The ORSphere Benchmark Evaluation and Its Potential Impact on Nuclear Criticality Safety

Energy Technology Data Exchange (ETDEWEB)

John D. Bess; Margaret A. Marshall; J. Blair Briggs

2013-10-01

In the early 1970’s, critical experiments using an unreflected metal sphere of highly enriched uranium (HEU) were performed with the focus to provide a “very accurate description…as an ideal benchmark for calculational methods and cross-section data files.” Two near-critical configurations of the Oak Ridge Sphere (ORSphere) were evaluated as acceptable benchmark experiments for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook). The results from those benchmark experiments were then compared with additional unmoderated and unreflected HEU metal benchmark experiment configurations currently found in the ICSBEP Handbook. For basic geometries (spheres, cylinders, and slabs) the eigenvalues calculated using MCNP5 and ENDF/B-VII.0 were within 3 of their respective benchmark values. There appears to be generally a good agreement between calculated and benchmark values for spherical and slab geometry systems. Cylindrical geometry configurations tended to calculate low, including more complex bare HEU metal systems containing cylinders. The ORSphere experiments do not calculate within their 1s uncertainty and there is a possibility that the effect of the measured uncertainties for the GODIVA I benchmark may need reevaluated. There is significant scatter in the calculations for the highly-correlated ORCEF cylinder experiments, which are constructed from close-fitting HEU discs and annuli. Selection of a nuclear data library can have a larger impact on calculated eigenvalue results than the variation found within calculations of a given experimental series, such as the ORCEF cylinders, using a single nuclear data set.

The International Criticality Safety Benchmark Evaluation Project

International Nuclear Information System (INIS)

Briggs, B. J.; Dean, V. F.; Pesic, M. P.

2001-01-01

In order to properly manage the risk of a nuclear criticality accident, it is important to establish the conditions for which such an accident becomes possible for any activity involving fissile material. Only when this information is known is it possible to establish the likelihood of actually achieving such conditions. It is therefore important that criticality safety analysts have confidence in the accuracy of their calculations. Confidence in analytical results can only be gained through comparison of those results with experimental data. The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in October of 1992 by the US Department of Energy. The project was managed through the Idaho National Engineering and Environmental Laboratory (INEEL), but involved 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 Flats Plant. An International Criticality Safety Data Exchange component was added to the project during 1994 and the project became what is currently known as the International Criticality Safety Benchmark Evaluation Project (ICSBEP). Representatives from the United Kingdom, France, Japan, the Russian Federation, Hungary, Kazakhstan, Korea, Slovenia, Yugoslavia, Spain, and Israel are now participating on the project In December of 1994, the ICSBEP became an official activity of the Organization for Economic Cooperation and Development - Nuclear Energy Agency's (OECD-NEA) Nuclear Science Committee. The United States currently remains the lead country, providing most of the administrative support. The purpose of the ICSBEP is to: (1) identify and evaluate a comprehensive set of critical benchmark data; (2) verify the data, to the extent possible, by reviewing original and subsequently revised documentation, and by talking with the

Pool critical assembly pressure vessel facility benchmark

International Nuclear Information System (INIS)

Remec, I.; Kam, F.B.K.

1997-07-01

This pool critical assembly (PCA) pressure vessel wall facility benchmark (PCA benchmark) is described and analyzed in this report. Analysis of the PCA benchmark can be used for partial fulfillment of the requirements for the qualification of the methodology for pressure vessel neutron fluence calculations, as required by the US Nuclear Regulatory Commission regulatory guide DG-1053. Section 1 of this report describes the PCA benchmark and provides all data necessary for the benchmark analysis. The measured quantities, to be compared with the calculated values, are the equivalent fission fluxes. In Section 2 the analysis of the PCA benchmark is described. Calculations with the computer code DORT, based on the discrete-ordinates method, were performed for three ENDF/B-VI-based multigroup libraries: BUGLE-93, SAILOR-95, and BUGLE-96. An excellent agreement of the calculated (C) and measures (M) equivalent fission fluxes was obtained. The arithmetic average C/M for all the dosimeters (total of 31) was 0.93 ± 0.03 and 0.92 ± 0.03 for the SAILOR-95 and BUGLE-96 libraries, respectively. The average C/M ratio, obtained with the BUGLE-93 library, for the 28 measurements was 0.93 ± 0.03 (the neptunium measurements in the water and air regions were overpredicted and excluded from the average). No systematic decrease in the C/M ratios with increasing distance from the core was observed for any of the libraries used

Benchmark calculation of SCALE-PC 4.3 CSAS6 module and burnup credit criticality analysis

Energy Technology Data Exchange (ETDEWEB)

Shin, Hee Sung; Ro, Seong Gy; Shin, Young Joon; Kim, Ik Soo [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-12-01

Calculation biases of SCALE-PC CSAS6 module for PWR spent fuel, metallized spent fuel and solution of nuclear materials have been determined on the basis of the benchmark to be 0.01100, 0.02650 and 0.00997, respectively. With the aid of the code system, nuclear criticality safety analysis for the spent fuel storage pool has been carried out to determine the minimum burnup of spent fuel required for safe storage. The criticality safety analysis is performed using three types of isotopic composition of spent fuel: ORIGEN2-calculated isotopic compositions; the conservative inventory obtained from the multiplication of ORIGEN2-calculated isotopic compositions by isotopic correction factors; the conservative inventory of only U, Pu and {sup 241}Am. The results show that the minimum burnup for three cases are 990,6190 and 7270 MWd/tU, respectively in the case of 5.0 wt% initial enriched spent fuel. (author). 74 refs., 68 figs., 35 tabs.

Compilation report of VHTRC temperature coefficient benchmark calculations

Energy Technology Data Exchange (ETDEWEB)

Yasuda, Hideshi; Yamane, Tsuyoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1995-11-01

A calculational benchmark problem has been proposed by JAERI to an IAEA Coordinated Research Program, `Verification of Safety Related Neutronic Calculation for Low-enriched Gas-cooled Reactors` to investigate the accuracy of calculation results obtained by using codes of the participating countries. This benchmark is made on the basis of assembly heating experiments at a pin-in block type critical assembly, VHTRC. Requested calculation items are the cell parameters, effective multiplication factor, temperature coefficient of reactivity, reaction rates, fission rate distribution, etc. Seven institutions from five countries have joined the benchmark works. Calculation results are summarized in this report with some remarks by the authors. Each institute analyzed the problem by applying the calculation code system which was prepared for the HTGR development of individual country. The values of the most important parameter, k{sub eff}, by all institutes showed good agreement with each other and with the experimental ones within 1%. The temperature coefficient agreed within 13%. The values of several cell parameters calculated by several institutes did not agree with the other`s ones. It will be necessary to check the calculation conditions again for getting better agreement. (J.P.N.).

EPRI depletion benchmark calculations using PARAGON

International Nuclear Information System (INIS)

Kucukboyaci, Vefa N.

2015-01-01

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

Criticality benchmark comparisons leading to cross-section upgrades

International Nuclear Information System (INIS)

Alesso, H.P.; Annese, C.E.; Heinrichs, D.P.; Lloyd, W.R.; Lent, E.M.

1993-01-01

For several years criticality benchmark calculations with COG. COG is a point-wise Monte Carlo code developed at Lawrence Livermore National Laboratory (LLNL). It solves the Boltzmann equation for the transport of neutrons and photons. The principle consideration in developing COG was that the resulting calculation would be as accurate as the point-wise cross-sectional data, since no physics computational approximations were used. The objective of this paper is to report on COG results for criticality benchmark experiments in concert with MCNP comparisons which are resulting in corrections an upgrades to the point-wise ENDL cross-section data libraries. Benchmarking discrepancies reported here indicated difficulties in the Evaluated Nuclear Data Livermore (ENDL) cross-sections for U-238 at thermal neutron energy levels. This led to a re-evaluation and selection of the appropriate cross-section values from several cross-section sets available (ENDL, ENDF/B-V). Further cross-section upgrades anticipated

Benchmark calculations for VENUS-2 MOX -fueled reactor dosimetry

International Nuclear Information System (INIS)

Kim, Jong Kung; Kim, Hong Chul; Shin, Chang Ho; Han, Chi Young; Na, Byung Chan

2004-01-01

As a part of a Nuclear Energy Agency (NEA) Project, it was pursued the benchmark for dosimetry calculation of the VENUS-2 MOX-fueled reactor. In this benchmark, the goal is to test the current state-of-the-art computational methods of calculating neutron flux to reactor components against the measured data of the VENUS-2 MOX-fuelled critical experiments. The measured data to be used for this benchmark are the equivalent fission fluxes which are the reaction rates divided by the U 235 fission spectrum averaged cross-section of the corresponding dosimeter. The present benchmark is, therefore, defined to calculate reaction rates and corresponding equivalent fission fluxes measured on the core-mid plane at specific positions outside the core of the VENUS-2 MOX-fuelled reactor. This is a follow-up exercise to the previously completed UO 2 -fuelled VENUS-1 two-dimensional and VENUS-3 three-dimensional exercises. The use of MOX fuel in LWRs presents different neutron characteristics and this is the main interest of the current benchmark compared to the previous ones

Effects of neutron data libraries and criticality codes on IAEA criticality benchmark problems

International Nuclear Information System (INIS)

Sarker, Md.M.; Takano, Makoto; Masukawa, Fumihiro; Naito, Yoshitaka

1993-10-01

In order to compare the effects of neutron data libraries and criticality codes to thermal reactors (LWR), the IAEA criticality benchmark calculations have been performed. The experiments selected in this study include TRX-1 and TRX-2 with a simple geometric configuration. Reactor lattice calculation codes WIMS-D/4, MCNP-4, JACS (MGCL, KENO), and SRAC were used in the present calculations. The TRX cores were analyzed by WIMS-D/4 using WIMS original library and also by MCNP-4, JACS (MGCL, KENO), and SRAC using the libraries generated from JENDL-3 and ENDF/B-IV nuclear data files. An intercomparison work for the above mentioned code systems and cross section libraries was performed by analyzing the LWR benchmark experiments TRX-1 and TRX-2. The TRX cores were also analyzed for supercritical and subcritical conditions and these results were compared. In the case of critical condition, the results were in good agreement. But for the supercritical and subcritical conditions, the difference of the results obtained by using the different cross section libraries become larger than for the critical condition. (author)

Assessment of the available {sup 233}U cross-section evaluations in the calculation of critical benchmark experiments

Energy Technology Data Exchange (ETDEWEB)

Leal, L.C.; Wright, R.Q.

1996-10-01

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 S{sub n} 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.

Assessment of the Available (Sup 233)U Cross Sections Evaluations in the Calculation of Critical Benchmark Experiments

Energy Technology Data Exchange (ETDEWEB)

Leal, L.C.

1993-01-01

In this report we investigate the adequacy of the available {sup 233}U cross-section data for calculation of experimental critical systems. The {sup 233}U evaluations provided in two evaluated nuclear data libraries, the U. S. Data Bank [ENDF/B (Evaluated Nuclear Data Files)] and the Japanese Data Bank [JENDL (Japanese Evaluated Nuclear Data Library)] are examined. Calculations were performed for six thermal and ten fast experimental critical systems using the Sn transport XSDRNPM code. To verify the performance of the {sup 233}U cross-section data for nuclear criticality safety application in which the neutron energy spectrum is predominantly in the epithermal energy range, calculations of four numerical benchmark systems with energy spectra in the intermediate energy range were done. These calculations serve only as an indication of the difference in calculated results that may be expected when the two {sup 233}U cross-section evaluations are used for problems with neutron spectra in the intermediate energy range. Additionally, comparisons of experimental and calculated central fission rate ratios were also made. The study has suggested that an ad hoc {sup 233}U evaluation based on the JENDL library provides better overall results for both fast and thermal experimental critical systems.

FENDL neutronics benchmark: Specifications for the calculational neutronics and shielding benchmark

International Nuclear Information System (INIS)

Sawan, M.E.

1994-12-01

During the IAEA Advisory Group Meeting on ''Improved Evaluations and Integral Data Testing for FENDL'' held in Garching near Munich, Germany in the period 12-16 September 1994, the Working Group II on ''Experimental and Calculational Benchmarks on Fusion Neutronics for ITER'' recommended that a calculational benchmark representative of the ITER design should be developed. This report describes the neutronics and shielding calculational benchmark available for scientists interested in performing analysis for this benchmark. (author)

OECD/NEA Burnup Credit Calculational Criticality Benchmark Phase I-B Results

International Nuclear Information System (INIS)

DeHart, M.D.

1993-01-01

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 149 Sm, 151 Sm, and 155 Gd

Monte Carlo code criticality benchmark comparisons for waste packaging

International Nuclear Information System (INIS)

Alesso, H.P.; Annese, C.E.; Buck, R.M.; Pearson, J.S.; Lloyd, W.R.

1992-07-01

COG is a new point-wise Monte Carlo code being developed and tested at Lawrence Livermore National Laboratory (LLNL). It solves the Boltzmann equation for the transport of neutrons and photons. The objective of this paper is to report on COG results for criticality benchmark experiments both on a Cray mainframe and on a HP 9000 workstation. COG has been recently ported to workstations to improve its accessibility to a wider community of users. COG has some similarities to a number of other computer codes used in the shielding and criticality community. The recently introduced high performance reduced instruction set (RISC) UNIX workstations provide computational power that approach mainframes at a fraction of the cost. A version of COG is currently being developed for the Hewlett Packard 9000/730 computer with a UNIX operating system. Subsequent porting operations will move COG to SUN, DEC, and IBM workstations. In addition, a CAD system for preparation of the geometry input for COG is being developed. In July 1977, Babcock ampersand Wilcox Co. (B ampersand W) was awarded a contract to conduct a series of critical experiments that simulated close-packed storage of LWR-type fuel. These experiments provided data for benchmarking and validating calculational methods used in predicting K-effective of nuclear fuel storage in close-packed, neutron poisoned arrays. Low enriched UO2 fuel pins in water-moderated lattices in fuel storage represent a challenging criticality calculation for Monte Carlo codes particularly when the fuel pins extend out of the water. COG and KENO calculational results of these criticality benchmark experiments are presented

HEU benchmark calculations and LEU preliminary calculations for IRR-1

International Nuclear Information System (INIS)

Caner, M.; Shapira, M.; Bettan, M.; Nagler, A.; Gilat, J.

2004-01-01

We performed neutronics calculations for the Soreq Research Reactor, IRR-1. The calculations were done for the purpose of upgrading and benchmarking our codes and methods. The codes used were mainly WIMS-D/4 for cell calculations and the three dimensional diffusion code CITATION for full core calculations. The experimental flux was obtained by gold wire activation methods and compared with our calculated flux profile. The IRR-1 is loaded with highly enriched uranium fuel assemblies, of the plate type. In the framework of preparation for conversion to low enrichment fuel, additional calculations were done assuming the presence of LEU fresh fuel. In these preliminary calculations we investigated the effect on the criticality and flux distributions of the increase of U-238 loading, and the corresponding uranium density.(author)

OECD/NEA Burnup Credit Calculational Criticality Benchmark Phase I-B Results

Energy Technology Data Exchange (ETDEWEB)

DeHart, M.D.

1993-01-01

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 {sup 149}Sm, {sup 151}Sm, and {sup 155}Gd.

RECENT ADDITIONS OF CRITICALITY SAFETY RELATED INTEGRAL BENCHMARK DATA TO THE ICSBEP AND IRPHEP HANDBOOKS

Energy Technology Data Exchange (ETDEWEB)

J. Blair Briggs; Lori Scott; Yolanda Rugama; Enrico Sartori

2009-09-01

High-quality integral benchmark experiments have always been a priority for criticality safety. However, interest in integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of future criticality safety needs to support next generation reactor and advanced fuel cycle concepts. The importance of drawing upon existing benchmark data is becoming more apparent because of dwindling availability of critical facilities worldwide and the high cost of performing new experiments. Integral benchmark data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the International Handbook of Reactor Physics Benchmark Experiments are widely used. Benchmark data have been added to these two handbooks since the last Nuclear Criticality Safety Division Topical Meeting in Knoxville, Tennessee (September 2005). This paper highlights these additions.

Recent additions of criticality safety related integral benchmark data to the ICSBEP and IRPHEP handbooks

International Nuclear Information System (INIS)

Briggs, J. B.; Scott, L.; Rugama, Y.; Sartori, E.

2009-01-01

High-quality integral benchmark experiments have always been a priority for criticality safety. However, interest in integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of future criticality safety needs to support next generation reactor and advanced fuel cycle concepts. The importance of drawing upon existing benchmark data is becoming more apparent because of dwindling availability of critical facilities worldwide and the high cost of performing new experiments. Integral benchmark data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the International Handbook of Reactor Physics Benchmark Experiments are widely used. Benchmark data have been added to these two handbooks since the last Nuclear Criticality Safety Division Topical Meeting in Knoxville, Tennessee (September 2005). This paper highlights these additions. (authors)

REcent Additions Of Criticality Safety Related Integral Benchmark Data To The Icsbep And Irphep Handbooks

International Nuclear Information System (INIS)

Briggs, J. Blair; Scott, Lori; Rugama, Yolanda; Sartori, Enrico

2009-01-01

High-quality integral benchmark experiments have always been a priority for criticality safety. However, interest in integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of future criticality safety needs to support next generation reactor and advanced fuel cycle concepts. The importance of drawing upon existing benchmark data is becoming more apparent because of dwindling availability of critical facilities worldwide and the high cost of performing new experiments. Integral benchmark data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the International Handbook of Reactor Physics Benchmark Experiments are widely used. Benchmark data have been added to these two handbooks since the last Nuclear Criticality Safety Division Topical Meeting in Knoxville, Tennessee (September 2005). This paper highlights these additions.

Criticality benchmarks for COG: A new point-wise Monte Carlo code

International Nuclear Information System (INIS)

Alesso, H.P.; Pearson, J.; Choi, J.S.

1989-01-01

COG is a new point-wise Monte Carlo code being developed and tested at LLNL for the Cray computer. It solves the Boltzmann equation for the transport of neutrons, photons, and (in future versions) charged particles. Techniques included in the code for modifying the random walk of particles make COG most suitable for solving deep-penetration (shielding) problems. However, its point-wise cross-sections also make it effective for a wide variety of criticality problems. COG has some similarities to a number of other computer codes used in the shielding and criticality community. These include the Lawrence Livermore National Laboratory (LLNL) codes TART and ALICE, the Los Alamos National Laboratory code MCNP, the Oak Ridge National Laboratory codes 05R, 06R, KENO, and MORSE, the SACLAY code TRIPOLI, and the MAGI code SAM. Each code is a little different in its geometry input and its random-walk modification options. Validating COG consists in part of running benchmark calculations against critical experiments as well as other codes. The objective of this paper is to present calculational results of a variety of critical benchmark experiments using COG, and to present the resulting code bias. Numerous benchmark calculations have been completed for a wide variety of critical experiments which generally involve both simple and complex physical problems. The COG results, which they report in this paper, have been excellent

OECD/NEA expert group on uncertainty analysis for criticality safety assessment: Results of benchmark on sensitivity calculation (phase III)

Energy Technology Data Exchange (ETDEWEB)

Ivanova, T.; Laville, C. [Institut de Radioprotection et de Surete Nucleaire IRSN, BP 17, 92262 Fontenay aux Roses (France); Dyrda, J. [Atomic Weapons Establishment AWE, Aldermaston, Reading, RG7 4PR (United Kingdom); Mennerdahl, D. [E Mennerdahl Systems EMS, Starvaegen 12, 18357 Taeby (Sweden); Golovko, Y.; Raskach, K.; Tsiboulia, A. [Inst. for Physics and Power Engineering IPPE, 1, Bondarenko sq., 249033 Obninsk (Russian Federation); Lee, G. S.; Woo, S. W. [Korea Inst. of Nuclear Safety KINS, 62 Gwahak-ro, Yuseong-gu, Daejeon 305-338 (Korea, Republic of); Bidaud, A.; Sabouri, P. [Laboratoire de Physique Subatomique et de Cosmologie LPSC, CNRS-IN2P3/UJF/INPG, Grenoble (France); Patel, A. [U.S. Nuclear Regulatory Commission (NRC), Washington, DC 20555-0001 (United States); Bledsoe, K.; Rearden, B. [Oak Ridge National Laboratory ORNL, M.S. 6170, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Gulliford, J.; Michel-Sendis, F. [OECD/NEA, 12, Bd des Iles, 92130 Issy-les-Moulineaux (France)

2012-07-01

The sensitivities of the k{sub eff} eigenvalue to neutron cross sections have become commonly used in similarity studies and as part of the validation algorithm for criticality safety assessments. To test calculations of the sensitivity coefficients, a benchmark study (Phase III) has been established by the OECD-NEA/WPNCS/EG UACSA (Expert Group on Uncertainty Analysis for Criticality Safety Assessment). This paper presents some sensitivity results generated by the benchmark participants using various computational tools based upon different computational methods: SCALE/TSUNAMI-3D and -1D, MONK, APOLLO2-MORET 5, DRAGON-SUSD3D and MMKKENO. The study demonstrates the performance of the tools. It also illustrates how model simplifications impact the sensitivity results and demonstrates the importance of 'implicit' (self-shielding) sensitivities. This work has been a useful step towards verification of the existing and developed sensitivity analysis methods. (authors)

IAEA sodium void reactivity benchmark calculations

International Nuclear Information System (INIS)

Hill, R.N.; Finck, P.J.

1992-01-01

In this paper, the IAEA-1 992 ''Benchmark Calculation of Sodium Void Reactivity Effect in Fast Reactor Core'' problem is evaluated. The proposed design is a large axially heterogeneous oxide-fueled fast reactor as described in Section 2; the core utilizes a sodium plenum above the core to enhance leakage effects. The calculation methods used in this benchmark evaluation are described in Section 3. In Section 4, the calculated core performance results for the benchmark reactor model are presented; and in Section 5, the influence of steel and interstitial sodium heterogeneity effects is estimated

One dimensional benchmark calculations using diffusion theory

International Nuclear Information System (INIS)

Ustun, G.; Turgut, M.H.

1986-01-01

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)

Criticality Benchmark Results Using Various MCNP Data Libraries

International Nuclear Information System (INIS)

Frankle, Stephanie C.

1999-01-01

A suite of 86 criticality benchmarks has been recently implemented in MCNPtrademark as part of the nuclear data validation effort. These benchmarks have been run using two sets of MCNP continuous-energy neutron data: ENDF/B-VI based data through Release 2 (ENDF60) and the ENDF/B-V based data. New evaluations were completed for ENDF/B-VI for a number of the important nuclides such as the isotopes of H, Be, C, N, O, Fe, Ni, 235,238 U, 237 Np, and 239,240 Pu. When examining the results of these calculations for the five manor categories of 233 U, intermediate-enriched 235 U (IEU), highly enriched 235 U (HEU), 239 Pu, and mixed metal assembles, we find the following: (1) The new evaluations for 9 Be, 12 C, and 14 N show no net effect on k eff ; (2) There is a consistent decrease in k eff for all of the solution assemblies for ENDF/B-VI due to 1 H and 16 O, moving k eff further from the benchmark value for uranium solutions and closer to the benchmark value for plutonium solutions; (3) k eff decreased for the ENDF/B-VI Fe isotopic data, moving the calculated k eff further from the benchmark value; (4) k eff decreased for the ENDF/B-VI Ni isotopic data, moving the calculated k eff closer to the benchmark value; (5) The W data remained unchanged and tended to calculate slightly higher than the benchmark values; (6) For metal uranium systems, the ENDF/B-VI data for 235 U tends to decrease k eff while the 238 U data tends to increase k eff . The net result depends on the energy spectrum and material specifications for the particular assembly; (7) For more intermediate-energy systems, the changes in the 235,238 U evaluations tend to increase k eff . For the mixed graphite and normal uranium-reflected assembly, a large increase in k eff due to changes in the 238 U evaluation moved the calculated k eff much closer to the benchmark value. (8) There is little change in k eff for the uranium solutions due to the new 235,238 U evaluations; and (9) There is little change in k eff

Benchmarking of HEU mental annuli critical assemblies with internally reflected graphite cylinder

Directory of Open Access Journals (Sweden)

Xiaobo Liu

2017-01-01

Full Text Available Three experimental configurations of critical assemblies, performed in 1963 at the Oak Ridge Critical Experiment Facility, which are assembled using three different diameter HEU annuli (15-9 inches, 15-7 inches and 13-7 inches metal annuli with internally reflected graphite cylinder are evaluated and benchmarked. The experimental uncertainties which are 0.00057, 0.00058 and 0.00057 respectively, and biases to the benchmark models which are − 0.00286, − 0.00242 and − 0.00168 respectively, were determined, and the experimental benchmark keff results were obtained for both detailed and simplified models. The calculation results for both detailed and simplified models using MCNP6-1.0 and ENDF/B-VII.1 agree well to the benchmark experimental results within difference less than 0.2%. The benchmarking results were accepted for the inclusion of ICSBEP Handbook.

Analysis and evaluation of critical experiments for validation of neutron transport calculations

International Nuclear Information System (INIS)

Bazzana, S.; Blaumann, H; Marquez Damian, J.I

2009-01-01

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. [es

INTEGRAL BENCHMARKS AVAILABLE THROUGH THE INTERNATIONAL REACTOR PHYSICS EXPERIMENT EVALUATION PROJECT AND THE INTERNATIONAL CRITICALITY SAFETY BENCHMARK EVALUATION PROJECT

Energy Technology Data Exchange (ETDEWEB)

J. Blair Briggs; Lori Scott; Enrico Sartori; Yolanda Rugama

2008-09-01

Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) continue to expand their efforts and broaden their scope to identify, evaluate, and provide integral benchmark data for method and data validation. Benchmark model specifications provided by these two projects are used heavily by the international reactor physics, nuclear data, and criticality safety communities. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. The status of the IRPhEP and ICSBEP is discussed in this paper, and the future of the two projects is outlined and discussed. Selected benchmarks that have been added to the IRPhEP and ICSBEP handbooks since PHYSOR’06 are highlighted, and the future of the two projects is discussed.

INTEGRAL BENCHMARKS AVAILABLE THROUGH THE INTERNATIONAL REACTOR PHYSICS EXPERIMENT EVALUATION PROJECT AND THE INTERNATIONAL CRITICALITY SAFETY BENCHMARK EVALUATION PROJECT

International Nuclear Information System (INIS)

J. Blair Briggs; Lori Scott; Enrico Sartori; Yolanda Rugama

2008-01-01

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

Validating analysis methodologies used in burnup credit criticality calculations

International Nuclear Information System (INIS)

Brady, M.C.; Napolitano, D.G.

1992-01-01

The concept of allowing reactivity credit for the depleted (or burned) state of pressurized water reactor fuel in the licensing of spent fuel facilities introduces a new challenge to members of the nuclear criticality community. The primary difference in this analysis approach is the technical ability to calculate spent fuel compositions (or inventories) and to predict their effect on the system multiplication factor. Isotopic prediction codes are used routinely for in-core physics calculations and the prediction of radiation source terms for both thermal and shielding analyses, but represent an innovation for criticality specialists. This paper discusses two methodologies currently being developed to specifically evaluate isotopic composition and reactivity for the burnup credit concept. A comprehensive approach to benchmarking and validating the methods is also presented. This approach involves the analysis of commercial reactor critical data, fuel storage critical experiments, chemical assay isotopic data, and numerical benchmark calculations

Benchmark calculation of nuclear design code for HCLWR

International Nuclear Information System (INIS)

Suzuki, Katsuo; Saji, Etsuro; Gakuhari, Kazuhiko; Akie, Hiroshi; Takano, Hideki; Ishiguro, Yukio.

1986-01-01

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)

Criticality safety benchmark experiment on 10% enriched uranyl nitrate solution using a 28-cm-thickness slab core

International Nuclear Information System (INIS)

Yamamoto, Toshihiro; Miyoshi, Yoshinori; Kikuchi, Tsukasa; Watanabe, Shouichi

2002-01-01

The second series of critical experiments with 10% enriched uranyl nitrate solution using 28-cm-thick slab core have been performed with the Static Experiment Critical Facility of the Japan Atomic Energy Research Institute. Systematic critical data were obtained by changing the uranium concentration of the fuel solution from 464 to 300 gU/l under various reflector conditions. In this paper, the thirteen critical configurations for water-reflected cores and unreflected cores are identified and evaluated. The effects of uncertainties in the experimental data on k eff are quantified by sensitivity studies. Benchmark model specifications that are necessary to construct a calculational model are given. The uncertainties of k eff 's included in the benchmark model specifications are approximately 0.1%Δk eff . The thirteen critical configurations are judged to be acceptable benchmark data. Using the benchmark model specifications, sample calculation results are provided with several sets of standard codes and cross section data. (author)

Benchmark calculation of subchannel analysis codes

International Nuclear Information System (INIS)

1996-02-01

In order to evaluate the analysis capabilities of various subchannel codes used in thermal-hydraulic design of light water reactors, benchmark calculations were performed. The selected benchmark problems and major findings obtained by the calculations were as follows: (1)As for single-phase flow mixing experiments between two channels, the calculated results of water temperature distribution along the flow direction were agreed with experimental results by tuning turbulent mixing coefficients properly. However, the effect of gap width observed in the experiments could not be predicted by the subchannel codes. (2)As for two-phase flow mixing experiments between two channels, in high water flow rate cases, the calculated distributions of air and water flows in each channel were well agreed with the experimental results. In low water flow cases, on the other hand, the air mixing rates were underestimated. (3)As for two-phase flow mixing experiments among multi-channels, the calculated mass velocities at channel exit under steady-state condition were agreed with experimental values within about 10%. However, the predictive errors of exit qualities were as high as 30%. (4)As for critical heat flux(CHF) experiments, two different results were obtained. A code indicated that the calculated CHF's using KfK or EPRI correlations were well agreed with the experimental results, while another code suggested that the CHF's were well predicted by using WSC-2 correlation or Weisman-Pei mechanistic model. (5)As for droplets entrainment and deposition experiments, it was indicated that the predictive capability was significantly increased by improving correlations. On the other hand, a remarkable discrepancy between codes was observed. That is, a code underestimated the droplet flow rate and overestimated the liquid film flow rate in high quality cases, while another code overestimated the droplet flow rate and underestimated the liquid film flow rate in low quality cases. (J.P.N.)

MCNP calculations for criticality-safety benchmarks with ENDF/B-V and ENDF/B-VI libraries

International Nuclear Information System (INIS)

Iverson, J.L.; Mosteller, R.D.

1995-01-01

The MCNP Monte Carlo code, in conjunction with its continuous-energy ENDF/B-V and ENDF/B-VI cross-section libraries, has been benchmarked against results from 27 different critical experiments. The predicted values of k eff are in excellent agreement with the benchmarks, except for the ENDF/B-V results for solutions of plutonium nitrate and, to a lesser degree, for the ENDF/B-V and ENDF/B-VI results for a bare sphere of 233 U

Benchmark calculations by KENO-Va using the JEF 2.2 library

Energy Technology Data Exchange (ETDEWEB)

Markova, L.

1994-12-01

This work has to be a contribution to the validation of the JEF2.2 neutron cross-section libarary, following the earlier published benchmark calculations having been performed to validate the previous version JEF1.1 of the libarary. Several simple calculational problems and one experimental problem were chosen for a criticality calculations. In addition also a realistic hexagonal arrangement of the VVER-440 fuel assemblies in a spent fuel cask were analyzed in a partly cylindrized model. All criticality calculations, carried out by the KENO-Va code using the JEF2.2 neutron cross-section library in 172 energy groups, resulted in multiplication factors (k{sub eff}) which were tabulated and compared with the results of other available calculations of the same problems. (orig.).

Benchmark calculation for water reflected STACY cores containing low enriched uranyl nitrate solution

International Nuclear Information System (INIS)

Miyoshi, Yoshinori; Yamamoto, Toshihiro; Nakamura, Takemi

2001-01-01

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)

Standard Guide for Benchmark Testing of Light Water Reactor Calculations

CERN Document Server

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 ...

Benchmark calculations of power distribution within assemblies

International Nuclear Information System (INIS)

Cavarec, C.; Perron, J.F.; Verwaerde, D.; West, J.P.

1994-09-01

The main objective of this Benchmark is to compare different techniques for fine flux prediction based upon coarse mesh diffusion or transport calculations. We proposed 5 ''core'' configurations including different assembly types (17 x 17 pins, ''uranium'', ''absorber'' or ''MOX'' assemblies), with different boundary conditions. The specification required results in terms of reactivity, pin by pin fluxes and production rate distributions. The proposal for these Benchmark calculations was made by J.C. LEFEBVRE, J. MONDOT, J.P. WEST and the specification (with nuclear data, assembly types, core configurations for 2D geometry and results presentation) was distributed to correspondents of the OECD Nuclear Energy Agency. 11 countries and 19 companies answered the exercise proposed by this Benchmark. Heterogeneous calculations and homogeneous calculations were made. Various methods were used to produce the results: diffusion (finite differences, nodal...), transport (P ij , S n , Monte Carlo). This report presents an analysis and intercomparisons of all the results received

Benchmark calculations on nuclear characteristics of JRR-4 HEU core by SRAC code system

International Nuclear Information System (INIS)

Arigane, Kenji

1987-04-01

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)

Criticality benchmark guide for light-water-reactor fuel in transportation and storage packages

International Nuclear Information System (INIS)

Lichtenwalter, J.J.; Bowman, S.M.; DeHart, M.D.; Hopper, C.M.

1997-03-01

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 calculations for fusion blanket development

International Nuclear Information System (INIS)

Sawan, M.E.; Cheng, E.T.

1985-01-01

Benchmark problems representing the leading fusion blanket concepts are presented. Benchmark calculations for self-cooled Li/sub 17/Pb/sub 83/ 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 TBR to group structure and weighting spectrum increases and Li enrichment decrease with up to 20% discrepancies for thin natural Li/sub 17/Pb/sub 83/ blankets

Benchmark assemblies of the Los Alamos critical assemblies facility

International Nuclear Information System (INIS)

Dowdy, E.J.

1986-01-01

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. (author)

Benchmark assemblies of the Los Alamos Critical Assemblies Facility

International Nuclear Information System (INIS)

Dowdy, E.J.

1985-01-01

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

Benchmark assemblies of the Los Alamos critical assemblies facility

International Nuclear Information System (INIS)

Dowdy, E.J.

1985-01-01

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

Benchmark calculations for fusion blanket development

International Nuclear Information System (INIS)

Sawan, M.L.; Cheng, E.T.

1986-01-01

Benchmark problems representing the leading fusion blanket concepts are presented. Benchmark calculations for self-cooled Li 17 Pb 83 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 Li 17 Pb 83 blankets. (author)

ENDF/B-VII.1 Neutron Cross Section Data Testing with Critical Assembly Benchmarks and Reactor Experiments

Energy Technology Data Exchange (ETDEWEB)

Kahler, A. [Los Alamos National Laboratory (LANL); Macfarlane, R E [Los Alamos National Laboratory (LANL); Mosteller, R D [Los Alamos National Laboratory (LANL); Kiedrowski, B C [Los Alamos National Laboratory (LANL); Frankle, S C [Los Alamos National Laboratory (LANL); Chadwick, M. B. [Los Alamos National Laboratory (LANL); Mcknight, R D [Argonne National Laboratory (ANL); Lell, R M [Argonne National Laboratory (ANL); Palmiotti, G [Idaho National Laboratory (INL); Hiruta, h [Idaho National Laboratory (INL); Herman, Micheal W [Brookhaven National Laboratory (BNL); Arcilla, r [Brookhaven National Laboratory (BNL); Mughabghab, S F [Brookhaven National Laboratory (BNL); Sublet, J C [Culham Science Center, Abington, UK; Trkov, A. [Jozef Stefan Institute, Slovenia; Trumbull, T H [Knolls Atomic Power Laboratory; Dunn, Michael E [ORNL

2011-01-01

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 423 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 unrnoderated and uranium reflected (235)U and (239)Pu 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 (236)U; (238,242)Pu and (241,243)Am capture in fast systems. Other deficiencies, such as the overprediction of Pu solution system critical

Performance assessment of new neutron cross section libraries using MCNP code and some critical benchmarks

International Nuclear Information System (INIS)

Bakkari, B El; Bardouni, T El.; Erradi, L.; Chakir, E.; Meroun, O.; Azahra, M.; Boukhal, H.; Khoukhi, T El.; Htet, A.

2007-01-01

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 [fr

New calculations for critical assemblies using MCNP4B

International Nuclear Information System (INIS)

Adams, A.A.; Frankle, S.C.; Little, R.C.

1997-07-01

A suite of 41 criticality benchmarks has been modeled using MCNP trademark (version 4B). Most of the assembly specifications were obtained from the Cross Section Evaluation Working Group (CSEWG) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) compendiums of experimental benchmarks. A few assembly specifications were obtained from experimental papers. The suite contains thermal and fast assemblies, bare and reflected assemblies, and emphasizes 233 U, 235 U, 238 U, and 239 Pu. The values of k eff for each assembly in the suite were calculated using MCNP libraries derived primarily from release 2 of ENDF/B-V and release 2 of ENDF/B-VI. The results show that the new ENDF/B-VI.2 evaluations for H, O, N, B, 235 U, 238 U, and 239 Pu can have a significant impact on the values of k eff . In addition to the integral quantity k eff , several additional experimental measurements were performed and documented. These experimental measurements include central fission and reaction-rate ratios for various isotopes, and neutron leakage and flux spectra. They provide more detailed information about the accuracy of the nuclear data than can k eff . Comparison calculations were performed using both ENDF/B-V.2 and ENDF/B-VI.2-based data libraries. The purpose of this paper is to compare the results of these additional calculations with experimental data, and to use these results to assess the quality of the nuclear data

Calculation of the 5th AER dynamic benchmark with APROS

International Nuclear Information System (INIS)

Puska, E.K.; Kontio, H.

1998-01-01

The model used for calculation of the 5th AER dynamic benchmark with APROS code is presented. In the calculation of the 5th AER dynamic benchmark the three-dimensional neutronics model of APROS was used. The core was divided axially into 20 nodes according to the specifications of the benchmark and each six identical fuel assemblies were placed into one one-dimensional thermal hydraulic channel. The five-equation thermal hydraulic model was used in the benchmark. The plant process and automation was described with a generic VVER-440 plant model created by IVO PE. (author)

Calculation of Upper Subcritical Limits for Nuclear Criticality in a Repository

International Nuclear Information System (INIS)

J.W. Pegram

1998-01-01

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, k eff 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 k eff . 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 k eff 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 (k eff ). 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

Benchmark calculations for critical experiments at FKBN-M facility with uranium-plutonium-polyethylene systems using JENDL-3.2 and MVP Monte-Carlo code

International Nuclear Information System (INIS)

Obara, Toru; Morozov, A.G.; Kevrolev, V.V.; Kuznetsov, V.V.; Treschalin, S.A.; Lukin, A.V.; Terekhin, V.A.; Sokolov, Yu.A.; Kravchenko, V.G.

2000-01-01

Benchmark calculations were performed for critical experiments at FKBN-M facility in RFNC-VNIITF, Russia using JENDL-3.2 nuclear data library and continuous energy Monte-Carlo code MVP. The fissile materials were high-enriched uranium and plutonium. Polyethylene was used as moderator. The neutron spectrum was changed by changing the geometry. Calculation results by MVP showed some errors. Discussion was made by reaction rates and η values obtained by MVP. It showed the possibility that cross sections of U-235 had different trend of error in fast and thermal energy region respectively. It also showed the possibility of some error of cross section of Pu-239 in high energy region. (author)

Validation of the Continuous-Energy Monte Carlo Criticality-Safety Analysis System MVP and JENDL-3.2 Using the Internationally Evaluated Criticality Benchmarks

International Nuclear Information System (INIS)

Mitake, Susumu

2003-01-01

Validation of the continuous-energy Monte Carlo criticality-safety analysis system, comprising the MVP code and neutron cross sections based on JENDL-3.2, was examined using benchmarks evaluated in the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments'. Eight experiments (116 configurations) for the plutonium solution and plutonium-uranium mixture systems performed at Valduc, Battelle Pacific Northwest Laboratories, and other facilities were selected and used in the studies. The averaged multiplication factors calculated with MVP and MCNP-4B using the same neutron cross-section libraries based on JENDL-3.2 were in good agreement. Based on methods provided in the Japanese nuclear criticality-safety handbook, the estimated criticality lower-limit multiplication factors to be used as a subcriticality criterion for the criticality-safety evaluation of nuclear facilities were obtained. The analysis proved the applicability of the MVP code to the criticality-safety analysis of nuclear fuel facilities, particularly to the analysis of systems fueled with plutonium and in homogeneous and thermal-energy conditions

OECD/NEA burnup credit criticality benchmarks phase IIIB: Burnup calculations of BWR fuel assemblies for storage and transport

International Nuclear Information System (INIS)

Okuno, Hiroshi; Naito, Yoshitaka; Suyama, Kenya

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. 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 ∞ 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)

OECD/NEA burnup credit criticality benchmarks phase IIIB. Burnup calculations of BWR fuel assemblies for storage and transport

Energy Technology Data Exchange (ETDEWEB)

Okuno, Hiroshi; Naito, Yoshitaka; Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-02-01

The report describes the final results of the Phase IIIB Benchmark conducted by the Expert Group on Burnup Credit Criticality Safety under the auspices of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD). The Benchmark was intended to compare the predictability of current computer code and data library combinations for the atomic number densities of an irradiated PWR fuel assembly model. The fuel assembly was irradiated under specific power of 25.6 MW/tHM up to 40 GWd/tHM and cooled for five years. The void fraction was assumed to be uniform throughout the channel box and constant, at 0, 40 and 70%, during burnup. In total, 16 results were submitted from 13 institutes of 7 countries. The calculated atomic number densities of 12 actinides and 20 fission product nuclides were found to be for the most part within a range of {+-}10% relative to the average, although some results, esp. {sup 155}Eu and gadolinium isotopes, exceeded the band, which will require further investigation. Pin-wise burnup results agreed well among the participants. The results in the infinite neutron multiplication factor k{sub {infinity}} also accorded well with each other for void fractions of 0 and 40%; however some results deviated from the averaged value noticeably for the void fraction of 70%. (author)

MCNP benchmark analyses of critical experiments for the Space Nuclear Thermal Propulsion program

International Nuclear Information System (INIS)

Selcow, E.C.; Cerbone, R.J.; Ludewig, H.; Mughabghab, S.F.; Schmidt, E.; Todosow, M.; Parma, E.J.; Ball, R.M.; Hoovler, G.S.

1993-01-01

Benchmark analyses have been performed of Particle Bed Reactor (PBR) critical experiments (CX) using the MCNP radiation transport code. The experiments have been conducted at the Sandia National Laboratory reactor facility in support of the Space Nuclear Thermal Propulsion (SNTP) program. The test reactor is a nineteen element water moderated and reflected thermal system. A series of integral experiments have been carried out to test the capabilities of the radiation transport codes to predict the performance of PBR systems. MCNP was selected as the preferred radiation analysis tool for the benchmark experiments. Comparison between experimental and calculational results indicate close agreement. This paper describes the analyses of benchmark experiments designed to quantify the accuracy of the MCNP radiation transport code for predicting the performance characteristics of PBR reactors

MCNP benchmark analyses of critical experiments for the Space Nuclear Thermal Propulsion program

Science.gov (United States)

Selcow, Elizabeth C.; Cerbone, Ralph J.; Ludewig, Hans; Mughabghab, Said F.; Schmidt, Eldon; Todosow, Michael; Parma, Edward J.; Ball, Russell M.; Hoovler, Gary S.

1993-01-01

Benchmark analyses have been performed of Particle Bed Reactor (PBR) critical experiments (CX) using the MCNP radiation transport code. The experiments have been conducted at the Sandia National Laboratory reactor facility in support of the Space Nuclear Thermal Propulsion (SNTP) program. The test reactor is a nineteen element water moderated and reflected thermal system. A series of integral experiments have been carried out to test the capabilities of the radiation transport codes to predict the performance of PBR systems. MCNP was selected as the preferred radiation analysis tool for the benchmark experiments. Comparison between experimental and calculational results indicate close agreement. This paper describes the analyses of benchmark experiments designed to quantify the accuracy of the MCNP radiation transport code for predicting the performance characteristics of PBR reactors.

Development and validation of a criticality calculation scheme based on French deterministic transport codes

International Nuclear Information System (INIS)

Santamarina, A.

1991-01-01

A criticality-safety calculational scheme using the automated deterministic code system, APOLLO-BISTRO, has been developed. The cell/assembly code APOLLO is used mainly in LWR and HCR design calculations, and its validation spans a wide range of moderation ratios, including voided configurations. Its recent 99-group library and self-shielded cross-sections has been extensively qualified through critical experiments and PWR spent fuel analysis. The PIC self-shielding formalism enables a rigorous treatment of the fuel double heterogeneity in dissolver medium calculations. BISTRO is an optimized multidimensional SN code, part of the modular CCRR package used mainly in FBR calculations. The APOLLO-BISTRO scheme was applied to the 18 experimental benchmarks selected by the OECD/NEACRP Criticality Calculation Working Group. The Calculation-Experiment discrepancy was within ± 1% in ΔK/K and always looked consistent with the experimental uncertainty margin. In the critical experiments corresponding to a dissolver type benchmark, our tools computed a satisfactory Keff. In the VALDUC fuel storage experiments, with hafnium plates, the computed Keff ranged between 0.994 and 1.003 for the various watergaps spacing the fuel clusters from the absorber plates. The APOLLO-KENOEUR statistic calculational scheme, based on the same self-shielded multigroup library, supplied consistent results within 0.3% in ΔK/K. (Author)

ENDF/B-VII.1 Neutron Cross Section Data Testing with Critical Assembly Benchmarks and Reactor Experiments

Energy Technology Data Exchange (ETDEWEB)

Kahler, A.C.; Herman, M.; Kahler,A.C.; MacFarlane,R.E.; Mosteller,R.D.; Kiedrowski,B.C.; Frankle,S.C.; Chadwick,M.B.; McKnight,R.D.; Lell,R.M.; Palmiotti,G.; Hiruta,H.; Herman,M.; Arcilla,R.; Mughabghab,S.F.; Sublet,J.C.; Trkov,A.; Trumbull,T.H.; Dunn,M.

2011-12-01

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 423 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 [M. B. Chadwick et al., 'ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data,' Nuclear Data Sheets, 112, 2887 (2011)]. 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 {sup 235}U and {sup 239}Pu 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

WIPP Benchmark calculations with the large strain SPECTROM codes

International Nuclear Information System (INIS)

Callahan, G.D.; DeVries, K.L.

1995-08-01

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

Analysis of the international criticality benchmark no 19 of a realistic fuel dissolver

International Nuclear Information System (INIS)

Smith, H.J.; Santamarina, A.

1991-01-01

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 238 U 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 density functional theory calculations for nanoscale conductance

DEFF Research Database (Denmark)

Strange, Mikkel; Bækgaard, Iben Sig Buur; Thygesen, Kristian Sommer

2008-01-01

We present a set of benchmark calculations for the Kohn-Sham elastic transmission function of five representative single-molecule junctions. The transmission functions are calculated using two different density functional theory methods, namely an ultrasoft pseudopotential plane-wave code...

Calculation of the fifth atomic energy research dynamic benchmark with APROS

International Nuclear Information System (INIS)

Puska Eija Karita; Kontio Harii

1998-01-01

The band-out presents the model used for calculation of the fifth atomic energy research dynamic benchmark with APROS code. In the calculation of the fifth atomic energy research dynamic benchmark the three-dimensional neutronics model of APROS was used. The core was divided axially into 20 nodes according to the specifications of the benchmark and each six identical fuel assemblies were placed into one one-dimensional thermal hydraulic channel. The five-equation thermal hydraulic model was used in the benchmark. The plant process and automation was described with a generic WWER-440 plant model created by IVO Power Engineering Ltd. - Finland. (Author)

The fifth AER dynamic benchmark calculation with hextran-smabre

International Nuclear Information System (INIS)

Haemaelaeinen, A.; Kyrki-Rajamaeki, R.

1998-01-01

The first AER benchmark for coupling of the thermohydraulic codes and three-dimensional reactordynamic core models is discussed. 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 VVER-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 1/6 symmetry is used in the core. 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)

Benchmark neutron porosity log calculations

International Nuclear Information System (INIS)

Little, R.C.; Michael, M.; Verghese, K.; Gardner, R.P.

1989-01-01

Calculations have been made for a benchmark neutron porosity log problem with the general purpose Monte Carlo code MCNP and the specific purpose Monte Carlo code McDNL. For accuracy and timing comparison purposes the CRAY XMP and MicroVax II computers have been used with these codes. The CRAY has been used for an analog version of the MCNP code while the MicroVax II has been used for the optimized variance reduction versions of both codes. Results indicate that the two codes give the same results within calculated standard deviations. Comparisons are given and discussed for accuracy (precision) and computation times for the two codes

Reactor calculation benchmark PCA blind test results

International Nuclear Information System (INIS)

Kam, F.B.K.; Stallmann, F.W.

1980-01-01

Further improvement in calculational procedures or a combination of calculations and measurements is necessary to attain 10 to 15% (1 sigma) accuracy for neutron exposure parameters (flux greater than 0.1 MeV, flux greater than 1.0 MeV, and dpa). The calculational modeling of power reactors should be benchmarked in an actual LWR plant to provide final uncertainty estimates for end-of-life predictions and limitations for plant operations. 26 references, 14 figures, 6 tables

Reactor calculation benchmark PCA blind test results

Energy Technology Data Exchange (ETDEWEB)

Kam, F.B.K.; Stallmann, F.W.

1980-01-01

Further improvement in calculational procedures or a combination of calculations and measurements is necessary to attain 10 to 15% (1 sigma) accuracy for neutron exposure parameters (flux greater than 0.1 MeV, flux greater than 1.0 MeV, and dpa). The calculational modeling of power reactors should be benchmarked in an actual LWR plant to provide final uncertainty estimates for end-of-life predictions and limitations for plant operations. 26 references, 14 figures, 6 tables.

JNC results of BN-600 benchmark calculation (phase 4)

International Nuclear Information System (INIS)

Ishikawa, Makoto

2003-01-01

The present work is the results of JNC, Japan, for the Phase 4 of the BN-600 core benchmark problem (Hex-Z fully MOX fuelled core model) organized by IAEA. The benchmark specification is based on 1) the RCM report of IAEA CRP on 'Updated Codes and Methods to Reduce the Calculational Uncertainties of LMFR Reactivity Effects, Action 3.12' (Calculations for BN-600 fully fuelled MOX core for subsequent transient analyses). JENDL-3.2 nuclear data library was used for calculating 70 group ABBN-type group constants. Cell models for fuel assembly and control rod calculations were applied: homogeneous and heterogeneous (cylindrical supercell) model. Basic diffusion calculation was three-dimensional Hex-Z model, 18 group (Citation code). Transport calculations were 18 group, three-dimensional (NSHEC code) based on Sn-transport nodal method developed at JNC. The generated thermal power per fission was based on Sher's data corrected on the basis of ENDF/B-IV data library. Calculation results are presented in Tables for intercomparison

WWER-1000 Burnup Credit Benchmark (CB5)

International Nuclear Information System (INIS)

Manolova, M.A.

2002-01-01

In the paper the specification of WWER-1000 Burnup Credit Benchmark first phase (depletion calculations), given. The second phase - criticality calculations for the WWER-1000 fuel pin cell, will be given after the evaluation of the results, obtained at the first phase. The proposed benchmark is a continuation of the WWER benchmark activities in this field (Author)

Preparation of a criticality benchmark based on experiments performed at the RA-6 reactor

International Nuclear Information System (INIS)

Bazzana, S.; Blaumann, H; Marquez Damian, J.I

2009-01-01

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 U 3 Si 2 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. [es

Validation of new 240Pu cross section and covariance data via criticality calculation

International Nuclear Information System (INIS)

Kim, Do Heon; Gil, Choong-Sup; Kim, Hyeong Il; Lee, Young-Ouk; Leal, Luiz C.; Dunn, Michael E.

2011-01-01

Recent collaboration between KAERI and ORNL has completed an evaluation for 240 Pu neutron cross section with covariance data. The new 240 Pu cross section data has been validated through 28 criticality safety benchmark problems taken from the ICSBEP and/or CSEWG specifications with MCNP calculations. The calculation results based on the new evaluation have been compared with those based on recent evaluations such as ENDF/B-VII.0, JEFF-3.1.1, and JENDL-4.0. In addition, the new 240 Pu covariance data has been tested for some criticality benchmarks via the DANTSYS/SUSD3D-based nuclear data sensitivity and uncertainty analysis of k eff . The k eff uncertainty estimates by the new covariance data has been compared with those by JENDL-4.0, JENDL-3.3, and Low-Fidelity covariance data. (author)

Criticality Analysis Of TCA Critical Lattices With MNCP-4C Monte Carlo Calculation

International Nuclear Information System (INIS)

Zuhair

2002-01-01

The use of uranium-plutonium mixed oxide (MOX) fuel in electric generation light water reactor (PWR, BWR) is being planned in Japan. Therefore, the accuracy evaluations of neutronic analysis code for MOX cores have been employed by many scientists and reactor physicists. Benchmark evaluations for TCA was done using various calculation methods. The Monte Carlo become the most reliable method to predict criticality of various reactor types. In this analysis, the MCNP-4C code was chosen because various superiorities the code has. All in all, the MCNP-4C calculation for TCA core with 38 MOX critical lattice configurations gave the results with high accuracy. The JENDL-3.2 library showed significantly closer results to the ENDF/B-V. The k eff values calculated with the ENDF/B-VI library gave underestimated results. The ENDF/B-V library gave the best estimation. It can be concluded that MCNP-4C calculation, especially with ENDF/B-V and JENDL-3.2 libraries, for MOX fuel utilized NPP design in reactor core is the best choice

Burn-up Credit Criticality Safety Benchmark Phase III-C. Nuclide Composition and Neutron Multiplication Factor of a Boiling Water Reactor Spent Fuel Assembly for Burn-up Credit and Criticality Control of Damaged Nuclear Fuel

International Nuclear Information System (INIS)

Suyama, K.; Uchida, Y.; Kashima, T.; Ito, T.; Miyaji, T.

2016-01-01

Criticality control of damaged nuclear fuel is one of the key issues in the decommissioning operation of the Fukushima Daiichi Nuclear Power Station accident. The average isotopic composition of spent nuclear fuel as a function of burn-up is required in order to evaluate criticality parameters of the mixture of damaged nuclear fuel with other materials. The NEA Expert Group on Burn-up Credit Criticality (EGBUC) has organised several international benchmarks to assess the accuracy of burn-up calculation methodologies. For BWR fuel, the Phase III-B benchmark, published in 2002, was a remarkable landmark that provided general information on the burn-up properties of BWR spent fuel based on the 8x8 type fuel assembly. Since the publication of the Phase III-B benchmark, all major nuclear data libraries have been revised; in Japan from JENDL-3.2 to JENDL-4, in Europe from JEF-2.2 to JEFF-3.1 and in the US from ENDF/B-VI to ENDF/B-VII.1. Burn-up calculation methodologies have been improved by adopting continuous-energy Monte Carlo codes and modern neutronics calculation methods. Considering the importance of the criticality control of damaged fuel in the Fukushima Daiichi Nuclear Power Station accident, a new international burn-up calculation benchmark for the 9 x 9 STEP-3 BWR fuel assemblies was organised to carry out the inter-comparison of the averaged isotopic composition in the interest of the burnup credit criticality safety community. Benchmark specifications were proposed and approved at the EGBUC meeting in September 2012 and distributed in October 2012. The deadline for submitting results was set at the end of February 2013. The basic model for the benchmark problem is an infinite two-dimensional array of BWR fuel assemblies consisting of a 9 x 9 fuel rod array with a water channel in the centre. The initial uranium enrichment of fuel rods without gadolinium is 4.9, 4.4, 3.9, 3.4 and 2.1 wt% and 3.4 wt% for the rods using gadolinium. The burn-up conditions are

Benchmarking ENDF/B-VII.0

International Nuclear Information System (INIS)

Marck, Steven C. van der

2006-01-01

The new major release VII.0 of the ENDF/B nuclear data library has been tested extensively using benchmark calculations. These were based upon MCNP-4C3 continuous-energy Monte Carlo neutronics simulations, together with nuclear data processed using the code NJOY. Three types of benchmarks were used, viz., criticality safety benchmarks (fusion) shielding benchmarks, and reference systems for which the effective delayed neutron fraction is reported. For criticality safety, more than 700 benchmarks from the International Handbook of Criticality Safety Benchmark Experiments were used. Benchmarks from all categories were used, ranging from low-enriched uranium, compound fuel, thermal spectrum ones (LEU-COMP-THERM), to mixed uranium-plutonium, metallic fuel, fast spectrum ones (MIX-MET-FAST). For fusion shielding many benchmarks were based on IAEA specifications for the Oktavian experiments (for Al, Co, Cr, Cu, LiF, Mn, Mo, Si, Ti, W, Zr), Fusion Neutronics Source in Japan (for Be, C, N, O, Fe, Pb), and Pulsed Sphere experiments at Lawrence Livermore National Laboratory (for 6 Li, 7 Li, Be, C, N, O, Mg, Al, Ti, Fe, Pb, D 2 O, H 2 O, concrete, polyethylene and teflon). For testing delayed neutron data more than thirty measurements in widely varying systems were used. Among these were measurements in the Tank Critical Assembly (TCA in Japan) and IPEN/MB-01 (Brazil), both with a thermal spectrum, and two cores in Masurca (France) and three cores in the Fast Critical Assembly (FCA, Japan), all with fast spectra. In criticality safety, many benchmarks were chosen from the category with a thermal spectrum, low-enriched uranium, compound fuel (LEU-COMP-THERM), because this is typical of most current-day reactors, and because these benchmarks were previously underpredicted by as much as 0.5% by most nuclear data libraries (such as ENDF/B-VI.8, JEFF-3.0). The calculated results presented here show that this underprediction is no longer there for ENDF/B-VII.0. The average over 257

BENCHMARKING ORTEC ISOTOPIC MEASUREMENTS AND CALCULATIONS

Energy Technology Data Exchange (ETDEWEB)

Dewberry, R; Raymond Sigg, R; Vito Casella, V; Nitin Bhatt, N

2008-09-29

This report represents a description of compiled benchmark tests conducted to probe and to demonstrate the extensive utility of the Ortec ISOTOPIC {gamma}-ray analysis computer program. The ISOTOPIC program performs analyses of {gamma}-ray spectra applied to specific acquisition configurations in order to apply finite-geometry correction factors and sample-matrix-container photon absorption correction factors. The analysis program provides an extensive set of preset acquisition configurations to which the user can add relevant parameters in order to build the geometry and absorption correction factors that the program determines from calculus and from nuclear g-ray absorption and scatter data. The Analytical Development Section field nuclear measurement group of the Savannah River National Laboratory uses the Ortec ISOTOPIC analysis program extensively for analyses of solid waste and process holdup applied to passive {gamma}-ray acquisitions. Frequently the results of these {gamma}-ray acquisitions and analyses are to determine compliance with facility criticality safety guidelines. Another use of results is to designate 55-gallon drum solid waste as qualified TRU waste3 or as low-level waste. Other examples of the application of the ISOTOPIC analysis technique to passive {gamma}-ray acquisitions include analyses of standard waste box items and unique solid waste configurations. In many passive {gamma}-ray acquisition circumstances the container and sample have sufficient density that the calculated energy-dependent transmission correction factors have intrinsic uncertainties in the range 15%-100%. This is frequently the case when assaying 55-gallon drums of solid waste with masses of up to 400 kg and when assaying solid waste in extensive unique containers. Often an accurate assay of the transuranic content of these containers is not required, but rather a good defensible designation as >100 nCi/g (TRU waste) or <100 nCi/g (low level solid waste) is required. In

An improved benchmark model for the Big Ten critical assembly - 021

International Nuclear Information System (INIS)

Mosteller, R.D.

2010-01-01

A new benchmark specification is developed for the BIG TEN uranium critical assembly. The assembly has a fast spectrum, and its core contains approximately 10 wt.% enriched uranium. Detailed specifications for the benchmark are provided, and results from the MCNP5 Monte Carlo code using a variety of nuclear-data libraries are given for this benchmark and two others. (authors)

Evaluation and validation of criticality codes for fuel dissolver calculations

International Nuclear Information System (INIS)

Santamarina, A.; Smith, H.J.; Whitesides, G.E.

1991-01-01

During the past ten years an OECD/NEA Criticality Working Group has examined the validity of criticality safety computational methods. International calculation tools which were shown to be valid in systems for which experimental data existed were demonstrated to be inadequate when extrapolated to fuel dissolver media. A theoretical study of the main physical parameters involved in fuel dissolution calculations was performed, i.e. range of moderation, variation of pellet size and the fuel double heterogeneity effect. The APOLLO/P IC method developed to treat this latter effect permits us to supply the actual reactivity variation with pellet dissolution and to propose international reference values. The disagreement among contributors' calculations was analyzed through a neutron balance breakdown, based on three-group microscopic reaction rates. The results pointed out that fast and resonance nuclear data in criticality codes are not sufficiently reliable. Moreover the neutron balance analysis emphasized the inadequacy of the standard self-shielding formalism to account for 238 U resonance mutual self-shielding in the pellet-fissile liquor interaction. The benchmark exercise has resolved a potentially dangerous inadequacy in dissolver calculations. (author)

Effects of existing evaluated nuclear data files on neutronics characteristics of the BFS-62-3A critical assembly benchmark model

International Nuclear Information System (INIS)

Semenov, Mikhail

2002-11-01

This report is continuation of studying of the experiments performed on BFS-62-3A critical assembly in Russia. The objective of work is definition of the cross section uncertainties on reactor neutronics parameters as applied to the hybrid core of the BN-600 reactor of Beloyarskaya NPP. Two-dimensional benchmark model of BFS-62-3A was created specially for these purposes and experimental values were reduced to it. Benchmark characteristics for this assembly are 1) criticality; 2) central fission rate ratios (spectral indices); and 3) fission rate distributions in stainless steel reflector. The effects of nuclear data libraries have been studied by comparing the results calculated using available modern data libraries - ENDF/B-V, ENDF/B-VI, ENDF/B-VI-PT, JENDL-3.2 and ABBN-93. All results were computed by Monte Carlo method with the continuous energy cross-sections. The checking of the cross sections of major isotopes on wide benchmark criticality collection was made. It was shown that ENDF/B-V data underestimate the criticality of fast reactor systems up to 2% Δk. As for the rest data, the difference between each other in criticality for BFS-62-3A is around 0.6% Δk. However, taking into account the results obtained for other fast reactor benchmarks (and steel-reflected also), it may conclude that the difference in criticality calculation results can achieve 1% Δk. This value is in a good agreement with cross section uncertainty evaluated for BN-600 hybrid core (±0.6% Δk). This work is related to the JNC-IPPE Collaboration on Experimental Investigation of Excess Weapons Grade Pu Disposition in BN-600 Reactor Using BFS-2 Facility. (author)

Monte Carlo benchmark calculations for 400MWTH PBMR core

International Nuclear Information System (INIS)

Kim, H. C.; Kim, J. K.; Kim, S. Y.; Noh, J. M.

2007-01-01

A large interest in high-temperature gas-cooled reactors (HTGR) has been initiated in connection with hydrogen production in recent years. In this study, as a part of work for establishing Monte Carlo computation system for HTGR core analysis, some benchmark calculations for pebble-type HTGR were carried out using MCNP5 code. The core of the 400MW t h Pebble-bed Modular Reactor (PBMR) was selected as a benchmark model. Recently, the IAEA CRP5 neutronics and thermal-hydraulics benchmark problem was proposed for the testing of existing methods for HTGRs to analyze the neutronics and thermal-hydraulic behavior for the design and safety evaluations of the PBMR. This study deals with the neutronic benchmark problems, for fresh fuel and cold conditions (Case F-1), and first core loading with given number densities (Case F-2), proposed for PBMR. After the detailed MCNP modeling of the whole facility, benchmark calculations were performed. Spherical fuel region of a fuel pebble is divided into cubic lattice element in order to model a fuel pebble which contains, on average, 15000 CFPs (Coated Fuel Particles). Each element contains one CFP at its center. In this study, the side length of each cubic lattice element to have the same amount of fuel was calculated to be 0.1635 cm. The remaining volume of each lattice element was filled with graphite. All of different 5 concentric shells of CFP were modeled. The PBMR annular core consists of approximately 452000 pebbles in the benchmark problems. In Case F-1 where the core was filled with only fresh fuel pebble, a BCC(body-centered-cubic) lattice model was employed in order to achieve the random packing core with the packing fraction of 0.61. The BCC lattice was also employed with the size of the moderator pebble increased in a manner that reproduces the specified F/M ratio of 1:2 while preserving the packing fraction of 0.61 in Case F-2. The calculations were pursued with ENDF/B-VI cross-section library and used sab2002 S(α, �

Reactor group constants and benchmark test

Energy Technology Data Exchange (ETDEWEB)

Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-08-01

The evaluated nuclear data files such as JENDL, ENDF/B-VI and JEF-2 are validated by analyzing critical mock-up experiments for various type reactors and assessing applicability for nuclear characteristics such as criticality, reaction rates, reactivities, etc. This is called Benchmark Testing. In the nuclear calculations, the diffusion and transport codes use the group constant library which is generated by processing the nuclear data files. In this paper, the calculation methods of the reactor group constants and benchmark test are described. Finally, a new group constants scheme is proposed. (author)

Benchmark testing calculations for 232Th

International Nuclear Information System (INIS)

Liu Ping

2003-01-01

The cross sections of 232 Th 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 K eff 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)

CSRL-V ENDF/B-V 227-group neutron cross-section library and its application to thermal-reactor and criticality safety benchmarks

International Nuclear Information System (INIS)

Ford, W.E. III; Diggs, B.R.; Knight, J.R.; Greene, N.M.; Petrie, L.M.; Webster, C.C.; Westfall, R.M.; Wright, R.Q.; Williams, M.L.

1982-01-01

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

Benchmark criticality experiments for fast fission configuration with high enriched nuclear fuel

International Nuclear Information System (INIS)

Sikorin, S.N.; Mandzik, S.G.; Polazau, S.A.; Hryharovich, T.K.; Damarad, Y.V.; Palahina, Y.A.

2014-01-01

Benchmark criticality experiments of fast heterogeneous configuration with high enriched uranium (HEU) nuclear fuel were performed using the 'Giacint' critical assembly of the Joint Institute for Power and Nuclear Research - Sosny (JIPNR-Sosny) of the National Academy of Sciences of Belarus. The critical assembly core comprised fuel assemblies without a casing for the 34.8 mm wrench. Fuel assemblies contain 19 fuel rods of two types. The first type is metal uranium fuel rods with 90% enrichment by U-235; the second one is dioxide uranium fuel rods with 36% enrichment by U-235. The total fuel rods length is 620 mm, and the active fuel length is 500 mm. The outer fuel rods diameter is 7 mm, the wall is 0.2 mm thick, and the fuel material diameter is 6.4 mm. The clad material is stainless steel. The side radial reflector: the inner layer of beryllium, and the outer layer of stainless steel. The top and bottom axial reflectors are of stainless steel. The analysis of the experimental results obtained from these benchmark experiments by developing detailed calculation models and performing simulations for the different experiments is presented. The sensitivity of the obtained results for the material specifications and the modeling details were examined. The analyses used the MCNP and MCU computer programs. This paper presents the experimental and analytical results. (authors)

TRX and UO2 criticality benchmarks with SAM-CE

International Nuclear Information System (INIS)

Beer, M.; Troubetzkoy, E.S.; Lichtenstein, H.; Rose, P.F.

1980-01-01

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

3-D extension C5G7 MOX benchmark calculation using threedant code

International Nuclear Information System (INIS)

Kim, H.Ch.; Han, Ch.Y.; Kim, J.K.; Na, B.Ch.

2005-01-01

It pursued the benchmark on deterministic 3-D MOX fuel assembly transport calculations without spatial homogenization (C5G7 MOX Benchmark Extension). The goal of this benchmark is to provide a more through test results for the abilities of current available 3-D methods to handle the spatial heterogeneities of reactor core. The benchmark requires solutions in the form of normalized pin powers as well as the eigenvalue for each of the control rod configurations; without rod, with A rods, and with B rods. In this work, the DANTSYS code package was applied to analyze the 3-D Extension C5G7 MOX Benchmark problems. The THREEDANT code within the DANTSYS code package, which solves the 3-D transport equation in x-y-z, and r-z-theta geometries, was employed to perform the benchmark calculations. To analyze the benchmark with the THREEDANT code, proper spatial and angular approximations were made. Several calculations were performed to investigate the effects of the different spatial approximations on the accuracy. The results from these sensitivity studies were analyzed and discussed. From the results, it is found that the 4*4 grid per pin cell is sufficiently refined so that very little benefit is obtained by increasing the mesh size. (authors)

MOx Depletion Calculation Benchmark

International Nuclear Information System (INIS)

San Felice, Laurence; Eschbach, Romain; Dewi Syarifah, Ratna; Maryam, Seif-Eddine; Hesketh, Kevin

2016-01-01

Under the auspices of the NEA Nuclear Science Committee (NSC), the Working Party on Scientific Issues of Reactor Systems (WPRS) has been established to study the reactor physics, fuel performance, radiation transport and shielding, and the uncertainties associated with modelling of these phenomena in present and future nuclear power systems. The WPRS has different expert groups to cover a wide range of scientific issues in these fields. The Expert Group on Reactor Physics and Advanced Nuclear Systems (EGRPANS) was created in 2011 to perform specific tasks associated with reactor physics aspects of present and future nuclear power systems. EGRPANS provides expert advice to the WPRS and the nuclear community on the development needs (data and methods, validation experiments, scenario studies) for different reactor systems and also provides specific technical information regarding: core reactivity characteristics, including fuel depletion effects; core power/flux distributions; Core dynamics and reactivity control. In 2013 EGRPANS published a report that investigated fuel depletion effects in a Pressurised Water Reactor (PWR). This was entitled 'International Comparison of a Depletion Calculation Benchmark on Fuel Cycle Issues' NEA/NSC/DOC(2013) that documented a benchmark exercise for UO 2 fuel rods. This report documents a complementary benchmark exercise that focused on PuO 2 /UO 2 Mixed Oxide (MOX) fuel rods. The results are especially relevant to the back-end of the fuel cycle, including irradiated fuel transport, reprocessing, interim storage and waste repository. Saint-Laurent B1 (SLB1) was the first French reactor to use MOx assemblies. SLB1 is a 900 MWe PWR, with 30% MOx fuel loading. The standard MOx assemblies, used in Saint-Laurent B1 reactor, include three zones with different plutonium enrichments, high Pu content (5.64%) in the center zone, medium Pu content (4.42%) in the intermediate zone and low Pu content (2.91%) in the peripheral zone

EA-MC Neutronic Calculations on IAEA ADS Benchmark 3.2

Energy Technology Data Exchange (ETDEWEB)

Dahlfors, Marcus [Uppsala Univ. (Sweden). Dept. of Radiation Sciences; Kadi, Yacine [CERN, Geneva (Switzerland). Emerging Energy Technologies

2006-01-15

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 {sup 129}I, {sup 237}Np and {sup 243}Am samples and of fission reaction rates in {sup 235}U, {sup 237}Np and {sup 243}Am 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.

Benchmark calculation programme concerning typical LMFBR structures

International Nuclear Information System (INIS)

Donea, J.; Ferrari, G.; Grossetie, J.C.; Terzaghi, A.

1982-01-01

This programme, which is part of a comprehensive activity aimed at resolving difficulties encountered in using design procedures based on ASME Code Case N-47, should allow to get confidence in computer codes which are supposed to provide a realistic prediction of the LMFBR component behaviour. The calculations started on static analysis of typical structures made of non linear materials stressed by cyclic loads. The fluid structure interaction analysis is also being considered. Reasons and details of the different benchmark calculations are described, results obtained are commented and future computational exercise indicated

FENDL-2 and associated benchmark calculations

International Nuclear Information System (INIS)

Pashchenko, A.B.; Muir, D.W.

1992-03-01

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

Immunotoxicity of perfluorinated alkylates: calculation of benchmark doses based on serum concentrations in children

DEFF Research Database (Denmark)

Grandjean, Philippe; Budtz-Joergensen, Esben

2013-01-01

BACKGROUND: Immune suppression may be a critical effect associated with exposure to perfluorinated compounds (PFCs), as indicated by recent data on vaccine antibody responses in children. Therefore, this information may be crucial when deciding on exposure limits. METHODS: Results obtained from...... 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...

The International Criticality Safety Benchmark Evaluation Project on the Internet

International Nuclear Information System (INIS)

Briggs, J.B.; Brennan, S.A.; Scott, L.

2000-01-01

The International Criticality Safety Benchmark Evaluation Project (ICSBEP) was initiated in October 1992 by the US Department of Energy's (DOE's) defense programs and is documented in the Transactions of numerous American Nuclear Society and International Criticality Safety Conferences. The work of the ICSBEP is documented as an Organization for Economic Cooperation and Development (OECD) handbook, International Handbook of Evaluated Criticality Safety Benchmark Experiments. The ICSBEP Internet site was established in 1996 and its address is http://icsbep.inel.gov/icsbep. A copy of the ICSBEP home page is shown in Fig. 1. The ICSBEP Internet site contains the five primary links. Internal sublinks to other relevant sites are also provided within the ICSBEP Internet site. A brief description of each of the five primary ICSBEP Internet site links is given

Method of characteristics - Based sensitivity calculations for international PWR benchmark

International Nuclear Information System (INIS)

Suslov, I. R.; Tormyshev, I. V.; Komlev, O. G.

2013-01-01

Method to calculate sensitivity of fractional-linear neutron flux functionals to transport equation coefficients is proposed. Implementation of the method on the basis of MOC code MCCG3D is developed. Sensitivity calculations for fission intensity for international PWR benchmark are performed. (authors)

Benchmark Calculations of Noncovalent Interactions of Halogenated Molecules

Czech Academy of Sciences Publication Activity Database

Řezáč, Jan; Riley, Kevin Eugene; Hobza, Pavel

2012-01-01

Roč. 8, č. 11 (2012), s. 4285-4292 ISSN 1549-9618 R&D Projects: GA ČR GBP208/12/G016 Institutional support: RVO:61388963 Keywords : halogenated molecules * noncovalent interactions * benchmark calculations Subject RIV: CF - Physical ; Theoretical Chemistry Impact factor: 5.389, year: 2012

Processing and benchmarking of evaluated nuclear data file/b-viii.0β4 cross-section library by analysis of a series of critical experimental benchmark using the monte carlo code MCNP(X and NJOY2016

Directory of Open Access Journals (Sweden)

Kabach Ouadie

2017-12-01

Full Text Available To validate the new Evaluated Nuclear Data File (ENDF/B-VIII.0β4 library, 31 different critical cores were selected and used for a benchmark test of the important parameter keff. The four utilized libraries are processed using Nuclear Data Processing Code (NJOY2016. The results obtained with the ENDF/B-VIII.0β4 library were compared against those calculated with ENDF/B-VI.8, ENDF/B-VII.0, and ENDF/B-VII.1 libraries using the Monte Carlo N-Particle (MCNP(X code. All the MCNP(X calculations of keff values with these four libraries were compared with the experimentally measured results, which are available in the International Critically Safety Benchmark Evaluation Project. The obtained results are discussed and analyzed in this paper.

COVE 2A Benchmarking calculations using NORIA

International Nuclear Information System (INIS)

Carrigan, C.R.; Bixler, N.E.; Hopkins, P.L.; Eaton, R.R.

1991-10-01

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

Criticality safety benchmark evaluation project: Recovering the past

Energy Technology Data Exchange (ETDEWEB)

Trumble, E.F.

1997-06-01

A very brief summary of the Criticality Safety Benchmark Evaluation Project of the Westinghouse Savannah River Company is provided in this paper. The purpose of the project is to provide a source of evaluated criticality safety experiments in an easily usable format. Another project goal is to search for any experiments that may have been lost or contain discrepancies, and to determine if they can be used. Results of evaluated experiments are being published as US DOE handbooks.

The fifth Atomic Energy Research dynamic benchmark calculation with HEXTRAN-SMABRE

International Nuclear Information System (INIS)

Haenaelaeinen, Anitta

1998-01-01

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)

Benchmark Calculations on Halden IFA-650 LOCA Test Results

International Nuclear Information System (INIS)

Ek, Mirkka; Kekkonen, Laura; Kelppe, Seppo; Stengaard, J.O.; Josek, Radomir; Wiesenack, Wolfgang; Aounallah, Yacine; Wallin, Hannu; Grandjean, Claude; Herb, Joachim; Lerchl, Georg; Trambauer, Klaus; Sonnenburg, Heinz-Guenther; Nakajima, Tetsuo; Spykman, Gerold; Struzik, Christine

2010-01-01

The assessment of the consequences of a loss-of-coolant accident (LOCA) is to a large extent based on calculations carried out with codes especially developed for addressing the phenomena occurring during the transient. Since the time of the first LOCA experiments, which were largely conducted with fresh fuel, changes in fuel design, the introduction of new cladding materials and in particular the move to high burnup have not only generated a need to re-examine the LOCA safety criteria and to verify their continued validity, but also to confirm that codes show an appropriate performance especially with respect to high burnup phenomena influencing LOCA fuel behaviour. As part of international efforts, the OECD Halden Reactor Project program implemented a test series to address particular LOCA issues. Based on recommendations of a group of experts from the US NRC, EPRI, EDF, FRAMATOME-ANP and GNF, the primary objective of the experiments were defined as 1. Measure the extent of fuel (fragment) relocation into the ballooned region and evaluate its possible effect on cladding temperature and oxidation. 2. Investigate the extent (if any) of 'secondary transient hydriding' on the inner side of the cladding above and below the burst region. The Halden LOCA series, using high burnup fuel segments, contains test cases well suited for checking the ability of LOCA analysis codes to predict or reproduce the measurements and to provide clues as to where the codes need to be improved. The NEA Working Group on Fuel Safety, WGFS, therefore decided to conduct a code benchmark based on the Halden LOCA test series. Emphasis was on the codes' ability to predict or reproduce the thermal and mechanical response of fuel and cladding. Before starting the benchmark, participants were given the opportunity to tune their codes to the experimental system applied in the Halden LOCA tests. To this end, the data from the two commissioning runs were made available. The first of these runs went

Stationary PWR-calculations by means of LWRSIM at the NEACRP 3D-LWRCT benchmark

International Nuclear Information System (INIS)

Van de Wetering, T.F.H.

1993-01-01

Within the framework of participation in an international benchmark, calculations were executed by means of an adjusted version of the computer code Light Water Reactor SIMulation (LWRSIM) for three-dimensional reactor core calculations of pressurized water reactors. The 3-D LWR Core Transient Benchmark was set up aimed at the comparison of 3-D computer codes for transient calculations in LWRs. Participation in the benchmark provided more insight in the accuracy of the code when applied for other pressurized water reactors than applied for the nuclear power plant Borssele in the Netherlands, for which the code has been developed and used originally

LCEs for Naval Reactor Benchmark Calculations

International Nuclear Information System (INIS)

W.J. Anderson

1999-01-01

The purpose of this engineering calculation is to document the MCNP4B2LV evaluations of Laboratory Critical Experiments (LCEs) performed as part of the Disposal Criticality Analysis Methodology program. LCE evaluations documented in this report were performed for 22 different cases with varied design parameters. Some of these LCEs (10) are documented in existing references (Ref. 7.1 and 7.2), but were re-run for this calculation file using more neutron histories. The objective of this analysis is to quantify the MCNP4B2LV code system's ability to accurately calculate the effective neutron multiplication factor (k eff ) for various critical configurations. These LCE evaluations support the development and validation of the neutronics methodology used for criticality analyses involving Naval reactor spent nuclear fuel in a geologic repository

Calculations of IAEA-CRP-6 Benchmark Case 1 through 7 for a TRISO-Coated Fuel Particle

International Nuclear Information System (INIS)

Kim, Young Min; Lee, Y. W.; Chang, J. H.

2005-01-01

IAEA-CRP-6 is a coordinated research program of IAEA on Advances in HTGR fuel technology. The CRP examines aspects of HTGR fuel technology, ranging from design and fabrication to characterization, irradiation testing, performance modeling, as well as licensing and quality control issues. The benchmark section of the program treats simple analytical cases, pyrocarbon layer behavior, single TRISO-coated fuel particle behavior, and benchmark calculations of some irradiation experiments performed and planned. There are totally seventeen benchmark cases in the program. Member countries are participating in the benchmark calculations of the CRP with their own developed fuel performance analysis computer codes. Korea is also taking part in the benchmark calculations using a fuel performance analysis code, COPA (COated PArticle), which is being developed in Korea Atomic Energy Research Institute. The study shows the calculational results of IAEACRP- 6 benchmark cases 1 through 7 which describe the structural behaviors for a single fuel particle

Criticality Benchmark Analysis of Water-Reflected Uranium Oxyfluoride Slabs

International Nuclear Information System (INIS)

Marshall, Margaret A.; Bess, John D.

2009-01-01

A series of twelve experiments were conducted in the mid 1950's at the Oak Ridge National Laboratory Critical Experiments Facility to determine the critical conditions of a semi-infinite water-reflected slab of aqueous uranium oxyfluoride (UO2F2). A different slab thickness was used for each experiment. Results from the twelve experiment recorded in the laboratory notebook were published in Reference 1. Seven of the twelve experiments were determined to be acceptable benchmark experiments for the inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. This evaluation will not only be available to handbook users for the validation of computer codes and integral cross-section data, but also for the reevaluation of experimental data used in the ANSI/ANS-8.1 standard. This evaluation is important as part of the technical basis of the subcritical slab limits in ANSI/ANS-8.1. The original publication of the experimental results was used for the determination of bias and bias uncertainties for subcritical slab limits, as documented by Hugh Clark's paper 'Subcritical Limits for Uranium-235 Systems'.

Benchmark test of JEF-1 evaluation by calculating fast criticalities

International Nuclear Information System (INIS)

Pelloni, S.

1986-06-01

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)

Analysis of the impact of correlated benchmark experiments on the validation of codes for criticality safety analysis

International Nuclear Information System (INIS)

Bock, M.; Stuke, M.; Behler, M.

2013-01-01

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 (k eff ). The analysis presented here is based on this proposal. (orig.)

Benchmarking ENDF/B-VII.1, JENDL-4.0 and JEFF-3.1

International Nuclear Information System (INIS)

Van Der Marck, S. C.

2012-01-01

Three nuclear data libraries have been tested extensively using criticality safety benchmark calculations. The three libraries are the new release of the US library ENDF/B-VII.1 (2011), the new release of the Japanese library JENDL-4.0 (2011), and the OECD/NEA library JEFF-3.1 (2006). All calculations were performed with the continuous-energy Monte Carlo code MCNP (version 4C3, as well as version 6-beta1). Around 2000 benchmark cases from the International Handbook of Criticality Safety Benchmark Experiments (ICSBEP) were used. The results were analyzed per ICSBEP category, and per element. Overall, the three libraries show similar performance on most criticality safety benchmarks. The largest differences are probably caused by elements such as Be, C, Fe, Zr, W. (authors)

Providing Nuclear Criticality Safety Analysis Education through Benchmark Experiment Evaluation

International Nuclear Information System (INIS)

Bess, John D.; Briggs, J. Blair; Nigg, David W.

2009-01-01

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.

Influence of the ab initio n–d cross sections in the critical heavy-water benchmarks

International Nuclear Information System (INIS)

Morillon, B.; Lazauskas, R.; Carbonell, J.

2013-01-01

Highlights: ► We solve the three nucleon problem using different NN potential (MT, AV18 and INOY) to calculate the Neutron–deuteron cross sections. ► These cross sections are compared to the existing experimental data and to international libraries. ► We describe the different sets of heavy water benchmarks for which the Monte Carlo simulations have been performed including our new Neutron–deuteron cross sections. ► The results obtained by the ab initio INOY potential have been compared with the calculations based on the international library cross sections and are found to be of the same quality. - Abstract: The n–d elastic and breakup cross sections are computed by solving the three-body Faddeev equations for realistic and semi-realistic nucleon–nucleon potentials. These cross sections are inserted in the Monte Carlo simulation of the nuclear processes considered in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook). The results obtained using thes ab initio n–d cross sections are compared with those provided by the most renown international libraries

Calculus of a reactor VVER-1000 benchmark; Calcul d'un benchmark de reacteur VVER-1000

Energy Technology Data Exchange (ETDEWEB)

Dourougie, C

1998-07-01

In the framework of the FMDP (Fissile Materials Disposition Program between the US and Russian, a benchmark was tested. The pin cells contain low enriched uranium (LEU) and mixed oxide fuels (MOX). The calculations are done for a wide range of temperatures and solute boron concentrations, in accidental conditions. (A.L.B.)

Evaluation and validation of criticality codes for fuel dissolver calculations

International Nuclear Information System (INIS)

Santamarina, A.; Smith, H.J.; Whitesides, G.E.

1991-01-01

During the past ten years an OECD/NEA Criticality Working Group has examined the validity of criticality safety computational methods. International calculation tools which were shown to be valid in systems for which experimental data existed were demonstrated to be inadequate when extrapolated to fuel dissolver media. The spread of the results in the international calculation amounted to ± 12,000 pcm in the realistic fuel dissolver exercise n degrees 19 proposed by BNFL, and to ± 25,000 pcm in the benchmark n degrees 20 in which fissile material in solid form is surrounded by fissile material in solution. A theoretical study of the main physical parameters involved in fuel dissolution calculations was performed, i.e. range of moderation, variation of pellet size and the fuel double heterogeneity effect. The APOLLO/P IC method developed to treat latter effect, permits us to supply the actual reactivity variation with pellet dissolution and to propose international reference values. The disagreement among contributors' calculations was analyzed through a neutron balance breakdown, based on three-group microscopic reaction rates solicited from the participants. The results pointed out that fast and resonance nuclear data in criticality codes are not sufficiently reliable. Moreover the neutron balance analysis emphasized the inadequacy of the standard self-shielding formalism (NITAWL in the international SCALE package) to account for 238 U resonance mutual self-shielding in the pellet-fissile liquor interaction. Improvements in the up-dated 1990 contributions, as do recent complementary reference calculations (MCNP, VIM, ultrafine slowing-down CGM calculation), confirm the need to use rigorous self-shielding methods in criticality design-oriented codes. 6 refs., 11 figs., 3 tabs

Comparisons of the MCNP criticality benchmark suite with ENDF/B-VI.8, JENDL-3.3, and JEFF-3.0

International Nuclear Information System (INIS)

Kim, Do Heon; Gil, Choong-Sup; Kim, Jung-Do; Chang, Jonghwa

2003-01-01

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 χ 2 differences between the MCNP results and experimental data were calculated for the libraries. (author)

HTR-PROTEUS benchmark calculations. Pt. 1. Unit cell results LEUPRO-1 and LEUPRO-2

International Nuclear Information System (INIS)

Hogenbirk, A.; Stad, R.C.L. van der; Janssen, A.J.; Klippel, H.T.; Kuijper, J.C.

1995-09-01

In the framework of the IAEA Co-ordinated Research Programme (CRP) on 'Validation of Safety Related Physics Calculations for Low-Enriched (LEU) HTGRs' calculational benchmarks are performed on the basis of LEU-HTR pebble-bed critical experiments carried out in the PROTEUS facility at PSI, Switzerland. Of special interest is the treatment of the double heterogeneity of the fuel and the spherical fuel elements of these pebble bed core configurations. Also of interest is the proper calculation of the safety related physics parameters like the effect of water ingress and control rod worth. This document describes the ECN results of the LEUPRO-1 and LEUPRO-2 unitcell calculations performed with the codes WIMS-E, SCALE-4 and MCNP4A. Results of the LEUPRO-1 unit cell with 20% water ingress in the void is also reported for both the single and the double heterogeneous case. Emphasis is put on the intercomparison of the results obtained by the deterministic codes WIMS-E and SCALE-4, and the Monte Carlo code MCNP4A. The LEUPRO whole core calculations will be reported later. (orig.)

VENUS-2 MOX Core Benchmark: Results of ORNL Calculations Using HELIOS-1.4

Energy Technology Data Exchange (ETDEWEB)

Ellis, RJ

2001-02-02

The Task Force on Reactor-Based Plutonium Disposition, now an Expert Group, was set up through the Organization for Economic Cooperation and Development/Nuclear Energy Agency to facilitate technical assessments of burning weapons-grade plutonium mixed-oxide (MOX) fuel in U.S. pressurized-water reactors and Russian VVER nuclear reactors. More than ten countries participated to advance the work of the Task Force in a major initiative, which was a blind benchmark study to compare code benchmark calculations against experimental data for the VENUS-2 MOX core at SCK-CEN in Mol, Belgium. At the Oak Ridge National Laboratory, the HELIOS-1.4 code was used to perform a comprehensive study of pin-cell and core calculations for the VENUS-2 benchmark.

Two-dimensional benchmark calculations for PNL-30 through PNL-35

International Nuclear Information System (INIS)

Mosteller, R.D.

1997-01-01

Interest in critical experiments with lattices of mixed-oxide (MOX) fuel pins has been revived by the possibility that light water reactors will be used for disposition of weapons-grade plutonium. A series of six experiments with MOX lattices, designated PNL-30 through PNL-35, was performed at Pacific Northwest Laboratories in 1975 and 1976, and a set of benchmark specifications for these experiments subsequently was adopted by the Cross Section Evaluation Working Group (CSEWG). Although there appear to be some problems with these experiments, they remain the only CSEWG benchmarks for MOX lattices. The number of fuel pins in these experiments is relatively low, corresponding to fewer than 4 typical pressurized-water-reactor fuel assemblies. Accordingly, they are more appropriate as benchmarks for lattice-physics codes than for reactor-core simulator codes. Unfortunately, the CSEWG specifications retain the full three-dimensional (3D) detail of the experiments, while lattice-physics codes almost universally are limited to two dimensions (2D). This paper proposes an extension of the benchmark specifications to include a 2D model, and it justifies that extension by comparing results from the MCNP Monte Carlo code for the 2D and 3D specifications

Validation of the criticality calculation for fuel elements using the Gamtec 2 - Keno 2 and 4

International Nuclear Information System (INIS)

Teixeira, M.C.C.; Andrade, M.C. de

1990-01-01

For criticality safety in the fabrication, storage and transportation of fuel assemblies, subcriticality analysis must be done. The calculations are performed at CDTN with the GAMTEC computer code, to homogenize the fuel assembly in order to create 16 group cross-section library, and with KENO code, for determining the multiplication factor. To validate the calculational method, suitable Benchmark experiments have been done. The results show that the calculational model overestimates kef when kef+ 2 σ was considered. (author) [pt

JNC results of BFS-62-3A benchmark calculation (CRP: Phase 5)

International Nuclear Information System (INIS)

Ishikawa, M.

2004-01-01

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

Validation of the EIR LWR calculation methods for criticality assessment of storage pools

International Nuclear Information System (INIS)

Grimm, P.; Paratte, J.M.

1986-11-01

The EIR code system for the calculation of light water reactors is presented and the methods used are briefly described. The application of the system to various types of critical experiments and benchmark problems proves its good accuracy, even for heterogeneous configurations containing strong neutron absorbers such as Boral. Since the multiplication factor k eff is normally somewhat overpredicted and the spread of the results is small, this code system is validated for the calculation of storage pools, taking into account a safety margins of 1.5% on k eff . (author)

Heavy nucleus resonant absorption calculation benchmarks

International Nuclear Information System (INIS)

Tellier, H.; Coste, H.; Raepsaet, C.; Van der Gucht, C.

1993-01-01

The calculation of the space and energy dependence of the heavy nucleus resonant absorption in a heterogeneous lattice is one of the hardest tasks in reactor physics. Because of the computer time and memory needed, it is impossible to represent finely the cross-section behavior in the resonance energy range for everyday computations. Consequently, reactor physicists use a simplified formalism, the self-shielding formalism. As no clean and detailed experimental results are available to validate the self-shielding calculations, Monte Carlo computations are used as a reference. These results, which were obtained with the TRIPOLI continuous-energy Monte Carlo code, constitute a set of numerical benchmarks than can be used to evaluate the accuracy of the techniques or formalisms that are included in any reactor physics codes. Examples of such evaluations, for the new assembly code APOLLO2 and the slowing-down code SECOL, are given for cases of 238 U and 232 Th fuel elements

Analysis of kyoto university reactor physics critical experiments using NCNSRC calculation methodology

International Nuclear Information System (INIS)

Amin, E.; Hathout, A.M.; Shouman, S.

1997-01-01

The kyoto university reactor physics experiments on the university critical assembly is used to benchmark validate the NCNSRC calculations methodology. This methodology has two lines, diffusion and Monte Carlo. The diffusion line includes the codes WIMSD4 for cell calculations and the two dimensional diffusion code DIXY2 for core calculations. The transport line uses the MULTIKENO-Code vax Version. Analysis is performed for the criticality, and the temperature coefficients of reactivity (TCR) for the light water moderated and reflected cores, of the different cores utilized in the experiments. The results of both Eigen value and TCR approximately reproduced the experimental and theoretical Kyoto results. However, some conclusions are drawn about the adequacy of the standard wimsd4 library. This paper is an extension of the NCNSRC efforts to assess and validate computer tools and methods for both Et-R R-1 and Et-MMpr-2 research reactors. 7 figs., 1 tab

CFD-calculations to a core catcher benchmark

International Nuclear Information System (INIS)

Willschuetz, H.G.

1999-04-01

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. The difficulty consists in the experimental simulation of the decay heat that can be neglected for the short-run course of events like relocation and spreading, which must, however, be considered during investigation of the long time behaviour. Therefore the German GRS, defined together with Battelle Ingenieurtechnik a benchmark problem in order to determine particular problems and differences of CFD codes simulating an expanded corium and from this, requirements for a reasonable measurement of experiments, that will be performed later. First the finite-volume-codes Comet 1.023, CFX 4.2 and CFX-TASCflow were used. To be able to make comparisons to a finite-element-code, now calculations are performed at the Institute of Safety Research at the Forschungszentrum Rossendorf with the code ANSYS/FLOTRAN. For the benchmark calculations of stage 1 a pure and liquid melt with internal heat sources was assumed uniformly distributed over the area of the planned core catcher of a EPR plant. Using the Standard-k-ε-turbulence model and assuming an initial state of a motionless superheated melt several large convection rolls will establish within the melt pool. The temperatures at the surface do not sink to a solidification level due to the enhanced convection heat transfer. The temperature gradients at the surface are relatively flat while there are steep gradients at the ground where the no slip condition is applied. But even at the ground no solidification temperatures are observed. Although the problem in the ANSYS-calculations is handled two-dimensional and not three-dimensional like in the finite-volume-codes, there are no fundamental deviations to the results of the other codes. (orig.)

Some comments on cold hydrogenous moderators, simple synthetic kernels and benchmark calculations

International Nuclear Information System (INIS)

Dorning, J.

1997-09-01

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.

JNC results of BN-600 benchmark calculation (phase 3)

International Nuclear Information System (INIS)

Ishikawa, M.

2002-01-01

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

Benchmark calculations of thermal reaction rates. I - Quantal scattering theory

Science.gov (United States)

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.

Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel: Joint U.S./Russian Progress Report for Fiscal Year 1997 Volume 2-Calculations Performed in the United States

Energy Technology Data Exchange (ETDEWEB)

Primm III, RT

2002-05-29

This volume of the progress report provides documentation of reactor physics and criticality safety studies conducted in the US during fiscal year 1997 and sponsored by the Fissile Materials Disposition Program of the US Department of Energy. Descriptions of computational and experimental benchmarks for the verification and validation of computer programs for neutron physics analyses are included. All benchmarks include either plutonium, uranium, or mixed uranium and plutonium fuels. Calculated physics parameters are reported for all of the computational benchmarks and for those experimental benchmarks that the US and Russia mutually agreed in November 1996 were applicable to mixed-oxide fuel cycles for light-water reactors.

Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel: Joint U.S./Russian Progress Report for Fiscal Year 1997 Volume 2-Calculations Performed in the United States

International Nuclear Information System (INIS)

Primm III, RT

2002-01-01

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

Impact of the 235U Covariance Data in Benchmark Calculations

International Nuclear Information System (INIS)

Leal, Luiz C.; Mueller, D.; Arbanas, G.; Wiarda, D.; Derrien, H.

2008-01-01

The error estimation for calculated quantities relies on nuclear data uncertainty information available in the basic nuclear data libraries such as the U.S. Evaluated Nuclear Data File (ENDF/B). The uncertainty files (covariance matrices) in the ENDF/B library are generally obtained from analysis of experimental data. In the resonance region, the computer code SAMMY is used for analyses of experimental data and generation of resonance parameters. In addition to resonance parameters evaluation, SAMMY also generates resonance parameter covariance matrices (RPCM). SAMMY uses the generalized least-squares formalism (Bayes method) together with the resonance formalism (R-matrix theory) for analysis of experimental data. Two approaches are available for creation of resonance-parameter covariance data. (1) During the data-evaluation process, SAMMY generates both a set of resonance parameters that fit the experimental data and the associated resonance-parameter covariance matrix. (2) For existing resonance-parameter evaluations for which no resonance-parameter covariance data are available, SAMMY can retroactively create a resonance-parameter covariance matrix. The retroactive method was used to generate covariance data for 235U. The resulting 235U covariance matrix was then used as input to the PUFF-IV code, which processed the covariance data into multigroup form, and to the TSUNAMI code, which calculated the uncertainty in the multiplication factor due to uncertainty in the experimental cross sections. The objective of this work is to demonstrate the use of the 235U covariance data in calculations of critical benchmark systems

Impact of the 235U covariance data in benchmark calculations

International Nuclear Information System (INIS)

Leal, Luiz; Mueller, Don; Arbanas, Goran; Wiarda, Dorothea; Derrien, Herve

2008-01-01

The error estimation for calculated quantities relies on nuclear data uncertainty information available in the basic nuclear data libraries such as the U.S. Evaluated Nuclear Data File (ENDF/B). The uncertainty files (covariance matrices) in the ENDF/B library are generally obtained from analysis of experimental data. In the resonance region, the computer code SAMMY is used for analyses of experimental data and generation of resonance parameters. In addition to resonance parameters evaluation, SAMMY also generates resonance parameter covariance matrices (RPCM). SAMMY uses the generalized least-squares formalism (Bayes' method) together with the resonance formalism (R-matrix theory) for analysis of experimental data. Two approaches are available for creation of resonance-parameter covariance data. (1) During the data-evaluation process, SAMMY generates both a set of resonance parameters that fit the experimental data and the associated resonance-parameter covariance matrix. (2) For existing resonance-parameter evaluations for which no resonance-parameter covariance data are available, SAMMY can retroactively create a resonance-parameter covariance matrix. The retroactive method was used to generate covariance data for 235 U. The resulting 235 U covariance matrix was then used as input to the PUFF-IV code, which processed the covariance data into multigroup form, and to the TSUNAMI code, which calculated the uncertainty in the multiplication factor due to uncertainty in the experimental cross sections. The objective of this work is to demonstrate the use of the 235 U covariance data in calculations of critical benchmark systems. (authors)

Benchmarking density-functional-theory calculations of rotational g tensors and magnetizabilities using accurate coupled-cluster calculations.

Science.gov (United States)

Lutnaes, Ola B; Teale, Andrew M; Helgaker, Trygve; Tozer, David J; Ruud, Kenneth; Gauss, Jürgen

2009-10-14

An accurate set of benchmark rotational g tensors and magnetizabilities are calculated using coupled-cluster singles-doubles (CCSD) theory and coupled-cluster single-doubles-perturbative-triples [CCSD(T)] theory, in a variety of basis sets consisting of (rotational) London atomic orbitals. The accuracy of the results obtained is established for the rotational g tensors by careful comparison with experimental data, taking into account zero-point vibrational corrections. After an analysis of the basis sets employed, extrapolation techniques are used to provide estimates of the basis-set-limit quantities, thereby establishing an accurate benchmark data set. The utility of the data set is demonstrated by examining a wide variety of density functionals for the calculation of these properties. None of the density-functional methods are competitive with the CCSD or CCSD(T) methods. The need for a careful consideration of vibrational effects is clearly illustrated. Finally, the pure coupled-cluster results are compared with the results of density-functional calculations constrained to give the same electronic density. The importance of current dependence in exchange-correlation functionals is discussed in light of this comparison.

Benchmarking the new JENDL-4.0 library on criticality experiments of a research reactor with oxide LEU (20 w/o) fuel, light water moderator and beryllium reflectors

International Nuclear Information System (INIS)

Liem, Peng Hong; Sembiring, Tagor Malem

2012-01-01

Highlights: ► Benchmark calculations of the new JENDL-4.0 library. ► Thermal research reactor with oxide LEU fuel, H 2 O moderator and Be reflector. ► JENDL-4.0 library shows better C/E values for criticality evaluations. - Abstract: Benchmark calculations of the new JENDL-4.0 library on the criticality experiments of a thermal research reactor with oxide low enriched uranium (LEU, 20 w/o) fuel, light water moderator and beryllium reflector (RSG GAS) have been conducted using a continuous energy Monte Carlo code, MVP-II. The JENDL-4.0 library shows better C/E values compared to the former library JENDL-3.3 and other world-widely used latest libraries (ENDF/B-VII.0 and JEFF-3.1).

Calculation of Single Cell and Fuel Assembly IRIS Benchmarks Using WIMSD5B and GNOMER Codes

International Nuclear Information System (INIS)

Pevec, D.; Grgic, D.; Jecmenica, R.

2002-01-01

IRIS reactor (an acronym for International Reactor Innovative and Secure) is a modular, integral, light water cooled, small to medium power (100-335 MWe/module) reactor, which addresses the requirements defined by the United States Department of Energy for Generation IV nuclear energy systems, i.e., proliferation resistance, enhanced safety, improved economics, and waste reduction. An international consortium led by Westinghouse/BNFL was created for development of IRIS reactor; it includes universities, institutes, commercial companies, and utilities. Faculty of Electrical Engineering and Computing, University of Zagreb joined the consortium in year 2001, with the aim to take part in IRIS neutronics design and safety analyses of IRIS transients. A set of neutronic benchmarks for IRIS reactor was defined with the objective to compare results of all participants with exactly the same assumptions. In this paper a calculation of Benchmark 44 for IRIS reactor is described. Benchmark 44 is defined as a core depletion benchmark problem for specified IRIS reactor operating conditions (e.g., temperatures, moderator density) without feedback. Enriched boron, inhomogeneously distributed in axial direction, is used as an integral fuel burnable absorber (IFBA). The aim of this benchmark was to enable a more direct comparison of results of different code systems. Calculations of Benchmark 44 were performed using the modified CORD-2 code package. The CORD-2 code package consists of WIMSD and GNOMER codes. WIMSD is a well-known lattice spectrum calculation code. GNOMER solves the neutron diffusion equation in three-dimensional Cartesian geometry by the Green's function nodal method. The following parameters were obtained in Benchmark 44 analysis: effective multiplication factor as a function of burnup, nuclear peaking factor as a function of burnup, axial offset as a function of burnup, core-average axial power profile, core radial power profile, axial power profile for selected

Statistical Analysis of Reactor Pressure Vessel Fluence Calculation Benchmark Data Using Multiple Regression Techniques

International Nuclear Information System (INIS)

Carew, John F.; Finch, Stephen J.; Lois, Lambros

2003-01-01

The calculated >1-MeV pressure vessel fluence is used to determine the fracture toughness and integrity of the reactor pressure vessel. It is therefore of the utmost importance to ensure that the fluence prediction is accurate and unbiased. In practice, this assurance is provided by comparing the predictions of the calculational methodology with an extensive set of accurate benchmarks. A benchmarking database is used to provide an estimate of the overall average measurement-to-calculation (M/C) bias in the calculations ( ). This average is used as an ad-hoc multiplicative adjustment to the calculations to correct for the observed calculational bias. However, this average only provides a well-defined and valid adjustment of the fluence if the M/C data are homogeneous; i.e., the data are statistically independent and there is no correlation between subsets of M/C data.Typically, the identification of correlations between the errors in the database M/C values is difficult because the correlation is of the same magnitude as the random errors in the M/C data and varies substantially over the database. In this paper, an evaluation of a reactor dosimetry benchmark database is performed to determine the statistical validity of the adjustment to the calculated pressure vessel fluence. Physical mechanisms that could potentially introduce a correlation between the subsets of M/C ratios are identified and included in a multiple regression analysis of the M/C data. Rigorous statistical criteria are used to evaluate the homogeneity of the M/C data and determine the validity of the adjustment.For the database evaluated, the M/C data are found to be strongly correlated with dosimeter response threshold energy and dosimeter location (e.g., cavity versus in-vessel). It is shown that because of the inhomogeneity in the M/C data, for this database, the benchmark data do not provide a valid basis for adjusting the pressure vessel fluence.The statistical criteria and methods employed in

A Critical Thinking Benchmark for a Department of Agricultural Education and Studies

Science.gov (United States)

Perry, Dustin K.; Retallick, Michael S.; Paulsen, Thomas H.

2014-01-01

Due to an ever changing world where technology seemingly provides endless answers, today's higher education students must master a new skill set reflecting an emphasis on critical thinking, problem solving, and communications. The purpose of this study was to establish a departmental benchmark for critical thinking abilities of students majoring…

A proposal of a benchmark for calculation of the power distribution next to the absorber

International Nuclear Information System (INIS)

Temesvari, E.; Hordosy, G.; Maraczy, Cs.; Hegyi, Gy.; Kereszturi, A.

1999-01-01

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. (Authors)

PHEBUS-FPTO Benchmark calculations

International Nuclear Information System (INIS)

Shepherd, I.; Ball, A.; Trambauer, K.; Barbero, F.; Olivar Dominguez, F.; Herranz, L.; Biasi, L.; Fermandjian, J.; Hocke, K.

1991-01-01

This report summarizes a set of pre-test predictions made for the first Phebus-FP test, FPT-O. There were many different calculations, performed by various organizations and they represent the first attempt to calculate the whole experimental sequence, from bundle to containment. Quantitative agreement between the various calculations was not good but the particular models in the code responsible for disagreements were mostly identified. A consensus view was formed as to how the test would proceed. It was found that a successful execution of the test will require a different operating procedure than had been assumed here. Critical areas which require close attention are the need to devize a strategy for the power and flow in the bundle that takes account of uncertainties in the modelling and the shroud conductivity and the necessity to develop a reliable method to achieve the desired thermalhydraulic conditions in the containment

Computer simulation of Masurca critical and subcritical experiments. Muse-4 benchmark. Final report

International Nuclear Information System (INIS)

2006-01-01

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

An integral nodal variational method for multigroup criticality calculations

International Nuclear Information System (INIS)

Lewis, E.E.; Tsoulfanidis, N.

2003-01-01

An integral formulation of the variational nodal method is presented and applied to a series of benchmark critically problems. The method combines an integral transport treatment of the even-parity flux within the spatial node with an odd-parity spherical harmonics expansion of the Lagrange multipliers at the node interfaces. The response matrices that result from this formulation are compatible with those in the VARIANT code at Argonne National Laboratory. Either homogeneous or heterogeneous nodes may be employed. In general, for calculations requiring higher-order angular approximations, the integral method yields solutions with comparable accuracy while requiring substantially less CPU time and memory than the standard spherical harmonics expansion using the same spatial approximations. (author)

Calculational benchmark comparisons for a low sodium void worth actinide burner core design

International Nuclear Information System (INIS)

Hill, R.N.; Kawashima, M.; Arie, K.; Suzuki, M.

1992-01-01

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

Update of KASHIL-E6 library for shielding analysis and benchmark calculations

International Nuclear Information System (INIS)

Kim, D. H.; Kil, C. S.; Jang, J. H.

2004-01-01

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

Production of neutron cross section library based on JENDL-4.0 to continuous-energy Monte Carlo code MVP and its application to criticality analysis of benchmark problems in the ICSBEP handbook

International Nuclear Information System (INIS)

Okumura, Keisuke; Nagaya, Yasunobu

2011-09-01

In May 2010, JENDL-4.0 was released from Japan Atomic Energy Agency as the updated Japanese Nuclear Data Library. It was processed by the nuclear data processing system LICEM and an arbitrary-temperature neutron cross section library MVPlib - nJ40 was produced for the neutron and photon transport calculation code MVP based on the continuous-energy Monte Carlo method. The library contains neutron cross sections for 406 nuclides on the free gas model, thermal scattering cross sections, and cross sections of pseudo fission products for burn-up calculations with MVP. Criticality benchmark calculations were carried out with MVP and MVPlib - nJ40 for about 1,000 cases of critical experiments stored in the hand book of International Criticality Safety Benchmark Evaluation Project (ICSBEP), which covers a wide variety of fuel materials, fuel forms, and neutron spectra. We report all comparison results (C/E values) of effective neutron multiplication factors between calculations and experiments to give a validation data for the prediction accuracy of JENDL-4.0 for criticalities. (author)

Validation of neutron-transport calculations in benchmark facilities for improved damage-fluence predictions

International Nuclear Information System (INIS)

Williams, M.L.; Stallmann, F.W.; Maerker, R.E.; Kam, F.B.K.

1983-01-01

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

MCNP simulation of the TRIGA Mark II benchmark experiment

International Nuclear Information System (INIS)

Jeraj, R.; Glumac, B.; Maucec, M.

1996-01-01

The complete 3D MCNP model of the TRIGA Mark II reactor is presented. It enables precise calculations of some quantities of interest in a steady-state mode of operation. Calculational results are compared to the experimental results gathered during reactor reconstruction in 1992. Since the operating conditions were well defined at that time, the experimental results can be used as a benchmark. It may be noted that this benchmark is one of very few high enrichment benchmarks available. In our simulations experimental conditions were thoroughly simulated: fuel elements and control rods were precisely modeled as well as entire core configuration and the vicinity of the core. ENDF/B-VI and ENDF/B-V libraries were used. Partial results of benchmark calculations are presented. Excellent agreement of core criticality, excess reactivity and control rod worths can be observed. (author)

2010 Criticality Accident Alarm System Benchmark Experiments At The CEA Valduc SILENE Facility

International Nuclear Information System (INIS)

Miller, Thomas Martin; Dunn, Michael E.; Wagner, John C.; McMahan, Kimberly L.; Authier, Nicolas; Jacquet, Xavier; Rousseau, Guillaume; Wolff, Herve; Piot, Jerome; Savanier, Laurence; Baclet, Nathalie; Lee, Yi-kang; Masse, Veronique; Trama, Jean-Christophe; Gagnier, Emmanuel; Naury, Sylvie; Lenain, Richard; Hunter, Richard; Kim, Soon; Dulik, George Michael; Reynolds, Kevin H.

2011-01-01

Several experiments were performed at the CEA Valduc SILENE reactor facility, which are intended to be published as evaluated benchmark experiments in the ICSBEP Handbook. These evaluated benchmarks will be useful for the verification and validation of radiation transport codes and evaluated nuclear data, particularly those that are used in the analysis of CAASs. During these experiments SILENE was operated in pulsed mode in order to be representative of a criticality accident, which is rare among shielding benchmarks. Measurements of the neutron flux were made with neutron activation foils and measurements of photon doses were made with TLDs. Also unique to these experiments was the presence of several detectors used in actual CAASs, which allowed for the observation of their behavior during an actual critical pulse. This paper presents the preliminary measurement data currently available from these experiments. Also presented are comparisons of preliminary computational results with Scale and TRIPOLI-4 to the preliminary measurement data.

Monte Carlo benchmarking: Validation and progress

International Nuclear Information System (INIS)

Sala, P.

2010-01-01

Document available in abstract form only. Full text of publication follows: Calculational tools for radiation shielding at accelerators are faced with new challenges from the present and next generations of particle accelerators. All the details of particle production and transport play a role when dealing with huge power facilities, therapeutic ion beams, radioactive beams and so on. Besides the traditional calculations required for shielding, activation predictions have become an increasingly critical component. Comparison and benchmarking with experimental data is obviously mandatory in order to build up confidence in the computing tools, and to assess their reliability and limitations. Thin target particle production data are often the best tools for understanding the predictive power of individual interaction models and improving their performances. Complex benchmarks (e.g. thick target data, deep penetration, etc.) are invaluable in assessing the overall performances of calculational tools when all ingredients are put at work together. A review of the validation procedures of Monte Carlo tools will be presented with practical and real life examples. The interconnections among benchmarks, model development and impact on shielding calculations will be highlighted. (authors)

Accuracy of WWR-M criticality calculations with code MCU-RFFI

International Nuclear Information System (INIS)

Petrov, Yu.V.; Erykalov, A.N.; Onegin, M.S.

1999-01-01

The scattering and deviation of fuel element parameters by manufacturing, approximations of the reactor structure in the computer model, the partly inadequate neutron cross sections in the computer codes etc. lead to a discrepancy between the reactivity computations and data. We have compared reactivity calculations using the MCU-RRFI Monte Carlo code of critical assemblies containing WWR-M2 (36 enriched) an WWR-M5 (90%) fuel elements with benchmark experiments. The agreement was about Δρ≅±0.3%. A strong influence of the water ratio on reactivity was shown and a significant heterogeneous effect was found. We have also investigated, by full scale reactor calculations for the RETR program, the contribution to the reactivity of the main reactor structure elements: beryllium reflector, experimental channels irradiation devices inside the core, etc. Calculations show the importance of a more thorough study of the contributions of products of the (n, α) reaction in the Be reflector to the reactivity. Ways of improving the accuracy of the calculations are discussed. (author)

Accuracy of WWR-M criticality calculations with code MCU-RFFI

Energy Technology Data Exchange (ETDEWEB)

Petrov, Yu V [Petersburg Nuclear Physics Institute RAS, 188350 Gatchina, St. Petersburg (Russian Federation); Erykalov, A N; Onegin, M S [Petersburg Nuclear Physics Institute RAS, 188350 Gatchina, St. Petersburg (Russian Federation)

1999-10-01

The scattering and deviation of fuel element parameters by manufacturing, approximations of the reactor structure in the computer model, the partly inadequate neutron cross sections in the computer codes etc. lead to a discrepancy between the reactivity computations and data. We have compared reactivity calculations using the MCU-RRFI Monte Carlo code of critical assemblies containing WWR-M2 (36 enriched) an WWR-M5 (90%) fuel elements with benchmark experiments. The agreement was about {delta}{rho}{approx_equal}{+-}0.3%. A strong influence of the water ratio on reactivity was shown and a significant heterogeneous effect was found. We have also investigated, by full scale reactor calculations for the RETR program, the contribution to the reactivity of the main reactor structure elements: beryllium reflector, experimental channels irradiation devices inside the core, etc. Calculations show the importance of a more thorough study of the contributions of products of the (n, {alpha}) reaction in the Be reflector to the reactivity. Ways of improving the accuracy of the calculations are discussed. (author)

Benchmark experiment to verify radiation transport calculations for dosimetry in radiation therapy; Benchmark-Experiment zur Verifikation von Strahlungstransportrechnungen fuer die Dosimetrie in der Strahlentherapie

Energy Technology Data Exchange (ETDEWEB)

Renner, Franziska [Physikalisch-Technische Bundesanstalt (PTB), Braunschweig (Germany)

2016-11-01

Monte Carlo simulations are regarded as the most accurate method of solving complex problems in the field of dosimetry and radiation transport. In (external) radiation therapy they are increasingly used for the calculation of dose distributions during treatment planning. In comparison to other algorithms for the calculation of dose distributions, Monte Carlo methods have the capability of improving the accuracy of dose calculations - especially under complex circumstances (e.g. consideration of inhomogeneities). However, there is a lack of knowledge of how accurate the results of Monte Carlo calculations are on an absolute basis. A practical verification of the calculations can be performed by direct comparison with the results of a benchmark experiment. This work presents such a benchmark experiment and compares its results (with detailed consideration of measurement uncertainty) with the results of Monte Carlo calculations using the well-established Monte Carlo code EGSnrc. The experiment was designed to have parallels to external beam radiation therapy with respect to the type and energy of the radiation, the materials used and the kind of dose measurement. Because the properties of the beam have to be well known in order to compare the results of the experiment and the simulation on an absolute basis, the benchmark experiment was performed using the research electron accelerator of the Physikalisch-Technische Bundesanstalt (PTB), whose beam was accurately characterized in advance. The benchmark experiment and the corresponding Monte Carlo simulations were carried out for two different types of ionization chambers and the results were compared. Considering the uncertainty, which is about 0.7 % for the experimental values and about 1.0 % for the Monte Carlo simulation, the results of the simulation and the experiment coincide.

Development of M3C code for Monte Carlo reactor physics criticality calculations

International Nuclear Information System (INIS)

Kumar, Anek; Kannan, Umasankari; Krishanani, P.D.

2015-06-01

The development of Monte Carlo code (M3C) for reactor design entails use of continuous energy nuclear data and Monte Carlo simulations for each of the neutron interaction processes. BARC has started a concentrated effort for developing a new general geometry continuous energy Monte Carlo code for reactor physics calculation indigenously. The code development required a comprehensive understanding of the basic continuous energy cross section sets. The important features of this code are treatment of heterogeneous lattices by general geometry, use of point cross sections along with unionized energy grid approach, thermal scattering model for low energy treatment, capability of handling the microscopic fuel particles dispersed randomly. The capability of handling the randomly dispersed microscopic fuel particles which is very useful for the modeling of High-Temperature Gas-Cooled reactor fuels which are composed of thousands of microscopic fuel particle (TRISO fuel particle), randomly dispersed in a graphite matrix. The Monte Carlo code for criticality calculation is a pioneering effort and has been used to study several types of lattices including cluster geometries. The code has been verified for its accuracy against more than 60 sample problems covering a wide range from simple (like spherical) to complex geometry (like PHWR lattice). Benchmark results show that the code performs quite well for the criticality calculation of the system. In this report, the current status of the code, features of the code, some of the benchmark results for the testing of the code and input preparation etc. are discussed. (author)

Hextran-Smabre calculation of the VVER-1000 coolant transient benchmark

Energy Technology Data Exchange (ETDEWEB)

Elina Syrjaelahti; Anitta Haemaelaeinen [VTT Processes, P.O.Box 1604, FIN-02044 VTT (Finland)

2005-07-01

Full text of publication follows: The VVER-1000 Coolant Transient benchmark is intended for validation of couplings of the thermal hydraulic codes and three dimensional neutron kinetic core models. It concerns a switching on a main coolant pump when the other three main coolant pumps are in operation. Problem is based on experiment performed in Kozloduy NPP in Bulgaria. In addition to the real plant transient, two extreme scenarios concerning control rod ejection after switching on a main coolant pump were calculated. In VTT the three-dimensional advanced nodal code HEXTRAN is used for the core kinetics and dynamics, and thermohydraulic system code SMABRE as a thermal hydraulic model for the primary and secondary loop. Parallelly coupled HEXTRAN-SMABRE code has been in production use since early 90's, and it has been extensively used for analysis of VVER NPPs. The SMABRE input model is based on the standard VVER-1000 input used in VTT. Last plant specific modifications to the input model have been made in EU projects. The whole core calculation is performed in the core with HEXTRAN. Also the core model is based on earlier VVER-1000 models. Nuclear data for the calculation was specified in the benchmark. The paper outlines the input models used for both codes. Calculated results are introduced both for the coupled core system with inlet and outlet boundary conditions and for the whole plant model. Sensitivity studies have been performed for selected parameters. (authors)

Hextran-Smabre calculation of the VVER-1000 coolant transient benchmark

International Nuclear Information System (INIS)

Elina Syrjaelahti; Anitta Haemaelaeinen

2005-01-01

Full text of publication follows: The VVER-1000 Coolant Transient benchmark is intended for validation of couplings of the thermal hydraulic codes and three dimensional neutron kinetic core models. It concerns a switching on a main coolant pump when the other three main coolant pumps are in operation. Problem is based on experiment performed in Kozloduy NPP in Bulgaria. In addition to the real plant transient, two extreme scenarios concerning control rod ejection after switching on a main coolant pump were calculated. In VTT the three-dimensional advanced nodal code HEXTRAN is used for the core kinetics and dynamics, and thermohydraulic system code SMABRE as a thermal hydraulic model for the primary and secondary loop. Parallelly coupled HEXTRAN-SMABRE code has been in production use since early 90's, and it has been extensively used for analysis of VVER NPPs. The SMABRE input model is based on the standard VVER-1000 input used in VTT. Last plant specific modifications to the input model have been made in EU projects. The whole core calculation is performed in the core with HEXTRAN. Also the core model is based on earlier VVER-1000 models. Nuclear data for the calculation was specified in the benchmark. The paper outlines the input models used for both codes. Calculated results are introduced both for the coupled core system with inlet and outlet boundary conditions and for the whole plant model. Sensitivity studies have been performed for selected parameters. (authors)

Elasto-plastic benchmark calculations. Step 1: verification of the numerical accuracy of the computer programs

International Nuclear Information System (INIS)

Corsi, F.

1985-01-01

In connection with the design of nuclear reactors components operating at elevated temperature, design criteria need a level of realism in the prediction of inelastic structural behaviour. This concept leads to the necessity of developing non linear computer programmes, and, as a consequence, to the problems of verification and qualification of these tools. Benchmark calculations allow to carry out these two actions, involving at the same time an increased level of confidence in complex phenomena analysis and in inelastic design calculations. With the financial and programmatic support of the Commission of the European Communities (CEE) a programme of elasto-plastic benchmark calculations relevant to the design of structural components for LMFBR has been undertaken by those Member States which are developing a fast reactor project. Four principal progressive aims were initially pointed out that brought to the decision to subdivide the Benchmark effort in a calculations series of four sequential steps: step 1 to 4. The present document tries to summarize Step 1 of the Benchmark exercise, to derive some conclusions on Step 1 by comparison of the results obtained with the various codes and to point out some concluding comments on the first action. It is to point out that even if the work was designed to test the capabilities of the computer codes, another aim was to increase the skill of the users concerned

Benchmarking ENDF/B-VII.1, JENDL-4.0 and JEFF-3.1.1 with MCNP6

International Nuclear Information System (INIS)

Marck, Steven C. van der

2012-01-01

Recent releases of three major world nuclear reaction data libraries, ENDF/B-VII.1, JENDL-4.0, and JEFF-3.1.1, have been tested extensively using benchmark calculations. The calculations were performed with the latest release of the continuous energy Monte Carlo neutronics code MCNP, i.e. MCNP6. Three types of benchmarks were used, viz. criticality safety benchmarks, (fusion) shielding benchmarks, and reference systems for which the effective delayed neutron fraction is reported. For criticality safety, more than 2000 benchmarks from the International Handbook of Criticality Safety Benchmark Experiments were used. Benchmarks from all categories were used, ranging from low-enriched uranium, compound fuel, thermal spectrum ones (LEU-COMP-THERM), to mixed uranium-plutonium, metallic fuel, fast spectrum ones (MIX-MET-FAST). For fusion shielding many benchmarks were based on IAEA specifications for the Oktavian experiments (for Al, Co, Cr, Cu, LiF, Mn, Mo, Si, Ti, W, Zr), Fusion Neutronics Source in Japan (for Be, C, N, O, Fe, Pb), and Pulsed Sphere experiments at Lawrence Livermore National Laboratory (for 6 Li, 7 Li, Be, C, N, O, Mg, Al, Ti, Fe, Pb, D2O, H2O, concrete, polyethylene and teflon). The new functionality in MCNP6 to calculate the effective delayed neutron fraction was tested by comparison with more than thirty measurements in widely varying systems. Among these were measurements in the Tank Critical Assembly (TCA in Japan) and IPEN/MB-01 (Brazil), both with a thermal spectrum, two cores in Masurca (France) and three cores in the Fast Critical Assembly (FCA, Japan), all with fast spectra. The performance of the three libraries, in combination with MCNP6, is shown to be good. The results for the LEU-COMP-THERM category are on average very close to the benchmark value. Also for most other categories the results are satisfactory. Deviations from the benchmark values do occur in certain benchmark series, or in isolated cases within benchmark series. Such

Benchmark calculations with simple phantom for neutron dosimetry (2)

International Nuclear Information System (INIS)

Yukio, Sakamoto; Shuichi, Tsuda; Tatsuhiko, Sato; Nobuaki, Yoshizawa; Hideo, Hirayama

2004-01-01

Benchmark calculations for high-energy neutron dosimetry were undertaken after SATIF-5. Energy deposition in a cylindrical phantom with 100 cm radius and 30 cm depth was calculated for the irradiation of neutrons from 100 MeV to 10 GeV. Using the ICRU four-element loft tissue phantom and four single-element (hydrogen, carbon, nitrogen and oxygen) phantoms, the depth distributions of deposition energy and those total at the central region of phantoms within l cm radius and at the whole region of phantoms within 100 cm radius were calculated. The calculated results of FLUKA, MCNPX, MARS, HETC-3STEP and NMTC/JAM codes were compared. It was found that FLUKA, MARS and NMTC/JAM showed almost the same results. For the high-energy neutron incident, the MCNP-X results showed the largest ones in the total deposition energy and the HETC-3STEP results show'ed smallest ones. (author)

Summary report on the international comparison of NEACRP burnup benchmark calculations for high conversion light water reactor lattices

International Nuclear Information System (INIS)

Akie, Hiroshi; Ishiguro, Yukio; Takano, Hideki

1988-10-01

The results of the NEACRP HCLWR cell burnup benchmark calculations are summarized in this report. Fifteen organizations from eight countries participated in this benchmark and submitted twenty solutions. Large differences are still observed among the calculated values of void reactivities and conversion ratios. These differences are mainly caused from the discrepancies in the reaction rates of U-238, Pu-239 and fission products. The physics problems related to these results are briefly investigated in the report. In the specialists' meeting on this benchmark calculations held in April 1988, it was recommended to perform continuous energy Monte Carlo calculations in order to obtain reference solutions for design codes. The conclusions resulted from the specialists' meeting are also presented. (author)

US/JAERI calculational benchmarks for nuclear data and codes intercomparison. Article 8

International Nuclear Information System (INIS)

Youssef, M.Z.; Jung, J.; Sawan, M.E.; Nakagawa, M.; Mori, T.; Kosako, K.

1986-01-01

Prior to analyzing the integral experiments performed at the FNS facility at JAERI, both US and JAERI's analysts have agreed upon four calculational benchmark problems proposed by JAERI to intercompare results based on various codes and data base used independently by both countries. To compare codes the same data base is used (ENDF/B-IV). To compare nuclear data libraries, common codes were applied. Some of the benchmarks chosen were geometrically simple and consisted of a single material to clearly identify sources of discrepancies and thus help in analysing the integral experiments

MCNP and OMEGA criticality calculations

International Nuclear Information System (INIS)

Seifert, E.

1998-04-01

The reliability of OMEGA criticality calculations is shown by a comparison with calculations by the validated and widely used Monte Carlo code MCNP. The criticality of 16 assemblies with uranium as fissionable is calculated with the codes MCNP (Version 4A, ENDF/B-V cross sections), MCNP (Version 4B, ENDF/B-VI cross sections), and OMEGA. Identical calculation models are used for the three codes. The results are compared mutually and with the experimental criticality of the assemblies. (orig.)

Benchmark calculation for the steady-state temperature distribution of the HTR-10 under full-power operation

International Nuclear Information System (INIS)

Chen Fubing; Dong Yujie; Zheng Yanhua; Shi Lei; Zhang Zuoyi

2009-01-01

Within the framework of a Coordinated Research Project on Evaluation of High Temperature Gas-Cooled Reactor Performance (CRP-5) initiated by the International Atomic Energy Agency (IAEA), the calculation of steady-state temperature distribution of the 10 MW High Temperature Gas-Cooled Reactor-Test Module (HTR-10) under its initial full power experimental operation has been defined as one of the benchmark problems. This paper gives the investigation results obtained by different countries who participate in solving this benchmark problem. The validation works of the THERMIX code used by the Institute of Nuclear and New Energy Technology (INET) are also presented. For the benchmark items defined in this CRP, various calculation results correspond well with each other and basically agree the experimental results. Discrepancies existing among various code results are preliminarily attributed to different methods, models, material properties, and so on used in the computations. Temperatures calculated by THERMIX for the measuring points in the reactor internals agree well with the experimental values. The maximum fuel center temperatures calculated by the participants are much lower than the limited value of 1,230degC. According to the comparison results of code-to-code as well as code-to-experiment, THERMIX is considered to reproduce relatively satisfactory results for the CRP-5 benchmark problem. (author)

Three-dimensional RAMA fluence methodology benchmarking

International Nuclear Information System (INIS)

Baker, S. P.; Carter, R. G.; Watkins, K. E.; Jones, D. B.

2004-01-01

This paper describes the benchmarking of the RAMA Fluence Methodology software, that has been performed in accordance with U. S. Nuclear Regulatory Commission Regulatory Guide 1.190. The RAMA Fluence Methodology has been developed by TransWare Enterprises Inc. through funding provided by the Electric Power Research Inst., Inc. (EPRI) and the Boiling Water Reactor Vessel and Internals Project (BWRVIP). The purpose of the software is to provide an accurate method for calculating neutron fluence in BWR pressure vessels and internal components. The Methodology incorporates a three-dimensional deterministic transport solution with flexible arbitrary geometry representation of reactor system components, previously available only with Monte Carlo solution techniques. Benchmarking was performed on measurements obtained from three standard benchmark problems which include the Pool Criticality Assembly (PCA), VENUS-3, and H. B. Robinson Unit 2 benchmarks, and on flux wire measurements obtained from two BWR nuclear plants. The calculated to measured (C/M) ratios range from 0.93 to 1.04 demonstrating the accuracy of the RAMA Fluence Methodology in predicting neutron flux, fluence, and dosimetry activation. (authors)

Standard problem exercise to validate criticality codes for spent LWR fuel transport container calculations

International Nuclear Information System (INIS)

Whitesides, G.H.; Stephens, M.E.

1984-01-01

During the past two years, a Working Group established by the Organization for Economic Co-Operation and Development's Nuclear Energy Agency (OECD-NEA) has been developing a set of criticality benchmark problems which could be used to help establish the validity of criticality safety computer programs and their associated nuclear data for calculation of ksub(eff) for spent light water reactor (LWR) fuel transport containers. The basic goal of this effort was to identify a set of actual critical experiments which would contain the various material and geometric properties present in spent LWR transport contrainers. These data, when used by the various computational methods, are intended to demonstrate the ability of each method to accurately reproduce the experimentally measured ksub(eff) for the parameters under consideration

International report to validate criticality safety calculations for fissile material transport

International Nuclear Information System (INIS)

Whitesides, G.E.

1984-01-01

During the past three years a Working Group established by the Organization for Economic Co-operation and Development's Nuclear Energy Agency (OECD-NEA) in Paris, France, has been studying the validity and applicability of a variety of criticality safety computer programs and their associated nuclear data for the computation of the neutron multiplication factor, k/sub eff/, for various transport packages used in the fuel cycle. The principal objective of this work has been to provide an internationally acceptable basis for the licensing authorities in a country to honor licensing approvals granted by other participating countries. Eleven countries participated in the initial study which consisted of examining criticality safety calculations for packages designed for spent light water reactor fuel transport. This paper presents a summary of this study which has been completed and reported in an OECD-NEA Report No. CSNI-71. The basic goal of this study was to outline a satisfactory validation procedure for this particular application. First, a set of actual critical experiments were chosen which contained the various material and geometric properties present in typical LWR transport containers. Secondly, calculations were made by each of the methods in order to determine how accurately each method reproduced the experimental values. This successful effort in developing a benchmark procedure for validating criticality calculations for spent LWR transport packages along with the successful intercomparison of a number of methods should provide increased confidence by licensing authorities in the use of these methods for this area of application. 4 references, 2 figures

Benchmark calculations on residue production within the EURISOL DS project; Part I: thin targets

CERN Document Server

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

CERN Document Server

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.

Determining the nuclear data uncertainty on MONK10 and WIMS10 criticality calculations

Science.gov (United States)

Ware, Tim; Dobson, Geoff; Hanlon, David; Hiles, Richard; Mason, Robert; Perry, Ray

2017-09-01

The ANSWERS Software Service is developing a number of techniques to better understand and quantify uncertainty on calculations of the neutron multiplication factor, k-effective, in nuclear fuel and other systems containing fissile material. The uncertainty on the calculated k-effective arises from a number of sources, including nuclear data uncertainties, manufacturing tolerances, modelling approximations and, for Monte Carlo simulation, stochastic uncertainty. For determining the uncertainties due to nuclear data, a set of application libraries have been generated for use with the MONK10 Monte Carlo and the WIMS10 deterministic criticality and reactor physics codes. This paper overviews the generation of these nuclear data libraries by Latin hypercube sampling of JEFF-3.1.2 evaluated data based upon a library of covariance data taken from JEFF, ENDF/B, JENDL and TENDL evaluations. Criticality calculations have been performed with MONK10 and WIMS10 using these sampled libraries for a number of benchmark models of fissile systems. Results are presented which show the uncertainty on k-effective for these systems arising from the uncertainty on the input nuclear data.

Gas cooled fast reactor benchmarks for JNC and Cea neutronic tools assessment

International Nuclear Information System (INIS)

Rimpault, G.; Sugino, K.; Hayashi, H.

2005-01-01

In order to verify the adequacy of JNC and Cea computational tools for the definition of GCFR (gas cooled fast reactor) core characteristics, GCFR neutronic benchmarks have been performed. The benchmarks have been carried out on two different cores: 1) a conventional Gas-Cooled fast Reactor (EGCR) core with pin-type fuel, and 2) an innovative He-cooled Coated-Particle Fuel (CPF) core. Core characteristics being studied include: -) Criticality (Effective multiplication factor or K-effective), -) Instantaneous breeding gain (BG), -) Core Doppler effect, and -) Coolant depressurization reactivity. K-effective and coolant depressurization reactivity at EOEC (End Of Equilibrium Cycle) state were calculated since these values are the most critical characteristics in the core design. In order to check the influence due to the difference of depletion calculation systems, a simple depletion calculation benchmark was performed. Values such as: -) burnup reactivity loss, -) mass balance of heavy metals and fission products (FP) were calculated. Results of the core design characteristics calculated by both JNC and Cea sides agree quite satisfactorily in terms of core conceptual design study. Potential features for improving the GCFR computational tools have been discovered during the course of this benchmark such as the way to calculate accurately the breeding gain. Different ways to improve the accuracy of the calculations have also been identified. In particular, investigation on nuclear data for steel is important for EGCR and for lumped fission products in both cores. The outcome of this benchmark is already satisfactory and will help to design more precisely GCFR cores. (authors)

MCNP neutron benchmarks

International Nuclear Information System (INIS)

Hendricks, J.S.; Whalen, D.J.; Cardon, D.A.; Uhle, J.L.

1991-01-01

Over 50 neutron benchmark calculations have recently been completed as part of an ongoing program to validate the MCNP Monte Carlo radiation transport code. The new and significant aspects of this work are as follows: These calculations are the first attempt at a validation program for MCNP and the first official benchmarking of version 4 of the code. We believe the chosen set of benchmarks is a comprehensive set that may be useful for benchmarking other radiation transport codes and data libraries. These calculations provide insight into how well neutron transport calculations can be expected to model a wide variety of problems

OECD/NEA benchmark for time-dependent neutron transport calculations without spatial homogenization

Energy Technology Data Exchange (ETDEWEB)

Hou, Jason, E-mail: jason.hou@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Ivanov, Kostadin N. [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Boyarinov, Victor F.; Fomichenko, Peter A. [National Research Centre “Kurchatov Institute”, Kurchatov Sq. 1, Moscow (Russian Federation)

2017-06-15

Highlights: • A time-dependent homogenization-free neutron transport benchmark was created. • The first phase, known as the kinetics phase, was described in this work. • Preliminary results for selected 2-D transient exercises were presented. - Abstract: A Nuclear Energy Agency (NEA), Organization for Economic Co-operation and Development (OECD) benchmark for the time-dependent neutron transport calculations without spatial homogenization has been established in order to facilitate the development and assessment of numerical methods for solving the space-time neutron kinetics equations. The benchmark has been named the OECD/NEA C5G7-TD benchmark, and later extended with three consecutive phases each corresponding to one modelling stage of the multi-physics transient analysis of the nuclear reactor core. This paper provides a detailed introduction of the benchmark specification of Phase I, known as the “kinetics phase”, including the geometry description, supporting neutron transport data, transient scenarios in both two-dimensional (2-D) and three-dimensional (3-D) configurations, as well as the expected output parameters from the participants. Also presented are the preliminary results for the initial state 2-D core and selected transient exercises that have been obtained using the Monte Carlo method and the Surface Harmonic Method (SHM), respectively.

Benchmark Evaluation of HTR-PROTEUS Pebble Bed Experimental Program

International Nuclear Information System (INIS)

Bess, John D.; Montierth, Leland; Köberl, Oliver

2014-01-01

Benchmark models were developed to evaluate 11 critical core configurations of the HTR-PROTEUS pebble bed experimental program. Various additional reactor physics measurements were performed as part of this program; currently only a total of 37 absorber rod worth measurements have been evaluated as acceptable benchmark experiments for Cores 4, 9, and 10. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the 235 U enrichment of the fuel, impurities in the moderator pebbles, and the density and impurity content of the radial reflector. Calculations of k eff with MCNP5 and ENDF/B-VII.0 neutron nuclear data are greater than the benchmark values but within 1% and also within the 3σ uncertainty, except for Core 4, which is the only randomly packed pebble configuration. Repeated calculations of k eff with MCNP6.1 and ENDF/B-VII.1 are lower than the benchmark values and within 1% (~3σ) except for Cores 5 and 9, which calculate lower than the benchmark eigenvalues within 4σ. The primary difference between the two nuclear data libraries is the adjustment of the absorption cross section of graphite. Simulations of the absorber rod worth measurements are within 3σ of the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments

Montecarlo calculation for a benchmark on interactive effects of Gadolinium poisoned pins in BWRs

International Nuclear Information System (INIS)

Borgia, M.G.; Casali, F.; Cepraga, D.

1985-01-01

K infinite and burn-up calculations have been done in the frame of a benchmark organized by Physic Reactor Committee of NEA. The calculations, performed by the Montecarlo code KIM, concerned BWR lattices having UO*L2 fuel rodlets with and without gadolinium oxide

HELIOS calculations for UO2 lattice benchmarks

International Nuclear Information System (INIS)

Mosteller, R.D.

1998-01-01

Calculations for the ANS UO 2 lattice benchmark have been performed with the HELIOS lattice-physics code and six of its cross-section libraries. The results obtained from the different libraries permit conclusions to be drawn regarding the adequacy of the energy group structures and of the ENDF/B-VI evaluation for 238 U. Scandpower A/S, the developer of HELIOS, provided Los Alamos National Laboratory with six different cross section libraries. Three of the libraries were derived directly from Release 3 of ENDF/B-VI (ENDF/B-VI.3) and differ only in the number of groups (34, 89 or 190). The other three libraries are identical to the first three except for a modification to the cross sections for 238 U in the resonance range

Proposal of a benchmark for core burnup calculations for a VVER-1000 reactor core

International Nuclear Information System (INIS)

Loetsch, T.; Khalimonchuk, V.; Kuchin, A.

2009-01-01

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)

Benchmark calculations on resonance absorption by 238U in a PWR pin-cell geometry

International Nuclear Information System (INIS)

Kruijf, W.J.M. de; Janssen, A.J.

1993-12-01

Very accurate Monte Carlo calculations with MCNP have been performed to serve as a reference for benchmark calculations on resonance absorption by 238 U in a typical PWR pin-cell geometry. Calculations with the energy-pointwise slowing down code ROLAIDS-CPM show that this code calculates the resonance absorption accurately. Calculations with the multigroup discrete ordinates code XSDRN show that accurate results can only be achieved with a very fine energy mesh. (orig.)

Benefits of the delta K of depletion benchmarks for burnup credit validation

International Nuclear Information System (INIS)

Lancaster, D.; Machiels, A.

2012-01-01

Pressurized Water Reactor (PWR) burnup credit validation is demonstrated using the benchmarks for quantifying fuel reactivity decrements, published as 'Benchmarks for Quantifying Fuel Reactivity Depletion Uncertainty,' EPRI Report 1022909 (August 2011). This demonstration uses the depletion module TRITON available in the SCALE 6.1 code system followed by criticality calculations using KENO-Va. The difference between the predicted depletion reactivity and the benchmark's depletion reactivity is a bias for the criticality calculations. The uncertainty in the benchmarks is the depletion reactivity uncertainty. This depletion bias and uncertainty is used with the bias and uncertainty from fresh UO 2 critical experiments to determine the criticality safety limits on the neutron multiplication factor, k eff . The analysis shows that SCALE 6.1 with the ENDF/B-VII 238-group cross section library supports the use of a depletion bias of only 0.0015 in delta k if cooling is ignored and 0.0025 if cooling is credited. The uncertainty in the depletion bias is 0.0064. Reliance on the ENDF/B V cross section library produces much larger disagreement with the benchmarks. The analysis covers numerous combinations of depletion and criticality options. In all cases, the historical uncertainty of 5% of the delta k of depletion ('Kopp memo') was shown to be conservative for fuel with more than 30 GWD/MTU burnup. Since this historically assumed burnup uncertainty is not a function of burnup, the Kopp memo's recommended bias and uncertainty may be exceeded at low burnups, but its absolute magnitude is small. (authors)

Critical Assessment of Metagenome Interpretation-a benchmark of metagenomics software.

Science.gov (United States)

Sczyrba, Alexander; Hofmann, Peter; Belmann, Peter; Koslicki, David; Janssen, Stefan; Dröge, Johannes; Gregor, Ivan; Majda, Stephan; Fiedler, Jessika; Dahms, Eik; Bremges, Andreas; Fritz, Adrian; Garrido-Oter, Ruben; Jørgensen, Tue Sparholt; Shapiro, Nicole; Blood, Philip D; Gurevich, Alexey; Bai, Yang; Turaev, Dmitrij; DeMaere, Matthew Z; Chikhi, Rayan; Nagarajan, Niranjan; Quince, Christopher; Meyer, Fernando; Balvočiūtė, Monika; Hansen, Lars Hestbjerg; Sørensen, Søren J; Chia, Burton K H; Denis, Bertrand; Froula, Jeff L; Wang, Zhong; Egan, Robert; Don Kang, Dongwan; Cook, Jeffrey J; Deltel, Charles; Beckstette, Michael; Lemaitre, Claire; Peterlongo, Pierre; Rizk, Guillaume; Lavenier, Dominique; Wu, Yu-Wei; Singer, Steven W; Jain, Chirag; Strous, Marc; Klingenberg, Heiner; Meinicke, Peter; Barton, Michael D; Lingner, Thomas; Lin, Hsin-Hung; Liao, Yu-Chieh; Silva, Genivaldo Gueiros Z; Cuevas, Daniel A; Edwards, Robert A; Saha, Surya; Piro, Vitor C; Renard, Bernhard Y; Pop, Mihai; Klenk, Hans-Peter; Göker, Markus; Kyrpides, Nikos C; Woyke, Tanja; Vorholt, Julia A; Schulze-Lefert, Paul; Rubin, Edward M; Darling, Aaron E; Rattei, Thomas; McHardy, Alice C

2017-11-01

Methods for assembly, taxonomic profiling and binning are key to interpreting metagenome data, but a lack of consensus about benchmarking complicates performance assessment. The Critical Assessment of Metagenome Interpretation (CAMI) challenge has engaged the global developer community to benchmark their programs on highly complex and realistic data sets, generated from ∼700 newly sequenced microorganisms and ∼600 novel viruses and plasmids and representing common experimental setups. Assembly and genome binning programs performed well for species represented by individual genomes but were substantially affected by the presence of related strains. Taxonomic profiling and binning programs were proficient at high taxonomic ranks, with a notable performance decrease below family level. Parameter settings markedly affected performance, underscoring their importance for program reproducibility. The CAMI results highlight current challenges but also provide a roadmap for software selection to answer specific research questions.

Neutron transport calculations of some fast critical assemblies

Energy Technology Data Exchange (ETDEWEB)

Martinez-Val Penalosa, J A

1976-07-01

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.

Neutron transport calculations of some fast critical assemblies

International Nuclear Information System (INIS)

Martinez-Val Penalosa, J. A.

1976-01-01

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

Benchmarking criticality analysis of TRIGA fuel storage racks.

Science.gov (United States)

Robinson, Matthew Loren; DeBey, Timothy M; Higginbotham, Jack F

2017-01-01

A criticality analysis was benchmarked to sub-criticality measurements of the hexagonal fuel storage racks at the United States Geological Survey TRIGA MARK I reactor in Denver. These racks, which hold up to 19 fuel elements each, are arranged at 0.61m (2 feet) spacings around the outer edge of the reactor. A 3-dimensional model was created of the racks using MCNP5, and the model was verified experimentally by comparison to measured subcritical multiplication data collected in an approach to critical loading of two of the racks. The validated model was then used to show that in the extreme condition where the entire circumference of the pool was lined with racks loaded with used fuel the storage array is subcritical with a k value of about 0.71; well below the regulatory limit of 0.8. A model was also constructed of the rectangular 2×10 fuel storage array used in many other TRIGA reactors to validate the technique against the original TRIGA licensing sub-critical analysis performed in 1966. The fuel used in this study was standard 20% enriched (LEU) aluminum or stainless steel clad TRIGA fuel. Copyright © 2016. Published by Elsevier Ltd.

Accelerator driven systems. ADS benchmark calculations. Results of stage 2. Radiotoxic waste transmutation

Energy Technology Data Exchange (ETDEWEB)

Freudenreich, W.E.; Gruppelaar, H

1998-12-01

This report contains the results of calculations made at ECN-Petten of a benchmark to study the neutronic potential of a modular fast spectrum ADS (Accelerator-Driven System) for radiotoxic waste transmutation. The study is focused on the incineration of TRans-Uranium elements (TRU), Minor Actinides (MA) and Long-Lived Fission Products (LLFP), in this case {sup 99}Tc. The benchmark exercise is made in the framework of an IAEA Co-ordinated Research Programme. A simplified description of an ADS, restricted to the reactor part, with TRU or MA fuel (k{sub eff}=0.96) has been analysed. All spectrum calculations have been performed with the Monte Carlo code MCNP-4A. The burnup calculations have been performed with the code FISPACT coupled to MCNP-4A by means of our OCTOPUS system. The cross sections are based upon JEF-2.2 for transport calculations and supplemented with EAF-4 data for inventory calculations. The determined quantities are: core dimensions, fuel inventories, system power, sensitivity on external source spectrum and waste transmutation rates. The main conclusions are: The MA-burner requires only a small accelerator current increase during burnup, in contrast to the TRU-burner. The {sup 99} Tc-burner has a large initial loading; a more effective design may be possible. 5 refs.

Accelerator driven systems. ADS benchmark calculations. Results of stage 2. Radiotoxic waste transmutation

International Nuclear Information System (INIS)

Freudenreich, W.E.; Gruppelaar, H.

1998-12-01

This report contains the results of calculations made at ECN-Petten of a benchmark to study the neutronic potential of a modular fast spectrum ADS (Accelerator-Driven System) for radiotoxic waste transmutation. The study is focused on the incineration of TRans-Uranium elements (TRU), Minor Actinides (MA) and Long-Lived Fission Products (LLFP), in this case 99 Tc. The benchmark exercise is made in the framework of an IAEA Co-ordinated Research Programme. A simplified description of an ADS, restricted to the reactor part, with TRU or MA fuel (k eff =0.96) has been analysed. All spectrum calculations have been performed with the Monte Carlo code MCNP-4A. The burnup calculations have been performed with the code FISPACT coupled to MCNP-4A by means of our OCTOPUS system. The cross sections are based upon JEF-2.2 for transport calculations and supplemented with EAF-4 data for inventory calculations. The determined quantities are: core dimensions, fuel inventories, system power, sensitivity on external source spectrum and waste transmutation rates. The main conclusions are: The MA-burner requires only a small accelerator current increase during burnup, in contrast to the TRU-burner. The 99 Tc-burner has a large initial loading; a more effective design may be possible. 5 refs

Calculational assessment of critical experiments with mixed-oxide fuel pin arrays moderated by organic solution

International Nuclear Information System (INIS)

Smolen, G.R.; Funabashi, H.

1987-01-01

Critical experiments have been conducted with organically moderated mixed-oxide (MOX) fuel pin assemblies at the Pacific Northwest Lab. Critical Mass Lab. These experiments are part of a joint exchange program between the US Dept. of Energy and the Power Reactor and Nuclear Fuel Development Corp. of Japan in the area of criticality data development. The purpose of these experiments is to benchmark computer codes and cross-section libraries and to assess the reactivity difference between systems moderated by water and those moderated by an organic solution. Past studies have indicated that some organic mixtures may be better moderators than water. This topic is of particular importance to the criticality safety of fuel processing plants where fissile material is dissolved in organic solutions during the solvent extraction process. In the past, it has been assumed that the codes and libraries benchmarked with water-moderated experiments were adequate when performing design and licensing studies of organically moderated systems. Calculations presented in this paper indicated that the Scale code system and the 27-energy-group cross-section library accurately compute k/sub eff/ for organically moderated MOX fuel pin assemblies. Furthermore, the reactivity of an organic solution with a 32 vol % TBP/68 vol% NPH mixture in a heterogeneous configuration is the same, for practical purposes, as water

Calculational assessment of critical experiments with mixed oxide fuel pin arrays moderated by organic solution

International Nuclear Information System (INIS)

Smolen, G.R.

1987-01-01

Critical experiments have been conducted with organic-moderated mixed oxide (MOX) fuel pin assemblies at the Pacific Northwest Laboratory (PNL) Critical Mass Laboratory (CML). These experiments are part of a joint exchange program between the United States Department of Energy (USDOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan in the area of criticality data development. The purpose of these experiments is to benchmark computer codes and cross-section libraries and to assess the reactivity difference between systems moderated by water and those moderated by an organic solution. Past studies have indicated that some organic mixtures may be better moderators than water. This topic is of particular importance to the criticality safety of fuel processing plants where fissile material is dissolved in organic solutions during the solvent extraction process. In the past, it has been assumed that the codes and libraries benchmarked with water-moderated experiments were adequate when performing design and licensing studies of organic-moderated systems. Calculations presented in this paper indicated that the SCALE code system and the 27-energy-group cross-section accurately compute k-effectives for organic moderated MOX fuel-pin assemblies. Furthermore, the reactivity of an organic solution with a 32-vol-% TBP/68-vol-% NPH mixture in a heterogeneous configuration is the same, for practical purposes, as water. 5 refs

Status on benchmark testing of CENDL-3

CERN Document Server

Liu Ping

2002-01-01

CENDL-3, the newest version of China Evaluated Nuclear Data Library has been finished, and distributed for some benchmarks analysis recently. The processing was carried out using the NJOY nuclear data processing code system. The calculations and analysis of benchmarks were done with Monte Carlo code MCNP and reactor lattice code WIMSD5A. The calculated results were compared with the experimental results based on ENDF/B6. In most thermal and fast uranium criticality benchmarks, the calculated k sub e sub f sub f values with CENDL-3 were in good agreements with experimental results. In the plutonium fast cores, the k sub e sub f sub f values were improved significantly with CENDL-3. This is duo to reevaluation of the fission spectrum and elastic angular distributions of sup 2 sup 3 sup 9 Pu and sup 2 sup 4 sup 0 Pu. CENDL-3 underestimated the k sub e sub f sub f values compared with other evaluated data libraries for most spherical or cylindrical assemblies of plutonium or uranium with beryllium

Determining the nuclear data uncertainty on MONK10 and WIMS10 criticality calculations

Directory of Open Access Journals (Sweden)

Ware Tim

2017-01-01

Full Text Available The ANSWERS Software Service is developing a number of techniques to better understand and quantify uncertainty on calculations of the neutron multiplication factor, k-effective, in nuclear fuel and other systems containing fissile material. The uncertainty on the calculated k-effective arises from a number of sources, including nuclear data uncertainties, manufacturing tolerances, modelling approximations and, for Monte Carlo simulation, stochastic uncertainty. For determining the uncertainties due to nuclear data, a set of application libraries have been generated for use with the MONK10 Monte Carlo and the WIMS10 deterministic criticality and reactor physics codes. This paper overviews the generation of these nuclear data libraries by Latin hypercube sampling of JEFF-3.1.2 evaluated data based upon a library of covariance data taken from JEFF, ENDF/B, JENDL and TENDL evaluations. Criticality calculations have been performed with MONK10 and WIMS10 using these sampled libraries for a number of benchmark models of fissile systems. Results are presented which show the uncertainty on k-effective for these systems arising from the uncertainty on the input nuclear data.

Thermal reactor benchmark tests on JENDL-2

International Nuclear Information System (INIS)

Takano, Hideki; Tsuchihashi, Keichiro; Yamane, Tsuyoshi; Akino, Fujiyoshi; Ishiguro, Yukio; Ido, Masaru.

1983-11-01

A group constant library for the thermal reactor standard nuclear design code system SRAC was produced by using the evaluated nuclear data JENDL-2. Furthermore, the group constants for 235 U were calculated also from ENDF/B-V. Thermal reactor benchmark calculations were performed using the produced group constant library. The selected benchmark cores are two water-moderated lattices (TRX-1 and 2), two heavy water-moderated cores (DCA and ETA-1), two graphite-moderated cores (SHE-8 and 13) and eight critical experiments for critical safety. The effective multiplication factors and lattice cell parameters were calculated and compared with the experimental values. The results are summarized as follows. (1) Effective multiplication factors: The results by JENDL-2 are considerably improved in comparison with ones by ENDF/B-IV. The best agreement is obtained by using JENDL-2 and ENDF/B-V (only 235 U) data. (2) Lattice cell parameters: For the rho 28 (the ratio of epithermal to thermal 238 U captures) and C* (the ratio of 238 U captures to 235 U fissions), the values calculated by JENDL-2 are in good agreement with the experimental values. The rho 28 (the ratio of 238 U to 235 U fissions) are overestimated as found also for the fast reactor benchmarks. The rho 02 (the ratio of epithermal to thermal 232 Th captures) calculated by JENDL-2 or ENDF/B-IV are considerably underestimated. The functions of the SRAC system have been continued to be extended according to the needs of its users. A brief description will be given, in Appendix B, to the extended parts of the SRAC system together with the input specification. (author)

Isopiestic density law of actinide nitrates applied to criticality calculations

International Nuclear Information System (INIS)

Leclaire, Nicolas; Anno, Jacques; Courtois, Gerard; Poullot, Gilles; Rouyer, Veronique

2003-01-01

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: UO 2 (NO 3 ) 2 , U(NO 3 ) 4 , Pu(NO 3 ) 4 , Pu(NO 3 ) 3 , Th(NO 3 ) 4 , Am(NO 3 ) 3 and HNO 3 . 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)

Testing of cross section libraries for TRIGA criticality benchmark

International Nuclear Information System (INIS)

Snoj, L.; Trkov, A.; Ravnik, M.

2007-01-01

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 (∼ 2 20 pcm) are from 235 U and Zr. (author)

ZZ ECN-BUBEBO, ECN-Petten Burnup Benchmark Book, Inventories, Afterheat

International Nuclear Information System (INIS)

Kloosterman, Jan Leen

1999-01-01

Description of program or function: Contains experimental benchmarks which can be used for the validation of burnup code systems and accompanied data libraries. Although the benchmarks presented here are thoroughly described in literature, it is in many cases not straightforward to retrieve unambiguously the correct input data and corresponding results from the benchmark Descriptions. Furthermore, results which can easily be measured, are sometimes difficult to calculate because of conversions to be made. Therefore, emphasis has been put to clarify the input of the benchmarks and to present the benchmark results in such a way that they can easily be calculated and compared. For more thorough Descriptions of the benchmarks themselves, the literature referred to here should be consulted. This benchmark book is divided in 11 chapters/files containing the following in text and tabular form: chapter 1: Introduction; chapter 2: Burnup Credit Criticality Benchmark Phase 1-B; chapter 3: Yankee-Rowe Core V Fuel Inventory Study; chapter 4: H.B. Robinson Unit 2 Fuel Inventory Study; chapter 5: Turkey Point Unit 3 Fuel Inventory Study; chapter 6: Turkey Point Unit 3 Afterheat Power Study; chapter 7: Dickens Benchmark on Fission Product Energy Release of U-235; chapter 8: Dickens Benchmark on Fission Product Energy Release of Pu-239; chapter 9: Yarnell Benchmark on Decay Heat Measurements of U-233; chapter 10: Yarnell Benchmark on Decay Heat Measurements of U-235; chapter 11: Yarnell Benchmark on Decay Heat Measurements of Pu-239

Benchmark calculation for GT-MHR using HELIOS/MASTER code package and MCNP

International Nuclear Information System (INIS)

Lee, Kyung Hoon; Kim, Kang Seog; Noh, Jae Man; Song, Jae Seung; Zee, Sung Quun

2005-01-01

The latest research associated with the very high temperature gas-cooled reactor (VHTR) is focused on the verification of a system performance and safety under operating conditions for the VHTRs. As a part of those, an international gas-cooled reactor program initiated by IAEA is going on. The key objectives of this program are the validation of analytical computer codes and the evaluation of benchmark models for the projected and actual VHTRs. New reactor physics analysis procedure for the prismatic VHTR is under development by adopting the conventional two-step procedure. In this procedure, a few group constants are generated through the transport lattice calculations using the HELIOS code, and the core physics analysis is performed by the 3-dimensional nodal diffusion code MASTER. We evaluated the performance of the HELIOS/MASTER code package through the benchmark calculations related to the GT-MHR (Gas Turbine-Modular Helium Reactor) to dispose weapon plutonium. In parallel, MCNP is employed as a reference code to verify the results of the HELIOS/MASTER procedure

Final results of the fifth three-dimensional dynamic Atomic Energy Research benchmark problem calculations

International Nuclear Information System (INIS)

Hadek, J.

1999-01-01

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)

Recent R and D around the Monte-Carlo code Tripoli-4 for criticality calculation

International Nuclear Information System (INIS)

Hugot, F.X.; Lee, Y.K.; Malvagi, F.

2008-01-01

TRIPOLI-4 [1] is the fourth generation of the TRIPOLI family of Monte Carlo codes developed from the 60's by CEA. It simulates the 3D transport of neutrons, photons, electrons and positrons as well as coupled neutron-photon propagation and electron-photons cascade showers. The code addresses radiation protection and shielding problems, as well as criticality and reactor physics problems through both critical and subcritical neutronics calculations. It uses full pointwise as well as multigroup cross-sections. The code has been validated through several hundred benchmarks as well as measurement campaigns. It is used as a reference tool by CEA as well as its industrial and institutional partners, and in the NURESIM [2] European project. Section 2 reviews its main features, with emphasis on the latest developments. Section 3 presents some recent R and D for criticality calculations. Fission matrix, Eigen-values and eigenvectors computations will be exposed. Corrections on the standard deviation estimator in the case of correlations between generation steps will be detailed. Section 4 presents some preliminary results obtained by the new mesh tally feature. The last section presents the interest of using XML format output files. (authors)

NUPEC BWR Full-size Fine-mesh Bundle Test (BFBT) Benchmark. Volume II: uncertainty and sensitivity analyses of void distribution and critical power - Specification

International Nuclear Information System (INIS)

Aydogan, F.; Hochreiter, L.; Ivanov, K.; Martin, M.; Utsuno, H.; Sartori, E.

2010-01-01

This report provides the specification for the uncertainty exercises of the international OECD/NEA, NRC and NUPEC BFBT benchmark problem including the elemental task. The specification was prepared jointly by Pennsylvania State University (PSU), USA and the Japan Nuclear Energy Safety (JNES) Organisation, in cooperation with the OECD/NEA and the Commissariat a l'energie atomique (CEA Saclay, France). The work is sponsored by the US NRC, METI-Japan, the OECD/NEA and the Nuclear Engineering Program (NEP) of Pennsylvania State University. This uncertainty specification covers the fourth exercise of Phase I (Exercise-I-4), and the third exercise of Phase II (Exercise II-3) as well as the elemental task. The OECD/NRC BFBT benchmark provides a very good opportunity to apply uncertainty analysis (UA) and sensitivity analysis (SA) techniques and to assess the accuracy of thermal-hydraulic models for two-phase flows in rod bundles. During the previous OECD benchmarks, participants usually carried out sensitivity analysis on their models for the specification (initial conditions, boundary conditions, etc.) to identify the most sensitive models or/and to improve the computed results. The comprehensive BFBT experimental database (NEA, 2006) leads us one step further in investigating modelling capabilities by taking into account the uncertainty analysis in the benchmark. The uncertainties in input data (boundary conditions) and geometry (provided in the benchmark specification) as well as the uncertainties in code models can be accounted for to produce results with calculational uncertainties and compare them with the measurement uncertainties. Therefore, uncertainty analysis exercises were defined for the void distribution and critical power phases of the BFBT benchmark. This specification is intended to provide definitions related to UA/SA methods, sensitivity/ uncertainty parameters, suggested probability distribution functions (PDF) of sensitivity parameters, and selected

Calculation of the Thermal Radiation Benchmark Problems for a CANDU Fuel Channel Analysis Using the CFX-10 Code

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyoung Tae; Park, Joo Hwan; Rhee, Bo Wook

2006-07-15

To justify the use of a commercial Computational Fluid Dynamics (CFD) code for a CANDU fuel channel analysis, especially for the radiation heat transfer dominant conditions, the CFX-10 code is tested against three benchmark problems which were used for the validation of a radiation heat transfer in the CANDU analysis code, a CATHENA. These three benchmark problems are representative of the CANDU fuel channel configurations from a simple geometry to whole fuel channel geometry. With assumptions of a non-participating medium completely enclosed with the diffuse, gray and opaque surfaces, the solutions of the benchmark problems are obtained by the concept of surface resistance to radiation accounting for the view factors and the emissivities. The view factors are calculated by the program MATRIX version 1.0 avoiding the difficulty of hand calculation for the complex geometries. For the solutions of the benchmark problems, the temperature or the net radiation heat flux boundary conditions are prescribed for each radiating surface to determine the radiation heat transfer rate or the surface temperature, respectively by using the network method. The Discrete Transfer Model (DTM) is used for the CFX-10 radiation model and its calculation results are compared with the solutions of the benchmark problems. The CFX-10 results for the three benchmark problems are in close agreement with these solutions, so it is concluded that the CFX-10 with a DTM radiation model can be applied to the CANDU fuel channel analysis where a surface radiation heat transfer is a dominant mode of the heat transfer.

Calculation of the Thermal Radiation Benchmark Problems for a CANDU Fuel Channel Analysis Using the CFX-10 Code

International Nuclear Information System (INIS)

Kim, Hyoung Tae; Park, Joo Hwan; Rhee, Bo Wook

2006-07-01

To justify the use of a commercial Computational Fluid Dynamics (CFD) code for a CANDU fuel channel analysis, especially for the radiation heat transfer dominant conditions, the CFX-10 code is tested against three benchmark problems which were used for the validation of a radiation heat transfer in the CANDU analysis code, a CATHENA. These three benchmark problems are representative of the CANDU fuel channel configurations from a simple geometry to whole fuel channel geometry. With assumptions of a non-participating medium completely enclosed with the diffuse, gray and opaque surfaces, the solutions of the benchmark problems are obtained by the concept of surface resistance to radiation accounting for the view factors and the emissivities. The view factors are calculated by the program MATRIX version 1.0 avoiding the difficulty of hand calculation for the complex geometries. For the solutions of the benchmark problems, the temperature or the net radiation heat flux boundary conditions are prescribed for each radiating surface to determine the radiation heat transfer rate or the surface temperature, respectively by using the network method. The Discrete Transfer Model (DTM) is used for the CFX-10 radiation model and its calculation results are compared with the solutions of the benchmark problems. The CFX-10 results for the three benchmark problems are in close agreement with these solutions, so it is concluded that the CFX-10 with a DTM radiation model can be applied to the CANDU fuel channel analysis where a surface radiation heat transfer is a dominant mode of the heat transfer

Depletion benchmarks calculation of random media using explicit modeling approach of RMC

International Nuclear Information System (INIS)

Liu, Shichang; She, Ding; Liang, Jin-gang; Wang, Kan

2016-01-01

Highlights: • Explicit modeling of RMC is applied to depletion benchmark for HTGR fuel element. • Explicit modeling can provide detailed burnup distribution and burnup heterogeneity. • The results would serve as a supplement for the HTGR fuel depletion benchmark. • The method of adjacent burnup regions combination is proposed for full-core problems. • The combination method can reduce memory footprint, keeping the computing accuracy. - Abstract: Monte Carlo method plays an important role in accurate simulation of random media, owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. Three stochastic geometry modeling methods including Random Lattice Method, Chord Length Sampling and explicit modeling approach with mesh acceleration technique, have been implemented in RMC to simulate the particle transport in the dispersed fuels, in which the explicit modeling method is regarded as the best choice. In this paper, the explicit modeling method is applied to the depletion benchmark for HTGR fuel element, and the method of combination of adjacent burnup regions has been proposed and investigated. The results show that the explicit modeling can provide detailed burnup distribution of individual TRISO particles, and this work would serve as a supplement for the HTGR fuel depletion benchmark calculations. The combination of adjacent burnup regions can effectively reduce the memory footprint while keeping the computational accuracy.

Critical Assessment of Metagenome Interpretation – a benchmark of computational metagenomics software

Science.gov (United States)

Sczyrba, Alexander; Hofmann, Peter; Belmann, Peter; Koslicki, David; Janssen, Stefan; Dröge, Johannes; Gregor, Ivan; Majda, Stephan; Fiedler, Jessika; Dahms, Eik; Bremges, Andreas; Fritz, Adrian; Garrido-Oter, Ruben; Jørgensen, Tue Sparholt; Shapiro, Nicole; Blood, Philip D.; Gurevich, Alexey; Bai, Yang; Turaev, Dmitrij; DeMaere, Matthew Z.; Chikhi, Rayan; Nagarajan, Niranjan; Quince, Christopher; Meyer, Fernando; Balvočiūtė, Monika; Hansen, Lars Hestbjerg; Sørensen, Søren J.; Chia, Burton K. H.; Denis, Bertrand; Froula, Jeff L.; Wang, Zhong; Egan, Robert; Kang, Dongwan Don; Cook, Jeffrey J.; Deltel, Charles; Beckstette, Michael; Lemaitre, Claire; Peterlongo, Pierre; Rizk, Guillaume; Lavenier, Dominique; Wu, Yu-Wei; Singer, Steven W.; Jain, Chirag; Strous, Marc; Klingenberg, Heiner; Meinicke, Peter; Barton, Michael; Lingner, Thomas; Lin, Hsin-Hung; Liao, Yu-Chieh; Silva, Genivaldo Gueiros Z.; Cuevas, Daniel A.; Edwards, Robert A.; Saha, Surya; Piro, Vitor C.; Renard, Bernhard Y.; Pop, Mihai; Klenk, Hans-Peter; Göker, Markus; Kyrpides, Nikos C.; Woyke, Tanja; Vorholt, Julia A.; Schulze-Lefert, Paul; Rubin, Edward M.; Darling, Aaron E.; Rattei, Thomas; McHardy, Alice C.

2018-01-01

In metagenome analysis, computational methods for assembly, taxonomic profiling and binning are key components facilitating downstream biological data interpretation. However, a lack of consensus about benchmarking datasets and evaluation metrics complicates proper performance assessment. The Critical Assessment of Metagenome Interpretation (CAMI) challenge has engaged the global developer community to benchmark their programs on datasets of unprecedented complexity and realism. Benchmark metagenomes were generated from ~700 newly sequenced microorganisms and ~600 novel viruses and plasmids, including genomes with varying degrees of relatedness to each other and to publicly available ones and representing common experimental setups. Across all datasets, assembly and genome binning programs performed well for species represented by individual genomes, while performance was substantially affected by the presence of related strains. Taxonomic profiling and binning programs were proficient at high taxonomic ranks, with a notable performance decrease below the family level. Parameter settings substantially impacted performances, underscoring the importance of program reproducibility. While highlighting current challenges in computational metagenomics, the CAMI results provide a roadmap for software selection to answer specific research questions. PMID:28967888

Regional Competitive Intelligence: Benchmarking and Policymaking

OpenAIRE

Huggins , Robert

2010-01-01

Benchmarking exercises have become increasingly popular within the sphere of regional policymaking in recent years. The aim of this paper is to analyse the concept of regional benchmarking and its links with regional policymaking processes. It develops a typology of regional benchmarking exercises and regional benchmarkers, and critically reviews the literature, both academic and policy oriented. It is argued that critics who suggest regional benchmarking is a flawed concept and technique fai...

VENUS-2 MOX Core Benchmark: Results of ORNL Calculations Using HELIOS-1.4 - Revised Report

Energy Technology Data Exchange (ETDEWEB)

Ellis, RJ

2001-06-01

The Task Force on Reactor-Based Plutonium Disposition (TFRPD) was formed by the Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA) to study reactor physics, fuel performance, and fuel cycle issues related to the disposition of weapons-grade (WG) plutonium as mixed-oxide (MOX) reactor fuel. To advance the goals of the TFRPD, 10 countries and 12 institutions participated in a major TFRPD activity: a blind benchmark study to compare code calculations to experimental data for the VENUS-2 MOX core at SCK-CEN in Mol, Belgium. At Oak Ridge National Laboratory, the HELIOS-1.4 code system was used to perform the comprehensive study of pin-cell and MOX core calculations for the VENUS-2 MOX core benchmark study.

Growth and Expansion of the International Criticality Safety Benchmark Evaluation Project and the Newly Organized International Reactor Physics Experiment Evaluation Project

International Nuclear Information System (INIS)

J. Blair Briggs; Lori Scott; Yolanda Rugama; Enrico Satori

2007-01-01

Since ICNC 2003, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) has continued to expand its efforts and broaden its scope. Criticality-alarm/shielding type benchmarks and fundamental physics measurements that are relevant to criticality safety applications are not only included in the scope of the project, but benchmark data are also included in the latest version of the handbook. A considerable number of improvements have been made to the searchable database, DICE and the criticality-alarm/shielding benchmarks and fundamental physics measurements have been included in the database. There were 12 countries participating on the ICSBEP in 2003. That number has increased to 18 with recent contributions of data and/or resources from Brazil, Czech Republic, Poland, India, Canada, and China. South Africa, Germany, Argentina, and Australia have been invited to participate. Since ICNC 2003, the contents of the ''International Handbook of Evaluated Criticality Safety Benchmark Experiments'' have increased from 350 evaluations (28,000 pages) containing benchmark specifications for 3070 critical or subcritical configurations to 442 evaluations (over 38,000 pages) containing benchmark specifications for 3957 critical or subcritical configurations, 23 criticality-alarm-placement/shielding configurations with multiple dose points for each, and 20 configurations that have been categorized as fundamental physics measurements that are relevant to criticality safety applications in the 2006 Edition of the ICSBEP Handbook. Approximately 30 new evaluations and 250 additional configurations are expected to be added to the 2007 Edition of the Handbook. Since ICNC 2003, a reactor physics counterpart to the ICSBEP, The International Reactor Physics Experiment Evaluation Project (IRPhEP) was initiated. Beginning in 1999, the IRPhEP was conducted as a pilot activity by the by the Organization of Economic Cooperation and Development (OECD) Nuclear Energy Agency

RB reactor as the U-D2O benchmark criticality system

International Nuclear Information System (INIS)

Pesic, M.

1998-01-01

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)

Analyses and results of the OECD/NEA WPNCS EGUNF benchmark phase II. Technical report; Analysen und Ergebnisse zum OECD/NEA WPNCS EGUNF Benchmark Phase II. Technischer Bericht

Energy Technology Data Exchange (ETDEWEB)

Hannstein, Volker; Sommer, Fabian

2017-05-15

The report summarizes the performed studies and results in the frame of the phase II benchmarks of the expert group of used nuclear fuel (EGUNF) of the working party of nuclear criticality safety (WPNCS) of the nuclear energy agency (NEA) of the organization for economic co-operation and development (OECD). The studies specified within the benchmarks have been realized to the full extent. The scope of the benchmarks was the comparison of a generic BWR fuel element with gadolinium containing fuel rods with several computer codes and cross section libraries of different international working groups and institutions. The used computational model allows the evaluation of the accuracy of fuel rod and their influence of the inventory calculations and the respective influence on BWR burnout credit calculations.

Analysis of the ITER computational shielding benchmark with the Monte Carlo TRIPOLI-4® neutron gamma coupled calculations

International Nuclear Information System (INIS)

Lee, Yi-Kang

2016-01-01

Highlights: • Verification and validation of TRIPOLI-4 radiation transport calculations for ITER shielding benchmark. • Evaluation of CEA-V5.1.1 and FENDL-3.0 nuclear data libraries on D–T fusion neutron continuous energy transport calculations. • Advances in nuclear analyses for nuclear heating and radiation damage in iron. • This work also demonstrates that the “safety factors” concept is necessary in the nuclear analyses of ITER. - Abstract: With the growing interest in using the continuous-energy TRIPOLI-4 ® Monte Carlo radiation transport code for ITER applications, a key issue that arises is whether or not the released TRIPOLI-4 code and its associated nuclear data libraries are verified and validated for the D–T fusion neutronics calculations. Previous published benchmark results of TRIPOLI-4 code on the ITER related activities have concentrated on the first wall loading, the reactor dosimetry, the nuclear heating, and the tritium breeding ratio. To enhance the TRIPOLI-4 verification and validation on neutron-gamma coupled calculations for fusion device application, the computational ITER shielding benchmark of M. E. Sawan was performed in this work by using the 2013 released TRIPOLI-4.9S code and the associated CEA-V5.1.1 data library. First wall, blanket, vacuum vessel and toroidal field magnet of the inboard and outboard components were fully modelled in this 1-D toroidal cylindrical benchmark. The 14.1 MeV source neutrons were sampled from a uniform isotropic distribution in the plasma zone. Nuclear responses including neutron and gamma fluxes, nuclear heating, and material damage indicator were benchmarked against previous published results. The capabilities of the TRIPOLI-4 code on the evaluation of above physics parameters were presented. The nuclear data library from the new FENDL-3.0 evaluation was also benchmarked against the CEA-V5.1.1 results for the neutron transport calculations. The results show that both data libraries can be

Validation of the WIMSD4M cross-section generation code with benchmark results

International Nuclear Information System (INIS)

Deen, J.R.; Woodruff, W.L.; Leal, L.E.

1995-01-01

The WIMSD4 code has been adopted for cross-section generation in support of the Reduced Enrichment Research and Test Reactor (RERTR) program at Argonne National Laboratory (ANL). Subsequently, the code has undergone several updates, and significant improvements have been achieved. The capability of generating group-collapsed micro- or macroscopic cross sections from the ENDF/B-V library and the more recent evaluation, ENDF/B-VI, in the ISOTXS format makes the modified version of the WIMSD4 code, WIMSD4M, very attractive, not only for the RERTR program, but also for the reactor physics community. The intent of the present paper is to validate the WIMSD4M cross-section libraries for reactor modeling of fresh water moderated cores. The results of calculations performed with multigroup cross-section data generated with the WIMSD4M code will be compared against experimental results. These results correspond to calculations carried out with thermal reactor benchmarks of the Oak Ridge National Laboratory (ORNL) unreflected HEU critical spheres, the TRX LEU critical experiments, and calculations of a modified Los Alamos HEU D 2 O moderated benchmark critical system. The benchmark calculations were performed with the discrete-ordinates transport code, TWODANT, using WIMSD4M cross-section data. Transport calculations using the XSDRNPM module of the SCALE code system are also included. In addition to transport calculations, diffusion calculations with the DIF3D code were also carried out, since the DIF3D code is used in the RERTR program for reactor analysis and design. For completeness, Monte Carlo results of calculations performed with the VIM and MCNP codes are also presented

Validation of the WIMSD4M cross-section generation code with benchmark results

Energy Technology Data Exchange (ETDEWEB)

Deen, J.R.; Woodruff, W.L. [Argonne National Lab., IL (United States); Leal, L.E. [Oak Ridge National Lab., TN (United States)

1995-01-01

The WIMSD4 code has been adopted for cross-section generation in support of the Reduced Enrichment Research and Test Reactor (RERTR) program at Argonne National Laboratory (ANL). Subsequently, the code has undergone several updates, and significant improvements have been achieved. The capability of generating group-collapsed micro- or macroscopic cross sections from the ENDF/B-V library and the more recent evaluation, ENDF/B-VI, in the ISOTXS format makes the modified version of the WIMSD4 code, WIMSD4M, very attractive, not only for the RERTR program, but also for the reactor physics community. The intent of the present paper is to validate the WIMSD4M cross-section libraries for reactor modeling of fresh water moderated cores. The results of calculations performed with multigroup cross-section data generated with the WIMSD4M code will be compared against experimental results. These results correspond to calculations carried out with thermal reactor benchmarks of the Oak Ridge National Laboratory (ORNL) unreflected HEU critical spheres, the TRX LEU critical experiments, and calculations of a modified Los Alamos HEU D{sub 2}O moderated benchmark critical system. The benchmark calculations were performed with the discrete-ordinates transport code, TWODANT, using WIMSD4M cross-section data. Transport calculations using the XSDRNPM module of the SCALE code system are also included. In addition to transport calculations, diffusion calculations with the DIF3D code were also carried out, since the DIF3D code is used in the RERTR program for reactor analysis and design. For completeness, Monte Carlo results of calculations performed with the VIM and MCNP codes are also presented.

The calculational VVER burnup Credit Benchmark No.3 results with the ENDF/B-VI rev.5 (1999)

Energy Technology Data Exchange (ETDEWEB)

Rodriguez Gual, Maritza [Centro de Tecnologia Nuclear, La Habana (Cuba). E-mail: mrgual@ctn.isctn.edu.cu

2000-07-01

The purpose of this papers to present the results of CB3 phase of the VVER calculational benchmark with the recent evaluated nuclear data library ENDF/B-VI Rev.5 (1999). This results are compared with the obtained from the other participants in the calculations (Czech Republic, Finland, Hungary, Slovaquia, Spain and the United Kingdom). The phase (CB3) of the VVER calculation benchmark is similar to the Phase II-A of the OECD/NEA/INSC BUC Working Group benchmark for PWR. The cases without burnup profile (BP) were performed with the WIMS/D-4 code. The rest of the cases have been carried with DOTIII discrete ordinates code. The neutron library used was the ENDF/B-VI rev. 5 (1999). The WIMS/D-4 (69 groups) is used to collapse cross sections from the ENDF/B-VI Rev. 5 (1999) to 36 groups working library for 2-D calculations. This work also comprises the results of CB1 (obtained with ENDF/B-VI rev. 5 (1999), too) and CB3 for cases with Burnup of 30 MWd/TU and cooling time of 1 and 5 years and for case with Burnup of 40 MWd/TU and cooling time of 1 year. (author)

Development and benchmark verification of a parallelized Monte Carlo burnup calculation program MCBMPI

International Nuclear Information System (INIS)

Yang Wankui; Liu Yaoguang; Ma Jimin; Yang Xin; Wang Guanbo

2014-01-01

MCBMPI, a parallelized burnup calculation program, was developed. The program is modularized. Neutron transport calculation module employs the parallelized MCNP5 program MCNP5MPI, and burnup calculation module employs ORIGEN2, with the MPI parallel zone decomposition strategy. The program system only consists of MCNP5MPI and an interface subroutine. The interface subroutine achieves three main functions, i.e. zone decomposition, nuclide transferring and decaying, data exchanging with MCNP5MPI. Also, the program was verified with the Pressurized Water Reactor (PWR) cell burnup benchmark, the results showed that it's capable to apply the program to burnup calculation of multiple zones, and the computation efficiency could be significantly improved with the development of computer hardware. (authors)

a Proposed Benchmark Problem for Scatter Calculations in Radiographic Modelling

Science.gov (United States)

Jaenisch, G.-R.; Bellon, C.; Schumm, A.; Tabary, J.; Duvauchelle, Ph.

2009-03-01

Code Validation is a permanent concern in computer modelling, and has been addressed repeatedly in eddy current and ultrasonic modeling. A good benchmark problem is sufficiently simple to be taken into account by various codes without strong requirements on geometry representation capabilities, focuses on few or even a single aspect of the problem at hand to facilitate interpretation and to avoid that compound errors compensate themselves, yields a quantitative result and is experimentally accessible. In this paper we attempt to address code validation for one aspect of radiographic modeling, the scattered radiation prediction. Many NDT applications can not neglect scattered radiation, and the scatter calculation thus is important to faithfully simulate the inspection situation. Our benchmark problem covers the wall thickness range of 10 to 50 mm for single wall inspections, with energies ranging from 100 to 500 keV in the first stage, and up to 1 MeV with wall thicknesses up to 70 mm in the extended stage. A simple plate geometry is sufficient for this purpose, and the scatter data is compared on a photon level, without a film model, which allows for comparisons with reference codes like MCNP. We compare results of three Monte Carlo codes (McRay, Sindbad and Moderato) as well as an analytical first order scattering code (VXI), and confront them to results obtained with MCNP. The comparison with an analytical scatter model provides insights into the application domain where this kind of approach can successfully replace Monte-Carlo calculations.

Growth and Expansion of the International Criticality Safety Benchmark Evaluation Project and the Newly Organized International Reactor Physics Experiment Evaluation Project

Energy Technology Data Exchange (ETDEWEB)

J. Blair Briggs; Lori Scott; Yolanda Rugama; Enrico Satori

2007-05-01

Since ICNC 2003, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) has continued to expand its efforts and broaden its scope. Criticality-alarm / shielding type benchmarks and fundamental physics measurements that are relevant to criticality safety applications are not only included in the scope of the project, but benchmark data are also included in the latest version of the handbook. A considerable number of improvements have been made to the searchable database, DICE and the criticality-alarm / shielding benchmarks and fundamental physics measurements have been included in the database. There were 12 countries participating on the ICSBEP in 2003. That number has increased to 18 with recent contributions of data and/or resources from Brazil, Czech Republic, Poland, India, Canada, and China. South Africa, Germany, Argentina, and Australia have been invited to participate. Since ICNC 2003, the contents of the “International Handbook of Evaluated Criticality Safety Benchmark Experiments” have increased from 350 evaluations (28,000 pages) containing benchmark specifications for 3070 critical or subcritical configurations to 442 evaluations (over 38,000 pages) containing benchmark specifications for 3957 critical or subcritical configurations, 23 criticality-alarm-placement / shielding configurations with multiple dose points for each, and 20 configurations that have been categorized as fundamental physics measurements that are relevant to criticality safety applications in the 2006 Edition of the ICSBEP Handbook. Approximately 30 new evaluations and 250 additional configurations are expected to be added to the 2007 Edition of the Handbook. Since ICNC 2003, a reactor physics counterpart to the ICSBEP, The International Reactor Physics Experiment Evaluation Project (IRPhEP) was initiated. Beginning in 1999, the IRPhEP was conducted as a pilot activity by the by the Organization of Economic Cooperation and Development (OECD) Nuclear Energy

Thermal lattice benchmarks for testing basic evaluated data files, developed with MCNP4B

International Nuclear Information System (INIS)

Maucec, M.; Glumac, B.

1996-01-01

The development of unit cell and full reactor core models of DIMPLE S01A and TRX-1 and TRX-2 benchmark experiments, using Monte Carlo computer code MCNP4B is presented. Nuclear data from ENDF/B-V and VI version of cross-section library were used in the calculations. In addition, a comparison to results obtained with the similar models and cross-section data from the EJ2-MCNPlib library (which is based upon the JEF-2.2 evaluation) developed in IRC Petten, Netherlands is presented. The results of the criticality calculation with ENDF/B-VI data library, and a comparison to results obtained using JEF-2.2 evaluation, confirm the MCNP4B full core model of a DIMPLE reactor as a good benchmark for testing basic evaluated data files. On the other hand, the criticality calculations results obtained using the TRX full core models show less agreement with experiment. It is obvious that without additional data about the TRX geometry, our TRX models are not suitable as Monte Carlo benchmarks. (author)

Calculations to an IAHR-benchmark test using the CFD-code CFX-4

Energy Technology Data Exchange (ETDEWEB)

Krepper, E

1998-10-01

The calculation concerns a test, which was defined as a benchmark for 3-D codes by the working group of advanced nuclear reactor types of IAHR (International Association of Hydraulic Research). The test is well documented and detailed measuring results are available. The test aims at the investigation of phenomena, which are important for heat removal at natural circulation conditions in a nuclear reactor. The task for the calculation was the modelling of the forced flow field of a single phase incompressible fluid with consideration of heat transfer and influence of gravity. These phenomena are typical also for other industrial processes. The importance of correct modelling of these phenomena also for other applications is a motivation for performing these calculations. (orig.)

DRY TRANSFER FACILITY CRITICALITY SAFETY CALCULATIONS

International Nuclear Information System (INIS)

C.E. Sanders

2005-01-01

This design calculation updates the previous criticality evaluation for the fuel handling, transfer, and staging operations to be performed in the Dry Transfer Facility (DTF) including the remediation area. The purpose of the calculation is to demonstrate that operations performed in the DTF and RF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Dry Transfer Facility Description Document'' (BSC 2005 [DIRS 173737], p. 3-8). A description of the changes is as follows: (1) Update the supporting calculations for the various Category 1 and 2 event sequences as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2005 [DIRS 171429], Section 7). (2) Update the criticality safety calculations for the DTF staging racks and the remediation pool to reflect the current design. This design calculation focuses on commercial spent nuclear fuel (SNF) assemblies, i.e., pressurized water reactor (PWR) and boiling water reactor (BWR) SNF. U.S. Department of Energy (DOE) Environmental Management (EM) owned SNF is evaluated in depth in the ''Canister Handling Facility Criticality Safety Calculations'' (BSC 2005 [DIRS 173284]) and is also applicable to DTF operations. Further, the design and safety analyses of the naval SNF canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. Also, note that the results for the Monitored Geologic Repository (MGR) Site specific Cask (MSC) calculations are limited to the

Development of common user data model for APOLLO3 and MARBLE and application to benchmark problems

International Nuclear Information System (INIS)

Yokoyama, Kenji

2009-07-01

A Common User Data Model, CUDM, has been developed for the purpose of benchmark calculations between APOLLO3 and MARBLE code systems. The current version of CUDM was designed for core calculation benchmark problems with 3-dimensional Cartesian, 3-D XYZ, geometry. CUDM is able to manage all input/output data such as 3-D XYZ geometry, effective macroscopic cross section, effective multiplication factor and neutron flux. In addition, visualization tools for geometry and neutron flux were included. CUDM was designed by the object-oriented technique and implemented using Python programming language. Based on the CUDM, a prototype system for a benchmark calculation, CUDM-benchmark, was also developed. The CUDM-benchmark supports input/output data conversion for IDT solver in APOLLO3, and TRITAC and SNT solvers in MARBLE. In order to evaluate pertinence of CUDM, the CUDM-benchmark was applied to benchmark problems proposed by T. Takeda, G. Chiba and I. Zmijarevic. It was verified that the CUDM-benchmark successfully reproduced the results calculated with reference input data files, and provided consistent results among all the solvers by using one common input data defined by CUDM. In addition, a detailed benchmark calculation for Chiba benchmark was performed by using the CUDM-benchmark. Chiba benchmark is a neutron transport benchmark problem for fast criticality assembly without homogenization. This benchmark problem consists of 4 core configurations which have different sodium void regions, and each core configuration is defined by more than 5,000 fuel/material cells. In this application, it was found that the results by IDT and SNT solvers agreed well with the reference results by Monte-Carlo code. In addition, model effects such as quadrature set effect, S n order effect and mesh size effect were systematically evaluated and summarized in this report. (author)

Static benchmarking of the NESTLE advanced nodal code

International Nuclear Information System (INIS)

Mosteller, R.D.

1997-01-01

Results from the NESTLE advanced nodal code are presented for multidimensional numerical benchmarks representing four different types of reactors, and predictions from NESTLE are compared with measured data from pressurized water reactors (PWRs). The numerical benchmarks include cases representative of PWRs, boiling water reactors (BWRs), CANDU heavy water reactors (HWRs), and high-temperature gas-cooled reactors (HTGRs). The measured PWR data include critical soluble boron concentrations and isothermal temperature coefficients of reactivity. The results demonstrate that NESTLE correctly solves the multigroup diffusion equations for both Cartesian and hexagonal geometries, that it reliably calculates k eff and reactivity coefficients for PWRs, and that--subsequent to the incorporation of additional thermal-hydraulic models--it will be able to perform accurate calculations for the corresponding parameters in BWRs, HWRs, and HTGRs as well

A benchmark test of computer codes for calculating average resonance parameters

International Nuclear Information System (INIS)

Ribon, P.; Thompson, A.

1983-01-01

A set of resonance parameters has been generated from known, but secret, average values; the parameters have then been adjusted to mimic experimental data by including the effects of Doppler broadening, resolution broadening and statistical fluctuations. Average parameters calculated from the dataset by various computer codes are compared with each other, and also with the true values. The benchmark test is fully described in the report NEANDC160-U (NEA Data Bank Newsletter No. 27 July 1982); the present paper is a summary of this document. (Auth.)

Analysis of the ITER computational shielding benchmark with the Monte Carlo TRIPOLI-4{sup ®} neutron gamma coupled calculations

Energy Technology Data Exchange (ETDEWEB)

Lee, Yi-Kang, E-mail: yi-kang.lee@cea.fr

2016-11-01

Highlights: • Verification and validation of TRIPOLI-4 radiation transport calculations for ITER shielding benchmark. • Evaluation of CEA-V5.1.1 and FENDL-3.0 nuclear data libraries on D–T fusion neutron continuous energy transport calculations. • Advances in nuclear analyses for nuclear heating and radiation damage in iron. • This work also demonstrates that the “safety factors” concept is necessary in the nuclear analyses of ITER. - Abstract: With the growing interest in using the continuous-energy TRIPOLI-4{sup ®} Monte Carlo radiation transport code for ITER applications, a key issue that arises is whether or not the released TRIPOLI-4 code and its associated nuclear data libraries are verified and validated for the D–T fusion neutronics calculations. Previous published benchmark results of TRIPOLI-4 code on the ITER related activities have concentrated on the first wall loading, the reactor dosimetry, the nuclear heating, and the tritium breeding ratio. To enhance the TRIPOLI-4 verification and validation on neutron-gamma coupled calculations for fusion device application, the computational ITER shielding benchmark of M. E. Sawan was performed in this work by using the 2013 released TRIPOLI-4.9S code and the associated CEA-V5.1.1 data library. First wall, blanket, vacuum vessel and toroidal field magnet of the inboard and outboard components were fully modelled in this 1-D toroidal cylindrical benchmark. The 14.1 MeV source neutrons were sampled from a uniform isotropic distribution in the plasma zone. Nuclear responses including neutron and gamma fluxes, nuclear heating, and material damage indicator were benchmarked against previous published results. The capabilities of the TRIPOLI-4 code on the evaluation of above physics parameters were presented. The nuclear data library from the new FENDL-3.0 evaluation was also benchmarked against the CEA-V5.1.1 results for the neutron transport calculations. The results show that both data libraries

Two-group k-eigenvalue benchmark calculations for planar geometry transport in a binary stochastic medium

International Nuclear Information System (INIS)

Davis, I.M.; Palmer, T.S.

2005-01-01

Benchmark calculations are performed for neutron transport in a two material (binary) stochastic multiplying medium. Spatial, angular, and energy dependence are included. The problem considered is based on a fuel assembly of a common pressurized water reactor. The mean chord length through the assembly is determined and used as the planar geometry system length. According to assumed or calculated material distributions, this system length is populated with alternating fuel and moderator segments of random size. Neutron flux distributions are numerically computed using a discretized form of the Boltzmann transport equation employing diffusion synthetic acceleration. Average quantities (group fluxes and k-eigenvalue) and variances are calculated from an ensemble of realizations of the mixing statistics. The effects of varying two parameters in the fuel, two different boundary conditions, and three different sets of mixing statistics are assessed. A probability distribution function (PDF) of the k-eigenvalue is generated and compared with previous research. Atomic mix solutions are compared with these benchmark ensemble average flux and k-eigenvalue solutions. Mixing statistics with large standard deviations give the most widely varying ensemble solutions of the flux and k-eigenvalue. The shape of the k-eigenvalue PDF qualitatively agrees with previous work. Its overall shape is independent of variations in fuel cross-sections for the problems considered, but its width is impacted by these variations. Statistical distributions with smaller standard deviations alter the shape of this PDF toward a normal distribution. The atomic mix approximation yields large over-predictions of the ensemble average k-eigenvalue and under-predictions of the flux. Qualitatively correct flux shapes are obtained in some cases. These benchmark calculations indicate that a model which includes higher statistical moments of the mixing statistics is needed for accurate predictions of binary

The OECD/NRC BWR full-size fine-mesh bundle tests benchmark (BFBT)-general description

International Nuclear Information System (INIS)

Sartori, Enrico; Hochreiter, L.E.; Ivanov, Kostadin; Utsuno, Hideaki

2004-01-01

The need to refine models for best-estimate calculations based on good-quality experimental data have been expressed in many recent meetings in the field of nuclear applications. The needs arising in this respect should not be limited to currently available macroscopic approaches but should be extended to next-generation approaches that focus on more microscopic processes. One most valuable database identified for the thermal-hydraulics modelling was developed by the Nuclear Power Engineering Corporation (NUPEC). Part of this database will be made available for an international benchmark exercise. This fine-mesh high-quality data encourages advancement in the insufficiently developed field of the two-phase flow theory. Considering that the present theoretical approach is relatively immature, the benchmark specification is designed so that it will systematically assess and compare the participants' numerical models on the prediction of detailed void distributions and critical powers. The development of truly mechanistic models for critical power prediction is currently underway. These innovative models should include elementary processes such as void distributions, droplet deposit, liquid film entrainment, etc. The benchmark problem includes both macroscopic and microscopic measurement data. In this context, the sub-channel grade void fraction data are regarded as the macroscopic data, and the digitized computer graphic images are the microscopic data. The proposed benchmark consists of two parts (phases), each part consisting of different exercises: Phase 1- Void distribution benchmark: Exercise 1- Steady-state sub-channel grade benchmark. Exercise 2- Steady-state microscopic grade benchmark. Exercise 3-Transient macroscopic grade benchmark. Phase 2-Critical power benchmark: Exercise 1-Steady-state benchmark. Exercise 2-Transient benchmark. (author)

Russian system of computerized analysis for licensing at atomic industry (SCALA) and its validation on ICSBEP handbook data and on some burnup calculations

International Nuclear Information System (INIS)

Ivanova, T.; Polyakov, A.; Saraeva, T.; Tsiboulia, A.

2001-01-01

Validation of criticality calculations using SCALA was performed using data presented in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. This paper contains the results of statistical analysis of discrepancies between calculated and benchmark-model k eff and conclusions about uncertainties of criticality prediction for different types of multiplying systems following from this analysis. (authors)

Calculations of different transmutation concepts. An international benchmark exercise

International Nuclear Information System (INIS)

2000-01-01

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)

Benchmark evaluation of the RELAP code to calculate boiling in narrow channels

International Nuclear Information System (INIS)

Kunze, J.F.; Loyalka, S.K.; McKibben, J.C.; Hultsch, R.; Oladiran, O.

1990-01-01

The RELAP code has been tested with benchmark experiments (such as the loss-of-fluid test experiments at the Idaho National Engineering Laboratory) at high pressures and temperatures characteristic of those encountered in loss-of-coolant accidents (LOCAs) in commercial light water power reactors. Application of RELAP to the LOCA analysis of a low pressure (< 7 atm) and low temperature (< 100 degree C), plate-type research reactor, such as the University of Missouri Research Reactor (MURR), the high-flux breeder reactor, high-flux isotope reactor, and Advanced Test Reactor, requires resolution of questions involving overextrapolation to very low pressures and low temperatures, and calculations of the pulsed boiling/reflood conditions in the narrow rectangular cross-section channels (typically 2 mm thick) of the plate fuel elements. The practical concern of this problem is that plate fuel temperatures predicted by RELAP5 (MOD2, version 3) during the pulsed boiling period can reach high enough temperatures to cause plate (clad) weakening, though not melting. Since an experimental benchmark of RELAP under such LOCA conditions is not available and since such conditions present substantial challenges to the code, it is important to verify the code predictions. The comparison of the pulsed boiling experiments with the RELAP calculations involves both visual observations of void fraction versus time and measurements of temperatures near the fuel plate surface

Reactor physics tests and benchmark analyses of STACY

International Nuclear Information System (INIS)

Miyoshi, Yoshinori; Umano, Takuya

1996-01-01

The Static Experiment Critical Facility, STACY in the Nuclear Fuel Cycle Safety Engineering Research Facility, NUCEF is a solution type critical facility to accumulate fundamental criticality data on uranyl nitrate solution, plutonium nitrate solution and their mixture. A series of critical experiments have been performed for 10 wt% enriched uranyl nitrate solution using a cylindrical core tank. In these experiments, systematic data of the critical height, differential reactivity of the fuel solution, kinetic parameter and reactor power were measured with changing the uranium concentration of the fuel solution from 313 gU/l to 225 gU/l. Critical data through the first series of experiments for the basic core are reported in this paper for evaluating the accuracy of the criticality safety calculation codes. Benchmark calculations of the neutron multiplication factor k eff for the critical condition were made using a neutron transport code TWOTRAN in the SRAC system and a continuous energy Monte Carlo code MCNP 4A with a Japanese evaluated nuclear data library, JENDL 3.2. (J.P.N.)

Preparation of data for criticality safety evaluation of nuclear fuel cycle facilities

International Nuclear Information System (INIS)

Okuno, Hiroshi; Suyama, Kenya; Yoshiyama, Hiroshi; Tonoike, Kotaro; Miyoshi, Yoshinori

2005-01-01

Nuclear Criticality Safety Handbook/Data Collection, Version 2 was submitted to the Ministry of Education, Culture, Sports, Science and Technology (MEXT) of Japan as a contract report. In this presentation paper, its outline and related recent works are presented. After an introduction in Chapter 1, useful information to obtain the atomic number densities was collected in Chapter 2. The nuclear characteristic parameters for 11 nuclear fuels were provided in Chapter 3, and subcriticality judgment graphs were given in Chapter 4. The estimated critical and estimated lower-limit critical values were supplied for the 11 nuclear fuels as results of calculations by using the Japanese Evaluated Nuclear Data Library, JENDL-3.2, and the continuous energy Monte Carlo neutron transport code MVP in Chapter 5. The results of benchmark calculations based on the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook were summarized into six fuel categories in Chapter 6. As for recent works, subcriticality judgment graphs for U-SiO 2 and Pu-SiO 2 were obtained. Benchmark calculations were made with the combination of the latest version of the library JENDL-3.3 and MVP code for a series of STACY experiments and the estimated critical and estimated lower-limit critical values of 10 wt%-enriched uranium nitrate solutions were calculated. (author)

VERA Pin and Fuel Assembly Depletion Benchmark Calculations by McCARD and DeCART

Energy Technology Data Exchange (ETDEWEB)

Park, Ho Jin; Cho, Jin Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Monte Carlo (MC) codes have been developed and used to simulate a neutron transport since MC method was devised in the Manhattan project. Solving the neutron transport problem with the MC method is simple and straightforward to understand. Because there are few essential approximations for the 6- dimension phase of a neutron such as the location, energy, and direction in MC calculations, highly accurate solutions can be obtained through such calculations. In this work, the VERA pin and fuel assembly (FA) depletion benchmark calculations are performed to examine the depletion capability of the newly generated DeCART multi-group cross section library. To obtain the reference solutions, MC depletion calculations are conducted using McCARD. Moreover, to scrutinize the effect by stochastic uncertainty propagation, uncertainty propagation analyses are performed using a sensitivity and uncertainty (S/U) analysis method and stochastic sampling (S.S) method. It is still expensive and challenging to perform a depletion analysis by a MC code. Nevertheless, many studies and works for a MC depletion analysis have been conducted to utilize the benefits of the MC method. In this study, McCARD MC and DeCART MOC transport calculations are performed for the VERA pin and FA depletion benchmarks. The DeCART depletion calculations are conducted to examine the depletion capability of the newly generated multi-group cross section library. The DeCART depletion calculations give excellent agreement with the McCARD reference one. From the McCARD results, it is observed that the MC depletion results depend on how to split the burnup interval. First, only to quantify the effect of the stochastic uncertainty propagation at 40 DTS, the uncertainty propagation analyses are performed using the S/U and S.S. method.

Benchmark comparisons of evaluated nuclear data files

International Nuclear Information System (INIS)

Resler, D.A.; Howerton, R.J.; White, R.M.

1994-05-01

With the availability and maturity of several evaluated nuclear data files, it is timely to compare the results of integral tests with calculations using these different files. We discuss here our progress in making integral benchmark tests of the following nuclear data files: ENDL-94, ENDF/B-V and -VI, JENDL-3, JEF-2, and BROND-2. The methods used to process these evaluated libraries in a consistent way into applications files for use in Monte Carlo calculations is presented. Using these libraries, we are calculating and comparing to experiment k eff for 68 fast critical assemblies of 233,235 U and 239 Pu with reflectors of various material and thickness

Criticality calculations for homogeneous mixtures of uranium and plutonium

International Nuclear Information System (INIS)

Spiegelberg, R. de S.H.

1981-05-01

Critical parameters were calculated using the one-dimensional multigroup transport theory. Calculations have been performed for water mixture of uranium metal and uranium oxides and plutonium nitrates to determine the dimensions of simple critical geometries (sphere and cylinder). The results of the calculations were plotted showing critical parameters (volume, radius or critical mass). The critical values obtained in Handbuch zur Kritikalitat were used to compare with critical parameters. A sensitivity study for the influences of mesh space size, multigroup structure and order of the S sub(n) approximation on the critical radius was carried out. The GAMTEC-II code was used to generate multigroup cross sections data. Critical radius were calculated using the one-dimensional multigroup transport code DTF-IV. (Author) [pt

Specification of phase 3 benchmark (Hex-Z heterogeneous and burnup calculation)

International Nuclear Information System (INIS)

Kim, Y.I.

2002-01-01

During the second RCM of the IAEA Co-ordinated Research Project Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects the following items were identified as important. Heterogeneity will affect absolute core reactivity. Rod worths could be considerably reduced by heterogeneity effects depending on their detailed design. Heterogeneity effects will affect the resonance self-shielding in the treatment of fuel Doppler, steel Doppler and sodium density effects. However, it was considered more important to concentrate on the sodium density effect in order to reduce the calculational effort required. It was also recognized that burnup effects will have an influence on fuel Doppler and sodium worths. A benchmark for the assessment of heterogeneity effect for Phase 3 was defined. It is to be performed for the Hex-Z model of the reactor only. No calculations will be performed for the R-Z model. For comparison with heterogeneous evaluations, the control rod worth will be calculated at the beginning of the equilibrium cycle, based on the homogeneous model. The definitions of rod raised and rod inserted for SHR are given, using the composition numbers

Benchmark calculation for radioactivity inventory using MAXS library based on JENDL-4.0 and JEFF-3.0/A for decommissioning BWR plants

Directory of Open Access Journals (Sweden)

Tanaka Ken-ichi

2016-01-01

Full Text Available We performed benchmark calculation for radioactivity activated in a Primary Containment Vessel (PCV of a Boiling Water Reactor (BWR by using MAXS library, which was developed by collapsing with neutron energy spectra in the PCV of the BWR. Radioactivities due to neutron irradiation were measured by using activation foil detector of Gold (Au and Nickel (Ni at thirty locations in the PCV. We performed activation calculations of the foils with SCALE5.1/ORIGEN-S code with irradiation conditions of each foil location as the benchmark calculation. We compared calculations and measurements to estimate an effectiveness of MAXS library.

Systematic approach to establishing criticality biases

International Nuclear Information System (INIS)

Larson, S.L.

1995-09-01

A systematic approach has been developed to determine benchmark biases and apply those biases to code results to meet the requirements of DOE Order 5480.24 regarding documenting criticality safety margins. Previously, validation of the code against experimental benchmarks to prove reasonable agreement was sufficient. However, DOE Order 5480.24 requires contractors to adhere to the requirements of ANSI/ANS-8.1 and establish subcritical margins. A method was developed to incorporate biases and uncertainties from benchmark calculations into a k eff value with quantifiable uncertainty. The method produces a 95% confidence level in both the k eff value of the scenario modeled and the distribution of the k eff S calculated by the Monte Carlo code. Application of the method to a group of benchmarks modeled using the KENO-Va code and the SCALE 27 group cross sections is also presented

Calculations of the IAEA-CRP-6 Benchmark Cases by Using the ABAQUS FE Model for a Comparison with the COPA Results

International Nuclear Information System (INIS)

Cho, Moon-Sung; Kim, Y. M.; Lee, Y. W.; Jeong, K. C.; Kim, Y. K.; Oh, S. C.

2006-01-01

The fundamental design for a gas-cooled reactor relies on an understanding of the behavior of a coated particle fuel. KAERI, which has been carrying out the Korean VHTR (Very High Temperature modular gas cooled Reactor) Project since 2004, is developing a fuel performance analysis code for a VHTR named COPA (COated Particle fuel Analysis). COPA predicts temperatures, stresses, a fission gas release and failure probabilities of a coated particle fuel in normal operating conditions. Validation of COPA in the process of its development is realized partly by participating in the benchmark section of the international CRP-6 program led by IAEA which provides comprehensive benchmark problems and analysis results obtained from the CRP-6 member countries. Apart from the validation effort through the CRP-6, a validation of COPA was attempted by comparing its benchmark results with the visco-elastic solutions obtained from the ABAQUS code calculations for the same CRP-6 TRISO coated particle benchmark problems involving creep, swelling, and pressure. The study shows the calculation results of the IAEA-CRP-6 benchmark cases 5 through 7 by using the ABAQUS FE model for a comparison with the COPA results

Evaluation of PWR and BWR assembly benchmark calculations. Status report of EPRI computational benchmark results, performed in the framework of the Netherlands` PINK programme (Joint project of ECN, IRI, KEMA and GKN)

Energy Technology Data Exchange (ETDEWEB)

Gruppelaar, H. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Klippel, H.T. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Kloosterman, J.L. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Hoogenboom, J.E. [Technische Univ. Delft (Netherlands). Interfacultair Reactor Instituut; Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Instituut; Verhagen, F.C.M. [Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands); Bruggink, J.C. [Gemeenschappelijke Kernenergiecentrale Nederland N.V., Dodewaard (Netherlands)

1993-11-01

Benchmark results of the Dutch PINK working group on calculational benchmarks on single pin cell and multipin assemblies as defined by EPRI are presented and evaluated. First a short update of methods used by the various institutes involved is given as well as an update of the status with respect to previous performed pin-cell calculations. Problems detected in previous pin-cell calculations are inspected more closely. Detailed discussion of results of multipin assembly calculations is given. The assembly consists of 9 pins in a multicell square lattice in which the central pin is filled differently, i.e. a Gd pin for the BWR assembly and a control rod/guide tube for the PWR assembly. The results for pin cells showed a rather good overall agreement between the four participants although BWR pins with high void fraction turned out to be difficult to calculate. With respect to burnup calculations good overall agreement for the reactivity swing was obtained, provided that a fine time grid is used. (orig.)

Evaluation of PWR and BWR assembly benchmark calculations. Status report of EPRI computational benchmark results, performed in the framework of the Netherlands' PINK programme (Joint project of ECN, IRI, KEMA and GKN)

International Nuclear Information System (INIS)

Gruppelaar, H.; Klippel, H.T.; Kloosterman, J.L.; Hoogenboom, J.E.; Bruggink, J.C.

1993-11-01

Benchmark results of the Dutch PINK working group on calculational benchmarks on single pin cell and multipin assemblies as defined by EPRI are presented and evaluated. First a short update of methods used by the various institutes involved is given as well as an update of the status with respect to previous performed pin-cell calculations. Problems detected in previous pin-cell calculations are inspected more closely. Detailed discussion of results of multipin assembly calculations is given. The assembly consists of 9 pins in a multicell square lattice in which the central pin is filled differently, i.e. a Gd pin for the BWR assembly and a control rod/guide tube for the PWR assembly. The results for pin cells showed a rather good overall agreement between the four participants although BWR pins with high void fraction turned out to be difficult to calculate. With respect to burnup calculations good overall agreement for the reactivity swing was obtained, provided that a fine time grid is used. (orig.)

Creation of a simplified benchmark model for the neptunium sphere experiment

International Nuclear Information System (INIS)

Mosteller, Russell D.; Loaiza, David J.; Sanchez, Rene G.

2004-01-01

Although neptunium is produced in significant amounts by nuclear power reactors, its critical mass is not well known. In addition, sizeable uncertainties exist for its cross sections. As an important step toward resolution of these issues, a critical experiment was conducted in 2002 at the Los Alamos Critical Experiments Facility. In the experiment, a 6-kg sphere of 237 Np was surrounded by nested hemispherical shells of highly enriched uranium. The shells were required in order to reach a critical condition. Subsequently, a detailed model of the experiment was developed. This model faithfully reproduces the components of the experiment, but it is geometrically complex. Furthermore, the isotopics analysis upon which that model is based omits nearly 1 % of the mass of the sphere. A simplified benchmark model has been constructed that retains all of the neutronically important aspects of the detailed model and substantially reduces the computer resources required for the calculation. The reactivity impact, of each of the simplifications is quantified, including the effect of the missing mass. A complete set of specifications for the benchmark is included in the full paper. Both the detailed and simplified benchmark models underpredict k eff by more than 1% Δk. This discrepancy supports the suspicion that better cross sections are needed for 237 Np.

Benchmarking of MCNP for calculating dose rates at an interim storage facility for nuclear waste.

Science.gov (United States)

Heuel-Fabianek, Burkhard; Hille, Ralf

2005-01-01

During the operation of research facilities at Research Centre Jülich, Germany, nuclear waste is stored in drums and other vessels in an interim storage building on-site, which has a concrete shielding at the side walls. Owing to the lack of a well-defined source, measured gamma spectra were unfolded to determine the photon flux on the surface of the containers. The dose rate simulation, including the effects of skyshine, using the Monte Carlo transport code MCNP is compared with the measured dosimetric data at some locations in the vicinity of the interim storage building. The MCNP data for direct radiation confirm the data calculated using a point-kernel method. However, a comparison of the modelled dose rates for direct radiation and skyshine with the measured data demonstrate the need for a more precise definition of the source. Both the measured and the modelled dose rates verified the fact that the legal limits (<1 mSv a(-1)) are met in the area outside the perimeter fence of the storage building to which members of the public have access. Using container surface data (gamma spectra) to define the source may be a useful tool for practical calculations and additionally for benchmarking of computer codes if the discussed critical aspects with respect to the source can be addressed adequately.

Criticality experiments to provide benchmark data on neutron flux traps

International Nuclear Information System (INIS)

Bierman, S.R.

1988-06-01

The experimental measurements covered by this report were designed to provide benchmark type data on water moderated LWR type fuel arrays containing neutron flux traps. The experiments were performed at the US Department of Energy Hanford Critical Mass Laboratory, operated by Pacific Northwest Laboratory. The experimental assemblies consisted of 2 /times/ 2 arrays of 4.31 wt % 235 U enriched UO 2 fuel rods, uniformly arranged in water on a 1.891 cm square center-to-center spacing. Neutron flux traps were created between the fuel units using metal plates containing varying amounts of boron. Measurements were made to determine the effect that boron loading and distance between the fuel and flux trap had on the amount of fuel required for criticality. Also, measurements were made, using the pulse neutron source technique, to determine the effect of boron loading on the effective neutron multiplications constant. On two assemblies, reaction rate measurements were made using solid state track recorders to determine absolute fission rates in 235 U and 238 U. 14 refs., 12 figs., 7 tabs

RB reactor benchmark cores

International Nuclear Information System (INIS)

Pesic, M.

1998-01-01

A selected set of the RB reactor benchmark cores is presented in this paper. The first results of validation of the well-known Monte Carlo MCNP TM code and adjoining neutron cross section libraries are given. They confirm the idea for the proposal of the new U-D 2 O criticality benchmark system and support the intention to include this system in the next edition of the recent OECD/NEA Project: International Handbook of Evaluated Criticality Safety Experiment, in near future. (author)

Riding Bare-Back on unstructured meshes for 21. century criticality calculations - 244

International Nuclear Information System (INIS)

Kelley, K.C.; Martz, R.L.; Crane, D.L.

2010-01-01

MCNP has a new capability that permits tracking of neutrons and photons on an unstructured mesh which is embedded as a mesh universe within its legacy geometry capability. The mesh geometry is created through Abaqus/CAE using its solid modeling capabilities. Transport results are calculated for mesh elements through a path length estimator while element to element tracking is performed on the mesh. The results from MCNP can be exported to Abaqus/CAE for visualization or other-physics analysis. The simple Godiva criticality benchmark problem was tested with this new mesh capability. Computer run time is proportional to the number of mesh elements used. Both first and second order polyhedrons are used. Models that used second order polyhedrons produced slightly better results without significantly increasing computer run time. Models that used first order hexahedrons had shorter runtimes than models that used first order tetrahedrons. (authors)

Review of microscopic integral cross section data in fundamental reactor dosimetry benchmark neutron fields

International Nuclear Information System (INIS)

Fabry, A.; McElroy, W.N.; Kellogg, L.S.; Lippincott, E.P.; Grundl, J.A.; Gilliam, D.M.; Hansen, G.E.

1976-01-01

This paper is intended to review and critically discuss microscopic integral cross section measurement and calculation data for fundamental reactor dosimetry benchmark neutron fields. Specifically the review covers the following fundamental benchmarks: the spontaneous californium-252 fission neutron spectrum standard field; the thermal-neutron induced uranium-235 fission neutron spectrum standard field; the (secondary) intermediate-energy standard neutron field at the center of the Mol-ΣΣ, NISUS, and ITN-ΣΣ facilities; the reference neutron field at the center of the Coupled Fast Reactor Measurement Facility; the reference neutron field at the center of the 10% enriched uranium metal, cylindrical, fast critical; the (primary) Intermediate-Energy Standard Neutron Field

Results of LWR core transient benchmarks

International Nuclear Information System (INIS)

Finnemann, H.; Bauer, H.; Galati, A.; Martinelli, R.

1993-10-01

LWR core transient (LWRCT) benchmarks, based on well defined problems with a complete set of input data, are used to assess the discrepancies between three-dimensional space-time kinetics codes in transient calculations. The PWR problem chosen is the ejection of a control assembly from an initially critical core at hot zero power or at full power, each for three different geometrical configurations. The set of problems offers a variety of reactivity excursions which efficiently test the coupled neutronic/thermal - hydraulic models of the codes. The 63 sets of submitted solutions are analyzed by comparison with a nodal reference solution defined by using a finer spatial and temporal resolution than in standard calculations. The BWR problems considered are reactivity excursions caused by cold water injection and pressurization events. In the present paper, only the cold water injection event is discussed and evaluated in some detail. Lacking a reference solution the evaluation of the 8 sets of BWR contributions relies on a synthetic comparative discussion. The results of this first phase of LWRCT benchmark calculations are quite satisfactory, though there remain some unresolved issues. It is therefore concluded that even more challenging problems can be successfully tackled in a suggested second test phase. (authors). 46 figs., 21 tabs., 3 refs

Preliminary Results for the OECD/NEA Time Dependent Benchmark using Rattlesnake, Rattlesnake-IQS and TDKENO

Energy Technology Data Exchange (ETDEWEB)

DeHart, Mark D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mausolff, Zander [Univ. of Florida, Gainesville, FL (United States); Weems, Zach [Univ. of Florida, Gainesville, FL (United States); Popp, Dustin [Univ. of Florida, Gainesville, FL (United States); Smith, Kristin [Univ. of Florida, Gainesville, FL (United States); Shriver, Forrest [Univ. of Florida, Gainesville, FL (United States); Goluoglu, Sedat [Univ. of Florida, Gainesville, FL (United States); Prince, Zachary [Texas A & M Univ., College Station, TX (United States); Ragusa, Jean [Texas A & M Univ., College Station, TX (United States)

2016-08-01

One goal of the MAMMOTH M&S project is to validate the analysis capabilities within MAMMOTH. Historical data has shown limited value for validation of full three-dimensional (3D) multi-physics methods. Initial analysis considered the TREAT startup minimum critical core and one of the startup transient tests. At present, validation is focusing on measurements taken during the M8CAL test calibration series. These exercises will valuable in preliminary assessment of the ability of MAMMOTH to perform coupled multi-physics calculations; calculations performed to date are being used to validate the neutron transport solver Rattlesnake\\cite{Rattlesnake} and the fuels performance code BISON. Other validation projects outside of TREAT are available for single-physics benchmarking. Because the transient solution capability of Rattlesnake is one of the key attributes that makes it unique for TREAT transient simulations, validation of the transient solution of Rattlesnake using other time dependent kinetics benchmarks has considerable value. The Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) has recently developed a computational benchmark for transient simulations. This benchmark considered both two-dimensional (2D) and 3D configurations for a total number of 26 different transients. All are negative reactivity insertions, typically returning to the critical state after some time.

Preliminary Results for the OECD/NEA Time Dependent Benchmark using Rattlesnake, Rattlesnake-IQS and TDKENO

International Nuclear Information System (INIS)

DeHart, Mark D.; Mausolff, Zander; Weems, Zach; Popp, Dustin; Smith, Kristin; Shriver, Forrest; Goluoglu, Sedat; Prince, Zachary; Ragusa, Jean

2016-01-01

One goal of the MAMMOTH M&S project is to validate the analysis capabilities within MAMMOTH. Historical data has shown limited value for validation of full three-dimensional (3D) multi-physics methods. Initial analysis considered the TREAT startup minimum critical core and one of the startup transient tests. At present, validation is focusing on measurements taken during the M8CAL test calibration series. These exercises will valuable in preliminary assessment of the ability of MAMMOTH to perform coupled multi-physics calculations; calculations performed to date are being used to validate the neutron transport solver Rattlesnake\\citelesnake) and the fuels performance code BISON. Other validation projects outside of TREAT are available for single-physics benchmarking. Because the transient solution capability of Rattlesnake is one of the key attributes that makes it unique for TREAT transient simulations, validation of the transient solution of Rattlesnake using other time dependent kinetics benchmarks has considerable value. The Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) has recently developed a computational benchmark for transient simulations. This benchmark considered both two-dimensional (2D) and 3D configurations for a total number of 26 different transients. All are negative reactivity insertions, typically returning to the critical state after some time.

Experiments for IFR fuel criticality in ZPPR-21

International Nuclear Information System (INIS)

Olsen, D.N.; Collins, P.J.; Carpenter, S.G.

1991-01-01

A series of benchmark measurements was made in ZPPR-21 to validate criticality calculations for fuel processing operations for Argonne's Integral Fast Reactor program. Six different mixtures of Pu/U/Zr fuel with a graphite reflector were built and criticality was determined by period measurements. The assemblies were isolated from room return neutrons by a lithium hydride shield. Analysis was done using a fully-detailed model with the VIM Monte Carlo code and ENDF/B-V.2 data. Sensitivity analysis was used to validate the measurements against other benchmark data. A simple RZ model was defined and used with the KENO code. Corrections to the RZ model were provided by the VIM calculations with low statistical uncertainty. (Author)

Benchmark Evaluation of Start-Up and Zero-Power Measurements at the High-Temperature Engineering Test Reactor

International Nuclear Information System (INIS)

Bess, John D.; Fujimoto, Nozomu

2014-01-01

Benchmark models were developed to evaluate six cold-critical and two warm-critical, zero-power measurements of the HTTR. Additional measurements of a fully-loaded subcritical configuration, core excess reactivity, shutdown margins, six isothermal temperature coefficients, and axial reaction-rate distributions were also evaluated as acceptable benchmark experiments. Insufficient information is publicly available to develop finely-detailed models of the HTTR as much of the design information is still proprietary. However, the uncertainties in the benchmark models are judged to be of sufficient magnitude to encompass any biases and bias uncertainties incurred through the simplification process used to develop the benchmark models. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the impurity content of the various graphite blocks that comprise the HTTR. Monte Carlo calculations of keff are between approximately 0.9 % and 2.7 % greater than the benchmark values. Reevaluation of the HTTR models as additional information becomes available could improve the quality of this benchmark and possibly reduce the computational biases. High-quality characterization of graphite impurities would significantly improve the quality of the HTTR benchmark assessment. Simulation of the other reactor physics measurements are in good agreement with the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments

On the thermal scattering law data for reactor lattice calculations

International Nuclear Information System (INIS)

Trkov, A.; Mattes, M.

2004-01-01

Thermal scattering law data for hydrogen bound in water, hydrogen bound in zirconium hydride and deuterium bound in heavy water have been re-evaluated. The influence of the thermal scattering law data on critical lattices has been studied with detailed Monte Carlo calculations and a summary of results is presented for a numerical benchmark and for the TRIGA reactor benchmark. Systematics for a large sequence of benchmarks analysed with the WIMS-D lattice code are also presented. (author)

Verification of NUREC Code Transient Calculation Capability Using OECD NEA/US NRC PWR MOX/UO2 Core Transient Benchmark Problem

International Nuclear Information System (INIS)

Joo, Hyung Kook; Noh, Jae Man; Lee, Hyung Chul; Yoo, Jae Woon

2006-01-01

In this report, we verified the NUREC code transient calculation capability using OECD NEA/US NRC PWR MOX/UO2 Core Transient Benchmark Problem. The benchmark problem consists of Part 1, a 2-D problem with given T/H conditions, Part 2, a 3-D problem at HFP condition, Part 3, a 3-D problem at HZP condition, and Part 4, a transient state initiated by a control rod ejection at HZP condition in Part 3. In Part 1, the results of NUREC code agreed well with the reference solution obtained from DeCART calculation except for the pin power distributions at the rodded assemblies. In Part 2, the results of NUREC code agreed well with the reference DeCART solutions. In Part 3, some results of NUREC code such as critical boron concentration and core averaged delayed neutron fraction agreed well with the reference PARCS 2G solutions. But the error of the assembly power at the core center was quite large. The pin power errors of NUREC code at the rodded assemblies was much smaller the those of PARCS code. The axial power distribution also agreed well with the reference solution. In Part 4, the results of NUREC code agreed well with those of PARCS 2G code which was taken as the reference solution. From the above results we can conclude that the results of NUREC code for steady states and transient states of the MOX loaded LWR core agree well with those of the other codes

Ensuring the validity of calculated subcritical limits

International Nuclear Information System (INIS)

Clark, H.K.

1977-01-01

The care taken at the Savannah River Laboratory and Plant to ensure the validity of calculated subcritical limits is described. Close attention is given to ANSI N16.1-1975, ''Validation of Calculational Methods for Nuclear Criticality Safety.'' The computer codes used for criticality safety computations, which are listed and are briefly described, have been placed in the SRL JOSHUA system to facilitate calculation and to reduce input errors. A driver module, KOKO, simplifies and standardizes input and links the codes together in various ways. For any criticality safety evaluation, correlations of the calculational methods are made with experiment to establish bias. Occasionally subcritical experiments are performed expressly to provide benchmarks. Calculated subcritical limits contain an adequate but not excessive margin to allow for uncertainty in the bias. The final step in any criticality safety evaluation is the writing of a report describing the calculations and justifying the margin

Researches on nuclear criticality safety evaluation

Energy Technology Data Exchange (ETDEWEB)

Okuno, Hiroshi; Suyama, Kenya; Nomura, Yasushi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2003-10-01

For criticality safety evaluation of burnup fuel, the general-purpose burnup calculation code, SWAT, was revised, and its precision was confirmed through comparison with other results from OECD/NEA's burnup credit benchmarks. Effect by replacing the evaluated nuclear data from JENDL-3.2 to ENDF/B-VI and JEF-2.2 was also studied. Correction factors were derived for conservative evaluation of nuclide concentrations obtained with the simplified burnup code ORIGEN2.1. The critical masses of curium were calculated and evaluated for nuclear criticality safety management of minor actinides. (author)

Researches on nuclear criticality safety evaluation

International Nuclear Information System (INIS)

Okuno, Hiroshi; Suyama, Kenya; Nomura, Yasushi

2003-01-01

For criticality safety evaluation of burnup fuel, the general-purpose burnup calculation code, SWAT, was revised, and its precision was confirmed through comparison with other results from OECD/NEA's burnup credit benchmarks. Effect by replacing the evaluated nuclear data from JENDL-3.2 to ENDF/B-VI and JEF-2.2 was also studied. Correction factors were derived for conservative evaluation of nuclide concentrations obtained with the simplified burnup code ORIGEN2.1. The critical masses of curium were calculated and evaluated for nuclear criticality safety management of minor actinides. (author)

Uranium systems to enhance benchmarks for use in the verification of criticality safety computer models. Final report, February 16, 1990--December 31, 1994

International Nuclear Information System (INIS)

Busch, R.D.

1995-01-01

Dr. Robert Busch of the Department of Chemical and Nuclear Engineering was the principal investigator on this project with technical direction provided by the staff in the Nuclear Criticality Safety Group at Los Alamos. During the period of the contract, he had a number of graduate and undergraduate students working on subtasks. The objective of this work was to develop information on uranium systems to enhance benchmarks for use in the verification of criticality safety computer models. During the first year of this project, most of the work was focused on setting up the SUN SPARC-1 Workstation and acquiring the literature which described the critical experiments. By august 1990, the Workstation was operational with the current version of TWODANT loaded on the system. MCNP, version 4 tape was made available from Los Alamos late in 1990. Various documents were acquired which provide the initial descriptions of the critical experiments under consideration as benchmarks. The next four years were spent working on various benchmark projects. A number of publications and presentations were made on this material. These are briefly discussed in this report

Experimental and computational benchmark tests

International Nuclear Information System (INIS)

Gilliam, D.M.; Briesmeister, J.F.

1994-01-01

A program involving principally NIST, LANL, and ORNL has been in progress for about four years now to establish a series of benchmark measurements and calculations related to the moderation and leakage of 252 Cf neutrons from a source surrounded by spherical aqueous moderators of various thicknesses and compositions. The motivation for these studies comes from problems in criticality calculations concerning arrays of multiplying components, where the leakage from one component acts as a source for the other components. This talk compares experimental and calculated values for the fission rates of four nuclides - 235 U, 239 Pu, 238 U, and 237 Np - in the leakage spectrum from moderator spheres of diameters 76.2 mm, 101.6 mm, and 127.0 mm, with either pure water or enriched B-10 solutions as the moderator. Very detailed Monte Carlo calculations were done with the MCNP code, using a open-quotes light waterclose quotes S(α,β) scattering kernel

Criticality calculations for a critical assembly, graphite moderate, using 20% enriched uranium

International Nuclear Information System (INIS)

Almeida Ferreira, A.C. de; Hukai, R.Y.

1975-01-01

The construction of a Zero Power Reactor (ZPR) at the Instituto de Energia Atomica in order to measure the neutron characteristics (parameters) of HTGR reactors is proposed. The necessary quantity fissile uranium for these measurements has been calculed. Criticality studies of graphite moderated critical assemblies containing thorium have been made and the critical mass of each of several typical commercial HTGR compositions has been calculated using computer codes HAMMER and CITATION. Assemblies investigated contained a central cylindrical core region, simulating a typical commercial HTGR composition, a uranium-graphite driver region and a outer pure graphite reflector region. It is concluded that a 10Kg inventory of fissile uranium will be required for a program of measurements utilizing each of the several calculated assemblies

Proposal on the accelerator driven molten-salt reactor (ATW concept) benchmark calculations. (STAGE 1 - without an external neutron source)

International Nuclear Information System (INIS)

Svarny, J.; Mikolas, P.

1999-01-01

The first stage of ATW neutronic benchmark (without an external source), based on the simple modelling of two component concept is presented. The simple model of two component concept of the ATW (graphite + molten salt system) was found. The main purpose of this benchmark is not only to provide the basic characteristics of given ADS but also to test codes in calculations of the rate of transmutation waste and to evaluate basic kinetics parameters and reactivity effects. (author)

The solution of the LEU and MOX WWER-1000 calculation benchmark with the CARATE - multicell code

International Nuclear Information System (INIS)

Hordosy, G.; Maraczy, Cs.

2000-01-01

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 calculations in multigroup and multidimensional time-dependent transport

International Nuclear Information System (INIS)

Ganapol, B.D.; Musso, E.; Ravetto, P.; Sumini, M.

1990-01-01

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

Reactor fuel depletion benchmark of TINDER

International Nuclear Information System (INIS)

Martin, W.J.; Oliveira, C.R.E. de; Hecht, A.A.

2014-01-01

Highlights: • A reactor burnup benchmark of TINDER, coupling MCNP6 to CINDER2008, was performed. • TINDER is a poor candidate for fuel depletion calculations using its current libraries. • Data library modification is necessary if fuel depletion is desired from TINDER. - Abstract: Accurate burnup calculations are key to proper nuclear reactor design, fuel cycle modeling, and disposal estimations. The TINDER code, originally designed for activation analyses, has been modified to handle full burnup calculations, including the widely used predictor–corrector feature. In order to properly characterize the performance of TINDER for this application, a benchmark calculation was performed. Although the results followed the trends of past benchmarked codes for a UO 2 PWR fuel sample from the Takahama-3 reactor, there were obvious deficiencies in the final result, likely in the nuclear data library that was used. Isotopic comparisons versus experiment and past code benchmarks are given, as well as hypothesized areas of deficiency and future work

Calculation of the minimum critical mass of fissile nuclides

International Nuclear Information System (INIS)

Wright, R.Q.; Hopper, Calvin Mitchell

2008-01-01

The OB-1 method for the calculation of the minimum critical mass of fissile actinides in metal/water systems was described in a previous paper. A fit to the calculated minimum critical mass data using the extended criticality parameter is the basis of the revised method. The solution density (grams/liter) for the minimum critical mass is also obtained by a fit to calculated values. Input to the calculation consists of the Maxwellian averaged fission and absorption cross sections and the thermal values of nubar. The revised method gives more accurate values than the original method does for both the minimum critical mass and the solution densities. The OB-1 method has been extended to calculate the uncertainties in the minimum critical mass for 12 different fissile nuclides. The uncertainties for the fission and capture cross sections and the estimated nubar uncertainties are used to determine the uncertainties in the minimum critical mass, either in percent or grams. Results have been obtained for U-233, U-235, Pu-236, Pu-239, Pu-241, Am-242m, Cm-243, Cm-245, Cf-249, Cf-251, Cf-253, and Es-254. Eight of these 12 nuclides are included in the ANS-8.15 standard.

Comparative analysis of nine structural codes used in the second WIPP benchmark problem

International Nuclear Information System (INIS)

Morgan, H.S.; Krieg, R.D.; Matalucci, R.V.

1981-11-01

In the Waste Isolation Pilot Plant (WIPP) Benchmark II study, various computer codes were compared on the basis of their capabilities for calculating the response of hypothetical drift configurations for nuclear waste experiments and storage demonstration. The codes used by participants in the study were ANSALT, DAPROK, JAC, REM, SANCHO, SPECTROM, STEALTH, and two different implementations of MARC. Errors were found in the preliminary results, and several calculations were revised. Revised solutions were in reasonable agreement except for the REM solution. The Benchmark II study allowed significant advances in understanding the relative behavior of computer codes available for WIPP calculations. The study also pointed out the possible need for performing critical design calculations with more than one code. Lastly, it indicated the magnitude of the code-to-code spread in results which is to be expected even when a model has been explicitly defined

Review of microscopic integral cross section data in fundamental reactor dosimetry benchmark neutron fields

International Nuclear Information System (INIS)

Fabry, A.; McElroy, W.N.; Kellogg, L.S.; Lippincott, E.P.; Grundl, J.A.; Gilliam, D.M.; Hansen, G.E.

1976-10-01

The paper is intended to review and critically discuss microscopic integral cross section measurement and calculation data for fundamental reactor dosimetry benchmark neutron fields. Specifically the review covers the following fundamental benchmarks: (1) the spontaneous californium-252 fission neutron spectrum standard field; (2) the thermal-neutron induced uranium-235 fission neutron spectrum standard field; (3) the (secondary) intermediate-energy standard neutron field at the center of the Mol-ΣΣ, NISUS, and ITN--ΣΣ facilities; (4) the reference neutron field at the center of the Coupled Fast Reactor Measurement Facility (CFRMF); (5) the reference neutron field at the center of the 10 percent enriched uranium metal, cylindrical, fast critical; and (6) the (primary) Intermediate-Energy Standard Neutron Field

Benchmarks for GADRAS performance validation

International Nuclear Information System (INIS)

Mattingly, John K.; Mitchell, Dean James; Rhykerd, Charles L. Jr.

2009-01-01

The performance of the Gamma Detector Response and Analysis Software (GADRAS) was validated by comparing GADRAS model results to experimental measurements for a series of benchmark sources. Sources for the benchmark include a plutonium metal sphere, bare and shielded in polyethylene, plutonium oxide in cans, a highly enriched uranium sphere, bare and shielded in polyethylene, a depleted uranium shell and spheres, and a natural uranium sphere. The benchmark experimental data were previously acquired and consist of careful collection of background and calibration source spectra along with the source spectra. The calibration data were fit with GADRAS to determine response functions for the detector in each experiment. A one-dimensional model (pie chart) was constructed for each source based on the dimensions of the benchmark source. The GADRAS code made a forward calculation from each model to predict the radiation spectrum for the detector used in the benchmark experiment. The comparisons between the GADRAS calculation and the experimental measurements are excellent, validating that GADRAS can correctly predict the radiation spectra for these well-defined benchmark sources.

Analysis of an OECD/NEA high-temperature reactor benchmark

International Nuclear Information System (INIS)

Hosking, J. G.; Newton, T. D.; Koeberl, O.; Morris, P.; Goluoglu, S.; Tombakoglu, T.; Colak, U.; Sartori, E.

2006-01-01

This paper describes analyses of the OECD/NEA HTR benchmark organized by the 'Working Party on the Scientific Issues of Reactor Systems (WPRS)', formerly the 'Working Party on the Physics of Plutonium Fuels and Innovative Fuel Cycles'. The benchmark was specifically designed to provide inter-comparisons for plutonium and thorium fuels when used in HTR systems. Calculations considering uranium fuel have also been included in the benchmark, in order to identify any increased uncertainties when using plutonium or thorium fuels. The benchmark consists of five phases, which include cell and whole-core calculations. Analysis of the benchmark has been performed by a number of international participants, who have used a range of deterministic and Monte Carlo code schemes. For each of the benchmark phases, neutronics parameters have been evaluated. Comparisons are made between the results of the benchmark participants, as well as comparisons between the predictions of the deterministic calculations and those from detailed Monte Carlo calculations. (authors)

Experiments for IFR fuel criticality in ZPPR-21

International Nuclear Information System (INIS)

Olsen, D.N.; Collins, P.J.; Carpenter, S.G.

1991-01-01

A series of benchmark measurements was made in ZPPR-21 to validate criticality calculations for fuel operations in Argonne's Integral Fast Reactor. Six different mixtures of Pu/U/Zr fuel with a graphite reflector were built and criticality was determined by period measurements. The assemblies were isolated from room return problems by a lithium hydride shield. Analysis was done using a fully-detailed model with the VIM Monte Carlo code and ENDF/B-V.2 data. Sensitivity analysis was used to validate the measurements against other benchmark data. A simple RZ model was defined the used with the KENO code. Corrections to the RZ model were provided by the VIM calculations with low statistical uncertainty. 7 refs., 5 figs., 5 tabs

Identification of critical parameters for PEMFC stack performance characterization and control strategies for reliable and comparable stack benchmarking

DEFF Research Database (Denmark)

Mitzel, Jens; Gülzow, Erich; Kabza, Alexander

2016-01-01

This paper is focused on the identification of critical parameters and on the development of reliable methodologies to achieve comparable benchmark results. Possibilities for control sensor positioning and for parameter variation in sensitivity tests are discussed and recommended options for the ...

Radiation Detection Computational Benchmark Scenarios

Energy Technology Data Exchange (ETDEWEB)

Shaver, Mark W.; Casella, Andrew M.; Wittman, Richard S.; McDonald, Ben S.

2013-09-24

Modeling forms an important component of radiation detection development, allowing for testing of new detector designs, evaluation of existing equipment against a wide variety of potential threat sources, and assessing operation performance of radiation detection systems. This can, however, result in large and complex scenarios which are time consuming to model. A variety of approaches to radiation transport modeling exist with complementary strengths and weaknesses for different problems. This variety of approaches, and the development of promising new tools (such as ORNL’s ADVANTG) which combine benefits of multiple approaches, illustrates the need for a means of evaluating or comparing different techniques for radiation detection problems. This report presents a set of 9 benchmark problems for comparing different types of radiation transport calculations, identifying appropriate tools for classes of problems, and testing and guiding the development of new methods. The benchmarks were drawn primarily from existing or previous calculations with a preference for scenarios which include experimental data, or otherwise have results with a high level of confidence, are non-sensitive, and represent problem sets of interest to NA-22. From a technical perspective, the benchmarks were chosen to span a range of difficulty and to include gamma transport, neutron transport, or both and represent different important physical processes and a range of sensitivity to angular or energy fidelity. Following benchmark identification, existing information about geometry, measurements, and previous calculations were assembled. Monte Carlo results (MCNP decks) were reviewed or created and re-run in order to attain accurate computational times and to verify agreement with experimental data, when present. Benchmark information was then conveyed to ORNL in order to guide testing and development of hybrid calculations. The results of those ADVANTG calculations were then sent to PNNL for

Core Benchmarks Descriptions

International Nuclear Information System (INIS)

Pavlovichev, A.M.

2001-01-01

Actual regulations while designing of new fuel cycles for nuclear power installations comprise a calculational justification to be performed by certified computer codes. It guarantees that obtained calculational results will be within the limits of declared uncertainties that are indicated in a certificate issued by Gosatomnadzor of Russian Federation (GAN) and concerning a corresponding computer code. A formal justification of declared uncertainties is the comparison of calculational results obtained by a commercial code with the results of experiments or of calculational tests that are calculated with an uncertainty defined by certified precision codes of MCU type or of other one. The actual level of international cooperation provides an enlarging of the bank of experimental and calculational benchmarks acceptable for a certification of commercial codes that are being used for a design of fuel loadings with MOX fuel. In particular, the work is practically finished on the forming of calculational benchmarks list for a certification of code TVS-M as applied to MOX fuel assembly calculations. The results on these activities are presented

Criticality and Its Uncertainty Analysis of Spent Fuel Storage Rack for Research Reactor

International Nuclear Information System (INIS)

Han, Tae Young; Park, Chang Je; Lee, Byung Chul

2011-01-01

For evaluating the criticality safety of spent fuel storage rack in an open pool type research reactor, a permissible upper limit of criticality should be determined. It can be estimated from the criticality upper limit presented by the regulatory guide and an uncertainty of criticality calculation. In this paper, criticalities for spent fuel storage rack are carried out at various conditions. The calculation uncertainty of MCNP system is evaluated from the calculation results for the benchmark experiments. Then, the upper limit of criticality is determined from the uncertainties and the calculated criticality of the spent fuel storage rack is evaluated

Benchmark for the qualification of gamma shielding calculation methods for light-water type reactor spent fuels

International Nuclear Information System (INIS)

Blum, P.; Cagnon, R.; Nimal, J.C.

1982-01-01

This report gives the results of a campaign of gamma dose rates measurement in the vicinity of a transport package loaded with 12 PWR spent fuel assemblies, so that the characteristics of the package and the fuel. It describes the measuring methods, and gives the accuracy of the data which will be usefull, as benchmarks, to the control of the calculation methods used to verify the gamma shielding of the packages. It shows how to calculate gamma dose rates from the data given on the package and the fuel, and gives the results of a calculation with the Mecure IV code and compares them to the measurements

RELAP5/MOD2 benchmarking study: Critical heat flux under low-flow conditions

International Nuclear Information System (INIS)

Ruggles, E.; Williams, P.T.

1990-01-01

Experimental studies by Mishima and Ishii performed at Argonne National Laboratory and subsequent experimental studies performed by Mishima and Nishihara have investigated the critical heat flux (CHF) for low-pressure low-mass flux situations where low-quality burnout may occur. These flow situations are relevant to long-term decay heat removal after a loss of forced flow. The transition from burnout at high quality to burnout at low quality causes very low burnout heat flux values. Mishima and Ishii postulated a model for the low-quality burnout based on flow regime transition from churn turbulent to annular flow. This model was validated by both flow visualization and burnout measurements. Griffith et al. also studied CHF in low mass flux, low-pressure situations and correlated data for upflows, counter-current flows, and downflows with the local fluid conditions. A RELAP5/MOD2 CHF benchmarking study was carried out investigating the performance of the code for low-flow conditions. Data from the experimental study by Mishima and Ishii were the basis for the benchmark comparisons

Analysis of the European results on the HTTR's core physics benchmarks

International Nuclear Information System (INIS)

Raepsaet, X.; Damian, F.; Ohlig, U.A.; Brockmann, H.J.; Haas, J.B.M. de; Wallerboss, E.M.

2002-01-01

Within the frame of the European contract HTR-N1 calculations are performed on the benchmark problems of the HTTR's start-up core physics experiments initially proposed by the IAEA in a Co-ordinated Research Programme. Three European partners, the FZJ in Germany, NRG and IRI in the Netherlands, and CEA in France, have joined this work package with the aim to validate their calculational methods. Pre-test and post-test calculational results, obtained by the partners, are compared with each other and with the experiment. Parts of the discrepancies between experiment and pre-test predictions are analysed and tackled by different treatments. In the case of the Monte Carlo code TRIPOLI4, used by CEA, the discrepancy between measurement and calculation at the first criticality is reduced to Δk/k∼0.85%, when considering the revised data of the HTTR benchmark. In the case of the diffusion codes, this discrepancy is reduced to: Δk/k∼0.8% (FZJ) and 2.7 or 1.8% (CEA). (author)

Validation of the ABBN/CONSYST constants system. Part 1: Validation through the critical experiments on compact metallic cores

International Nuclear Information System (INIS)

Ivanova, T.T.; Manturov, G.N.; Nikolaev, M.N.; Rozhikhin, E.V.; Semenov, M.Yu.; Tsiboulia, A.M.

1999-01-01

Worldwide compilation of criticality safety benchmark experiments, evaluated due to an activity of the International Criticality Safety Benchmark Evaluation Project (ICSBEP), discovers new possibilities for validation of the ABBN-93.1 cross section library for criticality safety analysis. Results of calculations of small assemblies with metal-fuelled cores are presented in this paper. It is concluded that ABBN-93.1 predicts criticality of such systems with required accuracy

Coupled fast-thermal core 'HERBE', as the benchmark experiment at the RB reactor

International Nuclear Information System (INIS)

Pesic, M.

2003-10-01

Validation of the well-known Monte Carlo code MCNP TM against measured criticality data for the coupled fast-thermal HERBE. System at the RB research reactor is shown in this paper. Experimental data are obtained for regular HERBE core and for the cases of controlled flooding of the neutron converter zone by heavy water. Earlier calculations of these criticality parameters, done by combination of transport and diffusion codes using 2D geometry model are also compared to new calculations carried out by the MCNP code in 3D geometry, applying new detailed 3D model of the HEU fuel slug, developed recently. Satisfactory agreements in comparison of the HERBE criticality calculation results with experimental data, in spite complex heterogeneous composition of the HERBE core, are obtained and confirmed that HERBE core could be used as a criticality benchmark for coupled fast-thermal core. (author)

Start-up of a cold loop in a VVER-440, the 7th AER benchmark calculation with HEXTRAN-SMABRE-PORFLO

International Nuclear Information System (INIS)

Hovi, Ville; Taivassalo, Veikko; Haemaelaeinen, Anitta; Raety, Hanna; Syrjaelahti, Elina

2017-01-01

The 7 th dynamic AER benchmark is the first in which three-dimensional thermal hydraulics codes are supposed to be applied. The aim is to get a more precise core inlet temperature profile than the sector temperatures available typically with system codes. The benchmark consists of a start-up of the sixth, isolated loop in a VVER-440 plant. The isolated loop initially contains cold water without boric acid and the start-up leads to a somewhat asymmetrical core power increase due to feedbacks in the core. In this study, the 7 th AER benchmark is calculated with the three-dimensional nodal reactor dynamics code HEXTRAN-SMABRE coupled with the porous computational fluid dynamics code PORFLO. These three codes are developed at VTT. A novel two-way coupled simulation of the 7 th AER benchmark was performed successfully demonstrating the feasibility and advantages of the new reactor analysis framework. The modelling issues for this benchmark are reported and some evaluation against the previously reported comparisons between the system codes is provided.

Attila calculations for the 3-D C5G7 benchmark extension

International Nuclear Information System (INIS)

Wareing, T.A.; McGhee, J.M.; Barnett, D.A.; Failla, G.A.

2005-01-01

The performance of the Attila radiation transport software was evaluated for the 3-D C5G7 MOX benchmark extension, a follow-on study to the MOX benchmark developed by the 'OECD/NEA Expert Group on 3-D Radiation Transport Benchmarks'. These benchmarks were designed to test the ability of modern deterministic transport methods to model reactor problems without spatial homogenization. Attila is a general purpose radiation transport software package with an integrated graphical user interface (GUI) for analysis, set-up and postprocessing. Attila provides solutions to the discrete-ordinates form of the linear Boltzmann transport equation on a fully unstructured, tetrahedral mesh using linear discontinuous finite-element spatial differencing in conjunction with diffusion synthetic acceleration of inner iterations. The results obtained indicate that Attila can accurately solve the benchmark problem without spatial homogenization. (authors)

IAEA coordinated research project (CRP) on 'Analytical and experimental benchmark analyses of accelerator driven systems'

International Nuclear Information System (INIS)

Abanades, Alberto; Aliberti, Gerardo; Gohar, Yousry; Talamo, Alberto; Bornos, Victor; Kiyavitskaya, Anna; Carta, Mario; Janczyszyn, Jerzy; Maiorino, Jose; Pyeon, Cheolho; Stanculescu, Alexander; Titarenko, Yury; Westmeier, Wolfram

2008-01-01

In December 2005, the International Atomic Energy Agency (IAEA) has started a Coordinated Research Project (CRP) on 'Analytical and Experimental Benchmark Analyses of Accelerator Driven Systems'. The overall objective of the CRP, performed within the framework of the Technical Working Group on Fast Reactors (TWGFR) of IAEA's Nuclear Energy Department, is to increase the capability of interested Member States in developing and applying advanced reactor technologies in the area of long-lived radioactive waste utilization and transmutation. The specific objective of the CRP is to improve the present understanding of the coupling of an external neutron source (e.g. spallation source) with a multiplicative sub-critical core. The participants are performing computational and experimental benchmark analyses using integrated calculation schemes and simulation methods. The CRP aims at integrating some of the planned experimental demonstration projects of the coupling between a sub-critical core and an external neutron source (e.g. YALINA Booster in Belarus, and Kyoto University's Critical Assembly (KUCA)). The objective of these experimental programs is to validate computational methods, obtain high energy nuclear data, characterize the performance of sub-critical assemblies driven by external sources, and to develop and improve techniques for sub-criticality monitoring. The paper summarizes preliminary results obtained to-date for some of the CRP benchmarks. (authors)

H.B. Robinson-2 pressure vessel benchmark

Energy Technology Data Exchange (ETDEWEB)

Remec, I.; Kam, F.B.K.

1998-02-01

The H. B. Robinson Unit 2 Pressure Vessel Benchmark (HBR-2 benchmark) is described and analyzed in this report. Analysis of the HBR-2 benchmark can be used as partial fulfillment of the requirements for the qualification of the methodology for calculating neutron fluence in pressure vessels, as required by the U.S. Nuclear Regulatory Commission Regulatory Guide DG-1053, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. Section 1 of this report describes the HBR-2 benchmark and provides all the dimensions, material compositions, and neutron source data necessary for the analysis. The measured quantities, to be compared with the calculated values, are the specific activities at the end of fuel cycle 9. The characteristic feature of the HBR-2 benchmark is that it provides measurements on both sides of the pressure vessel: in the surveillance capsule attached to the thermal shield and in the reactor cavity. In section 2, the analysis of the HBR-2 benchmark is described. Calculations with the computer code DORT, based on the discrete-ordinates method, were performed with three multigroup libraries based on ENDF/B-VI: BUGLE-93, SAILOR-95 and BUGLE-96. The average ratio of the calculated-to-measured specific activities (C/M) for the six dosimeters in the surveillance capsule was 0.90 {+-} 0.04 for all three libraries. The average C/Ms for the cavity dosimeters (without neptunium dosimeter) were 0.89 {+-} 0.10, 0.91 {+-} 0.10, and 0.90 {+-} 0.09 for the BUGLE-93, SAILOR-95 and BUGLE-96 libraries, respectively. It is expected that the agreement of the calculations with the measurements, similar to the agreement obtained in this research, should typically be observed when the discrete-ordinates method and ENDF/B-VI libraries are used for the HBR-2 benchmark analysis.

Benchmark for Strategic Performance Improvement.

Science.gov (United States)

Gohlke, Annette

1997-01-01

Explains benchmarking, a total quality management tool used to measure and compare the work processes in a library with those in other libraries to increase library performance. Topics include the main groups of upper management, clients, and staff; critical success factors for each group; and benefits of benchmarking. (Author/LRW)

Application of an integrated PC-based neutronics code system to criticality safety

International Nuclear Information System (INIS)

Briggs, J.B.; Nigg, D.W.

1991-01-01

An integrated system of neutronics and radiation transport software suitable for operation in an IBM PC-class environment has been under development at the Idaho National Engineering Laboratory (INEL) for the past four years. Four modules within the system are particularly useful for criticality safety applications. Using the neutronics portion of the integrated code system, effective neutron multiplication values (k eff values) have been calculated for a variety of benchmark critical experiments for metal systems (Plutonium and Uranium), Aqueous Systems (Plutonium and Uranium) and LWR fuel rod arrays. A description of the codes and methods used in the analysis and the results of the benchmark critical experiments are presented in this paper. In general, excellent agreement was found between calculated and experimental results. (Author)

Criticality calculation method for mixer-settlers

International Nuclear Information System (INIS)

Gonda, Kozo; Aoyagi, Haruki; Nakano, Ko; Kamikawa, Hiroshi.

1980-01-01

A new criticality calculation code MACPEX has been developed to evaluate and manage the criticality of the process in the extractor of mixer-settler type. MACPEX can perform the combined calculation with the PUREX process calculation code MIXSET, to get the neutron flux and the effective multiplication constant in the mixer-settlers. MACPEX solves one-dimensional diffusion equation by the explicit difference method and the standard source-iteration technique. The characteristics of MACPEX are as follows. 1) Group constants of 4 energy groups for the 239 Pu-H 2 O solution, water, polyethylene and SUS 28 are provided. 2) The group constants of the 239 Pu-H 2 O solution are given by the functional formulae of the plutonium concentration, which is less than 50 g/l. 3) Two boundary conditions of the vacuum condition and the reflective condition are available in this code. 4) The geometrical bucklings can be calculated for a certain energy group and/or region by using the three dimentional neutron flux profiles obtained by CITATION. 5) The buckling correction search can be carried out in order to get a desired k sub(eff). (author)

Benchmarking and Performance Management

Directory of Open Access Journals (Sweden)

Adrian TANTAU

2010-12-01

Full Text Available The relevance of the chosen topic is explained by the meaning of the firm efficiency concept - the firm efficiency means the revealed performance (how well the firm performs in the actual market environment given the basic characteristics of the firms and their markets that are expected to drive their profitability (firm size, market power etc.. This complex and relative performance could be due to such things as product innovation, management quality, work organization, some other factors can be a cause even if they are not directly observed by the researcher. The critical need for the management individuals/group to continuously improve their firm/company’s efficiency and effectiveness, the need for the managers to know which are the success factors and the competitiveness determinants determine consequently, what performance measures are most critical in determining their firm’s overall success. Benchmarking, when done properly, can accurately identify both successful companies and the underlying reasons for their success. Innovation and benchmarking firm level performance are critical interdependent activities. Firm level variables, used to infer performance, are often interdependent due to operational reasons. Hence, the managers need to take the dependencies among these variables into account when forecasting and benchmarking performance. This paper studies firm level performance using financial ratio and other type of profitability measures. It uses econometric models to describe and then propose a method to forecast and benchmark performance.

Benchmarking Benchmarks

NARCIS (Netherlands)

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.

CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

International Nuclear Information System (INIS)

C.E. Sanders

2005-01-01

This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for the current design of the CHF and may not reflect the ongoing design evolution of the facility

Verification of HELIOS-MASTER system through benchmark of critical experiments

International Nuclear Information System (INIS)

Kim, H. Y.; Kim, K. Y.; Cho, B. O.; Lee, C. C.; Zee, S. O.

1999-01-01

The HELIOS-MASTER code system is verified through the benchmark of the critical experiments that were performed by RRC 'Kurchatov Institute' with water-moderated hexagonally pitched lattices of highly enriched Uranium fuel rods (80w/o). We also used the same input by using the MCNP code that was described in the evaluation report, and compared our results with those of the evaluation report. HELIOS, developed by Scandpower A/S, is a two-dimensional transport program for the generation of group cross-sections, and MASTER, developed by KAERI, is a three-dimensional nuclear design and analysis code based on the two-group diffusion theory. It solves neutronics model with the AFEN (Analytic Function Expansion Nodal) method for hexagonal geometry. The results show that the HELIOS-MASTER code system is fast and accurate enough to be used as nuclear core analysis tool for hexagonal geometry

Calculational study for criticality safety data of fissionable actinides

International Nuclear Information System (INIS)

Nojiri, Ichiro; Fukasaku, Yasuhiro.

1997-01-01

This study has been carried out to obtain basic criticality safety characteristics of minor actinides nuclides. Criticality safety data of minor actinides nuclides have been surveyed through public literatures. Critical mass of seven nuclides, Np-237, Am-241, Am-242m, Am-243, Cm-243, Cm-244 and Cm-245, have been calculated by using two code systems of criticality safety analysis, SCALE-4 and MCNP4A, under some material and reflector conditions. Some applicable cross-section libraries have been used for each code systems. Calculated data have been compared with each other and with published data. The results of this comparison shows that there is no discrepancy within the computational codes and the calculated data is strongly depend on the cross-section library. (author)

Analysis of radially heterogeneous ZPPR-13A benchmark for investigating the spatial dependence of the calculated-to-experiment ratio for control rod worths

International Nuclear Information System (INIS)

Mahalakshmi, B.; Mohanakrishnan, P.

1993-01-01

Investigation were performed on the ZPPR-13A critical assembly to determine the cause of the radial variation of the calculated-to-experimental (C/E) ratio for control rod worth in large heterogeneous cores. The effects of errors in cross section, mesh size, group condensation, transport, and modeling were studied by studied by using two- and three-dimensional diffusion calculations and three-dimensional transport calculations. In that process, the cross-section set and the calculation scheme that are being used for fast reactor design in India have been revalidated. The cross-section set was found to yield satisfactory results. Three-dimensional calculations with adjusted and unadjusted cross sections confirmed that the error in cross sections was largely responsible for the radial dependence of the C/E ratios. The contributions from group condensation and mesh size errors were < 2%, and from modeling errors and transport correction, < 1%. The effect of these errors is insignificant when compared with the effect of the cross-section error. The analysis also showed that even without the adjustment in diffusion coefficient suggested in earlier studies, a satisfactory prediction is found, at least for this benchmark. The diffusion-to-transport correction for control rod worth was found to be -7%

Helium generation reaction rates for 6Li and 10B in benchmark facilities

International Nuclear Information System (INIS)

Farrar, Harry IV; Oliver, B.M.; Lippincott, E.P.

1980-01-01

The helium generation rates for 10 B and 6 Li have been measured in two benchmark reactor facilities having neutron spectra similar to those found in a breeder reactor. The irradiations took place in the Coupled Fast Reactivity Measurements Facility (CFRMF) and in the 10% enriched 235 U critical assembly, BIG-10. The helium reaction rates were obtained by precise high-sensitivity gas mass spectrometric analyses of the helium content of numerous small samples. Comparison of these reaction rates with other reaction rates measured in the same facilities, and with rates calculated from published cross sections and from best estimates of the neutron spectral shapes, indicate significant discrepancies in the calculated values. Additional irradiations in other benchmark facilities have been undertaken to better determine the energy ranges where the discrepancies lie

Benchmarking quantum mechanical calculations with experimental NMR chemical shifts of 2-HADNT

Science.gov (United States)

Liu, Yuemin; Junk, Thomas; Liu, Yucheng; Tzeng, Nianfeng; Perkins, Richard

2015-04-01

In this study, both GIAO-DFT and GIAO-MP2 calculations of nuclear magnetic resonance (NMR) spectra were benchmarked with experimental chemical shifts. The experimental chemical shifts were determined experimentally for carbon-13 (C-13) of seven carbon atoms for the TNT degradation product 2-hydroxylamino-4,6-dinitrotoluene (2-HADNT). Quantum mechanics GIAO calculations were implemented using Becke-3-Lee-Yang-Parr (B3LYP) and other six hybrid DFT methods (Becke-1-Lee-Yang-Parr (B1LYP), Becke-half-and-half-Lee-Yang-Parr (BH and HLYP), Cohen-Handy-3-Lee-Yang-Parr (O3LYP), Coulomb-attenuating-B3LYP (CAM-B3LYP), modified-Perdew-Wang-91-Lee-Yang-Parr (mPW1LYP), and Xu-3-Lee-Yang-Parr (X3LYP)) which use the same correlation functional LYP. Calculation results showed that the GIAO-MP2 method gives the most accurate chemical shift values, and O3LYP method provides the best prediction of chemical shifts among the B3LYP and other five DFT methods. Three types of atomic partial charges, Mulliken (MK), electrostatic potential (ESP), and natural bond orbital (NBO), were also calculated using MP2/aug-cc-pVDZ method. A reasonable correlation was discovered between NBO partial charges and experimental chemical shifts of carbon-13 (C-13).

Integral parameters for the Godiva benchmark calculated by using theoretical and adjusted fission spectra of 235U

International Nuclear Information System (INIS)

Caldeira, A.D.

1987-05-01

The theoretical and adjusted Watt spectrum representations for 235 U are used as weighting functions to calculate K eff and θ f 28 /θ f 25 for the benchmark Godiva. The results obtained show that the values of K eff and θ f 28 /θ f 25 are not affected by spectrum form change. (author) [pt

Impact of MCNP unresolved resonance probability-table treatment on uranium and plutonium benchmarks

International Nuclear Information System (INIS)

Mosteller, R.D.; Little, R.C.

1998-01-01

Versions of MCNP up through and including 4B have not accurately modeled neutron self-shielding effects in the unresolved resonance energy region. Recently, a probability-table treatment has been incorporated into a developmental version of MCNP. This paper presents MCNP results for a variety of uranium and plutonium critical benchmarks, calculated with and without the probability-table treatment

Dynamic benchmarking of simulation codes

International Nuclear Information System (INIS)

Henry, R.E.; Paik, C.Y.; Hauser, G.M.

1996-01-01

output includes a plot of the MAAP calculation and the plant data. For the large integral experiments, a major part, but not all of the MAAP code is needed. These use an experiment specific benchmark routine that includes all of the information and boundary conditions for performing the calculation, as well as the information of which parts of MAAP are unnecessary and can be 'bypassed'. Lastly, the separate effects tests only require a few MAAP routines. These are exercised through their own specific benchmark routine that includes the experiment specific information and boundary conditions. This benchmark routine calls the appropriate MAAP routines from the source code, performs the calculations, including integration where necessary and provide the comparison between the MAAP calculation and the experimental observations. (author)

3-D neutron transport benchmarks

International Nuclear Information System (INIS)

Takeda, T.; Ikeda, H.

1991-03-01

A set of 3-D neutron transport benchmark problems proposed by the Osaka University to NEACRP in 1988 has been calculated by many participants and the corresponding results are summarized in this report. The results of K eff , control rod worth and region-averaged fluxes for the four proposed core models, calculated by using various 3-D transport codes are compared and discussed. The calculational methods used were: Monte Carlo, Discrete Ordinates (Sn), Spherical Harmonics (Pn), Nodal Transport and others. The solutions of the four core models are quite useful as benchmarks for checking the validity of 3-D neutron transport codes

FRIB driver linac vacuum model and benchmarks

CERN Document Server

Durickovic, Bojan; Kersevan, Roberto; Machicoane, Guillaume

2014-01-01

The Facility for Rare Isotope Beams (FRIB) is a superconducting heavy-ion linear accelerator that is to produce rare isotopes far from stability for low energy nuclear science. In order to achieve this, its driver linac needs to achieve a very high beam current (up to 400 kW beam power), and this requirement makes vacuum levels of critical importance. Vacuum calculations have been carried out to verify that the vacuum system design meets the requirements. The modeling procedure was benchmarked by comparing models of an existing facility against measurements. In this paper, we present an overview of the methods used for FRIB vacuum calculations and simulation results for some interesting sections of the accelerator. (C) 2013 Elsevier Ltd. All rights reserved.

Application of MCNP in the criticality calculation for reactors

International Nuclear Information System (INIS)

Zhong Zhaopeng; Shi Gong; Hu Yongming

2003-01-01

The criticality calculation is carried out with 3-D Monte Carlo code (MCNP). The author focuses on the introduction of modelling of the core and reflector. The core description is simplified by using repetition structure function of MCNP. k eff in different control rods positions are calculated for the case of JRR3, and the results is consistent with that of the reference. This work shows that MCNP is applicable for reactor criticality calculation

A review of the current state-of-the-art methodology for handling bias and uncertainty in performing criticality safety evaluations. Final report

International Nuclear Information System (INIS)

Disney, R.K.

1994-10-01

The methodology for handling bias and uncertainty when calculational methods are used in criticality safety evaluations (CSE's) is a rapidly evolving technology. The changes in the methodology are driven by a number of factors. One factor responsible for changes in the methodology for handling bias and uncertainty in CSE's within the overview of the US Department of Energy (DOE) is a shift in the overview function from a ''site'' perception to a more uniform or ''national'' perception. Other causes for change or improvement in the methodology for handling calculational bias and uncertainty are; (1) an increased demand for benchmark criticals data to expand the area (range) of applicability of existing data, (2) a demand for new data to supplement existing benchmark criticals data, (3) the increased reliance on (or need for) computational benchmarks which supplement (or replace) experimental measurements in critical assemblies, and (4) an increased demand for benchmark data applicable to the expanded range of conditions and configurations encountered in DOE site restoration and remediation

Start-up of a cold loop in a VVER-440, the 7{sup th} AER benchmark calculation with HEXTRAN-SMABRE-PORFLO

Energy Technology Data Exchange (ETDEWEB)

Hovi, Ville; Taivassalo, Veikko; Haemaelaeinen, Anitta; Raety, Hanna; Syrjaelahti, Elina [VTT Technical Research Centre of Finland Ltd, VTT (Finland)

2017-09-15

The 7{sup th} dynamic AER benchmark is the first in which three-dimensional thermal hydraulics codes are supposed to be applied. The aim is to get a more precise core inlet temperature profile than the sector temperatures available typically with system codes. The benchmark consists of a start-up of the sixth, isolated loop in a VVER-440 plant. The isolated loop initially contains cold water without boric acid and the start-up leads to a somewhat asymmetrical core power increase due to feedbacks in the core. In this study, the 7{sup th} AER benchmark is calculated with the three-dimensional nodal reactor dynamics code HEXTRAN-SMABRE coupled with the porous computational fluid dynamics code PORFLO. These three codes are developed at VTT. A novel two-way coupled simulation of the 7{sup th} AER benchmark was performed successfully demonstrating the feasibility and advantages of the new reactor analysis framework. The modelling issues for this benchmark are reported and some evaluation against the previously reported comparisons between the system codes is provided.

239Pu prompt fission neutron spectra impact on a set of criticality and experimental reactor benchmarks

International Nuclear Information System (INIS)

Peneliau, Y.; Litaize, O.; Archier, P.; De Saint Jean, C.

2014-01-01

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 239 Pu cross sections and other nuclear data is crucial issue in order to reduce these sources of uncertainty. The Prompt Fission Neutron Spectra (PFNS) for 239 Pu 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

Power reactor pressure vessel benchmarks

International Nuclear Information System (INIS)

Rahn, F.J.

1978-01-01

A review is given of the current status of experimental and calculational benchmarks for use in understanding the radiation embrittlement effects in the pressure vessels of operating light water power reactors. The requirements of such benchmarks for application to pressure vessel dosimetry are stated. Recent developments in active and passive neutron detectors sensitive in the ranges of importance to embrittlement studies are summarized and recommendations for improvements in the benchmark are made. (author)

Vver-1000 Mox core computational benchmark

International Nuclear Information System (INIS)

2006-01-01

The NEA Nuclear Science Committee has established an Expert Group that deals with the status and trends of reactor physics, fuel performance and fuel cycle issues related to disposing of weapons-grade plutonium in mixed-oxide fuel. The objectives of the group are to provide NEA member countries with up-to-date information on, and to develop consensus regarding, core and fuel cycle issues associated with burning weapons-grade plutonium in thermal water reactors (PWR, BWR, VVER-1000, CANDU) and fast reactors (BN-600). These issues concern core physics, fuel performance and reliability, and the capability and flexibility of thermal water reactors and fast reactors to dispose of weapons-grade plutonium in standard fuel cycles. The activities of the NEA Expert Group on Reactor-based Plutonium Disposition are carried out in close co-operation (jointly, in most cases) with the NEA Working Party on Scientific Issues in Reactor Systems (WPRS). A prominent part of these activities include benchmark studies. At the time of preparation of this report, the following benchmarks were completed or in progress: VENUS-2 MOX Core Benchmarks: carried out jointly with the WPRS (formerly the WPPR) (completed); VVER-1000 LEU and MOX Benchmark (completed); KRITZ-2 Benchmarks: carried out jointly with the WPRS (formerly the WPPR) (completed); Hollow and Solid MOX Fuel Behaviour Benchmark (completed); PRIMO MOX Fuel Performance Benchmark (ongoing); VENUS-2 MOX-fuelled Reactor Dosimetry Calculation (ongoing); VVER-1000 In-core Self-powered Neutron Detector Calculational Benchmark (started); MOX Fuel Rod Behaviour in Fast Power Pulse Conditions (started); Benchmark on the VENUS Plutonium Recycling Experiments Configuration 7 (started). This report describes the detailed results of the benchmark investigating the physics of a whole VVER-1000 reactor core using two-thirds low-enriched uranium (LEU) and one-third MOX fuel. It contributes to the computer code certification process and to the

Definition and Analysis of Heavy Water Reactor Benchmarks for Testing New Wims-D Libraries

International Nuclear Information System (INIS)

Leszczynski, Francisco

2000-01-01

This work is part of the IAEA-WIMS Library Update Project (WLUP). A group of heavy water reactor benchmarks have been selected for testing new WIMS-D libraries, including calculations with WIMSD5B program and the analysis of results.These benchmarks cover a wide variety of reactors and conditions, from fresh fuels to high burnup, and from natural to enriched uranium.Besides, each benchmark includes variations in lattice pitch and in coolants (normally heavy water and void).Multiplication factors with critical experimental bucklings and other parameters are calculated and compared with experimental reference values.The WIMS libraries used for the calculations were generated with basic data from JEF-2.2 Rev.3 (JEF) and ENDF/B-VI iNReleaseln 5 (E6) Results obtained with WIMS-86 (W86) library, included with WIMSD5B package, from Windfrith, UK with adjusted data, are included also, for showing the improvements obtained with the new -not adjusted- libraries.The calculations with WIMSD5B were made with two methods (input program options): PIJ (two-dimension collision probability method) and DSN (one-dimension Sn method, with homogenization of materials by ring).The general conclusions are: the library based on JEF data and the DSN meted give the best results, that in average are acceptable

Definition and Analysis of Heavy Water Reactor Benchmarks for Testing New Wims-D Libraries; Definicion y Analisis de Benchmarks de Reactores de Agua Pesada para Pruebas de Nuevas Bibliotecas de Datos Wims-D

Energy Technology Data Exchange (ETDEWEB)

Leszczynski, Francisco [Comision Nacional de Energia Atomica, Centro Atomico Bariloche (Argentina)

2000-07-01

This work is part of the IAEA-WIMS Library Update Project (WLUP). A group of heavy water reactor benchmarks have been selected for testing new WIMS-D libraries, including calculations with WIMSD5B program and the analysis of results.These benchmarks cover a wide variety of reactors and conditions, from fresh fuels to high burnup, and from natural to enriched uranium.Besides, each benchmark includes variations in lattice pitch and in coolants (normally heavy water and void).Multiplication factors with critical experimental bucklings and other parameters are calculated and compared with experimental reference values.The WIMS libraries used for the calculations were generated with basic data from JEF-2.2 Rev.3 (JEF) and ENDF/B-VI iNReleaseln 5 (E6) Results obtained with WIMS-86 (W86) library, included with WIMSD5B package, from Windfrith, UK with adjusted data, are included also, for showing the improvements obtained with the new -not adjusted- libraries.The calculations with WIMSD5B were made with two methods (input program options): PIJ (two-dimension collision probability method) and DSN (one-dimension Sn method, with homogenization of materials by ring).The general conclusions are: the library based on JEF data and the DSN meted give the best results, that in average are acceptable.

CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

Energy Technology Data Exchange (ETDEWEB)

C.E. Sanders

2005-04-07

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 impact and applicability of critical experiment evaluations

Energy Technology Data Exchange (ETDEWEB)

Brewer, R. [Los Alamos National Lab., NM (United States)

1997-06-01

This paper very briefly describes a project to evaluate previously performed critical experiments. The evaluation is intended for use by criticality safety engineers to verify calculations, and may also be used to identify data which need further investigation. The evaluation process is briefly outlined; the accepted benchmark critical experiments will be used as a standard for verification and validation. The end result of the project will be a comprehensive reference document.

Research and design calculation of multipurpose critical assembly using moderated light water and low enriched fuel from 1.6 to 5.0% U-235

International Nuclear Information System (INIS)

Nguyen Kien Cuong; Vo Doan Hai Dang; Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Minh Tuan; Nguyen Manh Hung; Pham Quang Huy; Tran Quoc Duong; Tran Tri Vien

2015-01-01

Basing on the idea in ??using fuel of nuclear power plants such as PWR (AP-1000) and VVER-1000 with light water as moderation, design calculation of critical assembly was performed to confirm the possibility of using these fuels. Designed critical assembly has simple structure consisting of low enriched fuel from 1.6% to 5% U-235; water has functions as cooling, biological protection and control. Critical assembly is operated at nominal power 100 W with fuel pitch about 2.0 cm. Applications of the critical assembly are quite abundant in basic research, education and training with low investment cost compare with research reactor and easy in operation. So critical assembly can be used for university or training centre for nuclear engineering training. Main objectives of the project are: design calculation in neutronics, thermal hydraulics and safety analysis for critical configuration benchmarks using low enriched fuel; design in mechanical and auxiliary systems for critical assembly; determine technical specifications and estimate construction, installation cost of critical assembly. The process of design, fabrication, installation and construction of critical assembly will be considered with different implementation phases and localization capabilities in installation of critical assembly is highly feasibility. Cost estimation of construction and installation of critical assembly was implemented and showed that investment cost for critical assembly is much lower than research reactor and most of components, systems of critical assembly can be localized with current technique quality of the country. (author)

Benchmark analysis of SPERT-IV reactor with Monte Carlo code MVP

International Nuclear Information System (INIS)

Motalab, M.A.; Mahmood, M.S.; Khan, M.J.H.; Badrun, N.H.; Lyric, Z.I.; Altaf, M.H.

2014-01-01

Highlights: • MVP was used for SPERT-IV core modeling. • Neutronics analysis of SPERT-IV reactor was performed. • Calculation performed to estimate critical rod height, excess reactivity. • Neutron flux, time integrated neutron flux and Cd-ratio also calculated. • Calculated values agree with experimental data. - Abstract: The benchmark experiment of the SPERT-IV D-12/25 reactor core has been analyzed with the Monte Carlo code MVP using the cross-section libraries based on JENDL-3.3. The MVP simulation was performed for the clean and cold core. The estimated values of K eff at the experimental critical rod height and the core excess reactivity were within 5% with the experimental data. Thermal neutron flux profiles at different vertical and horizontal positions of the core were also estimated. Cadmium Ratio at different point of the core was also estimated. All estimated results have been compared with the experimental results. Generally good agreement has been found between experimentally determined and the calculated results

Interim results of the sixth three-dimensional AER dynamic benchmark problem calculation. Solution of problem with DYN3D and RELAP5-3D codes

International Nuclear Information System (INIS)

Hadek, J.; Kral, P.; Macek, J.

2001-01-01

The paper gives a brief survey of the 6 th three-dimensional AER dynamic benchmark calculation results received with the codes DYN3D and RELAPS-3D at NRI Rez. This benchmark was defined at the 10 th AER Symposium. Its initiating event is a double ended break in the steam line of steam generator No. I in a WWER-440/213 plant at the end of the first fuel cycle and in hot full power conditions. Stationary and burnup calculations as well as tuning of initial state before the transient were performed with the code DYN3D. Transient calculations were made with the system code RELAPS-3D.The KASSETA library was used for the generation of reactor core neutronic parameters. The detailed six loops model of NPP Dukovany was adopted for the 6 th AER dynamic benchmark purposes. The RELAPS-3D full core neutronic model was connected with seven coolant channels thermal-hydraulic model of the core (Authors)

Review for session K - benchmarks

International Nuclear Information System (INIS)

McCracken, A.K.

1980-01-01

Eight of the papers to be considered in Session K are directly concerned, at least in part, with the Pool Critical Assembly (P.C.A.) benchmark at Oak Ridge. The remaining seven papers in this session, the subject of this review, are concerned with a variety of topics related to the general theme of Benchmarks and will be considered individually

Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

Energy Technology Data Exchange (ETDEWEB)

Chiang, Min-Han; Wang, Jui-Yu [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Sheu, Rong-Jiun, E-mail: rjsheu@mx.nthu.edu.tw [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Liu, Yen-Wan Hsueh [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China)

2014-05-01

The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects.

Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

International Nuclear Information System (INIS)

Chiang, Min-Han; Wang, Jui-Yu; Sheu, Rong-Jiun; Liu, Yen-Wan Hsueh

2014-01-01

The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects

Benchmark calculation of APOLLO-2 and SLAROM-UF in a fast reactor lattice

International Nuclear Information System (INIS)

Hazama, T.

2009-07-01

A lattice cell benchmark calculation is carried out for APOLLO2 and SLAROM-UF on the infinite lattice of a simple pin cell featuring a fast reactor. The accuracy in k-infinity and reaction rates is investigated in their reference and standard level calculations. In the 1. reference level calculation, APOLLO2 and SLAROM-UF agree with the reference value of k-infinity obtained by a continuous energy Monte Carlo calculation within 50 pcm. However, larger errors are observed in a particular reaction rate and energy range. The major problem common to both codes is in the cross section library of 239 Pu in the unresolved energy range. In the 2. reference level calculation, which is based on the ECCO 1968 group structure, both results of k-infinity agree with the reference value within 100 pcm. The resonance overlap effect is observed by several percents in cross sections of heavy nuclides. In the standard level calculation based on the APOLLO2 library creation methodology, a discrepancy appears by more than 300 pcm. A restriction is revealed in APOLLO2. Its standard cross section library does not have a sufficiently small background cross section to evaluate the self shielding effect on 56 Fe cross sections. The restriction can be removed by introducing the mixture self-shielding treatment recently introduced to APOLLO2. SLAROM-UF original standard level calculation based on the JFS-3 library creation methodology is the best among the standard level calculations. Improvement from the SLAROM-UF standard level calculation is achieved mainly by use of a proper weight function for light or intermediate nuclides. (author)

Evaluation of AMPX-KENO benchmark calculations for high-density spent fuel storage racks

International Nuclear Information System (INIS)

Turner, S.E.; Gurley, M.K.

1981-01-01

The AMPX-KENO computer code package is commonly used to evaluate criticality in high-density spent fuel storage rack designs. Consequently, it is important to know the reliability that can be placed on such calculations and whether or not the results are conservative. This paper evaluates a series of AMPX-KENO calculations which have been made on selected critical experiments. The results are compared with similar analyses reported in the literature by the Oak Ridge National Laboratory and BandW. 8 refs

TRIGA criticality experiment for testing burn-up calculations

International Nuclear Information System (INIS)

Persic, Andreja; Ravnik, Matjaz; Zagar, Tomaz

1999-01-01

A criticality experiment with partly burned TRIGA fuel is described. 20 wt % enriched standard TRIGA fuel elements initially containing 12 wt % U are used. Their average burn-up is 1.4 MWd. Fuel element burn-up is calculated in 2-D four group diffusion approximation using TRIGLAV code. The burn-up of several fuel elements is also measured by reactivity method. The excess reactivity of several critical and subcritical core configurations is measured. Two core configurations contain the same fuel elements in the same arrangement as were used in the fresh TRIGA fuel criticality experiment performed in 1991. The results of the experiment may be applied for testing the computer codes used for fuel burn-up calculations. (author)

Calculation of Critical Temperatures by Empirical Formulae

Directory of Open Access Journals (Sweden)

Trzaska J.

2016-06-01

Full Text Available The paper presents formulas used to calculate critical temperatures of structural steels. Equations that allow calculating temperatures Ac1, Ac3, Ms and Bs were elaborated based on the chemical composition of steel. To elaborate the equations the multiple regression method was used. Particular attention was paid to the collection of experimental data which was required to calculate regression coefficients, including preparation of data for calculation. The empirical data set included more than 500 chemical compositions of structural steel and has been prepared based on information available in literature on the subject.

OECD/NEA International Benchmark exercises: Validation of CFD codes applied nuclear industry; OECD/NEA internatiion Benchmark exercices: La validacion de los codigos CFD aplicados a la industria nuclear

Energy Technology Data Exchange (ETDEWEB)

Pena-Monferrer, C.; Miquel veyrat, A.; Munoz-Cobo, J. L.; Chiva Vicent, S.

2016-08-01

In the recent years, due, among others, the slowing down of the nuclear industry, investment in the development and validation of CFD codes, applied specifically to the problems of the nuclear industry has been seriously hampered. Thus the International Benchmark Exercise (IBE) sponsored by the OECD/NEA have been fundamental to analyze the use of CFD codes in the nuclear industry, because although these codes are mature in many fields, still exist doubts about them in critical aspects of thermohydraulic calculations, even in single-phase scenarios. The Polytechnic University of Valencia (UPV) and the Universitat Jaume I (UJI), sponsored by the Nuclear Safety Council (CSN), have actively participated in all benchmark's proposed by NEA, as in the expert meetings,. In this paper, a summary of participation in the various IBE will be held, describing the benchmark itself, the CFD model created for it, and the main conclusions. (Author)

Benchmark experiments to test plutonium and stainless steel cross sections. Topical report

International Nuclear Information System (INIS)

Jenquin, U.P.; Bierman, S.R.

1978-06-01

The Nuclear Regulatory Commission (NRC) commissioned Battelle, Pacific Northwest Laboratory (PNL) to ascertain the accuracy of the neutron cross sections for the isotopes of plutonium and the constituents of stainless steel and determine if improvements can be made in application to criticality safety analysis. NRC's particular area of interest is in the transportation of light-water reactor spent fuel assemblies. The project was divided into two tasks. The first task was to define a set of integral experimental measurements (benchmarks). The second task is to use these benchmarks in neutronics calculations such that the accuracy of ENDF/B-IV plutonium and stainless steel cross sections can be assessed. The results of the first task are given in this report. A set of integral experiments most pertinent to testing the cross sections has been identified and the code input data for calculating each experiment has been developed

Dose Rate Experiment at JET for Benchmarking the Calculation Direct One Step Method

International Nuclear Information System (INIS)

Angelone, M.; Petrizzi, L.; Pillon, M.; Villari, R.; Popovichev, S.

2006-01-01

Neutrons produced by D-D and D-T plasmas induce the activation of tokamak materials and of components. The development of reliable methods to assess dose rates is a key issue for maintenance and operating nuclear machines, in normal and off-normal conditions. In the frame of the EFDA Fusion Technology work programme, a computational tool based upon MCNP Monte Carlo code has been developed to predict the dose rate after shutdown: it is called Direct One Step Method (D1S). The D1S is an innovative approach in which the decay gammas are coupled to the neutrons as in the prompt case and they are transported in one single step in the same run. Benchmarking of this new tool with experimental data taken in a complex geometry like that of a tokamak is a fundamental step to test the reliability of the D1S method. A dedicated benchmark experiment was proposed for the 2005-2006 experimental campaign of JET. Two irradiation positions have been selected for the benchmark: one inner position inside the vessel, not far from the plasma, called the 2 upper irradiation end (IE2), where neutron fluence is relatively high. The second position is just outside a vertical port in an external position (EX). Here the neutron flux is lower and the dose rate to be measured is not very far from the residual background. Passive detectors are used for in-vessel measurements: the high sensitivity Thermo Luminescent Dosimeters (TLDs) GR-200A (natural LiF), which ensure measurements down to environmental dose level. An active detector of Geiger-Muller (GM) type is used for out of vessel dose rate measurement. Before their use the detectors were calibrated in a secondary gamma-ray standard (Cs-137 and Co-60) facility in term of air-kerma. The background measurement was carried-out in the period July -September 2005 in the outside position EX using the GM tube and in September 2005 inside the vacuum vessel using TLD detectors located in the 2 Upper irradiation end IE2. In the present work

Parametric Criticality Safety Calculations for Arrays of TRU Waste Containers

Energy Technology Data Exchange (ETDEWEB)

Gough, Sean T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-10-26

The Nuclear Criticality Safety Division (NCSD) has performed criticality safety calculations for finite and infinite arrays of transuranic (TRU) waste containers. The results of these analyses may be applied in any technical area onsite (e.g., TA-54, TA-55, etc.), as long as the assumptions herein are met. These calculations are designed to update the existing reference calculations for waste arrays documented in Reference 1, in order to meet current guidance on calculational methodology.

Accelerator shielding benchmark problems

International Nuclear Information System (INIS)

Hirayama, H.; Ban, S.; Nakamura, T.

1993-01-01

Accelerator shielding benchmark problems prepared by Working Group of Accelerator Shielding in the Research Committee on Radiation Behavior in the Atomic Energy Society of Japan were compiled by Radiation Safety Control Center of National Laboratory for High Energy Physics. Twenty-five accelerator shielding benchmark problems are presented for evaluating the calculational algorithm, the accuracy of computer codes and the nuclear data used in codes. (author)

Shielding benchmark problems

International Nuclear Information System (INIS)

Tanaka, Shun-ichi; Sasamoto, Nobuo; Oka, Yoshiaki; Kawai, Masayoshi; Nakazawa, Masaharu.

1978-09-01

Shielding benchmark problems were prepared by the Working Group of Assessment of Shielding Experiments in the Research Comittee on Shielding Design of the Atomic Energy Society of Japan, and compiled by the Shielding Laboratory of Japan Atomic Energy Research Institute. Twenty-one kinds of shielding benchmark problems are presented for evaluating the calculational algorithm and the accuracy of computer codes based on the discrete ordinates method and the Monte Carlo method and for evaluating the nuclear data used in the codes. (author)

Assessing reactor physics codes capabilities to simulate fast reactors on the example of the BN-600 benchmark

Energy Technology Data Exchange (ETDEWEB)

Ivanov, Vladimir [Scientific and Engineering Centre for Nuclear and Radiation Safety (SES NRS), Moscow (Russian Federation); Bousquet, Jeremy [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

2016-11-15

This work aims to assess the capabilities of reactor physics codes (initially validated for thermal reactors) to simulate fast sodium cooled reactors. The BFS-62-3A critical experiment from the BN-600 Hybrid Core Benchmark Analyses was chosen for the investigation. Monte-Carlo codes (KENO from SCALE and SERPENT 2.1.23) and the deterministic diffusion code DYN3D-MG are applied to calculate the neutronic parameters. It was found that the multiplication factor and reactivity effects calculated by KENO and SERPENT using the ENDF/B-VII.0 continuous energy library are in a good agreement with each other and with the measured benchmark values. Few-groups macroscopic cross sections, required for DYN3D-MG, were prepared in applying different methods implemented in SCALE and SERPENT. The DYN3D-MG results of a simplified benchmark show reasonable agreement with results from Monte-Carlo calculations and measured values. The former results are used to justify DYN3D-MG implementation for sodium cooled fast reactors coupled deterministic analysis.

Uncertainties in criticality analysis which affect the storage and transportation of LWR fuel

International Nuclear Information System (INIS)

Napolitani, D.G.

1989-01-01

Satisfying the design criteria for subcriticality with uncertainties affects: the capacity of LWR storage arrays, maximum allowable enrichment, minimum allowable burnup and economics of various storage options. There are uncertainties due to: calculational method, data libraries, geometric limitations, modelling bias, the number and quality of benchmarks performed and mechanical uncertainties in the array. Yankee Atomic Electric Co. (YAEC) has developed and benchmarked methods to handle: high density storage rack designs, pin consolidation, low density moderation and burnup credit. The uncertainties associated with such criticality analysis are quantified on the basis of clean criticals, power reactor criticals and intercomparison of independent analysis methods

Benchmark thermal-hydraulic analysis with the Agathe Hex 37-rod bundle

International Nuclear Information System (INIS)

Barroyer, P.; Hudina, M.; Huggenberger, M.

1981-09-01

Different computer codes are compared, in prediction performance, based on the AGATHE HEX 37-rod bundle experimental results. The compilation of all available calculation results allows a critical assessment of the codes. For the time being, it is concluded which codes are best suited for gas cooled fuel element design purposes. Based on the positive aspects of these cooperative Benchmark exercises, an attempt is made to define a computer code verification procedure. (Auth.)

Criticality benchmark results for the ENDF60 library with MCNP trademark

International Nuclear Information System (INIS)

Keen, N.D.; Frankle, S.C.; MacFarlane, R.E.

1995-01-01

The continuous-energy neutron data library ENDF60, for use with the Monte Carlo N-Particle radiation transport code MCNP4A, was released in the fall of 1994. The ENDF60 library is comprised of 124 nuclide data files based on the ENDF/B-VI (B-VI) evaluations through Release 2. Fifty-two percent of these B-VI evaluations are translations from ENDF/B-V (B-V). The remaining forty-eight percent are new evaluations which have sometimes changed significantly. Among these changes are greatly increased use of isotopic evaluations, more extensive resonance-parameter evaluations, and energy-angle correlated distributions for secondary particles. In particular, the upper energy limit for the resolved resonance region of 235 U, 238 U and 239 Pu has been extended from 0.082, 4.0, and 0.301 keV to 2..25, 10.0, and 2.5 keV respectively. As regulatory oversight has advanced and performing critical experiments has become more difficult, there has been an increased reliance on computational methods. For the criticality safety community, the performance of the combined transport code and data library is of interest. The purpose of this abstract is to provide benchmarking results to aid the user in determining the best data library for their application

A Heterogeneous Medium Analytical Benchmark

International Nuclear Information System (INIS)

Ganapol, B.D.

1999-01-01

A benchmark, called benchmark BLUE, has been developed for one-group neutral particle (neutron or photon) transport in a one-dimensional sub-critical heterogeneous plane parallel medium with surface illumination. General anisotropic scattering is accommodated through the Green's Function Method (GFM). Numerical Fourier transform inversion is used to generate the required Green's functions which are kernels to coupled integral equations that give the exiting angular fluxes. The interior scalar flux is then obtained through quadrature. A compound iterative procedure for quadrature order and slab surface source convergence provides highly accurate benchmark qualities (4- to 5- places of accuracy) results

Availability of Neutronics Benchmarks in the ICSBEP and IRPhEP Handbooks for Computational Tools Testing

Energy Technology Data Exchange (ETDEWEB)

Bess, John D.; Briggs, J. Blair; Ivanova, Tatiana; Hill, Ian; Gulliford, Jim

2017-02-01

In the past several decades, numerous experiments have been performed worldwide to support reactor operations, measurements, design, and nuclear safety. Those experiments represent an extensive international investment in infrastructure, expertise, and cost, representing significantly valuable resources of data supporting past, current, and future research activities. Those valuable assets represent the basis for recording, development, and validation of our nuclear methods and integral nuclear data [1]. The loss of these experimental data, which has occurred all too much in the recent years, is tragic. The high cost to repeat many of these measurements can be prohibitive, if not impossible, to surmount. Two international projects were developed, and are under the direction of the Organisation for Co-operation and Development Nuclear Energy Agency (OECD NEA) to address the challenges of not just data preservation, but evaluation of the data to determine its merit for modern and future use. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) was established to identify and verify comprehensive critical benchmark data sets; evaluate the data, including quantification of biases and uncertainties; compile the data and calculations in a standardized format; and formally document the effort into a single source of verified benchmark data [2]. Similarly, the International Reactor Physics Experiment Evaluation Project (IRPhEP) was established to preserve integral reactor physics experimental data, including separate or special effects data for nuclear energy and technology applications [3]. Annually, contributors from around the world continue to collaborate in the evaluation and review of select benchmark experiments for preservation and dissemination. The extensively peer-reviewed integral benchmark data can then be utilized to support nuclear design and safety analysts to validate the analytical tools, methods, and data needed for next

Utilizing benchmark data from the ANL-ZPR diagnostic cores program

International Nuclear Information System (INIS)

Schaefer, R. W.; McKnight, R. D.

2000-01-01

The support of the criticality safety community is allowing the production of benchmark descriptions of several assemblies from the ZPR Diagnostic Cores Program. The assemblies have high sensitivities to nuclear data for a few isotopes. This can highlight limitations in nuclear data for selected nuclides or in standard methods used to treat these data. The present work extends the use of the simplified model of the U9 benchmark assembly beyond the validation of k eff . Further simplifications have been made to produce a data testing benchmark in the style of the standard CSEWG benchmark specifications. Calculations for this data testing benchmark are compared to results obtained with more detailed models and methods to determine their biases. These biases or corrections factors can then be applied in the use of the less refined methods and models. Data testing results using Versions IV, V, and VI of the ENDF/B nuclear data are presented for k eff , f 28 /f 25 , c 28 /f 25 , and β eff . These limited results demonstrate the importance of studying other integral parameters in addition to k eff in trying to improve nuclear data and methods and the importance of accounting for methods and/or modeling biases when using data testing results to infer the quality of the nuclear data files

Validation the methodology calculate critical position of control rods to the critical facility IPEN/MB-01

International Nuclear Information System (INIS)

Lopez Aldama, D.; Rodriguez Gual, R.

1998-01-01

Presently work intends to validate the models and programs used in the Nuclear Technology Center for calculating the critical position of control rods by means of the analysis of the measurements performed at the critical facility IPEN/MB-01. The lattice calculations were carried out with the WIMS/D4 code and for the global calculations the diffusion code SNAP-3D was used

Neutronic computational modeling of the ASTRA critical facility using MCNPX

International Nuclear Information System (INIS)

Rodriguez, L. P.; Garcia, C. R.; Milian, D.; Milian, E. E.; Brayner, C.

2015-01-01

The Pebble Bed Very High Temperature Reactor is considered as a prominent candidate among Generation IV nuclear energy systems. Nevertheless the Pebble Bed Very High Temperature Reactor faces an important challenge due to the insufficient validation of computer codes currently available for use in its design and safety analysis. In this paper a detailed IAEA computational benchmark announced by IAEA-TECDOC-1694 in the framework of the Coordinated Research Project 'Evaluation of High Temperature Gas Cooled Reactor (HTGR) Performance' was solved in support of the Generation IV computer codes validation effort using MCNPX ver. 2.6e computational code. In the IAEA-TECDOC-1694 were summarized a set of four calculational benchmark problems performed at the ASTRA critical facility. Benchmark problems include criticality experiments, control rod worth measurements and reactivity measurements. The ASTRA Critical Facility at the Kurchatov Institute in Moscow was used to simulate the neutronic behavior of nuclear pebble bed reactors. (Author)

Shielding benchmark problems, (2)

International Nuclear Information System (INIS)

Tanaka, Shun-ichi; Sasamoto, Nobuo; Oka, Yoshiaki; Shin, Kazuo; Tada, Keiko.

1980-02-01

Shielding benchmark problems prepared by Working Group of Assessment of Shielding Experiments in the Research Committee on Shielding Design in the Atomic Energy Society of Japan were compiled by Shielding Laboratory in Japan Atomic Energy Research Institute. Fourteen shielding benchmark problems are presented newly in addition to twenty-one problems proposed already, for evaluating the calculational algorithm and accuracy of computer codes based on discrete ordinates method and Monte Carlo method and for evaluating the nuclear data used in codes. The present benchmark problems are principally for investigating the backscattering and the streaming of neutrons and gamma rays in two- and three-dimensional configurations. (author)

RISKIND verification and benchmark comparisons

Energy Technology Data Exchange (ETDEWEB)

Biwer, B.M.; Arnish, J.J.; Chen, S.Y.; Kamboj, S.

1997-08-01

This report presents verification calculations and benchmark comparisons for RISKIND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the population from exposures associated with the transportation of spent nuclear fuel and other radioactive materials. Spreadsheet calculations were performed to verify the proper operation of the major options and calculational steps in RISKIND. The program is unique in that it combines a variety of well-established models into a comprehensive treatment for assessing risks from the transportation of radioactive materials. Benchmark comparisons with other validated codes that incorporate similar models were also performed. For instance, the external gamma and neutron dose rate curves for a shipping package estimated by RISKIND were compared with those estimated by using the RADTRAN 4 code and NUREG-0170 methodology. Atmospheric dispersion of released material and dose estimates from the GENII and CAP88-PC codes. Verification results have shown the program to be performing its intended function correctly. The benchmark results indicate that the predictions made by RISKIND are within acceptable limits when compared with predictions from similar existing models.

RISKIND verification and benchmark comparisons

International Nuclear Information System (INIS)

Biwer, B.M.; Arnish, J.J.; Chen, S.Y.; Kamboj, S.

1997-08-01

This report presents verification calculations and benchmark comparisons for RISKIND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the population from exposures associated with the transportation of spent nuclear fuel and other radioactive materials. Spreadsheet calculations were performed to verify the proper operation of the major options and calculational steps in RISKIND. The program is unique in that it combines a variety of well-established models into a comprehensive treatment for assessing risks from the transportation of radioactive materials. Benchmark comparisons with other validated codes that incorporate similar models were also performed. For instance, the external gamma and neutron dose rate curves for a shipping package estimated by RISKIND were compared with those estimated by using the RADTRAN 4 code and NUREG-0170 methodology. Atmospheric dispersion of released material and dose estimates from the GENII and CAP88-PC codes. Verification results have shown the program to be performing its intended function correctly. The benchmark results indicate that the predictions made by RISKIND are within acceptable limits when compared with predictions from similar existing models

Evaluation of neutron thermalization parameters and benchmark reactor calculations using a synthetic scattering function for molecular gases

International Nuclear Information System (INIS)

Gillete, V.H.; Patino, N.E.; Granada, J.E.; Mayer, R.E.

1988-01-01

Using a synthetic scattering function which describes the interaction of neutrons with molecular gases we provide analytical expressions for zero-and first-order scattering kernels, σ 0 (E 0 →E), σ 1 (E 0 →E), and total cross section σ 0 (E 0 ). Based on these quantities, we have performed calculations of thermalization parameters and transport coefficients for H 2 O, D 2 O, C 6 H 6 and (CH 2 ) n at room temperature. Comparasion of such values with available experimental data and other calculations is satisfactory. We also generated nuclear data libraries for H 2 O with 47 thermal groups at 300K and performed some benchmark calculations ( 235 U, 239 Pu, PWR cell and typical APWR cell); the resulting reactivities are compared with experimental data and ENDF/B-IV calculations. (author) [pt

Quantum computing applied to calculations of molecular energies: CH2 benchmark.

Science.gov (United States)

Veis, Libor; Pittner, Jiří

2010-11-21

Quantum computers are appealing for their ability to solve some tasks much faster than their classical counterparts. It was shown in [Aspuru-Guzik et al., Science 309, 1704 (2005)] that they, if available, would be able to perform the full configuration interaction (FCI) energy calculations with a polynomial scaling. This is in contrast to conventional computers where FCI scales exponentially. We have developed a code for simulation of quantum computers and implemented our version of the quantum FCI algorithm. We provide a detailed description of this algorithm and the results of the assessment of its performance on the four lowest lying electronic states of CH(2) molecule. This molecule was chosen as a benchmark, since its two lowest lying (1)A(1) states exhibit a multireference character at the equilibrium geometry. It has been shown that with a suitably chosen initial state of the quantum register, one is able to achieve the probability amplification regime of the iterative phase estimation algorithm even in this case.

Benchmarking of nuclear economics tools

International Nuclear Information System (INIS)

Moore, Megan; Korinny, Andriy; Shropshire, David; Sadhankar, Ramesh

2017-01-01

Highlights: • INPRO and GIF economic tools exhibited good alignment in total capital cost estimation. • Subtle discrepancies in the cost result from differences in financing and the fuel cycle assumptions. • A common set of assumptions was found to reduce the discrepancies to 1% or less. • Opportunities for harmonisation of economic tools exists. - Abstract: Benchmarking of the economics methodologies developed by the Generation IV International Forum (GIF) and the International Atomic Energy Agency’s International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO), was performed for three Generation IV nuclear energy systems. The Economic Modeling Working Group of GIF developed an Excel based spreadsheet package, G4ECONS (Generation 4 Excel-based Calculation Of Nuclear Systems), to calculate the total capital investment cost (TCIC) and the levelised unit energy cost (LUEC). G4ECONS is sufficiently generic in the sense that it can accept the types of projected input, performance and cost data that are expected to become available for Generation IV systems through various development phases and that it can model both open and closed fuel cycles. The Nuclear Energy System Assessment (NESA) Economic Support Tool (NEST) was developed to enable an economic analysis using the INPRO methodology to easily calculate outputs including the TCIC, LUEC and other financial figures of merit including internal rate of return, return of investment and net present value. NEST is also Excel based and can be used to evaluate nuclear reactor systems using the open fuel cycle, MOX (mixed oxide) fuel recycling and closed cycles. A Super Critical Water-cooled Reactor system with an open fuel cycle and two Fast Reactor systems, one with a break-even fuel cycle and another with a burner fuel cycle, were selected for the benchmarking exercise. Published data on capital and operating costs were used for economics analyses using G4ECONS and NEST tools. Both G4ECONS and

Criticality safety analysis for mockup facility

International Nuclear Information System (INIS)

Shin, Young Joon; Shin, Hee Sung; Kim, Ik Soo; Oh, Seung Chul; Ro, Seung Gy; Bae, Kang Mok

2000-03-01

Benchmark calculations for SCALE4.4 CSAS6 module have been performed for 31 UO 2 fuel, 15MOX fuel and 10 metal material criticality experiments and then calculation biases of the SCALE 4.4 CSAS6 module have been revealed to be 0.00982, 0.00579 and 0.02347, respectively. When CSAS6 is applied to the criticality safety analysis for the mockup facility in which several kinds of nuclear material components are included, the calculation bias of CSAS6 is conservatively taken to be 0.02347. With the aid of this benchmarked code system, criticality safety analyses for the mockup facility at normal and hypothetical accidental conditions have been carried out. It appears that the maximum K eff is 0.28356 well below than the critical limit, K eff =0.95 at normal condition. In a hypothetical accidental condition, the maximum K eff is found to be 0.73527 much lower than the subcritical limit. For another hypothetical accidental condition the nuclear material leaks out of container and spread or lump in the floor, it was assumed that the nuclear material is shaped into a slab and water exists in the empty space of the nuclear material. K eff has been calculated as function of slab thickness and the volume ratio of water to nuclear material. The result shows that the K eff increases as the water volume ratio increases. It is also revealed that the K eff reaches to the maximum value when water if filled in the empty space of nuclear material. The maximum K eff value is 0.93960 lower than the subcritical limit

Verification of RRC Ki code package for neutronic calculations of WWER core with GD

International Nuclear Information System (INIS)

Aleshin, S.S.; Bolshagin, S.N.; Lazarenko, A.P.; Markov, A.V.; Pavlov, V.I.; Pavlovitchev, A.M.; Sidorenko, V.D.; Tsvetkov, V.M.

2001-01-01

The report presented is concerned with verification results of TVS-M/PERMAK-A/BIPR-7A code package for WWERs neutronic calculation as applied to calculation of systems containing U-GD pins. The verification is based on corresponded benchmark calculations, data critical experiments and on operation data obtained WWER units with Gd. The comparison results are discussed (Authors)

Validation of calculational methods for nuclear criticality safety - approved 1975

International Nuclear Information System (INIS)

Anon.

1977-01-01

The American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors, N16.1-1975, states in 4.2.5: In the absence of directly applicable experimental measurements, the limits may be derived from calculations made by a method shown to be valid by comparison with experimental data, provided sufficient allowances are made for uncertainties in the data and in the calculations. There are many methods of calculation which vary widely in basis and form. Each has its place in the broad spectrum of problems encountered in the nuclear criticality safety field; however, the general procedure to be followed in establishing validity is common to all. The standard states the requirements for establishing the validity and area(s) of applicability of any calculational method used in assessing nuclear criticality safety

Variational Variance Reduction for Monte Carlo Criticality Calculations

International Nuclear Information System (INIS)

Densmore, Jeffery D.; Larsen, Edward W.

2001-01-01

A new variational variance reduction (VVR) method for Monte Carlo criticality calculations was developed. This method employs (a) a variational functional that is more accurate than the standard direct functional, (b) a representation of the deterministically obtained adjoint flux that is especially accurate for optically thick problems with high scattering ratios, and (c) estimates of the forward flux obtained by Monte Carlo. The VVR method requires no nonanalog Monte Carlo biasing, but it may be used in conjunction with Monte Carlo biasing schemes. Some results are presented from a class of criticality calculations involving alternating arrays of fuel and moderator regions

Catalog and history of the experiments of criticality Saclay (1958-1964) Valduc / Building 10 (1964-2003)

International Nuclear Information System (INIS)

Poullot, G.; Dumont, V.; Anno, J.; Cousinou, P.; Grivot, P.; Girault, E.; Fouillaud, P.; Barbry, F.

2003-01-01

The group ' International Criticality Safety Evaluation Benchmark evaluation project ' (I.C.S.B.E.P.) has for aim to supply to the international community experiments of benchmarks criticality, of certified quality, used to guarantee the qualification of criticality calculation codes. Have been defined: a structure of experiments classification, a format of standard presentation, a structure of work with evaluation, internal and external checks, presentation in plenary session. After favourable opinion of the work group, the synthesis document called evaluation is integrated to the general report I.C.S.B.E.P. (N.C.)

IAEA coordinated research project (CRP) on 'Analytical and experimental benchmark analyses of accelerator driven systems'

Energy Technology Data Exchange (ETDEWEB)

Abanades, Alberto [Universidad Politecnica de Madrid (Spain); Aliberti, Gerardo; Gohar, Yousry; Talamo, Alberto [ANL, Argonne (United States); Bornos, Victor; Kiyavitskaya, Anna [Joint Institute of Power Eng. and Nucl. Research ' Sosny' , Minsk (Belarus); Carta, Mario [ENEA, Casaccia (Italy); Janczyszyn, Jerzy [AGH-University of Science and Technology, Krakow (Poland); Maiorino, Jose [IPEN, Sao Paulo (Brazil); Pyeon, Cheolho [Kyoto University (Japan); Stanculescu, Alexander [IAEA, Vienna (Austria); Titarenko, Yury [ITEP, Moscow (Russian Federation); Westmeier, Wolfram [Wolfram Westmeier GmbH, Ebsdorfergrund (Germany)

2008-07-01

In December 2005, the International Atomic Energy Agency (IAEA) has started a Coordinated Research Project (CRP) on 'Analytical and Experimental Benchmark Analyses of Accelerator Driven Systems'. The overall objective of the CRP, performed within the framework of the Technical Working Group on Fast Reactors (TWGFR) of IAEA's Nuclear Energy Department, is to increase the capability of interested Member States in developing and applying advanced reactor technologies in the area of long-lived radioactive waste utilization and transmutation. The specific objective of the CRP is to improve the present understanding of the coupling of an external neutron source (e.g. spallation source) with a multiplicative sub-critical core. The participants are performing computational and experimental benchmark analyses using integrated calculation schemes and simulation methods. The CRP aims at integrating some of the planned experimental demonstration projects of the coupling between a sub-critical core and an external neutron source (e.g. YALINA Booster in Belarus, and Kyoto University's Critical Assembly (KUCA)). The objective of these experimental programs is to validate computational methods, obtain high energy nuclear data, characterize the performance of sub-critical assemblies driven by external sources, and to develop and improve techniques for sub-criticality monitoring. The paper summarizes preliminary results obtained to-date for some of the CRP benchmarks. (authors)

Preliminary analysis of the proposed BN-600 benchmark core

International Nuclear Information System (INIS)

John, T.M.

2000-01-01

The Indira Gandhi Centre for Atomic Research is actively involved in the design of Fast Power Reactors in India. The core physics calculations are performed by the computer codes that are developed in-house or by the codes obtained from other laboratories and suitably modified to meet the computational requirements. The basic philosophy of the core physics calculations is to use the diffusion theory codes with the 25 group nuclear cross sections. The parameters that are very sensitive is the core leakage, like the power distribution at the core blanket interface etc. are calculated using transport theory codes under the DSN approximations. All these codes use the finite difference approximation as the method to treat the spatial variation of the neutron flux. Criticality problems having geometries that are irregular to be represented by the conventional codes are solved using Monte Carlo methods. These codes and methods have been validated by the analysis of various critical assemblies and calculational benchmarks. Reactor core design procedure at IGCAR consists of: two and three dimensional diffusion theory calculations (codes ALCIALMI and 3DB); auxiliary calculations, (neutron balance, power distributions, etc. are done by codes that are developed in-house); transport theory corrections from two dimensional transport calculations (DOT); irregular geometry treated by Monte Carlo method (KENO); cross section data library used CV2M (25 group)

Design and Implementation of a Web-Based Reporting and Benchmarking Center for Inpatient Glucometrics

Science.gov (United States)

Schnipper, Jeffrey Lawrence; Messler, Jordan; Ramos, Pedro; Kulasa, Kristen; Nolan, Ann; Rogers, Kendall

2014-01-01

Background: Insulin is a top source of adverse drug events in the hospital, and glycemic control is a focus of improvement efforts across the country. Yet, the majority of hospitals have no data to gauge their performance on glycemic control, hypoglycemia rates, or hypoglycemic management. Current tools to outsource glucometrics reports are limited in availability or function. Methods: Society of Hospital Medicine (SHM) faculty designed and implemented a web-based data and reporting center that calculates glucometrics on blood glucose data files securely uploaded by users. Unit labels, care type (critical care, non–critical care), and unit type (eg, medical, surgical, mixed, pediatrics) are defined on upload allowing for robust, flexible reporting. Reports for any date range, care type, unit type, or any combination of units are available on demand for review or downloading into a variety of file formats. Four reports with supporting graphics depict glycemic control, hypoglycemia, and hypoglycemia management by patient day or patient stay. Benchmarking and performance ranking reports are generated periodically for all hospitals in the database. Results: In all, 76 hospitals have uploaded at least 12 months of data for non–critical care areas and 67 sites have uploaded critical care data. Critical care benchmarking reveals wide variability in performance. Some hospitals achieve top quartile performance in both glycemic control and hypoglycemia parameters. Conclusions: This new web-based glucometrics data and reporting tool allows hospitals to track their performance with a flexible reporting system, and provides them with external benchmarking. Tools like this help to establish standardized glucometrics and performance standards. PMID:24876426

Verification and validation benchmarks.

Energy Technology Data Exchange (ETDEWEB)

Oberkampf, William Louis; Trucano, Timothy Guy

2007-02-01

Verification and validation (V&V) are the primary means to assess the accuracy and reliability of computational simulations. V&V methods and procedures have fundamentally improved the credibility of simulations in several high-consequence fields, such as nuclear reactor safety, underground nuclear waste storage, and nuclear weapon safety. Although the terminology is not uniform across engineering disciplines, code verification deals with assessing the reliability of the software coding, and solution verification deals with assessing the numerical accuracy of the solution to a computational model. Validation addresses the physics modeling accuracy of a computational simulation by comparing the computational results with experimental data. Code verification benchmarks and validation benchmarks have been constructed for a number of years in every field of computational simulation. However, no comprehensive guidelines have been proposed for the construction and use of V&V benchmarks. For example, the field of nuclear reactor safety has not focused on code verification benchmarks, but it has placed great emphasis on developing validation benchmarks. Many of these validation benchmarks are closely related to the operations of actual reactors at near-safety-critical conditions, as opposed to being more fundamental-physics benchmarks. This paper presents recommendations for the effective design and use of code verification benchmarks based on manufactured solutions, classical analytical solutions, and highly accurate numerical solutions. In addition, this paper presents recommendations for the design and use of validation benchmarks, highlighting the careful design of building-block experiments, the estimation of experimental measurement uncertainty for both inputs and outputs to the code, validation metrics, and the role of model calibration in validation. It is argued that the understanding of predictive capability of a computational model is built on the level of

SIMCRI: a simple computer code for calculating nuclear criticality parameters

International Nuclear Information System (INIS)

Nakamaru, Shou-ichi; Sugawara, Nobuhiko; Naito, Yoshitaka; Katakura, Jun-ichi; Okuno, Hiroshi.

1986-03-01

This is a user's manual for a simple criticality calculation code SIMCRI. The code has been developed to facilitate criticality calculation on a single unit of nuclear fuel. SIMCRI makes an extensive survey with a little computing time. Cross section library MGCL for SIMCRI is the same one for the Monte Carlo criticality code KENOIV; it is, therefore, easy to compare the results of the two codes. SIMCRI solves eigenvalue problems and fixed source problems based on the one space point B 1 equation. The results include infinite and effective multiplication factor, critical buckling, migration area, diffusion coefficient and so on. SIMCRI is comprised in the criticality safety evaluation code system JACS. (author)

A benchmark on the calculation of kinetic parameters based on reactivity effect experiments in the CROCUS reactor

International Nuclear Information System (INIS)

Paratte, J.M.; Frueh, R.; Kasemeyer, U.; Kalugin, M.A.; Timm, W.; Chawla, R.

2006-01-01

Measurements in the CROCUS reactor at EPFL, Lausanne, are reported for the critical water level and the inverse reactor period for several different sets of delayed supercritical conditions. The experimental configurations were also calculated by four different calculation methods. For each of the supercritical configurations, the absolute reactivity value has been determined in two different ways, viz.: (i) through direct comparison of the multiplication factor obtained employing a given calculation method with the corresponding value for the critical case (calculated reactivity: ρ calc ); (ii) by application of the inhour equation using the kinetic parameters obtained for the critical configuration and the measured inverse reactor period (measured reactivity: ρ meas ). The calculated multiplication factors for the reference critical configuration, as well as ρ calc for the supercritical cases, are found to be in good agreement. However, the values of ρ meas produced by two of the applied calculation methods differ appreciably from the corresponding ρ calc values, clearly indicating deficiencies in the kinetic parameters obtained from these methods

Burn-up TRIGA Mark II benchmark experiment

International Nuclear Information System (INIS)

Persic, A.; Ravnik, M.; Zagar, T.

1998-01-01

Different reactor codes are used for calculations of reactor parameters. The accuracy of the programs is tested through comparison of the calculated values with the experimental results. Well-defined and accurately measured benchmarks are required. The experimental results of reactivity measurements, fuel element reactivity worth distribution and fuel-up measurements are presented in this paper. The experiments were performed with partly burnt reactor core. The experimental conditions were well defined, so that the results can be used as a burn-up benchmark test case for a TRIGA Mark II reactor calculations.(author)

Revaluering benchmarking - A topical theme for the construction industry

DEFF Research Database (Denmark)

Rasmussen, Grane Mikael Gregaard

2011-01-01

and questioning the concept objectively. This paper addresses the underlying nature of benchmarking, and accounts for the importance of focusing attention on the sociological impacts benchmarking has in organizations. To understand these sociological impacts, benchmarking research needs to transcend...... the perception of benchmarking systems as secondary and derivative and instead studying benchmarking as constitutive of social relations and as irredeemably social phenomena. I have attempted to do so in this paper by treating benchmarking using a calculative practice perspective, and describing how...

Evaluation of cross sections of 56Fe up to 3 GeV and integral benchmark calculation for thick target yield

International Nuclear Information System (INIS)

Yoshizawa, Nobuaki; Meigo, Shin-ichiro

2001-01-01

The neutron and proton cross sections of 56 Fe were evaluated up to 3 GeV. JENDL High Energy File of 56 Fe were developed for use in transport calculation. For neutrons, the high-energy data are merged with JENDL3.3-file. Integral benchmark calculations for thick target neutron yields (TTY) for 113 MeV and 256 MeV proton bombardment of Fe targets were performed using the evaluated libraries. Calculated TTY neutron spectra were compared with experimental data. For 113 MeV, calculated TTY at 7.5 degree underestimated in the emitted neutron energy range above 10 MeV. For 256 MeV, calculated TTY well agree with experimental data except below 10 MeV. (author)

Proposal on the accelerator driven molten-salt reactor (ATW-concept) benchmark calculation (stage-1 without an external neutron sources)

International Nuclear Information System (INIS)

Svarny, J.; Mikolas, P.

1999-01-01

The simple model of two component concept of the ATW (graphite + molten salt system) was found. The main purpose of this benchmark will be not only to provide the basic characteristics of given ADS but also to test codes in calculations of the rate of transmutation waste and to evaluate basic kinetics parameters and reactivity effects. (Authors)

KAERI results for BN600 full MOX benchmark (Phase 4)

International Nuclear Information System (INIS)

Lee, Kibog Lee

2003-01-01

The purpose of this document is to report the results of KAERI's calculation for the Phase-4 of BN-600 full MOX fueled core benchmark analyses according to the RCM report of IAEA CRP Action on U pdated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects. T he BN-600 full MOX core model is based on the specification in the document, F ull MOX Model (Phase4. doc ) . This document addresses the calculational methods employed in the benchmark analyses and benchmark results carried out by KAERI

Benchmark Calculation of Radial Expectation Value for Confined Hydrogen-Like Atoms and Isotropic Harmonic Oscillators

International Nuclear Information System (INIS)

Yu, Rong Mei; Zan, Li Rong; Jiao, Li Guang; Ho, Yew Kam

2017-01-01

Spatially confined atoms have been extensively investigated to model atomic systems in extreme pressures. For the simplest hydrogen-like atoms and isotropic harmonic oscillators, numerous physical quantities have been established with very high accuracy. However, the expectation value of which is of practical importance in many applications has significant discrepancies among calculations by different methods. In this work we employed the basis expansion method with cut-off Slater-type orbitals to investigate these two confined systems. Accurate values for several low-lying bound states were obtained by carefully examining the convergence with respect to the size of basis. A scaling law for was derived and it is used to verify the accuracy of numerical results. Comparison with other calculations show that the present results establish benchmark values for this quantity, which may be useful in future studies. (author)

SU-E-J-30: Benchmark Image-Based TCP Calculation for Evaluation of PTV Margins for Lung SBRT Patients

Energy Technology Data Exchange (ETDEWEB)

Li, M [Wayne State Univeristy, Detroit, MI (United States); Chetty, I [Henry Ford Health System, Detroit, MI (United States); Zhong, H [Henry Ford Hospital System, Detroit, MI (United States)

2014-06-01

Purpose: Tumor control probability (TCP) calculated with accumulated radiation doses may help design appropriate treatment margins. Image registration errors, however, may compromise the calculated TCP. The purpose of this study is to develop benchmark CT images to quantify registration-induced errors in the accumulated doses and their corresponding TCP. Methods: 4DCT images were registered from end-inhale (EI) to end-exhale (EE) using a “demons” algorithm. The demons DVFs were corrected by an FEM model to get realistic deformation fields. The FEM DVFs were used to warp the EI images to create the FEM-simulated images. The two images combined with the FEM DVF formed a benchmark model. Maximum intensity projection (MIP) images, created from the EI and simulated images, were used to develop IMRT plans. Two plans with 3 and 5 mm margins were developed for each patient. With these plans, radiation doses were recalculated on the simulated images and warped back to the EI images using the FEM DVFs to get the accumulated doses. The Elastix software was used to register the FEM-simulated images to the EI images. TCPs calculated with the Elastix-accumulated doses were compared with those generated by the FEM to get the TCP error of the Elastix registrations. Results: For six lung patients, the mean Elastix registration error ranged from 0.93 to 1.98 mm. Their relative dose errors in PTV were between 0.28% and 6.8% for 3mm margin plans, and between 0.29% and 6.3% for 5mm-margin plans. As the PTV margin reduced from 5 to 3 mm, the mean TCP error of the Elastix-reconstructed doses increased from 2.0% to 2.9%, and the mean NTCP errors decreased from 1.2% to 1.1%. Conclusion: Patient-specific benchmark images can be used to evaluate the impact of registration errors on the computed TCPs, and may help select appropriate PTV margins for lung SBRT patients.

SU-E-J-30: Benchmark Image-Based TCP Calculation for Evaluation of PTV Margins for Lung SBRT Patients

International Nuclear Information System (INIS)

Li, M; Chetty, I; Zhong, H

2014-01-01

Purpose: Tumor control probability (TCP) calculated with accumulated radiation doses may help design appropriate treatment margins. Image registration errors, however, may compromise the calculated TCP. The purpose of this study is to develop benchmark CT images to quantify registration-induced errors in the accumulated doses and their corresponding TCP. Methods: 4DCT images were registered from end-inhale (EI) to end-exhale (EE) using a “demons” algorithm. The demons DVFs were corrected by an FEM model to get realistic deformation fields. The FEM DVFs were used to warp the EI images to create the FEM-simulated images. The two images combined with the FEM DVF formed a benchmark model. Maximum intensity projection (MIP) images, created from the EI and simulated images, were used to develop IMRT plans. Two plans with 3 and 5 mm margins were developed for each patient. With these plans, radiation doses were recalculated on the simulated images and warped back to the EI images using the FEM DVFs to get the accumulated doses. The Elastix software was used to register the FEM-simulated images to the EI images. TCPs calculated with the Elastix-accumulated doses were compared with those generated by the FEM to get the TCP error of the Elastix registrations. Results: For six lung patients, the mean Elastix registration error ranged from 0.93 to 1.98 mm. Their relative dose errors in PTV were between 0.28% and 6.8% for 3mm margin plans, and between 0.29% and 6.3% for 5mm-margin plans. As the PTV margin reduced from 5 to 3 mm, the mean TCP error of the Elastix-reconstructed doses increased from 2.0% to 2.9%, and the mean NTCP errors decreased from 1.2% to 1.1%. Conclusion: Patient-specific benchmark images can be used to evaluate the impact of registration errors on the computed TCPs, and may help select appropriate PTV margins for lung SBRT patients

Implantation of a new calculation method of fuel depletion in the CITHAM code

International Nuclear Information System (INIS)

Alvarenga, M.A.B.

1985-01-01

It is evaluated the accuracy of the linear aproximation method used in the CITHAN code to obtain the solution of depletion equations. Results are compared with the Benchmark problem. The convenience of depletion chain before criticality calculations is analysed. The depletion calculation was modified using linear combination technic of linear chains. (M.C.K.) [pt

SCALE-4 analysis of pressurized water reactor critical configurations. Volume 1: Summary

International Nuclear Information System (INIS)

DeHart, M.D.

1995-03-01

The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit is to be taken for the reduced reactivity of burned or spent fuel relative to its original fresh composition, it is necessary to benchmark computational methods used in determining such reactivity worth against spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized water reactors (PWR). The analysis methodology utilized for all calculations in this report is based on the modules and data associated with the SCALE-4 code system. Each of the five volumes comprising this report provides an overview of the methodology applied. Subsequent volumes also describe in detail the approach taken in performing criticality calculations for these PWR configurations: Volume 2 describes criticality calculations for the Tennessee Valley Authority's Sequoyah Unit 2 reactor for Cycle 3; Volume 3 documents the analysis of Virginia Power's Surry Unit 1 reactor for the Cycle 2 core; Volume 4 documents the calculations performed based on GPU Nuclear Corporation's Three Mile Island Unit 1 Cycle 5 core; and, lastly, Volume 5 describes the analysis of Virginia Power's North Anna Unit 1 Cycle 5 core. Each of the reactor-specific volumes provides the details of calculations performed to determine the effective multiplication factor for each reactor core for one or more critical configurations using the SCALE-4 system; these results are summarized in this volume. Differences between the core designs and their possible impact on the criticality calculations are also discussed. Finally, results are presented for additional analyses performed to verify that solutions were sufficiently converged

Benchmarking

OpenAIRE

Meylianti S., Brigita

1999-01-01

Benchmarking has different meaning to different people. There are five types of benchmarking, namely internal benchmarking, competitive benchmarking, industry / functional benchmarking, process / generic benchmarking and collaborative benchmarking. Each type of benchmarking has its own advantages as well as disadvantages. Therefore it is important to know what kind of benchmarking is suitable to a specific application. This paper will discuss those five types of benchmarking in detail, includ...

Comparative sensitivity study of some criticality safety benchmark experiments using JEFF-3.1.2, JEFF-3.2T and ENDF/B-VII.1

International Nuclear Information System (INIS)

Kooyman, Timothee; Messaoudia, Nadia

2014-01-01

A sensitivity study on a set of evaluated criticality benchmarks with two versions of the JEFF nuclear data library, namely JEFF-3.1.2 and JEFF-3.2T, and ENDF/B-VII.1 was performed using MNCP(X) 2.6.0. As these benchmarks serve to estimate the upper safety limit for criticality risk analysis at SCK.CEN the sensitivity of their results to nuclear data is an important parameter to asses. Several nuclides were identified as being responsible for an evident change in the effective multiplication factor k eff : 235 U, 239 Pu, 240 Pu, 54 Fe, 56 Fe, 57 Fe and 208 Pb. A high sensitivity was found to the fission cross-section of all the fissile material in the study. Additionally, a smaller sensitivity to inelastic and capture cross-section of 235 U and 240 Pu was also found. Sensitivity to the scattering law for non-fissile material was postulated. The biggest change in the k eff due to non-fissile material was due to 208 Pb evaluation (±700 pcm), followed by 56 Fe (±360 pcm) for both versions of the JEFF library. Changes due to 235 U (±300 pcm) and Pu isotopes (±120 pcm for 239 Pu and ±80 pcm for 240 Pu) were found only with JEFF-3.1.2. 238 U was found to have no effect on the k eff . Significant improvements were identified between the two versions of the JEFF library. No further differences were found between the JEFF-3.2T and the ENDF/B-VII.1 calculations involving 235 U or Pu. (authors)

Links among available integral benchmarks and differential date evaluations, computational biases and uncertainties, and nuclear criticality safety biases on potential MOX production throughput

International Nuclear Information System (INIS)

Goluoglu, S.; Hopper, C.M.

2004-01-01

Through the use of Oak Ridge National Laboratory's recently developed and applied sensitivity and uncertainty computational analysis techniques, this paper presents the relevance and importance of available and needed integral benchmarks and differential data evaluations impacting potential MOX production throughput determinations relative to low-moderated MOX fuel blending operations. The relevance and importance in the availability of or need for critical experiment benchmarks and data evaluations are presented in terms of computational biases as influenced by computational and experimental sensitivities and uncertainties relative to selected MOX production powder blending processes. Recent developments for estimating the safe margins of subcriticality for assuring nuclear criticality safety for process approval are presented. In addition, the impact of the safe margins (due to computational biases and uncertainties) on potential MOX production throughput will also be presented. (author)

Criticality Calculations with MCNP6 - Practical Lectures

Energy Technology Data Exchange (ETDEWEB)

Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Methods, Codes, and Applications (XCP-3); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Methods, Codes, and Applications (XCP-3); Alwin, Jennifer Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Methods, Codes, and Applications (XCP-3)

2016-11-29

These slides are used to teach MCNP (Monte Carlo N-Particle) usage to nuclear criticality safety analysts. The following are the lecture topics: course information, introduction, MCNP basics, criticality calculations, advanced geometry, tallies, adjoint-weighted tallies and sensitivities, physics and nuclear data, parameter studies, NCS validation I, NCS validation II, NCS validation III, case study 1 - solution tanks, case study 2 - fuel vault, case study 3 - B&W core, case study 4 - simple TRIGA, case study 5 - fissile mat. vault, criticality accident alarm systems. After completion of this course, you should be able to: Develop an input model for MCNP; Describe how cross section data impact Monte Carlo and deterministic codes; Describe the importance of validation of computer codes and how it is accomplished; Describe the methodology supporting Monte Carlo codes and deterministic codes; Describe pitfalls of Monte Carlo calculations; Discuss the strengths and weaknesses of Monte Carlo and Discrete Ordinants codes; The diffusion theory model is not strictly valid for treating fissile systems in which neutron absorption, voids, and/or material boundaries are present. In the context of these limitations, identify a fissile system for which a diffusion theory solution would be adequate.

Criticality Calculations with MCNP6 - Practical Lectures

International Nuclear Information System (INIS)

Brown, Forrest B.; Rising, Michael Evan; Alwin, Jennifer Louise

2016-01-01

These slides are used to teach MCNP (Monte Carlo N-Particle) usage to nuclear criticality safety analysts. The following are the lecture topics: course information, introduction, MCNP basics, criticality calculations, advanced geometry, tallies, adjoint-weighted tallies and sensitivities, physics and nuclear data, parameter studies, NCS validation I, NCS validation II, NCS validation III, case study 1 - solution tanks, case study 2 - fuel vault, case study 3 - B&W core, case study 4 - simple TRIGA, case study 5 - fissile mat. vault, criticality accident alarm systems. After completion of this course, you should be able to: Develop an input model for MCNP; Describe how cross section data impact Monte Carlo and deterministic codes; Describe the importance of validation of computer codes and how it is accomplished; Describe the methodology supporting Monte Carlo codes and deterministic codes; Describe pitfalls of Monte Carlo calculations; Discuss the strengths and weaknesses of Monte Carlo and Discrete Ordinants codes; The diffusion theory model is not strictly valid for treating fissile systems in which neutron absorption, voids, and/or material boundaries are present. In the context of these limitations, identify a fissile system for which a diffusion theory solution would be adequate.

Validation of SCALE-4 criticality sequences using ENDF/B-V data

International Nuclear Information System (INIS)

Bowman, S.M.; Wright, R.Q.; DeHart, M.D.; Taniuchi, H.

1993-01-01

The SCALE code system developed at Oak Ridge National Laboratory contains criticality safety analysis sequences that include the KENO V.a Monte Carlo code for calculation of the effective multiplication factor. These sequences are widely used for criticality safety analyses performed both in the United States and abroad. The purpose of the current work is to validate the SCALE-4 criticality sequences with an ENDF/B-V cross-section library for future distribution with SCALE-4. The library used for this validation is a broad-group library (44 groups) collapsed from the 238-group SCALE library. Extensive data testing of both the 238-group and the 44-group libraries included 10 fast and 18 thermal CSEWG benchmarks and 5 other fast benchmarks. Both libraries contain approximately 300 nuclides and are, therefore, capable of modeling most systems, including those containing spent fuel or radioactive waste. The validation of the broad-group library used 93 critical experiments as benchmarks. The range of experiments included 60 light-water-reactor fuel rod lattices, 13 mixed-oxide fuel rod lattice, and 15 other low- and high-enriched uranium critical assemblies

Solution of the fifth dynamic Atomic Energy Research benchmark problem using the coupled code DIN3/ATHLET

International Nuclear Information System (INIS)

Kliem, S.

1998-01-01

The fifth dynamic benchmark is the first benchmark for coupled thermohydraulic system/three dimensional hexagonal neutron kinetic core models. In this benchmark the interaction between the components of a WWER-440 NPP with the reactor core has been investigated. The initiating event is a symmetrical break of the main steam header at the end of the first fuel cycle and the shutdown conditions with one control rod group s tucking. This break causes an overcooling of the primary circuit. During this overcooling the scram reactivity is compensated and the scrammed reactor becomes re critical. The calculation was continued until the highly-borated water from the high pressure injection system terminated the power excursion. Several aspects of the very complex and complicated benchmark problem are analyzed in detail. Sensitivity studies with different hydraulic parameters are made. The influence on the course of the transient and on the solution is discussed.(Author)

Analysis of benchmark experiments for testing the IKE multigroup cross-section libraries based on ENDF/B-III and IV

International Nuclear Information System (INIS)

Keinert, J.; Mattes, M.

1975-01-01

Benchmark experiments offer the most direct method for validation of nuclear cross-section sets and calculational methods. For 16 fast and thermal critical assemblies containing uranium and/or plutonium of different compositions we compared our calculational results with measured integral quantities, such as ksub(eff), central reaction rate ratios or fast and thermal activation (dis)advantage factors. Cause of the simple calculational modelling of these assemblies the calculations proved as a good test for the IKE multigroup cross-section libraries essentially based on ENDF/B-IV. In general, our calculational results are in excellent agreement with the measured values. Only with some critical systems the basic ENDF/B-IV data proved to be insufficient in calculating ksub(eff), probably due to Pu neutron data and U 238 fast capture cross-sections. (orig.) [de

Spectrum integrated (n,He) cross section comparisons and least squares analyses for 6Li and 10B in benchmark fields

International Nuclear Information System (INIS)

Schenter, R.E.; Oliver, B.M.; Farrar, H. IV.

1986-06-01

Spectrum integrated cross sections for 6 Li and 10 B from five benchmark fast reactor neutron fields are compared with calculated values obtained using the ENDF/B-V Cross Section Files. The benchmark fields include the Coupled Fast Reactivity Measurements Facility (CFRMF) at the Idaho National Engineering Laboratory, the 10% Enriched U-235 Critical Assembly (BIG-10) at Los Alamos National Laboratory, the Sigma-Sigma and Fission Cavity fields of the BR-1 reactor at CEN/SCK, and the Intermediate Energy Standard Neutron Field (ISNF) at the National Bureau of Standards. Results from least square analyses using the FERRET computer code to obtain adjusted cross section values and their uncertainties are presented. Input to these calculations include the above five benchmark data sets. These analyses indicate a need for revision in the ENDF/B-V files for the 10 B and 6 Li cross sections for energies above 50 keV

VENUS-2 Benchmark Problem Analysis with HELIOS-1.9

International Nuclear Information System (INIS)

Jeong, Hyeon-Jun; Choe, Jiwon; Lee, Deokjung

2014-01-01

Since there are reliable results of benchmark data from the OECD/NEA report of the VENUS-2 MOX benchmark problem, by comparing benchmark results users can identify the credibility of code. In this paper, the solution of the VENUS-2 benchmark problem from HELIOS 1.9 using the ENDF/B-VI library(NJOY91.13) is compared with the result from HELIOS 1.7 with consideration of the MCNP-4B result as reference data. The comparison contains the results of pin cell calculation, assembly calculation, and core calculation. The eigenvalues from those are considered by comparing the results from other codes. In the case of UOX and MOX assemblies, the differences from the MCNP-4B results are about 10 pcm. However, there is some inaccuracy in baffle-reflector condition, and relatively large differences were found in the MOX-reflector assembly and core calculation. Although HELIOS 1.9 utilizes an inflow transport correction, it seems that it has a limited effect on the error in baffle-reflector condition

Analysis of a molten salt reactor benchmark

International Nuclear Information System (INIS)

Ghosh, Biplab; Bajpai, Anil; Degweker, S.B.

2013-01-01

This paper discusses results of our studies of an IAEA molten salt reactor (MSR) benchmark. The benchmark, proposed by Japan, involves burnup calculations of a single lattice cell of a MSR for burning plutonium and other minor actinides. We have analyzed this cell with in-house developed burnup codes BURNTRAN and McBURN. This paper also presents a comparison of the results of our codes and those obtained by the proposers of the benchmark. (author)

Validating MCNP5 libraries and tracking the reason for differences between libraries in criticality calculations

International Nuclear Information System (INIS)

Hossny, K.

2015-01-01

The purpose of this work is to validate MCNP5 libraries by simulating 4 detailed benchmark experiments and comparing MCNP5 results (each library) with the experimental results and also the previously validated codes for the same experiments MORET 4.A coupled with APOLLO2 (France), and MONK8 (UK). The reasons for difference between libraries are also investigated in this work. Investigating the reason for the differences between libraries will be done by specifying a different library for specific part (clad, fuel, light water) and checking the result deviation than the previously calculated result (with all parts of the same library). The investigated benchmark experiments are of single fuel rods arrays that are water-moderated and water-reflected. Rods contained low-enriched (4.738 wt.% 92 235 U)uranium dioxide (UO 2 ) fuel were clad with aluminum alloy AGS. These experiments were subcritical approaches extrapolated to critical, with the multiplication factor reached being very close to 1.000 (within 0.1%); the subcritical approach parameter was the water level. The studied four cases differ from each other in pitch, number of fuel rods and of course critical height of water. The results show that although library ENDF/B-IV lacks light water treatment card, however its results can be reliable as light water treatment library does not have significant differences from library to another, so it will not be necessary to specify light water treatment card. The main reason for differences between ENDF/B-V and ENDF/B-VI is light water material, especially the Hydrogen element. Specifying the library of Uranium is necessary in case of using library ENDF/B-IV. On the other hand it is not necessary to specify library of cladding material whatever the used library. Validated libraries are ENDF/BIV, ENDF/B-V and ENDF/B-VI with codes in MCNP 42C, 50C and 60C respectively. The presentation slides have been added to the article

MCNP Perturbation Capability for Monte Carlo Criticality Calculations

International Nuclear Information System (INIS)

Hendricks, J.S.; Carter, L.L.; McKinney, G.W.

1999-01-01

The differential operator perturbation capability in MCNP4B has been extended to automatically calculate perturbation estimates for the track length estimate of k eff in MCNP4B. The additional corrections required in certain cases for MCNP4B are no longer needed. Calculating the effect of small design changes on the criticality of nuclear systems with MCNP is now straightforward

Verification and validation benchmarks

International Nuclear Information System (INIS)

Oberkampf, William Louis; Trucano, Timothy Guy

2007-01-01

Verification and validation (V and V) are the primary means to assess the accuracy and reliability of computational simulations. V and V methods and procedures have fundamentally improved the credibility of simulations in several high-consequence fields, such as nuclear reactor safety, underground nuclear waste storage, and nuclear weapon safety. Although the terminology is not uniform across engineering disciplines, code verification deals with assessing the reliability of the software coding, and solution verification deals with assessing the numerical accuracy of the solution to a computational model. Validation addresses the physics modeling accuracy of a computational simulation by comparing the computational results with experimental data. Code verification benchmarks and validation benchmarks have been constructed for a number of years in every field of computational simulation. However, no comprehensive guidelines have been proposed for the construction and use of V and V benchmarks. For example, the field of nuclear reactor safety has not focused on code verification benchmarks, but it has placed great emphasis on developing validation benchmarks. Many of these validation benchmarks are closely related to the operations of actual reactors at near-safety-critical conditions, as opposed to being more fundamental-physics benchmarks. This paper presents recommendations for the effective design and use of code verification benchmarks based on manufactured solutions, classical analytical solutions, and highly accurate numerical solutions. In addition, this paper presents recommendations for the design and use of validation benchmarks, highlighting the careful design of building-block experiments, the estimation of experimental measurement uncertainty for both inputs and outputs to the code, validation metrics, and the role of model calibration in validation. It is argued that the understanding of predictive capability of a computational model is built on the

Verification and validation benchmarks

International Nuclear Information System (INIS)

Oberkampf, William L.; Trucano, Timothy G.

2008-01-01

Verification and validation (V and V) are the primary means to assess the accuracy and reliability of computational simulations. V and V methods and procedures have fundamentally improved the credibility of simulations in several high-consequence fields, such as nuclear reactor safety, underground nuclear waste storage, and nuclear weapon safety. Although the terminology is not uniform across engineering disciplines, code verification deals with assessing the reliability of the software coding, and solution verification deals with assessing the numerical accuracy of the solution to a computational model. Validation addresses the physics modeling accuracy of a computational simulation by comparing the computational results with experimental data. Code verification benchmarks and validation benchmarks have been constructed for a number of years in every field of computational simulation. However, no comprehensive guidelines have been proposed for the construction and use of V and V benchmarks. For example, the field of nuclear reactor safety has not focused on code verification benchmarks, but it has placed great emphasis on developing validation benchmarks. Many of these validation benchmarks are closely related to the operations of actual reactors at near-safety-critical conditions, as opposed to being more fundamental-physics benchmarks. This paper presents recommendations for the effective design and use of code verification benchmarks based on manufactured solutions, classical analytical solutions, and highly accurate numerical solutions. In addition, this paper presents recommendations for the design and use of validation benchmarks, highlighting the careful design of building-block experiments, the estimation of experimental measurement uncertainty for both inputs and outputs to the code, validation metrics, and the role of model calibration in validation. It is argued that the understanding of predictive capability of a computational model is built on the

New Improved Nuclear Data for Nuclear Criticality and Safety

International Nuclear Information System (INIS)

Guber, Klaus H.; Leal, Luiz C.; Lampoudis, C.; Kopecky, S.; Schillebeeckx, P.; Emiliani, F.; Wynants, R.; Siegler, P.

2011-01-01

The Geel Electron Linear Accelerator (GELINA) was used to measure neutron total and capture cross sections of 182,183,184,186 W and 63,65 Cu in the energy range from 100 eV to ∼200 keV using the time-of-flight method. GELINA is the only high-power white neutron source with excellent timing resolution and ideally suited for these experiments. Concerns about the use of existing cross-section data in nuclear criticality calculations using Monte Carlo codes and benchmarks were a prime motivator for the new cross-section measurements. To support the Nuclear Criticality Safety Program, neutron cross-section measurements were initiated using GELINA at the EC-JRC-IRMM. Concerns about data deficiencies in some existing cross-section evaluations from libraries such as ENDF/B, JEFF, or JENDL for nuclear criticality calculations were the prime motivator for new cross-section measurements. Over the past years many troubles with existing nuclear data have emerged, such as problems related to proper normalization, neutron sensitivity backgrounds, poorly characterized samples, and use of improper pulse-height weighting functions. These deficiencies may occur in the resolved- and unresolved-resonance region and may lead to erroneous nuclear criticality calculations. An example is the use of the evaluated neutron cross-section data for tungsten in nuclear criticality safety calculations, which exhibit discrepancies in benchmark calculations and show the need for reliable covariance data. We measured the neutron total and capture cross sections of 182,183,184,186 W and 63,65 Cu in the neutron energy range from 100 eV to several hundred keV. This will help to improve the representation of the cross sections since most of the available evaluated data rely only on old measurements. Usually these measurements were done with poor experimental resolution or only over a very limited energy range, which is insufficient for the current application.

Criticality Model

International Nuclear Information System (INIS)

Alsaed, A.

2004-01-01

The ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2003) presents the methodology for evaluating potential criticality situations in the monitored geologic repository. As stated in the referenced Topical Report, the detailed methodology for performing the disposal criticality analyses will be documented in model reports. Many of the models developed in support of the Topical Report differ from the definition of models as given in the Office of Civilian Radioactive Waste Management procedure AP-SIII.10Q, ''Models'', in that they are procedural, rather than mathematical. These model reports document the detailed methodology necessary to implement the approach presented in the Disposal Criticality Analysis Methodology Topical Report and provide calculations utilizing the methodology. Thus, the governing procedure for this type of report is AP-3.12Q, ''Design Calculations and Analyses''. The ''Criticality Model'' is of this latter type, providing a process evaluating the criticality potential of in-package and external configurations. The purpose of this analysis is to layout the process for calculating the criticality potential for various in-package and external configurations and to calculate lower-bound tolerance limit (LBTL) values and determine range of applicability (ROA) parameters. The LBTL calculations and the ROA determinations are performed using selected benchmark experiments that are applicable to various waste forms and various in-package and external configurations. The waste forms considered in this calculation are pressurized water reactor (PWR), boiling water reactor (BWR), Fast Flux Test Facility (FFTF), Training Research Isotope General Atomic (TRIGA), Enrico Fermi, Shippingport pressurized water reactor, Shippingport light water breeder reactor (LWBR), N-Reactor, Melt and Dilute, and Fort Saint Vrain Reactor spent nuclear fuel (SNF). The scope of this analysis is to document the criticality computational method. The criticality

Russian system of computerized analysis for licensing at atomic industry (SCALA) and its validation on ICSBEP handbook data and some burnup calculations

International Nuclear Information System (INIS)

Ivanova, T.; Nikolaev, M.; Polyakov, A.; Saraeva, T.; Tsiboulia, A.

2000-01-01

The System of Computerized Analysis for Licensing at Atomic industry (SCALA) is a Russian analogue of the well-known SCALE system. For criticality evaluations the ABBN-93 system is used with TWODANT and with joined American KENO and Russian MMK Monte-Carlo code MMKKENO. Using the same cross sections and input models, all these codes give results that coincide within the statistical uncertainties (for Monte-Carlo codes). Validation of criticality calculations using SCALA was performed using data presented in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. Another task of the work was to test the burnup capability of SCALA system in complex geometry in compare with other codes. Benchmark models of VVER type reactor assemblies with UO 2 and MOX fuel including the cases with burnable gadolinium absorbers were calculated. KENO-VI and MMK codes were used for power distribution calculations, ORIGEN code was used for the isotopic kinetics calculations. (authors)

Criticality Calculations for a Typical Nuclear Fuel Fabrication Plant with Low Enriched Uranium

International Nuclear Information System (INIS)

Elsayed, Hade; Nagy, Mohamed; Agamy, Said; Shaat, Mohmaed

2013-01-01

The operations with the fissile materials such as U 235 introduce the risk of a criticality accident that may be lethal to nearby personnel and can lead the facility to shutdown. Therefore, the prevention of a nuclear criticality accident should play a major role in the design of a nuclear facility. The objectives of criticality safety are to prevent a self-sustained nuclear chain reaction and to minimize the consequences. Sixty criticality accidents were occurred in the world. These are accidents divided into two categories, 22 accidents occurred in process facilities and 38 accidents occurred during critical experiments or operations with research reactor. About 21 criticality accidents including Japan Nuclear Fuel Conversion Co. (JCO) accident took place with fuel solution or slurry and only one accident occurred with metal fuel. In this study the nuclear criticality calculations have been performed for a typical nuclear fuel fabrication plant producing nuclear fuel elements for nuclear research reactors with low enriched uranium up to 20%. The calculations were performed for both normal and abnormal operation conditions. The effective multiplication factor (k eff ) during the nuclear fuel fabrication process (Uranium hexafluoride - Ammonium Diuranate conversion process) was determined. Several accident scenarios were postulated and the criticalities of these accidents were evaluated. The computer code MCNP-4B which based on Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations were performed for the cases of, change of moderator to fuel ratio, solution density and concentration of the solute in order to prevent or mitigate criticality accidents during the nuclear fuel fabrication process. The calculation results are analyzed and discussed

Calculation of neutron importance function in fissionable assemblies using Monte Carlo method

International Nuclear Information System (INIS)

Feghhi, S.A.H.; Shahriari, M.; Afarideh, H.

2007-01-01

The purpose of the present work is to develop an efficient solution method for the calculation of neutron importance function in fissionable assemblies for all criticality conditions, based on Monte Carlo calculations. The neutron importance function has an important role in perturbation theory and reactor dynamic calculations. Usually this function can be determined by calculating the adjoint flux while solving the adjoint weighted transport equation based on deterministic methods. However, in complex geometries these calculations are very complicated. In this article, considering the capabilities of MCNP code in solving problems with complex geometries and its closeness to physical concepts, a comprehensive method based on the physical concept of neutron importance has been introduced for calculating the neutron importance function in sub-critical, critical and super-critical conditions. For this propose a computer program has been developed. The results of the method have been benchmarked with ANISN code calculations in 1 and 2 group modes for simple geometries. The correctness of these results has been confirmed for all three criticality conditions. Finally, the efficiency of the method for complex geometries has been shown by the calculation of neutron importance in Miniature Neutron Source Reactor (MNSR) research reactor

Spin-orbit couplings within the equation-of-motion coupled-cluster framework: Theory, implementation, and benchmark calculations

Energy Technology Data Exchange (ETDEWEB)

Epifanovsky, Evgeny [Department of Chemistry, University of Southern California, Los Angeles, California 90089-0482 (United States); Department of Chemistry, University of California, Berkeley, California 94720 (United States); Q-Chem Inc., 6601 Owens Drive, Suite 105, Pleasanton, California 94588 (United States); Klein, Kerstin; Gauss, Jürgen [Institut für Physikalische Chemie, Universität Mainz, D-55099 Mainz (Germany); Stopkowicz, Stella [Department of Chemistry, Centre for Theoretical and Computational Chemistry, University of Oslo, N-0315 Oslo (Norway); Krylov, Anna I. [Department of Chemistry, University of Southern California, Los Angeles, California 90089-0482 (United States)

2015-08-14

We present a formalism and an implementation for calculating spin-orbit couplings (SOCs) within the EOM-CCSD (equation-of-motion coupled-cluster with single and double substitutions) approach. The following variants of EOM-CCSD are considered: EOM-CCSD for excitation energies (EOM-EE-CCSD), EOM-CCSD with spin-flip (EOM-SF-CCSD), EOM-CCSD for ionization potentials (EOM-IP-CCSD) and electron attachment (EOM-EA-CCSD). We employ a perturbative approach in which the SOCs are computed as matrix elements of the respective part of the Breit-Pauli Hamiltonian using zeroth-order non-relativistic wave functions. We follow the expectation-value approach rather than the response-theory formulation for property calculations. Both the full two-electron treatment and the mean-field approximation (a partial account of the two-electron contributions) have been implemented and benchmarked using several small molecules containing elements up to the fourth row of the periodic table. The benchmark results show the excellent performance of the perturbative treatment and the mean-field approximation. When used with an appropriate basis set, the errors with respect to experiment are below 5% for the considered examples. The findings regarding basis-set requirements are in agreement with previous studies. The impact of different correlation treatment in zeroth-order wave functions is analyzed. Overall, the EOM-IP-CCSD, EOM-EA-CCSD, EOM-EE-CCSD, and EOM-SF-CCSD wave functions yield SOCs that agree well with each other (and with the experimental values when available). Using an EOM-CCSD approach that provides a more balanced description of the target states yields more accurate results.

Benchmark for evaluation and validation of reactor simulations (BEAVRS)

Energy Technology Data Exchange (ETDEWEB)

Horelik, N.; Herman, B.; Forget, B.; Smith, K. [Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 77 Massachusetts Avenue, Cambridge, MA 02139 (United States)

2013-07-01

Advances in parallel computing have made possible the development of high-fidelity tools for the design and analysis of nuclear reactor cores, and such tools require extensive verification and validation. This paper introduces BEAVRS, a new multi-cycle full-core Pressurized Water Reactor (PWR) depletion benchmark based on two operational cycles of a commercial nuclear power plant that provides a detailed description of fuel assemblies, burnable absorbers, in-core fission detectors, core loading patterns, and numerous in-vessel components. This benchmark enables analysts to develop extremely detailed reactor core models that can be used for testing and validation of coupled neutron transport, thermal-hydraulics, and fuel isotopic depletion. The benchmark also provides measured reactor data for Hot Zero Power (HZP) physics tests, boron letdown curves, and three-dimensional in-core flux maps from fifty-eight instrumented assemblies. Initial comparisons between calculations performed with MIT's OpenMC Monte Carlo neutron transport code and measured cycle 1 HZP test data are presented, and these results display an average deviation of approximately 100 pcm for the various critical configurations and control rod worth measurements. Computed HZP radial fission detector flux maps also agree reasonably well with the available measured data. All results indicate that this benchmark will be extremely useful in validation of coupled-physics codes and uncertainty quantification of in-core physics computational predictions. The detailed BEAVRS specification and its associated data package is hosted online at the MIT Computational Reactor Physics Group web site (http://crpg.mit.edu/), where future revisions and refinements to the benchmark specification will be made publicly available. (authors)

Calculation of neutron importance function in fissionable assemblies using Monte Carlo method

International Nuclear Information System (INIS)

Feghhi, S. A. H.; Afarideh, H.; Shahriari, M.

2007-01-01

The purpose of the present work is to develop an efficient solution method to calculate neutron importance function in fissionable assemblies for all criticality conditions, using Monte Carlo Method. The neutron importance function has a well important role in perturbation theory and reactor dynamic calculations. Usually this function can be determined by calculating adjoint flux through out solving the Adjoint weighted transport equation with deterministic methods. However, in complex geometries these calculations are very difficult. In this article, considering the capabilities of MCNP code in solving problems with complex geometries and its closeness to physical concepts, a comprehensive method based on physical concept of neutron importance has been introduced for calculating neutron importance function in sub-critical, critical and supercritical conditions. For this means a computer program has been developed. The results of the method has been benchmarked with ANISN code calculations in 1 and 2 group modes for simple geometries and their correctness has been approved for all three criticality conditions. Ultimately, the efficiency of the method for complex geometries has been shown by calculation of neutron importance in MNSR research reactor

Comparison of results from the MCNP criticality validation suite using ENDF/B-VI and preliminary ENDF/B-VII nuclear data

Energy Technology Data Exchange (ETDEWEB)

Mosteller, R. D. (Russell D.)

2004-01-01

The MCNP Criticality Validation Suite is a collection of 31 benchmarks taken from the International Handbook of Evaluated Criticality Safety Benchmark Experiments. MCNP5 calculations clearly demonstrate that, overall, nuclear data for a preliminary version of ENDFB-VII produce better agreement with the benchmarks in the suite than do corresponding data from ENDF/B-VI. Additional calculations identify areas where improvements in the data still are needed. Based on results for the MCNP Criticality Validation Suite, the Pre-ENDF/B-VII nuclear data produce substantially better overall results than do their ENDF/B-VI counterparts. The calculated values for k{sub eff} for bare metal spheres and for an IEU cylinder reflected by normal uranium are in much better agreement with the benchmark values. In addition, the values of k{sub eff} for the bare metal spheres are much more consistent with those for corresponding metal spheres reflected by normal uranium or water. In addition, a long-standing controversy about the need for an ad hoc adjustment to the {sup 238}U resonance integral for thermal systems may finally be resolved. On the other hand, improvements still are needed in a number of areas. Those areas include intermediate-energy cross sections for {sup 235}U, angular distributions for elastic scattering in deuterium, and fast cross sections for {sup 237}Np.

Implications of Monte Carlo Statistical Errors in Criticality Safety Assessments

International Nuclear Information System (INIS)

Pevey, Ronald E.

2005-01-01

Most criticality safety calculations are performed using Monte Carlo techniques because of Monte Carlo's ability to handle complex three-dimensional geometries. For Monte Carlo calculations, the more histories sampled, the lower the standard deviation of the resulting estimates. The common intuition is, therefore, that the more histories, the better; as a result, analysts tend to run Monte Carlo analyses as long as possible (or at least to a minimum acceptable uncertainty). For Monte Carlo criticality safety analyses, however, the optimization situation is complicated by the fact that procedures usually require that an extra margin of safety be added because of the statistical uncertainty of the Monte Carlo calculations. This additional safety margin affects the impact of the choice of the calculational standard deviation, both on production and on safety. This paper shows that, under the assumptions of normally distributed benchmarking calculational errors and exact compliance with the upper subcritical limit (USL), the standard deviation that optimizes production is zero, but there is a non-zero value of the calculational standard deviation that minimizes the risk of inadvertently labeling a supercritical configuration as subcritical. Furthermore, this value is shown to be a simple function of the typical benchmarking step outcomes--the bias, the standard deviation of the bias, the upper subcritical limit, and the number of standard deviations added to calculated k-effectives before comparison to the USL

Benchmark critical experiments on low-enriched uranium oxide systems with H/U = 0.77

International Nuclear Information System (INIS)

Tuck, G.; Oh, I.

1979-08-01

Ten benchmark experiments were performed at the Critical Mass Laboratory at Rockwell International's Rocky Flats Plant, Golden, Colorado, for the US Nuclear Regulatory Commission. They provide accurate criticality data for low-enriched damp uranium oxide (U 3 O 8 ) systems. The core studied consisted of 152 mm cubical aluminum cans containing an average of 15,129 g of low-enriched (4.46% 235 U) uranium oxide compacted to a density of 4.68 g/cm 3 and with an H/U atomic ratio of 0.77. One hundred twenty five (125) of these cans were arranged in an approx. 770 mm cubical array. Since the oxide alone cannot be made critical in an array of this size, an enriched (approx. 93% 235 U) metal or solution driver was used to achieve criticality. Measurements are reported for systems having the least practical reflection and for systems reflected by approx. 254-mm-thick concrete or plastic. Under the three reflection conditions, the mass of the uranium metal driver ranged from 29.87 kg to 33.54 kg for an oxide core of 1864.6 kg. For an oxide core of 1824.9 kg, the weight of the high concentration (351.2 kg U/m 3 ) solution driver varied from 14.07 kg to 16.14 kg, and the weight of the low concentration (86.4 kg U/m 3 ) solution driver from 12.4 kg to 14.0 kg

Criticality calculations with MCNP trademark: A primer

International Nuclear Information System (INIS)

Harmon, C.D. II; Busch, R.D.; Briesmeister, J.F.; Forster, R.A.

1994-01-01

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

Benchmark calculations on fluid coupled co-axial cylinders typical to LMFBR structures

International Nuclear Information System (INIS)

Dostal, M.; Descleve, P.; Gantenbein, F.; Lazzeri, L.

1983-01-01

This paper describes a joint effort promoted and funded by the Commission of European Community under the umbrella of Fast Reactor Co-ordinating Committee and working group on Codes and Standards No. 2 with the purpose to test several programs currently used for dynamic analysis of fluid-coupled structures. The scope of the benchmark calculations is limited to beam type modes of vibration, small displacement of the structures and small pressure variation such as encountered in seismic or flow induced vibration problems. Five computer codes were used: ANSYS, AQUAMODE, NOVAX, MIAS/SAP4 and ZERO where each program employs a different structural-fluid formulation. The calculations were performed for four different geometrical configurations of concentric cylinders where the effect of gap size, water level, and support conditions were considered. The analytical work was accompanied by experiments carried out on a purpose-built rig. The test rig consisted of two concentric cylinders independently supported on flexible cantilevers. A geometrical simplicity and attention in the rig design to eliminate the structural coupling between the cylinders lead to unambiguous test results. Only the beam natural frequencies, in phase and out of phase were measured. The comparison of different analytical methods and experimental results is presented and discussed. The degree of agreement varied between very good and unacceptable. (orig./GL)

Benchmarking of human resources management

Directory of Open Access Journals (Sweden)

David M. Akinnusi

2008-11-01

Full Text Available This paper reviews the role of human resource management (HRM which, today, plays a strategic partnership role in management. The focus of the paper is on HRM in the public sector, where much hope rests on HRM as a means of transforming the public service and achieving much needed service delivery. However, a critical evaluation of HRM practices in the public sector reveals that these services leave much to be desired. The paper suggests the adoption of benchmarking as a process to revamp HRM in the public sector so that it is able to deliver on its promises. It describes the nature and process of benchmarking and highlights the inherent difficulties in applying benchmarking in HRM. It concludes with some suggestions for a plan of action. The process of identifying “best” practices in HRM requires the best collaborative efforts of HRM practitioners and academicians. If used creatively, benchmarking has the potential to bring about radical and positive changes in HRM in the public sector. The adoption of the benchmarking process is, in itself, a litmus test of the extent to which HRM in the public sector has grown professionally.

Qualification of coupled 3D neutron kinetic/thermal hydraulic code systems by the calculation of a VVER-440 benchmark. Re-connection of an isolated loop

Energy Technology Data Exchange (ETDEWEB)

Kotsarev, Alexander; Lizorkin, Mikhail [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation); Bencik, Marek; Hadek, Jan [UJV Rez, a.s., Rez (Czech Republic); Kozmenkov, Yaroslav; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany)

2016-09-15

The 7th AER dynamic benchmark is a continuation of the efforts to validate the codes systematically for the estimation of the transient behavior of VVER type nuclear power plants. The main part of the benchmark is the simulation of the re-connection of an isolated circulation loop with low temperature in a VVER-440 plant. This benchmark was calculated by the National Research Centre ''Kurchatov Institute'' (with the code ATHLET/BIPR-VVER), UJV Rez (with the code RELAP5-3D {sup copyright}) and HZDR (with the code DYN3D/ATHLET). The paper gives an overview of the behavior of the main thermal hydraulic and neutron kinetic parameters in the provided solutions.

INTEGRAL BENCHMARK DATA FOR NUCLEAR DATA TESTING THROUGH THE ICSBEP AND THE NEWLY ORGANIZED IRPHEP

Energy Technology Data Exchange (ETDEWEB)

J. Blair Briggs; Lori Scott; Yolanda Rugama; Enrico Satori

2007-04-01

The status of the International Criticality Safety Benchmark Evaluation Project (ICSBEP) was last reported in a nuclear data conference at the International Conference on Nuclear Data for Science and Technology, ND-2004, in Santa Fe, New Mexico. Since that time the number and type of integral benchmarks have increased significantly. Included in the ICSBEP Handbook are criticality-alarm / shielding and fundamental physic benchmarks in addition to the traditional critical / subcritical benchmark data. Since ND 2004, a reactor physics counterpart to the ICSBEP, the International Reactor Physics Experiment Evaluation Project (IRPhEP) was initiated. The IRPhEP is patterned after the ICSBEP, but focuses on other integral measurements, such as buckling, spectral characteristics, reactivity effects, reactivity coefficients, kinetics measurements, reaction-rate and power distributions, nuclide compositions, and other miscellaneous-type measurements in addition to the critical configuration. The status of these two projects is discussed and selected benchmarks highlighted in this paper.

Criticality criteria for submissions based on calculations

International Nuclear Information System (INIS)

Burgess, M.H.

1975-06-01

Calculations used in criticality clearances are subject to errors from various sources, and allowance must be made for these errors is assessing the safety of a system. A simple set of guidelines is defined, drawing attention to each source of error, and recommendations as to its application are made. (author)

Calculation of HTR-10 first criticality with MVP

International Nuclear Information System (INIS)

Xie Jiachun; Yao Lianying

2015-01-01

The first criticality of 10 MW pebble-bed high temperature gas-cooled reactor-test module (HTR-10) was calculated with MVP. According to the characteristics of HTR-10, the Statistical Geometry Model of MVP was employed to describe the random arrangement of coated fuel particles in the fuel pebbles and the random distribution of the fuel and dummy pebbles in the core. Compared with previous results from VSOP and MCNP, the MVP results with JENDL-3.3 library were little more different, but the results with ENDF/B-Ⅵ.8 library were very close. The relative errors were less than 0.7%, compared with the first criticality experimental results. The study shows that MVP could be used in the physics calculations for pebble bed high temperature gas-cooled reactors. (authors)

Spectrum integrated (n,He) cross section comparison and least squares analysis for /sup 6/Li and /sup 10/B in benchmark fields

International Nuclear Information System (INIS)

Schenter, R.E.; Oliver, B.M.; Farrar, H. IV

1987-01-01

Spectrum integrated cross sections for /sup 6/Li and /sup 10/B from five benchmark fast reactor neutron fields are compared with calculated values obtained using the ENDF/B-V Cross Section Files. The benchmark fields include the Coupled Fast Reactivity Measurements Facility (CFRMF) at the Idaho National Engineering Laboratory, the 10% Enriched U-235 Critical Assembly (BIG-10) at Los Alamos National Laboratory, the Sigma Sigma and Fission Cavity fields of the BR-1 reactor at CEN/SCK, and the Intermediate-Energy Standard Neutron Field (ISNF) at the National Bureau of Standards. Results from least square analyses using the FERRET computer code to obtain adjusted cross section values and their uncertainties are presented. Input to these calculations include the above five benchmark data sets. These analyses indicate a need for revision in the ENDF/B-V files for the /sup 10/B cross section for energies above 50 keV

Spectral measurements in critical assemblies: MCNP specifications and calculated results

Energy Technology Data Exchange (ETDEWEB)

Stephanie C. Frankle; Judith F. Briesmeister

1999-12-01

Recently, a suite of 86 criticality benchmarks for the Monte Carlo N-Particle (MCNP) transport code was developed, and the results of testing the ENDF/B-V and ENDF/B-VI data (through Release 2) were published. In addition to the standard k{sub eff} measurements, other experimental measurements were performed on a number of these benchmark assemblies. In particular, the Cross Section Evaluation Working Group (CSEWG) specifications contain experimental data for neutron leakage and central-flux measurements, central-fission ratio measurements, and activation ratio measurements. Additionally, there exists another set of fission reaction-rate measurements performed at the National Institute of Standards and Technology (NIST) utilizing a {sup 252}Cf source. This report will describe the leakage and central-flux measurements and show a comparison of experimental data to MCNP simulations performed using the ENDF/B-V and B-VI (Release 2) data libraries. Central-fission and activation reaction-rate measurements will be described, and the comparison of experimental data to MCNP simulations using available data libraries for each reaction of interest will be presented. Finally, the NIST fission reaction-rate measurements will be described. A comparison of MCNP results published previously with the current MCNP simulations will be presented for the NIST measurements, and a comparison of the current MCNP simulations to the experimental measurements will be presented.