Benchmark calculations of sodium fast critical experiments
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
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
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
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
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)
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
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
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
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
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
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)
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)
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
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
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.)
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)
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
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
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)
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
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
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)
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'
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
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
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)
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
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
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
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
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)
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
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
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
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
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
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
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
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
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
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
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
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)
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)
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
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
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
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
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.).
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
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
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
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.
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)
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
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 ...
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
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
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
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
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
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
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)
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
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
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
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
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
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
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
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
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
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.
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
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
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
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
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
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
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
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
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
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
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
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)
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
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
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
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
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
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
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
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
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
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)
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
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)
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
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
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
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
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
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.
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
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
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
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
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
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
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
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
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
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)
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)
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
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
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
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
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
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)
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
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
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
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
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
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
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
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).
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
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
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)
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
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)
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.
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
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
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
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
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)
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
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
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
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
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.)
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
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)
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
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.)
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
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.
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)
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)
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
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
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
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
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
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
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
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
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
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
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
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
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)
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.
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.
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.
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
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
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
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
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)
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
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
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
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
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
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
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...
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
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
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
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
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
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
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
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
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)
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
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
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
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)
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
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)
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
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
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...
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
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
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)
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
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.
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
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)
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
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
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
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
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)
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
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)
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
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)
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
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
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
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%.
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
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
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
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
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
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)
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