experiments completes the presentation. 17 tabs
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EDF R and D is presently developing a new, state-of-the-art calculation chain called ANDROMEDE including the APOLLO2/JEFF3-based CEA multigroup library/REL2005 scheme package for assembly computations and COCAGNE 3D code for core computations. The goal of this paper is to validate the calculation chain and its methodologies on a numerical benchmark of a small PWR which has been loaded with mixed fuel, KAIST 1A. The latter is challenging, being highly heterogeneous as it has assemblies with burnable poison, offers a rodded configuration and includes both UOX-MOX and core-reflector interfaces. Thus, we will test the capabilities of the models used in ANDROMEDE to compute such cores. The validation methodology employed is as follows: stochastic calculations are used to validate the ability of assembly schemes SHEM-MOC and REL2005 for the computation of 2D full cores. Afterwards, industrial two-group diffusion calculations were set up. Reactivity coefficients and pin-by-pin power distributions were compared with those obtained from REL2005. Finally, the last section gives the prospects of the use of multigroup SPn for industrial calculations. They raise several questions such as the energy meshes to be used as well the 2D reflector model to be applied. A reflector model is set up to test the SPn solver on full-core calculations with results compared to those of the REL2005 scheme. (author)
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Small-sample reactivity experiments are relevant to provide accurate information on the integral cross sections of materials. One of the specificities of these experiments is that the measured reactivity worth generally ranges between 1 and 10 pcm, which precludes the use of Monte Carlo for the analysis. As a consequence, several papers have been devoted to deterministic calculation routes, implying spatial and/or energetic discretization which could involve calculation bias. Within the Expert Group on Burn-Up Credit of the OECD/NEA, a benchmark was proposed to compare different calculation codes and methods for the analysis of these experiments. In four Sub-Phases with geometries ranging from a single cell to a full 3D core model, participants were asked to evaluate the reactivity worth due to the addition of small quantities of separated fission products and actinides into a UO2 fuel. Fourteen institutes using six different codes have participated in the Benchmark. For reactivity worth of more than a few tens of pcm, the Monte-Carlo approach based on the eigen-value difference method appears clearly as the reference method. However, in the case of reactivity worth as low as 1 pcm, it is concluded that the deterministic approach based on the exact perturbation formalism is more accurate and should be preferred. Promising results have also been reported using the newly available exact perturbation capability, developed in the Monte Carlo code TRIPOLI4, based on the calculation of a continuous energy adjoint flux in the reference situation, convoluted to the forward flux of the perturbed situation. (author)
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The Monte Carlo (MC)-discrete ordinates (SN) coupled method is an efficient approach to solve shielding calculations of nuclear device with complex geometries and deep penetration. The 3D MC-SN coupled method has been used for PWR shielding calculation for the first time. According to characteristics of NUREG/CR-6115 PWR model, the thermal shield is specified as the common surface to link the Monte Carlo complex geometrical model and the deep penetration SN model. 3D Monte Carlo code is employed to accurately simulate the structure from core to thermal shield. The neutron tracks crossing the thermal shield inner surface are recorded by MC code. The SN boundary source is generated by the interface program and used by the 3D SN code to treat the calculation from thermal shield to pressure vessel. The calculation results include the circular distributions of fast neutron flux at pressure vessel inner wall, pressure vessel T/4 and lower weld locations. The calculation results are performed with comparison to MCNP and DORT solutions of benchmark report and satisfactory agreements are obtained. The validity of the method and the correctness of the programs are proved. (authors)
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In May 2010, JENDL-4.0 was released from Japan Atomic Energy Agency as the updated Japanese Nuclear Data Library. It was processed by the nuclear data processing system LICEM and an arbitrary-temperature neutron cross section library MVPlib-nJ40 was produced for the neutron and photon transport calculation code MVP based on the continuous-energy Monte Carlo method. The library contains neutron cross sections for 406 nuclides on the free gas model, thermal scattering cross sections, and cross sections of pseudo fission products for burn-up calculations with MVP. Criticality benchmark calculations were carried out with MVP and MVPlib-nJ40 for about 1,000 cases of critical experiments stored in the hand book of International Criticality Safety Benchmark Evaluation Project (ICSBEP), which covers a wide variety of fuel materials, fuel forms, and neutron spectra. We report all comparison results (C/E values) of effective neutron multiplication factors between calculations and experiments to give a validation data for the prediction accuracy of JENDL-4.0 for criticalities. (author)
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Over the past several years, plant-life extension programs have been implemented at many U.S. plants. One method of pressure vessel (PV) fluence rate reduction being used in several of the older reactors involves partial replacement of the oxide fuel with metallic rods in those peripheral assemblies located at critical azimuths. This substitution extends axially over a region that depends on the individual plant design, but covers the most critical PV weld and plate locations, which may be subject to pressurized thermal shock. In order to analyze the resulting PV dosimetry using these partial-length shield assemblies (PLSA), a relatively simple but accurate method needs to be formulated and qualified that treats the axially asymmetric core leakage. Accordingly, an experiment was devised and performed at the VENUS critical facility in Mol, Belgium. The success of the proposed method bodes well for the accuracy of future analyses of on-line plants using PLSAs
Řezáč, Jan; Huang, Yuanhang; Hobza, Pavel; Beran, Gregory J O
2015-07-14
Many-body noncovalent interactions are increasingly important in large and/or condensed-phase systems, but the current understanding of how well various models predict these interactions is limited. Here, benchmark complete-basis set coupled cluster singles, doubles, and perturbative triples (CCSD(T)) calculations have been performed to generate a new test set for three-body intermolecular interactions. This "3B-69" benchmark set includes three-body interaction energies for 69 total trimer structures, consisting of three structures from each of 23 different molecular crystals. By including structures that exhibit a variety of intermolecular interactions and packing arrangements, this set provides a stringent test for the ability of electronic structure methods to describe the correct physics involved in the interactions. Both MP2.5 (the average of second- and third-order Møller-Plesset perturbation theory) and spin-component-scaled CCSD for noncovalent interactions (SCS-MI-CCSD) perform well. MP2 handles the polarization aspects reasonably well, but it omits three-body dispersion. In contrast, many widely used density functionals corrected with three-body D3 dispersion correction perform comparatively poorly. The primary difficulty stems from the treatment of exchange and polarization in the functionals rather than from the dispersion correction, though the three-body dispersion may also be moderately underestimated by the D3 correction. PMID:26575743
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Li, M [Wayne State Univeristy, Detroit, MI (United States); Chetty, I [Henry Ford Health System, Detroit, MI (United States); Zhong, H [Henry Ford Hospital System, Detroit, MI (United States)
2014-06-01
Purpose: Tumor control probability (TCP) calculated with accumulated radiation doses may help design appropriate treatment margins. Image registration errors, however, may compromise the calculated TCP. The purpose of this study is to develop benchmark CT images to quantify registration-induced errors in the accumulated doses and their corresponding TCP. Methods: 4DCT images were registered from end-inhale (EI) to end-exhale (EE) using a “demons” algorithm. The demons DVFs were corrected by an FEM model to get realistic deformation fields. The FEM DVFs were used to warp the EI images to create the FEM-simulated images. The two images combined with the FEM DVF formed a benchmark model. Maximum intensity projection (MIP) images, created from the EI and simulated images, were used to develop IMRT plans. Two plans with 3 and 5 mm margins were developed for each patient. With these plans, radiation doses were recalculated on the simulated images and warped back to the EI images using the FEM DVFs to get the accumulated doses. The Elastix software was used to register the FEM-simulated images to the EI images. TCPs calculated with the Elastix-accumulated doses were compared with those generated by the FEM to get the TCP error of the Elastix registrations. Results: For six lung patients, the mean Elastix registration error ranged from 0.93 to 1.98 mm. Their relative dose errors in PTV were between 0.28% and 6.8% for 3mm margin plans, and between 0.29% and 6.3% for 5mm-margin plans. As the PTV margin reduced from 5 to 3 mm, the mean TCP error of the Elastix-reconstructed doses increased from 2.0% to 2.9%, and the mean NTCP errors decreased from 1.2% to 1.1%. Conclusion: Patient-specific benchmark images can be used to evaluate the impact of registration errors on the computed TCPs, and may help select appropriate PTV margins for lung SBRT patients.
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In this paper, the results of the investigations on the nodalization effects for the ATHLET code are presented and discussed in details on the basis of experimental data for the VVER-1000 Coolant Transient Benchmark with different operating modes of four main coolant pumps. ATHLET calculations with different nodalization and their impact was analyzed. The work studied the influence of annular outlet nodalization on calculation of coolant temperature. By comparing the test data versus calculated by ATHLET we showed a good agreement between the experimental data and simulation results for analyzed parameters. Keywords: VVER-1000, coolant transient benchmark, ATHLET, nodalization
HEXTRAN-SMABRE calculation of the 6th AER Benchmark, main steam line break in a WWER-440 NPP
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The sixth AER benchmark is the second AER benchmark for couplings of the thermal hydraulic codes and three dimensional neutron kinetic core models. It concerns a double end break of one main steam line in a WWER-440 plant. The core is at the end of its first cycle in full power conditions. In VTT HEXTRAN2.9 is used for the core kinetics and dynamics and SMABRE4.8 as a thermal hydraulic model for the primary and secondary loop. The plant model for SMABRE consists mainly of two input models, Loviisa model and a standard WWER-440/213 plant model. The primary loop includes six separate loops, the pressure vessel is divided into six parallel channels in SMABRE and the whole core calculation is performed in the core with HEXTRAN. The horizontal steam generators are modelled with heat transfer tubes in five levels and vertically with two parts, riser and downcomer. With this kind of detailed modelling of steam generators there occurs strong flashing after break opening. As a sequence of the main steam line break at nominal power level, the reactor trip is followed quite soon. The liquid temperature continues to decrease in one core inlet sector which may lead to recriticality and neuron power increase. The situation is very sensitive to small changes in the steam generator and break flow modelling and therefore several sensitivity calculations have been done. Also two stucked control rods have been assumed. Due to boric acid concentration in the high pressure safety injection subcriticality is finally guaranteed in the transient (Authors)
Epifanovsky, Evgeny; Klein, Kerstin; Stopkowicz, Stella; Gauss, Jürgen; Krylov, Anna I
2015-08-14
We present a formalism and an implementation for calculating spin-orbit couplings (SOCs) within the EOM-CCSD (equation-of-motion coupled-cluster with single and double substitutions) approach. The following variants of EOM-CCSD are considered: EOM-CCSD for excitation energies (EOM-EE-CCSD), EOM-CCSD with spin-flip (EOM-SF-CCSD), EOM-CCSD for ionization potentials (EOM-IP-CCSD) and electron attachment (EOM-EA-CCSD). We employ a perturbative approach in which the SOCs are computed as matrix elements of the respective part of the Breit-Pauli Hamiltonian using zeroth-order non-relativistic wave functions. We follow the expectation-value approach rather than the response-theory formulation for property calculations. Both the full two-electron treatment and the mean-field approximation (a partial account of the two-electron contributions) have been implemented and benchmarked using several small molecules containing elements up to the fourth row of the periodic table. The benchmark results show the excellent performance of the perturbative treatment and the mean-field approximation. When used with an appropriate basis set, the errors with respect to experiment are below 5% for the considered examples. The findings regarding basis-set requirements are in agreement with previous studies. The impact of different correlation treatment in zeroth-order wave functions is analyzed. Overall, the EOM-IP-CCSD, EOM-EA-CCSD, EOM-EE-CCSD, and EOM-SF-CCSD wave functions yield SOCs that agree well with each other (and with the experimental values when available). Using an EOM-CCSD approach that provides a more balanced description of the target states yields more accurate results. PMID:26277122
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Epifanovsky, Evgeny [Department of Chemistry, University of Southern California, Los Angeles, California 90089-0482 (United States); Department of Chemistry, University of California, Berkeley, California 94720 (United States); Q-Chem Inc., 6601 Owens Drive, Suite 105, Pleasanton, California 94588 (United States); Klein, Kerstin; Gauss, Jürgen [Institut für Physikalische Chemie, Universität Mainz, D-55099 Mainz (Germany); Stopkowicz, Stella [Department of Chemistry, Centre for Theoretical and Computational Chemistry, University of Oslo, N-0315 Oslo (Norway); Krylov, Anna I. [Department of Chemistry, University of Southern California, Los Angeles, California 90089-0482 (United States)
2015-08-14
We present a formalism and an implementation for calculating spin-orbit couplings (SOCs) within the EOM-CCSD (equation-of-motion coupled-cluster with single and double substitutions) approach. The following variants of EOM-CCSD are considered: EOM-CCSD for excitation energies (EOM-EE-CCSD), EOM-CCSD with spin-flip (EOM-SF-CCSD), EOM-CCSD for ionization potentials (EOM-IP-CCSD) and electron attachment (EOM-EA-CCSD). We employ a perturbative approach in which the SOCs are computed as matrix elements of the respective part of the Breit-Pauli Hamiltonian using zeroth-order non-relativistic wave functions. We follow the expectation-value approach rather than the response-theory formulation for property calculations. Both the full two-electron treatment and the mean-field approximation (a partial account of the two-electron contributions) have been implemented and benchmarked using several small molecules containing elements up to the fourth row of the periodic table. The benchmark results show the excellent performance of the perturbative treatment and the mean-field approximation. When used with an appropriate basis set, the errors with respect to experiment are below 5% for the considered examples. The findings regarding basis-set requirements are in agreement with previous studies. The impact of different correlation treatment in zeroth-order wave functions is analyzed. Overall, the EOM-IP-CCSD, EOM-EA-CCSD, EOM-EE-CCSD, and EOM-SF-CCSD wave functions yield SOCs that agree well with each other (and with the experimental values when available). Using an EOM-CCSD approach that provides a more balanced description of the target states yields more accurate results.
Epifanovsky, Evgeny; Klein, Kerstin; Stopkowicz, Stella; Gauss, Jürgen; Krylov, Anna I.
2015-08-01
We present a formalism and an implementation for calculating spin-orbit couplings (SOCs) within the EOM-CCSD (equation-of-motion coupled-cluster with single and double substitutions) approach. The following variants of EOM-CCSD are considered: EOM-CCSD for excitation energies (EOM-EE-CCSD), EOM-CCSD with spin-flip (EOM-SF-CCSD), EOM-CCSD for ionization potentials (EOM-IP-CCSD) and electron attachment (EOM-EA-CCSD). We employ a perturbative approach in which the SOCs are computed as matrix elements of the respective part of the Breit-Pauli Hamiltonian using zeroth-order non-relativistic wave functions. We follow the expectation-value approach rather than the response-theory formulation for property calculations. Both the full two-electron treatment and the mean-field approximation (a partial account of the two-electron contributions) have been implemented and benchmarked using several small molecules containing elements up to the fourth row of the periodic table. The benchmark results show the excellent performance of the perturbative treatment and the mean-field approximation. When used with an appropriate basis set, the errors with respect to experiment are below 5% for the considered examples. The findings regarding basis-set requirements are in agreement with previous studies. The impact of different correlation treatment in zeroth-order wave functions is analyzed. Overall, the EOM-IP-CCSD, EOM-EA-CCSD, EOM-EE-CCSD, and EOM-SF-CCSD wave functions yield SOCs that agree well with each other (and with the experimental values when available). Using an EOM-CCSD approach that provides a more balanced description of the target states yields more accurate results.
International Nuclear Information System (INIS)
We present a formalism and an implementation for calculating spin-orbit couplings (SOCs) within the EOM-CCSD (equation-of-motion coupled-cluster with single and double substitutions) approach. The following variants of EOM-CCSD are considered: EOM-CCSD for excitation energies (EOM-EE-CCSD), EOM-CCSD with spin-flip (EOM-SF-CCSD), EOM-CCSD for ionization potentials (EOM-IP-CCSD) and electron attachment (EOM-EA-CCSD). We employ a perturbative approach in which the SOCs are computed as matrix elements of the respective part of the Breit-Pauli Hamiltonian using zeroth-order non-relativistic wave functions. We follow the expectation-value approach rather than the response-theory formulation for property calculations. Both the full two-electron treatment and the mean-field approximation (a partial account of the two-electron contributions) have been implemented and benchmarked using several small molecules containing elements up to the fourth row of the periodic table. The benchmark results show the excellent performance of the perturbative treatment and the mean-field approximation. When used with an appropriate basis set, the errors with respect to experiment are below 5% for the considered examples. The findings regarding basis-set requirements are in agreement with previous studies. The impact of different correlation treatment in zeroth-order wave functions is analyzed. Overall, the EOM-IP-CCSD, EOM-EA-CCSD, EOM-EE-CCSD, and EOM-SF-CCSD wave functions yield SOCs that agree well with each other (and with the experimental values when available). Using an EOM-CCSD approach that provides a more balanced description of the target states yields more accurate results
Some benchmark calculations for VVER-1000 assemblies by WIMS-7B code
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Our aim in this report is to compare of calculation results, obtained with the use of different libraries, which are in the variant of the WIMS7B code. We had the three libraries: the 1986 library is based on the UKNDL files, the two 1996 libraries are based on the JEF-2.2 files, the one having the 69 group approximation, the other having the 172 group approximation. We wanted also to have some acquaintance with the new option of WIMS-7B - CACTUS. The variant of WIMS-7B was placed at our disposal by the code authors for a temporal use for 9 months. It was natural to make at comparisons with analogous values of TVS-M, MCU, Apollo-2, Casmo-4, Conkemo, MCNP, HELIOS codes, where the other different libraries were used. In accordance with our aims the calculations of unprofiled and profiled assemblies of the VVER-1000 reactor have been carried out by the option CACTUS. This option provides calculations by the characteristics method. The calculation results have been compared with the K∞ values obtained by other codes in work. The conclusion from this analysis is such: the methodical parts of errors of these codes have nearly the same values. Spacing for Keff values can be explained of the library microsections differences mainly. Nevertheless, the more detailed analysis of the results obtained is required. In conclusion the calculation of a depletion of VVER-1000 cell has been carried out. The comparison of the dependency of the multiply factor from the depletion obtained by WIMS-7B with different libraries and by the TVS-M, MCU, HELIOS and WIMS-ABBN codes in work has been performed. (orig.)
Neutron and photon shielding benchmark calculations by MCNP on the LR-0 experimental facility.
Hordósy, G
2005-01-01
In the framework of the REDOS project, the space-energy distribution of the neutron and photon flux has been calculated over the pressure vessel simulator thickness of the LR-0 experimental reactor, Rez, Czech Republic. The results calculated by the Monte Carlo code MCNP4C are compared with the measurements performed in the Nuclear Research Institute, Rez. The spectra have been measured at the barrel, in front of, inside and behind the pressure vessel in different configurations. The neutron measurements were performed in the energy range 0.1-10 MeV. This work has been done in the frame of the 5th Frame Work Programme of the European Community 1998-2002. PMID:16604591
500-MeV electron beam bench-mark experiments and calculations
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Experiments measuring the energy deposited by electron beams were performed to provide bench marks against which to evaluate our HANDYL76 electron beam computer code. The experiments, done at Stanford's Mk III accelerator, measured dose vs depth and dose vs radius profiles induced in layered aluminum targets by 500-MeV electrons. The dose was measured by passive thermoluminescence and photographic film placed between aluminum plates. The calculations predict a dose vs radius profile that forward-peaks on axis after the beam passes through a 200-cm air gap; the experimental measurements do not show this peak. This discrepancy indicates there may be a problem in using HANDYL76 to calculate deep penetration of a target with a large gap
Benchmark Calculation for the VHTR 2-D Core by Using the DeCART Code
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Cho, Jin-Young; Kim, Kang-Seog; Lee, Chung-Chan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
2006-07-01
Recently, a hexagonal module has been equipped to the DeCART (Deterministic Core Analysis based on Ray Tracing) whole core code for a hexagonal core analysis. The equipment includes a ray tracing module to solve the 2-D whole-core transport problem and a multi-group CMFD module to perform an efficient transport calculation. In this paper, the capability of the DeCART hexagonal module is examined by solving VHTR core problems.
Benchmark Calculation for the VHTR 2-D Core by Using the DeCART Code
International Nuclear Information System (INIS)
Recently, a hexagonal module has been equipped to the DeCART (Deterministic Core Analysis based on Ray Tracing) whole core code for a hexagonal core analysis. The equipment includes a ray tracing module to solve the 2-D whole-core transport problem and a multi-group CMFD module to perform an efficient transport calculation. In this paper, the capability of the DeCART hexagonal module is examined by solving VHTR core problems
DEFF Research Database (Denmark)
Cai, Xiao-Xiao; Llamas-Jansa, Isabel; Mullet, Steven;
2013-01-01
Geant4 is an open source general purpose simulation toolkit for particle transportation in matter. Since the extension of the thermal scattering model in Geant4.9.5 and the availability of the IAEA HP model cross section libraries, it is now possible to extend the application area of Geant4 to re...... models and the G4NDL library. However, cross sections of those missing isotopes were made available recently through the IAEA project “new evaluated neutron cross section libraries for Geant4”....
A proposal for a new U-D2O criticality benchmark: RB reactor core 39/1978
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Pešić Milan P.
2012-01-01
Full Text Available In 1958, the experimental RB reactor was designed as a heavy water critical assembly with natural uranium metal rods. It was the first nuclear fission critical facility at the Boris Kidrič (now Vinča Institute of Nuclear Sciences in the former Yugoslavia. The first non-reflected, unshielded core was assembled in an aluminium tank, at a distance of around 4 m from all adjacent surfaces, so as to achieve as low as possible neutron back reflection to the core. The 2% enriched uranium metal and 80% enriched uranium dioxide (dispersed in aluminum fuel elements (known as slugs were obtained from the USSR in 1960 and 1976, respectively. The so-called “clean” cores of the RB reactor were assembled from a single type of fuel elements. The “mixed” cores of the RB reactor, assembled from two or three types of different fuel elements, were also positioned in heavy water. Both types of cores can be composed as square lattices with different pitches, covering a range of 7 cm to 24 cm. A radial heavy water reflector of various thicknesses usually surrounds the cores. Up to 2006, four sets of clean cores (44 core configurations have been accepted as criticality benchmarks and included into the OECD ICSBEP Handbook. The RB mixed core 39/1978 was made of 31 natural uranium metal rods positioned in heavy water, in a lattice with a pitch of 8√2 cm and 78
Full CI benchmark calculations for several states of the same symmetry
Bauschlicher, Charles W., Jr.; Taylor, Peter R.
1987-01-01
Full CI (FCI) wave functions are used to compute energies for several electronic states of the same symmetry for SiH2, CH2, and CH2(+). It is found that CASSCF/multireference CI wave functions yield results very similar to FCI, irrespective of whether the CASSCF MOs are optimized independently for each state or using an average of the CASSCF energies for all desired states. The ionization potentials and excitation energies obtained from the FCI calculations should help calibrate methods (such as Green's function approaches, equations of motion and propagator methods, and cluster expansions) in which energy differences are computed directly.
Sihver, L.; Mancusi, D.; Niita, K.; Sato, T.; Townsend, L.; Farmer, C.; Pinsky, L.; Ferrari, A.; Cerutti, F.; Gomes, I.
Particles and heavy ions are used in various fields of nuclear physics, medical physics, and material science, and their interactions with different media, including human tissue and critical organs, have therefore carefully been investigated both experimentally and theoretically since the 1930s. However, heavy-ion transport includes many complex processes and measurements for all possible systems, including critical organs, would be impractical or too expensive; e.g. direct measurements of dose equivalents to critical organs in humans cannot be performed. A reliable and accurate particle and heavy-ion transport code is therefore an essential tool in the design study of accelerator facilities as well as for other various applications. Recently, new applications have also arisen within transmutation and reactor science, space and medicine, especially radiotherapy, and several accelerator facilities are operating or planned for construction. Accurate knowledge of the physics of interaction of particles and heavy ions is also necessary for estimating radiation damage to equipment used on space vehicles, to calculate the transport of the heavy ions in the galactic cosmic ray (GCR) through the interstellar medium, and the evolution of the heavier elements after the Big Bang. Concerns about the biological effect of space radiation and space dosimetry are increasing rapidly due to the perspective of long-duration astronaut missions, both in relation to the International Space Station and to manned interplanetary missions in near future. Radiation protection studies for crews of international flights at high altitude have also received considerable attention in recent years. There is therefore a need to develop accurate and reliable particle and heavy-ion transport codes. To be able to calculate complex geometries, including production and transport of protons, neutrons, and alpha particles, 3-dimensional transport using Monte Carlo (MC) technique must be used. Today
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The motivation to conduct this benchmark exercise, a summary of the results, and a discussion of and conclusions from the intercomparison are given in Section 5.2. This section contains further details of the results of the calculations and intercomparisons, illustrated by tables and figures, but avoiding repetition of Section 5.2 as far as possible. (author)
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Monte Carlo N-Particle Transport Code System (MCNP) criticality calculations were performed on a library of critical benchmark experiments to obtain preliminary bias values and subcritical margins to be utilized in licensing calculations for high-level radioactive waste disposal. The critical experiments library includes a broad range of system physical and neutronic characteristics that are representative of a range of potential criticality configurations relevant to long-term deep geological disposal. Two hundred and eighty-nine critical benchmark experiments were selected and grouped into 20 critical experiment classifications. From the results of this study, an applicable subcritical margin or maximum allowable keff can be selected for preliminary repository criticality analysis based on the similarity between the physical and neutronic characteristics of the system being analyzed and the relevant library classification. The results of this study provide quantification of both the confidence associated with the MCNP code and the presented conservative method for performing criticality evaluations relevant to repository emplacement of high-level radioactive waste
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Graphical abstract: The performance of the SOPPA(CC2) method for the calculation of indirect nuclear carbon-carbon spin-spin coupling constants is tested on 197 coupling constants in 41 carbocycles. Research highlights: → Benchmarking of SOPPA(CC2) for carbon-carbon coupling constants in carbocycles. → SOPPA(CC2) scales as SOPPA. → SOPPA(CC2) performs well for indirect carbon-carbon coupling constants. → SOPPA(CC2) gives mean absolute errors of 1.11 Hz relative to experimental values. → SOPPA(CC2) performs better than SOPPA for couplings across more than one bond. - Abstract: We investigate the performance of the newly implemented SOPPA(CC2) method for the calculation of indirect carbon-carbon spin-spin coupling constants. SOPPA(CC2) scales as SOPPA, but has previously been shown to improve the accuracy of spin-spin coupling constants relative to CCSD. We compare the results of SOPPA(CC2) with SOPPA, SOPPA(CCSD), and available experimental values for a wide range of saturated carbocycles (in total 41 carbocycles and 197 coupling constants). It follows that SOPPA(CC2) performs better than SOPPA for couplings across more than one bond, while the two methods performs equally well for the one-bond couplings relatively to SOPPA(CCSD).
Moraitis, K; Georgoulis, M K; Archontis, V
2014-01-01
In earlier works we introduced and tested a nonlinear force-free (NLFF) method designed to self-consistently calculate the free magnetic energy and the relative magnetic helicity budgets of the corona of observed solar magnetic structures. The method requires, in principle, only a single, photospheric or low-chromospheric, vector magnetogram of a quiet-Sun patch or an active region and performs calculations in the absence of three-dimensional magnetic and velocity-field information. In this work we strictly validate this method using three-dimensional coronal magnetic fields. Benchmarking employs both synthetic, three-dimensional magnetohydrodynamic simulations and nonlinear force-free field extrapolations of the active-region solar corona. We find that our time-efficient NLFF method provides budgets that differ from those of more demanding semi-analytical methods by a factor of ~3, at most. This difference is expected from the physical concept and the construction of the method. Temporal correlations show mo...
International Nuclear Information System (INIS)
In 2005 the Argentine Government took the decision to complete the construction of the Atucha-II nuclear power plant, which has been progressing slowly during the last ten years. Atucha-II is a 745 MWe nuclear station moderated and cooled with heavy water, of German (Siemens) design located in Argentina. It has a pressure vessel design with 451 vertical coolant channels and the fuel assemblies (FA) are clusters of 37 natural UO2 rods with an active length of 530 cm. For the reactor physics area, a revision and update of reactor physics calculation methods and models was recently carried out covering cell, supercell (control rod) and core calculations. This paper presents benchmark comparisons of core parameters of a slightly idealized model of the Atucha-I core obtained with the PUMA reactor code with MCNP5. The Atucha-I core was selected because it is smaller, similar from a neutronic point of view, more symmetric than Atucha-II, and has some experimental data available. To validate the new models benchmark comparisons of k-effective, channel power and axial power distributions obtained with PUMA and MCNP5 have been performed. In addition, a simple cell heterogeneity correction recently introduced in PUMA is presented, which improves significantly the agreement of calculated channel powers with MCNP5. To complete the validation, the calculation of some of the critical configurations of the Atucha-I reactor measured during the experiments performed at first criticality is also presented. (authors)
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The purpose of this work is to validate MCNP5 libraries by simulating 4 detailed benchmark experiments and comparing MCNP5 results (each library) with the experimental results and also the previously validated codes for the same experiments MORET 4.A coupled with APOLLO2 (France), and MONK8 (UK). The reasons for difference between libraries are also investigated in this work. Investigating the reason for the differences between libraries will be done by specifying a different library for specific part (clad, fuel, light water) and checking the result deviation than the previously calculated result (with all parts of the same library). The investigated benchmark experiments are of single fuel rods arrays that are water-moderated and water-reflected. Rods contained low-enriched (4.738 wt.% 92235U)uranium dioxide (UO2) fuel were clad with aluminum alloy AGS. These experiments were subcritical approaches extrapolated to critical, with the multiplication factor reached being very close to 1.000 (within 0.1%); the subcritical approach parameter was the water level. The studied four cases differ from each other in pitch, number of fuel rods and of course critical height of water. The results show that although library ENDF/B-IV lacks light water treatment card, however its results can be reliable as light water treatment library does not have significant differences from library to another, so it will not be necessary to specify light water treatment card. The main reason for differences between ENDF/B-V and ENDF/B-VI is light water material, especially the Hydrogen element. Specifying the library of Uranium is necessary in case of using library ENDF/B-IV. On the other hand it is not necessary to specify library of cladding material whatever the used library. Validated libraries are ENDF/BIV, ENDF/B-V and ENDF/B-VI with codes in MCNP 42C, 50C and 60C respectively. The presentation slides have been added to the article
Influence of the FLUKA geometrical model on the ADS demonstration facility criticality calculations
International Nuclear Information System (INIS)
Criticality calculations for a typical ADS have been performed by means of the codes chain FLUKA-CMCS (formerly FLUKA - MC2). the FLUKA code calculates the neutron source for the subsequent neutron transport calculations performed by CMCS. In this paper it is investigated how the geometrical model used in FLUKA can impact the CMCS criticality calculations. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Bessette, Gregory Carl
2004-09-01
Modeling the response of buried reinforced concrete structures subjected to close-in detonations of conventional high explosives poses a challenge for a number of reasons. Foremost, there is the potential for coupled interaction between the blast and structure. Coupling enters the problem whenever the structure deformation affects the stress state in the neighboring soil, which in turn, affects the loading on the structure. Additional challenges for numerical modeling include handling disparate degrees of material deformation encountered in the structure and surrounding soil, modeling the structure details (e.g., modeling the concrete with embedded reinforcement, jointed connections, etc.), providing adequate mesh resolution, and characterizing the soil response under blast loading. There are numerous numerical approaches for modeling this class of problem (e.g., coupled finite element/smooth particle hydrodynamics, arbitrary Lagrange-Eulerian methods, etc.). The focus of this work will be the use of a coupled Euler-Lagrange (CEL) solution approach. In particular, the development and application of a CEL capability within the Zapotec code is described. Zapotec links two production codes, CTH and Pronto3D. CTH, an Eulerian shock physics code, performs the Eulerian portion of the calculation, while Pronto3D, an explicit finite element code, performs the Lagrangian portion. The two codes are run concurrently with the appropriate portions of a problem solved on their respective computational domains. Zapotec handles the coupling between the two domains. The application of the CEL methodology within Zapotec for modeling coupled blast/structure interaction will be investigated by a series of benchmark calculations. These benchmarks rely on data from the Conventional Weapons Effects Backfill (CONWEB) test series. In these tests, a 15.4-lb pipe-encased C-4 charge was detonated in soil at a 5-foot standoff from a buried test structure. The test structure was composed of a
Using ORIGEN and MCNP to calculate reactor criticals and burnup effects
International Nuclear Information System (INIS)
The purpose of this modeling effort was to verify the applicability of using ORIGEN-S and MCNP to the analysis of spent fuel of various enrichments and burnups. By comparing the results of criticality studies using MCNP and ORIGEN-S with the measured keff of 1.0, the suitability of the coupled ORIGEN-S/ MCNP package was determined. This study presents the results of the benchmark modeling of five pressurized water reactor (PWR) critical configurations. For these analyses, a combination of ORIGEN-S and MCNP was used to analyze the fuel depletion and criticality of five power reactor core configuration
International Nuclear Information System (INIS)
MPI parallelism are implemented on a SUN Workstation for running MCNPX and on the High Performance Computing Facility (HPC) for running MCNP5. 23 input less obtained from MCNP Criticality Validation Suite are utilized for the purpose of evaluating the amount of speed up achievable by using the parallel capabilities of MPI. More importantly, we will study the economics of using more processors and the type of problem where the performance gain are obvious. This is important to enable better practices of resource sharing especially for the HPC facilities processing time. Future endeavours in this direction might even reveal clues for best MCNP5/ MCNPX coding practices for optimum performance of MPI parallelisms. (author)
Five radionuclide vadose zone models with different degrees of complexity (CHAIN, MULTIMED_DP, FECTUZ, HYDRUS, and CHAIN 2D) were selected for use in soil screening level (SSL) calculations. A benchmarking analysis between the models was conducted for a radionuclide (99Tc) rele...
Calculation of Heterogeneity Effects in Fast Critical Assemblies
International Nuclear Information System (INIS)
The physical properties of the fast power reactor cores with respect to the neutron transport are usually not essentially influenced by the heterogeneity of the core structure. Nevertheless, when modelling the cores of these reactors by experimental critical assemblies the heterogeneity of the structure becomes essentially greater and cannot be neglected in the analysis of experiments. In connection with this fact a further development of the previously presented method of homogenization of slab lattices with respect to the chain reaction with fast neutrons is contributed. Attention is restricted to bare assemblies made up of a periodic lattice of parallel slabs of two different materials, a so-called ''sandwich'' reactor of materials 1 and 2. The geometry considered is (a) the plane geometry with boundaries parallel to the slabs, (b) the plane geometry with boundaries perpendicular to the slabs, and (c) the cylindrical geometry with boundaries perpendicular to the slabs with infinite or finite height. The restriction to two different media is made because of a simplification only and is not substantial. The starting point is the integral transport theory based on the solution of the kinetic Boltzmann equation. The solution is sought in the form of a product of asymptotic transport theory solution characterized by the buckling B2 and representing the general trend of the neutron emission density with the fine structure, taking account of the heterogeneous structure of alternative layers of media 1 and 2. As a result the value of the buckling B2 of the heterogeneous assembly together with energy spectra in each medium is obtained. As an illustration of the method the chain reaction with fast neutrons in a heterogeneous medium consisting of 235U and 238U plates placed in turn is investigated. For 16 energy groups the buckling B2 and the energy spectrum in 235U and 238U have been calculated for a 10%, 20% and 30% enrichment with 235U for different cell dimensions of
Swedish analysis of NEA/CSNI benchmark problems for criticality codes
International Nuclear Information System (INIS)
The Monte Carlo methods used by the members of the working group are adequate for calculations on large arrays. The differences in the results from different codes are probably caused by the differences in cross sections. The previous difficulties in obtaining good results for bare arrays of UNH-solution are, at least to some extent, explained by incomplete information. The neutron reflection from walls, ceiling and floor has earlier been neglected. The inclusion of these in the input to the Monte Carlo codes appears to lead to adequate results. The basis for the IAEA rules of calculating allowable numbers of fissile packages mixed with other packages (fissile or not) during transport, does not seem justified. This has been demonstrated for theoretical package designs. It has not been confirmed by the other members of the working group and no conclusion was drawn by the group. It is very likely that a mix of real packages can be found that supports the mentioned theoretical demonstration. (author)
Reactor group constants and benchmark test
International Nuclear Information System (INIS)
The evaluated nuclear data files such as JENDL, ENDF/B-VI and JEF-2 are validated by analyzing critical mock-up experiments for various type reactors and assessing applicability for nuclear characteristics such as criticality, reaction rates, reactivities, etc. This is called Benchmark Testing. In the nuclear calculations, the diffusion and transport codes use the group constant library which is generated by processing the nuclear data files. In this paper, the calculation methods of the reactor group constants and benchmark test are described. Finally, a new group constants scheme is proposed. (author)
Merger of Nuclear Data with Criticality Safety Calculations
International Nuclear Information System (INIS)
In this paper we report on current activities related to the merger of differential/integral data (especially in the resolved-resonance region) with nuclear criticality safety computations. Techniques are outlined for closer coupling of many processes measurement, data reduction, differential-data analysis, integral-data analysis, generating multigroup cross sections, data-testing, criticality computations which in the past have been treated independently
International Nuclear Information System (INIS)
Benchmark results of the Dutch PINK working group on calculational benchmarks on single pin cell and multipin assemblies as defined by EPRI are presented and evaluated. First a short update of methods used by the various institutes involved is given as well as an update of the status with respect to previous performed pin-cell calculations. Problems detected in previous pin-cell calculations are inspected more closely. Detailed discussion of results of multipin assembly calculations is given. The assembly consists of 9 pins in a multicell square lattice in which the central pin is filled differently, i.e. a Gd pin for the BWR assembly and a control rod/guide tube for the PWR assembly. The results for pin cells showed a rather good overall agreement between the four participants although BWR pins with high void fraction turned out to be difficult to calculate. With respect to burnup calculations good overall agreement for the reactivity swing was obtained, provided that a fine time grid is used. (orig.)
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Critical mass experiments were performed using assemblies which simulated one-dimensional lattice consisting of shielding containers with metal fissile materials. Calculations of the criticality of the above assemblies were carried out using the KLAN program with the BAS neutron constants. Errors in the calculations of the criticality for one-, two-, and three-dimensional lattices are estimated. 3 refs.; 1 tab
DEFF Research Database (Denmark)
Cismondi, Martin; Michelsen, Michael Locht
2007-01-01
A general strategy for global phase equilibrium calculations (GPEC) in binary mixtures is presented in this work along with specific methods for calculation of the different parts involved. A Newton procedure using composition, temperature and Volume as independent variables is used for calculation...... of critical lines. Each calculated point is analysed for stability by means of the tangent plane distance, and the occurrence of an unstable point is used to determine a critical endpoint (CEP). The critical endpoint, in turn, is used as the starting point for constructing the three-phase line. The...... equations for the critical endpoint, as well as for points on the three-phase line, are also solved using Newton's method with temperature, molar volume and composition as the independent variables. The different calculations are integrated into a general procedure that allows us to automatically trace...
International Nuclear Information System (INIS)
Over 50 neutron benchmark calculations have recently been completed as part of an ongoing program to validate the MCNP Monte Carlo radiation transport code. The new and significant aspects of this work are as follows: These calculations are the first attempt at a validation program for MCNP and the first official benchmarking of version 4 of the code. We believe the chosen set of benchmarks is a comprehensive set that may be useful for benchmarking other radiation transport codes and data libraries. These calculations provide insight into how well neutron transport calculations can be expected to model a wide variety of problems
Comparison of MCNPX and Albedo method in criticality calculation
International Nuclear Information System (INIS)
This study aims to conduct a computer simulation that will calculate the reactivity of a homogeneous reactor and compare the results with the calculations made by the albedo method. The simulation will be developed using the MCNPX. The study compared the results calculated for a hypothetical reactor by the albedo method for four groups of energy with those obtained by the MCNPX simulation. The design of the reactor is spherical and homogeneous with a reflector of finite thickness. The value obtained for the neutron effective multiplication factor - keff will be compared. Different situations were simulated in order to obtain results closer to the compared method and reality. The was Good consistency could be noticed between the calculated results. (author)
International Nuclear Information System (INIS)
The purpose of this calculation note is to provide the basis for criticality consequences for the Tank Farm Safety Analysis Report (FSAR). Criticality scenario is developed and details and description of the analysis methods are provided
VHTRC temperature coefficient benchmark problem
International Nuclear Information System (INIS)
As an activity of IAEA Coordinated Research Programme, a benchmark problem is proposed for verifications of neutronic calculation codes for a low enriched uranium fuel high temperature gas-cooled reactor. Two problems are given on the base of heating experiments at the VHTRC which is a pin-in-block type core critical assembly loaded mainly with 4% enriched uranium coated particle fuel. One problem, VH1-HP, asks to calculate temperature coefficient of reactivity from the subcritical reactivity values at five temperature steps between an room temperature where the assembly is nearly at critical state and 200degC. The other problem, VH1-HC, asks to calculate the effective multiplication factor of nearly critical loading cores at the room temperature and 200degC. Both problems further ask to calculate cell parameters such as migration area and spectral indices. Experimental results corresponding to main calculation items are also listed for comparison. (author)
Benchmark testing of 233U evaluations
International Nuclear Information System (INIS)
In this paper we investigate the adequacy of available 233U cross-section data (ENDF/B-VI and JENDL-3) for calculation of critical experiments. An ad hoc revised 233U evaluation is also tested and appears to give results which are improved relative to those obtained with either ENDF/B-VI or JENDL-3 cross sections. Calculations of keff were performed for ten fast benchmarks and six thermal benchmarks using the three cross-section sets. Central reaction-rate-ratio calculations were also performed
International Nuclear Information System (INIS)
Theoretical consideration is made for possibility to accelerate and judge convergence of a conventional Monte Carlo iterative calculation when it is used for a weak neutron interaction problem. And the clue for this consideration is rendered with some application analyses using the OECD/NEA source convergence benchmark problems. Some practical procedures are proposed to realize these acceleration and judgment methods in practical application using a Monte Carlo code. (author)
International Nuclear Information System (INIS)
Incorporating full three-dimensional (3-D) models of the reactor core into system transient codes allows for a 'best-estimate' calculation of interactions between the core behavior and plant dynamics. Recent progress in computer technology has made the development of coupled thermal-hydraulic (T-H) and neutron kinetics code systems feasible. Considerable efforts have been made in various countries and organizations in this direction. Appropriate benchmarks need to be developed that will permit testing of two particular aspects. One is to verify the capability of the coupled codes to analyze complex transients with coupled core-plant interactions. The second is to test fully the neutronics/T-H coupling. One such benchmark is the Pressurized Water Reactor Main Steam Line Break (MSLB) Benchmark problem. It was sponsored by the Organization for Economic Cooperation and Development, U.S. Nuclear Regulatory Commission, and The Pennsylvania State University. The benchmark problem uses a 3-D neutronics core model that is based on real plant design and operational data for the Three Mile Island Unit 1 nuclear power plant. The purpose of this benchmark is threefold: to verify the capability of system codes for analyzing complex transients with coupled core-plant interactions; to test fully the 3-D neutronics/T-H coupling; and to evaluate discrepancies among the predictions of coupled codes in best-estimate transient simulations. The purposes of the benchmark are met through the application of three exercises: a point kinetics plant simulation (exercise 1), a coupled 3-D neutronics/core T-H evaluation of core response (exercise 2), and a best-estimate coupled core-plant transient model (exercise 3).In this paper we present the three exercises of the MSLB benchmark, and we summarize the findings of the participants with regard to the current numerical and computational issues of coupled calculations. In addition, this paper reviews in some detail the sensitivity studies on
Criticality calculation of the nuclear material warehouse of the ININ
International Nuclear Information System (INIS)
In this work the conditions of nuclear safety were determined as much in normal conditions as in the accident event of the nuclear fuel warehouse of the reactor TRIGA Mark III of the Instituto Nacional de Investigaciones Nucleares (ININ). The warehouse contains standard fuel elements Leu - 8.5/20, a control rod with follower of standard fuel type Leu - 8.5/20, fuel elements Leu - 30/20, and the reactor fuel Sur-100. To check the subcritical state of the warehouse the effective multiplication factor (keff) was calculated. The keff calculation was carried out with the code MCNPX. (Author)
Quantum mechanical cluster calculations of critical scintillation processes
Derenzo, Stephen E.; Klintenberg, Mattias K.; Weber, Marvin J.
2000-01-01
This paper describes the use of commercial quantum chemistry codes to simu-late several critical scintillation processes. The crystal is modeled as a cluster of typically 50 atoms embedded in an array of typically 5,000 point charges designed to reproduce the electrostatic field of the infinite crystal. The Schrodinger equation is solved for the ground, ionized, and excited states of the system to determine the energy and electron wavefunction. Computational methods for the following cri...
CRISTAL V1: Criticality package for burn up credit calculations
International Nuclear Information System (INIS)
The first version of the CRISTAL package, created and validated as part of a joint project between IRSN, COGEMA and CEA, was delivered to users in November 1999. This fruitful cooperation between IRSN, COGEMA and CEA has been pursued until 2003 with the development and the validation of the package CRISTAL V1, whose main objectives are to improve the criticality safety studies including the Burn up Credit effect. (author)
International Nuclear Information System (INIS)
The NEA Working Party on Nuclear Criticality Safety established an Expert Group on Criticality Excursion Analysis in 2001 to explore the performance of various transient codes to evaluate criticality accidents in a fissile solution. Inter-code comparison exercises among four transient codes (AGNES, CRITEX, INCTAC and TRACE) have been carried out with typical transient experiments using uranyl nitrate fuel solution. Two sets of benchmarks were carried out based on experimental programmes performed in the Tracy reactor in Japan, and the Silene reactor in France. Tracy and Silene have the same geometrical features: an annular cylinder with a central void tube for a transient rod and similar operational modes for reactivity insertion. The experiments selected are representative benchmarks for low- and high-enriched uranyl nitrate solution, about 10 wt% for Tracy and 93 wt% for the Silene core. This report provides an analysis of the benchmark results obtained with four different codes. It will be of particular interest to criticality safety practitioners developing transient codes, notably since little experimental data is available and the existing transient codes are presently unavailable to the public. (authors)
Cross section library based discrepancies in MCNP criticality calculations
International Nuclear Information System (INIS)
In nuclear engineering several reactor physics problems can be approached using Monte Carlo neutron transport techniques, which usually give reliable results when properly used. The quality of the results is largely determined by the accuracy of the geometry model and the statistical uncertainty of the Monte Carlo calculation. There is, however, another potential source of error, namely the cross section data used with the Monte Carlo codes. It has been shown in several studies that there may be significant discrepancies between results calculated using cross section libraries based on different evaluated nuclear data files. These discrepancies are well known to the evaluators of nuclear data but less acknowledged by reactor physicists, who often rely on a single cross section library in their calculations. In this study, discrepancies originating from base nuclear data were investigated in a systematic manner using the MCNP4C code. Calculations on simplified UOX and MOX fuelled LWR lattices were carried out using cross section libraries based on ENDF/B-VI.8, JEFF-3.0, JENDL-3.3, JEF-2.2 and JENDL-3.2 evaluated data files. The neutron spectrum of the system was varied over a wide range by changing the ratio of hydrogen to heavy metal atoms. The essential isotopes underlying the discrepancies were identified and the roles of fission and absorption cross sections of the most important nuclides assessed. The results confirm that there are large systematic differences up to a few per cent in the multiplication factors of LWR lattices. The discrepancies are strongly dependent on material compositions and neutron spectra, and largely originate from U-238 and the primary fissile isotopes. It is concluded that these discrepancies should be taken into account in all reactor physics calculations, and that reactor physicists should not rely on results based on a single cross section library. (author)
International Nuclear Information System (INIS)
In this paper, the results of the investigations on the nodalization effects for the ATHLET code are presented and discussed in details on the basis of experimental data for the VVER-1000 Coolant Transient Benchmark with different operating modes of four main coolant pumps. ATHLET calculations with different nodalization and their impact was analyzed. The work studied the influence of annular outlet nodalization on calculation of coolant temperature. By comparing the test data versus calculated by ATHLET we showed a good agreement between the experimental data and simulation results for analyzed parameters.
International Nuclear Information System (INIS)
The Reactor Physics Committee of the Nuclear Energy Agency has set up a series of benchmark calculations to compare the performance of the various codes used in shielding calculations for fuel transport flasks. For one benchmark the calculations are to be compared with dose-rates measured outside a French TN-12 flask loaded with 12 irradiated fuel elements from a PWR. Neutron dose-rate measurements were made with a Helium-3 detector encased in paraffin with an unknown response. It did not produce measurements of equivalent dose-rates therefore standard flux to dose conversion factors could not be used in the calculations. A 1-dimensional adjoint calculation was carried out by the CEA at Saclay to determine the response function for the detector and this was used to calculate neutron dose-rates for the flask. However the dose-rates calculated away from the surface of the flask were underestimated suggesting that there was an angular dependence of the response. This report describes MCBEND calculations which have been performed to produce an angular response function for the detector, which was used to provide revised dose-rates. (author)
Energy Technology Data Exchange (ETDEWEB)
Locke, H.F.
1991-11-01
The Reactor Physics Committee of the Nuclear Energy Agency has set up a series of benchmark calculations to compare the performance of the various codes used in shielding calculations for fuel transport flasks. For one benchmark the calculations are to be compared with dose-rates measured outside a French TN-12 flask loaded with 12 irradiated fuel elements from a PWR. Neutron dose-rate measurements were made with a Helium-3 detector encased in paraffin with an unknown response. It did not produce measurements of equivalent dose-rates therefore standard flux to dose conversion factors could not be used in the calculations. A 1-dimensional adjoint calculation was carried out by the CEA at Saclay to determine the response function for the detector and this was used to calculate neutron dose-rates for the flask. However the dose-rates calculated away from the surface of the flask were underestimated suggesting that there was an angular dependence of the response. This report describes MCBEND calculations which have been performed to produce an angular response function for the detector, which was used to provide revised dose-rates. (author).
Quantum mechanical cluster calculations of critical scintillation processes
International Nuclear Information System (INIS)
This paper describes the use of commercial quantum chemistry codes to simulate several critical scintillation processes. The crystal is modeled as a cluster of typically 50 atoms embedded in an array of typically 5,000 point charges designed to reproduce the electrostatic field of the infinite crystal. The Schrodinger equation is solved for the ground, ionized, and excited states of the system to determine the energy and electron wave function. Computational methods for the following critical processes are described: (1) the formation and diffusion of relaxed holes, (2) the formation of excitons, (3) the trapping of electrons and holes by activator atoms, (4) the excitation of activator atoms, and (5) thermal quenching. Examples include hole diffusion in CsI, the exciton in CsI, the excited state of CsI:Tl, the energy barrier for the diffusion of relaxed holes in CaF2 and PbF2, and prompt hole trapping by activator atoms in CaF2:Eu and CdS:Te leading to an ultra-fast (<50ps) scintillation rise time.
Validation of KENO-based criticality calculations at Rocky Flats
International Nuclear Information System (INIS)
In the absence of experimental data, it is necessary to rely on computer-based computational methods in evaluating the criticality condition of a nuclear system. The validity of the computer codes is established in a two-part procedure as outlined in ANSI/ANS 8.1. The first step, usually the responsibility of the code developer, involves verification that the algorithmic structure of the code is performing the intended mathematical operations correctly. The second step involves an assessment of the code's ability to realistically portray the governing physical processes in question. This is accomplished by determining the code's bias, or systematic error, through a comparison of computational results to accepted values obtained experimentally. In this paper, the authors discuss the validation process for KENO and the Hansen-Roach cross sections in use at EG and G Rocky Flats. The validation process at Rocky Flats consists of both global and local techniques. The global validation resulted in a maximum keff limit of 0.95 for the limiting-accident scanarios of a criticality evaluation
International Nuclear Information System (INIS)
Continuous responses (e.g. body weight) are widely used in risk assessment for determining the benchmark dose (BMD) which is used to derive a U.S. EPA reference dose. One critical question that is not often addressed in dose–response assessments is whether to model the continuous data as normally or log-normally distributed. Additionally, if lognormality is assumed, and only summarized response data (i.e., mean ± standard deviation) are available as is usual in the peer-reviewed literature, the BMD can only be approximated. In this study, using the “hybrid” method and relative deviation approach, we first evaluate six representative continuous dose–response datasets reporting individual animal responses to investigate the impact on BMD/BMDL estimates of (1) the distribution assumption and (2) the use of summarized versus individual animal data when a log-normal distribution is assumed. We also conduct simulation studies evaluating model fits to various known distributions to investigate whether the distribution assumption has influence on BMD/BMDL estimates. Our results indicate that BMDs estimated using the hybrid method are more sensitive to the distribution assumption than counterpart BMDs estimated using the relative deviation approach. The choice of distribution assumption has limited impact on the BMD/BMDL estimates when the within dose-group variance is small, while the lognormality assumption is a better choice for relative deviation method when data are more skewed because of its appropriateness in describing the relationship between mean and standard deviation. Additionally, the results suggest that the use of summarized data versus individual response data to characterize log-normal distributions has minimal impact on BMD estimates. - Highlights: • We investigate to what extent the distribution assumption can affect BMD estimates. • Both real data analysis and simulation study are conducted. • BMDs estimated using hybrid method are more
International Nuclear Information System (INIS)
The assessment of the uncertainties of COBRA-IIIC thermal-hydraulic analyses of rod bundles is performed for a 5-by-5 bundle representing a PWR fuel assembly. In the first part of the paper the modeling uncertainties are evaluated in the term of the uncertainty of the turbulent mixing factor using the OECD NEA/NRC PSBT benchmark data. After that the uncertainties of the COBRA calculations are discussed performing Monte-Carlo type statistical analyses taking into account the modeling uncertainties and other uncertainties prescribed in the OECD NEA UAM benchmark specification. Both steady-state and transient cases are investigated. The target quantities are the uncertainties of the void distribution, the moderator density, the moderator temperature and the DNBR. We will see that - beyond the uncertainties of the geometry and the boundary conditions - it is very important to take into account the modeling uncertainties in case of bundle or sub-channel thermo-hydraulic calculations.
Research on GPU Acceleration for Monte Carlo Criticality Calculation
Xu, Qi; Yu, Ganglin; Wang, Kan
2014-06-01
The Monte Carlo neutron transport method can be naturally parallelized by multi-core architectures due to the dependency between particles during the simulation. The GPU+CPU heterogeneous parallel mode has become an increasingly popular way of parallelism in the field of scientific supercomputing. Thus, this work focuses on the GPU acceleration method for the Monte Carlo criticality simulation, as well as the computational efficiency that GPUs can bring. The "neutron transport step" is introduced to increase the GPU thread occupancy. In order to test the sensitivity of the MC code's complexity, a 1D one-group code and a 3D multi-group general purpose code are respectively transplanted to GPUs, and the acceleration effects are compared. The result of numerical experiments shows considerable acceleration effect of the "neutron transport step" strategy. However, the performance comparison between the 1D code and the 3D code indicates the poor scalability of MC codes on GPUs.
Critical evaluation of German regulatory specifications for calculating radiological exposure
International Nuclear Information System (INIS)
The assessment of radiological exposure of the public is an issue at the interface between scientific findings, juridical standard setting and political decision. The present work revisits the German regulatory specifications for calculating radiological exposure, like the already existing calculation model General Administrative Provision (AVV) for planning and monitoring nuclear facilities. We address the calculation models for the recent risk assessment regarding the final disposal of radioactive waste in Germany. To do so, a two-pronged approach is pursued. One part deals with radiological examinations of the groundwater-soil-transfer path of radionuclides into the biosphere. Processes at the so-called geosphere-biosphere-interface are examined, especially migration of I-129 in the unsaturated zone. This is necessary, since the German General Administrative Provision does not consider radionuclide transport via groundwater from an underground disposal facility yet. Especially data with regard to processes in the vadose zone are scarce. Therefore, using I-125 as a tracer, immobilization and mobilization of iodine is investigated in two reference soils from the German Federal Environment Agency. The second part of this study examines how scientific findings but also measures and activities of stakeholders and concerned parties influence juridical standard setting, which is necessary for risk management. Risk assessment, which is a scientific task, includes identification and investigation of relevant sources of radiation, possible pathways to humans, and maximum extent and duration of exposure based on dose-response functions. Risk characterization identifies probability and severity of health effects. These findings have to be communicated to authorities, who have to deal with the risk management. Risk management includes, for instance, taking into account acceptability of the risk, actions to reduce, mitigate, substitute or monitor the hazard, the setting of
Critical evaluation of German regulatory specifications for calculating radiological exposure
Energy Technology Data Exchange (ETDEWEB)
Koenig, Claudia; Walther, Clemens [Hannover Univ. (Germany). Inst. of Radioecology; Smeddinck, Ulrich [Technische Univ. Braunschweig (Germany). Inst. of Law
2015-07-01
The assessment of radiological exposure of the public is an issue at the interface between scientific findings, juridical standard setting and political decision. The present work revisits the German regulatory specifications for calculating radiological exposure, like the already existing calculation model General Administrative Provision (AVV) for planning and monitoring nuclear facilities. We address the calculation models for the recent risk assessment regarding the final disposal of radioactive waste in Germany. To do so, a two-pronged approach is pursued. One part deals with radiological examinations of the groundwater-soil-transfer path of radionuclides into the biosphere. Processes at the so-called geosphere-biosphere-interface are examined, especially migration of I-129 in the unsaturated zone. This is necessary, since the German General Administrative Provision does not consider radionuclide transport via groundwater from an underground disposal facility yet. Especially data with regard to processes in the vadose zone are scarce. Therefore, using I-125 as a tracer, immobilization and mobilization of iodine is investigated in two reference soils from the German Federal Environment Agency. The second part of this study examines how scientific findings but also measures and activities of stakeholders and concerned parties influence juridical standard setting, which is necessary for risk management. Risk assessment, which is a scientific task, includes identification and investigation of relevant sources of radiation, possible pathways to humans, and maximum extent and duration of exposure based on dose-response functions. Risk characterization identifies probability and severity of health effects. These findings have to be communicated to authorities, who have to deal with the risk management. Risk management includes, for instance, taking into account acceptability of the risk, actions to reduce, mitigate, substitute or monitor the hazard, the setting of
International Nuclear Information System (INIS)
The paper gives a brief survey of the 6th three-dimensional AER dynamic benchmark calculation results received with the codes DYN3D and RELAP5-3D at NRI Rez. This benchmark was defined at the 10th AER Symposium. Its initiating event is a double ended break in the steam line of steam generator No. 1 in a WWER-440/213 plant at the end of the first fuel cycle and in hot full power conditions. Stationary and burnup calculations as well as tuning of initial state before the transient were performed with the code DYN3D. Transient calculations were made with the system code RELAP5-3D. The KASSETA library was used for the generation of reactor core neutronic parameters. The detailed six loops model of NPP Dukovany was adopted for the 6th AER dynamic benchmark purposes. The RELAP5-3D full core neutronic model was connected with 37 coolant channels thermal-hydraulic model of the core, 6-sector nodalization of reactor downcomer, lower and upper plenum was used. Mixing in lower and upper plenum was simulated. The first part of paper contains a brief characteristic of RELAP5 -3D system code and a short description of NPP input deck and reactor core model. The second part shows the time dependencies of important global and local parameters (Authors)
International Nuclear Information System (INIS)
The lack of suitable benchmark problems makes it difficult to test sensitivity codes with a covariance library. A benchmark problem has therefore been defined for one- and two-dimensional sensitivity and uncertainity analysis codes and code systems. The problem, representative of a fusion reactor blanket, has a simple, three-zone /tau/-z geometry containing a D-T fusion neutron source distributed in a central void region surrounded by a thick 6LiH annulus. The response of interest is the 6Li tritium production per source neutron, T6. The calculation has been performed with SENSIBL using other codes from the AARE code system as a test of both SENSIBL and the linked, modular system. The caluclation was performed using the code system in the standard manner with a covariance data library in the COVFILS-2 format but modified to contain specifically tailored covariance data for H and 6Li (Path A). The calculation was also performed by a second method which uses specially perturbed H and Li cross sections (Path B). This method bypasses SENSIBL and allows a hand calculation of the benchmark T6 uncertainties. The results of Path A and Path B were total uncertainties in T6 of 0.21% and 0.19%, respectively. The closeness of the results for this challenging test gives confidence that SENSIBL and the AARE system will perform well for realistic sensitivity and uncertainty analyses
Directory of Open Access Journals (Sweden)
Maria Avramova
2013-01-01
Full Text Available Over the last few years, the Pennsylvania State University (PSU under the sponsorship of the US Nuclear Regulatory Commission (NRC has prepared, organized, conducted, and summarized two international benchmarks based on the NUPEC data—the OECD/NRC Full-Size Fine-Mesh Bundle Test (BFBT Benchmark and the OECD/NRC PWR Sub-Channel and Bundle Test (PSBT Benchmark. The benchmarks’ activities have been conducted in cooperation with the Nuclear Energy Agency/Organization for Economic Co-operation and Development (NEA/OECD and the Japan Nuclear Energy Safety (JNES Organization. This paper presents an application of the joint Penn State University/Technical University of Madrid (UPM version of the well-known sub-channel code COBRA-TF (Coolant Boiling in Rod Array-Two Fluid, namely, CTF, to the steady state critical power and departure from nucleate boiling (DNB exercises of the OECD/NRC BFBT and PSBT benchmarks. The goal is two-fold: firstly, to assess these models and to examine their strengths and weaknesses; and secondly, to identify the areas for improvement.
International Nuclear Information System (INIS)
It was determined that the criticality hazard associated with the Slagging Pyrolysis Incinerator (SPI) Facility would be minimal if a three-level criticality-hazard prevention program were implemented. The first strategy consists of screening all incoming wastes for fissile content. The second prevention level is provided by introducing a small concentration of a neutron-absorbing compound, such as B2O3, into the input waste stream. The third prevention level is provided by direct criticality-hazard monitoring using sensitive neutron detectors in all regions of the facility where a significant hazard has been identified - principally the drying, pyrolysis, and slag regions. The facility could be shut down rapidly for cleanout if the measurements indicate an unsafe condition is developing. The criticality safety provided by the product of these three independent measures should reduce the hazard to a negligible level
International Nuclear Information System (INIS)
Recent findings indicate that gamma radiation can contribute to the embrittlement of reactor materials. On this background an experimental benchmark programme at two low power reactors was started to measure both, neutron and gamma spectral fluences behind and inside of transmission modules consisting of variable iron and water slabs using a NE213 scintillation spectrometer and partly a HPGe detector. The experimental results are used to validate Monte Carlo calculation methods for coupled neutron/gamma problems. The experiment and results of a first series of measurements and comparisons to MCNP calculations for neutron and gamma energy spectra are presented. (author)
Prociuk, Alexander; Van Kuiken, Ben; Dunietz, Barry D
2006-11-28
Electronic transmission through a metal-molecule-metal system is calculated by employing a Green's function formalism in the scattering based scheme. Self-energy models representing the bulk and the potential bias are used to describe electron transport through the molecular system. Different self-energies can be defined by varying the partition between device and bulk regions of the metal-molecule-metal model system. In addition, the self-energies are calculated with different representations of the bulk through its Green's function. In this work, the dependence of the calculated transmission on varying the self-energy subspaces is benchmarked. The calculated transmission is monitored with respect to the different choices defining the self-energy model. In this report, we focus on one-dimensional model systems with electronic structures calculated at the density functional level of theory. PMID:17144733
International Nuclear Information System (INIS)
A new executable, identified as NJOY99.0 has been created to generate the 69-group cross-section library for the reactor lattice transport code WIMS. The new code incorporates modifications in the WIMSR module of NJOY to generate the 69-group library, which will be used for TRIGA reactor calculations. The basic evaluated nuclear data file JEF-2.2 was used to generate the 69-group cross-section library in WIMS format. The results for TRX-1, TRX-2, BAPL-1, BAPL-2, and BAPL-3 benchmarks obtained by using the generated 69-group cross-section library from JEF-2.2 were analyzed. The following integral parameters were considered for the validation of the 69-group library: finite medium effective multiplication factor (keff), Ratio of epithermal to thermal 238U captures (ρ28), Ratio of epithermal to thermal 235U fission (δ25), Ratio of 238U fission to 235U fission (δ28) and Ratio of 238U captures to 235U fissions (C*). The TRX and BAPL benchmark lattices were modeled with optimized inputs, which were suggested in the final report of the WIMS Library Update Project (WLUP) Stage-I by Ravnik. The calculated results of the integral parameters of TRX and BAPL Benchmark Lattices obtained by using the new version of code WIMSD-5B were found to be in good agreement with the experimental values. Besides, The TRX and BAPL calculation results showed that JEF-2.2 is reliable for thermal reactor calculations and validated the 69-group library, which will be used for the neutronic calculation of the TRIGA Mark-II research reactor at AERE, Savar, Dhaka, Bangladesh. (authors)
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To those Member States who have or have had significant fast reactor development programmes, it is of the utmost importance to have validated up-to-date codes and methods for fast reactor core physics analysis in support of R and D activities in the area of actinide utilization and incineration. They have recently focused on fast reactor systems for minor actinide transmutation and on cores optimized for consuming rather than breeding plutonium; the physics of the breeder reactor cycle having already been widely investigated. Plutonium burning systems may have an important role in managing plutonium stocks until the time when major programmes of self-sufficient fast breeder reactors are established. For assessing the safety of these systems it is important to determine the prediction accuracy of transient simulations and their associated reactivity coefficients. In response to Member States' expressed interest, the IAEA sponsored a Coordinated Research Project (CRP) on Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects. This CRP was started in November 1999 and at the first meeting the members of the CRP endorsed a benchmark on the BN-600 hybrid core for consideration in its first studies. Benchmark analyses of the BN-600 hybrid core were performed during the first three phases of the CRP investigating different nuclear data and levels of approximations in the calculation of, safety related reactivity effects and their influence on uncertainties in transient analysis predictions. In an additional phase of the benchmark studies experimental data was used for the validation and verification of nuclear data libraries and methods in support of the previous three phases. This report presents the results of the benchmark analyses of the hybrid UOX/MOX fuelled BN-600 reactor core. The aim of this report is to contribute to the reduction in uncertainties associated with reactivity coefficients and their influence on LMFR
ZZ ECN-BUBEBO, ECN-Petten Burnup Benchmark Book, Inventories, Afterheat
International Nuclear Information System (INIS)
Description of program or function: Contains experimental benchmarks which can be used for the validation of burnup code systems and accompanied data libraries. Although the benchmarks presented here are thoroughly described in literature, it is in many cases not straightforward to retrieve unambiguously the correct input data and corresponding results from the benchmark Descriptions. Furthermore, results which can easily be measured, are sometimes difficult to calculate because of conversions to be made. Therefore, emphasis has been put to clarify the input of the benchmarks and to present the benchmark results in such a way that they can easily be calculated and compared. For more thorough Descriptions of the benchmarks themselves, the literature referred to here should be consulted. This benchmark book is divided in 11 chapters/files containing the following in text and tabular form: chapter 1: Introduction; chapter 2: Burnup Credit Criticality Benchmark Phase 1-B; chapter 3: Yankee-Rowe Core V Fuel Inventory Study; chapter 4: H.B. Robinson Unit 2 Fuel Inventory Study; chapter 5: Turkey Point Unit 3 Fuel Inventory Study; chapter 6: Turkey Point Unit 3 Afterheat Power Study; chapter 7: Dickens Benchmark on Fission Product Energy Release of U-235; chapter 8: Dickens Benchmark on Fission Product Energy Release of Pu-239; chapter 9: Yarnell Benchmark on Decay Heat Measurements of U-233; chapter 10: Yarnell Benchmark on Decay Heat Measurements of U-235; chapter 11: Yarnell Benchmark on Decay Heat Measurements of Pu-239
Calculation of critical parameters for uranium and/or plutonium nitrate solution, 2
International Nuclear Information System (INIS)
For the purpose of criticality experiments of the nitrate solution systems composed of uranium and/or plutonium, the designs of Criticality Safety Experimental Facility(CSEF) are under-going in JAERI. In this report, by using the developed calculational code for the atomic number density of the nitrate solution of U/Pu mixture, the ratio H/Fissile has been calculated as a function of fuel concentration and acidity. Critical parameters such as infinite multiplication factor, effective multiplication factor, and the critical diameter of infinite cylinder have been evaluated with Monte Carlo code KENO-IV of JACS system. The dependence of multiplication factors on the formula for the atomic number density is described by comparing the calculated results with our formula and those with ARH-600 formula. Based on our formula, the effects of acid molality and Plutonium valence state on the criticality of nitrate solution have been discussed. (author)
Criticality calculations for the spent fuel storage pools for Etrr1 and Etrr2 reactors
International Nuclear Information System (INIS)
A criticality analysis of two spent fuel storage pools for Etrr1 and Etrr2 research reactors was performed. The multiplication factor for the pools was calculated as a function of relevant lattice physics parameters. Monte Carlo code MNCP-4A code was used in the criticality calculations. The results were compared with those given by CITATION code and results obtained formerly during the design phase of the pools with the MONK 6.3 code. Safety of the pools was confirmed. (author)
Influence of Metal Shells around Fuel Assemblies on Criticality Calculations for a Fuel Storage Pool
Babichev, L.; Khmialeuski, A.
2012-01-01
Influence of metal shells and size of cells in a fuel storage pool on the value of the effective neutron multiplication factor was studied. Monte Carlo code MCU-FREE was used in the criticality calculations. A criticality analysis of spent fuel storage pools for different degrees of packing of cells in the rack was performed.
Evaluation of approaches to calculate critical metal loads for forest ecosystems
International Nuclear Information System (INIS)
This paper evaluates approaches to calculate acceptable loads for metal deposition to forest ecosystems, distinguishing between critical loads, stand-still loads and target loads. We also evaluated the influence of including the biochemical metal cycle on the calculated loads. Differences are illustrated by examples of Cd, Cu, Pb and Zn for a deciduous forest on five major soil types in the Netherlands. Stand-still loads are generally lower than critical loads, which in turn are lower than the target loads indicating that present levels are below critical levels. Uncertainties in the calculated critical loads are mainly determined by the uncertainty in the critical limits and the chemical speciation model. Including the metal cycle has a small effect on the calculated critical loads. Results are discussed in view of the applicability of the critical load concept for metals in future protocols on the reduction in metal emissions. - Critical load methods for metals can be used to assess future risks due to metal inputs.
SPENT NUCLEAR FUEL NUMBER DENSITIES FOR MULTI-PURPOSE CANISTER CRITICALITY CALCULATIONS
International Nuclear Information System (INIS)
The purpose of this analysis is to calculate the number densities for spent nuclear fuel (SNF) to be used in criticality evaluations of the Multi-Purpose Canister (MPC) waste packages. The objective of this analysis is to provide material number density information which will be referenced by future MPC criticality design analyses, such as for those supporting the Conceptual Design Report
Energy Technology Data Exchange (ETDEWEB)
Mielke, Steven L.; Schwenke, David; Peterson, Kirk A.
2005-06-08
We present a detailed ab initio study of the effect that the Born–Oppenheimer diagonal correction (BODC) has on the saddle point properties of the H3 system and its isotopomers. Benchmark values are presented that are estimated to be within 0.1 cm–1 of the complete configuration interaction limit. We consider the basis set and correlation treatment requirements for accurate BODC calculations, and both are observed to be more favorable than for the Born–Oppenheimer energies. The BODC raises the H + H2 barrier height by 0.1532 kcal/mol and slightly narrows the barrier—with the imaginary frequency increasing by ~2%.
International Nuclear Information System (INIS)
As part of the Initial Feasibility Study of the Fast Mixed Spectrum Reactor, a series of benchmark calculations were made to determine the sensitivity of the physics analysis to differences in methods and data. Argonne National Laboratory (ANL), the Massachusetts Institute of Technology (MIT), and Oak Ridge National Laboratory (ORNL) were invited to participate with Brookhaven National Laboratory in the analysis of a FMSR model prescribed by BNL. Detailed comparisons are made including a comprehensive study on the adequacy of the fission product treatments
International Nuclear Information System (INIS)
This document presents the OKBM contribution to the analysis of a benchmark of BN-600 reactor hybrid core with simultaneous loading of uranium fuel and MOX fuel within the framework of the international IAEA Co-ordinated Research Project (CRP) on 'Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects'. The purpose of the present document is the comparison of some obtained for the Phase 2 results using the different energy groups number and the different mesh point size. The CRP participants at calculation of a benchmark used different planar mesh point sizes. The axial mesh point size was not stipulated, but the mesh sizes were specified for the desired results representation. Therefore in some cases there was possible the application of rather large axial mesh size. The discrepancy in results because the different mesh point size using should be estimated. The results of some participants were obtained using the relatively small energy groups number - 6, 9, and 12. The influence of the energy group number on value of the obtained reactivity coefficients is analyzed in case of the OKBM calculations results. Besides in case of the sodium density reactivity coefficient the OKBM used method of the choice for the few group optimal division of the energy scale is shown. The probable additional uncertainties as the consequence of the baseless group division are estimated by the comparison of the group division schemes applied by different CRP participants
Granatier, Jaroslav; Lazar, Petr; Otyepka, Michal; Hobza, Pavel
2011-01-01
The adsorption of Ag, Au, and Pd atoms on benzene, coronene, and graphene has been studied using post Hartree–Fock wave function theory (CCSD(T), MP2) and density functional theory (M06-2X, DFT-D3, PBE, vdW-DF) methods. The CCSD(T) benchmark binding energies for benzene–M (M = Pd, Au, Ag) complexes are 19.7, 4.2, and 2.3 kcal/mol, respectively. We found that the nature of binding of the three metals is different: While silver binds predominantly through dispersion interactions, the binding of...
Evaluation of approaches to calculate critical metal loads for forest soils
Vries, de W.; Groenenberg, J.E.
2009-01-01
This paper evaluates approaches to calculate acceptable loads for metal deposition to forest ecosystems, distinguishing between critical loads, stand-still loads and target loads. We also evaluated the influence of including the biochemical metal cycle on the calculated loads. Differences are illust
International Nuclear Information System (INIS)
Two methods of calculating criticality are available in the 3D generalised geometry Monte Carlo particle transport code SPARTAN (Bending and Heffer, 1975). The first is a matrix technique in which the multiplication constant and source distribution of the system under study are calculated from estimates of fission probabilities and the second a method in which the multiplication constant is inferred from estimates of changes in neutron population over a number of neutron generations. Modifications are described which have been made to the way in which these methods are used in SPARTAN in order to improve the efficiency of criticality calculations. (author)
Benchmark Evaluation of HTR-PROTEUS Pebble Bed Experimental Program
International Nuclear Information System (INIS)
Benchmark models were developed to evaluate 11 critical core configurations of the HTR-PROTEUS pebble bed experimental program. Various additional reactor physics measurements were performed as part of this program; currently only a total of 37 absorber rod worth measurements have been evaluated as acceptable benchmark experiments for Cores 4, 9, and 10. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the 235U enrichment of the fuel, impurities in the moderator pebbles, and the density and impurity content of the radial reflector. Calculations of keff with MCNP5 and ENDF/B-VII.0 neutron nuclear data are greater than the benchmark values but within 1% and also within the 3σ uncertainty, except for Core 4, which is the only randomly packed pebble configuration. Repeated calculations of keff with MCNP6.1 and ENDF/B-VII.1 are lower than the benchmark values and within 1% (~3σ) except for Cores 5 and 9, which calculate lower than the benchmark eigenvalues within 4σ. The primary difference between the two nuclear data libraries is the adjustment of the absorption cross section of graphite. Simulations of the absorber rod worth measurements are within 3σ of the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments
Mohammadi, A; Hassanzadeh, M; Gharib, M
2016-02-01
In this study, shielding calculation and criticality safety analysis were carried out for general material testing reactor (MTR) research reactors interim storage and relevant transportation cask. During these processes, three major terms were considered: source term, shielding, and criticality calculations. The Monte Carlo transport code MCNP5 was used for shielding calculation and criticality safety analysis and ORIGEN2.1 code for source term calculation. According to the results obtained, a cylindrical cask with body, top, and bottom thicknesses of 18, 13, and 13 cm, respectively, was accepted as the dual-purpose cask. Furthermore, it is shown that the total dose rates are below the normal transport criteria that meet the standards specified. PMID:26720262
International Nuclear Information System (INIS)
The purpose of this paper is to discuss the theories, techniques and computer codes that are frequently used in numerical reactor criticality and burnup calculations. It is a part of an integrated nuclear reactor calculation scheme conducted by the Reactors Department, Inshas Nuclear Research Centre. The crude part in numerical reactor criticality and burnup calculations includes the determination of neutron flux distribution which can be obtained in principle as a solution of Boltzmann transport equation. Numerical methods used for solving transport equations are discussed. Emphasis are made on numerical techniques based on multigroup diffusion theory. These numerical techniques include nodal, modal, and finite difference ones. The most commonly known computer codes utilizing these techniques are reviewed. Some of the main computer codes that have been already developed at the Reactors Department and related to numerical reactor criticality and burnup calculations have been presented
International Nuclear Information System (INIS)
The aim of this paper is to present the validation of evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 through the analysis of the integral parameters of TRX and BAPL benchmark lattices of thermal reactors for neutronics analysis of TRIGA Mark-II Research Reactor at AERE, Bangladesh. In this process, the 69-group cross-section library for lattice code WIMS was generated using the basic evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 with the help of nuclear data processing code NJOY99.0. Integral measurements on the thermal reactor lattices TRX-1, TRX-2, BAPL-UO2-1, BAPL-UO2-2 and BAPL-UO2-3 served as standard benchmarks for testing nuclear data files and have also been selected for this analysis. The integral parameters of the said lattices were calculated using the lattice transport code WIMSD-5B based on the generated 69-group cross-section library. The calculated integral parameters were compared to the measured values as well as the results of Monte Carlo Code MCNP. It was found that in most cases, the values of integral parameters show a good agreement with the experiment and MCNP results. Besides, the group constants in WIMS format for the isotopes U-235 and U-238 between two data files have been compared using WIMS library utility code WILLIE and it was found that the group constants are identical with very insignificant difference. Therefore, this analysis reflects the validation of evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 through benchmarking the integral parameters of TRX and BAPL lattices and can also be essential to implement further neutronic analysis of TRIGA Mark-II research reactor at AERE, Dhaka, Bangladesh.
Directory of Open Access Journals (Sweden)
Wonkyeong Kim
2015-01-01
Full Text Available A high-leakage core has been known to be a challenging problem not only for a two-step homogenization approach but also for a direct heterogeneous approach. In this paper the DIMPLE S06 core, which is a small high-leakage core, has been analyzed by a direct heterogeneous modeling approach and by a two-step homogenization modeling approach, using contemporary code systems developed for reactor core analysis. The focus of this work is a comprehensive comparative analysis of the conventional approaches and codes with a small core design, DIMPLE S06 critical experiment. The calculation procedure for the two approaches is explicitly presented in this paper. Comprehensive comparative analysis is performed by neutronics parameters: multiplication factor and assembly power distribution. Comparison of two-group homogenized cross sections from each lattice physics codes shows that the generated transport cross section has significant difference according to the transport approximation to treat anisotropic scattering effect. The necessity of the ADF to correct the discontinuity at the assembly interfaces is clearly presented by the flux distributions and the result of two-step approach. Finally, the two approaches show consistent results for all codes, while the comparison with the reference generated by MCNP shows significant error except for another Monte Carlo code, SERPENT2.
Ford, Donald J.
1993-01-01
Discusses benchmarking, the continuous process of measuring one's products, services, and practices against those recognized as leaders in that field to identify areas for improvement. Examines ways in which benchmarking can benefit human resources functions. (JOW)
International Nuclear Information System (INIS)
Highlights: ► To validate the SRAC2006 code system for TRIGA neutronics calculations. ► TRX and BAPL lattices are treated as standard benchmarks for this purpose. ► To compare the calculated results with experiment as well as MCNP values in this study. ► The study demonstrates a good agreement with the experiment and the MCNP results. ► Thus, this analysis reflects the validation study of the SRAC2006 code system. - Abstract: The goal of this study is to present the validation study of the SRAC2006 code system based on evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3 for neutronics analysis of TRIGA Mark-II Research Reactor at AERE, Bangladesh. This study is achieved through the analysis of integral parameters of TRX and BAPL benchmark lattices of thermal reactors. In integral measurements, the thermal reactor lattices TRX-1, TRX-2, BAPL-UO2-1, BAPL-UO2-2 and BAPL-UO2-3 are treated as standard benchmarks for validating/testing the SRAC2006 code system as well as nuclear data libraries. The integral parameters of the said lattices are calculated using the collision probability transport code PIJ of the SRAC2006 code system at room temperature 20 °C based on the above libraries. The calculated integral parameters are compared to the measured values as well as the MCNP values based on the Chinese evaluated nuclear data library CENDL-3.0. It was found that in most cases, the values of integral parameters demonstrate a good agreement with the experiment and the MCNP results. In addition, the group constants in SRAC format for TRX and BAPL lattices in fast and thermal energy range respectively are compared between the above libraries and it was found that the group constants are identical with very insignificant difference. Therefore, this analysis reflects the validation study of the SRAC2006 code system based on evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 and can also be essential to implement further neutronics calculations of
Campbell, Akiko
2016-01-01
Benchmarking is a process of comparison between performance characteristics of separate, often competing organizations intended to enable each participant to improve its own performance in the marketplace (Kay, 2007). Benchmarking sets organizations’ performance standards based on what “others” are achieving. Most widely adopted approaches are quantitative and reveal numerical performance gaps where organizations lag behind benchmarks; however, quantitative benchmarking on its own rarely yi...
International Nuclear Information System (INIS)
To evaluate calculation codes and the nuclear data in the energy region from 20 MeV to 100 MeV, intercomparisons of benchmark calculations with the MORSE-CG, modified HETC-KFA2 and MCNP4A codes were carried out for the transmission of quasi-monoenergetic neutrons generated by 43- and 68-MeV protons through iron and concrete shields. The comparisons between the calculations and the experiments show that the spectra on the axis of the neutron beam calculated by the MORSE-CG and the MCNP4A codes with the DLC-119/HILO86 and HILO86R are in good agreement with those measured. The spectra calculated for the thin shields by the modified HETC-KFA2 code agree well with those measured, while those for the thick shields are higher than measured ones. The spectra at the off-axis positions calculated by the MORSE-CG code agree well with those measured, though the modified HETC-KFA2 code greatly underestimates the measured spectra. (author)
Development of common user data model for APOLLO3 and MARBLE and application to benchmark problems
International Nuclear Information System (INIS)
A Common User Data Model, CUDM, has been developed for the purpose of benchmark calculations between APOLLO3 and MARBLE code systems. The current version of CUDM was designed for core calculation benchmark problems with 3-dimensional Cartesian, 3-D XYZ, geometry. CUDM is able to manage all input/output data such as 3-D XYZ geometry, effective macroscopic cross section, effective multiplication factor and neutron flux. In addition, visualization tools for geometry and neutron flux were included. CUDM was designed by the object-oriented technique and implemented using Python programming language. Based on the CUDM, a prototype system for a benchmark calculation, CUDM-benchmark, was also developed. The CUDM-benchmark supports input/output data conversion for IDT solver in APOLLO3, and TRITAC and SNT solvers in MARBLE. In order to evaluate pertinence of CUDM, the CUDM-benchmark was applied to benchmark problems proposed by T. Takeda, G. Chiba and I. Zmijarevic. It was verified that the CUDM-benchmark successfully reproduced the results calculated with reference input data files, and provided consistent results among all the solvers by using one common input data defined by CUDM. In addition, a detailed benchmark calculation for Chiba benchmark was performed by using the CUDM-benchmark. Chiba benchmark is a neutron transport benchmark problem for fast criticality assembly without homogenization. This benchmark problem consists of 4 core configurations which have different sodium void regions, and each core configuration is defined by more than 5,000 fuel/material cells. In this application, it was found that the results by IDT and SNT solvers agreed well with the reference results by Monte-Carlo code. In addition, model effects such as quadrature set effect, Sn order effect and mesh size effect were systematically evaluated and summarized in this report. (author)
MCNP simulation of the TRIGA Mark II benchmark experiment
International Nuclear Information System (INIS)
The complete 3D MCNP model of the TRIGA Mark II reactor is presented. It enables precise calculations of some quantities of interest in a steady-state mode of operation. Calculational results are compared to the experimental results gathered during reactor reconstruction in 1992. Since the operating conditions were well defined at that time, the experimental results can be used as a benchmark. It may be noted that this benchmark is one of very few high enrichment benchmarks available. In our simulations experimental conditions were thoroughly simulated: fuel elements and control rods were precisely modeled as well as entire core configuration and the vicinity of the core. ENDF/B-VI and ENDF/B-V libraries were used. Partial results of benchmark calculations are presented. Excellent agreement of core criticality, excess reactivity and control rod worths can be observed. (author)
Energy Technology Data Exchange (ETDEWEB)
Miller, Thomas Martin [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Celik, Cihangir [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dunn, Michael E [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wagner, John C [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McMahan, Kimberly L [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Authier, Nicolas [French Atomic Energy Commission (CEA), Centre de Valduc, Is-sur-Tille (France); Jacquet, Xavier [French Atomic Energy Commission (CEA), Centre de Valduc, Is-sur-Tille (France); Rousseau, Guillaume [French Atomic Energy Commission (CEA), Centre de Valduc, Is-sur-Tille (France); Wolff, Herve [French Atomic Energy Commission (CEA), Centre de Valduc, Is-sur-Tille (France); Savanier, Laurence [French Atomic Energy Commission (CEA), Centre de Valduc, Is-sur-Tille (France); Baclet, Nathalie [French Atomic Energy Commission (CEA), Centre de Valduc, Is-sur-Tille (France); Lee, Yi-kang [French Atomic Energy Commission (CEA), Centre de Saclay, Gif sur Yvette (France); Trama, Jean-Christophe [French Atomic Energy Commission (CEA), Centre de Saclay, Gif sur Yvette (France); Masse, Veronique [French Atomic Energy Commission (CEA), Centre de Saclay, Gif sur Yvette (France); Gagnier, Emmanuel [French Atomic Energy Commission (CEA), Centre de Saclay, Gif sur Yvette (France); Naury, Sylvie [French Atomic Energy Commission (CEA), Centre de Saclay, Gif sur Yvette (France); Blanc-Tranchant, Patrick [French Atomic Energy Commission (CEA), Centre de Saclay, Gif sur Yvette (France); Hunter, Richard [Babcock International Group (United Kingdom); Kim, Soon [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dulik, George Michael [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Reynolds, Kevin H. [Y-12 National Security Complex, Oak Ridge, TN (United States)
2015-01-01
In October 2010, a series of benchmark experiments were conducted at the French Commissariat a l'Energie Atomique et aux Energies Alternatives (CEA) Valduc SILENE facility. These experiments were a joint effort between the United States Department of Energy Nuclear Criticality Safety Program and the CEA. The purpose of these experiments was to create three benchmarks for the verification and validation of radiation transport codes and evaluated nuclear data used in the analysis of criticality accident alarm systems. This series of experiments consisted of three single-pulsed experiments with the SILENE reactor. For the first experiment, the reactor was bare (unshielded), whereas in the second and third experiments, it was shielded by lead and polyethylene, respectively. The polyethylene shield of the third experiment had a cadmium liner on its internal and external surfaces, which vertically was located near the fuel region of SILENE. During each experiment, several neutron activation foils and thermoluminescent dosimeters (TLDs) were placed around the reactor. Nearly half of the foils and TLDs had additional high-density magnetite concrete, high-density barite concrete, standard concrete, and/or BoroBond shields. CEA Saclay provided all the concrete, and the US Y-12 National Security Complex provided the BoroBond. Measurement data from the experiments were published at the 2011 International Conference on Nuclear Criticality (ICNC 2011) and the 2013 Nuclear Criticality Safety Division (NCSD 2013) topical meeting. Preliminary computational results for the first experiment were presented in the ICNC 2011 paper, which showed poor agreement between the computational results and the measured values of the foils shielded by concrete. Recently the hydrogen content, boron content, and density of these concrete shields were further investigated within the constraints of the previously available data. New computational results for the first experiment are now available
International Nuclear Information System (INIS)
All of the three exercises of the Organization for Economic Cooperation and Development/Nuclear Regulatory Commission pressurized water reactor main steam line break (PWR MSLB) benchmark were calculated at VTT, the Technical Research Centre of Finland. For the first exercise, the plant simulation with point-kinetic neutronics, the thermal-hydraulics code SMABRE was used. The second exercise was calculated with the three-dimensional reactor dynamics code TRAB-3D, and the third exercise with the combination TRAB-3D/SMABRE. VTT has over ten years' experience of coupling neutronic and thermal-hydraulic codes, but this benchmark was the first time these two codes, both developed at VTT, were coupled together. The coupled code system is fast and efficient; the total computation time of the 100-s transient in the third exercise was 16 min on a modern UNIX workstation. The results of all the exercises are similar to those of the other participants. In order to demonstrate the effect of secondary circuit modeling on the results, three different cases were calculated. In case 1 there is no phase separation in the steam lines and no flow reversal in the aspirator. In case 2 the flow reversal in the aspirator is allowed, but there is no phase separation in the steam lines. Finally, in case 3 the drift-flux model is used for the phase separation in the steam lines, but the aspirator flow reversal is not allowed. With these two modeling variations, it is possible to cover a remarkably broad range of results. The maximum power level reached after the reactor trip varies from 534 to 904 MW, the range of the time of the power maximum being close to 30 s. Compared to the total calculated transient time of 100 s, the effect of the secondary side modeling is extremely important
VVER-related burnup credit calculations
International Nuclear Information System (INIS)
The calculations related to a VVER burnup credit calculational benchmark proposed to the Eastern and Central European research community in collaboration with the OECD/NEA/NSC Burnup Credit Criticality Benchmark Working Group (working under WPNCS - Working Party on Nuclear Criticality Safety) are described. The results of a three-year effort by analysts from the Czech Republic, Finland, Germany, Hungary, Russia, Slovakia and the United Kingdom are summarized and commented on. (author)
International Nuclear Information System (INIS)
Monte Carlo neutronics calculations can estimate accurate nuclear parameters from continuous energy nuclear library and detailed geometry. The continuous energy nuclear library for Monte Carlo simulations can be generated from several evaluated nuclear data files - ENDF/B-VI.8, JENDL-3.3, JEFF-3.0, etc . by NJOY 99. The objective of this paper is to quantify effects of evaluated nuclear data files on nuclear parameters estimated by Monte Carlo calculations for various critical experiment problems. In this study, Monte Carlo calculations are conducted by the MCCARD which is designed exclusively for the neutron transport calculation
Final evaluation of the CB3+burnup credit benchmark addition
International Nuclear Information System (INIS)
In 1966 a series of benchmarks focused on the application of burnup credit in WWER spent fuel management system was launched by L.Markova (1). The four phases of the proposed benchmark series corresponded to the phases of the Burnup Credit Criticality Benchmark organised by the OECD/NEA.These phases referred as CB1, CB2, CB3 and CB4 benchmarks were designed to investigate the main features of burnup credit in WWER spent fuel management systems. In the CB1 step, the multiplication factor of an infinite array of spent fuel rods was calculated taking the burnup, cooling time and different group of nuclides as parameters. The fuel compositions was given in the benchmark specification (Authors)
Energy Technology Data Exchange (ETDEWEB)
Ohta, Masayuki, E-mail: ohta.masayuki@jaea.go.jp [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan); Takakura, Kosuke; Ochiai, Kentaro; Sato, Satoshi; Konno, Chikara [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan)
2013-10-15
In order to examine a basic performance of the TRIPOLI code, two types of analyses were carried out with TRIPOLI-4.4 and MCNP5-1.40; one is a simple model calculation and the other is an analysis of iron fusion neutronics experiments with DT neutrons at the Fusion Neutronics Source (FNS) facility in Japan Atomic Energy Agency (JAEA). In the simple model calculation, we adopted a sphere of 0.5 m in radius with a 20 MeV neutron source in the center and calculated leakage neutron spectra from the sphere. We also analyzed in situ and Time-of-Flight (TOF) experiments for iron at JAEA/FNS. For the in situ experiment, neutron spectra and reaction rates for dosimetry reactions were calculated for several points inside the assembly. For the TOF experiment, angular neutron leakage spectra from the assembly were calculated. Results with TRIPOLI were comparable to those with MCNP in most calculations, but a difference between TRIPOLI and MCNP calculation results, probably caused by inadequate treatment of inelastic scattering data in TRIPOLI, appears in some calculations.
Calculation of criticality parameters for uranium and/or plutonium nitrate solutions, 1
International Nuclear Information System (INIS)
For the purpose of computing criticality parameters of the solutions with uranium and plutonium, the atomic number density was formulated and programmed in the computer. The amount of solvent in the solution was calculated from solute concentration and density of the solution. The numerical expression of the solution density was based on the literature data for aqueous nitrate solution, and on theoretical consideration for 30% TBP-n. dodecane solution. The calculated values of solution density were discussed compared with that in ARH-600, the criticality handbook in the United States. (author)
Larsson, Cecilia
2010-01-01
A few years ago Westinghouse started the development of a new method for criticality calculations for spent nuclear fuel storage pools called “PHOENIX-to–MCNP” (PHX2MCNP). PHX2MCNP transfers burn-up data from the code PHOENIX to use in MCNP in order to calculate the criticality. This thesis describes a work with the purpose to further validate the new method first by validating the software MCNP5 at higher water temperatures than room temperature and, in a second step, continue the developmen...
Calculation of critical experiment parameters for the High Flux Isotope Reactor
International Nuclear Information System (INIS)
Six critical experiments were performed shortly before the initial ascension to power of the High Flux Isotope Reactor (HFIR). Critical configurations were determined at various control rod positions by varying the soluble boron content in the light water coolant. Calculated k-effective was 2% high at beginning-of-life (BOL) typical conditions, but was 1.0 at end-of-life (EOL) typical conditions. Axially averaged power distributions for a given radial location were frequently within experimental error. At specific r,z locations with the core, the calculated power densities were significantly different from the experimentally derived values. A reassessment of the foil activation data seems desirable
Monte Carlo calculations of the REBUS critical experiment for validation of burnup credit
International Nuclear Information System (INIS)
The application of burnup credit (BUC) to criticality safety analysis for Spent Nuclear Fuel (SNF) configurations requires the implementation of both estimation of the SNF composition with the aid of depletion calculation tools and estimation of the SNF reactivity with the aid of criticality calculation tools. Amongst the several experimental programs dedicated to the validation of both calculation tools, REBUS is distinguished by a combination of chemical analysis and critical experiment. In addition to detailed assays of irradiated fuel, the reactivity worth of the fuel rods under investigation is measured both before and after irradiation. Since a whole bundle of fuel rods is used in the experiment, the change in reactivity is significant enough to be observable by Monte Carlo calculations. Thus, the calculation tools which see the most widespread use in SNF critical safety applications can be validated directly. Apart from the effective neutron multiplication factor keff, REBUS also provides measurements of the flux and fission rate distributions. While the program comprises investigation of commercial UO2 fuel rods and mixed oxide (MOX) fuel from a research reactor, the presentation will focus on the commercial UO2 fuel with an overview of the experimental setup and first results from the analysis. (author)
Experimental and computational benchmark tests
International Nuclear Information System (INIS)
A program involving principally NIST, LANL, and ORNL has been in progress for about four years now to establish a series of benchmark measurements and calculations related to the moderation and leakage of 252Cf neutrons from a source surrounded by spherical aqueous moderators of various thicknesses and compositions. The motivation for these studies comes from problems in criticality calculations concerning arrays of multiplying components, where the leakage from one component acts as a source for the other components. This talk compares experimental and calculated values for the fission rates of four nuclides - 235U, 239Pu, 238U, and 237Np - in the leakage spectrum from moderator spheres of diameters 76.2 mm, 101.6 mm, and 127.0 mm, with either pure water or enriched B-10 solutions as the moderator. Very detailed Monte Carlo calculations were done with the MCNP code, using a open-quotes light waterclose quotes S(α,β) scattering kernel
TRIGA Mark II benchmark experiment
International Nuclear Information System (INIS)
The experimental results of startup tests after reconstruction and modification of the TRIGA Mark II reactor in Ljubljana are presented. The experiments were performed with a completely fresh, compact, and uniform core. The operating conditions were well defined and controlled, so that the results can be used as a benchmark test case for TRIGA reactor calculations. Both steady-state and pulse mode operation were tested. In this paper, the following steady-state experiments are treated: critical core and excess reactivity, control rod worths, fuel element reactivity worth distribution, fuel temperature distribution, and fuel temperature reactivity coefficient
International Nuclear Information System (INIS)
The calculations performed for the Almaraz Unit 2 PWR using the code packages of the Atomic Energy Corporation of South Africa Ltd. are summarized. These calculations were done as part of the IAEA Coordinated Research Programme on In-Core Fuel Management Code Package Validation for LWRs. A brief description of the one-dimensional cross section generation package as well as of the Level II (scoping type) global core calculational package which was used is given. Detailed results are presented in several appendices. 29 figs., 20 tabs., 10 refs
Criticality calculations of the high-density spent fuel storage of the water-pool type
International Nuclear Information System (INIS)
High-density spent fuel racks increase the capacity of spent fuel pit for several times. The minimum cell pitch is limited mainly by the allowed multiplication factor of the system. Detailed criticality calculations have to be performed in order to determine the minimum allowable cell pitch. In this report are given the reactivity calculations of the spent fuel pit in dependence of cell pitch, cooling water density, boration and temperature.(author)
Assessment of formulas for calculating critical concentration by the agar diffusion method.
Drugeon, H.B.; Juvin, M E; Caillon, J.; Courtieu, A L
1987-01-01
The critical concentration of antibiotic was calculated by using the agar diffusion method with disks containing different charges of antibiotic. It is currently possible to use different calculation formulas (based on Fick's law) devised by Cooper and Woodman (the best known) and by Vesterdal. The results obtained with the formulas were compared with the MIC results (obtained by the agar dilution method). A total of 91 strains and two cephalosporins (cefotaxime and ceftriaxone) were studied....
Calculation of Henry constant on the base of critical parameters of adsorbable gas
International Nuclear Information System (INIS)
Calculation of Henry constant using correlation between critical parameters Psub(c), Tsub(c) and adsorption energy, determined by the value of internal pressure in molecular field of adsorbent, has been made. The calculated Henry constants for Ar, Kr and Xe, adsorbed by MoS2 and zeolite NaX, are compared with the experimental ones. The state of the molecules adsorbed is evaluated
Criticality Safety Code Validation with LWBR’s SB Cores
Energy Technology Data Exchange (ETDEWEB)
Putman, Valerie Lee
2003-01-01
The first set of critical experiments from the Shippingport Light Water Breeder Reactor Program included eight, simple geometry critical cores built with 233UO2-ZrO2, 235UO2-ZrO2, ThO2, and ThO2-233UO2 nuclear materials. These cores are evaluated, described, and modeled to provide benchmarks and validation information for INEEL criticality safety calculation methodology. In addition to consistency with INEEL methodology, benchmark development and nuclear data are consistent with International Criticality Safety Benchmark Evaluation Project methodology.Section 1 of this report introduces the experiments and the reason they are useful for validating some INEEL criticality safety calculations. Section 2 provides detailed experiment descriptions based on currently available experiment reports. Section 3 identifies criticality safety validation requirement sources and summarizes requirements that most affect this report. Section 4 identifies relevant hand calculation and computer code calculation methodologies used in the experiment evaluation, benchmark development, and validation calculations. Section 5 provides a detailed experiment evaluation. This section identifies resolutions for currently unavailable and discrepant information. Section 5 also reports calculated experiment uncertainty effects. Section 6 describes the developed benchmarks. Section 6 includes calculated sensitivities to various benchmark features and parameters. Section 7 summarizes validation results. Appendices describe various assumptions and their bases, list experimenter calculations results for items that were independently calculated for this validation work, report other information gathered and developed by SCIENTEC personnel while evaluating these same experiments, and list benchmark sample input and miscellaneous supplementary data.
Weterings, Peter J J M; Loftus, Christine; Lewandowski, Thomas A
2016-08-22
Potential adverse effects of chemical substances on thyroid function are usually examined by measuring serum levels of thyroid-related hormones. Instead, recent risk assessments for thyroid-active chemicals have focussed on iodine uptake inhibition, an upstream event that by itself is not necessarily adverse. Establishing the extent of uptake inhibition that can be considered de minimis, the chosen benchmark response (BMR), is therefore critical. The BMR values selected by two international advisory bodies were 5% and 50%, a difference that had correspondingly large impacts on the estimated risks and health-based guidance values that were established. Potential treatment-related inhibition of thyroidal iodine uptake is usually determined by comparing thyroidal uptake of radioactive iodine (RAIU) during treatment with a single pre-treatment RAIU value. In the present study it is demonstrated that the physiological intra-individual variation in iodine uptake is much larger than 5%. Consequently, in-treatment RAIU values, expressed as a percentage of the pre-treatment value, have an inherent variation, that needs to be considered when conducting dose-response analyses. Based on statistical and biological considerations, a BMR of 20% is proposed for benchmark dose analysis of human thyroidal iodine uptake data, to take the inherent variation in relative RAIU data into account. Implications for the tolerated daily intakes for perchlorate and chlorate, recently established by the European Food Safety Authority (EFSA), are discussed. PMID:27268963
Criticality coefficient calculation for a small PWR using Monte Carlo Transport Code
Energy Technology Data Exchange (ETDEWEB)
Trombetta, Debora M.; Su, Jian, E-mail: dtrombetta@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Chirayath, Sunil S., E-mail: sunilsc@tamu.edu [Department of Nuclear Engineering and Nuclear Security Science and Policy Institute, Texas A and M University, TX (United States)
2015-07-01
Computational models of reactors are increasingly used to predict nuclear reactor physics parameters responsible for reactivity changes which could lead to accidents and losses. In this work, preliminary results for criticality coefficient calculation using the Monte Carlo transport code MCNPX were presented for a small PWR. The computational modeling developed consists of the core with fuel elements, radial reflectors, and control rods inside a pressure vessel. Three different geometries were simulated, a single fuel pin, a fuel assembly and the core, with the aim to compare the criticality coefficients among themselves.The criticality coefficients calculated were: Doppler Temperature Coefficient, Coolant Temperature Coefficient, Coolant Void Coefficient, Power Coefficient, and Control Rod Worth. The coefficient values calculated by the MCNP code were compared with literature results, showing good agreement with reference data, which validate the computational model developed and allow it to be used to perform more complex studies. Criticality Coefficient values for the three simulations done had little discrepancy for almost all coefficients investigated, the only exception was the Power Coefficient. Preliminary results presented show that simple modelling as a fuel assembly can describe changes at almost all the criticality coefficients, avoiding the need of a complex core simulation. (author)
The critical behavior of hadronic matter: Comparison of lattice and bootstrap model calculations
Turko, L.
2015-01-01
Statistical bootstrap model and the related concept of the limiting temperature begun the discussion about phase transitions in the hadronic matter. This was also the origin of the quark-gluon plazma concept. We discuss here to which extend lattice studies of QCD critical behavior at non-zero chemical potential are compatible with the statistical bootstrap model calculations.
CB2 result evaluation (VVER-440 burnup credit benchmark)
International Nuclear Information System (INIS)
The second portion of the four-piece international calculational benchmark on the VVER burnup credit (CB2) prepared in the collaboration with the OECD/NEA/NSC Burnup Credit Criticality Benchmarks Working Group and proposed to the AER research community has been evaluated. The evaluated results of calculations performed by analysts from Cuba, the Czech Republic, Finland, Germany, Russia, Slovakia and the United Kingdom are presented. The goal of this study is to compare isotopic concentrations calculated by the participants using various codes and libraries for depletion of the VVER-440 fuel pin cell. No measured values were available for the comparison. (author)
Mielke, Steven L.; Schwenke, David W.; Peterson, Kirk A.
2005-06-01
We present a detailed ab initio study of the effect that the Born-Oppenheimer diagonal correction (BODC) has on the saddle-point properties of the H3 system and its isotopomers. Benchmark values are presented that are estimated to be within 0.1cm-1 of the complete configuration-interaction limit. We consider the basis set and correlation treatment requirements for accurate BODC calculations, and both are observed to be more favorable than for the Born-Oppenheimer energies. The BODC raises the H+H2 barrier height by 0.1532kcal/mol and slightly narrows the barrier—with the imaginary frequency increasing by ˜2%.
International Nuclear Information System (INIS)
A series of tests investigating dynamic pulse buckling of a cylindrical shell under axial impact is compared to several 2D and 3D finite element simulations of the event. The purpose of the work is to investigate the performance of various analysis codes and element types on a problem which is applicable to radioactive material transport packages, and ultimately to develop a benchmark problem to qualify finite element analysis codes for the transport package design industry. During the pulse buckling tests, a buckle formed at each end of the cylinder, and one of the two buckles became unstable and collapsed. Numerical simulations of the test were performed using PRONTO, a Sandia developed transient dynamics analysis code, and ABAQUS/Explicit with both shell and continuum elements. The calculations are compared to the tests with respect to deformed shape and impact load history
International Nuclear Information System (INIS)
Validation of criticality calculations using SCALA was performed using data presented in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. This paper contains the results of statistical analysis of discrepancies between calculated and benchmark-model keff and conclusions about uncertainties of criticality prediction for different types of multiplying systems following from this analysis. (authors)
Verification and validation benchmarks.
Energy Technology Data Exchange (ETDEWEB)
Oberkampf, William Louis; Trucano, Timothy Guy
2007-02-01
Verification and validation (V&V) are the primary means to assess the accuracy and reliability of computational simulations. V&V methods and procedures have fundamentally improved the credibility of simulations in several high-consequence fields, such as nuclear reactor safety, underground nuclear waste storage, and nuclear weapon safety. Although the terminology is not uniform across engineering disciplines, code verification deals with assessing the reliability of the software coding, and solution verification deals with assessing the numerical accuracy of the solution to a computational model. Validation addresses the physics modeling accuracy of a computational simulation by comparing the computational results with experimental data. Code verification benchmarks and validation benchmarks have been constructed for a number of years in every field of computational simulation. However, no comprehensive guidelines have been proposed for the construction and use of V&V benchmarks. For example, the field of nuclear reactor safety has not focused on code verification benchmarks, but it has placed great emphasis on developing validation benchmarks. Many of these validation benchmarks are closely related to the operations of actual reactors at near-safety-critical conditions, as opposed to being more fundamental-physics benchmarks. This paper presents recommendations for the effective design and use of code verification benchmarks based on manufactured solutions, classical analytical solutions, and highly accurate numerical solutions. In addition, this paper presents recommendations for the design and use of validation benchmarks, highlighting the careful design of building-block experiments, the estimation of experimental measurement uncertainty for both inputs and outputs to the code, validation metrics, and the role of model calibration in validation. It is argued that the understanding of predictive capability of a computational model is built on the level of
Shielding benchmark problems, (2)
International Nuclear Information System (INIS)
Shielding benchmark problems prepared by Working Group of Assessment of Shielding Experiments in the Research Committee on Shielding Design in the Atomic Energy Society of Japan were compiled by Shielding Laboratory in Japan Atomic Energy Research Institute. Fourteen shielding benchmark problems are presented newly in addition to twenty-one problems proposed already, for evaluating the calculational algorithm and accuracy of computer codes based on discrete ordinates method and Monte Carlo method and for evaluating the nuclear data used in codes. The present benchmark problems are principally for investigating the backscattering and the streaming of neutrons and gamma rays in two- and three-dimensional configurations. (author)
Sakamoto, Y
2002-01-01
In the prevention of nuclear disaster, there needs the information on the dose equivalent rate distribution inside and outside the site, and energy spectra. The three dimensional radiation transport calculation code is a useful tool for the site specific detailed analysis with the consideration of facility structures. It is important in the prediction of individual doses in the future countermeasure that the reliability of the evaluation methods of dose equivalent rate distribution and energy spectra by using of Monte Carlo radiation transport calculation code, and the factors which influence the dose equivalent rate distribution outside the site are confirmed. The reliability of radiation transport calculation code and the influence factors of dose equivalent rate distribution were examined through the analyses of critical accident at JCO's uranium processing plant occurred on September 30, 1999. The radiation transport calculations including the burn-up calculations were done by using of the structural info...
Enrico Fermi Fast Reactor Spent Nuclear Fuel Criticality Calculations: Degraded Mode
International Nuclear Information System (INIS)
The objective of this calculation is to characterize the nuclear criticality safety concerns associated with the codisposal of the Department of Energy's (DOE) Enrico Fermi (EF) Spent Nuclear Fuel (SNF) in a 5-Defense High-Level Waste (5-DHLW) Waste Package (WP) and placed in a Monitored Geologic Repository (MGR). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (keff) for the degraded mode internal configurations of the codisposal WP. The results of this calculation and those of Ref. 8 will be used to evaluate criticality issues and support the analysis that will be performed to demonstrate the viability of the codisposal concept for the Monitored Geologic Repository
Li, Xiaoyu; Xu, Xuefei; You, Xiaoqing; Truhlar, Donald G
2016-06-16
It is important to determine an appropriate computational method for obtaining accurate thermochemical properties of large biodiesel molecules such as methyl linolenate. In this study, we use Kohn-Sham density functional theory (DFT) and coupled cluster theory to calculate bond dissociation enthalpies (BDEs) of seven fragment molecules of methyl linolenate, in particular, propene, methyl formate, cis-3-hexene, 1,4-pentadiene, 1-pentene, butane, and methyl butanoate. The results are compared to BDEs obtained from experiments and to Oyeyemi et al.'s multireference averaged coupled pair functional (MRACPF2) calculations. We found that with extrapolation to the complete basis set (CBS) limit, the BDEs derived from coupled cluster calculations with single, double, and triple excitations (CCSDT) and from CCSDT with a perturbative treatment of connected quadruple excitations, CCSDT(2)Q/CBS, are closer to the available experimental values than those obtained by MRACPF2 for propene and methyl formate. The CCSDT/CBS calculations were chosen as the reference for validating the DFT methods. Among the density functionals, we found that M08-HX has the best performance with a mean unsigned deviation (MUD) from CCSDT/CBS of only 1.0 kcal/mol, whereas the much more expensive MRACPF2 has an MUD of 1.1 kcal/mol. We then used the most successfully validated density functionals to calculate the BDEs of methyl linolenate and compared the results with the MRACPF2 BDEs. The present study identifies several Kohn-Sham exchange-correlation functionals that should be useful for modeling ester combustion, especially the M08-HX, M06-2X, M05-2X, M08-SO, and MPWB1K global-hybrid meta functionals, the M11 and MN12-SX range-separated-hybrid meta functionals, the ωB97 range-separated hybrid gradient approximation functional, and the SOGGA11-X global-hybrid gradient approximation functional. PMID:27191950
Boldyreva, Anna
2014-01-01
This bachelor's thesis is focused on financial benchmarking of TULIPA PRAHA s.r.o. The aim of this work is to evaluate financial situation of the company, identify its strengths and weaknesses and to find out how efficient is the performance of this company in comparison with top companies within the same field by using INFA benchmarking diagnostic system of financial indicators. The theoretical part includes the characteristic of financial analysis, which financial benchmarking is based on a...
Energy Technology Data Exchange (ETDEWEB)
Koponen, B.L.; Hampel, V.E.
1982-10-21
This compilation contains 688 complete summaries of papers on nuclear criticality safety as presented at meetings of the American Nuclear Society (ANS). The selected papers contain criticality parameters for fissile materials derived from experiments and calculations, as well as criticality safety analyses for fissile material processing, transport, and storage. The compilation was developed as a component of the Nuclear Criticality Information System (NCIS) now under development at the Lawrence Livermore National Laboratory. The compilation is presented in two volumes: Volume 1 contains a directory to the ANS Transaction volume and page number where each summary was originally published, the author concordance, and the subject concordance derived from the keyphrases in titles. Volume 2 contains - in chronological order - the full-text summaries, reproduced here by permission of the American Nuclear Society from their Transactions, volumes 1-41.
International Nuclear Information System (INIS)
This compilation contains 688 complete summaries of papers on nuclear criticality safety as presented at meetings of the American Nuclear Society (ANS). The selected papers contain criticality parameters for fissile materials derived from experiments and calculations, as well as criticality safety analyses for fissile material processing, transport, and storage. The compilation was developed as a component of the Nuclear Criticality Information System (NCIS) now under development at the Lawrence Livermore National Laboratory. The compilation is presented in two volumes: Volume 1 contains a directory to the ANS Transaction volume and page number where each summary was originally published, the author concordance, and the subject concordance derived from the keyphrases in titles. Volume 2 contains - in chronological order - the full-text summaries, reproduced here by permission of the American Nuclear Society from their Transactions, volumes 1-41
International Nuclear Information System (INIS)
Critical masses of the 20%-enriched graphite moderated SHE assembly were measured in seven cores of different core configurations. The C/235U atomic ratio in the core region is from 2226 to 6628 in changing combination of fuel and graphite disks. The seven cores are cylindrical, surrounded with a radial graphite reflector, with the exceptions of one also loaded with slight thorium oxide and another with an annular core having inner and outer radial graphite reflectors. Calculation of the critical masses was made with 34-group one-dimensional diffusion computer code TUD, assuming that the core region is homogeneous for neutron transport calculation, disregarding the semihomogeneous as it is. The group constants for thermal neutrons were determined by an ordinary procedure considering the spatial dependence of neutron thermalization in use of Young and Koppel's scattering kernel (ENDF/A) for graphite, with computer codes PIXSE and TUD. The absorption and fission cross sections of 235U, 238U and C are from ENDF/B-III file. The epithermal energy region from 10 MeV to 0.683 eV is in 33 groups. Epithermal group constants were calculated with FAXSE code from GAM-I file. The resonance absorption and fission integrals of 235U and 238U were calculated with RICM code. Agreement in the critical masses between measurement and calculation is good, the discrepancies being 2.6% on average and 4.4% in maximum. In conclusion, the calculational technique used is adequate for predicting critical masses of the graphite moderated cores, loaded with 20%-enriched uranium, such as the present semi-homogeneous cores of SHE-5 -- SHE-9. (auth.)
Accelerator shielding benchmark problems
International Nuclear Information System (INIS)
Accelerator shielding benchmark problems prepared by Working Group of Accelerator Shielding in the Research Committee on Radiation Behavior in the Atomic Energy Society of Japan were compiled by Radiation Safety Control Center of National Laboratory for High Energy Physics. Twenty-five accelerator shielding benchmark problems are presented for evaluating the calculational algorithm, the accuracy of computer codes and the nuclear data used in codes. (author)
Accelerator shielding benchmark problems
Energy Technology Data Exchange (ETDEWEB)
Hirayama, H.; Ban, S.; Nakamura, T. [and others
1993-01-01
Accelerator shielding benchmark problems prepared by Working Group of Accelerator Shielding in the Research Committee on Radiation Behavior in the Atomic Energy Society of Japan were compiled by Radiation Safety Control Center of National Laboratory for High Energy Physics. Twenty-five accelerator shielding benchmark problems are presented for evaluating the calculational algorithm, the accuracy of computer codes and the nuclear data used in codes. (author).
A hybrid Monte Carlo and response matrix Monte Carlo method in criticality calculation
International Nuclear Information System (INIS)
Full core calculations are very useful and important in reactor physics analysis, especially in computing the full core power distributions, optimizing the refueling strategies and analyzing the depletion of fuels. To reduce the computing time and accelerate the convergence, a method named Response Matrix Monte Carlo (RMMC) method based on analog Monte Carlo simulation was used to calculate the fixed source neutron transport problems in repeated structures. To make more accurate calculations, we put forward the RMMC method based on non-analog Monte Carlo simulation and investigate the way to use RMMC method in criticality calculations. Then a new hybrid RMMC and MC (RMMC+MC) method is put forward to solve the criticality problems with combined repeated and flexible geometries. This new RMMC+MC method, having the advantages of both MC method and RMMC method, can not only increase the efficiency of calculations, also simulate more complex geometries rather than repeated structures. Several 1-D numerical problems are constructed to test the new RMMC and RMMC+MC method. The results show that RMMC method and RMMC+MC method can efficiently reduce the computing time and variations in the calculations. Finally, the future research directions are mentioned and discussed at the end of this paper to make RMMC method and RMMC+MC method more powerful. (authors)
International Nuclear Information System (INIS)
A methodology for criticality safety analysis of spent fuel casks with possibilities for burnup credit implementation is presented. This methodology includes the world well-known and applied program systems: NESSEL-NUKO for depletion and SCALE-4.4 for criticality calculations. The abilities of this methodology to analyze storage and transportation casks with different type of spent fuel are demonstrated on the base of various tests. The depletion calculations have been carried out for the power reactors (WWER-440 and WWER-1000) and the research reactor IRT-2000 (C-36) fuel assemblies. The criticality calculation models have been developed on the basis of real fuel casks, designed by the leading international companies (for WWER-440 and WWER-1000 spent fuel assemblies), as well as for real a WWER-440 storage cask, applied at the 'Kozloduy' NPP. The results obtained show that the criticality safety criterion Keff less than 0.95 is satisfied for both: fresh and spent fuel. Besides the implementation of burnup credit allows to account for the reduced reactivity of spent fuel and to evaluate the conservatism of the fresh fuel assumption. (author)
International Nuclear Information System (INIS)
A series of criticality benchmark experiments with a small LWR-type core, reflected by 30 cm of lead, was defined jointly by SEC (Service d'Etude de Criticite), Fontenay-aux-Roses, and SRD (Safety and Reliability Directorate). These experiments are very representative of the reflecting effect of lead, since the contribution of the lead to the reactivity was assessed as about 30% in Δ K. The experiments were carried out by SRSC (Service de Recherche en Surete et Criticite), Valduc, in December 1983 in the sub-critical facility called APPARATUS B. In addition, they confirmed and measured the effect on reactivity of a water gap between the core and the lead reflector; with a water gap of less than 1 cm, the reactivity can be greater than that of the core directly reflected the lead or by over 20 cm of water. The experimental results were to a large extent made use of by SRD with the aid of the MONK Monte Carlo code and to some extent by SEC with the aid of the MORET Monte Carlo Code. All the results obtained are presented in the summary tables. These experiments allowed to compare the different libraries of cross sections available
Critical point calculation for binary mixtures of symmetric non-additive hard disks
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W.T. Gózdz
2016-02-01
Full Text Available We have calculated the values of critical packing fractions for the mixtures of symmetric non-additive hard disks. An interesting feature of the model is the fact that the internal energy is zero and the phase transitions are entropically driven. A cluster algorithm for Monte Carlo simulations in a semigrand ensemble was used. The finite size scaling analysis was employed to compute the critical packing fractions for infinite systems with high accuracy for a range of non-additivity parameters wider than in the previous studies.
International Nuclear Information System (INIS)
A real-space renormalization group approach for the bond percolation problem in a square lattice with first- and second- neighbour bonds is proposed. The respective probabilities are treated, as independent variables. Two types of cells are constructed. In one of them the lattice is considered as two interpenetrating sublattices, first-neighbour bonds playing the role of intersublattice links. This allows the calculation of both critical exponents ν and γ, without resorting to any external field. Values found for the critical indices are in good agreement with data available in the literature. The phase diagram in parameter space is also obtained in each case. (Author)
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Dielectronic satellite spectra of helium-like argon, recorded with a high-resolution X-ray crystal spectrometer at the National Spherical Torus Experiment, were found to be inconsistent with existing predictions resulting in unacceptable values for the power balance and suggesting the unlikely existence of non-Maxwellian electron energy distributions. These problems were resolved with calculations from a new atomic code. It is now possible to perform reliable electron temperature measurements and to eliminate the uncertainties associated with determinations of non-Maxwellian distributions
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The measurement of the stationarity of Monte Carlo fission source distributions in keff calculations plays a central role in the ability to discriminate between fake and 'true' convergence (in the case of a high dominant ratio or in case of loosely coupled systems). Recent theoretical developments have been made in the study of source convergence diagnostics, using Shannon entropy. We will first recall those results, and we will then generalize them using the expression of Boltzmann entropy, highlighting the gain in terms of the various physical problems that we can treat. Finally we will present the results of several OECD/NEA benchmarks using the Tripoli-4 Monte Carlo code, enhanced with this new criterion. (authors)
Alize 3 - first critical experiment for the franco-german high flux reactor - calculations
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The results of experiments in the light water cooled D2O reflected critical assembly ALIZE III have been compared to calculations. A diffusion model was used with 3 fast and epithermal groups and two overlapping thermal groups, which leads to good agreement of calculated and measured power maps, even in the case of strong variations of the neutron spectrum in the core. The difference of calculated and measured keff was smaller than 0.5 per cent δk/k. Calculations of void and structure material coefficients of the reactivity of 'black' rods in the reflector, of spectrum variations (Cd-ratio, Pu-U-ratio) and to the delayed photoneutron fraction in the D2O reflector were made. Measurements of the influence of beam tubes on reactivity and flux distribution in the reflector were interpreted with regard to an optimum beam tube arrangement for the Franco- German High Flux Reactor. (author)
DEFF Research Database (Denmark)
Hougaard, Jens Leth; Tvede, Mich
2002-01-01
Within a production theoretic framework, this paper considers an axiomatic approach to benchmark selection. It is shown that two simple and weak axioms; efficiency and comprehensive monotonicity characterize a natural family of benchmarks which typically becomes unique. Further axioms are added in...... order to obtain a unique selection...
DEFF Research Database (Denmark)
Lawson, Lartey; Nielsen, Kurt
2005-01-01
We discuss individual learning by interactive benchmarking using stochastic frontier models. The interactions allow the user to tailor the performance evaluation to preferences and explore alternative improvement strategies by selecting and searching the different frontiers using directional...... suggested benchmarking tool. The study investigates how different characteristics on dairy farms influences the technical efficiency....
International Nuclear Information System (INIS)
Recently, there has been a new word added to our vocabulary - benchmarking. Because of benchmarking, our colleagues travel to power plants all around the world and guests from the European power plants visit us. We asked Marek Niznansky from the Nuclear Safety Department in Jaslovske Bohunice NPP to explain us this term. (author)
Bauschlicher, Charles W., Jr.; Langhoff, Stephen R.
1987-01-01
Full configuration interaction (CI) calculations on the ground states of N2, NO, and O2 using a DZP Gaussian basis are compared with single-reference SDCI and coupled pair approaches (CPF), as well as with CASSCF multireference CI approaches. The CASSCF/MRCI technique is found to describe multiple bonds as well as single bonds. Although the coupled pair functional approach gave chemical accuracy (1 kcal/mol) for bonds involving hydrogen, larger errors occur in the CPF approach for the multiple bonded systems considered here. CI studies on the 1Sigma(g +) state of N2, including all single, double, triple, and quadruple excitations show that triple excitations are very important for the multiple bond case, and accounts for most of the deficiency in the coupled pair functional methods.
Criticality calculation and control rods for the Westinghouse Reactor Evaluation Center facility
International Nuclear Information System (INIS)
This work evaluates two clean critical cores by WIMS-TRACA/CITATION codes calculation, 4 energy groups and bi dimension geometry. The first core is composed of U O2 with a clad of stainless steel and 20 absorbers Ag-In-Cd absorbers rods, the second is composed of U O2 with a clad of Zircaloy and 12 B4 C absorbers rods. (author)
Improvement of the skeleton tables for calculation of the critical heat load
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Paper presents analysis of drawbacks of the skeleton tables of the critical heat flows applied in calculated heat and hydraulic codes. Paper demonstrates the necessity to take account of specific nature of mechanisms of dryout crisis, of boiling crisis at slow mass rates and the range of small underheatings up to temperature of saturation. Attention is drawn to necessity of detailed account of the natural limitations of the application field of the skeleton tables
Benchmark Evaluation of the NRAD Reactor LEU Core Startup Measurements
Energy Technology Data Exchange (ETDEWEB)
J. D. Bess; T. L. Maddock; M. A. Marshall
2011-09-01
The Neutron Radiography (NRAD) reactor is a 250-kW TRIGA-(Training, Research, Isotope Production, General Atomics)-conversion-type reactor at the Idaho National Laboratory; it is primarily used for neutron radiography analysis of irradiated and unirradiated fuels and materials. The NRAD reactor was converted from HEU to LEU fuel with 60 fuel elements and brought critical on March 31, 2010. This configuration of the NRAD reactor has been evaluated as an acceptable benchmark experiment and is available in the 2011 editions of the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) and the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Significant effort went into precisely characterizing all aspects of the reactor core dimensions and material properties; detailed analyses of reactor parameters minimized experimental uncertainties. The largest contributors to the total benchmark uncertainty were the 234U, 236U, Er, and Hf content in the fuel; the manganese content in the stainless steel cladding; and the unknown level of water saturation in the graphite reflector blocks. A simplified benchmark model of the NRAD reactor was prepared with a keff of 1.0012 {+-} 0.0029 (1s). Monte Carlo calculations with MCNP5 and KENO-VI and various neutron cross section libraries were performed and compared with the benchmark eigenvalue for the 60-fuel-element core configuration; all calculated eigenvalues are between 0.3 and 0.8% greater than the benchmark value. Benchmark evaluations of the NRAD reactor are beneficial in understanding biases and uncertainties affecting criticality safety analyses of storage, handling, or transportation applications with LEU-Er-Zr-H fuel.
International Nuclear Information System (INIS)
We introduce a database (HAB11) of electronic coupling matrix elements (Hab) for electron transfer in 11 π-conjugated organic homo-dimer cations. High-level ab inito calculations at the multireference configuration interaction MRCI+Q level of theory, n-electron valence state perturbation theory NEVPT2, and (spin-component scaled) approximate coupled cluster model (SCS)-CC2 are reported for this database to assess the performance of three DFT methods of decreasing computational cost, including constrained density functional theory (CDFT), fragment-orbital DFT (FODFT), and self-consistent charge density functional tight-binding (FODFTB). We find that the CDFT approach in combination with a modified PBE functional containing 50% Hartree-Fock exchange gives best results for absolute Hab values (mean relative unsigned error = 5.3%) and exponential distance decay constants β (4.3%). CDFT in combination with pure PBE overestimates couplings by 38.7% due to a too diffuse excess charge distribution, whereas the economic FODFT and highly cost-effective FODFTB methods underestimate couplings by 37.6% and 42.4%, respectively, due to neglect of interaction between donor and acceptor. The errors are systematic, however, and can be significantly reduced by applying a uniform scaling factor for each method. Applications to dimers outside the database, specifically rotated thiophene dimers and larger acenes up to pentacene, suggests that the same scaling procedure significantly improves the FODFT and FODFTB results for larger π-conjugated systems relevant to organic semiconductors and DNA
International Nuclear Information System (INIS)
The interaction of Ar with H2 and HCl has been studied using Moeller - Plesset perturbation theory (MP2, MP3, MP4) and coupled-cluster [CCSD, CCSD(T)] methods with augmented correlation consistent basis sets. Basis sets as large as triply augmented quadruple zeta quality were used to investigate the convergence trends. Interaction energies were determined using the supermolecule approach with the counterpoise correction to account for basis set superposition error. Comparison with the available empirical potentials finds excellent agreement for both binding energies and transition state. For Ar - H2, the estimated complete basis set (CBS) limits for the binding energies of the two equivalent minima and the connecting transition state (TS) are, respectively, 55 and 47cm-1 at the MP4 level and 54 and 46cm-1 at the CCSD(T) level, respectively [the XC(fit) empirical potential of Bissonnette et al. [J. Chem. Phys. 105, 2639 (1996)] yields 56.6 and 47.8cm-1 for H2 (v=0)]. The estimated CBS limits for the binding energies of the two minima and transition state of Ar - HCl are 185, 155, and 109cm-1 at the MP4 level and 176, 147, and 105cm-1 at the CCSD(T) level, respectively [the H6(4,3,0) empirical potential of Hutson [J. Phys. Chem. 96, 4237 (1992)] yields 176.0, 148.3, and 103.3cm-1 for HCl (v=0)]. Basis sets containing diffuse functions of (dfg) symmetries were found to be essential for accurately modeling these two complexes, which are largely bound by dispersion and induction forces. Highly correlated wave functions were also required for accurate results. This was found to be particularly true for ArHCl, where significant differences in calculated binding energies were observed between MP2, MP4, and CCSD(T). copyright 1998 American Institute of Physics
Dujko, S; White, R D; Petrović, Z Lj; Robson, R E
2010-04-01
A multiterm solution of the Boltzmann equation has been developed and used to calculate transport coefficients of charged-particle swarms in gases under the influence of electric and magnetic fields crossed at arbitrary angles when nonconservative collisions are present. The hierarchy resulting from a spherical-harmonic decomposition of the Boltzmann equation in the hydrodynamic regime is solved numerically by representing the speed dependence of the phase-space distribution function in terms of an expansion in Sonine polynomials about a Maxwellian velocity distribution at an internally determined temperature. Results are given for electron swarms in certain collisional models for ionization and attachment over a range of angles between the fields and field strengths. The implicit and explicit effects of ionization and attachment on the electron-transport coefficients are considered using physical arguments. It is found that the difference between the two sets of transport coefficients, bulk and flux, resulting from the explicit effects of nonconservative collisions, can be controlled either by the variation in the magnetic field strengths or by the angles between the fields. In addition, it is shown that the phenomena of ionization cooling and/or attachment cooling/heating previously reported for dc electric fields carry over directly to the crossed electric and magnetic fields. The results of the Boltzmann equation analysis are compared with those obtained by a Monte Carlo simulation technique. The comparison confirms the theoretical basis and numerical integrity of the moment method for solving the Boltzmann equation and gives a set of well-established data that can be used to test future codes and plasma models. PMID:20481843
Energy Technology Data Exchange (ETDEWEB)
Kubas, Adam; Blumberger, Jochen, E-mail: j.blumberger@ucl.ac.uk [Department of Physics and Astronomy, University College London, Gower Street, London WC1E 6BT (United Kingdom); Hoffmann, Felix [Department of Physics and Astronomy, University College London, Gower Street, London WC1E 6BT (United Kingdom); Lehrstuhl für Theoretische Chemie, Ruhr-Universität Bochum, Universitätsstr. 150, 44801 Bochum (Germany); Heck, Alexander; Elstner, Marcus [Institute of Physical Chemistry, Karlsruhe Institute of Technology, Fritz-Haber-Weg 6, 76131 Karlsruhe (Germany); Oberhofer, Harald [Department of Chemistry, Technical University of Munich, Lichtenbergstr. 4, 85747 Garching (Germany)
2014-03-14
We introduce a database (HAB11) of electronic coupling matrix elements (H{sub ab}) for electron transfer in 11 π-conjugated organic homo-dimer cations. High-level ab inito calculations at the multireference configuration interaction MRCI+Q level of theory, n-electron valence state perturbation theory NEVPT2, and (spin-component scaled) approximate coupled cluster model (SCS)-CC2 are reported for this database to assess the performance of three DFT methods of decreasing computational cost, including constrained density functional theory (CDFT), fragment-orbital DFT (FODFT), and self-consistent charge density functional tight-binding (FODFTB). We find that the CDFT approach in combination with a modified PBE functional containing 50% Hartree-Fock exchange gives best results for absolute H{sub ab} values (mean relative unsigned error = 5.3%) and exponential distance decay constants β (4.3%). CDFT in combination with pure PBE overestimates couplings by 38.7% due to a too diffuse excess charge distribution, whereas the economic FODFT and highly cost-effective FODFTB methods underestimate couplings by 37.6% and 42.4%, respectively, due to neglect of interaction between donor and acceptor. The errors are systematic, however, and can be significantly reduced by applying a uniform scaling factor for each method. Applications to dimers outside the database, specifically rotated thiophene dimers and larger acenes up to pentacene, suggests that the same scaling procedure significantly improves the FODFT and FODFTB results for larger π-conjugated systems relevant to organic semiconductors and DNA.
Fission source convergence of Monte Carlo criticality calculations in weakly coupled fissile arrays
International Nuclear Information System (INIS)
Anomalous fission source convergence in a Monte Carlo criticality calculation for a weakly coupled array of two fissile material units are demonstrated. Introducing coupling coefficients among array units, it is quantitatively explained that this anomaly is caused by an insufficient restoring force to the true distribution and its large statistical uncertainty, especially, in a symmetric system. A new approach for estimating the fission source intensity ratio in an array is proposed by obtaining the eigenvector of a coupling coefficient matrix. This method also gives the uncertainty of the ratio as well as the ratio, which is available for evaluating the accuracy of the obtained ratio. The correlation between a calculated keff and the fission source intensity ratio is formulated. It is illustrated theoretically and empirically that there is no significant correlation in a symmetric two-unit array system. In general, care should be taken that a calculated keff may be biased by an incorrect fission source distribution, especially, in a slightly asymmetric system. A regionwise weight adjustment method is developed such that the fission source intensity ratio is forced to converge to a predetermined ratio. Using this method, a satisfactory convergence can be achieved. A larger number of neutrons per generation is recommended for a Monte Carlo criticality calculation of a weakly coupled array of units. (author)
International Nuclear Information System (INIS)
The principle of corresponding state on the fluctuation structure, which is the spatial distribution of various clusters of molecules caused by density fluctuations, in supercritical states around the critical points has been investigated. In this paper, we performed Molecular Dynamics (MD) simulation to extract the fluctuation structure around the critical points of 2-Center-Lennard-Jones (2CLJ) fluids, whose characteristics change by their molecular elongations. First, we indentified some critical points of 2CLJ fluids with comparatively shorter elongations applying Lotfi's function, which correctly describes the liquid-vapor coexistence line of Lennard-Jones (LJ) fluid, and successfully defined each critical point. Next, two methods were applied in the estimation of the fluctuation structure: one is the evaluation of the dispersion of the number of molecules at a certain domain, and the other is the calculation of static structure factor. As a result, in 2CLJ fluids which have shorter molecular elongations comparatively, the principle of corresponding state is satisfied because of the small differences in the fluctuation structures extracted in the present two methods. On the other hand, some results imply that the fluctuation may decrease in 2CLJ fluids which have the longer molecular elongations although more accurate evaluation of the critical points in those fluids is necessary for the further investigation. (author)
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The BFS-62 critical experiments are currently used as 'benchmark' for verification of IPPE codes and nuclear data, which have been used in the study of loading a significant amount of Pu in fast reactors. The BFS-62 experiments have been performed at BFS-2 critical facility of IPPE (Obninsk). The experimental program has been arranged in such a way that the effect of replacement of uranium dioxied blanket by the steel reflector as well as the effect of replacing UOX by MOX on the main characteristics of the reactor model was studied. Wide experimental program, including measurements of the criticality-keff, spectral indices, radial and axial fission rate distributions, control rod mock-up worth, sodium void reactivity effect SVRE and some other important nuclear physics parameters, was fulfilled in the core. Series of 4 BFS-62 critical assemblies have been designed for studying the changes in BN-600 reactor physics from existing state to hybrid core. All the assemblies are modeling the reactor state prior to refueling, i.e. with all control rod mock-ups withdrawn from the core. The following items are chosen for the analysis in this report: Description of the critical assembly BFS-62-3A as the 3rd assembly in a series of 4 BFS critical assemblies studying BN-600 reactor with MOX-UOX hybrid zone and steel reflector; Development of a 3D homogeneous calculation model for the BFS-62-3A critical experiment as the mock-up of BN-600 reactor with hybrid zone and steel reflector; Evaluation of measured nuclear physics parameters keff and SVRE (sodium void reactivity effect); Preparation of adjusted equivalent measured values for keff and SVRE. Main series of calculations are performed using 3D HEX-Z diffusion code TRIGEX in 26 groups, with the ABBN-93 cross-section set. In addition, precise calculations are made, in 299 groups and Ps-approximation in scattering, by Monte-Carlo code MMKKENO and discrete ordinate code TWODANT. All calculations are based on the common system
Intact and Degraded Component Criticality Calculations of N Reactor Spent Nuclear Fuel
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The objective of this calculation is to perform intact and degraded mode criticality evaluations of the Department of Energy's (DOE) N Reactor Spent Nuclear Fuel codisposed in a 2-Defense High-Level Waste (2-DHLW)/2-Multi-Canister Overpack (MCO) Waste Package (WP) and emplaced in a monitored geologic repository (MGR) (see Attachment I). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (keff) for both intact and degraded mode internal configurations of the codisposal waste package. This calculation will support the analysis that will be performed to demonstrate the technical viability for disposing of U-metal (N Reactor) spent nuclear fuel in the potential MGR
Evaluation of the accuracy of group calculations for reactor criticality perturbations
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For calculations of criticality perturbations it is necessary to use group constants which take into account not only the peculiarities of the intra-group flux but also those of the behaviour of the adjoint flux. A new method is proposed for obtaining bilinear-averaged constants of this type on the basis of the resonance characteristics of the importance function and the difference between the value of neutron importance at the group boundary and the group-averaged value (the bsup(+j) factor). A number of calculations are made for the ratios of reactivity coefficients in the BFS assemblies. Values have been obtained for the difference between the results of calculation with bilinear-averaged constants and those averaged conventionally (over flux). In many cases, this difference exceeds the experimental error. (author)
International Nuclear Information System (INIS)
To increase the accuracy of finite element simulations in daily practice the local German and Austrian Deep Drawing Research Groups of IDDRG founded a special Working Group in year 2000. The main objective of this group was the continuously ongoing study and discussion of numerical / material effects in simulation jobs and to work out possible solutions. As a first theme of this group the intensive study of small die radii and the possibility of detecting material failure in these critical forming positions was selected. The part itself is a fictional body panel outside in which the original door handle of the VW Golf A4 has been constructed, a typical position of possible material necking or rupture in the press shop. All conditions to do a successful simulation have been taken care of in advance, material data, boundary conditions, friction, FLC and others where determined for the two materials in investigation - a mild steel and a dual phase steel HXT500X. The results of the experiments have been used to design the descriptions of two different benchmark runs for the simulation. The simulations with different programs as well as with different parameters showed on one hand negligible and on the other hand parameters with strong impact on the result - thereby having a different impact on a possible material failure prediction
International Nuclear Information System (INIS)
The present paper summarizes calculation results for an international benchmark proposed by the Sodium-cooled Fast Reactor core Feed-back and transient response (SFR-FT) under the framework of the Working Party on scientific issues of Reactor Systems (WPRS) of the Nuclear Energy Agency of the OECD. It focuses on the large size oxide-fueled SFR. Library effect for core performance characteristics and reactivity feedback coefficients is analyzed using sensitivity analysis. The effect of ultra-fine energy group calculation in effective cross section generation is also analyzed. The discrepancy is about 0.4% for a neutron multiplication factor by changing JENDL-4.0 with JEFF-3.1. That is about -0.1% by changing JENDL-4.0 with ENDF/B-VII.1. The main contributions to the discrepancy between JENDL-4.0 and ENDF/B-VII.1 are 240Pu capture, 238U inelastic scattering and 239Pu fission. Those to the discrepancy between JENDL-4.0 and JEFF-3.1 are 23Na inelastic scattering, 56Fe inelastic scattering, 238Pu fission, 240Pu capture, 240Pu fission, 238U inelastic scattering, 239Pu fission and 239Pu nu-value. As for the sodium void reactivity, JEFF-3.1 and ENDF/B-VII.1 underestimate by about 8% compared with JENDL-4.0. The main contributions to the discrepancy between JENDL-4.0 and ENDF/B-VII.1 and 23Na elastic scattering, 23Na inelastic scattering and 239Pu fission. That to the discrepancy between JENDL-4.0 and JEFF-3.1 is 23Na inelastic scattering. The ultra-fine energy group calculation increases the sodium void reactivity by 2%. (author)
HELIOS-2: Benchmarking against hexagonal lattices
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The critical experiments performed at the Hungarian ZR-6 reactor and experiments performed at General Purpose P critical facility at RRC 'Kurchatov Institute' are used to benchmark HELIOS-2, as part of its ongoing validation and verification activities. The comparisons presented in this paper are based on, ZR6-WWER-EXP-001, Vol. 1 (2007), LEU-COMP-THERM-015, Vol. 4 (2005), and LEU-COMP-THERM-061, Vol. 4 (2002). On ZR-6, single cell, macro cell, and 2D calculations on selected experiments, regular and perturbed, are made. In the 2D calculations, the radial leakage is treated by including the reflector in the calculations, representing only axial leakage by a measured axial buckling. keff and RMS reaction rates comparisons are presented. On General Purpose P critical facility, the entire experiment is modelled as 2D. Comparisons of keff and fission rates are presented and the effect of the axial buckling on keff is investigated. (Author)
International Nuclear Information System (INIS)
The IAEA-WIMS Library Update Project (WLUP) is on the end stage. The final library will be released on 2002. It is a result of research and development made by more than ten investigators during 10 years. The organization of benchmarks for testing and choosing the best set of data has been coordinated by the author of this paper. It is presented the organization, name conventions, contents and documentation of WLUP benchmarks, and an updated list of the main parameters for all cases. First, the benchmarks objectives and types are given. Then, comparisons of results from different WIMSD libraries are included. Finally it is described the program QVALUE for analysis and plot of results. Some examples are given. The set of benchmarks implemented on this work is a fundamental tool for testing new multigroup libraries. (author)
Adaptation of the B1 leakage model to Monte Carlo criticality calculations
International Nuclear Information System (INIS)
This paper presents an attempt to consistently adapt the B1 homogeneous leakage model within Monte Carlo criticality calculations based on the power iteration method. Unlike deterministic lattice codes, most of Monte Carlo-based reactor physics codes perform lattice calculations without introducing leakage models. The critical flux is however required to accurately compute homogenized cross sections and diffusion coefficients in the context of lattice physics computation. In our proposed approach, a fundamental mode approximation is introduced in the Monte Carlo K-effective power iteration method. Similarly to the deterministic implementation of the lattice code DRAGON (typically the collision probability method), B1 equations are solved at each cycle, leading to Monte Carlo estimates for the critical buckling B2 and for the group-dependent leakage rates. These leakage reactions are then introduced in the neutron random walk. This approach is discussed on legacy PWR pin cell cases, by direct comparison with results obtained by the collision probability method. This approach leads to consistent results between the Monte Carlo and the deterministic computational ways of the DRAGON code. (author)
International Nuclear Information System (INIS)
On the recommendation of the IAEA International Working Group on Gas Cooled Reactors, the IAEA established a Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low-Enriched High Temperature Gas Cooled Reactors (HTGRs) in 1990. The objective of the CRP was to provide safety-related physics data for low-enriched uranium (LEU) fueled HTGRs for use in validating reactor physics codes used by the participating countries for analyses of their designs. Experience on low-enriched uranium, graphite-moderated reactor systems from research institutes and critical facilities in participating countries were brought into the CRP and shared among participating institutes. The status of experimental data and code validation for HTGRs and the remaining needs at the initiation of this CRP were addressed in detail at the IAEA Specialists Meeting on Uncertainties in Physics Calculations for HTGR Cores held at the Paul Scherrer Institute (PSI), Villigen, Switzerland in May, 1990. The main activities of the CRP were conducted within an international project at the PROTEUS critical experiment facility at the Paul Scherrer Institute, Villigen, Switzerland. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. Fuel for the experiments was provided by the KFA Research Center, Juelich, Germany. Initial criticality was achieved on July 7, 1992. These experiments were conducted over a range of experimental parameters such as carbon-to-uranium ratio, core height-to-diameter ratio, and simulated moisture concentration. To assure that the experiments being conducted are appropriate for the design of the participants, specialists from each of the countries have participated
Critical analysis of fragment-orbital DFT schemes for the calculation of electronic coupling values
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Schober, Christoph; Reuter, Karsten; Oberhofer, Harald, E-mail: harald.oberhofer@ch.tum.de [Chair for Theoretical Chemistry, Technische Universität München, Lichtenbergstr. 4, D-85747 Garching (Germany)
2016-02-07
We present a critical analysis of the popular fragment-orbital density-functional theory (FO-DFT) scheme for the calculation of electronic coupling values. We discuss the characteristics of different possible formulations or “flavors” of the scheme which differ by the number of electrons in the calculation of the fragments and the construction of the Hamiltonian. In addition to two previously described variants based on neutral fragments, we present a third version taking a different route to the approximate diabatic state by explicitly considering charged fragments. In applying these FO-DFT flavors to the two molecular test sets HAB7 (electron transfer) and HAB11 (hole transfer), we find that our new scheme gives improved electronic couplings for HAB7 (−6.2% decrease in mean relative signed error) and greatly improved electronic couplings for HAB11 (−15.3% decrease in mean relative signed error). A systematic investigation of the influence of exact exchange on the electronic coupling values shows that the use of hybrid functionals in FO-DFT calculations improves the electronic couplings, giving values close to or even better than more sophisticated constrained DFT calculations. Comparing the accuracy and computational cost of each variant, we devise simple rules to choose the best possible flavor depending on the task. For accuracy, our new scheme with charged-fragment calculations performs best, while numerically more efficient at reasonable accuracy is the variant with neutral fragments.
The validity of the transport approximation in critical-size and reactivity calculations
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The validity of the transport approximation in critical-size and reactivity calculations. Elastically scattered neutrons are, in general, not distributed isotropically in the laboratory system, and a convenient way of taking this into account in neutron- transport calculations is to use the transport approximation. In this, the elastic cross-section is replaced by an elastic transport cross-section with an isotropic angular distribution. This leads to a considerable simplification in the neutron-transport calculation. In the present paper, the theoretical bases of the transport approximation in both one-group and many-group formalisms are given. The accuracy of the approximation is then studied in the multi-group case for a number of typical systems by means of the Sn method using the isotropic and anisotropic versions of the method, which exist as alternative options of the machine code SAINT written at Aldermaston for use on IBM-709/7090 machines. The dependence of the results of the anisotropic calculations on the number of moments used to represent the angular distributions is also examined. The results of the various calculations are discussed, and an indication is given of the types of system for which the transport approximation is adequate and of those for which it is inadequate. (author)
Critical analysis of fragment-orbital DFT schemes for the calculation of electronic coupling values
International Nuclear Information System (INIS)
We present a critical analysis of the popular fragment-orbital density-functional theory (FO-DFT) scheme for the calculation of electronic coupling values. We discuss the characteristics of different possible formulations or “flavors” of the scheme which differ by the number of electrons in the calculation of the fragments and the construction of the Hamiltonian. In addition to two previously described variants based on neutral fragments, we present a third version taking a different route to the approximate diabatic state by explicitly considering charged fragments. In applying these FO-DFT flavors to the two molecular test sets HAB7 (electron transfer) and HAB11 (hole transfer), we find that our new scheme gives improved electronic couplings for HAB7 (−6.2% decrease in mean relative signed error) and greatly improved electronic couplings for HAB11 (−15.3% decrease in mean relative signed error). A systematic investigation of the influence of exact exchange on the electronic coupling values shows that the use of hybrid functionals in FO-DFT calculations improves the electronic couplings, giving values close to or even better than more sophisticated constrained DFT calculations. Comparing the accuracy and computational cost of each variant, we devise simple rules to choose the best possible flavor depending on the task. For accuracy, our new scheme with charged-fragment calculations performs best, while numerically more efficient at reasonable accuracy is the variant with neutral fragments
Energy Technology Data Exchange (ETDEWEB)
Pruet, J; Brown, D A; Descalle, M
2006-05-22
The authors describe tools developed by the Computational Nuclear Physics group for testing the quality of internally developed nuclear data and the fidelity of translations from ENDF formatted data to ENDL formatted data used by Livermore. These tests include S{sub n} calculations for the effective k value characterizing critical assemblies and for replacement coefficients of different materials embedded in the Godiva and Jezebel critical assemblies. For those assemblies and replacement materials for which reliable experimental information is available, these calculations provide an integral check on the quality of data. Because members of the ENDF and reactor communities use calculations for these same assemblies in their validation process, a comparison between their results with ENDF formatted data and their results with data translated into the ENDL format provides a strong check on the accuracy of translations. As a first application of the test suite they present a study comparing ENDL 99 and ENDF/B-V. They also consider the quality of the ENDF/B-V translation previously done by the Computational Nuclear Physics group. No significant errors are found.
Three-dimensional RAMA fluence methodology benchmarking
International Nuclear Information System (INIS)
This paper describes the benchmarking of the RAMA Fluence Methodology software, that has been performed in accordance with U. S. Nuclear Regulatory Commission Regulatory Guide 1.190. The RAMA Fluence Methodology has been developed by TransWare Enterprises Inc. through funding provided by the Electric Power Research Inst., Inc. (EPRI) and the Boiling Water Reactor Vessel and Internals Project (BWRVIP). The purpose of the software is to provide an accurate method for calculating neutron fluence in BWR pressure vessels and internal components. The Methodology incorporates a three-dimensional deterministic transport solution with flexible arbitrary geometry representation of reactor system components, previously available only with Monte Carlo solution techniques. Benchmarking was performed on measurements obtained from three standard benchmark problems which include the Pool Criticality Assembly (PCA), VENUS-3, and H. B. Robinson Unit 2 benchmarks, and on flux wire measurements obtained from two BWR nuclear plants. The calculated to measured (C/M) ratios range from 0.93 to 1.04 demonstrating the accuracy of the RAMA Fluence Methodology in predicting neutron flux, fluence, and dosimetry activation. (authors)
Reference calculations on critical assemblies with Apollo2 code working with a fine multigroup mesh
International Nuclear Information System (INIS)
The objective of this thesis is to add to the multigroup transport code APOLLO2 the capability to perform deterministic reference calculations, for any type of reactor, using a very fine energy mesh of several thousand groups. This new reference tool allows us to validate the self-shielding model used in industrial applications, to perform depletion calculations, differential effects calculations, critical buckling calculations or to evaluate precisely data required by the self shielding model. At its origin, APOLLO2 was designed to perform routine calculations with energy meshes around one hundred groups. That is why, in the current format of cross sections libraries, almost each value of the multigroup energy transfer matrix is stored. As this format is not convenient for a high number of groups (concerning memory size), we had to search out a new format for removal matrices and consequently to modify the code. In the new format we found, only some values of removal matrices are kept (these values depend on a reconstruction precision choice), the other ones being reconstructed by a linear interpolation, what reduces the size of these matrices. Then we had to show that APOLLO2 working with a fine multigroup mesh had the capability to perform reference calculations on any assembly geometry. For that, we successfully carried out the validation with several calculations for which we compared APOLLO2 results (obtained with the universal mesh of 11276 groups) to results obtained with Monte Carlo codes (MCNP, TRIPOLI4). Physical analysis led with this new tool have been very fruitful and show a great potential for such an R and D tool. (author)
On the thermal scattering law data for reactor lattice calculations
International Nuclear Information System (INIS)
Thermal scattering law data for hydrogen bound in water, hydrogen bound in zirconium hydride and deuterium bound in heavy water have been re-evaluated. The influence of the thermal scattering law data on critical lattices has been studied with detailed Monte Carlo calculations and a summary of results is presented for a numerical benchmark and for the TRIGA reactor benchmark. Systematics for a large sequence of benchmarks analysed with the WIMS-D lattice code are also presented. (author)
Energy Technology Data Exchange (ETDEWEB)
Mueller, Don [ORNL; Rearden, Bradley T [ORNL; Hollenbach, Daniel F [ORNL
2009-02-01
The Radiochemical Development Facility at Oak Ridge National Laboratory has been storing solid materials containing 233U for decades. Preparations are under way to process these materials into a form that is inherently safe from a nuclear criticality safety perspective. This will be accomplished by down-blending the {sup 233}U materials with depleted or natural uranium. At the request of the U.S. Department of Energy, a study has been performed using the SCALE sensitivity and uncertainty analysis tools to demonstrate how these tools could be used to validate nuclear criticality safety calculations of selected process and storage configurations. ISOTEK nuclear criticality safety staff provided four models that are representative of the criticality safety calculations for which validation will be needed. The SCALE TSUNAMI-1D and TSUNAMI-3D sequences were used to generate energy-dependent k{sub eff} sensitivity profiles for each nuclide and reaction present in the four safety analysis models, also referred to as the applications, and in a large set of critical experiments. The SCALE TSUNAMI-IP module was used together with the sensitivity profiles and the cross-section uncertainty data contained in the SCALE covariance data files to propagate the cross-section uncertainties ({Delta}{sigma}/{sigma}) to k{sub eff} uncertainties ({Delta}k/k) for each application model. The SCALE TSUNAMI-IP module was also used to evaluate the similarity of each of the 672 critical experiments with each application. Results of the uncertainty analysis and similarity assessment are presented in this report. A total of 142 experiments were judged to be similar to application 1, and 68 experiments were judged to be similar to application 2. None of the 672 experiments were judged to be adequately similar to applications 3 and 4. Discussion of the uncertainty analysis and similarity assessment is provided for each of the four applications. Example upper subcritical limits (USLs) were
Benchmarking and Performance Management
Directory of Open Access Journals (Sweden)
Adrian TANTAU
2010-12-01
Full Text Available The relevance of the chosen topic is explained by the meaning of the firm efficiency concept - the firm efficiency means the revealed performance (how well the firm performs in the actual market environment given the basic characteristics of the firms and their markets that are expected to drive their profitability (firm size, market power etc.. This complex and relative performance could be due to such things as product innovation, management quality, work organization, some other factors can be a cause even if they are not directly observed by the researcher. The critical need for the management individuals/group to continuously improve their firm/company’s efficiency and effectiveness, the need for the managers to know which are the success factors and the competitiveness determinants determine consequently, what performance measures are most critical in determining their firm’s overall success. Benchmarking, when done properly, can accurately identify both successful companies and the underlying reasons for their success. Innovation and benchmarking firm level performance are critical interdependent activities. Firm level variables, used to infer performance, are often interdependent due to operational reasons. Hence, the managers need to take the dependencies among these variables into account when forecasting and benchmarking performance. This paper studies firm level performance using financial ratio and other type of profitability measures. It uses econometric models to describe and then propose a method to forecast and benchmark performance.
The Criticality Calculation Of Fission Yield Of U-235 Solution And Its Radiation Dose
International Nuclear Information System (INIS)
The calculation assesment of fission yield of U-235 solution in the extraction and evaporation units has been performed for the prediction of that when the criticality accident occurs in the production of fuel element for the research reactor. The Grover Tuck and fission distribution probability methods are used in this case. The calculation result using the fission distribution probability methods show the fission of 2,7 x 1018 for the uranium concentration of 200 grams/litre and that of 2,5 x 1018 fissions for U of 40 grams/litre in the extraction unit. The calculation results from the evaporation unit revealed the fission of 3,1 x 1018 for 400 grams/litre uranium and 1,77 x 1018 fissions for 80 grams/litre uranium. Using the Grover Tuck calculation method give results that 8,267 x 1017 fissions and 2,878 x 1017 fissions respectively. Radiation dose of 200 gram/litre solution is about 1450,29 Rad for neutron and 4785,96 Rad for gamma ray
Criticality calculations for a spent fuel storage pool for a BWR type reactor
International Nuclear Information System (INIS)
In this work, the methodology for the calculation of the constant of effective multiplication for the arrangement of spent fuel assemblies in the pool of a BWR type reactor is shown. Calculations were done for the pool of spent fuel specified in FSAR and for the assemblies that is thought a conservative composition of high enrichment and without Gadolinium, giving credit to the stainless steel boxes of the frames that keep the assemblies. To carry out this simulation, RECORD and MIXQUIC codes were used. With record code, macroscopic cross sections, two energy groups, for the characteristics of the thought assemblies were obtained. Cross sections, as well as the dimensions of the frames that keep the fuel assemblies were used as input data for MIXQUIC code. With this code, criticality calculations in two dimensions were done, supposing that there is not leak of neutrons along the axial of the main line. Additional calculations, supposing changes in the temperature, distance among fuel assemblies and the thickness of the stainless steel box of the frame were done. The obtained results, including the effect in tolerances due to temperature, weight and thickness, show that the arrangement in the pool, when frames are fully charged, is subcritical by less than 5% in δK. (Author)
International Nuclear Information System (INIS)
Using Mixed Oxide (MOX) fuel in traditional Pressurized Water Reactor (PWR) assemblies has been researched at length and has shown to provide the benefit of transmutation and targets the amount and toxicity of high level waste needed to be managed. Advanced MOX concepts using enriched Uranium Dioxide (UO2) are required for multiple recycling of plutonium. The use of MOX and ordinary UO2 fuel in the same assembly as well as unfueled rods and assembly edge effects contrasts with the unit cell computational assumption of a uniform infinite array of rods. While a deterministic method of calculating the Dancoff factor has traditionally been employed in fuel assembly analysis due to the lighter computational and modeling requirements, this research seeks to determine the validity of the uniform, infinite lattice assumption with respect to Dancoff factor and determine the magnitude of the impact of nonuniform lattice effects on fuel assembly criticality calculations as well as transuranic isotope production and transmutation. This research explored the pin-to-pin interaction in a non-uniform lattice of MOX fuel rods and UO2 fuel rods through the impact of the calculated Dancoff factors from the deterministic method used in SCALE versus the Monte Carlo method used in the code DANCOFF-MC. Using the Monte Carlo method takes into account the non-uniform lattice effects of having neighboring fuel rods with different cross-sectional spectra whereas the Dancoff factor calculated by SCALE assumes a uniform, infinite lattice of one fuel rod type. Differences in eigenvalue calculations as a function of burnup are present between the two methods of Dancoff factor calculation. The percent difference is greatest at low burnup and then becomes smaller throughout the cycle. Differences in the transmutation rate of transuranic isotopes in the MOX fuel are also present between the Dancoff factor calculation methods. The largest difference is in Pu-239, Pu-242, and Am-241 composition
Design floor response spectra calculation research for 18-5 critical device building
International Nuclear Information System (INIS)
Background: Seismic test analysis of a cabinet in 18-5 critical device building is required for safety inspection of zero power reactor, however, the floor response spectra at the height of 3.5 m where the cabinet is located are lacking. Purpose: For seismic test analysis of the cabinet in 18-5 critical device building, dynamic analysis is performed for the building, with the purpose to obtain the floor response spectra at the height of 3.5 m where the cabinet is located. Methods: The displacement time history and acceleration time history on the given height floor are calculated by using transient dynamic time history method with the software ANSYS under the load of foundation displacement time history, then acceleration response spectrum on the floor that the cabinet locates is calculated with the corresponding acceleration time history obtained on the cabinet floor. Results: The design floor response spectra under load OBE and SSE at the height of 3.5 m are offered with damping ratios of 2%, 4%, 5% and 7%, respectively. Conclusions: The design floor response spectra are credible and have been adopted by test analysis of the cabinet. (authors)
Directory of Open Access Journals (Sweden)
Daniela Niculescu
2016-02-01
Full Text Available Organisational culture and employee engagement have been the focus of recent broad-based research efforts. Adding this concern to the revealed importance of performance indicators on human capital, and that their use is getting momentum, in order to attach financial values to knowledge management assets, it becomes more and more critical to measure human capital value. Key for Romanian FSO’s managers becomes to consider that both human and financial values have a focus on adding value in every process and function in the organisation, and to perpetuate organisational profitability by the corporate culture, on the one hand, where culture is a powerful factor that helps a company to engage, on the other hand, talented people. There is a substantial concern on using ROI on Learning and Development programmes, but whilst this is still declared, Romanian FSOs do not yet have a consistent method to measure it. This study is showing the criticality of connecting people to financial results and data analysis suggests that ROI calculation has a positive impact on creating and fostering a powerful organisational culture and that employees’ awareness of ROI values within their organisation has a powerful effect on their sense of engagement. Our findings have a more practical implication for the analysed industry by shaping a formal ROI measurement mechanisms blueprint, an ROI calculation model for the Romanian FSOs, in the form of a mechanism that could be employed when considering the design of an ROI Methodology for Romanian FSOs.
Podymaka, Valerii I.; Novohretskyi, Serhii M.
2014-01-01
The dynamic stability calculation method of the parallel operation of the marine synchronous generators is examined. The improved calculations algorithm of the critical time of the short circuit switch-off is offeted which takes into account salient-pole rotor of the synchronous generator. The results of calculations which prove efficiency of using the proposed algorithm, are presented.
Critical groups vs. representative person: dose calculations due to predicted releases from USEXA
Energy Technology Data Exchange (ETDEWEB)
Ferreira, N.L.D., E-mail: nelson.luiz@ctmsp.mar.mil.br [Centro Tecnologico da Marinha (CTM/SP), Sao Paulo, SP (Brazil); Rochedo, E.R.R., E-mail: elainerochedo@gmail.com [Instituto de Radiprotecao e Dosimetria (lRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Mazzilli, B.P., E-mail: mazzilli@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)
2013-07-01
The critical group cf Centro Experimental Aramar (CEA) site was previously defined based 00 the effluents releases to the environment resulting from the facilities already operational at CEA. In this work, effective doses are calculated to members of the critical group considering the predicted potential uranium releases from the Uranium Hexafluoride Production Plant (USEXA). Basically, this work studies the behavior of the resulting doses related to the type of habit data used in the analysis and two distinct situations are considered: (a) the utilization of average values obtained from official institutions (IBGE, IEA-SP, CNEN, IAEA) and from the literature; and (b) the utilization of the 95{sup tb} percentile of the values derived from distributions fit to the obtained habit data. The first option corresponds to the way that data was used for the definition of the critical group of CEA done in former assessments, while the second one corresponds to the use of data in deterministic assessments, as recommended by ICRP to estimate doses to the so--called 'representative person' . (author)
Critical groups vs. representative person: dose calculations due to predicted releases from USEXA
International Nuclear Information System (INIS)
The critical group cf Centro Experimental Aramar (CEA) site was previously defined based 00 the effluents releases to the environment resulting from the facilities already operational at CEA. In this work, effective doses are calculated to members of the critical group considering the predicted potential uranium releases from the Uranium Hexafluoride Production Plant (USEXA). Basically, this work studies the behavior of the resulting doses related to the type of habit data used in the analysis and two distinct situations are considered: (a) the utilization of average values obtained from official institutions (IBGE, IEA-SP, CNEN, IAEA) and from the literature; and (b) the utilization of the 95tb percentile of the values derived from distributions fit to the obtained habit data. The first option corresponds to the way that data was used for the definition of the critical group of CEA done in former assessments, while the second one corresponds to the use of data in deterministic assessments, as recommended by ICRP to estimate doses to the so--called 'representative person' . (author)
Yu, Jen-Shiang K; Hwang, Jenn-Kang; Tang, Chuan Yi; Yu, Chin-Hui
2004-01-01
A number of recently released numerical libraries including Automatically Tuned Linear Algebra Subroutines (ATLAS) library, Intel Math Kernel Library (MKL), GOTO numerical library, and AMD Core Math Library (ACML) for AMD Opteron processors, are linked against the executables of the Gaussian 98 electronic structure calculation package, which is compiled by updated versions of Fortran compilers such as Intel Fortran compiler (ifc/efc) 7.1 and PGI Fortran compiler (pgf77/pgf90) 5.0. The ifc 7.1 delivers about 3% of improvement on 32-bit machines compared to the former version 6.0. Performance improved from pgf77 3.3 to 5.0 is also around 3% when utilizing the original unmodified optimization options of the compiler enclosed in the software. Nevertheless, if extensive compiler tuning options are used, the speed can be further accelerated to about 25%. The performances of these fully optimized numerical libraries are similar. The double-precision floating-point (FP) instruction sets (SSE2) are also functional on AMD Opteron processors operated in 32-bit compilation, and Intel Fortran compiler has performed better optimization. Hardware-level tuning is able to improve memory bandwidth by adjusting the DRAM timing, and the efficiency in the CL2 mode is further accelerated by 2.6% compared to that of the CL2.5 mode. The FP throughput is measured by simultaneous execution of two identical copies of each of the test jobs. Resultant performance impact suggests that IA64 and AMD64 architectures are able to fulfill significantly higher throughput than the IA32, which is consistent with the SpecFPrate2000 benchmarks. PMID:15032545
International Nuclear Information System (INIS)
Computer development has a bearing on the choice of methods and their possible uses. The authors discuss the possible uses of the diffusion and transport theories and their limitations. Most of the problems encountered in regard to criticality involve fissile materials in simple or multiple assemblies. These entail the use of methods of calculation based on different principles. There are approximate methods of calculation, but very often, for economic reasons or with a view to practical application, a high degree of accuracy is required in determining the reactivity of the assemblies in question, and the methods based on the Monte Carlo principle are then the most valid. When these methods are used, accuracy is linked with the calculation time, so that the usefulness of the codes derives from their speed. With a view to carrying out the work in the best conditions, depending on the geometry and the nature of the materials involved, various codes must be used. Four principal codes are described, as are their variants; some typical possibilities and certain fundamental results are presented. Finally the accuracies of the various methods are compared. (author)
International Nuclear Information System (INIS)
The ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2003) presents the methodology for evaluating potential criticality situations in the monitored geologic repository. As stated in the referenced Topical Report, the detailed methodology for performing the disposal criticality analyses will be documented in model reports. Many of the models developed in support of the Topical Report differ from the definition of models as given in the Office of Civilian Radioactive Waste Management procedure AP-SIII.10Q, ''Models'', in that they are procedural, rather than mathematical. These model reports document the detailed methodology necessary to implement the approach presented in the Disposal Criticality Analysis Methodology Topical Report and provide calculations utilizing the methodology. Thus, the governing procedure for this type of report is AP-3.12Q, ''Design Calculations and Analyses''. The ''Criticality Model'' is of this latter type, providing a process evaluating the criticality potential of in-package and external configurations. The purpose of this analysis is to layout the process for calculating the criticality potential for various in-package and external configurations and to calculate lower-bound tolerance limit (LBTL) values and determine range of applicability (ROA) parameters. The LBTL calculations and the ROA determinations are performed using selected benchmark experiments that are applicable to various waste forms and various in-package and external configurations. The waste forms considered in this calculation are pressurized water reactor (PWR), boiling water reactor (BWR), Fast Flux Test Facility (FFTF), Training Research Isotope General Atomic (TRIGA), Enrico Fermi, Shippingport pressurized water reactor, Shippingport light water breeder reactor (LWBR), N-Reactor, Melt and Dilute, and Fort Saint Vrain Reactor spent nuclear fuel (SNF). The scope of this analysis is to document the criticality computational method. The criticality