Status of multigroup cross-section data for shielding applications

International Nuclear Information System (INIS)

Roussin, R.W.; Maskewitz, B.F.; Trubey, D.K.

1983-01-01

Multigroup cross-section libraries for shielding applications in formats for direct use in discrete ordinates or Monte Carlo codes have long been a part of the Data Library Collection (DLC) of the Radiation Shielding Information Center (RSIC). In recent years libraries in more flexible and comprehensive formats, which allow the user to derive his own problem-dependent sets, have been added to the collection. The current status of both types is described, as well as projections for adding data libraries based on ENDF/B-V

Correction of multigroup cross sections for resolved resonance interference in mixed absorbers

International Nuclear Information System (INIS)

Williams, M.L.

1982-07-01

The effect that interference between resolved resonances has on averaging multigroup cross sections is examined for thermal reactor-type problems. A simple and efficient numerical scheme is presented to correct a preprocessed multigroup library for interference effects. The procedure is implemented in a design oriented lattice physics computer code and compared with rigorous numerical calculations. The approximate method for computing resonance interference correction factors is applied to obtaining fine-group cross sections for a homogeneous uranium-plutonium mixture and a uranium oxide lattice. It was found that some fine group cross sections are changed by more than 40% due to resonance interference. The change in resonance interference correction factors due to burnup of a PWR fuel pin is examined and found to be small. The effect of resolved resonance interference on collapsed broad-group cross sections for thermal reactor calculations is discussed

ZZ ANSLV, Multigroup Cross Sections Library for ANS Reactor Design Studies

International Nuclear Information System (INIS)

2000-01-01

A - Description of program or function: - Format: AMPX Master Interface Library format. Number of groups: Fine Group (99 energy groups) General Purpose Neutron Library. Materials: H, He, Be, B, Graphite, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Kr, Zr, Mo, Tc, Ru, Ag, Cd, Cs, Ce, Pr, Pm, Sm, Eu, Hf, Ta, U, C, F, Cu, Sn, Pb, Rh, I, Xe, Nd, Th, Np, Pu, Am, Cm, Bk, Cf, Es, MAFP, WAFP. Origin: ENDF/B-V. - Format: AMPX Master Interface Library format. Number of groups: Broad Group (39 energy groups) General Purpose Neutron Library. Materials: H, He, Be, B, Graphite, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Kr, Zr, Mo, Tc, Ru, Ag, Cd, Cs, Ce, Pr, Pm, Sm, Eu, Hf, Ta, U, C, F, Cu, Sn, Pb, Rh, I, Xe, Nd, Th, Np, Pu, Am, Cm, Bk, Cf, Es, MAFP, WAFP. Origin: ENDF/B-V. - Format: AMPX Master Interface Library format. Number of groups: Gamma-Ray Interaction (GRI) Library in 44-groups. Materials: H, He, Be, B, C, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zr, Mo, Ag, Cd, Xe, Sm, Eu, Hf, Ta, Ir, Pb, Th, U, Pu. Origin: ENDF/B-V; LENDL-V evaluations for 12 materials. - Format: AMPX Master Interface Library format. Number of groups: Coupled Library containing (CNG) 99-group neutron and 44-group gamma-ray data. Materials: H, Be, B, C, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zr, Mo, Ag, Cd, Eu, Hf, Ta, Pb, Th, U, Pu. Origin: ENDF/B-V. - Format: AMPX Master Interface Library format. Number of groups: Coupled neutron-gamma (CNG) Library containing 39-group, and 44-group gamma-ray data. Materials: H, Be, B, C, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zr, Mo, Ag, Cd, Eu, Hf, Ta, Pb, Th, U, Pu. Origin: ENDF/B-V. Weighting spectrum: Maxwellian 300 K + 1/(E*sigma-total) + fission spectrum4 types of boundaries have been used depending isotope and library type (see report). Pseudo-problem-independent, multigroup cross section libraries were generated to support the Advanced Neutron source (ANS) reactor design studies. The ANS was

Generation and performance of a multigroup coupled neutron-gamma cross-section library for deterministic and Monte Carlo borehole logging analysis

International Nuclear Information System (INIS)

Kodeli, I.; Aldama, D. L.; De Leege, P. F. A.; Legrady, D.; Hoogenboom, J. E.; Cowan, P.

2004-01-01

As part of the IRTMBA (Improved Radiation Transport Modelling for Borehole Applications) project of the EU community's 5. framework program a special purpose multigroup cross-section library was prepared for use in deterministic and Monte Carlo oil well logging particle transport calculations. This library is expected to improve the prediction of the neutron and gamma spectra at the detector positions of the logging tool, and their use for the interpretation of the neutron logging measurements was studied. Preparation and testing of this library is described. (authors)

MC²-3: Multigroup Cross Section Generation Code for Fast Reactor Analysis

Energy Technology Data Exchange (ETDEWEB)

Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Yang, W. S. [Argonne National Lab. (ANL), Argonne, IL (United States)

2013-11-08

The MC²-3 code is a Multigroup Cross section generation Code for fast reactor analysis, developed by improving the resonance self-shielding and spectrum calculation methods of MC²-2 and integrating the one-dimensional cell calculation capabilities of SDX. The code solves the consistent P1 multigroup transport equation using basic neutron data from ENDF/B data files to determine the fundamental mode spectra for use in generating multigroup neutron cross sections. A homogeneous medium or a heterogeneous slab or cylindrical unit cell problem is solved in ultrafine (~2000) or hyperfine (~400,000) group levels. In the resolved resonance range, pointwise cross sections are reconstructed with Doppler broadening at specified isotopic temperatures. The pointwise cross sections are directly used in the hyperfine group calculation whereas for the ultrafine group calculation, self-shielded cross sections are prepared by numerical integration of the pointwise cross sections based upon the narrow resonance approximation. For both the hyperfine and ultrafine group calculations, unresolved resonances are self-shielded using the analytic resonance integral method. The ultrafine group calculation can also be performed for two-dimensional whole-core problems to generate region-dependent broad-group cross sections. Multigroup cross sections are written in the ISOTXS format for a user-specified group structure. The code is executable on UNIX, Linux, and PC Windows systems, and its library includes all isotopes of the ENDF/BVII. 0 data.

RGENDF - An interface program between the NJOY code and codes using multigroup cross-sections

International Nuclear Information System (INIS)

Chalhoub, E.S.; Anaf, J.

1988-02-01

An interface program for reformatting multigroup cross-section libraries generated by NJOY into ENDF/B-V format and the EXPANDA, PFCOND and COMPAR input formats is presented. (author). 7 refs, 1 fig., 1 tab

AMPX: a modular code system for generating coupled multigroup neutron-gamma libraries from ENDF/B

Energy Technology Data Exchange (ETDEWEB)

Greene, N.M.; Lucius, J.L.; Petrie, L.M.; Ford, W.E. III; White, J.E.; Wright, R.Q.

1976-03-01

AMPX is a modular system for producing coupled multigroup neutron-gamma cross section sets. Basic neutron and gamma cross-section data for AMPX are obtained from ENDF/B libraries. Most commonly used operations required to generate and collapse multigroup cross-section sets are provided in the system. AMPX is flexibly dimensioned; neutron group structures, and gamma group structures, and expansion orders to represent anisotropic processes are all arbitrary and limited only by available computer core and budget. The basic processes provided will (1) generate multigroup neutron cross sections; (2) generate multigroup gamma cross sections; (3) generate gamma yields for gamma-producing neutron interactions; (4) combine neutron cross sections, gamma cross sections, and gamma yields into final ''coupled sets''; (5) perform one-dimensional discrete ordinates transport or diffusion theory calculations for neutrons and gammas and, on option, collapse the cross sections to a broad-group structure, using the one-dimensional results as weighting functions; (6) plot cross sections, on option, to facilitate the ''evaluation'' of a particular multigroup set of data; (7) update and maintain multigroup cross section libraries in such a manner as to make it not only easy to combine new data with previously processed data but also to do it in a single pass on the computer; and (8) output multigroup cross sections in convenient formats for other codes. (auth)

12G: code for conversion of isotope-ordered cross-section libraries into group-ordered cross-section libraries

International Nuclear Information System (INIS)

Resnik, W.M. II; Bosler, G.E.

1977-09-01

Many current reactor physics codes accept cross-section libraries in an isotope-ordered form, convert them with internal preprocessing routines to a group-ordered form, and then perform calculations using these group-ordered data. Occasionally, because of storage and time limitations, the preprocessing routines in these codes cannot convert very large multigroup isotope-ordered libraries. For this reason, the I2G code, i.e., ISOTXS to GRUPXS, was written to convert externally isotope-ordered cross section libraries in the standard file format called ISOTXS to group-ordered libraries in the standard format called GRUPXS. This code uses standardized multilevel data management routines which establish a strategy for the efficient conversion of large libraries. The I2G code is exportable contingent on access to, and an intimate familiarization with, the multilevel routines. These routines are machine dependent, and therefore must be provided by the importing facility. 6 figures, 3 tables

Multigroup cross section collapsing optimization of a He-3 detector assembly model using deterministic transport techniques

International Nuclear Information System (INIS)

Huang, Mi; Yi, Ce; Manalo, Kevin L.; Sjoden, Glenn E.

2011-01-01

Multigroup optimization is performed on a neutron detector assembly to examine the validity of transport response in forward and adjoint modes. For SN transport simulations, we discuss the multigroup collapse of an 80 group library to 40, 30, and 16 groups, constructed from using the 3-D parallel PENTRAN and macroscopic cross section collapsing with YGROUP contribution weighting. The difference in using P_1 and P_3 Legendre order in scattering cross sections is investigated; also, associated forward and adjoint transport responses are calculated. We conclude that for the block analyzed, a 30 group cross section optimizes both computation time and accuracy relative to the 80 group transport calculations. (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

MCNP and MATXS cross section libraries based on JENDL-3.3

International Nuclear Information System (INIS)

Kosako, Kazuaki; Konno, Chikara; Fukahori, Tokio; Shibata, Keiichi

2003-01-01

The continuous energy cross section library for the Monte Carlo transport code MCNP-4C, FSXLIB-J33, has been generated from the latest version of JENDL-3.3. The multigroup cross section library with the MATXS format, MATXS-J33, has been generated also from JENDL-3.3. Both libraries contain all nuclides in JENDL-3.3 and are processed at 300 K by the nuclear data processing system NJOY99. (author)

ESELEM 4: a code for calculating fine neutron spectrum and multi-group cross sections in plate lattice

International Nuclear Information System (INIS)

Nakagawa, Masayuki; Katsuragi, Satoru; Narita, Hideo.

1976-07-01

The multi-group treatment has been used in the design study of fast reactors and analysis of experiments at fast critical assemblies. The accuracy of the multi-group cross sections therefore affects strongly the results of these analyses. The ESELEM 4 code has been developed to produce multi-group cross sections with an advanced method from the nuclear data libraries used in the JAERI Fast set. ESELEM 4 solves integral transport equation by the collision probability method in plate lattice geometry to obtain the fine neutron spectrum. A typical fine group mesh width is 0.008 in lethargy unit. The multi-group cross sections are calculated by weighting the point data with the fine structure neutron flux. Some devices are applied to reduce computation time and computer core storage required for the calculation. The slowing down sources are calculated with the use of a recurrence formula derived for elastic and inelastic scattering. The broad group treatment is adopted above 2 MeV for dealing with both light any heavy elements. Also the resonance cross sections of heavy elements are represented in a broad group structure, for which we use the values of the JAERI Fast set. The library data are prepared by the PRESM code from ENDF/A type nuclear data files. The cross section data can be compactly stored in the fast computer core memory for saving the core storage and data processing time. The programme uses the variable dimensions to increase its flexibility. The users' guide for ESELEM 4 and PRESM is also presented in this report. (auth.)

Preparation of multigroup lumped fission product cross-sections from ENDF/B-VI for FBRs

International Nuclear Information System (INIS)

Devan, K.; Gopalakrishnan, V.; Mohanakrishnan, P.; Sridharan, M.S.

1997-01-01

Multigroup pseudo fission product cross-sections were computed from the American evaluated nuclear data library ENDF/B-VI, corresponding to various burnups of the proposed 500 MWe prototype fast breeder reactor (PFBR), in India. The data were derived from the cross-sections of 111 selected fission products that account for almost complete capture of fission products in an FBR. The dependence of burnup on the pseudo fission product cross-sections, and comparison with other data sets, viz. JNDC, ENDF/B-IV and ABBN, are discussed. (author)

FENDL/MG. Library of multigroup cross sections in GENDF and MATXS format for neutron-photon transport calculations. Version 1.1 of March 1995. Summary documentation

International Nuclear Information System (INIS)

Pashchenko, A.B.; Wienke, H.; Ganesan, S.

1996-01-01

Selected neutron reaction nuclear data evaluations and photon-atomic interaction cross section libraries for elements of interest to the IAEA's program on Fusion Evaluated Nuclear Data Library (FENDL) have been processed into GENDF and MATXS format using the NJOY system by R.E. MacFarlane, in VITAMIN-J group structure with VITAMIN-E weighting spectrum. This document summarizes the resulting multigroup data library FENDL/MG version 1.1. The data are available costfree, upon request from the IAEA Nuclear Data Section, online or on magnetic tape. (author). 7 refs, 1 tab

Research of the application of multi-group libraries based on ENDF/B-VII library in the reactor design

International Nuclear Information System (INIS)

Mi Aijun; Li Junjie

2010-01-01

In this paper the multi-group libraries were constructed by processing ENDF/B-VII neutron incident files into multi-group structure, and the application of the multi-group libraries in the pressurized-water reactor(PWR) design was studied. The construction of the multi-group library is realized by using the NJOY nuclear data processing system. The code can process the neutron cross section files form ENDF format to MATXS format which was required in SN code. Two dimension transport theory code of discrete ordinates DORT was used to verify the multi-group libraries and the method of the construction by comparing calculations for some representative benchmarks. We made the PWR shielding calculation by using the multi-group libraries and studied the influence of the parameters involved during the construction of the libraries such as group structure, temperatures and weight functions on the shielding design of the PWR. This work is the preparation for the construction of the multi-group library which will be used in PWR shielding design in engineering. (authors)

Verification and validation of multi-group library MUSE1.0 created from ENDF/B-VII.0

International Nuclear Information System (INIS)

Chen Yixue; Wu Jun; Yang Shouhai; Zhang Bin; Lu Daogang; Chen Chaobin

2010-01-01

A multi-group library set named MUSE1.0 with 172-neutron group and 42-photon group is produced based on ENDF/B-VII.0 using NJOY code. Weight function of the multi-group library set is taken from the Vitanim-e library and the max legendre order of scattering matrix is six. All the nuclides have thermal scattering data created using free-gas scattering law and 10 Bondarenko background cross sections se lected to generate the self-shielded multi-group cross sections. The final libraries have GENDF-format, MATXS-format and ACE-multi-group sub-libraries and each sub-library generated under 4 temperatures(293 K,600 K,800 K and 900 K). This paper provides a summary of the procedure to produce the library set and a detail description of the validation of the multi-group library set by several critical benchmark devices and shielding benchmark devices using MCNP code. The ability to handle the thermal neutron transport and resonance self-shielding problems are investigated specially. In the end, we draw the conclusion that the multi-group libraries produced is credible and can be used in the R and D process of Supercritical Water Reactor Design. (authors)

MUXS: a code to generate multigroup cross sections for sputtering calculations

International Nuclear Information System (INIS)

Hoffman, T.J.; Robinson, M.T.; Dodds, H.L. Jr.

1982-10-01

This report documents MUXS, a computer code to generate multigroup cross sections for charged particle transport problems. Cross sections generated by MUXS can be used in many multigroup transport codes, with minor modifications to these codes, to calculate sputtering yields, reflection coefficients, penetration distances, etc

Optimization of multi-group cross sections for fast reactor analysis

International Nuclear Information System (INIS)

Chin, M. R.; Manalo, K. L.; Edgar, C. A.; Paul, J. N.; Molinar, M. P.; Redd, E. M.; Yi, C.; Sjoden, G. E.

2013-01-01

The selection of the number of broad energy groups, collapsed broad energy group boundaries, and their associated evaluation into collapsed macroscopic cross sections from a general 238-group ENDF/B-VII library dramatically impacted the k eigenvalue for fast reactor analysis. An analysis was undertaken to assess the minimum number of energy groups that would preserve problem physics; this involved studies using the 3D deterministic transport parallel code PENTRAN, the 2D deterministic transport code SCALE6.1, the Monte Carlo based MCNP5 code, and the YGROUP cross section collapsing tool on a spatially discretized MOX fuel pin comprised of 21% PUO 2 -UO 2 with sodium coolant. The various cases resulted in a few hundred pcm difference between cross section libraries that included the 238 multi-group reference, and cross sections rendered using various reaction and adjoint weighted cross sections rendered by the YGROUP tool, and a reference continuous energy MCNP case. Particular emphasis was placed on the higher energies characteristic of fission neutrons in a fast spectrum; adjoint computations were performed to determine the average per-group adjoint fission importance for the MOX fuel pin. This study concluded that at least 10 energy groups for neutron transport calculations are required to accurately predict the eigenvalue for a fast reactor system to within 250 pcm of the 238 group case. In addition, the cross section collapsing/weighting schemes within YGROUP that provided a collapsed library rendering eigenvalues closest to the reference were the contribution collapsed, reaction rate weighted scheme. A brief analysis on homogenization of the MOX fuel pin is also provided, although more work is in progress in this area. (authors)

FCXSEC: multigroup cross-section libraries for nuclear fuel cycle shielding calculations

International Nuclear Information System (INIS)

Ford, W.E. III; Webster, C.C.; Diggs, B.R.; Pevey, R.E.; Croff, A.G.

1980-05-01

Starting with the pseudo-composition-independent VITAMIN-C cross-sectin library, composition-dependent fine-(171n-36γ) and broad-group (22n-21γ) self-shielded AMPX master, broad-group microscopic ANISN-formatted, and broad-group macroscopic ANISN-formatted cross-section libraries were generated to be used for nuclear fuel cycle shielding calculations. The specifications for the data and the procedure used to prepare the libraries are described

One-, two- and three-dimensional transport codes using multi-group double-differential form cross sections

International Nuclear Information System (INIS)

Mori, Takamasa; Nakagawa, Masayuki; Sasaki, Makoto.

1988-11-01

We have developed a group of computer codes to realize the accurate transport calculation by using the multi-group double-differential form cross section. This type of cross section can correctly take account of the energy-angle correlated reaction kinematics. Accordingly, the transport phenomena in materials with highly anisotropic scattering are accurately calculated by using this cross section. They include the following four codes or code systems: PROF-DD : a code system to generate the multi-group double-differential form cross section library by processing basic nuclear data file compiled in the ENDF / B-IV or -V format, ANISN-DD : a one-dimensional transport code based on the discrete ordinate method, DOT-DD : a two-dimensional transport code based on the discrete ordinate method, MORSE-DD : a three-dimensional transport code based on the Monte Carlo method. In addition to these codes, several auxiliary codes have been developed to process calculated results. This report describes the calculation algorithm employed in these codes and how to use them. (author)

Neutron-photon multigroup cross sections for neutron energies up to 400 MeV: HILO86R

International Nuclear Information System (INIS)

Kotegawa, Hiroshi; Nakane, Yoshihiro; Hasegawa, Akira; Tanaka, Shun-ichi

1993-02-01

A macroscopic multigroup cross section library of 66 neutron and 22 photon groups for neutron energies up to 400 MeV: HILO86R is prepared for 10 typical shielding materials; water, concrete, iron, air, graphite, polyethylene, heavy concrete, lead, aluminum and soil. The library is a revision of the DLC-119/HILO86, in which only the cross sections below 19.6 MeV have been exchanged with a group cross section processed from the JENDL-3 microscopic cross section library. In the HILO86R library, self shielding factors are used to produce effective cross sections for neutrons less than 19.6 MeV considering rather coarse energy meshes. Energy spectra and dose attenuation in water, concrete and iron have been compared among the HILO, HILO86 and HILO86R libraries for different energy neutron sources. Significant discrepancy has been observed in the energy spectra less than a couple of MeV energy in iron among the libraries, resulting large difference in the dose attenuation. The difference was attributed to the effect of self-shielding factor, namely to the difference between infinite dilution and effective cross sections. Even for 400 MeV neutron source the influence of the self-shielding factor is significant, nevertheless only the cross sections below 19.6 MeV are exchanged. (author)

Multi-Group Library Generation with Explicit Resonance Interference Using Continuous Energy Monte Carlo Calculation

Energy Technology Data Exchange (ETDEWEB)

Park, Ho Jin; Cho, Jin Young [KAERI, Daejeon (Korea, Republic of); Kim, Kang Seog [Oak Ridge National Laboratory, Oak Ridge (United States); Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

2016-05-15

In this study, multi-group cross section libraries for the DeCART code were generated using a new procedure. The new procedure includes generating the RI tables based on the MC calculations, correcting the effective fission product yield calculations, and considering most of the fission products as resonant nuclides. KAERI (Korea Atomic Energy Research Institute) has developed the transport lattice code KARMA (Kernel Analyzer by Ray-tracing Method for fuel Assembly) and DeCART (Deterministic Core Analysis based on Ray Tracing) for a multi-group neutron transport analysis of light water reactors (LWRs). These codes adopt the method of characteristics (MOC) to solve the multi-group transport equation and resonance fixed source problem, the subgroup and the direct iteration method with resonance integral tables for resonance treatment. With the development of the DeCART and KARMA code, KAERI has established its own library generation system for a multi-group transport calculation. In the KAERI library generation system, the multi-group average cross section and resonance integral (RI) table are generated and edited using PENDF (point-wise ENDF) and GENDF (group-wise ENDF) produced by the NJOY code. The new method does not need additional processing because the MC method can handle any geometry information and material composition. In this study, the new method is applied to the dominant resonance nuclide such as U{sup 235} and U{sup 238} and the conventional method is applied to the minor resonance nuclides. To examine the newly generated multi-group cross section libraries, various benchmark calculations such as pin-cell, FA, and core depletion problem are performed and the results are compared with the reference solutions. Overall, the results by the new method agree well with the reference solution. The new procedure based on the MC method were verified and provided the multi-group library that can be used in the SMR nuclear design analysis.

Nuclear cross section library for oil well logging analysis

International Nuclear Information System (INIS)

Kodeli, I.; Kitsos, S.; Aldama, D.L.; Zefran, B.

2003-01-01

As part of the IRTMBA (Improved Radiation Transport Modelling for Borehole Applications) Project of the EU Community's 5 th Programme a special purpose multigroup cross section library to be used in the deterministic (as well as Monte Carlo) oil well logging particle transport calculations was prepared. This library is expected to improve the prediction of the neutron and gamma spectra at the detector positions of the logging tool, and their use for the interpretation of the neutron logging measurements was studied. Preparation and testing of this library is described. (author)

Nuclear data processing and multigroup cross section generation

International Nuclear Information System (INIS)

Kim, Jeong Do; Kil, Chung Sub

1996-01-01

The multigroup constants for WIMS/CASMO were updated with ENDF/B-VI and were tested. The continuous energy MCNP library developed last year was validated against the LWR-simulated critical experiments. The MCNP library will be used to design and analyze nuclear and shielding facilities. The system for generation of MATXS multigroup library and TRANSX code, which is able to prepare the data for the discrete ordinates and diffusion codes from the MATXS library, was established. The MATXS libraries for analyses of thermal and fast critical experiments were generated and tested. The MATXS/TRANSX system for the discrete ordinates and diffusion codes will be useful for nuclear analyses. 10 tabs., 5 figs., 17 refs. (Author)

Status of standard cross section library and future plan

International Nuclear Information System (INIS)

Zukeran, Atsushi

2001-01-01

JSSTDL-300 multi-group cross section library with 300 neutron energy groups coupled with 104 group γ-ray cross sections was developed for general users in nuclear reactor physics and/or design, whose source data is the evaluated nuclear data library JENDL-3.2. For the purpose of a standard or common use, several famous cross section libraries worldwide used, i.e., ABBN-25, GAM-123, VITAMIN-C/J(E+C), MGCL-137, BERMUDA-12 and FNS-125 for neutron, and LANL-12, -24-, -48, and CSEWG-94 for γ-ray, are consulted about setting the common energy group structure. Furthermore, in order to expand the applicability, the top energy is set on 20 MeV and the lowest energy is 10 -5 eV. In the thermal neutron energy region, the JSSTDL-300 has about 20 energy groups. Besides, many utility codes for group collapsing and for data format transformation are provided for general users. (author)

Generation of broad-group neutron/photon cross-section libraries for shielding applications

International Nuclear Information System (INIS)

Ingersoll, D.T.; Roussin, R.W.; Fu, C.Y.; White, J.E.

1989-01-01

The generation and use of multigroup cross-section libraries with broad energy group structures is primarily for the economy of computer resources. Also, the establishment of reference broad-group libraries is desirable in order to avoid duplication of effort, both in terms of the data generation and verification, and to assure a common data base for all participants in a specific project. Uncertainties are inevitably introduced into the broad-group cross sections due to approximations in the grouping procedure. The dominant uncertainty is generally with regard to the energy weighting function used to average the pointwise or fine-group data within a single broad group. Intelligent choice of the weighting functions can reduce such uncertainties. Also, judicious selection of the energy group structure can help to reduce the sensitivity of the computed responses to the weighting function, at least for a selected set of problems. Two new multigroup cross section libraries have been recently generated from ENDF/B-V data for two specific shielding applications. The first library was prepared for use in sodium-cooled reactor systems and is available in both broad-group structures. The second library, just recently completed, was prepared for use in air-over-ground environments and is available in a broad-group (46-neutron, 23-photon) energy structure. The selection of the specific group structures and weighting functions was an important part of the generation of both libraries

ERRFILS: a preliminary library of 30-group multigroup covariance data for use in CTR sensitivity studies

International Nuclear Information System (INIS)

LaBauve, R.J.; Muir, D.W.

1978-01-01

A library of 30-group multigroup covariance data was prepared from preliminary ENDF/B-V data with the NJOY code. Data for Fe, Cr, Ni, 10 B, C, Cu, H, and Pb are included in this library. Reactions include total cross sections, elastic and inelastic scattering cross sections, and the most important absorption cross sections. Typical data from the file are shown. 3 tables

BUGLE-93 (ENDF/B-VI) cross-section library data testing using shielding benchmarks

International Nuclear Information System (INIS)

Hunter, H.T.; Slater, C.O.; White, J.E.

1994-01-01

Several integral shielding benchmarks were selected to perform data testing for new multigroup cross-section libraries compiled from the ENDF/B-VI data for light water reactor (LWR) shielding and dosimetry. The new multigroup libraries, BUGLE-93 and VITAMIN-B6, were studied to establish their reliability and response to the benchmark measurements by use of radiation transport codes, ANISN and DORT. Also, direct comparisons of BUGLE-93 and VITAMIN-B6 to BUGLE-80 (ENDF/B-IV) and VITAMIN-E (ENDF/B-V) were performed. Some benchmarks involved the nuclides used in LWR shielding and dosimetry applications, and some were sensitive specific nuclear data, i.e. iron due to its dominant use in nuclear reactor systems and complex set of cross-section resonances. Five shielding benchmarks (four experimental and one calculational) are described and results are presented

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

Validation of multigroup neutron cross sections and calculational methods for the advanced neutron source against the FOEHN critical experiments measurements

International Nuclear Information System (INIS)

Smith, L.A.; Gallmeier, F.X.; Gehin, J.C.

1995-05-01

The FOEHN critical experiment was analyzed to validate the use of multigroup cross sections and Oak Ridge National Laboratory neutronics computer codes in the design of the Advanced Neutron Source. The ANSL-V 99-group master cross section library was used for all the calculations. Three different critical configurations were evaluated using the multigroup KENO Monte Carlo transport code, the multigroup DORT discrete ordinates transport code, and the multigroup diffusion theory code VENTURE. The simple configuration consists of only the fuel and control elements with the heavy water reflector. The intermediate configuration includes boron endplates at the upper and lower edges of the fuel element. The complex configuration includes both the boron endplates and components in the reflector. Cross sections were processed using modules from the AMPX system. Both 99-group and 20-group cross sections were created and used in two-dimensional models of the FOEHN experiment. KENO calculations were performed using both 99-group and 20-group cross sections. The DORT and VENTURE calculations were performed using 20-group cross sections. Because the simple and intermediate configurations are azimuthally symmetric, these configurations can be explicitly modeled in R-Z geometry. Since the reflector components cannot be modeled explicitly using the current versions of these codes, three reflector component homogenization schemes were developed and evaluated for the complex configuration. Power density distributions were calculated with KENO using 99-group cross sections and with DORT and VENTURE using 20-group cross sections. The average differences between the measured values and the values calculated with the different computer codes range from 2.45 to 5.74%. The maximum differences between the measured and calculated thermal flux values for the simple and intermediate configurations are ∼ 13%, while the average differences are < 8%

Integral tests of coupled multigroup neutron and gamma cross sections with fission and fusion sources

International Nuclear Information System (INIS)

Schriewer, J.; Hehn, G.; Mattes, M.; Pfister, G.; Keinert, J.

1978-01-01

Calculations were made for different benchmark experiments in order to test the coupled multigroup neutron and gamma library EURLIB-3 with 100 neutron groups and 20 gamma groups. In cooperation with EURATOM, Ispra, we produced this shielding library recently from ENDF/B-IV data for application in fission and fusion technology. Integral checks were performed for natural lithium, carbon, oxygen, and iron. Since iron is the most important structural material in nuclear technology, we started with calculations of iron benchmark experiments. Most of them are integral experiments of INR, Karlsruhe, but comparisons were also done with benchmark experiments from USA and Japan. For the experiments with fission sources we got satisfying results. All details of the resonances cannot be checked with flux measurements and multigroup cross sections used. But some averaged resonance behaviour of the measured and calculated fluxes can be compared and checked within the error limits given. We get greater differences in the calculations of benchmark experiments with 14 MeV neutron sources. For iron the group cross sections of EURLIB-3 produce an underestimation of the neutron flux in a broad energy region below the source energy. The conclusion is that the energy degradation by inelastic scattering is too strong. For fusion application the anisotropy of the inelastic scatter process must be taken into account, which isn't done by the processing codes at present. If this effect isn't enough, additional corrections have to be applied to the inelastic cross sections of iron in ENDF/B-IV. (author)

ZZ BOREHOLE-EB6.8-MG, multi group cross-section library for deterministic and Monte Carlo codes

International Nuclear Information System (INIS)

Kodeli, Ivo; Aldama, Daniel L.; Leege, Piet F.A. de; Legrady, David; Hoogenboom, J. Eduard

2007-01-01

1 - Description: Format: MATXS and ACE; Number of groups: 175 neutron, 45 gamma-ray; Nuclides: H-1, C-12, O-16, Na-23, Mg-nat, Al-27, Si-28, -29, -30, S-nat, Cl-35, -37, K-nat, Ca-nat, Mn-55, Fe-54, -56, -57, -58, I-127, W-nat. Origin: ENDF/B-VI.8; Weighting spectrum: Fission and fusion peak at high energies and a 1/E + thermal Maxwellian extension at low energies. The following materials/nuclides are included in the library: H-1, C-12, O-16, Na-23, Mg-nat, Al-27, Si-28, -29, -30, S-nat, Cl-35, -37, K-nat, Ca-nat, Fe-54, -56, -57, -58, Mn-55, I-127, W-nat. ZZ-BOREHOLE-EB6.8-MG is a multigroup cross section library for deterministic (DOORS, DANTSYS) and Monte Carlo (MCNP) transport codes developed for the oil well logging applications. The library is based on the ENDF/B-VI.8 evaluation and was processed by the NJOY-99 code. The cross sections are given in the 175 neutron and 45 gamma ray group structure. The MATXS format library can be directly used in TRANSX code to prepare the multigroup self-shielded cross sections for deterministic discrete ordinates codes like DOORS and DANTSYS. The data provided in the GROUPR and GAMINR format were converted to the MCNP ACE format by the NSLINK, SCALE and CRSRD codes. IAEA1398/03: Multigroup cross section data for Mn-55 were added in TRANSX format

RZ calculations for self shielded multigroup cross sections

Energy Technology Data Exchange (ETDEWEB)

Li, M.; Sanchez, R.; Zmijarevic, I.; Stankovski, Z. [Commissariat a l' Energie Atomique CEA, Direction de l' Energie Nucleaire, DEN/DM2S/SERMA/LENR, 91191 Gif-sur-Yvette Cedex (France)

2006-07-01

A collision probability method has been implemented for RZ geometries. The method accounts for white albedo, specular and translation boundary condition on the top and bottom surfaces of the geometry and for a white albedo condition on the outer radial surface. We have applied the RZ CP method to the calculation of multigroup self shielded cross sections for Gadolinia absorbers in BWRs. (authors)

RZ calculations for self shielded multigroup cross sections

International Nuclear Information System (INIS)

Li, M.; Sanchez, R.; Zmijarevic, I.; Stankovski, Z.

2006-01-01

A collision probability method has been implemented for RZ geometries. The method accounts for white albedo, specular and translation boundary condition on the top and bottom surfaces of the geometry and for a white albedo condition on the outer radial surface. We have applied the RZ CP method to the calculation of multigroup self shielded cross sections for Gadolinia absorbers in BWRs. (authors)

Review of uncertainty files and improved multigroup cross section files for FENDL

International Nuclear Information System (INIS)

Ganesan, S.

1994-03-01

The IAEA Nuclear Data Section, in co-operation with several national nuclear data centers and research groups, is creating an internationally available Fusion Evaluated Nuclear Data Library (FENDL), which will serve as a comprehensive source of processed and tested nuclear data tailored to the requirements of the Engineering and Development Activities (EDA) of the International Thermonuclear Experimental Reactor (ITER) Project and other fusion-related development projects. The FENDL project of the International Atomic Energy Agency has the task of coordination with the goal of assembling, processing and testing a comprehensive, fusion-relevant Fusion Evaluated Nuclear Data Library with unrestricted international distribution. The present report contains the summary of the IAEA Advisory Group Meeting on ''Review of Uncertainty Files and Improved Multigroup Cross Section Files for FENDL'', held during 8-12 November 1993 at the Tokai Research Establishment, JAERI, Japan, organized in cooperation with the Japan Atomic Energy Research Institute. The report presents the current status of the FENDL activity and the future work plans in the form of conclusions and recommendations of the four Working Groups of the Advisory Group Meeting on (1) experimental and calculational benchmarks, (2) preparation processed libraries for FENDL/ITER, (3) specifying procedures for improving FENDL and (4) selection of activation libraries for FENDL. (author). 1 tab

Adjustement of multigroup cross sections using fast reactor integral data

International Nuclear Information System (INIS)

Renke, C.A.C.

1982-01-01

A methodology for the adjustment of multigroup cross section is presented, structured with aiming to compatibility the limitated number of measured values of integral parameters known and disponible, and the great number of cross sections to be adjusted the group of cross section used is that obtained from the Carnaval II calculation system, understanding as formular the sets of calculation methods and data bases. The adjustment is realized, using the INCOAJ computer code, developed in function of one statistical formulation, structural from the bayer considerations, taking in account the measurement processes of cross section and integral parameters defined on statistical bases. (E.G.) [pt

MINX, Multigroup Cross-Sections and Self-Shielding Factors from ENDF/B for Program SPHINX

International Nuclear Information System (INIS)

Soran, P.D.; MacFarlane, R.E.; Harris, D.R.; LaBauve, R.J.; Hendricks, J.S.; Kidman, R.B.; Weisbin, C.R.; White, J.E.

1977-01-01

1 - Description of problem or function: MINX calculates fine-group averaged infinitely diluted cross sections and self-shielding factors from ENDF/B-IV data. Its primary purpose is to generate a pseudo-composition-independent multigroup library which is input to the SPHINX space-energy collapse program (2) (PSR-0129) through standard CCCC-III (8) interfaces. MINX incorporates and improves upon the resonance capabilities of existing codes such as ETOX (5) (NESC0388) and ENDRUN (9) and the high-order group-to-group transfer matrices of SUPERTOG (10) (PSR-0013) and ETOG (11). Fine group energy boundaries, Legendre expansion order, gross spectral shape component (in the Bondarenko flux model), temperatures and dilutions can all be used specifically. 2 - Method of solution: Infinitely dilute, un-broadened point cross sections are obtained from resolved resonance parameters using a modified version of the RESEND program (3) (NESC0465). The SIGMA1 (4) (IAEA0854) kernel-broadening method is used to Doppler broaden and thin the tabulated linearized pointwise cross sections at 0 K (outside of the unresolved energy region). Effective temperature- dependent self-shielded pointwise cross sections are derived from the formulation in the ETOX code. The primary modification to the ETOX algorithm is associated with the numerical quadrature scheme used to establish the mean values of the fluctuation intervals. The selection of energy mesh points, at which the effective cross sections are calculated, has been modified to include the energy points given in the ENDF/B file or, if the energy-independent formalism was employed, points at half-lethargy intervals. Infinitely dilute group cross sections and self-shielding factors are generated using the Bondarenko flux weighting model with the gross spectral shape under user control. The integral over energy for each group is divided into a set of panels defined by the union of the grid points describing the total cross section, the

Production and testing of the ENEA-Bologna VITJEFF32.BOLIB (JEFF-3.2) multi-group (199 n + 42 γ) cross section library in AMPX format for nuclear fission applications

Science.gov (United States)

Pescarini, Massimo; Orsi, Roberto; Frisoni, Manuela

2017-09-01

The ENEA-Bologna Nuclear Data Group produced the VITJEFF32.BOLIB multi-group coupled neutron/photon (199 n + 42 γ) cross section library in AMPX format, based on the OECD-NEA Data Bank JEFF-3.2 evaluated nuclear data library. VITJEFF32.BOLIB was conceived for nuclear fission applications as European counterpart of the ORNL VITAMIN-B7 similar library (ENDF/B-VII.0 data). VITJEFF32.BOLIB has the same neutron and photon energy group structure as the former ORNL VITAMIN-B6 reference library (ENDF/B-VI.3 data) and was produced using similar data processing methodologies, based on the LANL NJOY-2012.53 nuclear data processing system for the generation of the nuclide cross section data files in GENDF format. Then the ENEA-Bologna 2007 Revision of the ORNL SCAMPI nuclear data processing system was used for the conversion into the AMPX format. VITJEFF32.BOLIB contains processed cross section data files for 190 nuclides, obtained through the Bondarenko (f-factor) method for the treatment of neutron resonance self-shielding and temperature effects. Collapsed working libraries of self-shielded cross sections in FIDO-ANISN format, used by the deterministic transport codes of the ORNL DOORS system, can be generated from VITJEFF32.BOLIB through the cited SCAMPI version. This paper describes the methodology and specifications of the data processing performed and presents some results of the VITJEFF32.BOLIB validation.

Production and testing of the ENEA-Bologna VITJEFF32.BOLIB (JEFF-3.2 multi-group (199 n + 42 γ cross section library in AMPX format for nuclear fission applications

Directory of Open Access Journals (Sweden)

Pescarini Massimo

2017-01-01

Full Text Available The ENEA-Bologna Nuclear Data Group produced the VITJEFF32.BOLIB multi-group coupled neutron/photon (199 n + 42 γ cross section library in AMPX format, based on the OECD-NEA Data Bank JEFF-3.2 evaluated nuclear data library. VITJEFF32.BOLIB was conceived for nuclear fission applications as European counterpart of the ORNL VITAMIN-B7 similar library (ENDF/B-VII.0 data. VITJEFF32.BOLIB has the same neutron and photon energy group structure as the former ORNL VITAMIN-B6 reference library (ENDF/B-VI.3 data and was produced using similar data processing methodologies, based on the LANL NJOY-2012.53 nuclear data processing system for the generation of the nuclide cross section data files in GENDF format. Then the ENEA-Bologna 2007 Revision of the ORNL SCAMPI nuclear data processing system was used for the conversion into the AMPX format. VITJEFF32.BOLIB contains processed cross section data files for 190 nuclides, obtained through the Bondarenko (f-factor method for the treatment of neutron resonance self-shielding and temperature effects. Collapsed working libraries of self-shielded cross sections in FIDO-ANISN format, used by the deterministic transport codes of the ORNL DOORS system, can be generated from VITJEFF32.BOLIB through the cited SCAMPI version. This paper describes the methodology and specifications of the data processing performed and presents some results of the VITJEFF32.BOLIB validation.

Methodological study of the adjustment of multigroup cross sections for fast reactors, by means of integral data

International Nuclear Information System (INIS)

Bastos, H.F.B.N.

1979-01-01

In this work a study of the methodology of the adjustment of multigroup cross sections by means of integral data is presented. A synthesis of the principal methods existent and the mathematical development of the adaptation of one of them are made. A calculational system is built from this reference method, with the basic conditions for the operation of the process of adjustment. In order to test the system developed and analyze several problems related to the adjustment, a series of trial adjustments was made with the value of the U 235 fission cross section from the infinite dilution library used in the calculational system for fast reactors of the Instituto de Engenharia Nuclear. (author)

AMZ, multigroup constant library for EXPANDA code, generated by NJOY code from ENDF/B-IV

International Nuclear Information System (INIS)

Chalhoub, E.S.; Moraes, Marisa de

1985-01-01

It is described a library of multigroup constants with 70 energy groups and 37 isotopes to fast reactor calculation. The cross sections, scattering matrices and self-shielding factors were generated by NJOY code and RGENDF interface program, from ENDF/B-IV'S evaluated data. The library is edited in adequated format to be used by EXPANDA code. (M.C.K.) [pt

MENDF71x. Multigroup Neutron Cross Section Data Tables Based upon ENDF/B-VII.1

International Nuclear Information System (INIS)

Conlin, Jeremy Lloyd; Parsons, Donald Kent; Gardiner, Steven J.; Gray, Mark Girard; Lee, Mary Beth; White, Morgan Curtis

2015-01-01

A new multi-group neutron cross section library has been released along with the release of NDI version 2.0.20. The library is named MENDF71x and is based upon the evaluations released in ENDF/B-VII.1 which was made publicly available in December 2011. ENDF/B-VII.1 consists of 423 evaluations of which ten are excited states evaluations and 413 are ground state evaluations. MENDF71x was created by processing the 423 evaluations into 618-group, downscatter only NDI data tables. The ENDF/B evaluation files were processed using NJOY version 99.393 with the exception of 35 Cl and 233 U. Those two isotopes had unique properties that required that we process the evaluation using NJOY version 2012. The MENDF71x library was only processed to room temperature, i.e., 293.6 K. In the future, we plan on producing a multi-temperature library based on ENDF/B-VII.1 and compatible with MENDF71x.

A code system to generate multigroup cross-sections using basic data

International Nuclear Information System (INIS)

Garg, S.B.; Kumar, Ashok

1978-01-01

For the neutronic studies of nuclear reactors, multigroup cross-sections derived from the basic energy point data are needed. In order to carry out the design based studies, these cross-sections should also incorporate the temperature and fuel concentration effects. To meet these requirements, a code system comprising of RESRES, UNRES, FIGERO, INSCAT, FUNMO, AVER1 and BGPONE codes has been adopted. The function of each of these codes is discussed. (author)

Angular finite volume method for solving the multigroup transport equation with piecewise average scattering cross sections

International Nuclear Information System (INIS)

Calloo, A.; Vidal, J.F.; Le Tellier, R.; Rimpault, G.

2011-01-01

This paper deals with the solving of the multigroup integro-differential form of the transport equation for fine energy group structure. In that case, multigroup transfer cross sections display strongly peaked shape for light scatterers and the current Legendre polynomial expansion is not well-suited to represent them. Furthermore, even if considering an exact scattering cross sections representation, the scattering source in the discrete ordinates method (also known as the Sn method) being calculated by sampling the angular flux at given directions, may be wrongly computed due to lack of angular support for the angular flux. Hence, following the work of Gerts and Matthews, an angular finite volume solver has been developed for 2D Cartesian geometries. It integrates the multigroup transport equation over discrete volume elements obtained by meshing the unit sphere with a product grid over the polar and azimuthal coordinates and by considering the integrated flux per solid angle element. The convergence of this method has been compared to the S_n method for a highly anisotropic benchmark. Besides, piecewise-average scattering cross sections have been produced for non-bound Hydrogen atoms using a free gas model for thermal neutrons. LWR lattice calculations comparing Legendre representations of the Hydrogen scattering multigroup cross section at various orders and piecewise-average cross sections for this same atom are carried out (while keeping a Legendre representation for all other isotopes). (author)

Development of multi-group xs libraries for the gfr 2400 reactor

International Nuclear Information System (INIS)

Cerba, Š.; Vrban, B.; Lüley, J.; Necas, V.

2016-01-01

GFR 2400 is considered as a conceptual design of the large scale GEN IV Gas-Cooled Fast Reactor. In general, the GEN IV technologies are seen as reliable but also very challenging reactor concepts. Since GFR 2400 lacks any experimental data, the questions on its safety are even more complex and the assessment of its performance could be made only based on computational experience. The paper deals with the development process of multi-group XS libraries based on a hybrid deterministic-Stochastic methodology, using the NJOY99, TRANSX, DIF3D, PARTISN and MCNP5 codes. A new optimized 25 group SBJ E 71 2 5G cross section library was developed based on ENDF/B-VII.1 evaluated data, ZZ-KAFAX-E70 background cross sections and GFR 2400 neutron spectrum. The created library was validated through integral experiments evaluated on the HEX-Z deterministic models in DIF3D. The results were also compared with MCNP5 calculations. (authors)

Angular finite volume method for solving the multigroup transport equation with piecewise average scattering cross sections

Energy Technology Data Exchange (ETDEWEB)

Calloo, A.; Vidal, J.F.; Le Tellier, R.; Rimpault, G., E-mail: ansar.calloo@cea.fr, E-mail: jean-francois.vidal@cea.fr, E-mail: romain.le-tellier@cea.fr, E-mail: gerald.rimpault@cea.fr [CEA, DEN, DER/SPRC/LEPh, Saint-Paul-lez-Durance (France)

2011-07-01

This paper deals with the solving of the multigroup integro-differential form of the transport equation for fine energy group structure. In that case, multigroup transfer cross sections display strongly peaked shape for light scatterers and the current Legendre polynomial expansion is not well-suited to represent them. Furthermore, even if considering an exact scattering cross sections representation, the scattering source in the discrete ordinates method (also known as the Sn method) being calculated by sampling the angular flux at given directions, may be wrongly computed due to lack of angular support for the angular flux. Hence, following the work of Gerts and Matthews, an angular finite volume solver has been developed for 2D Cartesian geometries. It integrates the multigroup transport equation over discrete volume elements obtained by meshing the unit sphere with a product grid over the polar and azimuthal coordinates and by considering the integrated flux per solid angle element. The convergence of this method has been compared to the S{sub n} method for a highly anisotropic benchmark. Besides, piecewise-average scattering cross sections have been produced for non-bound Hydrogen atoms using a free gas model for thermal neutrons. LWR lattice calculations comparing Legendre representations of the Hydrogen scattering multigroup cross section at various orders and piecewise-average cross sections for this same atom are carried out (while keeping a Legendre representation for all other isotopes). (author)

COMPAR, NJOY, GROUPIE, FLANGE-2, ETOG-3, XLACS Multigroup Cross-Sections General Comparison

International Nuclear Information System (INIS)

Anaf, Jaime; Chalhoub, E.S.

1990-01-01

1 - Description of program or function: A system for comparing multigroup cross sections generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS. This system comprises the COMPAR program and interface (auxiliary) programs developed for each of the programs under consideration. These are REDCOMP for GROUPIE, FLACOMP for FLANGE-II, ETOCOMP for ETOG-3 and XLACOMP for XLACS. For the NJOY program there is RGENDF, a program developed apart from this system. It is a modular system in which the inclusion of new multigroup cross section generating program requires no more than the development of a new interface module. 2 - Method of solution: Refer to comments in main routine. 3 - Restrictions on the complexity of the problem: Refer to comments in main routine

ENEA-Bologna production and testing of Jeff-3.1 multi-group cross section libraries for nuclear fission applications

International Nuclear Information System (INIS)

Pescarini, M.; Orsi, R.; Sinitsa, V.

2008-01-01

The ENEA-Bologna Nuclear Data Group produced the JEFF-3.1 VITJEFF31.BOLIB and MATJEFF31. BOLIB fine-group coupled neutron and photon (199 n + 42 γ) cross section libraries for nuclear fission applications, respectively in AMPX and MATXS format, with the same specifications and energy group structure of the Endf/B-VI-3 VITAMIN-B6 American library. Each library, containing 181 nuclide cross section files, was generated from the same set of cross section data files in GENDF format, obtained through the Bondarenko (f-factor) method, with an ENEA-Bologna revised version of the GROUPR module of the NJOY-99.160 system. Collapsed working libraries of self-shielded cross sections in FIDO-ANISN format, used by the deterministic transport codes of the DANTSYS and DOORS systems, can be generated from VITJEFF31.BOLIB and MATJEFF31.BOLIB through, respectively, further data processing with an ENEA-Bologna revised version of the SCAMPI system and with the TRANSX code. This paper describes the methodology and specifications of the data processing performed and presents some results of the VITJEFF31.BOLIB validation. (authors)

ENEA-Bologna production and testing of JEF-2.2 multi-group cross section libraries for nuclear fission applications

International Nuclear Information System (INIS)

Pescarini, M.; Orsi, R.; Martinelli, T.; Sinitsa, V.; Blokhin, A.I.

2005-01-01

The ENEA-Bologna Nuclear Data Group produced the VITJEF22.BOLIB (NEA-1699/01 ZZ VITJEF22.BOLIB) and MATJEF22.BOLIB (NEA-1740/01 ZZ MATJEF22.BOLIB) fine-group coupled neutron and photon (199 n + 42 γ) cross section libraries for nuclear fission applications, respectively in AMPX and MATXS format and based on the JEF-2.2 European nuclear data file. Both the libraries were produced from the same set of cross section files in GENDF format, generated with the NJOY-94.66 nuclear data processing system. The present libraries can be considered as European counterparts of the VITAMIN-B6 (DLC-0184 ZZ VITAMIN-B6) American library in AMPX format, based on the ENDF/B-VI Release 3 American nuclear data file. In fact they have the same general features and the same neutron and photon energy group structures as VITAMIN-B6. In particular, all these libraries are pseudo-problem-independent and based on the Bondarenko (f-factor) method for the treatment of neutron resonance self-shielding and temperature effects. Each ENEA-Bologna library contains a set of 133 nuclide cross section files processed at 4 temperatures (300 K, 600 K, 1000 K and 2100 K) and obtained for the most part with 6 to 8 values of the background cross section σ 0 . Thermal scattering cross sections were processed at all the temperatures available in the JEF-2.2 thermal scattering law data file for 5 additional bound nuclides: H-1 in light water, H-1 in polyethylene, H-2 in heavy water, C in graphite and Be in beryllium metal. Collapsed working libraries of self-shielded cross sections in the formats used by the deterministic transport codes of the DANTSYS and DOORS systems can be generated from VITJEF22.BOLIB and MATJEF22.BOLIB through, respectively, further problem-dependent data processing with the AMPX or SCAMPI nuclear data processing systems and with the TRANSX code. (authors)

Procedure to Generate the MPACT Multigroup Library

Energy Technology Data Exchange (ETDEWEB)

Kim, Kang Seog [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-12-17

The CASL neutronics simulator MPACT is under development for the neutronics and T-H coupled simulation for the light water reactor. The objective of this document is focused on reviewing the current procedure to generate the MPACT multigroup library. Detailed methodologies and procedures are included in this document for further discussion to improve the MPACT multigroup library.

AFCI-2.0 Neutron Cross Section Covariance Library

Energy Technology Data Exchange (ETDEWEB)

Herman, M.; Herman, M; Oblozinsky, P.; Mattoon, C.M.; Pigni, M.; Hoblit, S.; Mughabghab, S.F.; Sonzogni, A.; Talou, P.; Chadwick, M.B.; Hale, G.M.; Kahler, A.C.; Kawano, T.; Little, R.C.; Yount, P.G.

2011-03-01

The cross section covariance library has been under development by BNL-LANL collaborative effort over the last three years. The project builds on two covariance libraries developed earlier, with considerable input from BNL and LANL. In 2006, international effort under WPEC Subgroup 26 produced BOLNA covariance library by putting together data, often preliminary, from various sources for most important materials for nuclear reactor technology. This was followed in 2007 by collaborative effort of four US national laboratories to produce covariances, often of modest quality - hence the name low-fidelity, for virtually complete set of materials included in ENDF/B-VII.0. The present project is focusing on covariances of 4-5 major reaction channels for 110 materials of importance for power reactors. The work started under Global Nuclear Energy Partnership (GNEP) in 2008, which changed to Advanced Fuel Cycle Initiative (AFCI) in 2009. With the 2011 release the name has changed to the Covariance Multigroup Matrix for Advanced Reactor Applications (COMMARA) version 2.0. The primary purpose of the library is to provide covariances for AFCI data adjustment project, which is focusing on the needs of fast advanced burner reactors. Responsibility of BNL was defined as developing covariances for structural materials and fission products, management of the library and coordination of the work; LANL responsibility was defined as covariances for light nuclei and actinides. The COMMARA-2.0 covariance library has been developed by BNL-LANL collaboration for Advanced Fuel Cycle Initiative applications over the period of three years, 2008-2010. It contains covariances for 110 materials relevant to fast reactor R&D. The library is to be used together with the ENDF/B-VII.0 central values of the latest official release of US files of evaluated neutron cross sections. COMMARA-2.0 library contains neutron cross section covariances for 12 light nuclei (coolants and moderators), 78 structural

AFCI-2.0 Neutron Cross Section Covariance Library

International Nuclear Information System (INIS)

Herman, M.; Oblozinsky, P.; Mattoon, C.M.; Pigni, M.; Hoblit, S.; Mughabghab, S.F.; Sonzogni, A.; Talou, P.; Chadwick, M.B.; Hale, G.M.; Kahler, A.C.; Kawano, T.; Little, R.C.; Yount, P.G.

2011-01-01

The cross section covariance library has been under development by BNL-LANL collaborative effort over the last three years. The project builds on two covariance libraries developed earlier, with considerable input from BNL and LANL. In 2006, international effort under WPEC Subgroup 26 produced BOLNA covariance library by putting together data, often preliminary, from various sources for most important materials for nuclear reactor technology. This was followed in 2007 by collaborative effort of four US national laboratories to produce covariances, often of modest quality - hence the name low-fidelity, for virtually complete set of materials included in ENDF/B-VII.0. The present project is focusing on covariances of 4-5 major reaction channels for 110 materials of importance for power reactors. The work started under Global Nuclear Energy Partnership (GNEP) in 2008, which changed to Advanced Fuel Cycle Initiative (AFCI) in 2009. With the 2011 release the name has changed to the Covariance Multigroup Matrix for Advanced Reactor Applications (COMMARA) version 2.0. The primary purpose of the library is to provide covariances for AFCI data adjustment project, which is focusing on the needs of fast advanced burner reactors. Responsibility of BNL was defined as developing covariances for structural materials and fission products, management of the library and coordination of the work; LANL responsibility was defined as covariances for light nuclei and actinides. The COMMARA-2.0 covariance library has been developed by BNL-LANL collaboration for Advanced Fuel Cycle Initiative applications over the period of three years, 2008-2010. It contains covariances for 110 materials relevant to fast reactor R and D. The library is to be used together with the ENDF/B-VII.0 central values of the latest official release of US files of evaluated neutron cross sections. COMMARA-2.0 library contains neutron cross section covariances for 12 light nuclei (coolants and moderators), 78

A broad-group cross-section library based on ENDF/B-VII.0 for fast neutron dosimetry Applications

Energy Technology Data Exchange (ETDEWEB)

Alpan, F.A. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2011-07-01

A new ENDF/B-VII.0-based coupled 44-neutron, 20-gamma-ray-group cross-section library was developed to investigate the latest evaluated nuclear data file (ENDF) ,in comparison to ENDF/B-VI.3 used in BUGLE-96, as well as to generate an objective-specific library. The objectives selected for this work consisted of dosimetry calculations for in-vessel and ex-vessel reactor locations, iron atom displacement calculations for reactor internals and pressure vessel, and {sup 58}Ni(n,{gamma}) calculation that is important for gas generation in the baffle plate. The new library was generated based on the contribution and point-wise cross-section-driven (CPXSD) methodology and was applied to one of the most widely used benchmarks, the Oak Ridge National Laboratory Pool Critical Assembly benchmark problem. In addition to the new library, BUGLE-96 and an ENDF/B-VII.0-based coupled 47-neutron, 20-gamma-ray-group cross-section library was generated and used with both SNLRML and IRDF dosimetry cross sections to compute reaction rates. All reaction rates computed by the multigroup libraries are within {+-} 20 % of measurement data and meet the U. S. Nuclear Regulatory Commission acceptance criterion for reactor vessel neutron exposure evaluations specified in Regulatory Guide 1.190. (authors)

ANSL-V: ENDF/B-V based multigroup cross-section libraries for Advanced Neutron Source (ANS) reactor studies. Supplement 1

Energy Technology Data Exchange (ETDEWEB)

Wright, R.Q.; Renier, J.P.; Bucholz, J.A.

1995-08-01

The original ANSL-V cross-section libraries (ORNL-6618) were developed over a period of several years for the physics analysis of the ANS reactor, with little thought toward including the materials commonly needed for shielding applications. Materials commonly used for shielding applications include calcium barium, sulfur, phosphorous, and bismuth. These materials, as well as {sup 6}Li, {sup 7}Li, and the naturally occurring isotopes of hafnium, have been added to the ANSL-V libraries. The gamma-ray production and gamma-ray interaction cross sections were completely regenerated for the ANSL-V 99n/44g library which did not exist previously. The MALOCS module was used to collapse the 99n/44g coupled library to the 39n/44g broad- group library. COMET was used to renormalize the two-dimensional (2- D) neutron matrix sums to agree with the one-dimensional (1-D) averaged values. The FRESH module was used to adjust the thermal scattering matrices on the 99n/44g and 39n/44g ANSL-V libraries. PERFUME was used to correct the original XLACS Legendre polynomial fits to produce acceptable distributions. The final ANSL-V 99n/44g and 39n/44g cross-section libraries were both checked by running RADE. The AIM module was used to convert the master cross-section libraries from binary coded decimal to binary format (or vice versa).

Extension and Verification of the Cross-Section Library for the VVER-1000 Surveillance Specimen Region

International Nuclear Information System (INIS)

Kirilova, D.; Belousov, S.; Ilieva, K.

2011-01-01

The objective of this work is a generation of new version of the BGL multigroup cross-section to extend the region of its applicability. The existing library version is problem oriented for VVER-1000 type of reactors and was generated by collapsing of the VITAMIN-B6 problem independent cross-section fine-group library applying the VVER-1000 reactor middle plane spectrum in cylindrical geometry. The new version BGLex additionally contains cross-sections averaged on the corresponding spectra of the surveillance specimen's (SS) region for VVER-1000 type of reactors. Comparative analysis of the neutron spectra for different one-dimensional geometry models that could be applied for the cross-section collapsing using the software package SCALE, showed a high sensitivity of the results to the geometry model. That is why a neutron importance assessment was done for the SS region using the adjoint solution calculated by the two-dimensional code DORT and problem-independent library VITAMIN-B6. The one-dimensional geometry model applied to the cross-section collapsing were determined by the material limits above the reactor core in axial direction z as for every material a homogenization in radial direction was done. The material homogenization in radial direction was done by material weighing taking into account the adjoint solution as well as the neutron source. The one-dimensional geometry model comprising the homogenized weighed materials was applied for the cross-section generation of the fine-group library VITAMIN-B6 to the broad-group structure of BGL library. The new version BGLex was extended with cross-sections for the SS region. Verification and validation of the new version BGLex is forthcoming. It includes comparison between the calculated results with the new version BGLex and the libraries BGL and VITAMIN-B6 and comparison with experimental results. (author)

Extension and Verification of the Cross-Section Library for the VVER- 1000 Surveillance Specimen Region

International Nuclear Information System (INIS)

Kirilova, D.; Belousov, S.; Ilieva, K.

2011-01-01

The objective of this work is a generation of new version of the BGL multigroup cross-section to extend the region of its applicability. The existing library version is problem oriented for VVER-1000 type of reactors and was generated by collapsing of the VITAMIN-B6 problem independent cross-section fine-group library applying the VVER-1000 reactor middle plane spectrum in cylindrical geometry. The new version BGLex additionally contains cross-sections averaged on the corresponding spectra of the surveillance specimen's (SS) region for VVER-1000 type of reactors. Comparative analysis of the neutron spectra for different one-dimensional geometry models that could be applied for the cross-section collapsing using the software package SCALE, showed a high sensitivity of the results to the geometry model. That is why a neutron importance assessment was done for the SS region using the adjoint solution calculated by the two-dimensional code DORT and problem-independent library VITAMIN-B6. The one-dimensional geometry model applied to the cross-section collapsing were determined by the material limits above the reactor core in axial direction z as for every material a homogenization in radial direction was done. The material homogenization in radial direction was done by material weighing taking into account the adjoint solution as well as the neutron source. The one-dimensional geometry model comprising the homogenized weighed materials was applied for the cross-section generation of the fine-group library VITAMIN-B6 to the broad-group structure of BGL library. The new version BGLex was extended with cross-sections for the SS region. Verification and validation of the new version BGLex is forthcoming. It includes comparison between the calculated results with the new version BGLex and the libraries BGL and VITAMIN-B6 and comparison with experimental results. (author)

Status report on multigroup cross section generation code development for high-fidelity deterministic neutronics simulation system

International Nuclear Information System (INIS)

Yang, W.S.; Lee, C.H.

2008-01-01

Under the fast reactor simulation program launched in April 2007, development of an advanced multigroup cross section generation code was initiated in July 2007, in conjunction with the development of the high-fidelity deterministic neutron transport code UNIC. The general objectives are to simplify the existing multi-step schemes and to improve the resolved and unresolved resonance treatments. Based on the review results of current methods and the fact that they have been applied successfully to fast critical experiment analyses and fast reactor designs for last three decades, the methodologies of the ETOE-2/MC 2 -2/SDX code system were selected as the starting set of methodologies for multigroup cross section generation for fast reactor analysis. As the first step for coupling with the UNIC code and use in a parallel computing environment, the MC 2 -2 code was updated by modernizing the memory structure and replacing old data management package subroutines and functions with FORTRAN 90 based routines. Various modifications were also made in the ETOE-2 and MC 2 -2 codes to process the ENDF/B-VII.0 data properly. Using the updated ETOE-2/MC 2 -2 code system, the ENDF/B-VII.0 data was successfully processed for major heavy and intermediate nuclides employed in sodium-cooled fast reactors. Initial verification tests of the MC 2 -2 libraries generated from ENDF/B-VII.0 data were performed by inter-comparison of twenty-one group infinite dilute total cross sections obtained from MC 2 -2, VIM, and NJOY. For almost all nuclides considered, MC 2 -2 cross sections agreed very well with those from VIM and NJOY. Preliminary validation tests of the ENDF/B-VII.0 libraries of MC 2 -2 were also performed using a set of sixteen fast critical benchmark problems. The deterministic results based on MC 2 -2/TWODANT calculations were in good agreement with MCNP solutions within ∼0.25% Δρ, except a few small LANL fast assemblies. Relative to the MCNP solution, the MC 2 -2/TWODANT

Status report on multigroup cross section generation code development for high-fidelity deterministic neutronics simulation system.

Energy Technology Data Exchange (ETDEWEB)

Yang, W. S.; Lee, C. H. (Nuclear Engineering Division)

2008-05-16

Under the fast reactor simulation program launched in April 2007, development of an advanced multigroup cross section generation code was initiated in July 2007, in conjunction with the development of the high-fidelity deterministic neutron transport code UNIC. The general objectives are to simplify the existing multi-step schemes and to improve the resolved and unresolved resonance treatments. Based on the review results of current methods and the fact that they have been applied successfully to fast critical experiment analyses and fast reactor designs for last three decades, the methodologies of the ETOE-2/MC{sup 2}-2/SDX code system were selected as the starting set of methodologies for multigroup cross section generation for fast reactor analysis. As the first step for coupling with the UNIC code and use in a parallel computing environment, the MC{sup 2}-2 code was updated by modernizing the memory structure and replacing old data management package subroutines and functions with FORTRAN 90 based routines. Various modifications were also made in the ETOE-2 and MC{sup 2}-2 codes to process the ENDF/B-VII.0 data properly. Using the updated ETOE-2/MC{sup 2}-2 code system, the ENDF/B-VII.0 data was successfully processed for major heavy and intermediate nuclides employed in sodium-cooled fast reactors. Initial verification tests of the MC{sup 2}-2 libraries generated from ENDF/B-VII.0 data were performed by inter-comparison of twenty-one group infinite dilute total cross sections obtained from MC{sup 2}-2, VIM, and NJOY. For almost all nuclides considered, MC{sup 2}-2 cross sections agreed very well with those from VIM and NJOY. Preliminary validation tests of the ENDF/B-VII.0 libraries of MC{sup 2}-2 were also performed using a set of sixteen fast critical benchmark problems. The deterministic results based on MC{sup 2}-2/TWODANT calculations were in good agreement with MCNP solutions within {approx}0.25% {Delta}{rho}, except a few small LANL fast assemblies

THEMIS-4: a coherent punctual and multigroup cross section library for Monte Carlo and SN codes from ENDF/B4

International Nuclear Information System (INIS)

Dejonghe, G.; Gonnord, J.; Monnier, A.; Nimal, J.C.

1983-05-01

The THEMIS cross section processing system has been developped to produce punctual data for MONTE CARLO and coherent multigroup data for SN codes from ENDF/B. The THEMIS-4 data base has been generated from ENDF/B4 using the system and can be accessed by the 3-D Monte Carlo system TRIPOLI-2 and by the SN codes ANISN and DOT. An interpretation of ORNL fusion shielding benchmark is presented

Generation of multigroup cross-sections from micro-group ones in code system SUHAM-U used for VVER-1000 reactor core calculations with MOX loading

Energy Technology Data Exchange (ETDEWEB)

Boyarinov, V.F.; Davidenko, V.D.; Polismakov, A.A.; Tsybulsky, V.F. [RRC Kurchatov Institute, Moscow (Russian Federation)

2005-07-01

At the present time, the new code system SUHAM-U for calculation of the neutron-physical processes in nuclear reactor core with triangular and square lattices based both on the modern micro-group (about 7000 groups) cross-sections library of code system UNK and on solving the multigroup (up to 89 groups) neutron transport equation by Surface Harmonics Method is elaborated. In this paper the procedure for generation of multigroup cross-sections from micro-group ones for calculation of VVER-1000 reactor core with MOX loading is described. The validation has consisted in computing VVER-1000 fuel assemblies with uranium and MOX fuel and has shown enough high accuracy under corresponding selection of the number and boundaries of the energy groups. This work has been fulfilled in the frame of ISTC project 'System Analyses of Nuclear Safety for VVER Reactors with MOX Fuels'.

Nuclear data and multigroup methods in fast reactor calculations

International Nuclear Information System (INIS)

Gur, Y.

1975-03-01

The work deals with fast reactor multigroup calculations, and the efficient treatment of basic nuclear data, which serves as raw material for the calculations. Its purpose is twofold: to build a computer code system that handles a large, detailed library of basic neutron cross section data, (such as ENDF/B-III) and yields a compact set of multigroup cross sections for reactor calculations; to use the code system for comparative analysis of different libraries, in order to discover basic uncertainties that still exist in the measurement of neutron cross sections, and to determine their influence upon uncertainties in nuclear calculations. A program named NANICK which was written in two versions is presented. The first handles the American basic data library, ENDF/B-III, while the second handles the German basic data library, KEDAK. The mathematical algorithm is identical in both versions, and only the file management is different. This program calculates infinitely diluted multigroup cross sections and scattering matrices. It is complemented by the program NASIF that calculates shielding factors from resonance parameters. Different versions of NASIF were written to handle ENDF/B-III or KEDAK. New methods for evaluating in reactor calculations the long term behavior of the neutron flux as well as its fine structure are described and an efficient calculation of the shielding factors from resonance parameters is offered. (B.G.)

ZZ TEMPEST/MUFT, Thermal Neutron and Fast Neutron Multigroup Cross-Section Library for Program LEOPARD

International Nuclear Information System (INIS)

Kim, Jung-Do; Lee, Jong Tai

1986-01-01

Description of problem or function: Format: TEMPEST and MUFT; Number of groups: 246 thermal groups in TEMPEST Format and 54 fast groups in MUFT Format. From this library, the program SPOTS4 generates a 172-54 group library as input to the code LEOPARD. Nuclides: H, O, Zr, C, Fe, Ni, Al, Cr, Mn, U, Pu, Th, Pa, Xe, Sm, B and D. Origin: ENDF/B-4; Weighting spectrum: 1/E + U 235 fission spectrum. Data library of thermal and fast neutron group Cross sections to generate input to the program LEOPARD. The data is based on ENDF/B-4 and consists of two parts: (1) 246 thermal groups in TEMPEST Format. (2) 54 fast groups in MUFT Format. From this library, the program SPOTS4 generates a 172-54 group library as input to the code LEOPARD (NESC0279)

MC2-2, Calculation of Fast Neutron Spectra and Multigroup Cross-Sections from ENDF/B Data

International Nuclear Information System (INIS)

2001-01-01

1 - Description of program or function: MC 2 -2 solves the neutron slowing-down equations using basic neutron data derived from ENDF/B data files to determine fundamental mode spectra for use in generating multigroup neutron cross sections. The current edition includes the ability to treat all ENDF/B-V and -VI data representations. It accommodates high-order P scattering representations and provides numerous capabilities such as isotope mixing, delayed neutron processing, free-format input, and flexibility in output data selection. This edition supersedes previous releases of the MC22 program and the earlier MC2 program. Improved physics algorithms and increased computational efficiency are incorporated. Input data files required by MC2-2 may be generated from ENDF/B data by the code ETOE-2. The hyper-fine-group integral transport theory module of MC2-2, RABANL, is an improved version of the RABBLE/RABID codes. Many of the MC2-2 modules are used in the SDX code. 2 - Methods: The extended transport P1, B1, consistent P1, and consistent B1 fundamental mode ultra-fine-group equations are solved using continuous slowing-down theory and multigroup methods. Fast and accurate resonance integral methods are used in the narrow resonance resolved and unresolved resonance treatments. A fundamental mode homogeneous unit cell calculation is performed using either a multigroup or a continuous slowing-down treatment. Multigroup neutron homogeneous cross sections are generated in an ISOTXS format for an arbitrary group structure. A hyper-fine-group integral transport slowing down calculation (RABANL) is available as an option. RABANL performs a homogeneous or heterogeneous (pin or slab) unit cell calculation over the resonance region (resolved and unresolved) and generates multigroup neutron cross sections in an ISOTXS format. Neutron cross sections are generated by RABANL for the homogeneous unit cell and for each heterogeneous region in the pin or slab unit cell calculation

AMZ, library of multigroup constants for EXPANDA computer codes, generated by NJOY computer code from ENDF/B-IV

International Nuclear Information System (INIS)

Chalhoub, E.S.; Moraes, M. de.

1984-01-01

A 70-group, 37-isotope library of multigroup constants for fast reactor nuclear design calculations is described. Nuclear cross sections, transfer matrices, and self-shielding factors were generated with NJOY code and an auxiliary program RGENDF using evaluated data from ENDF/B-IV. The output is being issued in a format suitable for EXPANDA code. Comparisons with JFS-2 library, as well as, test resuls for 14 CSEWG benchmark critical assemblies are presented. (Author) [pt

MC2-2: a code to calculate fast neutron spectra and multigroup cross sections

International Nuclear Information System (INIS)

Henryson, H. II; Toppel, B.J.; Stenberg, C.G.

1976-06-01

MC 2 -2 is a program to solve the neutron slowing down problem using basic neutron data derived from the ENDF/B data files. The spectrum calculated by MC 2 -2 is used to collapse the basic data to multigroup cross sections for use in standard reactor neutronics codes. Four different slowing down formulations are used by MC 2 -2: multigroup, continuous slowing down using the Goertzel-Greuling or Improved Goertzel-Greuling moderating parameters, and a hyper-fine-group integral transport calculation. Resolved and unresolved resonance cross sections are calculated accounting for self-shielding, broadening and overlap effects. This document provides a description of the MC 2 -2 program. The physics and mathematics of the neutron slowing down problem are derived and detailed information is provided to aid the MC 2 -2 user in preparing input for the program and implementation of the program on IBM 370 or CDC 7600 computers

MIRANDA - a module based on multiregion resonance theory for generating cross sections within the AUS neutronics code system

International Nuclear Information System (INIS)

Robinson, G.S.

1985-12-01

MIRANDA is the cross-section generation module of the AUS neutronics code system used to prepare multigroup cross-section data which are pertinent to a particular study from a general purpose multigroup library of cross sections. Libraries have been prepared from ENDF/B which are suitable for thermal and fast fission reactors and for fusion blanket studies. The libraries include temperature dependent data, resonance cross sections represented by subgroup parameters and may contain photon as well as neutron data. The MIRANDA module includes a multiregion resonance calculation in slab, cylinder or cluster geometry, a homogeneous B L flux solution, and a group condensation facility. This report documents the modifications to an earlier version of MIRANDA and provides a complete user's manual

System THEMIS. Cross sections processing system from ENDF/B

Energy Technology Data Exchange (ETDEWEB)

Gonnord, J.

1983-09-01

The THEMIS system allowed to prepare a self punctual and multigroup library for codes solving the TRIPOLI-PROMETHEE transport equation, allowing comparisons with different methods and approximations. The contents of the THEMIS data base was fixed from its use by the PROMETHEE system (punctual Monte Carlo calculations, multigroup calculations, uncertainties analysis and sensitivity studies). The main characteristics of the THEMIS cross section processing system are briefly presented.

Obtaining incremental multigroup cross sections for CANDU super cells with reactivity devices

International Nuclear Information System (INIS)

Balaceanu, V.; Constantin, M.

2001-01-01

In the last 20 years a multigroup methodology WIMS - PIJXYZ (WP) was developed and validated at INR Pitesti for obtaining incremental cross sections for reactivity devices in CANDU reactors. This is an alternate methodology to the CANDU classic methodology (experimentally adjusted) based on the POWDERPUFS and MULTICELL computer codes. The 2D supercell calculation performed with the WIMS code, that is a NEA Data Bank transport code, and which produces multigroup cross sections (on 18 energy groups) for CANDU supercell material (standard and perturbed, with and without reactivity devices). To obtain an as correct as possible 3D modelling for the CANDU supercells containing reactivity devices, the WIMS cross sections are used as input data for the PIJXYZ code, thus obtaining homogenized cross sections for CANDU supercells. PIJXYZ is an integral transport code based on the formalism of the first collision probabilities. It is analogue to the SHETAN code and it was created for neutron analyzes at cell level for CANDU type reactors were the reactivity devices are perpendicular to the fuel channels. The coordinate system used in PIJXYZ is a mixed one, namely a rectangular-cylindrical system. The geometric model used in PIJXYZ is presented. The fuel beam is represented by a horizontal cylinder and the reactivity device by a vertical one both cylinders being immersed in the moderator. Two supercell types were considered: a perturbed supercell (containing a reactivity device) and the standard supercell were the place of reactivity device is occupied by the moderator. The incremental cross sections for reactivity device are obtained as differences between the homogenized over supercell cross sections (with reactivity device) and homogenized over standards supercell (without device) cross sections. The PIJXYZ computation may be done on an energy cutting with 2 up to 18 groups. The validation of VIMS - PIJXYZ was done on the basis of several benchmark and by comparison with

A comparison of the BUGLE-80, SAILOR, and ELXSIR neutron cross-section libraries for PWR pressure vessels surveillance dosimetry and shielding applications

International Nuclear Information System (INIS)

Basha, H.S.; Manahan, M.P.

1992-01-01

In this paper three multigroup neutron cross-section libraries are used in synthesized three-dimensional discrete ordinates transport analyses to investigate their similarities, differences, and results for pressurized water reactor (PWR) pressure vessel surveillance dosimetry and shielding applications. The calculated-to-experimental (C/E) rations and the calculated reaction rates of several fast reactions are compared for the BUGLE-80, SAILOR, and ELXSIR cross-section libraries at the 97-deg surveillance capsule of the San Onofre Nuclear Generation Station Unit 2 (SONGS-2) and at the 90- and 97-deg (C/E ratios only) cavity dosimetry locations for another PWR (referred to as Reactor X)

System THEMIS. Cross sections processing system from ENDF/B

International Nuclear Information System (INIS)

Gonnord, J.

1983-09-01

The THEMIS system allowed to prepare a self punctual and multigroup library for codes solving the TRIPOLI-PROMETHEE transport equation, allowing comparisons with different methods and approximations. The contents of the THEMIS data base was fixed from its use by the PROMETHEE system (punctual Monte Carlo calculations, multigroup calculations, uncertainties analysis and sensitivity studies). The main characteristics of the THEMIS cross section processing system are briefly presented [fr

Comparison of integral cross section values of several cross section libraries in the SAND-II format

International Nuclear Information System (INIS)

Zijp, W.L.; Nolthenius, H.J.

1976-09-01

A comparison of some integral cross-section values for several cross-section libraries in the SAND-II format is presented. The integral cross-section values are calculated with the aid of the spectrum functions for a Watt fission spectrum, a 1/E spectrum and a Maxwellian spectrum. The libraries which are considered here are CCC-112B, ENDF/B-IV, DETAN74, LAPENAS and CESNEF. These 5 cross-section libraries used have all the SAND-II format. Discrepancies between cross-sections in the different libraries are indicated but not discussed

COMPAR: system to compare multigroup cross sections generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 AND XLACS

International Nuclear Information System (INIS)

Anaf, J.; Chalhoub, E.S.

1987-11-01

A system, composed by the computer programs COMPAR and its interfaces, developed for comparing multigroup cross sections calculated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS, is presented. (author)

Verification of KARMA GEOM/TRPT Module with Given Multi-group Cross Sections

International Nuclear Information System (INIS)

Koo, Bon Seung; Hong, Ser Gi; Song, Jae Seung

2009-01-01

KAERI has developed a two-dimensional multigroup transport theory code KARMA (Kernel Analyzer by Ray-tracing Method for Fuel Assembly). KARMA uses CMFD (Coarse Mesh Finite Difference) accelerated MOC (Method of Characteristics) method for burnup calculation on a single fuel pin, a fuel assembly and a core consisting of rectangular array of fuel pins. KARMA code intends to be employed as a nuclear design tool for the Korean commercial pressurizer water reactor. Prior to the application to actual assembly designs, the code has to be approved by regularity agency. Therefore, it is essential that the reliability of KARMA code should be sufficiently evaluated against well-defined benchmark problems. In this paper, verification of GEOM/TRPT modules of KARMA was performed to confirm a reliability of the KARMA transport solution via comparisons with Monte Carlo calculations by using a consistent set of multi-group macroscopic cross-sections

Comparison of integral cross section values of several cross section libraries in the SAND-II format

International Nuclear Information System (INIS)

Zijp, W.L.; Nolthenius, H.J.

1978-01-01

A comparison of some integral cross section values for several cross section libraries in the SAND-II format is presented. The integral cross section values are calculated with aid of the spectrum functions for a Watt fission spectrum, a 1/E spectrum and a Maxwellian spectrum. The libraries which are considered here are CCC-112B, ENDF/B-IV, DETAN74, LAPENAS and CESNEF. These 5 cross section libraries used have all the SAND-II format. (author)

Collapsing of multigroup cross sections in optimization problems solved by means of the maximum principle of Pontryagin

International Nuclear Information System (INIS)

Anton, V.

1979-05-01

A new formulation of multigroup cross section collapsing based on the conservation of point or zone value of hamiltonian is presented. This attempt is proper to optimization problems solved by means of maximum principle of Pontryagin. (author)

Generation, Testing, and Validation of a WIMS-D/4 Multigroup Cross-Section Library Based on the JENDL-3.2 Nuclear Data

International Nuclear Information System (INIS)

Rahman, Mafizur; Takano, Hideki

2001-01-01

A new 69-group library of multigroup constants for the lattice code WIMS-D/4 has been generated with an improved resonance treatment, processing nuclear data from JENDL-3.2 by NJOY91.108. A parallel ENDF/B-VI based library has also been constructed for intercomparison of results. Benchmark calculations for a number of thermal reactor critical assemblies of both uranium and plutonium fuels have been performed with the code WIMS-D/4.1 with its three different libraries: the original WIMS library (NEA-0329/10) and the new ENDF/B-VI and JENDL-3.2 based libraries. The results calculated with both ENDF and JENDL based libraries show a similar tendency and are found in better agreement with the experimental values. Benchmark parameters are further calculated with the comprehensive lattice code SRAC95. The results from SRAC95 and WIMS-D/4.1 (both using JENDL-3.2 based libraries) agree well with each other. The new library is also verified for its applicability to mixed-oxide cores of varying plutonium contents

SCAMPI: A code package for cross-section processing

International Nuclear Information System (INIS)

Parks, C.V.; Petrie, L.M.; Bowman, S.M.; Broadhead, B.L.; Greene, N.M.; White, J.E.

1996-01-01

The SCAMPI code package consists of a set of SCALE and AMPX modules that have been assembled to facilitate user needs for preparation of problem-specific, multigroup cross-section libraries. The function of each module contained in the SCANTI code package is discussed, along with illustrations of their use in practical analyses. Ideas are presented for future work that can enable one-step processing from a fine-group, problem-independent library to a broad-group, problem-specific library ready for a shielding analysis

SCAMPI: A code package for cross-section processing

Energy Technology Data Exchange (ETDEWEB)

Parks, C.V.; Petrie, L.M.; Bowman, S.M.; Broadhead, B.L.; Greene, N.M.; White, J.E.

1996-04-01

The SCAMPI code package consists of a set of SCALE and AMPX modules that have been assembled to facilitate user needs for preparation of problem-specific, multigroup cross-section libraries. The function of each module contained in the SCANTI code package is discussed, along with illustrations of their use in practical analyses. Ideas are presented for future work that can enable one-step processing from a fine-group, problem-independent library to a broad-group, problem-specific library ready for a shielding analysis.

COMPAR: A system for comparing multigroup cross-sections generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS

International Nuclear Information System (INIS)

Anaf, J.; Chalhoub, E.S.

1988-02-01

A system consisting of the COMPAR computer program and its interfaces which was developed for comparing multigroup cross-sections generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS is presented. (author). 13 refs

Cross section library DOSCROS77 (in the SAND-II format)

International Nuclear Information System (INIS)

Zijp, W.L.; Nolthenius, H.J.; Borg, N.J.C.M. van der.

1977-08-01

The dosimetry cross section library DOSCROS77 is documented with tables, plots and cross section values averaged over a few reference spectra. This library is based on the ENDF/B-IV dosimetry file, supplemented with some other evaluations. The total number of reaction cross section sets incorporated in this library is 49 (+3 cover cross sections sets). The cross section data are available in a format which is suitable for the program SAND-II

Use of CPXSD for generation of effective fast multigroup libraries for pressure vessel fluence calculations

International Nuclear Information System (INIS)

Alpan, F. Arzu; Haghighat, Alireza

2008-01-01

Multigroup (i.e., broad-group) libraries play a significant role in the accuracy of transport calculations. There are several broad-group libraries available for particular applications. For example the 47-neutron (26 fast groups), 20-gamma-group BUGLE libraries are commonly used for light water reactor shielding and pressure vessel dosimetry problems. However, there is no publicly available methodology to construct group structures for a problem and objective of interest. Therefore, we have developed the Contribution and Point-wise Cross-Section Driven (CPXSD) methodology, which constructs effective fine-and broad-group structures. In this paper, we use the CPXSD methodology to construct broad-group structures for fast neutron dosimetry problems. It is demonstrated that the broad-group libraries generated from CPXSD constructed group structures, while only 14 groups (rather than 26 groups) in the fast energy range are in good agreement (similar to 1 %-2 %) with the fine-group library from which they were derived, in reaction rate calculations.

The LAW library

International Nuclear Information System (INIS)

Green, N.M.; Parks, C.V.; Arwood, J.W.

1989-01-01

The 238 group LAW library is a new multigroup library based on ENDF/B-V data. It contains data for 302 materials and will be distributed by the Radiation Shielding Information Center, located at Oak Ridge National Laboratory. It was generated for use in neutronics calculations required in radioactive waste analyses, though it has equal utility in any study requiring multigroup neutron cross sections

Neutron-photon multigroup cross sections for neutron energies less than or equal to400 MeV. Revision 1

International Nuclear Information System (INIS)

Alsmiller, R.G. Jr.; Barnes, J.M.; Drischler, J.D.

1986-02-01

Multigroup cross sections (66 neutron groups and 22 photon groups) are described for neutron energies from thermal to 400 MeV. The elements considered are hydrogen, 10 B, 11 B, carbon, nitrogen, oxygen, sodium, magnesium, aluminum, silicon, sulfur, potassium, calcium, chromium, iron, nickel, tungsten, and lead. The cross section data presented are a revision of similar data presented previously. In the case of iron, transport calculations using the earlier and the revised cross sections are presented and compared, and significant differences are found. The revised cross sections are available from the Radiation Shielding information Center of the Oak Ridge National Laboratory. 32 refs., 5 figs., 3 tabs

MINX: a multigroup interpretation of nuclear X-sections from ENDF/B

International Nuclear Information System (INIS)

Weisbin, C.R.; Soran, P.D.; MacFarlane, R.E.; Harris, D.R.; LaBauve, R.J.; Hendricks, J.S.; White, J.E.; Kidman, R.B.

1976-09-01

MINX calculates fine-group averaged infinitely dilute cross sections, self-shielding factors, and group-to-group transfer matrices from ENDF/B-IV data. Its primary purpose is to generate pseudo-composition independent multigroup libraries in the standard CCCC-III interface formats for use in the design and analysis of nuclear systems. MINX incorporates and improves upon the resonance capabilities of existing codes such as ETOX and ENDRUN and the high-Legendre-order transfer matrices of ETOG and SUPERTOG. Group structure, Legendre order, weight function, temperature, dilutions, and processing tolerances are all under user control. Paging and variable dimensioning allow very large problems to be run. Both CDC and IBM versions of MINX are available

Preparation of lumped fission product (FP) cross sections for a multigroup library

International Nuclear Information System (INIS)

Ono, S.; Corcuera, R.P.

1984-01-01

A method for the calculation of lumped Fission Product (FP) cross sections has been developed. The group constants fo each nuclide are generated by NJOY code, based on ENDF/B-V data. In this first version, cross section of 28 nuclides are lumped for typical characteristics of Binary Breeder Reactor (BBR). One energy group calculations are made for a 1000 MWe fast reactor to verify the influence of burnup, number of FP and fuel composition on the lumped fission product cross sections. (Author) [pt

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)

SERKON program for compiling a multigroup library to be used in BETTY calculation

International Nuclear Information System (INIS)

Nguyen Phuoc Lan.

1982-11-01

A SERKON-type program was written to compile data sets generated by FEDGROUP-3 into a multigroup library for BETTY calculation. A multigroup library was generated from the ENDF/B-IV data file and tested against the TRX-1 and TRX-2 lattices with good results. (author)

Group cross-section processing method and common nuclear group cross-section library based on JENDL-3 nuclear data file

International Nuclear Information System (INIS)

Hasegawa, Akira

1991-01-01

A common group cross-section library has been developed in JAERI. This system is called 'JSSTDL-295n-104γ (neutron:295 gamma:104) group constants library system', which is composed of a common 295n-104γ group cross-section library based on JENDL-3 nuclear data file and its utility codes. This system is applicable to fast and fusion reactors. In this paper, firstly outline of group cross-section processing adopted in Prof. GROUCH-G/B system is described in detail which is a common step for all group cross-section library generation. Next available group cross-section libraries developed in Japan based on JENDL-3 are briefly reviewed. Lastly newly developed JSSTDL library system is presented with some special attention to the JENDL-3 data. (author)

MARS-ORNL, Processing Program Collection for AMPX, CCCC, ANISN, DOT, MORSE Format Library. LINX, MINX Library Utility, Data Merge. BINX, MINX Utility and SPHINX Utility, BCD to BIN Library Conversion. CINX, MINX Utility and SPHINX Utility, Library Data Collapsing

International Nuclear Information System (INIS)

2001-01-01

Description of problem or function: MARS-ORNL is a selection of computer codes for the generation of problem-dependent multigroup cross section libraries. They are selected modules from the AMPX-2 system for AMPX interface format libraries, LASL codes for CCCC interfaces, and processing codes for libraries to be used by ANISN, DOT, or MORSE codes. The codes in the collection are used in connection with the following DLC data libraries: ZZ-LIB-IV (DLC-0040), ZZ-VITAMIN-C (DLC-0041), VITAMIN-4C (DLC-0053), ZZ-CLEAR/42B (DLC-0042), ZZ-CSRL/43B (DLC-0043), and EPRMASTER (DLC-0052). The functions of these processing codes are briefly described: A. AMPX Modules: AIM: Converts AMPX Master Interface Files from EBCDIC to binary form and back. AJAX: Merges, collects, assembles, re-orders, joins, and copies selected nuclides from AMPX Master Interfaces. BONAMI: Accesses Bondarenko factors from an AMPX Master Library and performs resonance self-shielding calculations. CHOX: Produces a coupled interface library in AMPX format by combining neutron libraries (generated by module XLACS), gamma libraries (generated by module SMUG), and photon production libraries (generated by module LAPHNGAS). CHOXM: Combines self-shielding factors as generated by the code SPHINX (PSR-0129) and an infinite dilution neutron master interface (generated by XLACS) to generate a self-shielded neutron AMPX Interface File. The interface produced by CHOXM is an input to the NITAWL module of AMPX. CHOXM is a modified version of CHOX. COMAND: Collapses ANISN cross section libraries. DIAL: Produces edits from AMPX Master Interfaces. ICE-II: Accepts cross sections from an AMPX working library and produces mixed cross sections in four formats: (1) AMPX working library format; (2) ANISN format; (3) group-independent ANISN format; (4) Monte Carlo processed cross section library format. NITAWL: Produces self-shielded and working cross section libraries in the formats required by the ANISN, DOT, or MORSE codes

Validation of multigroup neutron cross sections for the Advanced Neutron Source against the FOEHN critical experimental measurements

International Nuclear Information System (INIS)

Smith, L.A.; Gehin, J.C.; Worley, B.A.; Renier, J.P.

1994-01-01

The FOEHN critical experiments were analyzed to validate the use of multigroup cross sections in the design of the Advanced Neutron Source. Eleven critical configurations were evaluated using the KENO, DORT, and VENTURE neutronics codes. Eigenvalue and power density profiles were computed and show very good agreement with measured values

FENDL/MG-2.0 and FENDL/MC-2.0. The processed cross-section libraries for neutron photon transport calculations. Version 1, March 1997. Summary documentation

International Nuclear Information System (INIS)

Wienke, H.; Herman, M.

1998-01-01

Evaluated neutron reaction data and photon-atom interaction cross sections for materials contained in the general purpose Fusion Evaluated Nuclear Data Library (FENDL/E2.0) have been processed with the NJOY code system into VITAMIN-J multigroup structure, for use in discrete-ordinates transport codes, and into continuous energy ACE format, for use in the Monte Carlo transport code MCNP. This document summarizes the resulting data libraries FENDL/MG-2.0 version 1 and FENDL/MC-2.0 version 1. The data are available costfree from the IAEA Nuclear Data Section online or on magnetic tape. (author)

Calculation of multigroup constants in WIMS format with programs fedgroup and flange and comparison of the results obtained using different evaluated libraries

International Nuclear Information System (INIS)

Trkov, A.; Budnar, M.; Copic, M.; Perdan, A.; Ravnik, M.

1982-01-01

Multigroup constants for 1-H-1, 92-U-235, and 92-U-238 have been calculated. Averaged cross-sections and other constants have been prepared in the WIMS 69-group format. Comparison has been made between group constants obtained with several evaluated libraries (KEDAK-3 1975, 1979, ENDF/B-4, ENDF/B-5) and the WIMS-D library. Observed differences are most pronounced in the resonance and fast region. From test runs on fuel cell with the WIMS program it can be deduced that these differences affect the fewgroup constants significantly. (author)

Neutron-photon multigroup cross sections for neutron energies less than or equal to400 MeV. Revision 1

International Nuclear Information System (INIS)

Alsmiller, R.G. Jr.; Barnes, J.M.; Drischler, J.D.

1986-01-01

For a variety of applications, e.g., accelerator shielding design, neutrons in radiotherapy, radiation damage studies, etc., it is necessary to carry out transport calculations involving medium-energy (greater than or equal to20 MeV) neutrons. A previous paper described neutron-photon multigroup cross sections in the ANISN format for neutrons from thermal to 400 MeV. In the present paper the cross-section data presented previously have been revised to make them agree with available experimental data. 7 refs., 1 fig

Development of modern CANDU PHWR cross-section libraries for SCALE

International Nuclear Information System (INIS)

Shoman, Nathan T.; Skutnik, Steven E.

2016-01-01

Highlights: • New ORIGEN libraries for CANDU 28 and 37-element fuel assemblies have been created. • These new reactor data libraries are based on modern ENDF/B-VII.0 cross-section data. • The updated CANDU data libraries show good agreement with radiochemical assay data. • Eu-154 overestimated when using ENDF-VII.0 due to a lower thermal capture cross-section. - Abstract: A new set of SCALE fuel lattice models have been developed for the 28-element and 37-element CANDU fuel assembly designs using modern cross-section data from ENDF-B/VII.0 in order to produce new reactor data libraries for SCALE/ORIGEN depletion analyses. These new libraries are intended to provide users with a convenient means of evaluating depletion of CANDU fuel assemblies using ORIGEN through pre-generated cross sections based on SCALE lattice physics calculations. The performance of the new CANDU ORIGEN libraries in depletion analysis benchmarks to radiochemical assay data were compared to the previous version of the CANDU libraries provided with SCALE (based on WIMS-AECL models). Benchmark comparisons with available radiochemical assay data indicate that the new cross-section libraries perform well at matching major actinide species (U/Pu), which are generally within 1–4% of experimental values. The library also showed similar or better results over the WIMS-AECL library regarding fission product species and minor actinoids (Np, Am, and Cm). However, a notable exception was in calculated inventories of "1"5"4Eu and "1"5"5Eu, where the new library employing modern nuclear data (ENDF/B-VII.0) performed substantially poorer than the previous WIMS-AECL library (which used ENDF-B/VI.8 cross-sections for these species). The cause for this discrepancy appears to be due to differences in the "1"5"4Eu thermal capture cross-section between ENDF/B-VI.8 and ENDF/B-VII.0, an effect which is exacerbated by the highly thermalized flux of a CANDU heavy water reactor compared to that of a typical

AMPX-77: A modular code system for generating coupled multigroup neutron-gamma cross-section libraries from ENDF/B-IV and/or ENDF/B-V

Energy Technology Data Exchange (ETDEWEB)

Greene, N.M.; Ford, W.E. III; Petrie, L.M.; Arwood, J.W.

1992-10-01

AMPX-77 is a modular system of computer programs that pertain to nuclear analyses, with a primary emphasis on tasks associated with the production and use of multigroup cross sections. AH basic cross-section data are to be input in the formats used by the Evaluated Nuclear Data Files (ENDF/B), and output can be obtained in a variety of formats, including its own internal and very general formats, along with a variety of other useful formats used by major transport, diffusion theory, and Monte Carlo codes. Processing is provided for both neutron and gamma-my data. The present release contains codes all written in the FORTRAN-77 dialect of FORTRAN and wig process ENDF/B-V and earlier evaluations, though major modules are being upgraded in order to process ENDF/B-VI and will be released when a complete collection of usable routines is available.

AMPX-77: A modular code system for generating coupled multigroup neutron-gamma cross-section libraries from ENDF/B-IV and/or ENDF/B-V

International Nuclear Information System (INIS)

Greene, N.M.; Ford, W.E. III; Petrie, L.M.; Arwood, J.W.

1992-10-01

AMPX-77 is a modular system of computer programs that pertain to nuclear analyses, with a primary emphasis on tasks associated with the production and use of multigroup cross sections. AH basic cross-section data are to be input in the formats used by the Evaluated Nuclear Data Files (ENDF/B), and output can be obtained in a variety of formats, including its own internal and very general formats, along with a variety of other useful formats used by major transport, diffusion theory, and Monte Carlo codes. Processing is provided for both neutron and gamma-my data. The present release contains codes all written in the FORTRAN-77 dialect of FORTRAN and wig process ENDF/B-V and earlier evaluations, though major modules are being upgraded in order to process ENDF/B-VI and will be released when a complete collection of usable routines is available

AMPX: a modular system for multigroup cross-section generation and manipulation

International Nuclear Information System (INIS)

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

1978-01-01

The AMPX system, developed at the Oak Ridge National Laboratory over the past seven years, is a collection of computer programs in a modular arrangement. Starting with ENDF-formatted nuclear data files, the system includes a full range of features needed to produce and use multigroup neutron, gamma-ray production, and gamma-ray interaction cross-section data. The balance between production and analysis is roughly even; thus, the system serves a wide variety of needs. The modularity is particularly attractive, since it allows the user to choose an arbitrary execution sequence from the approximately 40 to 50 modules available in the system. The modularity also allows selection from different treatments; e.g., the Nordheim method, a full-blown integral transport calculation, the Bondarenko method, or other alternative can be selected for resonance shielding. 2 figures

Differences between cross-section libraries for neutron dosimetry

International Nuclear Information System (INIS)

Tardelli, T.C.; Stecher, L.C.; Coelho, T.S.; Castro, V.A. De; Cavalieri, T.A.; Menzel, F.; Giarola, R.S.; Domingos, D.B.; Yoriyaz, H.

2013-01-01

Absorbed dose calculations depend on a consistent set of nuclear data used in simulations in computer codes. Nuclear data are stored in libraries, however, the information available about the differences in dose caused by different libraries are rare. The libraries are processed by a computer system to be able to be used by a radiation transport code. One of the systems capable of processing nuclear data is the NJOY system. The objective of this study is to evaluate the nuclear data libraries for neutrons available in the literature, and to quantify the differences in absorbed dose obtained using the libraries JENDL 4.0, JEFF 3.3.1 and ENDF/B.VII. The absorbed dose calculation was performed on a simple geometric model, as spheres, and in anthropomorphic model of the human body based on the ICRP-110 for neutron transport simulation using the MCNP5 code. The results were compared with literature data. The results obtained with cross sections from the libraries JEFF and ENDF/B.VII have shown to be identical in most cases, except for one case where the difference has exceeded 10%. The results obtained with JENDL library has shown to be considerably different in most cases comparing to other two libraries. Some differences were over 200%. The dose calculations showed differences between the libraries, which is justified by differences in the cross sections. It has been observed that the cross sections values of certain nuclides assume quite different values in different libraries. These differences in turn cause considerable differences in dose calculations. (author)

Cross-section library and processing techniques within the SCALE system

International Nuclear Information System (INIS)

Westfall, R.M.

1986-01-01

A summary of each of the SCALE system features involved in problem-dependent cross section processing is presented. These features include criticality libraries, shielding libraries, the Standard Composition Library, the SCALE functional modules: BONAMI-S, NITAWL-S, XSDRNPM-S, ICE-S, and the Material Information Processor. The automated procedure for cross-section processing is described with examples. 15 refs

Cross-section libraries and kerma factors

International Nuclear Information System (INIS)

Little, R.C.; MacFarlane, R.E.; Seamon, R.E.

1991-01-01

A large amount of data is required in order to accurately simulate various aspects of Cold Neutron Sources using radiation transport codes such as MCNP and TWODANT. In particular, the following types of data are needed: couple neutron/photon transport libraries, neutron thermal S(α,β) data, response function data (including energy deposition), and proton interaction data. This paper concentrates on the coupled neutron/photon transport libraries and energy deposition. Data libraries available to radiation transport codes are obtained as a result of efforts in many areas, including differential and integral measurements, theoretical model codes, data evaluations, data processing, and data testing. A wide variety of data libraries are available to users of radiation transport codes, including pointwise and multigroup libraries. At Los Alamos, the authors generally recommend the use of data libraries derived from ENDF/B-V. It is often important to know how much energy is deposited in various regions of a device. This problem is typically modeled in radiation transport codes by folding the calculated fluences with an energy-dependent 'heating number'. The heating number represents the average energy deposited locally per collision. Calculation of these heating numbers from evaluated data libraries is fraught with difficulty. Many past difficulties related to energy deposition should be resolved by the release of ENDF/B-VI

Benchmarking of multigroup neutron cross sections libraries on neutron transmission through WWER-440 vessel

International Nuclear Information System (INIS)

Ilieva, K.; Belousov, S.; Apostolov, T.

1998-01-01

The verification of calculated neutron fluence onto the WWER-440/230 pressure vessel is very topical task in particular referring that some of this type of reactors have been operated the major part of its design lifetime. Since the induced activity from the neutron irradiation onto the elements is a simple response of neutron flux the neutron fluence verification usually is done using the measured activity of radionuclides produced during reactor operation. Calculational and experimental results of 54 Mn induced activity of scraps from inner wall of Unit 1 reactor pressure vessel after 18th cycle and detectors irradiated behind the vessel during the 18th cycle of Unit 1 at Kozloduy NPP as well as neutron flux attenuation through the WWER-440/230 pressure vessel are presented. Neutron cross sections libraries generated on the base of ENDF/B-IV and ENDF/B-VI have been used in the calculations. The comparative analysis of evaluated activities and attenuation coefficient demonstrates the better reliability of the neutron fluence calculations by the libraries based on ENDF/B-VI than by ones on ENDF/B-IV. The extreme rarity of data for the activity of scraps from the WWER-440 reactor vessel and its combination with the data for the detectors irradiated behind the vessel makes them especially attractive for verification of calculational methods of neutron fluence onto the WWER-440 vessel with dummy cassettes loading. (author)

Problem Oriented Neutron-Gamma Cross Sections Libraries for WWER-440 and WWER-1000 Shielding and Reactor Vessel Dosimetry Application

International Nuclear Information System (INIS)

Belousov, S.; Antonov, S.; Ilieva, K.

1997-01-01

The 47 neutron and 20 gamma group libraries BGL-440 and BGL-1000 for the shielding and reactor vessel dosimetry application have been generated for WWER-440 and WWER-1000 by collapsing the VITAMIN-B6 library (199 neutron and 42 gamma groups on the base of ENDF/B-6). The first parts of the libraries for neutron-gamma transport calculation, BGL-440-1 (150 nuclides) and BGL-1000-1 (140 nuclides), have been generated by a modified version of SAS1X control module of the SCALE system. The appropriate zone-average neutron flux had been used for these sub-libraries collapsing. The BGL-440-2 and BGL-1000-2 sub-libraries consist of cross sections for all 120 nuclides of VITAMIN-B6, for calculation of the transport through non-reactor materials of dosimeters, capsules, specimens which may be placed in the cavity behind the reactor vessel. The neutron spectrum just beyond the RPV had been used for this collapsing. As the first test the comparative calculations of the neutron flux on/behind the WWER-1000 reactor vessel have been realised using the libraries BGL-1000 and BUGLE, intended for the American PWR reactors. The integral neutron flux values by BGL-1000 and BUGLE differ by 3% onto the vessel, and 5% behind the vessel. This result shows that the calculations of the neutron flux responses for the WWER vessel surveillance, especially in locations behind the WWER vessel have to be done by the appropriate BGL library. Key words: neutron transport, multigroup neutron cross section libraries

MOX Cross-Section Libraries for ORIGEN-ARP

International Nuclear Information System (INIS)

Gauld, I.C.

2003-01-01

The use of mixed-oxide (MOX) fuel in commercial nuclear power reactors operated in Europe has expanded rapidly over the past decade. The predicted characteristics of MOX fuel such as the nuclide inventories, thermal power from decay heat, and radiation sources are required for design and safety evaluations, and can provide valuable information for non-destructive safeguards verification activities. This report describes the development of computational methods and cross-section libraries suitable for the analysis of irradiated MOX fuel with the widely-used and recognized ORIGEN-ARP isotope generation and depletion code of the SCALE (Standardized Computer Analyses for Licensing Evaluation) code system. The MOX libraries are designed to be used with the Automatic Rapid Processing (ARP) module of SCALE that interpolates appropriate values of the cross sections from a database of parameterized cross-section libraries to create a problem-dependent library for the burnup analysis. The methods in ORIGEN-ARP, originally designed for uranium-based fuels only, have been significantly upgraded to handle the larger number of interpolation parameters associated with MOX fuels. The new methods have been incorporated in a new version of the ARP code that can generate libraries for low-enriched uranium (LEU) and MOX fuel types. The MOX data libraries and interpolation algorithms in ORIGEN-ARP have been verified using a database of declared isotopic concentrations for 1042 European MOX fuel assemblies. The methods and data are validated using a numerical MOX fuel benchmark established by the Organization for Economic Cooperation and Development (OECD) Working Group on burnup credit and nuclide assay measurements for irradiated MOX fuel performed as part of the Belgonucleaire ARIANE International Program

Neutron cross-section library for SAND-2 and its service program

International Nuclear Information System (INIS)

Berzonis, M.A.; Bondars, Kh.Ya.; Lapenas, A.A.

1978-01-01

The logical structure of the neutron cross-section library used in the SAND-2 program complex is considered. The organization of the DSIG01 program creating and servicing the neutron cross section library is described. The DSIG 01 program is written on FORTRAN and permits to create the neutron cross section library on the ES computer magnetic discs operating under the control of the ES operating system and to perform certain manipulations therewith

SHAMSI, 48 group cross-section library for fusion nucleonics analysis

International Nuclear Information System (INIS)

Ponti, C.; Abbas, Tayyab.

1982-01-01

A P 3 48 group coupled neutron gamma-ray (34 N - 14 G) cross-section library is produced and validated for neutronic studies in fusion reactor blanket/shield. This report describes the library content, the procedure adopted and the results of the calculations performed for testing the cross sections

Development and testing of multigroup library with correction of self-shielding effects in fusion-fission hybrid reactor

International Nuclear Information System (INIS)

Zou Jun; He Zhaozhong; Zeng Qin; Qiu Yuefeng; Wang Minghuang

2010-01-01

A multigroup library HENDL2.1/SS (Hybrid Evaluated Nuclear Data Library/Self-Shielding) based on ENDF/B-VII.0 evaluate data has been generated using Bondarenko and flux calculator method for the correction of self-shielding effect of neutronics analyses. To validate the reliability of the multigroup library HENDL2.1/SS, transport calculations for fusion-fission hybrid system FDS-I were performed in this paper. It was verified that the calculations with the HENDL2.1/SS gave almost the same results with MCNP calculations and were better than calculations with the HENDL2.0/MG which is another multigroup library without self-shielding correction. The test results also showed that neglecting resonance self-shielding caused underestimation of the K eff , neutron fluxes and waste transmutation ratios in the multigroup calculations of FDS-I.

Remarks on the comparison of cross section libraries for neutron metrology

International Nuclear Information System (INIS)

Zijp, W.L.; Nolthenius, H.J.; Appelman, K.H.

1977-01-01

Cross section libraries in a 620 group structure were available from different origin: CCC-112B, DETAN-74 and ENDF/B-IV. For a few well known neutron spectra (CFRMF spectrum, ΣΣ spectrum, fission neutron spectrum, HFR neutron spectrum) a comparison was made of the available experimental reaction rates in foil detectors and the reaction rates as calculated with the different cross section libraries. This investigation is dealing with the consistency of cross section data within a library, and the consistency of activity data in actual reaction rate determinations. Some preliminary conclusions are given

Comparative analysis among several cross section sets

International Nuclear Information System (INIS)

Caldeira, A.D.

1983-01-01

Critical parameters were calculated using the one dimensional multigroup transport theory for several cross section sets. Calculations have been performed for water mixtures of uranium metal, plutonium metal and uranium-thorium oxide, and for metallics systems, to determine the critical dimensions of geometries (sphere and cylinder). For this aim, the following cross section sets were employed: 1) multigroup cross section sets obtained from the GAMTEC-II code; 2) the HANSEN-ROACH cross section sets; 3) cross section sets from the ENDF/B-IV, processed by the NJOY code. Finally, we have also calculated the corresponding critical radius using the one dimensional multigroup transport DTF-IV code. The numerical results agree within a few percent with the critical values obtained in the literature (where the greatest discrepancy occured in the critical dimensions of water mixtures calculated with the values generated by the NJOY code), a very good results in comparison with similar works. (Author) [pt

Production and testing of the VITAMIN-B6 fine-group and the BUGLE-93 broad-group neutron/photon cross-section libraries derived from ENDF/B-VI nuclear data

International Nuclear Information System (INIS)

Ingersoll, D.T.; White, J.E.; Wright, R.Q.; Hunter, H.T.; Slater, C.O.; Greene, N.M.; MacFarlane, R.E.

1993-01-01

A new multigroup cross-section library based on ENDF/B-VI data has been produced and tested for light water reactor shielding and reactor pressure vessel dosimetry applications. The broad-group library is designated BUGLE-93. The processing methodology is consistent with ANSI/ANS 6.1.2, since the ENDF data were first processed into a fine-group, ''pseudo problem-independent'' format and then collapsed into the final broad-group format. The fine-group library is designated VITAMIN-B6. An extensive integral data testing effort was also performed. In general, results using the new data show significant improvements relative to earlier ENDF data

Preparation and benchmarking of ANSL-V cross sections for advanced neutron source reactor studies

International Nuclear Information System (INIS)

Arwood, J.W.; Ford, W.E. III; Greene, N.M.; Petrie, L.M.; Primm, R.T. III; Waddell, M.W.; Webster, C.C.; Westfall, R.M.; Wright, R.Q.

1987-01-01

Research and development for the advanced neutron source (ANS) reactor is being funded by the US Dept. of Energy. This reactor is to provide the world's most intense steady-state source of low-energy neutrons for a national experimental user facility. Pseudo-problem-independent, multigroup cross-section libraries were generated to support ANS design work. The libraries, designated ANSL-V, are data bases in AMPX master format for subsequent generation of problem-dependent cross sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, and MORSE. Included in ANSL-V are 123-material P 3 neutron, 46-material P 0 or P 6 secondary gamma-ray production (SGRP), and 34-material P 6 gamma-ray interaction (GRI) libraries

ACT-1000. Group activation cross-section library for WWER-1000 type reactors

Energy Technology Data Exchange (ETDEWEB)

Zolotarev, K I; Pashchenko, A B [National Research Centre - A.I. Leipunsky Institute for Physics and Power Engineering, Obninsk (Russian Federation)

2001-10-01

The ACT-1000, a problem-oriented library of group-averaged activation cross-sections for WWER-1000 type reactors, is based on evaluated microscopic cross-section data files. The ACT-1000 data library was designed for calculating induced activity for the main dose-generated nuclides contained in WWER-1000 structural materials. In preparing the ACT-1000 library, 47 group-averaged cross-section data for the 10{sup -9}-17.33 MeV energy range were used to calculate the spatial-energy neutron flux distribution. (author)

Cross sections in 25 groups obtained from ENDF/B-IV and ENDL/78 libraries, processed with GALAXY and NJOY computer codes

International Nuclear Information System (INIS)

Chalhoub, E.S.; Corcuera, R.P.

1982-01-01

The discrepancies existing between ENDF/B-IV and ENDL/78 libraries, in diferent energy regions are identified, and the order of the differences in multigroup sections are determined, when GALAXY or NJOY computer codes are used. (E.G.) [pt

Graphs of the cross sections in the recommended Monte Carlo cross-section library at the Los Alamos Scientific Laboratory

International Nuclear Information System (INIS)

Soran, P.D.; Seamon, R.E.

1980-05-01

Graphs of all neutron cross sections and photon production cross sections on the Recommended Monte Carlo Cross Section (RMCCS) library have been plotted along with local neutron heating numbers. Values for anti ν, the average number of neutrons per fission, are also given

COVFILS: 30-group covariance library based on ENDF/B-V

International Nuclear Information System (INIS)

Muir, D.W.; LaBauve, R.J.

1981-03-01

A library of 30-group cross sections and covariances called COVFILS has been prepared from ENDF/B-V data using the NJOY code system. COVFILS includes data on the total cross section, scattering cross sections, and the most important absorption cross sections for 1 H, 10 B, C, 16 O, Cr, Fe, Ni, Cu, and Pb. This report contains detailed descriptions of various features of the library, a listing of a FORTRAN retrieval program, and 143 plots of the multigroup cross-section uncertainties and their correlations

A new modelling of the multigroup scattering cross section in deterministic codes for neutron transport

International Nuclear Information System (INIS)

Calloo, A.A.

2012-01-01

In reactor physics, calculation schemes with deterministic codes are validated with respect to a reference Monte Carlo code. The remaining biases are attributed to the approximations and models induced by the multigroup theory (self-shielding models and expansion of the scattering law using Legendre polynomials) to represent physical phenomena (resonant absorption and scattering anisotropy respectively). This work focuses on the relevance of a polynomial expansion to model the scattering law. Since the outset of reactor physics, the latter has been expanded on a truncated Legendre polynomial basis. However, the transfer cross sections are highly anisotropic, with non-zero values for a very small range of the cosine of the scattering angle. Besides, the finer the energy mesh and the lighter the scattering nucleus, the more exacerbated is the peaked shape of this cross section. As such, the Legendre expansion is less suited to represent the scattering law. Furthermore, this model induces negative values which are non-physical. In this work, various scattering laws are briefly described and the limitations of the existing model are pointed out. Hence, piecewise-constant functions have been used to represent the multigroup scattering cross section. This representation requires a different model for the diffusion source. The discrete ordinates method which is widely employed to solve the transport equation has been adapted. Thus, the finite volume method for angular discretization has been developed and implemented in Paris environment which hosts the S n solver, Snatch. The angular finite volume method has been compared to the collocation method with Legendre moments to ensure its proper performance. Moreover, unlike the latter, this method is adapted for both the Legendre moments and the piecewise-constant functions representations of the scattering cross section. This hybrid-source method has been validated for different cases: fuel cell in infinite lattice

Graphs of the cross sections in the Alternate Monte Carlo Cross Section library at the Los Alamos Scientific Laboratory

International Nuclear Information System (INIS)

Seamon, R.E.; Soran, P.D.

1980-06-01

Graphs of all neutron cross sections and photon production cross sections on the Alternate Monte Carlo Cross Section (AMCCS) library have been plotted along with local neutron heating numbers. The values of ν-bar, the average number of neutrons per fission, are also plotted for appropriate isotopes

DOSCROS81. ECN Cross-Section Library for neutron dosimetry. Summary of contents and documentation

International Nuclear Information System (INIS)

Lemmel, H.D.

1982-01-01

This document summarizes the contents and documentation of the Cross Section Library DOSCROS81 (640 groups in an extended SAND-II format). The library is based on ENDF/B-5 dosimetry file, supplemented with some other evaluations. The total number of reaction cross section sets incorporated in this library is 70 (+3 cover cross section sets). The entire library can be obtained free of charge from the IAEA Nuclear Data Section. A revised version is called DOSCROS81A. (author)

Creation and validation of a neutron-gamma coupled multigroup cross section library

International Nuclear Information System (INIS)

Devan, K.; Gopalakrishnan, V.; Lee, S.M.

1995-01-01

The task of creating our own neutron-gamma coupled library was taken up. By using 1985 version of NJOY code system, a coupled set called IGC-DE4-S1 in ANISN format for 25 nuclides has been arrived at based on ENDF/B-IV neutron library and DLC-99 gamma library, with Legendre order of up to 5. The flow chart for the creation of coupled set is given. 5 refs, 1 fig., 3 tabs

Formats and processing of evaluated nuclear data into multigroup cross-sections

International Nuclear Information System (INIS)

Motta, M.

1984-01-01

The first part of these lectures concerns the data in nuclear files and their manipulation. The structure of the data files as divided into the resonance region (subdivided into the resolved and the unresolved regions) and the continuum region is presented. The reactions concerned are the elastic scattering; the radiative capture and the fission methods for averaging the cross sections are given. Then, the group averaging formulas and the self-shielding factors are presented in some detail. The second part concerns a presentation of nuclear data files handling and conversion. The main libraries are listed and several maintenance computer codes presented. The way the conversion among different files is handled is also presented. The listings of several BASIC programs for different cross section calculations are given. These codes are self-guided

CLUB - a multigroup integral transport theory code for lattice calculations of PHWR cells

International Nuclear Information System (INIS)

Krishnani, P.D.

1992-01-01

The computer code CLUB has been developed to calculate lattice parameters as a function of burnup for a pressurised heavy water reactor (PHWR) lattice cell containing fuel in the form of cluster. It solves the multigroup integral transport equation by the method based on combination of small scale collision probability (CP) method and large scale interface current technique. The calculations are performed by using WIMS 69 group cross section library or its condensed versions of 27 or 28 group libraries. It can also compute Keff from the given geometrical buckling in the input using multigroup diffusion theory in fundamental mode. The first order differential burnup equations can be solved by either Trapezoidal rule or Runge-Kutta method. (author). 17 refs., 2 figs

Multi-Group Covariance Data Generation from Continuous-Energy Monte Carlo Transport Calculations

International Nuclear Information System (INIS)

Lee, Dong Hyuk; Shim, Hyung Jin

2015-01-01

The sensitivity and uncertainty (S/U) methodology in deterministic tools has been utilized for quantifying uncertainties of nuclear design parameters induced by those of nuclear data. The S/U analyses which are based on multi-group cross sections can be conducted by an simple error propagation formula with the sensitivities of nuclear design parameters to multi-group cross sections and the covariance of multi-group cross section. The multi-group covariance data required for S/U analysis have been produced by nuclear data processing codes such as ERRORJ or PUFF from the covariance data in evaluated nuclear data files. However in the existing nuclear data processing codes, an asymptotic neutron flux energy spectrum, not the exact one, has been applied to the multi-group covariance generation since the flux spectrum is unknown before the neutron transport calculation. It can cause an inconsistency between the sensitivity profiles and the covariance data of multi-group cross section especially in resolved resonance energy region, because the sensitivities we usually use are resonance self-shielded while the multi-group cross sections produced from an asymptotic flux spectrum are infinitely-diluted. In order to calculate the multi-group covariance estimation in the ongoing MC simulation, mathematical derivations for converting the double integration equation into a single one by utilizing sampling method have been introduced along with the procedure of multi-group covariance tally

Energy-balance check for continuous energy cross section library CENACE-1.0

International Nuclear Information System (INIS)

Zhao Qiujuan; Wu Haicheng; Ge Zhigang

2014-01-01

In order to verify the reliability of the multiple-temperature continuous energy cross section library CENACE-1.0 when used for calculating nuclear heating in reactor core, NJOY99/HEATR module and auxiliary code chkACEheat developed locally were used to perform energy-balance check for all materials in the library. The test results show that the pass rate of KERMA factors and heat production cross sections of the CENACE-1.0 library is better than that of the other ACE libraries used as comparison. However, unreasonable KERMA factors still exist in various evaluation libraries, and methods to directly revise the calculation results of KERMA factors need to be developed. (authors)

New evaluated neutron cross section libraries for the GEANT4 code

International Nuclear Information System (INIS)

Mendoza, E.; Cano-Ott, D.; Guerrero, C.; Capote, R.

2012-04-01

The so-called High Precision neutron physics model implemented in the GEANT4 simulation package allows simulating the transport of neutrons with energies up to 20 MeV. It relies on the G4NDL cross section libraries, prepared by the GEANT4 collaboration from evaluated cross section files and distributed freely together with the code. Even though the performance of the G4NDL library has been improved over the time, users running complex simulations which involve the transport of neutrons do need more flexibility, in particular when assessing the uncertainties in the simulation results due to the neutron (and hence the nuclear) data library used. For this reason, a software tool has been developed for transforming any evaluated neutron cross section library in the ENDF-6 format into the G4NDL format. Furthermore, eight different releases of ENDF-B, JEFF, JENDL, CENDL and BROND national libraries have been translated into the G4NDL format and are distributed by the IAEA nuclear data service at www-nds.iaea.org/geant4. In this way, GEANT4 users have access to the complete list of standard evaluated neutron data libraries when performing Monte Carlo simulations with GEANT4. Consistency checks and a first validation of the libraries have been made following the methods described in this report. (author)

A revision of photon interaction data in the UKAEA nuclear data library

International Nuclear Information System (INIS)

Knipe, A.D.

1975-10-01

Photon interaction data in the UKAEA Nuclear Data Library have been updated and extended to cover all elements up to Atomic Number 94. Cross-sections for the photoelectric effect, Compton scattering, pair-production, and the total cross-section, are stored at 40 energy points in the range 0.01 MeV to 20 MeV. The angular distribution for Compton scattering is also included in the library. This report describes the derivation and accuracy of the data, and tabulates the cross-sections and angular distribution in the appendices. The preparation of multigroup cross-sections from the library's data is also discussed. (author)

Comparative analysis of the neutron cross-sections of iron from various evaluated data libraries

International Nuclear Information System (INIS)

Bychkov, V.M.; Vozyakov, V.V.; Manokhin, V.N.; Smoll, F.; Resner, P.; Seeliger, D.; Hermsdorf, D.

1983-09-01

The comparative analysis of neutron cross-sections of iron from evaluated nuclear data libraries SOKRATOR, KEDAK, ENDL is done in energy interval from 0.025 eV to 20 MeV. Some of iron cross-sections from SOKRATOR library are revised and new data, which are obtained by using new experimental data and more comprehensive theoretical methods, are recommended. As a result the new version of the iron neutron cross-section file (BNF-2012) is produced for SOKRATOR library. (author)

SENSIT: a cross-section and design sensitivity and uncertainty analysis code

International Nuclear Information System (INIS)

Gerstl, S.A.W.

1980-01-01

SENSIT computes the sensitivity and uncertainty of a calculated integral response (such as a dose rate) due to input cross sections and their uncertainties. Sensitivity profiles are computed for neutron and gamma-ray reaction cross sections of standard multigroup cross section sets and for secondary energy distributions (SEDs) of multigroup scattering matrices. In the design sensitivity mode, SENSIT computes changes in an integral response due to design changes and gives the appropriate sensitivity coefficients. Cross section uncertainty analyses are performed for three types of input data uncertainties: cross-section covariance matrices for pairs of multigroup reaction cross sections, spectral shape uncertainty parameters for secondary energy distributions (integral SED uncertainties), and covariance matrices for energy-dependent response functions. For all three types of data uncertainties SENSIT computes the resulting variance and estimated standard deviation in an integral response of interest, on the basis of generalized perturbation theory. SENSIT attempts to be more comprehensive than earlier sensitivity analysis codes, such as SWANLAKE

ZZ CANDULIB-AECL, Burnup-Dependent ORIGEN-S Cross-Section Libraries for Candu Reactor Fuels

International Nuclear Information System (INIS)

2002-01-01

1 - Historical background and information: - 28-element fuel cross-section library: Format: Designed for use with the ORIGEN-S isotope generation and depletion code. Materials: Co, Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, Nb, Mo, Tc, Ru, Rh, Pd, Ag, Cd, In, Sn, Sb, Te, I, Xe, Cs, Ba, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy, Ho, Er, Lu, Ta, W, Re, Au, Th, Pa, U, Np, Pu, Am, Cm. Origin: ENDSF, ENDF/B-IV, -V and -VI Weighting spectrum: determined using WIMS-AECL transport code. - 37-element fuel cross-section library: Format: Designed for use with the ORIGEN-S isotope generation and depletion code. Materials: Co, Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, Nb, Mo, Tc, Ru, Rh, Pd, Ag, Cd, In, Sn, Sb, Te, I, Xe, Cs, Ba, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy, Ho, Er, Lu, Ta, W, Re, Au, Th, Pa, U, Np, Pu, Am, Cm. Origin: ENDSF, ENDF/B-IV, -V and -VI Weighting spectrum: determined using WIMS-AECL transport code. In 1995, updated ORIGEN-S cross-section libraries were created as part of a program to upgrade and standardize the computer codes and nuclear data employed for used fuel characterization. This effort was funded through collaboration between Atomic Energy of Canada Limited and the Canadian Nuclear Power Utilities, under the Candu Owners Group (COG). The updated cross sections were generated using the WIMS-AECL lattice code and ENDF/B-V and -VI based data to provide cross section consistency with reactor physics codes. 2 - Application of the data: The libraries in this data collection are designed for characterising used fuel from Candu pressurized heavy water reactors. Two libraries are provided: one for the standard 28-element fuel bundle design, the other for the 37-element fuel bundle design. The libraries were generated for typical reactor operating conditions. The libraries are designed for use with the ORIGEN-S isotope generation and depletion code. 3 - Source and scope of data: The Candu libraries are updated with cross sections from a variety of different sources. Capture

Up to date cross sections library for Thermos and Record codes

International Nuclear Information System (INIS)

Hernandez Lopez, H.

1993-01-01

Reactor cell analysis is the first step in determining reactor core behavior and is required in the reload licensing process. For best results, reactor cell analysis should be carried out with libraries of up to date, accurate cross sections produced with well described methods from standard evaluated nuclear data. At first step in this work were determined the library structure for RECORD and THERMOS and were prepared the cross sections libraries using the NJOY nuclear data processing system and the ENDF-B/IV evaluated nuclear data. These libraries were used by the codes and some samples were perform, the result show some differences against the results obtained using the previous libraries. By other hand the libraries contain various adjustments to correct for deficiencies in nuclear data or analytical methods. These adjustments doesn't have any documentation, although some of them were identified in this work. (Author). 25 refs, 78 figs, 55 tabs

Sensitivity of 238U resonance absorption to library multigroup structure as calculated by WIMS-AECL

International Nuclear Information System (INIS)

Laughton, P.J.; Donnelly, J.V.

1995-01-01

In simulations of the TRX-1 experimental lattice, WIMS-AECL overpredicts, relative to MCNP, resonance absorption in neutron-energy groups containing the three large, low-lying resonances of 238 U when a standard ENDF/B-V-based library is used. A total excess in these groups of 4.0 neutron captures by 238 U per thousand fission neutrons has been observed. Similar comparisons are made in this work for the MIT-4 experimental lattice and simplified CANDU lattice cells containing 37-element fuel, with and without heavy-water coolant. Eleven different 89-group cross-section libraries were constructed for WIMS-AECL from ENDF/B-V data: only the neutron-energy-group boundaries used in generating multigroup cross sections and the Goldstein-Cohen correction factors differ from one library to the next. The first library uses the original 89-group structure, and the other ten involve energy groups of varying widths centred on the three large, low-lying resonances of 238 U. For TRX-1, some reduction in total discrepancy in 238 U capture can be achieved by using a new structure, although the improvement is small. The discrepancies in 238 U capture are of the same order for the MIT-4 case as those observed for TRX-1 for both the original group structure and the ten new structures. The WIMS-AECL calculation of 238 U resonance absorption in the same ranges of energy for the simplified CANDU 37-element lattice are in better agreement with MCNP than they are for TRX-1 and MIT-4: when the original structure is used, WIMS-AECL underpredicts total capture rate by 238 U in the energy range of interest by only 0.56 per thousand fission neutrons (coolant present) and 0.88 per thousand fission neutrons (voided coolant channel). The discrepancies are reduced when some of the new structures are used. For almost all of the cases considered here-TRX-1, MIT-4 and CANDU with coolant-better group-by-group agreement of 238 U capture around the 6.67-eV resonance is achieved by using a new library

Cross sections of the lumped fission products for the AMZ library

International Nuclear Information System (INIS)

Ono, S.; Corcueca, R.P.; Nascimento, J.A.

1985-01-01

The preparation of the lumped fission product cross section for the AMZ library is described. For this purpose 100 nuclides were selected. The cross sections for each nuclide were generated by the NJOY code with evaluated nuclear data from ENDF/B-V, complemented with ENDF/B-IV data. A comparison is performed between the data obtained and the lumped fission product cross section of JFS-II [pt

Library of neutron cross sections of the Thermos code

International Nuclear Information System (INIS)

Alonso V, G.; Hernandez L, H.

1991-10-01

The present work is the complement of the IT.SN/DFR-017 report in which the structure and the generation of the library of the Thermos code is described. In this report the comparison among the values of the cross sections that has the current library of the Thermos code and those generated by means of the ENDF-B/NJOY it is shown. (Author)

Implementation of the rapid cross section adjustment approach at General Electric

International Nuclear Information System (INIS)

Cowan, C.L.; Kujawski, E.; Protsik, R.

1978-01-01

The General Electric rapid cross section adjustment approach was developed to use the shielding factor method for formulating multigroup cross sections. In this approach, space- and composition-dependent cross sections for a particular reactor or shield design are prepared from a generalized cross section library by the use of resonance self-shielding factors, and by the adjustment of elastic scattering cross sections for the local neutron flux spectra. The principal tool in the cross section adjustment package is the data processing code TDOWN. This code was specified to give the user a high degree of flexibility in the analysis of advanced reactor designs. Of particular interest in the analysis of critical experiments is the ability to carry out cell heterogeneity self-shielding calculations using a multiregion equivalence relationship, and the homogenization of the cross sections over the specified cell with the flux weighting obtained from transport theory calculations. Extensive testing of the rapid cross section adjustment approach, including comparisons with Monte Carlo methods, indicated that this approach can be utilized with a high degree of confidence in the design analysis of complex fast reactor systems. 2 figures, 1 table

C4P cross-section libraries for safety analyses with SIMMER and related studies

International Nuclear Information System (INIS)

Rineiski, A.; Sinitsa, V.; Gabrielli, F.; Maschek, W.

2011-01-01

A code and data system, C 4 P, is under development at KIT. It includes fine-group master libraries and tools for generating problem-oriented cross-section libraries, primarily for safety studies with the SIMMER code and related analyses. In the paper, the 560-group master library and problem oriented 40-group and 72-group cross-section libraries, for thermal and fast systems, respectively, are described and their performances are investigated. (author)

Improvement of decay and cross-section data libraries for activation calculations

International Nuclear Information System (INIS)

Attaya, H.

1993-01-01

A new decay data library has been completed. The new library contains up-to-date decay information (half-lives, branching ratios, decay energies, γ's energies and intensities). Activation responses such as the air and water biological hazard potentials, the waste disposal rating, and the biological dose are also included in this library. Recently developed cross-section libraries have been acquired to be used together with the decay data library

Mixed Legendre moments and discrete scattering cross sections for anisotropy representation

International Nuclear Information System (INIS)

Calloo, A.; Vidal, J. F.; Le Tellier, R.; Rimpault, G.

2012-01-01

This paper deals with the resolution of the integro-differential form of the Boltzmann transport equation for neutron transport in nuclear reactors. In multigroup theory, deterministic codes use transfer cross sections which are expanded on Legendre polynomials. This modelling leads to negative values of the transfer cross section for certain scattering angles, and hence, the multigroup scattering source term is wrongly computed. The first part compares the convergence of 'Legendre-expanded' cross sections with respect to the order used with the method of characteristics (MOC) for Pressurised Water Reactor (PWR) type cells. Furthermore, the cross section is developed using piecewise-constant functions, which better models the multigroup transfer cross section and prevents the occurrence of any negative value for it. The second part focuses on the method of solving the transport equation with the above-mentioned piecewise-constant cross sections for lattice calculations for PWR cells. This expansion thereby constitutes a 'reference' method to compare the conventional Legendre expansion to, and to determine its pertinence when applied to reactor physics calculations. (authors)

COMBINE7.0 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

Energy Technology Data Exchange (ETDEWEB)

Woo Y. Yoon; David W. Nigg

2008-09-01

COMBINE7.0 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.0 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 finegroup cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko selfshielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those selfshielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.0 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a onedimensional, discrete

COMBINE7.0 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

International Nuclear Information System (INIS)

Yoon, Woo Y.; Nigg, David W.

2008-01-01

COMBINE7.0 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.0 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 finegroup cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko selfshielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those selfshielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.0 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a onedimensional, discrete

Description of the ENDF-NJOY system for the generation of cross sections libraries

International Nuclear Information System (INIS)

Alonso V, G.

1991-01-01

The physics of nuclear reactors requires of a great number of data to be able to evaluate the different phenomena that happen in a nuclear reactor; these data are mainly the microscopic neutron cross sections, but it is also required of data of radioactive decay and of nuclear structure for a great number of materials as well as of the cross sections of the photons and the production of these for the neutron interaction. These data group in nuclear databases, being the main ones: ENDF Nuclear Evaluated File, ENDL Dates Nuclear Evaluated Library it Dates (of the Laboratory Lawrence Livermore). JENDL Japanese Nuclear Evaluated Library Dates. Soviet SOKRATOR Nuclear Evaluated KEDAF Nuclear Karlsruhe File Dates. JEF Join Evaluated File (coordinated by NEA Data Bank). The existent codes that execute neutron and photon calculations require libraries of data that are very different some of other and of the databases. Of here that it is required of a series of processing codes that use the database like enter and its generate a secondary library of cross sections, which is read as enter for a code of spectra generation. Generally average cross sections by group are obtained; this library is that it is used in the codes that execute neutron calculations. (Author)

FENDL2/A-MCNP, FENDL2/A-VITJE and FENDL2/A-VITJFLAT. The processed FENDL-2 neutron activation cross-section data files. Summary documentation

International Nuclear Information System (INIS)

Pashchenko, A.B.; Wienke, H.

1997-01-01

This document summarizes the libraries of neutron activation cross-section data processed into the following three formats: continuous energy format as used by the Monte Carlo neutron/photon transport code MCNP4A; VITAMIN-J 175 multigroup format weighted with the VITAMIN-E weighting spectrum as used by the transmutation codes REAC*2/3 and FOUR ACES; VITAMIN-J 175 multigroup ENDF-6 format, with a flat weighting spectrum. The data are available from the IAEA Nuclear Data Section online via INTERNET by FTP command, or on magnetic tape. (author)

MCFT: a program for calculating fast and thermal neutron multigroup constants

International Nuclear Information System (INIS)

Yang Shunhai; Sang Xinzeng

1993-01-01

MCFT is a program for calculating the fast and thermal neutron multigroup constants, which is redesigned from some codes for generation of thermal neutron multigroup constants and for fast neutron multigroup constants adapted on CYBER 825 computer. It uses indifferently as basic input with the evaluated nuclear data contained in the ENDF/B (US), KEDAK (Germany) and UK (United Kingdom) libraries. The code includes a section devoted to the generation of resonant Doppler broadened cross section in the framework of single-or multi-level Breit-Wigner formalism. The program can compute the thermal neutron scattering law S (α, β, T) as the input data in tabular, free gas or diffusion motion form. It can treat up to 200 energy groups and Legendre moments up to P 5 . The output consists of various reaction multigroup constants in all neutron energy range desired in the nuclear reactor design and calculation. Three options in input file can be used by the user. The output format is arbitrary and defined by user with a minimum of program modification. The program includes about 15,000 cards and 184 subroutines. FORTRAN 5 computer language is used. The operation system is under NOS 2 on computer CYBER 825

FENDL/E-2.0. Evaluated nuclear data library of neutron-nucleus interaction cross sections and photon production cross sections and photon-atom interaction cross sections for fusion applications. Version 1, March 1997. Summary documentation

International Nuclear Information System (INIS)

Pashchenko, A.B.; Wienke, H.

1998-01-01

This document presents the description of a physical tape containing the basic evaluated nuclear data library of neutron-nucleus interaction cross sections, photon production cross sections and photon-atom interaction cross sections for fusion applications. It is part of the evaluated nuclear data library for fusion applications FENDL-2. The data are available cost-free from the Nuclear Data Section upon request. The data can also be retrieved by the user via online access through international computer networks. (author)

FENDL/E. Evaluated nuclear data library of neutron nuclear interaction cross-sections and photon production cross-sections and photon-atom interaction cross sections for fusion applications. Version 1.1 of November 1994

International Nuclear Information System (INIS)

Pashchenko, A.B.; Wienke, H.; Ganesan, S.; McLaughlin, P.K.

1996-01-01

This document presents the description of a physical tape containing the basic evaluated nuclear data library of neutron nuclear interaction cross-sections and photon production cross-sections and photon-atom interaction cross-sections for fusion applications. It is part of FENDL, the evaluated nuclear data library for fusion applications. The nuclear data are available cost-free for distribution to interested scientists upon request. The data can also be retrieved by the user via online access through international computer networks. (author). 11 refs, 1 tab

Development of the CPXSD Methodology for Generation of Fine-Group Libraries for Shielding Applications

International Nuclear Information System (INIS)

Alpan, F. Arzu; Haghighat, Alireza

2005-01-01

Multigroup cross sections are one of the major factors that cause uncertainties in the results of deterministic transport calculations. Thus, it is important to prepare effective cross-section libraries that include an appropriate group structure and are based on an appropriate spectrum. There are several multigroup cross-section libraries available for particular applications. For example, the 47-neutron, 20-gamma group BUGLE library that is derived from the 199-neutron, 42-gamma group VITAMIN-B6 library is widely used for light water reactor (LWR) shielding and pressure vessel dosimetry applications. However, there is no publicly available methodology that can construct problem-dependent libraries. Thus, the authors have developed the Contributon and Point-wise Cross Section Driven (CPXSD) methodology for constructing effective fine- and broad-group structures. In this paper, new fine-group structures were constructed using the CPXSD, and new fine-group cross-section libraries were generated. The 450-group LIB450 and 589-group LIB589 libraries were developed for neutrons sensitive to the fast and thermal energy ranges, respectively, for LWR shielding problems. As compared to a VITAMIN-B6-like library, the new fine-group library developed for fast neutron dosimetry calculations resulted in closer agreement to the continuous-energy predictions. For example, for the fast neutron cavity dosimetry, ∼4% improvement was observed for the 237 Np(n,f) reaction rate. For the thermal neutron 1 H(n, γ) reaction, a maximum improvement of ∼14% was observed in the reaction rate at the middowncomer position

Analysis of fusion neutronics calculations and appraisal of UW cross-section library

International Nuclear Information System (INIS)

Xie Jianping; Li Xingzhong; Ying Chuntong

1989-01-01

A series of calculations for different cases (especially for the values of tritium breeding ratio T, and the fuel breeding ratio F in the blanket of a hybrid reactor) were carried out by using ANISN program and UW cross-section library. The comparison with other results in China and abroad kalso was done. It was shownwn that the installation and execution of ANISN program on ELXSI machine at Tsinghua University are successful, and the UW cross-section library is reliable. It may be used for fusion neutronics calculation in the future. The paper also points out that the difference between the calculations and by the authors are due to jthe different in cross-section data used

ZZ DOSCROS, Neutron Cross-Section Library for Spectra Unfolding and Integral Parameter Evaluation

International Nuclear Information System (INIS)

Zijp, Willem L.; Nolthenius, Henk J.; Rieffe, Henk Ch.

1987-01-01

1 - Description of problem or function: Format: SAND-II; Number of groups: 640 fine group cross section values; Nuclides: Li, B, F, Na, Mg, Al, S, Sc, Ti, Cr, Mn, Fe, Co, Ni, Cu, Zn, As, Br, Nb, Mo, Rh, Pd, Ag, In, Sb, I, Cs, La, Eu, Sm, Dy, Lu, Ta, W, Re, Au, Th, U, Np, Pu. Origin: ENDF/B-V mainly, ENDF/B-IV, INDL/V. This library forms in combination with the DAMSIG81 library a convenient source of evaluated energy dependent cross section sets which may be used in the determination of neutron spectra by means of adjustment (or unfolding) procedures or which can be used for the determination of integral parameters (such as damage-to-activation ratio) useful in characterising the neutron spectra. The energy dependent fine group cross section data are presented in a 640 group structure of the SAND-II type. This group structure has 45 energy groups per energy decade below 1 MeV and a group width of 100 KeV above 1 MeV. The total energy span of this group structure is from 10 -10 MeV to 20 MeV. The library has the SAND-II format, which implies that a special part of the library has to contain cover cross section data sets. These cross section data sets are required in the SAND-II program for taking into account the influence of special detector surroundings which may be used during an irradiation. 2 - Method of solution: The selection of the reactions from the evaluated nuclear data libraries was determined by various properties of the reactions for neutron metrology. For this reason all the well- known reactions of the ENDF/B-V dosimetry file are included but these data are supplemented with cross section sets for less well known metrology reactions which may become of interest

TOPICS-B, Neutron and Gamma Cross-Sections Library Handling in FIDO Format

International Nuclear Information System (INIS)

Wasastjerna, Frej

2003-01-01

1 - Description of program or function: The program is intended to manipulate working format neutron and/or gamma cross section libraries, carrying out such operations as mixing materials, deleting unneeded groups, inserting response cross sections or whatever the user may require. It has been designed to make it easy to include new modules to cope with new requirements. The cross section libraries involved should preferably be in ANISN format, but if they are not, this too can be handled by adding new modules as needed. This program is intended to supersede TOPICS (NEA-1406). TOPICS was intended for interactive use, but experience has shown that using it is somewhat difficult. Therefore it was redesigned for batch use (the input is written to a file and the program is then run using that file, instead of reading input directly from the keyboard). 2 - Method of solution: Each required operation is performed by a separate module (a set of subprograms). 3 - Restrictions on the complexity of the problem: Essentially none, variable dimensioning is used. However, TOPICS-B is not intended to be applied to basic nuclear data libraries (such as the ENDF/B series) or to flexible format libraries (e.g., the VITAMIN series). It is intended only for working format libraries like the BUGLE series

Role of ''standard'' fine-group cross section libraries in shielding analysis

International Nuclear Information System (INIS)

Weisbin, C.R.; Roussin, R.W.; Oblow, E.M.; Cullen, D.E.; White, J.E.; Wright, R.Q.

1977-01-01

The Divisions of Magnetic Fusion Energy (DMFE) and Reactor Development and Demonstration (DRDD) of the United States Energy Research and Development Administration (ERDA) have jointly sponsored the development of a 171 neutron, 36 gamma ray group pseudo composition independent cross section library based upon ENDF/B-IV. This library (named VITAMIN-C and packaged by RSIC as DLC-41) is intended to be generally applicable to fusion blanket and LMFBR core and shield analysis. The purpose of this paper is to evaluate this library as a possible candidate for specific designation as a ''standard'' in light of American Nuclear Society standards for fine-group cross section data sets. The rationale and qualification procedure for such a standard are discussed. Finally, current limitations and anticipated extensions to this processed data file are described

Nuclear Data Uncertainty Propagation in Depletion Calculations Using Cross Section Uncertainties in One-group or Multi-group

Energy Technology Data Exchange (ETDEWEB)

Díez, C.J., E-mail: cj.diez@upm.es [Dpto. de Ingeníera Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain); Cabellos, O. [Dpto. de Ingeníera Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain); Instituto de Fusión Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain); Martínez, J.S. [Dpto. de Ingeníera Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain)

2015-01-15

Several approaches have been developed in last decades to tackle nuclear data uncertainty propagation problems of burn-up calculations. One approach proposed was the Hybrid Method, where uncertainties in nuclear data are propagated only on the depletion part of a burn-up problem. Because only depletion is addressed, only one-group cross sections are necessary, and hence, their collapsed one-group uncertainties. This approach has been applied successfully in several advanced reactor systems like EFIT (ADS-like reactor) or ESFR (Sodium fast reactor) to assess uncertainties on the isotopic composition. However, a comparison with using multi-group energy structures was not carried out, and has to be performed in order to analyse the limitations of using one-group uncertainties.

Nuclear Data Uncertainty Propagation in Depletion Calculations Using Cross Section Uncertainties in One-group or Multi-group

International Nuclear Information System (INIS)

Díez, C.J.; Cabellos, O.; Martínez, J.S.

2015-01-01

Several approaches have been developed in last decades to tackle nuclear data uncertainty propagation problems of burn-up calculations. One approach proposed was the Hybrid Method, where uncertainties in nuclear data are propagated only on the depletion part of a burn-up problem. Because only depletion is addressed, only one-group cross sections are necessary, and hence, their collapsed one-group uncertainties. This approach has been applied successfully in several advanced reactor systems like EFIT (ADS-like reactor) or ESFR (Sodium fast reactor) to assess uncertainties on the isotopic composition. However, a comparison with using multi-group energy structures was not carried out, and has to be performed in order to analyse the limitations of using one-group uncertainties

Nuclear Data Uncertainty Propagation in Depletion Calculations Using Cross Section Uncertainties in One-group or Multi-group

Science.gov (United States)

Díez, C. J.; Cabellos, O.; Martínez, J. S.

2015-01-01

Several approaches have been developed in last decades to tackle nuclear data uncertainty propagation problems of burn-up calculations. One approach proposed was the Hybrid Method, where uncertainties in nuclear data are propagated only on the depletion part of a burn-up problem. Because only depletion is addressed, only one-group cross sections are necessary, and hence, their collapsed one-group uncertainties. This approach has been applied successfully in several advanced reactor systems like EFIT (ADS-like reactor) or ESFR (Sodium fast reactor) to assess uncertainties on the isotopic composition. However, a comparison with using multi-group energy structures was not carried out, and has to be performed in order to analyse the limitations of using one-group uncertainties.

Requests on domestic nuclear data library from BWR design

International Nuclear Information System (INIS)

Maruyama, Hiromi

2003-01-01

Requests on the domestic nuclear data library JENDL and activities of the Nuclear Data Center have been presented from the perspective of BWR design and design code development. The requests include a standard multi-group cross section library, technical supports, and clarification of advantage of JENDL as well as requests from physical aspects. (author)

Neutron cross section library production code system for continuous energy Monte Carlo code MVP. LICEM

International Nuclear Information System (INIS)

Mori, Takamasa; Nakagawa, Masayuki; Kaneko, Kunio.

1996-05-01

A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author)

Neutron cross section library production code system for continuous energy Monte Carlo code MVP. LICEM

Energy Technology Data Exchange (ETDEWEB)

Mori, Takamasa; Nakagawa, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio

1996-05-01

A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author).

Point 2004 A Temperature Dependent ENDF/B-VI, Release 8 Cross Section Library

International Nuclear Information System (INIS)

Cullen, D E

2004-01-01

The ENDF/B data library has recently been updated and is now freely available through the National Nuclear Data Center (NNDC), Brookhaven National Laboratory. This most recent library is identified as ENDF/B-VI, Release 8. Release 8 completely supersedes all preceding releases. Release 8 will be the last release of ENDF/B-VI; the next release of ENDF/B data will be for the new ENDF/B-VII library. As distributed the ENDF/B-VI, Release 8 data includes cross sections represented in the form of a combination of resonance parameters and/or tabulated energy dependent cross sections, nominally at 0 Kelvin temperature. For use in applications this library has been processed into the form of temperature dependent cross sections at eight neutron reactor like temperatures, between 0 and 2100 Kelvin, in steps of 300 Kelvin. It has also been processed to five astrophysics like temperatures, 1, 10, 100 eV, 1 and 10 keV. For reference purposes, 300 Kelvin is approximately 1/40 eV, so that 1 eV is approximately 12,000 Kelvin. At each temperature the cross sections are tabulated and linearly interpolable in energy. All results are in the computer independent ENDF/B-VI character format [1], which allows the data to be easily transported between computers. In its processed form this library is approximately 4.3 gigabyte in size and is distributed on a single DVD

Status of recent fast capture cross section evaluations for important fission product nuclides

International Nuclear Information System (INIS)

Gruppelaar, H.

1982-01-01

A comparison is made between recent evaluations of fission-product cross sections as given in the CNEN/CEA, ENDF/B-IV, ENDF/V-V, JENDL-1, RCN-2 and RCN-3 data libraries. The intercomparison is restricted to 24 important fission products in a fast power reactor. The evaluation methods used to obtain the various data files are reviewed and possible shortcomings are indicated. A survey is given of the experimental data based used in the various evaluations. Some graphs are included showing the new ENDF/B-V and RCN-3 fastcapture cross-section evaluations. Further intercomparisons are made by means of multi-group and one-group cross sections. It is shown that lumped fission-product cross sections calculated from the most recent versions of the data files are in quite good agreement with each other. This review concludes with a discussion on observed discrepancies and requests for new measurements. 78 references

ORACLE: an adjusted cross-section and covariance library for fast-reactor analysis

International Nuclear Information System (INIS)

Yeivin, Y.; Marable, J.H.; Weisbin, C.R.; Wagschal, J.J.

1980-01-01

Benchmark integral-experiment values from six fast critical-reactor assemblies and two standard neutron fields are combined with corresponding calculations using group cross sections based on ENDF/B-V in a least-squares data adjustment using evaluated covariances from ENDF/B-V and supporting covariance evaluations. Purpose is to produce an adjusted cross-section and covariance library which is based on well-documented data and methods and which is suitable for fast-reactor design. By use of such a library, data- and methods-related biases of calculated performance parameters should be reduced and uncertainties of the calculated values minimized. Consistency of the extensive data base is analyzed using the chi-square test. This adjusted library ORACLE will be available shortly

AXMIX, ANISN Cross-Sections Mixing, Transport Corrections, Data Library Management

International Nuclear Information System (INIS)

2002-01-01

1 - Nature of physical problem solved: Mixing, changing table length, adjoining, making scattering order adjustments (PN delta function subtraction), and transport corrections of ANISN-type cross sections, and management of cross section data sets and libraries. 2 - Method of solution: The number of energy groups which will fit into the core allocated is determined first. If all groups will fit, the solution is straightforward. If not, then the maximum number of groups which will fit is processed repeatedly using direct access I/O and storage disks. 3 - Restrictions on the complexity of the problem: Some flexibility in applying AXMIX is lost when cross sections to be processed contain up-scatter. A special section on up-scatter is therefore included in the report

ZZ DLC-11 RITTS, 121-Group Coupled Cross-Section for ANISN, DOT, MORSE

International Nuclear Information System (INIS)

1970-01-01

A - Nature of physical problem solved: Format: ANISN, DTF-4, DOT and MORSE. Number of groups: 100 neutron energy groups (14.92 MeV to thermal) 21 gamma-ray energy groups (14.0 to 0.01 MeV) Nuclides: H, C, O, N, Na, Mg, P, S, Cl, K, and Ca, (microscopic cross sections) and 9 organic materials including 11-element standard man, 4-element standard man, skin, bone, tissue, brain, lung, red marrow, and muscle (macroscopic cross sections). Origin: ENDF/B for H, C, N, O, Na, and Mg; O5R library for Ca, S, and K; GAM-2 library for Cl; Evaluation by J.J. Ritts for P. Weighting spectrum: 1/E for the top 99 groups and Maxwellian for the thermal group values. DLC-11 data is suitable for neutron, gamma-ray, or coupled neutron and gamma-ray transport calculations. It is intended for use in multigroup discrete ordinates or Monte Carlo transport codes which treat anisotropic scattering by Legendre expansion up to order P3. DLC-11 is a collection of multigroup cross section data which were compiled by J. J. Ritts for use in depth-dose calculations in anthropomorphic phantoms. For convenience the data are grouped as follows - 1. A coupled 121-group (100 neutron, 21 gamma-ray) set of data for the 11 elements H, C, O, N. Na, Mg, P, S, Cl, K, and Ca. This set includes P3 coupled 121-group microscopic cross sections plus 121-group kerma factors for the 11 elements. 2. A 100-group set of neutron cross sections for the 11 elements. 3. A coupled 121-group set of macroscopic cross sections for 9 organic materials including 11-element standard man, 4-element standard man, skin, bone, tissue, brain, lung, red marrow, and muscle. B - Method of solution: The basic data sources were ENDF/B for H, C, N, O, Na, and Mg, the O5R library for Ca, S, and K, the GAM-2 library for Cl and an evaluation by Ritts for P. A 1/E spectrum was assumed for averaging the top 99 groups and a Maxwellian for averaging the thermal group values. The gamma-ray cross sections were computed from DLC-3/HPIC using MUG. The

Preliminary assessment of Geant4 HP models and cross section libraries by reactor criticality benchmark calculations

DEFF Research Database (Denmark)

Cai, Xiao-Xiao; Llamas-Jansa, Isabel; Mullet, Steven

2013-01-01

Geant4 is an open source general purpose simulation toolkit for particle transportation in matter. Since the extension of the thermal scattering model in Geant4.9.5 and the availability of the IAEA HP model cross section libraries, it is now possible to extend the application area of Geant4......, U and O in uranium dioxide, Al metal, Be metal, and Fe metal. The native HP cross section library G4NDL does not include data for elements with atomic number larger than 92. Therefore, transuranic elements, which have impacts for a realistic reactor, can not be simulated by the combination of the HP...... models and the G4NDL library. However, cross sections of those missing isotopes were made available recently through the IAEA project “new evaluated neutron cross section libraries for Geant4”....

Performance of JEF2.2 based continuous energy cross sections in predicting the multiplication factor of critical systems

International Nuclear Information System (INIS)

John, T.M.; de Leege, P.F.A.; Hoogenboom, J.E.

1996-01-01

The continuous energy representation of cross sections for neutronics calculations avoids the requirement of resonance self shielding and the assumptions about the neutron spectrum used for weighing cross sections, required in the preparation of a multigroup cross sections library. The cross sections library prepared for a particular temperature of the nuclide is valid irrespective of the environment of the nuclide and can be used in calculations for many types of reactors. It is comparatively easier to incorporate them in Monte Carlo simulation of neutron transport. The Monte Carlo code MCNP is capable of using a continuous energy representation of nuclear cross sections in simulation of neutron or photon transport. The ACER module of NJOY is able to generate the continuous energy cross section of any nuclide in a format that can be used by MCNP, from any evaluated data file in ENDF/B format. Continuous energy cross sections prepared from the evaluated data file JEF2.2 was used to analyse some standard critical benchmarks and also the critical configuration of the HOR, a 2 MW research reactor at Delft, the Netherlands. Results show that continuous energy cross sections prepared from JEF2.2 evaluated file predicts the multiplication factor of critical systems very close to unity. (author). 6 refs., 2 tabs., 1 fig

COMBINE7.1 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

Energy Technology Data Exchange (ETDEWEB)

Woo Y. Yoon; David W. Nigg

2009-08-01

COMBINE7.1 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.1 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 fine-group cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko self-shielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those self-shielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.1 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a one-dimensional, discrete

COMBINE7.1 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

International Nuclear Information System (INIS)

Yoon, Woo Y.; Nigg, David W.

2009-01-01

COMBINE7.1 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.1 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 fine-group cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko self-shielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those self-shielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.1 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a one-dimensional, discrete

Generation of multigroup cross sections from ENDF/B-IV nuclear data library

International Nuclear Information System (INIS)

Chapot, J.L.C.; Thome Filho, Z.D.

1980-04-01

The generation of nuclear data compacted in energy groups is made. The nuclear data library ENDF/B-IV, Evaluated Nuclear Data File, and the new version of the codes ETOG-3 and ETOT-3 are utilized. The data obtained are compared with data from other sources. (L.F.) [pt

Library of neutron reaction cross-sections in the ABBN-93 constant system

International Nuclear Information System (INIS)

Zabrodskaya, S.V.; Korchagina, Zh.A.; Koshcheev, V.N.; Nikolaev, M.N.; Tsibulya, A.M.

2001-01-01

The library of neutron reaction group cross-sections in the ABBN-93 constant set is described. The format used for data representation, the content and purpose of the sub-libraries and their practical application in the SCALE criticality safety estimation system are discussed. (author)

ZZ AIRFEWG, Gamma, Neutron Transport Calculation in Air Using FEWG1 Cross-Section

International Nuclear Information System (INIS)

1985-01-01

1 - Description of program or function: Format: ANISN; Number of groups: 37 neutron / 21 gamma-ray; Nuclides: air (79% N and 21% O); Origin: DLC-0031/FEWG1 cross sections (ENDF/B-IV). Weighting spectrum: 1/E. The AIRFEWG library has been generated by an ANISN multigroup calculation of gamma-ray, neutron, and secondary gamma-ray transport in infinite homogeneous air using DLC-0031/FEWG1 cross sections. 2 - Method of solution: The results were generated with a P3, ANISN run with a source in a single energy group. Thus, 58 such runs were required. For sources in the 37 neutron groups, both neutron and secondary gamma-ray fluence results were calculated. For gamma-ray sources only gamma-ray fluences were calculated

EJ2-MCNPlib. Contents of the JEF-2.2 based neutron cross-section library for MCNP4A

International Nuclear Information System (INIS)

Hogenbirk, A.; Oppe, J.

1995-05-01

In this report a description is given of the EJ2-MCNPlib library. The EJ2-MCNPlib library is to be used for reactivity/critically calculations and general neutron/photon transport calculations with the Monte Carlo code MCNP4A. The library is based on the European JEF-2.2 nuclear data evaluation and contains data for all (i.e. 313) nuclides available on this evaluation.The cross-section data were generated using the NJOY cross-section processing code system, version 91.118. For easy reference cross-section plots are given in this report for the total, elastic and absorption cross sections for all nuclides on the EJ2-MCNPlib library. Furthermore, for verification purposes a graphical intercomparison is given of the results of standard benchmark calculations performed with JEF-2.2 cross-section data and with ENDF/B-V cross-section data (whenever available). 6 refs

ARP: A PC-compatible scheme for generating ORIGEN-S cross section library

International Nuclear Information System (INIS)

Leal, L.C.; Hermann, O.W.; Parks, C.V.

1995-01-01

The SAS2H sequence of the SCALE code system has been widely used for treating problems related to the characterization of nuclear systems for disposal, storage, and shipment. The calculations, in general, consist of determining the isotope compositions of the different materials present in the problem as a function of time, which subsequently enable determination of the heat generation and radiation source terms. In the SAS2H scheme, time-dependent material concentrations are obtained using the ORIGEN-S code based on a point-depletion calculation that utilizes problem-dependent cross-section libraries generated by distinct codes of the SAS2H sequence. In this paper we will be concerned with the methodology utilized in the SAS2H control module to create cross-section libraries for point-depletion calculations with the ORIGEN-S code. A brief description of the SAS2H scheme will be given, and a new capability, the automatic rapid processing (ARP), for generating problem-dependent ORIGEN-S cross-section libraries will be presented. Use of ARP can enable execution of ORIGEN-S on a personal computer with identical accuracy to that obtained with SAS2H

MICROX-2 cross section library based on ENDF/B-VII

International Nuclear Information System (INIS)

Hou, J.; Ivanov, K.; Choi, H.

2012-01-01

New cross section libraries of a neutron transport code MICROX-2 have been generated for advanced reactor design and fuel cycle analyses. A total of 386 nuclides were processed, including 10 thermal scattering nuclides, which are available in ENDF/B-VII release 0 nuclear data. The NJOY system and MICROR code were used to process nuclear data and convert them into MICROX-2 format. The energy group structure of the new library was optimized for both the thermal and fast neutron spectrum reactors based on Contributon and Point-wise Cross Section Driven (CPXSD) method, resulting in a total of 1173 energy groups. A series of lattice cell level benchmark calculations have been performed against both experimental measurements and Monte Carlo calculations for the effective/infinite multiplication factor and reaction rate ratios. The results of MICROX-2 calculation with the new library were consistent with those of 15 reference cases. The average errors of the infinite multiplication factor and reaction rate ratio were 0.31% δk and 1.9%, respectively. The maximum error of reaction rate ratio was 8% for 238 U-to- 235 U fission of ZEBRA lattice against the reference calculation done by MCNP5. (authors)

How to Use Benchmark and Cross-section Studies to Improve Data Libraries and Models

Science.gov (United States)

Wagner, V.; Suchopár, M.; Vrzalová, J.; Chudoba, P.; Svoboda, O.; Tichý, P.; Krása, A.; Majerle, M.; Kugler, A.; Adam, J.; Baldin, A.; Furman, W.; Kadykov, M.; Solnyshkin, A.; Tsoupko-Sitnikov, S.; Tyutyunikov, S.; Vladimirovna, N.; Závorka, L.

2016-06-01

Improvements of the Monte Carlo transport codes and cross-section libraries are very important steps towards usage of the accelerator-driven transmutation systems. We have conducted a lot of benchmark experiments with different set-ups consisting of lead, natural uranium and moderator irradiated by relativistic protons and deuterons within framework of the collaboration “Energy and Transmutation of Radioactive Waste”. Unfortunately, the knowledge of the total or partial cross-sections of important reactions is insufficient. Due to this reason we have started extensive studies of different reaction cross-sections. We measure cross-sections of important neutron reactions by means of the quasi-monoenergetic neutron sources based on the cyclotrons at Nuclear Physics Institute in Řež and at The Svedberg Laboratory in Uppsala. Measurements of partial cross-sections of relativistic deuteron reactions were the second direction of our studies. The new results obtained during last years will be shown. Possible use of these data for improvement of libraries, models and benchmark studies will be discussed.

Effects of space-dependent cross sections on core physics parameters for compact fast spectrum space power reactors

International Nuclear Information System (INIS)

Lell, R.M.; Hanan, N.A.

1987-01-01

Effects of multigroup neutron cross section generation procedures on core physics parameters for compact fast spectrum reactors have been examined. Homogeneous and space-dependent multigroup cross section sets were generated in 11 and 27 groups for a representative fast reactor core. These cross sections were used to compute various reactor physics parameters for the reference core. Coarse group structure and neglect of space-dependence in the generation procedure resulted in inaccurate computations of reactor flux and power distributions and in significant errors regarding estimates of core reactivity and control system worth. Delayed neutron fraction was insensitive to cross section treatment, and computed reactivity coefficients were only slightly sensitive. However, neutron lifetime was found to be very sensitive to cross section treatment. Deficiencies in multigroup cross sections are reflected in core nuclear design and, consequently, in system mechanical design

Generation of neutron cross sections library for the Thermos code of the Fuel management System (FMS)

International Nuclear Information System (INIS)

Alonso V, G.; Viais J, J.

1990-10-01

There is developed a method to generate the library of neutron cross sections for the Thermos code by means of the database ENDF-B/IV and the NJOY code. The obtained results are compared with the version previous of the library of neutron cross sections which was processed using the version ENDF-B/III. (Author)

A punched-card library of neutron cross-sections and its use in the mechanized preparation of group cross-sections for use in Monte Carlo, Carlson S{sub n} and other multi-group neutronics calculations on high-speed computers

Energy Technology Data Exchange (ETDEWEB)

Parker, K [Atomic Weapons Research Establishment, Aldermaston (United Kingdom)

1962-03-15

The AWRE punched-card library of neutron cross-sections is described together with associated IBM-7090 programmes which process this data to give group-averaged cross-sections for use in Monte Carlo, Carlson S{sub n} and other multi-group neutronics calculations. The methods developed to deal with both isotropic and anisotropic elastic scattering are described. These include the multi-group transport approximation and the full treatment of anisotropic scattering using the Legendre polynomial moments of the scattering transfer matrix. The principles of group-constant formation are considered and illustrated by describing systems of group constants suitable for fast-reactor calculations. Practical problems such as the empirical adjustment of group constants to reproduce integral results and the collapsing of a many-group set of constants to give a few-group set are discussed. (author) [French] L'auteur decrit le fichier de cartes perforees sur lesquelles on enregistre a l'Atomic Weapons Research Establishment (AWRE) les sections efficaces neutroniques ainsi que les programmes IBM-7090 associes qui sont employes pour le traitement de ces informations, en vue d'obtenir des sections efficaces moyennes par groupe pouvant servir aux calculs de neutroniques a plusieurs groupes, effectues a l'aide des methodes de Monte-Carlo, S{sub n} de Carlson et autres methodes. L'auteur expose ensuite les methodes mises au point roda etudier la diffusion elastique, tant isotrope qu'anisotrope. Elles comprennent l'approximation de transport a plusieurs groupes, ainsi que le traitement complet de la diffusion anisotrope par les moments polynomiaux de Legendre de la matrice de transfert de la diffusion. L'auteur examine les principes de la formation des constantes de groupes; a titre d'illustration, il decrit les systemes de constantes de groupes qui se pretent aux calculs de reacteurs a neutrons rapides. Il expose quelques problemes pratiques, tels que l'ajustement empirique des

Nuclear libraries for SCALE5.1 system

International Nuclear Information System (INIS)

Vertes, P.

2009-01-01

Codes for preparing master and working AMPX libraries and point-wise nuclear libraries for SCALE5.1 system have been created. Master and working libraries are constructed from multigroup library in matxs form which are produced by means of the NJOY code. The point-wise cross-section library is derived from pend files obtained also by NJOY. The AMPX libraries may contain neutron, gamma production and gamma transport data, as well. The produced master libraries can be used either with stand-alone functional modules or with control modules. An assistant package of programs also has been developed in order to facilitate the usage of NJOY. (Authors)

Nuclear libraries for SCALE5.1 system

International Nuclear Information System (INIS)

Vertes, P.

2009-01-01

Codes for preparing master and working AMPX libraries and point-wise (PW) nuclear libraries for SCALE5.1 system have been created. Master and working libraries are constructed from multigroup library in matxs form which are produced by means of the NJOY code. The PW cross-section library is derived from pend files obtained also by NJOY. The AMPX libraries may contain neutron, gamma production and gamma transport data, as well. The produced master libraries can be used either with stand-alone functional modules or with control modules. An assistant package of programs also has been developed in order to facilitate the usage of NJOY. (author)

Comparative evaluation of photon cross section libraries for materials of interest in PET Monte Carlo simulations

CERN Document Server

Zaidi, H

1999-01-01

the many applications of Monte Carlo modelling in nuclear medicine imaging make it desirable to increase the accuracy and computational speed of Monte Carlo codes. The accuracy of Monte Carlo simulations strongly depends on the accuracy in the probability functions and thus on the cross section libraries used for photon transport calculations. A comparison between different photon cross section libraries and parametrizations implemented in Monte Carlo simulation packages developed for positron emission tomography and the most recent Evaluated Photon Data Library (EPDL97) developed by the Lawrence Livermore National Laboratory was performed for several human tissues and common detector materials for energies from 1 keV to 1 MeV. Different photon cross section libraries and parametrizations show quite large variations as compared to the EPDL97 coefficients. This latter library is more accurate and was carefully designed in the form of look-up tables providing efficient data storage, access, and management. Toge...

POINT 2011: ENDF/B-VII.1 Beta2 Temperature Dependent Cross Section Library

Energy Technology Data Exchange (ETDEWEB)

Cullen, D E

2011-04-07

This report is one in the series of 'POINT' reports that over the years have presented temperature dependent cross sections for the then current version of ENDF/B. In each case I have used my personal computer at home and publicly available data and codes. I have used these in combination to produce the temperature dependent cross sections used in applications and presented in this report. I should mention that today anyone with a personal computer can produce these results. The latest ENDF/B-VII.1 beta2 data library was recently and is now freely available through the National Nuclear Data Center (NNDC), Brookhaven National Laboratory. This release completely supersedes all preceding releases of ENDF/B. As distributed the ENDF/B-VII.1 data includes cross sections represented in the form of a combination of resonance parameters and/or tabulated energy dependent cross sections, nominally at 0 Kelvin temperature. For use in our applications the ENDF/B-VII.1 library has been processed into cross sections at eight neutron reactor like temperatures, between 0 and 2100 Kelvin, in steps of 300 Kelvin (the exception being 293.6 Kelvin, for exact room temperature at 20 Celsius). It has also been processed to five astrophysics like temperatures, 1, 10, 100 eV, 1 and 10 keV. For reference purposes, 300 Kelvin is approximately 1/40 eV, so that 1 eV is approximately 12,000 Kelvin. At each temperature the cross sections are tabulated and linearly interpolable in energy. All results are in the computer independent ENDF-6 character format [R2], which allows the data to be easily transported between computers. In its processed form the POINT 2011 library is approximately 16 gigabyte in size and is distributed on one compressed DVDs (see, below for the details of the contents of each DVD).

Development of the V4.2m5 and V5.0m0 Multigroup Cross Section Libraries for MPACT for PWR and BWR

Energy Technology Data Exchange (ETDEWEB)

Kim, Kang Seog [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Clarno, Kevin T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gentry, Cole [Univ. of Tennessee, Knoxville, TN (United States); Wiarda, Dorothea [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Williams, Mark L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kochunas, Brendan [Univ. of Michigan, Ann Arbor, MI (United States); Liu, Yuxuan [Univ. of Michigan, Ann Arbor, MI (United States); Palmtag, Scott [Core Physics, Inc., Wilmington, NC (United States); Godfrey, Andrew T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-03-01

The MPACT neutronics module of the Consortium for Advanced Simulation of Light Water Reactors (CASL) core simulator is a 3-D whole core transport code being developed for the CASL toolset, Virtual Environment for Reactor Analysis (VERA). Key characteristics of the MPACT code include (1) a subgroup method for resonance selfshielding and (2) a whole-core transport solver with a 2-D/1-D synthesis method. The MPACT code requires a cross section library to support all the MPACT core simulation capabilities which would be the most influencing component for simulation accuracy.

AXMIX program for cross section mixing and library arrangement

International Nuclear Information System (INIS)

Haynes, G.C.

1976-03-01

AXMIX is a FORTRAN IV computer code written to provide the user a tool for creating cross-section data sets for ANISN and DOT from cross-section sets already available on cards, nuclide-organized libraries, and group-independent data sets. Numerous options, including adjointing, P/sub n/ adjustments, and changing table length, are available to give the user broad flexibility. The number of energy groups which will fit into the core allocated is determined first. If all groups will fit, the solution is straightforward; if not, then the maximum number of groups which will fit is processed repeatedly by using direct access I/O and storage disks. Some flexibility in applying AXMIX is lost when cross sections to be processed contain upscatter. A special section on upscatter is included in the report. AXMIX is written for IBM System 360 computers with at least 150K bytes of memory. Problems of a practical nature require from 2 to 20 seconds of CPU time on a 360/91 computer. Running time is inversely proportional to the number of groups of data which will fit into core memory. I/O time is 50 to 100 times CPU time

Evaluated cross section libraries

International Nuclear Information System (INIS)

Maqurno, B.A.

1976-01-01

The dosimetry tape (ENDF/B-IV tape 412) was issued in a general CSEWG distribution, August 1974. The pointwise cross section data file was tested with specified reference spectra. A group averaged cross section data file (620 groups based on tape 412) was tested with the above spectra and the results are presented in this report

SENSIT: a cross-section and design sensitivity and uncertainty analysis code. [In FORTRAN for CDC-7600, IBM 360

Energy Technology Data Exchange (ETDEWEB)

Gerstl, S.A.W.

1980-01-01

SENSIT computes the sensitivity and uncertainty of a calculated integral response (such as a dose rate) due to input cross sections and their uncertainties. Sensitivity profiles are computed for neutron and gamma-ray reaction cross sections of standard multigroup cross section sets and for secondary energy distributions (SEDs) of multigroup scattering matrices. In the design sensitivity mode, SENSIT computes changes in an integral response due to design changes and gives the appropriate sensitivity coefficients. Cross section uncertainty analyses are performed for three types of input data uncertainties: cross-section covariance matrices for pairs of multigroup reaction cross sections, spectral shape uncertainty parameters for secondary energy distributions (integral SED uncertainties), and covariance matrices for energy-dependent response functions. For all three types of data uncertainties SENSIT computes the resulting variance and estimated standard deviation in an integral response of interest, on the basis of generalized perturbation theory. SENSIT attempts to be more comprehensive than earlier sensitivity analysis codes, such as SWANLAKE.

COMBINE/PC - a portable neutron spectrum and cross-section generation program

International Nuclear Information System (INIS)

Nigg, D.W.; Grimesey, R.A.; Curtis, R.L.

1990-01-01

Use of personal computers and engineering workstations for complex scientific computations has expanded rapidly in the past few years. This trend is expected to continue in the future with the introduction of increasingly sophisticated microprocessors and microcomputer systems. In response to this, an integrated system of neutronics and radiation transport software suitable for operation in an IBM personal computer (PC)-class environment has been under development at the Idaho National Engineering Laboratory (INEL) for the past 3 years. A key component of this system will be module to produce application-specific multigroup cross-section libraries that can be used in various neutron transport and diffusion theory code modules. This software module, referred to as COMBINE/PC, was recently completed at INEL and is the subject of this paper. COMBINE/PC was developed to provide an ENDF/B-based neutron cross-section generation capability of sufficient sophistication to handle a wide variety of practical fission and fusion-related applications while maintaining a compact machine-independent structure

Multigroup or multipoint thermal neutron data preparation. Programme SIGMA

International Nuclear Information System (INIS)

Matausek, M.V.; Kunc, M.

1974-01-01

When calculating the space energy distribution of thermal neutrons in reactor lattices, in either the multigroup or the multipoint approximation, it is convenient to divide the problem into two independent parts. Firstly, for all material regions of the given reactor lattice cell, the group or the point values of cross sections, scattering kernel and the outer source of thermal neutrons are calculated by a data preparation programme. These quantities are then used as input, by the programme which solves multigroup or multipoint transport equations, to generate the space energy neutron spectra in the cell considered and to determine the related integral quantities, namely the different reaction rates. The present report deals with the first part of the problem. An algorithm for constructing a set of thermal neutron input data, to be used with the multigroup or multipoint version of the code MULTI /1,2,3/, is presented and the new version of the programme SIGMA /4/, written in FORTRAN IV for the CDC-3600 computer, is described. For a given reactor cell material, composed of a number of different isotopes, this programme calculates the group or the point values of the scattering macroscopic absorption cross section, macroscopic scattering cross section, kernel and the outer source of thermal neutrons. Numerous options are foreseen in the programme, concerning the energy variation of cross sections and a scattering kernel, concerning the weighting spectrum in multigroup scheme or the procedure for constructing the scattering matrix in the multipoint scheme and, finally, concerning the organization of output. The details of the calculational algorithm are presented in Section 2 of the paper. Section 3 contains the description of the programme and the instructions for its use (author)

ECNJEFI. A JEFI based 219-group neutron cross-section library: User's manual

International Nuclear Information System (INIS)

Stad, R.C.L. van der; Gruppelaar, H.

1992-07-01

This manual describes the contents of the ECNJEF1 library. The ECNJEF1 library is a JEF1.1 based 219-group AMPX-Master library for reactor calculations with the AMPX/SCALE-system, e.g. the PASC-3 system as implemented at the Netherlands Energy Research Foundation in Petten, Netherlands. The group cross-section data were generated with NJOY and NPTXS/XLACS-2 from the AMPX system. The data on the ECNJEF1 library allows resolved-resonance treatment by NITAWL and/or unresolved resonance self-shielding by BONAMI. These codes are based upon the Nordheim and Bondarenko methods, respectively. (author). 10 refs., 7 tabs

FENDL/A-2.0. Neutron activation cross section data library for fusion applications

International Nuclear Information System (INIS)

Pashchenko, A.B.; Wienke, H.; Kopecky, J.; Sublet, J.C. Sublet; Forrest, R.A.

1997-01-01

This document describes the contents of a comprehensive neutron cross section data library for 13,006 neutron activation reactions with 739 target nuclides from H (A=1,Z=1) to Cm (A=248,Z=96), in the incident energy range up to 20 MeV. FENDL/A-2 is a sublibrary of FENDL-2, the second revision of the evaluated nuclear data library for fusion applications. It is supplemented by a decay data library FENDL/D-2 in ENDF-6 format for 1867 nuclides. The data are available from the IAEA Nuclear Data Section online via INTERNET by FTP command, or on magnetic tape upon request. (author)

Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Boiling Water Reactor Benchmark Problem

Directory of Open Access Journals (Sweden)

Kulesza Joel A.

2016-01-01

Full Text Available This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.

BARC 75 - A 75 group neutron-photon coupled cross-section library with P5- anisotropic scattering matrices

International Nuclear Information System (INIS)

Garg, S.B.

1990-01-01

A 75 group neutron-photon coupled cross-section library has been developed for 42 reactor nuclides utilizing the basic cross-section files - ENDF/B-IV for neutrons and DLC-7F for photons. 50 neutron energy groups and gamma energy groups are included in this library which should be well suited to carry out safety, shielding and core physics studies of nuclear reactors based on fission or fusion processes. This library is also adequate for oil logging and mineral exploration investigations. (author). 11 refs., 3 tabs

Basis calculation of phase cross section library in a low power fast reactor neutronic simulation

International Nuclear Information System (INIS)

Jachic, J.

1993-09-01

In order to implement the utilization of the efficient multidimensional cubic SPLINE interpolation, we determine the phase library bases for net like relevant state components. A generic cubic surface and a weighted plane pertinent alternative interpolating methods used capable to generate cross sections values for fixed coordinates from cell code calculated data points is used. It is verified that the phase library bases increases or decrease smoothly and monotonically with the spectrum asymmetry and total flux buckling. This justifies its use in cross section updating avoiding cell calculations. (author)

Multigroup and coupled forward-adjoint Monte Carlo calculation efficiencies for secondary neutron doses from proton beams

International Nuclear Information System (INIS)

Kelsey IV, Charles T.; Prinja, Anil K.

2011-01-01

We evaluate the Monte Carlo calculation efficiency for multigroup transport relative to continuous energy transport using the MCNPX code system to evaluate secondary neutron doses from a proton beam. We consider both fully forward simulation and application of a midway forward adjoint coupling method to the problem. Previously we developed tools for building coupled multigroup proton/neutron cross section libraries and showed consistent results for continuous energy and multigroup proton/neutron transport calculations. We observed that forward multigroup transport could be more efficient than continuous energy. Here we quantify solution efficiency differences for a secondary radiation dose problem characteristic of proton beam therapy problems. We begin by comparing figures of merit for forward multigroup and continuous energy MCNPX transport and find that multigroup is 30 times more efficient. Next we evaluate efficiency gains for coupling out-of-beam adjoint solutions with forward in-beam solutions. We use a variation of a midway forward-adjoint coupling method developed by others for neutral particle transport. Our implementation makes use of the surface source feature in MCNPX and we use spherical harmonic expansions for coupling in angle rather than solid angle binning. The adjoint out-of-beam transport for organs of concern in a phantom or patient can be coupled with numerous forward, continuous energy or multigroup, in-beam perturbations of a therapy beam line configuration. Out-of-beam dose solutions are provided without repeating out-of-beam transport. (author)

Creation of the equilibrium core PBMR ORIGEN-S cross section library

International Nuclear Information System (INIS)

Stoker, C.C.; Reitsma, F.; Karriem, Z.

2002-01-01

As part of the design calculations for the Pebble Bed Modular Reactor (PBMR), fuel inventories, neutron and gamma sources and decay heat needs to be determined for the fuel spheres. Using the SCALE4.4 code system, a PBMR specific cross section library was created for the ORIGEN-S depletion calculations, assuming a 10-pass refueling system for the PBMR. In this paper the rationale for the creation of the PBMR library is evaluated in terms of the spectrum dependence due to burn-up. The ORIGEN-S PBMR library was further evaluated comparing the results for different parameters calculated with the reactor analysis diffusion code VSOP and the Monte Carlo code MCNP4C. (author)

NJOY processed multigroup library for fast reactor applications and point data library for MCNP - Experience and validation

International Nuclear Information System (INIS)

Kim Jung-Do; Gil Choong-Sup

1996-01-01

JEF-1-based 50-group cross section library for fast reactor applications and point data library for continuous-energy Monte Carlo code MCNP have been generated using NJOY91.38 system. They have been examined by analyzing measured integral quantities such as criticality and central reaction rate ratios for 8 small fast critical assemblies. (author). 9 refs, 2 figs, 10 tabs

Generation of the library of neutron cross sections for the Record code of the Fuel Management System (FMS)

International Nuclear Information System (INIS)

Alonso V, G.; Hernandez L, H.

1991-11-01

On the basis of the library structure of the RECORD code a method to generate the neutron cross sections by means of the ENDF-B/IV database and the NJOY code has been developed. The obtained cross sections are compared with those of the current library which was processed using the ENDF-B/III version. (Author)

Production and testing of HENDL-2.1/CG coarse-group cross-section library based on ENDF/B-VII.0

International Nuclear Information System (INIS)

Xu Dezheng; He Zhaozhong; Zou Jun; Zeng Qin

2010-01-01

A coarse-group coupled neutron and photon (27n + 21γ) cross-section library HENDL-2.1/CG, based on ENDF/B-VII.0 evaluate data source, has been produced by FDS Team in Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). HENDL-2.1/CG containing 350 nuclide cross-section files (from 1 H to 252 Cf) was generated in MATXS format with the NJOY processing system and then by compiling coarse-group problem-dependent format using the TRANSX code. In order to verify the availability and reliability of the HENDL-2.1/CG data library, requisite benchmark calculations were performed and compared with HENDL-2.0/MG fine-group coupled neutron and photon (175n + 42γ) cross-section library. In general, results using the coarse-group library showed similarly believable as fine-group library.

ZZ COVFILS, 30-Group Covariance Library from ENDF/B-5 for Sensitivity Studies

International Nuclear Information System (INIS)

Muir, D.W.

1997-01-01

1 - Description of program or function: Format: ENDB/F; Number of groups: 30-Group Covariance Library; Nuclides: H-1, B-10, C, O-16, Cr, Fe, Ni, Cu, Pb. Origin: ENDF/B-V. COVFILS is a 30-Group Covariance Library. It contains neutron cross sections, and their uncertainties and correlation in multigroup form. These data can be used, in conjunction with sensitivity information, to estimate the data-related uncertainty in calculated integral quantities such as radiation-damage or heating. 2 - Method of solution: COVFILS was obtained by processing evaluations from ENDF/B-V with ERRORR module of the NJOY nuclear data processing system (LA-9303-M, Vols. 1).The group structure is the Los Alamos 30-group structure which is listed in 'File 1' of each multigroup data set in the library

Process of cross section generation for radiation shielding calculations, using the NJOY code

International Nuclear Information System (INIS)

Ono, S.; Corcuera, R.P.

1986-10-01

The process of multigroup cross sections generation for radiation shielding calculations, using the NJOY code, is explained. Photon production cross sections, processed by the GROUPR module, and photon interaction cross sections processed by the GAMINR are given. These data are compared with the data produced by the AMPX system and published data. (author) [pt

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.

Neutron cross-section libraries in the AMPX master interface format for thermal and fast reactors

International Nuclear Information System (INIS)

Bjerke, M.A.; Webster, C.C.

1981-12-01

Neutron cross-section libraries in the AMPX master interface format have been created for three reactor types. Included are an 84-group library for use with light-water reactors, a 27-group library for use with heavy-water CANDU reactors and a 126-group library for use with liquid metal fast breeder reactors. In general, ENDF/B data were used in the creation of these libraries, and the nuclides included in each library should be sufficient for most neutronic analyses of reactors of that type. Each library has been used successfully in fuel depletion calculations

Multigroup P8 - elastic scattering matrices of main reactor elements

International Nuclear Information System (INIS)

Garg, S.B.; Shukla, V.K.

1979-01-01

To study the effect of anisotropic scattering phenomenon on shielding and neutronics of nuclear reactors multigroup P8-elastic scattering matrices have been generated for H, D, He, 6 Li, 7 Li, 10 B, C, N, O, Na, Cr, Fe, Ni, 233 U, 235 U, 238 U, 239 Pu, 240 Pu, 241 Pu and 242 Pu using their angular distribution, Legendre coefficient and elastic scattering cross-section data from the basic ENDF/B library. Two computer codes HSCAT and TRANS have been developed to complete this task for BESM-6 and CDC-3600 computers. These scattering matrices can be directly used as input to the transport theory codes ANISN and DOT. (auth.)

Status of neutron dosimetry cross sections

International Nuclear Information System (INIS)

Griffin, P.J.; Kelly, J.G.

1992-01-01

Several new cross section libraries, such as ENDF/B-VI(release 2), IRDF-90,JEF-2.2, and JENDL-3 Dosimetry, have recently been made available to the dosimetry community. the Sandia National Laboratories (SNL) Radiation Metrology Laboratory (RML) has worked with these libraries since pre-release versions were available. this paper summarizes the results of the intercomparison and testing of dosimetry cross sections. As a result of this analysis, a compendium of the best dosimetry cross sections was assembled from the available libraries for use within the SNL RML. this library, referred to as the SNLRML Library, contains 66 general dosimetry sensors and 3 special dosimeters unique to the RML sensor inventory. The SNLRML cross sections have been put into a format compatible with commonly used spectrum determination codes

Bonderenko self-shielded cross sections and multiband parameters derived from the LLL Evaluated-Nuclear-Data Library (ENDL)

International Nuclear Information System (INIS)

Cullen, D.E.

1978-01-01

Bonderenko self-shielded cross sections and multiband parameters from the Lawrence Livermore Laboratory Evaluated-Nuclear-Data Library (ENDL) as of July 4, 1978 are presented. These data include total, elastic, capture, and fission cross sections in the TART 175 group structure. Multiband parameters are listed. Bonderenko self-shielded cross section and the multiband parameters are presented on microfiche

Calculated neutron-activation cross sections for E/sub n/ /le/ 100 MeV for a range of accelerator materials

International Nuclear Information System (INIS)

Bozoian, M.; Arthur, E.D.; Perry, R.T.; Wilson, W.B.; Young, P.G.

1988-01-01

Activation problems associated with particle accelerators are commonly dominated by reactions of secondary neutrons produced in reactions of beam particles with accelerator or beam stop materials. Measured values of neutron-activation cross sections above a few MeV are sparse. Calculations with the GNASH code have been made for neutrons incident on all stable nuclides of a range of elements common to accelerator materials. These elements include B, C, N, O, Ne, Mg, Al, Si, P, S, Ar, K, Ca, Cr, Mn, Fe, Co, Ni, Cu, Zn, Zr, Mo, Nd, and Sm. Calculations were made for a grid of incident neutron energies extending to 100 MeV. Cross sections leading to the direct production of as many as 87 activation products for each of 84 target nuclide were tabulated on this grid of neutron energies, each beginning with the threshold for the product nuclide's formation. Multigrouped values of these cross sections have been calculated and are being integrated into the cross-section library of the REAC-2 neutron activation code. Illustrative cross sections are presented. 20 refs., 6 figs., 1 tab

Experience in developing and using the VITAMIN-C 171-neutron, 36-gamma-ray group cross-section library

International Nuclear Information System (INIS)

Roussin, R.W.; Weisbin, C.R.; White, J.E.; Wright, R.Q.; Greene, N.M.; Ford, W.E. III; Wright, J.B.; Diggs, B.R.

1978-01-01

The Department of Energy (DOE) Division of Magnetic Fusion Energy (DMFE) and Reactor Research and Technology (DRRT) jointly sponsored the development of a coupled, fine-group cross-section library. The 171-neutron, 36-gamma-ray group library is intended to be applicable to fusion reactor neutronics and LMFBR core and shield analysis. Versions of the library are available from the Radiation Shielding Information Center (RSIC) at Oak Ridge National Laboratory in both AMPX and CCCC formats. Computer codes for energy group collapsing, interpolation on Bondarenko factors for resonance self-shielding and temperature corrections, and various other useful data manipulations are available. The experience gained in the utilization of this library is discussed. Indications are that this venture, which is designed to allow users to derive problem-dependent cross sections from a fine-group master library, has been a success

REX1-87, Multigroup Neutron Cross-Sections from ENDF/B

International Nuclear Information System (INIS)

Gopalakrishnan, V.; Ganesan, S.

1988-01-01

1 - Description of program or function: The program calculates self- shielding factors for reactor applications from a pre-processed (linearized) evaluated nuclear data file in the ENDF/B format. 2 - Method of solution: Bondarenko definition of multigroup self- shielding factors invoking narrow resonance treatment is used. 3 - Restrictions on the complexity of the problem: a) Maximum no. of energy group is 620. b) Only the built-in forms of the weighting functions can be chosen. c) The program is strictly limited to resolved resonance region from physical considerations

Calculation of self-shielding factors for cross-sections in the unresolved resonance region using the GRUCON applied program package

International Nuclear Information System (INIS)

Sinitsa, V.V.

1984-11-01

The author gives a scheme for the calculation of the self-shielding factors in the unresolved resonance region using the GRUCON applied program package. This package is especially created to be used in the conversion of evaluated neutron cross-section data, as available in existing data libraries, into multigroup microscopic constants. A detailed description of the formulae and algorithms used in the programs is given. Some typical examples of calculation are considered and the results are compared with those of other authors. The calculation accuracy is better than 2%

Evaluated cross-section libraries and kerma factors for neutrons up to 100 MeV on 12C

International Nuclear Information System (INIS)

Chadwick, M.B.; Blann, M.; Cox, L.; Young, P.G.; Meigooni, A.

1995-01-01

A program is being carried out at Lawrence Livermore National Laboratory to develop high-energy evaluated nuclear data libraries for use in Monte Carlo simulations of cancer radiation therapy. In this report we describe evaluated cross sections and kerma factors for neutrons with incident energies up to 100 MeV on 12 C. The aim of this effort is to incorporate advanced nuclear physics modeling methods, with new experimental measurements, to generate cross section libraries needed for an accurate simulation of dose deposition in fast neutron therapy. The evaluated libraries are based mainly on nuclear model calculations, benchmarked to experimental measurements where they exist. We use the GNASH code system, which includes Hauser-Feshbach, preequilibrium, and direct reaction mechanisms. The libraries tabulate elastic and nonelastic cross sections, angle-energy correlated production spectra for light ejectiles with A≤and kinetic energies given to light ejectiles and heavy recoil fragments. The major steps involved in this effort are: (1) development and validation of nuclear models for incident energies up to 100 MeV; (2) collation of experimental measurements, including new results from Louvain-la-Nueve and Los Alamos; (3) extension of the Livermore ENDL formats for representing high-energy data; (4) calculation and evaluation of nuclear data; and (5) validation of the libraries. We describe the evaluations in detail, with particular emphasis on our new high-energy modeling developments. Our evaluations agree well with experimental measurements of integrated and differential cross sections. We compare our results with the recent ENDF/B-VI evaluation which extends up to 32 MeV

Neutron Cross Section Libraries for Cryogenic Aromatic Moderator Materials

International Nuclear Information System (INIS)

Cantargi, Florencia; Granada, J.R.; Sbaffoni, Maria Monica

2008-01-01

The dynamics of a set of aromatic hydrocarbons, such as benzene, toluene, mesitylene and a 3:2 mixture (by volume) of mesitylene and toluene, all of them in solid phase, was studied as potential moderator materials for cold neutron sources. Cross section libraries were generated for hydrogen bounded in those materials, at several temperatures in ACE format, and they were used in MCNP calculations to analyze their neutron production compared with traditional materials like solid methane and liquid hydrogen. In particular, cross section libraries were generated at 20 K, which is the operating temperature of the majority of the existing cold neutron sources. Although solid methane is the best moderator in terms of cold neutron production, it has very poor radiation resistance, causing spontaneous burping even at fairly low doses. Such effect is considerably reduced in the aromatic hydrocarbons. On the other hand, all of them show a similar and significant neutron production, with the exception of benzene. Even though those aromatic materials are very easy to handle, the solid phases that produce an enhanced flux of cold neutrons correspond to amorphous structures rich in low-energy excitations, and they can be created through lengthy cooling processes requiring in many cases additional annealing stages. The 3:2 mesitylene-toluene mixture, that forms in a simple and direct manner the appropriate disordered structure, constitutes an excellent cryogenic moderator material, as it is able to produce an intense flux of cold neutrons while presenting high resistance to radiation, thus conforming a new and advantageous alternative to traditional moderator materials. (authors)

Conception and development of an adaptive energy mesher for multigroup library generation of the transport codes

International Nuclear Information System (INIS)

Mosca, P.

2009-12-01

The deterministic transport codes solve the stationary Boltzmann equation in a discretized energy formalism called multigroup. The transformation of continuous data in a multigroup form is obtained by averaging the highly variable cross sections of the resonant isotopes with the solution of the self-shielding models and the remaining ones with the coarse energy spectrum of the reactor type. So far the error of such an approach could only be evaluated retrospectively. To remedy this, we studied in this thesis a set of methods to control a priori the accuracy and the cost of the multigroup transport computation. The energy mesh optimisation is achieved using a two step process: the creation of a reference mesh and its optimized condensation. In the first stage, by refining locally and globally the energy mesh, we seek, on a fine energy mesh with subgroup self-shielding, a solution equivalent to a reference solver (Monte Carlo or pointwise deterministic solver). In the second step, once fixed the number of groups, depending on the acceptable computational cost, and chosen the most appropriate self-shielding models to the reactor type, we look for the best bounds of the reference mesh minimizing reaction rate errors by the particle swarm optimization algorithm. This new approach allows us to define new meshes for fast reactors as accurate as the currently used ones, but with fewer groups. (author)

Generation and Testing of the ENDF/B-VI Continuous-Energy Cross-Section Library for Use with Continuous-Energy Versions of KENO

International Nuclear Information System (INIS)

Goluoglu, Sedat; Dunn, Michael E.; Greene, Norman Maurice; Petrie Jr, Lester M.; Hollenbach, Daniel F.

2007-01-01

KENO V.a and KENO-VI are Monte Carlo codes that solve the multigroup form of the Boltzmann transport equation. These codes are part of the SCALE system of codes and are used for performing criticality calculations of systems with fissionable material. In general, continuous-energy Monte Carlo methods are preferred because such an approach avoids many of the assumptions inherent in the multigroup treatment. On the other hand, continuous-energy treatment is much more demanding in terms of computer storage space for data, memory requirements, and calculation speed. Continuous-energy versions of KENO V.a and KENO-VI have been created and are being extensively tested. Generation of ENDF/B-VI continuous-energy cross sections is explained, and the results of the validation and verification of the codes and the data are presented

Code implementation of partial-range angular scattering cross sections: GAMMER and MORSE

International Nuclear Information System (INIS)

Ward, J.T. Jr.

1978-01-01

A partial-range (finite-element) method has been previously developed for representing multigroup angular scattering in Monte Carlo photon transport. Computer application of the method, with preliminary quantitative results is discussed here. A multigroup photon cross section processing code, GAMMER, was written which utilized ENDF File 23 point data and the Klein--Nishina formula for Compton scattering. The cross section module of MORSE, along with several execution routines, were rewritten to permit use of the method with photon transport. Both conventional and partial-range techniques were applied for comparison to calculating angular and spectral penetration of 6-MeV photons through a six-inch iron slab. GAMMER was found to run 90% faster than SMUG, with further improvement evident for multiple-media situations; MORSE cross section storage was reduced by one-third; cross section processing, greatly simplified; and execution time, reduced by 15%. Particle penetration was clearly more forward peaked, as moment accuracy is retained to extremly high order. This method of cross section treatment offers potential savings in both storage and handling, as well as improved accuracy and running time in the actual execution phase. 3 figures, 4 tables

Improvements on burnup chain model and group cross section library in the SRAC system

International Nuclear Information System (INIS)

Akie, Hiroshi; Okumura, Keisuke; Takano, Hideki; Ishiguro, Yukio; Kaneko, Kunio.

1992-01-01

Data and functions of the cell burnup calculation of the SRAC system were revised to improve mainly the accuracy of the burnup calculation of high conversion light water reactors (HCLWRs). New burnup chain models were developed in order to treat fission products (FPs) and actinide nuclides in detail. Group cross section library, SRACLIB-JENDL2, was generated based on JENDL-2 nuclear data file. In generating this library, emphasis was placed on FPs and actinides. Also revised were the data such as the average energy release per fission for various actinides. These improved data were verified by performing the burnup analysis of PWR spent fuels. Some new functions were added to the SRAC system for the convenience to yield macroscopic cross sections used in the core burnup process. (author)

PUFF-IV, Code System to Generate Multigroup Covariance Matrices from ENDF/B-VI Uncertainty Files

International Nuclear Information System (INIS)

2007-01-01

1 - Description of program or function: The PUFF-IV code system processes ENDF/B-VI formatted nuclear cross section covariance data into multigroup covariance matrices. PUFF-IV is the newest release in this series of codes used to process ENDF uncertainty information and to generate the desired multi-group correlation matrix for the evaluation of interest. This version includes corrections and enhancements over previous versions. It is written in Fortran 90 and allows for a more modular design, thus facilitating future upgrades. PUFF-IV enhances support for resonance parameter covariance formats described in the ENDF standard and now handles almost all resonance parameter covariance information in the resolved region, with the exception of the long range covariance sub-subsections. PUFF-IV is normally used in conjunction with an AMPX master library containing group averaged cross section data. Two utility modules are included in this package to facilitate the data interface. The module SMILER allows one to use NJOY generated GENDF files containing group averaged cross section data in conjunction with PUFF-IV. The module COVCOMP allows one to compare two files written in COVERX format. 2 - Methods: Cross section and flux values on a 'super energy grid,' consisting of the union of the required energy group structure and the energy data points in the ENDF/B-V file, are interpolated from the input cross sections and fluxes. Covariance matrices are calculated for this grid and then collapsed to the required group structure. 3 - Restrictions on the complexity of the problem: PUFF-IV cannot process covariance information for energy and angular distributions of secondary particles. PUFF-IV does not process covariance information in Files 34 and 35; nor does it process covariance information in File 40. These new formats will be addressed in a future version of PUFF

Generation of ENDF/B-IV based 35 group neutron cross-section library and its application in criticality studies

International Nuclear Information System (INIS)

Garg, S.B.; Sinha, A.

1985-01-01

A 35 group cross-section library with P/sub 3/-anisotropic scattering matrices and resonance self-shielding factors has been generated from the basic ENDF/B-IV cross-section files for 57 elements. This library covers the neutron energy range from 0.005 ev to 15 MeV and is well suited for the neutronics and safety analysis of fission, fusion and hybrid systems. The library is contained in two well known files, namely, ISOTXS and BRKOXS. In order to test the efficacy of this library and to bring out the importance of resonance self-shielding, a few selected fast critical assemblies representing large dilute oxide and carbide fueled uranium and plutonium based systems have been analysed. These assemblies include ZPPR/sub 2/, ZPR-3-48, ZPR-3-53, ZPR-6-6A, ZPR-6-7, ZPR-9-31 and ZEBRA-2 and are amongst those recommended by the US Nuclear Data Evaluation Working Group for testing the accuracy of cross-sections. The evaluated multiplication constants of these assemblies compare favourably with those calculated by others

Development of ANJOYMC Program for Automatic Generation of Monte Carlo Cross Section Libraries

International Nuclear Information System (INIS)

Kim, Kang Seog; Lee, Chung Chan

2007-03-01

The NJOY code developed at Los Alamos National Laboratory is to generate the cross section libraries in ACE format for the Monte Carlo codes such as MCNP and McCARD by processing the evaluated nuclear data in ENDF/B format. It takes long time to prepare all the NJOY input files for hundreds of nuclides with various temperatures, and there can be some errors in the input files. In order to solve these problems, ANJOYMC program has been developed. By using a simple user input deck, this program is not only to generate all the NJOY input files automatically, but also to generate a batch file to perform all the NJOY calculations. The ANJOYMC program is written in Fortran90 and can be executed under the WINDOWS and LINUX operating systems in Personal Computer. Cross section libraries in ACE format can be generated in a short time and without an error by using a simple user input deck

Improvement of Modeling HTGR Neutron Physics by Uncertainty Analysis with the Use of Cross-Section Covariance Information

Science.gov (United States)

Boyarinov, V. F.; Grol, A. V.; Fomichenko, P. A.; Ternovykh, M. Yu

2017-01-01

This work is aimed at improvement of HTGR neutron physics design calculations by application of uncertainty analysis with the use of cross-section covariance information. Methodology and codes for preparation of multigroup libraries of covariance information for individual isotopes from the basic 44-group library of SCALE-6 code system were developed. A 69-group library of covariance information in a special format for main isotopes and elements typical for high temperature gas cooled reactors (HTGR) was generated. This library can be used for estimation of uncertainties, associated with nuclear data, in analysis of HTGR neutron physics with design codes. As an example, calculations of one-group cross-section uncertainties for fission and capture reactions for main isotopes of the MHTGR-350 benchmark, as well as uncertainties of the multiplication factor (k∞) for the MHTGR-350 fuel compact cell model and fuel block model were performed. These uncertainties were estimated by the developed technology with the use of WIMS-D code and modules of SCALE-6 code system, namely, by TSUNAMI, KENO-VI and SAMS. Eight most important reactions on isotopes for MHTGR-350 benchmark were identified, namely: 10B(capt), 238U(n,γ), ν5, 235U(n,γ), 238U(el), natC(el), 235U(fiss)-235U(n,γ), 235U(fiss).

A subroutine for the calculation of resonance cross sections of U-238 in HTR fuel elements

Energy Technology Data Exchange (ETDEWEB)

Cuniberti, R; Marullo, G C

1971-02-15

In this paper, a survey of the codes used at Ispra for the calculations of resonance absorption in HTR fuel elements is presented and a subroutine for the calculation of resonance cross-sections, in a seven groups energy structure, for a HTR lattice of annular type is described. A library of homogeneous resonance integrals and a wide tabulation of lump and kernel Bell factors, and moderators efficiency is given. This paper deals mainly with the problem of taking into account the correct slowing down of neutrons in the graphite and with the derivation of Bell factors to be used in a multigroup calculation scheme.

WIMS library up-date project: first stage results

International Nuclear Information System (INIS)

Prati, A.; Claro, L.H.

1990-01-01

The following benchmarks: TRX1, TRX2, BAPL-UO sub(2)-1, BAPL-UO sub (2)-2, BAPL-UO sub(2)-3 have been calculated with the WIMSD/4 code, as a contribution of CTA/IEAv, to the first stage of the WIMS Library Update Project, coordinated by the International Atomic Energy Agency. The card image input for each benchmark has been attached and the major input options/parameters are commented. The version of the WIMSD/4 code and its multigroup cross section library used to run the benchmarks are specified. Results from the major integral parameters are presented and discussed. (author)

MVP/GMVP II, MC Codes for Neutron and Photon Transport Calc. based on Continuous Energy and Multigroup Methods

International Nuclear Information System (INIS)

2005-01-01

A - Description of program or function: (1) Problems to be solved: MVP/GMVP can solve eigenvalue and fixed-source problems. The multigroup code GMVP can solve forward and adjoint problems for neutron, photon and neutron-photon coupled transport. The continuous-energy code MVP can solve only the forward problems. Both codes can also perform time-dependent calculations. (2) Geometry description: MVP/GMVP employs combinatorial geometry to describe the calculation geometry. It describes spatial regions by the combination of the 3-dimensional objects (BODIes). Currently, the following objects (BODIes) can be used. - BODIes with linear surfaces: half space, parallelepiped, right parallelepiped, wedge, right hexagonal prism; - BODIes with quadratic surface and linear surfaces: cylinder, sphere, truncated right cone, truncated elliptic cone, ellipsoid by rotation, general ellipsoid; - Arbitrary quadratic surface and torus. The rectangular and hexagonal lattice geometry can be used to describe the repeated geometry. Furthermore, the statistical geometry model is available to treat coated fuel particles or pebbles for high temperature reactors. (3) Particle sources: The various forms of energy-, angle-, space- and time-dependent distribution functions can be specified. (4) Cross sections: The ANISN-type PL cross sections or the double-differential cross sections can be used in the multigroup code GMVP. On the other hand, the specific cross section libraries are used in the continuous-energy code MVP. The libraries are generated from the evaluated nuclear data (JENDL-3.3, ENDF/B-VI, JEF-3.0 etc.) by using the LICEM code. The neutron cross sections in the unresolved resonance region are described by the probability table method. The neutron cross sections at arbitrary temperatures are available for MVP by just specifying the temperatures in the input data. (5) Boundary conditions: Vacuum, perfect reflective, isotropic reflective (white), periodic boundary conditions can be

Differences between cross-section libraries for neutron dosimetry; Diferencas entre bibliotecas de secoes de choque para dosimetria de neutrons

Energy Technology Data Exchange (ETDEWEB)

Tardelli, T.C.; Stecher, L.C.; Coelho, T.S.; Castro, V.A. De; Cavalieri, T.A.; Menzel, F.; Giarola, R.S.; Domingos, D.B.; Yoriyaz, H., E-mail: tiago.tardelli@gmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear

2013-08-15

Absorbed dose calculations depend on a consistent set of nuclear data used in simulations in computer codes. Nuclear data are stored in libraries, however, the information available about the differences in dose caused by different libraries are rare. The libraries are processed by a computer system to be able to be used by a radiation transport code. One of the systems capable of processing nuclear data is the NJOY system. The objective of this study is to evaluate the nuclear data libraries for neutrons available in the literature, and to quantify the differences in absorbed dose obtained using the libraries JENDL 4.0, JEFF 3.3.1 and ENDF/B.VII. The absorbed dose calculation was performed on a simple geometric model, as spheres, and in anthropomorphic model of the human body based on the ICRP-110 for neutron transport simulation using the MCNP5 code. The results were compared with literature data. The results obtained with cross sections from the libraries JEFF and ENDF/B.VII have shown to be identical in most cases, except for one case where the difference has exceeded 10%. The results obtained with JENDL library has shown to be considerably different in most cases comparing to other two libraries. Some differences were over 200%. The dose calculations showed differences between the libraries, which is justified by differences in the cross sections. It has been observed that the cross sections values of certain nuclides assume quite different values in different libraries. These differences in turn cause considerable differences in dose calculations. (author)

Production and Testing of the VITAMIN-B7 Fine-Group and BUGLE-B7 Broad-Group Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data

Energy Technology Data Exchange (ETDEWEB)

Risner, J. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wiarda, D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dunn, M. E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Miller, T. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Peplow, D. E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Patton, B. W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2011-09-30

New coupled neutron-gamma cross-section libraries have been developed for use in light water reactor (LWR) shielding applications, including pressure vessel dosimetry calculations. The libraries, which were generated using Evaluated Nuclear Data File/B Version VII Release 0 (ENDF/B-VII.0), use the same fine-group and broad-group energy structures as the VITAMIN-B6 and BUGLE-96 libraries. The processing methodology used to generate both libraries is based on the methods used to develop VITAMIN-B6 and BUGLE-96 and is consistent with ANSI/ANS 6.1.2. The ENDF data were first processed into the fine-group pseudo-problem-independent VITAMIN-B7 library and then collapsed into the broad-group BUGLE-B7 library. The VITAMIN-B7 library contains data for 391 nuclides. This represents a significant increase compared to the VITAMIN-B6 library, which contained data for 120 nuclides. The BUGLE-B7 library contains data for the same nuclides as BUGLE-96, and maintains the same numeric IDs for those nuclides. The broad-group data includes nuclides which are infinitely dilute and group collapsed using a concrete weighting spectrum, as well as nuclides which are self-shielded and group collapsed using weighting spectra representative of important regions of LWRs. The verification and validation of the new libraries includes a set of critical benchmark experiments, a set of regression tests that are used to evaluate multigroup crosssection libraries in the SCALE code system, and three pressure vessel dosimetry benchmarks. Results of these tests confirm that the new libraries are appropriate for use in LWR shielding analyses and meet the requirements of Regulatory Guide 1.190.

FENDL/A-MCNP and FENDL/A-VITJE. The processed neutron activation cross-section data files of the FENDL project. Version 1.1 of March 1995. Summary documentation

International Nuclear Information System (INIS)

Pashchenko, A.B.; Wienke, H.; Ganesan, S.

1996-01-01

This document summarizes a neutron activation cross-section database processed in two formats as generated by F.M. Mann within the project of the Fusion Evaluated Nuclear Data Library (FENDL): in continuous energy format as used by the Monte Carlo neutron/photon transport code MCNP; and in 175 group multigroup format with VIT-E weighting spectrum, as used by the transmutation code REAC*2/3. The data are available from the IAEA Nuclear Data Section online via INTERNET by FTP command, or on magnetic tape. (author). 2 refs, 1 tab

Multi-group transport methods for high-resolution neutron activation analysis

International Nuclear Information System (INIS)

Burns, K. A.; Smith, L. E.; Gesh, C. J.; Shaver, M. W.

2009-01-01

The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples. In these applications, high-resolution gamma-ray spectrometers are used to preserve as much information as possible about the emitted photon flux, which consists of both continuum and characteristic gamma rays with discrete energies. Monte Carlo transport is the most commonly used modeling tool for this type of problem, but computational times for many problems can be prohibitive. This work explores the use of multi-group deterministic methods for the simulation of neutron activation problems. Central to this work is the development of a method for generating multi-group neutron-photon cross-sections in a way that separates the discrete and continuum photon emissions so that the key signatures in neutron activation analysis (i.e., the characteristic line energies) are preserved. The mechanics of the cross-section preparation method are described and contrasted with standard neutron-gamma cross-section sets. These custom cross-sections are then applied to several benchmark problems. Multi-group results for neutron and photon flux are compared to MCNP results. Finally, calculated responses of high-resolution spectrometers are compared. Preliminary findings show promising results when compared to MCNP. A detailed discussion of the potential benefits and shortcomings of the multi-group-based approach, in terms of accuracy, and computational efficiency, is provided. (authors)

Generation of seven group cross section library for TRIGA LEU fuel in CITATION format and benchmarking some experimental and operational data

International Nuclear Information System (INIS)

Sarker, M.M.; Bhuiyan, S.I.; Akramuzzaman, M.

2007-01-01

The principal objective of this study is to validate the seven group cross section library in CITATION format for TRIGA LEU Fuel. This presentation deals with the 'generation of a cross section library for the CITATION and its validation. We used WIMSD-5B version for the generation of all group constants. The overall strategy is: (1) use WIMS package to generate few group neutron macroscopic cross section (cell constants) for all of the materials in the core and its immediate neighborhood (2) use 3-D code CITATION to perform the global analysis of the core to study: multiplication factor, neutron flux distribution and power peaking factors. Various options available in WIMS program were studied in depth to finalize the models to generate the most appropriate group constants. For the global analysis the code CITATION and a post processing program FCAP were chosen. Thus a seven group cross section library for the calculations of TRIGA Research Reactor was generated. To investigate the validity of the generated library a critical experiment of the TRIGA research reactor was benchmarked. (author)

Testing of a JEF-1 based WIMS-D cross section library for migration area and k-infinity predictions for LWHCR lattices

International Nuclear Information System (INIS)

Pelloni, S.; Stepanek, J.

1987-01-01

The cell code WIMSD4 is used for the analysis of PROTEUS-LWHCR experiments. A library for this code which is based on the European evaluation JEF-1 was produced at EIR using the Los Alamos NJOY system with its module WIMSR and the Canadian management code WILMA. In general, this library delivered more accurate eigenvalues and reaction rates than the WIMS-Standard and WIMS81 libraries did in comparison to experimental values from PROTEUS-LWHCR Cores 1-3. However, large discrepancies (up to about 10%) occured between calculated migration areas (M 2 ). Additional investigations have been undertaken to clarify this problem, since theoretical M 2 -values are needed for deducing k-infinity in the experiments. This has been done in the context of calculations for a reference LWHCR test lattice. The following major reasons for these deviations were found. First, the self-scattering term in non-moderators (P 0 matrix) in the JEF-1 library was not transport corrected. Second, Standard and JEF-1 libraries use infinite dilute cross sections for 238 U, whereas the WIMS81 library uses fully shielded cross sections. Third, the standard library uses the 'row' formula for the transport correction, whereas the 'inflow' formula is applied in the case of JEF-1 and WIMS81 libraries. Lastly, oxygen and 238 U scattering cross sections in the fast energy range are smaller in the case of the WIMS81 library. Differences in calculated k-infinity values between the currently used library and WIMS81 (up to 3%) come (in order of importance for the reference LWHCR lattice) mainly from resonance cross sections for 240 Pu capture, 238 U capture and 239 Pu fission. Recommendations have been made for generating a new JEF-1 library using updated versions of WIMSR and WILMA. (author)

New Tools to Prepare ACE Cross-section Files for MCNP Analytic Test Problems

International Nuclear Information System (INIS)

Brown, Forrest B.

2016-01-01

Monte Carlo calculations using one-group cross sections, multigroup cross sections, or simple continuous energy cross sections are often used to: (1) verify production codes against known analytical solutions, (2) verify new methods and algorithms that do not involve detailed collision physics, (3) compare Monte Carlo calculation methods with deterministic methods, and (4) teach fundamentals to students. In this work we describe 2 new tools for preparing the ACE cross-section files to be used by MCNP ® for these analytic test problems, simple a ce.pl and simple a ce m g.pl.

VITAMIN E: a multipurpose ENDF/B-V coupled neutron-gamma cross section library

International Nuclear Information System (INIS)

Barhen, J.; Cacuci, D.G.; Ford, W.E. III; Roussin, R.W.; Wagschal, J.J.; Weisbin, C.R.; White, J.E.; Wright, R.Q.

1979-01-01

The US Department of Energy Office of Fusion Energy and the Division of Reactor Research and Technology jointly sponsored the development of a coupled fine-group cross section library (VITAMIN-C). The experience gained in the generation, validation, and utilization of the VITAMIN-C library along with its broad range of applicability has led to the request for updating this data set using ENDF/B-V. Additional support in this regard has been provided by the Defense Nuclear Agency (DNA) and by EPRI in support of weapons analyses and light water reactor shielding and dosimetry problems, respectively. The rationale for developing the multipurpose ENDF/B-V-based VITAMIN-E library is presented, with special emphasis on new models used in the data generation algorithms. The library specifications and testing procedures are also discussed in detail. The distribution of the VITAMIN-E library is currently subject to the same restrictions as the distribution of the ENDF/B-V data. 2 tables

Development and testing of the VITAMIN-B7/BUGLE-B7 coupled neutron-gamma multigroup cross-section libraries

Energy Technology Data Exchange (ETDEWEB)

Risner, J.M.; Wiarda, D.; Miller, T.M.; Peplow, D.E.; Patton, B.W.; Dunn, M.E. [Oak Ridge National Laboratory, MS 6170, P.O. Box 2008, Oak Ridge, TN 37831-6170 (United States); Parks, B.T. [U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Mail Stop O10-B3, 11555 Rockville Pike, Rockville, MD 20852 (United States)

2011-07-01

The U.S. Nuclear Regulatory Commission's Regulatory Guide 1.190 states that calculational methods used to estimate reactor pressure vessel (RPV) fluence should use the latest version of the evaluated nuclear data file (ENDF). The VITAMIN-B6 fine-group library and BUGLE-96 broad-group library, which are widely used for RPV fluence calculations, were generated using ENDF/B-VI.3 data, which was the most current data when Regulatory Guide 1.190 was issued. We have developed new fine-group (VITAMIN-B7) and broad-group (BUGLE-B7) libraries based on ENDF/B-VII.0. These new libraries, which were processed using the AMPX code system, maintain the same group structures as the VITAMIN-B6 and BUGLE-96 libraries. Verification and validation of the new libraries were accomplished using diagnostic checks in AMPX, 'unit tests' for each element in VITAMIN-B7, and a diverse set of benchmark experiments including critical evaluations for fast and thermal systems, a set of experimental benchmarks that are used for SCALE regression tests, and three RPV fluence benchmarks. The benchmark evaluation results demonstrate that VITAMIN-B7 and BUGLE-B7 are appropriate for use in RPV fluence calculations and meet the calculational uncertainty criterion in Regulatory Guide 1.190. (authors)

Development and Testing of the VITAMIN-B7/BUGLE-B7 Coupled Neutron-Gamma Multigroup Cross-Section Libraries

International Nuclear Information System (INIS)

Risner, Joel M.; Wiarda, Dorothea; Miller, Thomas Martin; Peplow, Douglas E.; Patton, Bruce W.; Dunn, Michael E.; Parks, Benjamin T.

2011-01-01

The U.S. Nuclear Regulatory Commission's Regulatory Guide 1.190 states that calculational methods used to estimate reactor pressure vessel (RPV) fluence should use the latest version of the Evaluated Nuclear Data File (ENDF). The VITAMIN-B6 fine-group library and BUGLE-96 broad-group library, which are widely used for RPV fluence calculations, were generated using ENDF/B-VI data, which was the most current data when Regulatory Guide 1.190 was issued. We have developed new fine-group (VITAMIN-B7) and broad-group (BUGLE-B7) libraries based on ENDF/B-VII. These new libraries, which were processed using the AMPX code system, maintain the same group structures as the VITAMIN-B6 and BUGLE-96 libraries. Verification and validation of the new libraries was accomplished using diagnostic checks in AMPX, unit tests for each element in VITAMIN-B7, and a diverse set of benchmark experiments including critical evaluations for fast and thermal systems, a set of experimental benchmarks that are used for SCALE regression tests, and three RPV fluence benchmarks. The benchmark evaluation results demonstrate that VITAMIN-B7 and BUGLE-B7 are appropriate for use in LWR shielding applications, and meet the calculational uncertainty criterion in Regulatory Guide 1.190.

Validation of KENO V.a. and two cross-section libraries for criticality calculations of low-enriched uranium systems

International Nuclear Information System (INIS)

Easter, M.E.

1985-07-01

The SCALE code system, utilizing the Monte Carlo computer code KENO V.a, was employed to calculate 37 critical experiments. The critical assemblies had 235 U enrichments of 5% or less and cover a variety of geometries and materials. Values of k/sub eff/ were calculated using two different results using either of the cross-section libraries. The 16-energy-group Hansen-Roach and the 27-energy-group ENDF/B-IV cross-section libraries, available in SCALE, were used in this validation study, and both give good results for the experiments considered. It is concluded that the code and cross sections are adequate for low-enriched uranium systems and that reliable criticality safety calculations can be made for such systems provided the limits of validated applicability are not exceeded

An energy recondensation method using the discrete generalized multigroup energy expansion theory

International Nuclear Information System (INIS)

Zhu Lei; Forget, Benoit

2011-01-01

Highlights: → Discrete-generalized multigroup method was implemented as a recondensation scheme. → Coarse group cross-sections were recondensed from core-level solution. → Neighboring effect of reflector and MOX bundle was improved. → Methodology was shown to be fully consistent when a flat angular flux approximation is used. - Abstract: In this paper, the discrete generalized multigroup (DGM) method was used to recondense the coarse group cross-sections using the core level solution, thus providing a correction for neighboring effect found at the core level. This approach was tested using a discrete ordinates implementation in both 1-D and 2-D. Results indicate that 2 or 3 iterations can substantially improve the flux and fission density errors associated with strong interfacial spectral changes as found in the presence of strong absorbers, reflector of mixed-oxide fuel. The methodology is also proven to be fully consistent with the multigroup methodology as long as a flat-flux approximation is used spatially.

Recommended activation detector cross sections (RNDL-82)

International Nuclear Information System (INIS)

Bondars, Kh.Ya.; Lapenas, A.A.

1984-01-01

The results of the comparison between measured and calculated average cross sections in 5 benchmark experiments are presented. Calculations have been based on the data from 10 libraries of evaluated cross sections. The recommended library (RNDL-82) of the activation detector cross sections has been created on the basis of the comparison. RNDL-82, including 26 reactions, and the basic characteristics of the detectors are presented. (author)

RADHEAT-V3, a code system for generating coupled neutron and gamma-ray group constants and analyzing radiation transport

International Nuclear Information System (INIS)

Koyama, Kinji; Taji, Yukichi; Miyasaka, Shun-ichi; Minami, Kazuyoshi.

1977-07-01

The modular code system RADHEAT is for producing coupled multigroup neutron and gamma-ray cross section sets, analyzing the neutron and gamma-ray transport, and calculating the energy deposition and atomic displacements due to these radiations in a nuclear reactor or shield. The basic neutron cross sections and secondary gamma-ray production data are taken from ENDF/B and POPOP4 libraries respectively. The system (1) generates multigroup neutron cross sections, energy deposition coefficients and atomic displacement factors due to neutron reactions, (2) generates multigroup gamma-ray cross sections and energy transfer coefficients, (3) generates secondary gamma-ray production cross sections, (4) combines these cross sections into the coupled set, (5) outputs and updates the multigroup cross section libraries in convenient formats for other transport codes, (6) analyzes the neutron and gamma-ray transport and calculates the energy deposition and the number density of atomic displacements in a medium, (7) collapses the cross sections to a broad-group structure, by option, using the weighting functions obtained by one-dimensional transport calculation, and (8) plots, by option, multigroup cross sections, and neutron and gamma-ray distributions. Definitions of the input data required in various options of the code system are also given. (auth.)

IAEA nuclear data for applications: Cross section standards and the reference input parameter library (RIPL)

International Nuclear Information System (INIS)

Capote Noy, Roberto; Nichols, Alan L.; Pronyaev, Vladimir G.

2003-01-01

develop a library of validated nuclear-model input parameters, referred to as the Reference Input Parameter Library (RIPL). The first stage of this work was initiated in 1994 and the second step began in 1998, both as IAEA CRPs. A consistent library of recommended nuclear theoretical input parameters is now available (RIPL-2) that includes a large amount of theoretical information suitable for nuclear reaction calculations, along with a number of computer codes for parameter retrieval and related calculations. A third further phase of this project has been recently initiated in order to extend the applicability of the RIPL library to cross sections for reactions on nuclei far from the line of stability, incident energies up to 200 MeV, and reactions induced by charged particles. (authors)

WIMSD5, Deterministic Multigroup Reactor Lattice Calculations

International Nuclear Information System (INIS)

2004-01-01

1 - Description of program or function: The Winfrith improved multigroup scheme (WIMS) is a general code for reactor lattice cell calculation on a wide range of reactor systems. In particular, the code will accept rod or plate fuel geometries in either regular arrays or in clusters and the energy group structure has been chosen primarily for thermal calculations. The basic library has been compiled with 14 fast groups, 13 resonance groups and 42 thermal groups, but the user is offered the choice of accurate solutions in many groups or rapid calculations in few groups. Temperature dependent thermal scattering matrices for a variety of scattering laws are included in the library for the principal moderators which include hydrogen, deuterium, graphite, beryllium and oxygen. WIMSD5 is a successor version of WIMS-D/4. 2 - Method of solution: The treatment of resonances is based on the use of equivalence theorems with a library of accurately evaluated resonance integrals for equivalent homogeneous systems at a variety of temperatures. The collision theory procedure gives accurate spectrum computations in the 69 groups of the library for the principal regions of the lattice using a simplified geometric representation of complicated lattice cells. The computed spectra are then used for the condensation of cross-sections to the number of groups selected for solution of the transport equation in detailed geometry. Solution of the transport equation is provided either by use of the Carlson DSN method or by collision probability methods. Leakage calculations including an allowance for streaming asymmetries may be made using either diffusion theory or the more elaborate B1-method. The output of the code provides Eigenvalues for the cases where a simple buckling mode is applicable or cell-averaged parameters for use in overall reactor calculations. Various reaction rate edits are provided for direct comparison with experimental measurements. 3 - Restrictions on the complexity of

ZZ FCXSEC, Coupled Cross-Section Library for Shielding from VITAMIN-C in AMPX, ANISN Format

International Nuclear Information System (INIS)

1985-01-01

1 - Description of problem or function: Format: (a) and (b) AMPX, (c) and (d) ANISN; Number of groups: (a) Fine-group 171 neutron and 36 gamma-ray; (b) Broad-group 22 neutron and 21 gamma-ray; (c) Broad-group microscopic (22n-21 gamma); (d) Broad-group macroscopic; Nuclides: Mixtures: H 2 O, Borated water, Concrete, D 2 O, Lithium hydride, Boral, Dry air, Nitric acid, Uranium dioxide, S 3 0 4 , UF 6 TBP in dodecane, Sm 2 O 3 , Eu 2 O 3 , Gd 2 O 3 , Gd(NO 3 ) 3 in water, WB2, Spen fuel oxide, Thorium oxide, Uranium metal, Silver zeolite. Individual materials: C, Na, Al, Fe, Zircaloy, Cd Nb, Mo, Pb, Be, Ti, V, Mn, Co, Cu, Sn, Ta. Origin: VITAMIN-C; Weighting spectrum: From 1.1109+5 eV to 1.7333+7 eV → 239 Pu thermal fission; From 4.1399-1 eV to 1.1109+5 eV → 1/E; From 1.0000-5 eV to 4.1399-1 eV → Maxwellian. FSXSEC is a collection of cross section libraries to be used for nuclear fuel cycle shielding calculations, generated from the pseudo-composition-independent VITAMIN-C cross section library: (a) A composition-dependent self-shielded fine-group library with 171 neutron groups and 36 gamma groups, and a broad-group library with 22 neutron and 21 gamma groups for AMPX. (b) A broad-group microscopic and a broad-group macroscopic library in ANISN format. 2 - Method of solution: To generate library (a), AMPX modules BONAMI, CHOX, and MALOCS were used. To generate library (b), AMPX modules NITAWL and AXMIX were used

CSRL-V: processed ENDF/B-V 227-neutron-group and pointwise cross-section libraries for criticality safety, reactor, and shielding studies

International Nuclear Information System (INIS)

Ford, W.E. III; Diggs, B.R.; Petrie, L.M.; Webster, C.C.; Westfall, R.M.

1982-01-01

A P 3 227-neutron-group cross-section library has been processed for the subsequent generation of problem-dependent fine- or broad-group cross sections for a broad range of applications, including shipping cask calculations, general criticality safety analyses, and reactor core and shielding analyses. The energy group structure covers the range 10 -5 eV - 20 MeV, including 79 thermal groups below 3 eV. The 129-material library includes processed data for all materials in the ENDF/B-V General Purpose File, several data sets prepared from LENDL data, hydrogen with water- and polyethyelene-bound thermal kernels, deuterium with C 2 O-bound thermal kernels, carbon with a graphite thermal kernel, a special 1/V data set, and a dose factor data set. The library, which is in AMPX master format, is designated CSRL-V (Criticality Safety Reference Library based on ENDF/B-V data). Also included in CSRL-V is a pointwise total, fission, elastic scattering, and (n,γ) cross-section library containing data sets for all ENDF/B-V resonance materials. Data in the pointwise library were processed with the infinite dilute approximation at a temperature of 296 0 K

Implementing of AMPX-II system for a univac computer neutron cross-section libraries

International Nuclear Information System (INIS)

Sancho, J.; Verdu, G.; Serradell, V.

1984-01-01

The AMPX-II system, developed at ORNL, is constituted by a modular set of computer programs, for generation and handling of several nuclear data libraries. The processing starts from ENDF/B library. Along this paper, we refer mainly to the modules related with neutron cross section libraries: master, working and weighted. These modules have been implemented recently for a UNIVAC 1100/60 computer in the Universidad Politecnica de Valencia (Spain). In order to run the programs in that machine it has been necessary to introduce a number of modifications into their programing structure. The main difficulties found in this work and the need of verification for the new versions are also pointed out. We also refer to the results obtained from the execution of a set of little sample problems. (author)

Energy meshing techniques for processing ENDF/B-VI cross sections using the AMPX code system

International Nuclear Information System (INIS)

Dunn, M.E.; Greene, N.M.; Leal, L.C.

1999-01-01

Modern techniques for the establishment of criticality safety for fissile systems invariably require the use of neutronic transport codes with applicable cross-section data. Accurate cross-section data are essential for solving the Boltzmann Transport Equation for fissile systems. In the absence of applicable critical experimental data, the use of independent calculational methods is crucial for the establishment of subcritical limits. Moreover, there are various independent modern transport codes available to the criticality safety analyst (e.g., KENO V.a., MCNP, and MONK). In contrast, there is currently only one complete software package that processes data from the Version 6 format of the Evaluated Nuclear Data File (ENDF) to a format useable by criticality safety codes. To facilitate independent cross-section processing, Oak Ridge National Laboratory (ORNL) is upgrading the AMPX code system to enable independent processing of Version 6 formats using state-of-the-art procedures. The AMPX code system has been in continuous use at ORNL since the early 1970s and is the premier processor for providing multigroup cross sections for criticality safety analysis codes. Within the AMPX system, the module POLIDENT is used to access the resonance parameters in File 2 of an ENDF/B library, generate point cross-section data, and combine the cross sections with File 3 point data. At the heart of any point cross-section processing code is the generation of a suitable energy mesh for representing the data. The purpose of this work is to facilitate the AMPX upgrade through the development of a new and innovative energy meshing technique for processing point cross-section data

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

Nuclear data libraries for Tripoli-3.5 code

International Nuclear Information System (INIS)

Vergnaud, Th.

2001-01-01

The TRIPOLI-3 code uses multigroup nuclear data libraries generated using the NJOY-THEMIS suite of modules: for neutrons, they are produced from the ENDF/B-VI evaluations and cover the range between 20 MeV and 10 -5 eV, either in 315 groups and for one temperature, or in 3209 groups and for five temperatures; for gamma-rays, they are from JEF2 and are processed in groups between 14 MeV and keV. The probability tables used for the neutron transport calculations have been derived from the ENDF/B-VI evaluations using the CALENDF code. Cross sections for gamma production by neutron interaction (fission, capture or inelastic scattering) have been derived from ENDF/B-VI in 315 neutron groups and 75 gamma groups. The code also uses two response function libraries: for neutrons; based on several sources, in particular the dosimetry libraries IRDF/85 and IRDF/90; for gamma-rays it is based on the JEF2 evaluation and contains the kerma factors for all the elements and cross sections for all interactions. (author)

Development and benchmark of high energy continuous-energy neutron cross Section library HENDL-ADS/MC

International Nuclear Information System (INIS)

Chen Chong; Wang Minghuang; Zou Jun; Xu Dezheng; Zeng Qin

2012-01-01

The ADS (accelerator driven sub-critical system) has great energy spans, complex energy spectrum structures and strong physical effects. Hence, the existing nuclear data libraries can't fully meet the needs of nuclear analysis in ADS. In order to do nuclear analysis for ADS system, a point-wise data library HENDL-ADS/MC (hybrid evaluated nuclear data library) was produced by FDS team. Meanwhile, to test the availability and reliability of the HENDL-ADS/MC data library, a series of shielding and critical safety benchmarks were performed. To validate and qualify the reliability of the high-energy cross section for HENDL-ADS/MC library further, a series of high neutronics integral experiments have been performed. The testing results confirm the accuracy and reliability of HENDL-ADS/MC. (authors)

Calculation of atom displacement cross section for structure material

International Nuclear Information System (INIS)

Liu Ping; Xu Yiping

2015-01-01

The neutron radiation damage in material is an important consideration of the reactor design. The radiation damage of materials mainly comes from atom displacements of crystal structure materials. The reaction cross sections of charged particles, cross sections of displacements per atom (DPA) and KERMA are the basis of radiation damage calculation. In order to study the differences of DPA cross sections with different codes and different evaluated nuclear data libraries, the DPA cross sections for structure materials were calculated with UNF and NJOY codes, and the comparisons of results were given. The DPA cross sections from different evaluated nuclear data libraries were compared. And the comparison of DPA cross sections between NJOY and Monte Carlo codes was also done. The results show that the differences among these evaluated nuclear data libraries exist. (authors)

Evaluation of cross sections of Th-232 and U-233

International Nuclear Information System (INIS)

Dias, A.M.

1978-01-01

The cross sections in multigroups of Th-232 and U-233 are evaluated by comparison of theoretical results and experimental data obtained through experiments on the fast reactors IBR-I, EBR-II, BR-I and AETR. The deviation between calculated values and experimental results is about 10%. They are therefore satisfatory for neutronic calculations [pt

ROSFOND based heating-damage cross sections sub-library: Preliminary uncertainty assessment

International Nuclear Information System (INIS)

Sinitsa, V.V.

2016-01-01

The accuracy of radiation damage calculations for the most important LWR component, the reactor pressure vessel (RPV), directly linked with the RPV End-of-Life (EoL) prediction which is in its turn connected with fundamental nuclear safety aspects and relevant economic impacts. In this connection, for nearly ten years the ENEA-Bologna Nuclear Data Group conducts the nuclear data processing and validation activities addressed to update the specialized broad-group coupled neutron/photon working cross section libraries for shielding and radiation damage calculations through NJOY and Bologna revised version of SCAMPI data processing systems. A number of working group-wise data libraries has been prepared and transferred to the ENEA Data Bank for dissemination. Several years ago the NRC ”Kurchatov Institute” has reset the GRUCON project, originally designed to provide group constants for fast nuclear reactor calculations [12], with aim to expand its application area and to use in the WWER safety tasks, in particular, in the RPV radiation damage analyses. By means of updated GRUCON and NJOY-99 processing codes, and calculation procedure, developed in the NDG of ENEA Bologna, a sample of kerma&damage energy point-wise data sub-libraries from different evaluated data libraries has been generated. On the base of this sample, the quantitative assessment of kerma/dpa data precision in the RPV calculations is obtained

Development of automatic cross section compilation system for MCNP

International Nuclear Information System (INIS)

Maekawa, Fujio; Sakurai, Kiyoshi

1999-01-01

A development of a code system to automatically convert cross-sections for MCNP is in progress. The NJOY code is, in general, used to convert the data compiled in the ENDF format (Evaluated Nuclear Data Files by BNL) into the cross-section libraries required by various reactor physics codes. While the cross-section library: FSXLIB-J3R2 was already converted from the JENDL-3.2 version of Japanese Evaluated Nuclear Data Library for a continuous energy Monte Carlo code MCNP, the library keeps only the cross-sections at room temperature (300 K). According to the users requirements which want to have cross-sections at higher temperature, say 600 K or 900 K, a code system named 'autonj' is under development to provide a set of cross-section library of arbitrary temperature for the MCNP code. This system can accept any of data formats adopted JENDL that may not be treated by NJOY code. The input preparation that is repeatedly required at every nuclide on NJOY execution is greatly reduced by permitting the conversion process of as many nuclides as the user wants in one execution. A few MCNP runs were achieved for verification purpose by using two libraries FSXLIB-J3R2 and the output of autonj'. The almost identical MCNP results within the statistical errors show the 'autonj' output library is correct. In FY 1998, the system will be completed, and in FY 1999, the user's manual will be published. (K. Tsuchihashi)

AUS module MIRANDA - a data preparation code based on multiregion resonance theory

International Nuclear Information System (INIS)

Robinson, G.S.

1977-07-01

MIRANDA is a data preparation module of the AUS reactor neutronics scheme and is used to prepare multigroup cross-section data which are pertinent to a particular reactor system from a general purpose multigroup library of cross sections. The cross-section library has been prepared from ENDF/B and includes temperature dependent data and resonance cross sections represented by subgroup parameters. The MIRANDA module includes a multiregion resonance calculation in slab, cylinder or cluster geometry, a homogeneous Bsub(L) flux solution, and a group condensation facility. Interaction with other AUS modules, particularly for burnup calculations, is provided. (Author)

Shielding Factor Method for producing effective cross sections: MINX/SPHINX and the CCCC interface system

International Nuclear Information System (INIS)

MacFarlane, R.E.; Weisbin, C.R.; Paik, N.C.

1978-01-01

The Shielding Factor Method (SFM) is an economical designer-oriented method for producing the coarse-group space and energy self-shielded cross sections needed for reactor-core analysis. Extensive experience with the ETOX/1DX and ENDRUN/TDOWN systems has made the SFM the method of choice for most US fast-reactor design activities. The MINX/SPHINX system was designed to expand upon the capabilities of the older SFM codes and to incorporate the new standard interfaces for fast-reactor cross sections specified by the Committee for Computer Code Coordination (CCCC). MINX is the cross-section processor. It generates multigroup cross sections, shielding factors, and group-to-group transfer matriccs from ENDF/B-IV and writes them out as CCCC ISOTXS and BRKOXS files. It features detailed pointwise resonance reconstruction, accurate Doppler broadening, and an efficient treatment of anisotropic scattering. SPHINX is the space-and-energy shielding code. It uses specific mixture and geometry information together with equivalence principles to construct shielded macroscopic multigroup cross sections in as many as 240 groups. It then makes a flux calculation by diffusion or transport methods and collapses to an appropriate set of cell-averaged coarse-group effective cross sections. The integration of MINX and SPHINX with the CCCC interface system provides an efficient, accurate, and convenient system for producing effective cross sections for use in fast-reactor problems. The system has also proved useful in shielding and CTR applications. 3 figures, 4 tables

VITAMIN-J/COVA/EFF-3 cross-section covariance matrix library and its use to analyse benchmark experiments in sinbad database

International Nuclear Information System (INIS)

Kodeli, Ivan-Alexander

2005-01-01

The new cross-section covariance matrix library ZZ-VITAMIN-J/COVA/EFF3 intended to simplify and encourage sensitivity and uncertainty analysis was prepared and is available from the NEA Data Bank. The library is organised in a ready-to-use form including both the covariance matrix data as well as processing tools:-Cross-section covariance matrices from the EFF-3 evaluation for five materials: 9 Be, 28 Si, 56 Fe, 58 Ni and 60 Ni. Other data will be included when available. -FORTRAN program ANGELO-2 to extrapolate/interpolate the covariance matrices to a users' defined energy group structure. -FORTRAN program LAMBDA to verify the mathematical properties of the covariance matrices, like symmetry, positive definiteness, etc. The preparation, testing and use of the covariance matrix library are presented. The uncertainties based on the cross-section covariance data were compared with those based on other evaluations, like ENDF/B-VI. The collapsing procedure used in the ANGELO-2 code was compared and validated with the one used in the NJOY system

The Suppression of Energy Discretization Errors in Multigroup Transport Calculations

International Nuclear Information System (INIS)

Larsen, Edward

2013-01-01

The Objective of this project is to develop, implement, and test new deterministric methods to solve, as efficiently as possible, multigroup neutron transport problems having an extremely large number of groups. Our approach was to (i) use the standard CMFD method to 'coarsen' the space-angle grid, yielding a multigroup diffusion equation, and (ii) use a new multigrid-in-space-and-energy technique to efficiently solve the multigroup diffusion problem. The overall strategy of (i) how to coarsen the spatial an energy grids, and (ii) how to navigate through the various grids, has the goal of minimizing the overall computational effort. This approach yields not only the fine-grid solution, but also coarse-group flux-weighted cross sections that can be used for other related problems.

Comparison of CASMO and NESSEL few group cross section libraries and their usage in DYN3D

International Nuclear Information System (INIS)

Kuchin, A.; Ovdiyenko, Y.; Loetsch, T.

2007-01-01

This work presents comparative analysis of two group diffusion cross section libraries which were generated by NESSEL-4 and CASMO-4 lattice codes. Diffusion parameters were calculated for VVER-1000 fuel assemblies with stainless steel spacing grids and guiding tubes. These cross section sets were introduced into reactor core code DYN3D and tested on the base of real reactor core states. In this case operation data of the first three fuel cycles of 6-th unit of Zaporizhzhya NPP were used

Testing of the IRDF-90 cross-section library in benchmark neutron spectra

International Nuclear Information System (INIS)

Nolthenius, H.J.; Zsolnay, E.M.; Szondi, E.J.

1993-09-01

The new version of the International Reactor Dosimetry File IRDF-90 (called ''Version April 1993'') has been tested by calculation of average cross-sections and their uncertainties in a coarse three energy group structure and by neutron spectrum adjustments in reference neutron spectra. This paper presents the results obtained and compares them with the corresponding ones of the old IRDF-85 and with the data of the Nuclear Data Guide for Reactor Neutron Metrology. The applicability of the new library in the field of neutron metrology is discussed. (orig.)

Generation of SCALE 6 Input Data File for Cross Section Library of PWR Spent Fuel

International Nuclear Information System (INIS)

Jeong, Chang Joon; Cho, Dong Keun

2010-11-01

In order to obtain the cross section libraries of the Korean Pressurized water reactor (PWR) spent fuel (SF), SCALE 6 code input files have been generated. The PWR fuel data were obtained from the nuclear design report (NDR) of the current operating PWRs. The input file were prepared for 16 fuel types such as 4 types of Westinghouse 14x14, 3 types of OPR-1000 16x16, 4 types of Westinghouse 16x16, and 6 types of Westinghouse 17x17. For each fuel type, 5 kinds of fuel enrichments have been considered such as 1.5, 2.0 ,3.0, 4.0 and 5.0 wt%. In the SCALE 6 calculation, a ENDF-V 44 group was used. The 25 burnup step until 72000 MWD/T was used. A 1/4 symmetry model was used for 16x16 and 17x17 fuel assembly, and 1/2 symmetry model was used for 14x14 fuel assembly The generated cross section libraries will be used for the source-term analysis of the PWR SF

Sensitivity of LWR fuel cycle costs to uncertainties in detailed thermal cross sections

International Nuclear Information System (INIS)

Ryskamp, J.M.; Becker, M.; Harris, D.R.

1979-01-01

Cross sections averaged over the thermal energy (< 1 or 2 eV) group have been shown to have an important economic role for light-water reactors. Cost implications of thermal cross section uncertainties at the few-group level were reported earlier. When it has been determined that costs are sensitive to a specific thermal-group cross section, it becomes desirable to determine how specific energy-dependent cross sections influence fuel cycle costs. Multigroup cross-section sensitivity coefficients vary with fuel exposure. By changing the shape of a cross section displayed on a view-tube through an interactive graphics system, one can compute the change in few-group cross section using the exposure dependent sensitivity coefficients. With the changed exposure dependent few-group cross section, a new fuel cycle cost is computed by a sequence of batch depletion, core analysis, and fuel batch cost code modules. Fuel cycle costs are generally most sensitive to cross section uncertainties near the peak of the hardened Maxwellian flux

Resonance self-shielding method using resonance interference factor library for practical lattice physics computations of LWRs

International Nuclear Information System (INIS)

Choi, Sooyoung; Khassenov, Azamat; Lee, Deokjung

2016-01-01

This paper presents a new method of resonance interference effect treatment using resonance interference factor for high fidelity analysis of light water reactors (LWRs). Although there have been significant improvements in the lattice physics calculations over the several decades, there exist still relatively large errors in the resonance interference treatment, in the order of ∼300 pcm in the reactivity prediction of LWRs. In the newly developed method, the impact of resonance interference to the multi-group cross-sections has been quantified and tabulated in a library which can be used in lattice physics calculation as adjustment factors of multi-group cross-sections. The verification of the new method has been performed with Mosteller benchmark, UO_2 and MOX pin-cell depletion problems, and a 17×17 fuel assembly loaded with gadolinia burnable poison, and significant improvements were demonstrated in the accuracy of reactivity and pin power predictions, with reactivity errors down to the order of ∼100 pcm. (author)

Design and producing of fine-group cross section library HENDL3.0/FG for subcritical system

International Nuclear Information System (INIS)

Zou, J.; Zeng, Q.; Xu, D.; Hu, L.; Long, P.

2012-01-01

To improve the accuracy of the neutron analyses for subcritical system with thermal fission blanket, a coupled neutron and photon (315 n + 42γ) fine-group cross section library HENDL3.0/FG based on ENDF/B-VII, JEFF3.1 and JENDL3.3 was produced by FDS team. In order to test the availability and reliability of the HENDL3.0/FG data library, shielding and critical safety benchmarks were performed with VisualBUS code. The testing results indicated that the discrepancy between calculation and experimental values of nuclear parameters fell in a reasonable range. It showed that the nuclear data library had accuracy and availability. (authors)

Neutronic calculations in heavy water moderated multiplying media using GGC-3 library nuclear data

International Nuclear Information System (INIS)

Boado, H.J.; Gho, C.J.; Abbate, M.J.

1981-01-01

Differences in obtaining transference matrices between GGC-3 code and the system to produce multigroup cross sections using GGC-3 library, recently implemented at the Neutrons and Reactors Division, have been analized. Neutronic calculations in multiplicative systems containing heavy water have been made using both methods. From the obtained results, it is concluded that the new method is more appropriate to deal with systems including moderators other than light water. (author) [es

ZZ ENDL82, Evaluated Charged Particle, Neutron, Photon Cross-Section Library

International Nuclear Information System (INIS)

2001-01-01

Description of program or function: - Format: Described in the manual; - Number of groups: (energies between 100 eV and 100 MeV); - Nuclides: 94 (Z 1 to 99); - Origin: LLNL Evaluated Nuclear Data Library. ENDL82 is a collection of evaluated data for neutron-induced reactions, photon interactions with matter, and charged-particle-induced reactions. It is maintained in a computer-oriented system. All interpolable quantities for neutron-induced reactions are presented so that linear interpolation between successive entries yields values that are consistent with stated experimental errors, where experiments exist, or that adhere to an assumed law, such as 1/v energy dependence, within a small fraction (typically 1%). In the case of an assumed energy-dependence law for cross sections, this is accomplished by creating a large number of (energy, cross section) pairs by computer and subsequently thinning the points to a specified accuracy, using the subroutine THINER. All angular distributions are differential probabilities normalized to an integral of unity over the cosine of the scattering angle. All energy distributions of secondary particles are presented as normalized Legendre polynomial representations. The linear interpolation will construct an acceptable angular distribution at an intermediate energy

FENDL-3.0: Processing the Evaluated Nuclear Data Library for Fusion Applications

International Nuclear Information System (INIS)

Lopez Aldama, D.; Noy, R. Capote

2011-12-01

A description of the work undertaken towards the development of a new version of the neutron-induced part of the Fusion Evaluated Nuclear Data Library (FENDL) for applications is summarized. The main issues related to the selection and processing of evaluated nuclear data files using the NJOY-99 and PREPRO-2010 processing systems are described. The new version of FENDL for applications, termed FENDL-3.0, includes the evaluated nuclear data files in ENDF-6 format, the continuous-energy cross section files in ACE format for the MCNP family of Monte Carlo codes and the multi-group data library in MATXS format for deterministic transport calculations up to 55 MeV for 180 isotopes. Further, additional data are supplied in GENDF format for sensitivity studies. The library is freely available from the Nuclear Data Section at the International Atomic Energy Agency. (author)

Library of neutron cross sections of the Thermos code; Biblioteca de secciones eficaces de neutrones del codigo Thermos

Energy Technology Data Exchange (ETDEWEB)

Alonso V, G; Hernandez L, H [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

1991-10-15

The present work is the complement of the IT.SN/DFR-017 report in which the structure and the generation of the library of the Thermos code is described. In this report the comparison among the values of the cross sections that has the current library of the Thermos code and those generated by means of the ENDF-B/NJOY it is shown. (Author)

Propagation of cross section uncertainties in combined Monte Carlo neutronics and burnup calculations

Energy Technology Data Exchange (ETDEWEB)

Kuijper, J.C.; Oppe, J.; Klein Meulekamp, R.; Koning, H. [NRG - Fuels, Actinides and Isotopes group, Petten (Netherlands)

2005-07-01

Some years ago a methodology was developed at NRG for the calculation of 'density-to-density' and 'one-group cross section-to-density' sensitivity matrices and covariance matrices for final nuclide densities for burnup schemes consisting of multiple sets of flux/spectrum and burnup calculations. The applicability of the methodology was then demonstrated by calculations of BR3 MOX pin irradiation experiments employing multi-group cross section uncertainty data from the EAF4 data library. A recent development is the extension of this methodology to enable its application in combination with the OCTOPUS-MCNP-FISPACT/ORIGEN Monte Carlo burnup scheme. This required some extensions to the sensitivity matrix calculation tool CASEMATE. The extended methodology was applied on the 'HTR Plutonium Cell Burnup Benchmark' to calculate the uncertainties (covariances) in the final densities, as far as these uncertainties are caused by uncertainties in cross sections. Up to 600 MWd/kg these uncertainties are larger than the differences between the code systems. However, it should be kept in mind that the calculated uncertainties are based on EAF4 uncertainty data. It is not exactly clear on beforehand what a proper set of associated (MCNP) cross sections and covariances would yield in terms of final uncertainties in calculated densities. This will be investigated, by the same formalism, once these data becomes available. It should be noted that the studies performed up till the present date are mainly concerned with the influence of uncertainties in cross sections. The influence of uncertainties in the decay constants, although included in the formalism, is not considered further. Also the influence of other uncertainties (such as -geometrical- modelling approximations) has been left out of consideration for the time being. (authors)

Propagation of cross section uncertainties in combined Monte Carlo neutronics and burnup calculations

International Nuclear Information System (INIS)

Kuijper, J.C.; Oppe, J.; Klein Meulekamp, R.; Koning, H.

2005-01-01

Some years ago a methodology was developed at NRG for the calculation of 'density-to-density' and 'one-group cross section-to-density' sensitivity matrices and covariance matrices for final nuclide densities for burnup schemes consisting of multiple sets of flux/spectrum and burnup calculations. The applicability of the methodology was then demonstrated by calculations of BR3 MOX pin irradiation experiments employing multi-group cross section uncertainty data from the EAF4 data library. A recent development is the extension of this methodology to enable its application in combination with the OCTOPUS-MCNP-FISPACT/ORIGEN Monte Carlo burnup scheme. This required some extensions to the sensitivity matrix calculation tool CASEMATE. The extended methodology was applied on the 'HTR Plutonium Cell Burnup Benchmark' to calculate the uncertainties (covariances) in the final densities, as far as these uncertainties are caused by uncertainties in cross sections. Up to 600 MWd/kg these uncertainties are larger than the differences between the code systems. However, it should be kept in mind that the calculated uncertainties are based on EAF4 uncertainty data. It is not exactly clear on beforehand what a proper set of associated (MCNP) cross sections and covariances would yield in terms of final uncertainties in calculated densities. This will be investigated, by the same formalism, once these data becomes available. It should be noted that the studies performed up till the present date are mainly concerned with the influence of uncertainties in cross sections. The influence of uncertainties in the decay constants, although included in the formalism, is not considered further. Also the influence of other uncertainties (such as -geometrical- modelling approximations) has been left out of consideration for the time being. (authors)

Generation and testing of the shielding data library EURLIB for fission and fusion technology

International Nuclear Information System (INIS)

Caglioti, E.; Hehn, G.; Herrnberger, V.; Mattes, M.; Nicks, R.; Penkuhn, H.

1977-01-01

For the common field of core physics and shielding, the CSEWG group structure of 239 fast neutron groups had been proposed, of which the 100 neutron groups of the EURLIB Library is a sub-set for shielding. This standard group Library EURLIB had been initiated by the NEA-specialist group on shielding benchmarks in 1974. The wide acceptance of the Library for interpretation of benchmarks in the NEA program represents an important step forward in the standardization of group data which is the basic requirement for a useful collaboration. On the other side the interpretation of a series of different benchmark experiments with the EURLIB Library provides the best check of the cross section data for neutron and gamma-rays showing the needs for further improvements. The paper describes the joint work of IKE, Stuttgart and EURATOM, Ispra in generating multigroup libraries for neutron and gamma-rays. Special effort has been devoted to improve the flux weighting for both types of radiation and proper treatment of thermal neutrons. The coupled multigroup Library of 100 neutron and 20 gamma groups is collapsed into few group structures for typical designs of LWR, LMFBR, gas cooled and thermonuclear reactors. The work for optimal few group representation is done in cooperation with EIR, Wurenlingen. The testing of the EURLIB Library is a common effort of several institutions participating in the NEA shielding benchmark program

Proposal to extend CSEWG neutron and photon multigroup structures for wider applications

International Nuclear Information System (INIS)

LaBauve, R.J.; Wilson, W.B.

1976-02-01

The 239-group neutron multigroup structure recommended by the Codes and Formats Subcommittee of the cross section evaluation working group (CSEWG) for use in LMFBR design is not well suited for application in certain other areas, particularly thermal reactor design. This report describes a proposal for a neutron group structure consisting of 347 groups, which is an extension of the CSEWG group structure into the thermal range, and also includes more detail in other energy ranges important in LWR, HTGR, GCFR, and CTR design. Similarly, a proposed extension of the CSEWG 94-group photon multigroup structure to 103 groups is described. A subset of the neutron multigroup structure, consisting of 154 groups and for use in power reactor studies, is also presented

Determination of Intermediate Resonance Parameter with RMET21 for nTRACER

International Nuclear Information System (INIS)

Sohail, Muhammad; Kim, Myung Hyun

2012-01-01

Ray Tracing based code nTRACER is being developed in Seoul National University that has the capability of 3-dimensional whole core neutron transport calculation. As a part of development of multi-group neutron cross section library for nTRACER, the current work is intended to accurately determine intermediate resonance parameters. Beside the systematic calculation of subgroup parameters for resonance self shielding calculation, intermediate resonance parameters itself can be as important as the multi-group neutron cross section in the library and its overall accuracy. In this paper lambda factors were computed using RMET21 from ENDF/B-VII.1 for nTRACER to investigate its dependence on temperature and background cross section and replaced with lambda factors from HELIOS multi-group library. The procedure used for determining the intermediate resonance parameter for the isotope under study is introduced in the next section. Oxygen being one of the primary nuclide in PWR fuel has been selected for intermediate resonance parameters calculation

Development of Indian cross section data files for Th-232 and U-233 and integral validation studies

International Nuclear Information System (INIS)

Ganesan, S.

1988-01-01

This paper presents an overview of the tasks performed towards the development of Indian cross section data files for Th-232 and U-233. Discrepancies in various neutron induced reaction cross sections in various available evaluated data files have been obtained by processing the basic data into multigroup form and intercomparison of the latter. Interesting results of integral validation studies for capture, fission and (n,2n) cross sections for Th-232 by analyses of selected integral measurements are presented. In the resonance range, energy regions where significant differences in the calculated self-shielding factors for Th-232 occur have been identified by a comparison of self-shielded multigroup cross sections derived from two recent evaluated data files, viz., ENDF/B-V (Rev.2) and JENDL-2, for several dilutions and temperatures. For U-233, the three different basic data files ENDF/B-IV, JENDL-2 and ENDL-84 were intercompared. Interesting observations on the predictional capability of these files for the criticality of the spherical metal U-233 system are given. The current status of Indian data file is presented. (author) 62 ref

ZZ FENDL-2, Evaluated Nuclear Data Library for Fusion Neutronics Applications

International Nuclear Information System (INIS)

2005-01-01

Description: FENDL: Fusion Evaluated Nuclear Data Library. Materials/nuclides: H 1 , H 2 , H 3 , He 3 , He 4 , Li 6 , Li 7 , Be 9 , B 10 , B 11 , C 12 , N 14 , N 15 , O 16 , F 19 , Na 23 , Mg nat , Al 27 , Si 28 , Si 29 , Si 30 , P 31 , S nat , Cl 35 , Cl 37 , K nat , Ca nat , Ti 46 , Ti 47 , Ti 48 , Ti 49 , Ti 50 , V nat , Cr 50 , Cr 52 , Cr 53 , Cr 54 , Mn 55 , Fe 54 , Fe 56 , Fe 57 , Fe 58 , Co 59 , Ni 58 , Ni 60 , Ni 61 , Ni 62 , Ni 64 , Cu 63 , Cu 65 , Ga nat , Zr nat , Nb 93 , Mo 92 , Mo 94 , Mo 95 , Mo 96 , Mo 97 , Mo 98 , Mo 100 , Sn nat , Ta 181 , W 182 , W 183 , W 184 , W 186 , Au 197 , Pb 206 , Pb 207 , Pb 208 , Bi 209 . Photo-atomic data. IAEA1364/02: FENDL version 2.0 consists of the following sub-libraries: - ACTIVATION (FENDL/A-2.0)- neutron activation cross sections for 13006 reactions on 739 targets ranging from 1-H up to 248-Cm at incident energies up to 20 MeV. Pointwise and processed data in different formats are included. Plots are available. - DECAY (FENDL/D-2.0) - decay properties (decay type, decay energy, and half life) for 1867 nuclides and isomers. FENDL/D-2.0 sub-library is complementary to the activation sub-library. Pointwise and processed data are included. - DOSIMETRY (FENDL/DS-2.0) - neutron cross sections to be used for reactor neutron dosimetry by foil activation, radiation damage cross-sections, and benchmark neutron spectra. This sub-library is identical to the International Reactor Dosimetry File (IRDF-90). Pointwise and processed data are included. - FUSION (FENDL/C-2.0) - charged-particle cross sections for the following fusion reactions: 1-H 2 (d,n)2-He 3 , 1-H 2 (d,p)1-H 3 , 2-He 3 (d,p)2-He 4 , 1-H-3(t,2n)2-He 4 , and 1-H 3 (d,n)2-He 4 . Pointwise and processed data are included. - TRANSPORT - validated basic nuclear data (neutron-nucleus interaction including photon production, and photon-atom interaction cross sections) for 57 nuclides relevant for fusion. In addition to the pointwise data (FENDL/E-2.0), the sub-library

Cross section sensitivity study for fusion blankets incorporating lead neutron multiplier

International Nuclear Information System (INIS)

Pelloni, S.; Cheng, E.T.

1983-01-01

In the recent European INTOR design, lead has been considered for incorporation in the blanket as either an explicit or implicit neutron multiplier. The blanket employs either Li 2 SiO 3 or Li 17 Pb 83 as tritium breeding material. Nucleonic analysis was performed for this blanket using the DLC37 and DLC41 cross section libraries. The reaction rates were estimated using the reaction cross sections provided with both libraries. In addition to that, they were estimated using the MACKLIB-IV response library. The calculated tritium breeding ratio was found to be 5% less and 15% more in the calculations with DLC41 and DLC41 plus MACKLIB-IV libraries, respectively, than in the calculation with the DLC37 library. The Fe, Pb, and Li cross sections given by the ENDF/B-IV and V were reviewed. A sensitivity study of these cross section uncertainties shows that the tritium breeding ratio is relatively insensitive to the above mentioned partial cross sections. The calculated tritium breeding ratio can be known within +-2%. (Auth.)

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

Proposal to extend CSEWG neutron and photon multigroup structures for wider applications. [Tables

Energy Technology Data Exchange (ETDEWEB)

LaBauve, R.J.; Wilson, W.B.

1976-02-01

The 239-group neutron multigroup structure recommended by the Codes and Formats Subcommittee of the cross section evaluation working group (CSEWG) for use in LMFBR design is not well suited for application in certain other areas, particularly thermal reactor design. This report describes a proposal for a neutron group structure consisting of 347 groups, which is an extension of the CSEWG group structure into the thermal range, and also includes more detail in other energy ranges important in LWR, HTGR, GCFR, and CTR design. Similarly, a proposed extension of the CSEWG 94-group photon multigroup structure to 103 groups is described. A subset of the neutron multigroup structure, consisting of 154 groups and for use in power reactor studies, is also presented.

Reference calculations on critical assemblies with Apollo2 code working with a fine multigroup mesh; Calculs de reference avec un maillage multigroupe fin sur des assemblages critiques par Apollo2

Energy Technology Data Exchange (ETDEWEB)

Aggery, A

1999-12-01

The objective of this thesis is to add to the multigroup transport code APOLLO2 the capability to perform deterministic reference calculations, for any type of reactor, using a very fine energy mesh of several thousand groups. This new reference tool allows us to validate the self-shielding model used in industrial applications, to perform depletion calculations, differential effects calculations, critical buckling calculations or to evaluate precisely data required by the self shielding model. At its origin, APOLLO2 was designed to perform routine calculations with energy meshes around one hundred groups. That is why, in the current format of cross sections libraries, almost each value of the multigroup energy transfer matrix is stored. As this format is not convenient for a high number of groups (concerning memory size), we had to search out a new format for removal matrices and consequently to modify the code. In the new format we found, only some values of removal matrices are kept (these values depend on a reconstruction precision choice), the other ones being reconstructed by a linear interpolation, what reduces the size of these matrices. Then we had to show that APOLLO2 working with a fine multigroup mesh had the capability to perform reference calculations on any assembly geometry. For that, we successfully carried out the validation with several calculations for which we compared APOLLO2 results (obtained with the universal mesh of 11276 groups) to results obtained with Monte Carlo codes (MCNP, TRIPOLI4). Physical analysis led with this new tool have been very fruitful and show a great potential for such an R and D tool. (author)

Generation of neutron scattering cross sections for silicon dioxide

International Nuclear Information System (INIS)

Ramos, R; Marquez Damian, J.I; Granada, J.R.; Cantargi, F

2009-01-01

A set of neutron scattering cross sections for silicon and oxygen bound in silicon dioxide were generated and validated. The cross sections were generated in the ACE format for MCNP using the nuclear data processing system NJOY, and the validation was done with published experimental data. This cross section library was applied to the calculation of five critical configurations published in the benchmark Critical Experiments with Heterogeneous Compositions of Highly Enriched Uranium, Silicon Dioxide and Polyethylene. The original calculations did not use the thermal scattering libraries generated in this work and presented significant differences with the experimental results. For this reason, the newly generated library was added to the input and the multiplication factor for each configuration was recomputed. The utilization of the thermal scattering libraries did not result in an improvement of the computational results. Based on this we conclude that integral experiments to validate this type of thermal cross sections need to be designed with a higher influence of thermal scattering in the measured result, and the experiments have to be performed under more controlled conditions. [es

Integral-capture measurements and cross-section adjustments for Nd, Sm, and Eu

International Nuclear Information System (INIS)

Anderl, R.A.; Schmittroth, F.; Harker, Y.D.

1981-07-01

Integral-capture reaction rates are reported for 143 Nd, 144 Nd, 145 Nd, 147 Sm, 151 Eu, 152 Eu, 153 Eu, and 154 Eu irradiated in different neutron spectra in EBR-II. These reaction rates are based primarily on mass-spectrometric measurements of the isotopic atom ratios of the capture product to the target nuclide. The neutron spectra are characterized using passive neutron dosimetry and spectrum-unfolding with the FERRET least-squares data analysis code. Reaction rates for the neutron spectrum monitors were determined by the radiometric technique using Ge(Li) spectrometers. These rates are also reported here. The integral data for the rare-earth samples and for the spectrum monitors were used in multigroup flux/cross-section adtustment analyses with FERRET to generate adjustments to 47 group representations of the ENDF/B-IV capture cross sections for the rare-earth isotopes. These adjusted cross sections are in good agreement with recent differential data and with adjusted cross sections based on STEK integral data. Examples are given of the use of the adjusted cross sections and covariance matrices for cross-section evaluation

Annotated bibliography covering generation and use of evaluated cross section uncertainty files

International Nuclear Information System (INIS)

Peelle, R.W.; Burrows, T.W.

1983-03-01

Literature references related to definition, generation, and use of evaluated cross section uncertainty (variance-covariance) files are listed with comments intended primarily to guide the reader toward materials of immediate interest. Papers are also cited that cover covariance information for individual experiments and that relate to production and use of multigroup covariance matrices. Titles are divided among several major categories

Development on hybrid evaluated nuclear data library HENDL1.0/MG/MC

International Nuclear Information System (INIS)

Xu Dezheng; Gao Chunjing; Zheng Shanliang; Liu Haibo; Zhu Xiaoxiang; Li Jingjing; Wu Yican

2004-01-01

A Hybrid Evaluated Nuclear Data Library (HENDL) named as HENDL1.0 has been developed by Fusion Design Study (FDS) team of Institute of Plasma Physics, Academia Sinica (ASIPP) to take into account the requirements in design and research relevant to fusion, fission and fusion-fission sub-critical hybrid reactor. HENDLI1.0 contains one basic evaluated sub-library naming HENDL1.0/E and to processed working sub-libraries naming HENDL1.0/MG and HENDL1.0/MC, respectively. Through carefully comparing, distinguishing and choosing, HENDL1.0/E integrated basic evaluated neutron data files of 213 nuclides from the several main data libraries for evaluated neutron reaction cross sections including ENDF/B-VI (USA), JEF-2.2 (OECD/NEA, Europe), JENDL-3.2 (Japan), CENDL-2 (China), BROND-2 (Russia) and FENDL-2 (IAEA/NDS, ITER program). Based on this, 175-group neutron and 42-group photon neutron-photon coupled multi-group working library HENDL1.0/MG used for discrete ordinate Sn method transport calculation (such as ANISN code) and a compact ENDF form (ACE), continuous energy structure (pointwise) neutron cross section library HENDL1.0/MC for Monte Carlo method transport simulation (as MCMP code) can be attainable with the current group constants processing system NJOY and transport cross section preparation code TRANSX referring to the Vitamin-J energy group structure. In addition, two special bases i.e. transmutation (burnup) library BURNUP. DAT and response function library RESPONSE.DAT, have been also made for fuel cycle calculation and reactivity analyses of nuclear reactor. The relevant sample testing, benchmark checking and primary confirmation are also carried out to assess the validity of multi-purpose data library HENDL1.0. (authors)

Evaluated cross section libraries and kerma factors for neutrons up to 100 MeV on {sup 16}O and {sup 14}N

Energy Technology Data Exchange (ETDEWEB)

Chadwick, M.B.; Young, P.G.

1995-07-01

We present evaluations of the interaction of 20 to 100 MeV neutrons with oxygen and nitrogen nuclei, which follows on from our previous work on carbon. Our aim is to accurately represent integrated cross sections, inclusive emission spectra, and kerma factors, in a data library which can be used in radiation transport calculations. We apply the FKK-GNASH nuclear model code, which includes Hauser-Feshbach, preequilibrium, and direct reaction mechanisms, and use experimental measurements to optimize the calculations. We determine total, elastic, and nonelastic cross sections, angle-energy correlated emission spectra, for light ejectiles with A{<=}4 and gamma-rays, and average energy depositions. Our results for charged-particle emission spectra agree well with the measurements of Subramanian et al.. We compare kerma factors derived from our evaluated cross sections with experimental data, providing an integral benchmarking of our work. The evaluated data libraries are available as electronic files.

IAEA consultants' meeting on selection of evaluations for the FENDL/A-2 activation cross section library. Summary report

International Nuclear Information System (INIS)

Pashchenko, A.B.

1996-02-01

FENDL/A is a nuclear data library of neutron activation cross-sections prepared for use in nuclear fusion reactor development. The present report contains recommendations for the creation of a second improved version of FENDL/A, including a list of 400 neutron reactions to be considered with priority. (author)

New Standard Evaluated Neutron Cross Section Libraries for the GEANT4 Code and First Verification

CERN Document Server

Mendoza, Emilio; Koi, Tatsumi; Guerrero, Carlos

2014-01-01

The Monte Carlo simulation of the interaction of neutrons with matter relies on evaluated nuclear data libraries and models. The evaluated libraries are compilations of measured physical parameters (such as cross sections) combined with predictions of nuclear model calculations which have been adjusted to reproduce the experimental data. The results obtained from the simulations depend largely on the accuracy of the underlying nuclear data used, and thus it is important to have access to the nuclear data libraries available, either of general use or compiled for specific applications, and to perform exhaustive validations which cover the wide scope of application of the simulation code. In this paper we describe the work performed in order to extend the capabilities of the GEANT4 toolkit for the simulation of the interaction of neutrons with matter at neutron energies up to 20 MeV and a first verification of the results obtained. Such a work is of relevance for applications as diverse as the simulation of a n...

Preparation of next generation set of group cross sections. 3

International Nuclear Information System (INIS)

Kaneko, Kunio

2002-03-01

This fiscal year, based on the examination result about the evaluation energy range of heavy element unresolved resonance cross sections, the upper energy limit of the energy range, where ultra-fine group cross sections are produced, was raised to 50 keV, and an improvement of the group cross section processing system was promoted. At the same time, reflecting the result of studies carried out till now, a function producing delayed neutron data was added to the general-purpose group cross section processing system , thus the preparation of general purpose group cross section processing system has been completed. On the other hand, the energy structure, data constitution and data contents of next generation group cross section set were determined, and the specification of a 151 groups next generation group cross section set was defined. Based on the above specification, a concrete library format of the next generation cross section set has been determined. After having carried out the above-described work, using the general-purpose group cross section processing system , which was complete in this study, with use of the JENDL-3. 2 evaluated nuclear data, the 151 groups next generation group cross section of 92 nuclides and the ultra fine group resonance cross section library for 29 nuclides have been prepared. Utilizing the 151 groups next generation group cross section set and the ultra-fine group resonance cross-section library, a bench mark test calculation of fast reactors has been performed by using an advanced lattice calculation code. It was confirmed, by comparing the calculation result with a calculation result of continuous energy Monte Carlo code, that the 151 groups next generation cross section set has sufficient accuracy. (author)

Establishment of the BOSPOR-80 machine library of evaluated threshold reaction cross-sections and its testing by means of integral experiments

International Nuclear Information System (INIS)

Bychkov, V.M.; Zolotarev, K.I.; Pashchenko, A.B.; Plyaskin, V.I.

1982-08-01

A paper was published in 1979 containing a compilation of experimental data on the cross-sections of (n,p), (n,α) and (n,2n) threshold reactions and recommended excitation functions. A further paper considered the development of evaluation methods based on the use of theoretical model calculations, an increase in the number of recommended excitation functions, correction of the recommended cross-sections on the basis of integral experiments and allowance for recent experimental data. To satisfy the wide circle of users, BOSPOR-80 - a machine library of evaluated threshold reaction cross-sections - was set up

Specifications for adjusted cross section and covariance libraries based upon CSEWG fast reactor and dosimetry benchmarks

International Nuclear Information System (INIS)

Weisbin, C.R.; Marable, J.H.; Collins, P.J.; Cowan, C.L.; Peelle, R.W.; Salvatores, M.

1979-06-01

The present work proposes a specific plan of cross section library adjustment for fast reactor core physics analysis using information from fast reactor and dosimetry integral experiments and from differential data evaluations. This detailed exposition of the proposed approach is intended mainly to elicit review and criticism from scientists and engineers in the research, development, and design fields. This major attempt to develop useful adjusted libraries is based on the established benchmark integral data, accurate and well documented analysis techniques, sensitivities, and quantified uncertainties for nuclear data, integral experiment measurements, and calculational methodology. The adjustments to be obtained using these specifications are intended to produce an overall improvement in the least-squares sense in the quality of the data libraries, so that calculations of other similar systems using the adjusted data base with any credible method will produce results without much data-related bias. The adjustments obtained should provide specific recommendations to the data evaluation program to be weighed in the light of newer measurements, and also a vehicle for observing how the evaluation process is converging. This report specifies the calculational methodology to be used, the integral experiments to be employed initially, and the methods and integral experiment biases and uncertainties to be used. The sources of sensitivity coefficients, as well as the cross sections to be adjusted, are detailed. The formulae for sensitivity coefficients for fission spectral parameters are developed. A mathematical formulation of the least-square adjustment problem is given including biases and uncertainties in methods

TEMPEST-2, Thermalization Program for Neutron Spectra and Multigroup Cross-Sections

International Nuclear Information System (INIS)

Gowins, G.

1984-01-01

Description of problem or function: TEMPEST2 is a neutron thermalization program based upon the Wigner-Wilkins approximation for light moderators and the Wilkins approximation for heavy moderators. A Maxwellian distribution may also be used. The model used may be selected as a function of energy. The second-order differential equations are integrated directly rather than transformed to the Riccati equation. The program provides microscopic and macroscopic cross-section averages over the thermal neutron spectrum

Curves and tables of neutron cross sections

International Nuclear Information System (INIS)

Nakagawa, Tsuneo; Asami, Tetsuo; Yoshida, Tadashi

1990-07-01

Neutron cross-section curves from the Japanese Evaluated Nuclear Data Library version 3, JENDL-3, are presented in both graphical and tabular form for users in a wide range of application areas in the nuclear energy field. The contents cover cross sections for all the main reactions induced by neutrons with an energy below 20 MeV including; total, elastic scattering, capture, and fission, (n,n'), (n,2n), (n,3n), (n,α), (n,p) reactions. The 2200 m/s cross-section values, resonance integrals, and Maxwellian- and fission-spectrum averaged cross sections are also tabulated. (author)

Validation of evaluated neutron standard cross sections

International Nuclear Information System (INIS)

Badikov, S.; Golashvili, T.

2008-01-01

Some steps of the validation and verification of the new version of the evaluated neutron standard cross sections were carried out. In particular: -) the evaluated covariance data was checked for physical consistency, -) energy-dependent evaluated cross-sections were tested in most important neutron benchmark field - 252 Cf spontaneous fission neutron field, -) a procedure of folding differential standard neutron data in group representation for preparation of specialized libraries of the neutron standards was verified. The results of the validation and verification of the neutron standards can be summarized as follows: a) the covariance data of the evaluated neutron standards is physically consistent since all the covariance matrices of the evaluated cross sections are positive definite, b) the 252 Cf spectrum averaged standard cross-sections are in agreement with the evaluated integral data (except for 197 Au(n,γ) reaction), c) a procedure of folding differential standard neutron data in group representation was tested, as a result a specialized library of neutron standards in the ABBN 28-group structure was prepared for use in reactor applications. (authors)

SIRIUS - A one-dimensional multigroup analytic nodal diffusion theory code

Energy Technology Data Exchange (ETDEWEB)

Forslund, P. [Westinghouse Atom AB, Vaesteraas (Sweden)

2000-09-01

In order to evaluate relative merits of some proposed intranodal cross sections models, a computer code called Sirius has been developed. Sirius is a one-dimensional, multigroup analytic nodal diffusion theory code with microscopic depletion capability. Sirius provides the possibility of performing a spatial homogenization and energy collapsing of cross sections. In addition a so called pin power reconstruction method is available for the purpose of reconstructing 'heterogeneous' pin qualities. consequently, Sirius has the capability of performing all the calculations (incl. depletion calculations) which are an integral part of the nodal calculation procedure. In this way, an unambiguous numerical analysis of intranodal cross section models is made possible. In this report, the theory of the nodal models implemented in sirius as well as the verification of the most important features of these models are addressed.

Reference calculations on critical assemblies with Apollo2 code working with a fine multigroup mesh

International Nuclear Information System (INIS)

Aggery, A.

1999-12-01

The objective of this thesis is to add to the multigroup transport code APOLLO2 the capability to perform deterministic reference calculations, for any type of reactor, using a very fine energy mesh of several thousand groups. This new reference tool allows us to validate the self-shielding model used in industrial applications, to perform depletion calculations, differential effects calculations, critical buckling calculations or to evaluate precisely data required by the self shielding model. At its origin, APOLLO2 was designed to perform routine calculations with energy meshes around one hundred groups. That is why, in the current format of cross sections libraries, almost each value of the multigroup energy transfer matrix is stored. As this format is not convenient for a high number of groups (concerning memory size), we had to search out a new format for removal matrices and consequently to modify the code. In the new format we found, only some values of removal matrices are kept (these values depend on a reconstruction precision choice), the other ones being reconstructed by a linear interpolation, what reduces the size of these matrices. Then we had to show that APOLLO2 working with a fine multigroup mesh had the capability to perform reference calculations on any assembly geometry. For that, we successfully carried out the validation with several calculations for which we compared APOLLO2 results (obtained with the universal mesh of 11276 groups) to results obtained with Monte Carlo codes (MCNP, TRIPOLI4). Physical analysis led with this new tool have been very fruitful and show a great potential for such an R and D tool. (author)

Status of pseudo fission product cross sections for fast reactors. Results of the SWG 17, International working party on evaluation coordination of the nuclear science committee, NEA- OECD

International Nuclear Information System (INIS)

Gruppelaar, H.; Kloosterman, J.L.; Pijlgroms, B.J.; Rimpault, G.; Smith, P.; Ignatyuk, A.; Koshcheev, V.; Nikolaev, M.; Thsiboulia, A.; Kawai, M.; Nakagawa, T.; Watanabe, T.; Zukeran, A.; Nakajima, Y.; Matsunobu, H.

1998-08-01

Within the framework of the SWG17 benchmark organized by a Working Party of the Nuclear Science Committee of the Nuclear Energy Agency (NEA), a comparison of lumped or pseudo fission product cross sections for fast reactors has been made. Four institutions participated with data libraries based on the JEF2.2, EAF-4.2, BROND-2, FONDL-2.1, ADL-3 and JENDL-3.2 evaluated nuclear data files. Several parameters have been compared with each other: the one-group cross sections and reactivity worths of the lumped nuclide for several partial absorption and scattering cross sections, and the one-group cross sections of the individual fission products. Also graphs of the multi-group cross sections of the lumped nuclide have been compared, as well as graphs of capture cross sections for 27 nuclides. From two contributions based on JEF2.2, it can be concluded that the data processing influences the capture cross section by about 1% and the inelastic scattering cross section by 2%. The differences between the lumped cross sections of the different data libraries are surprisingly small: maximum 6% for capture and 9% for the inelastic scattering. Similar results are obtained for the reactivity effects. Since the reactivity worth of the lumped nuclide is dominated by the capture reaction, the maximum spread in the total reactivity worth is still only 5.3%. There is a systematic difference between total, elastic and capture cross sections of JENDL-3.2 and JEF2.2 of the same order of magnitude. Possible reasons for this discrepancy have been indicated. The one-group capture and inelastic scattering cross sections of most of the important individual fission products differ by less than 10% (root mean square values). Larger differences are observed for unstable nuclides where there is a lack of experimental data. For the (n,2n) group cross sections, which are rather sensitive to the weighting spectrum in the fast energy range, these differences are several tens of percents. The final

Qualification of the calculational methods of the fluence in the pressurised water reactors. Improvement of the cross sections treatment by the probability table method; Qualification des methodes de calculs de fluence dans les reacteurs a eau pressurisee. Amelioration du traitement des sections efficaces par la methode des tables de probabilite

Energy Technology Data Exchange (ETDEWEB)

Zheng, S H

1994-01-01

It is indispensable to know the fluence on the nuclear reactor pressure vessel. The cross sections and their treatment have an important rule to this problem. In this study, two ``benchmarks`` have been interpreted by the Monte Carlo transport program TRIPOLI to qualify the calculational method and the cross sections used in the calculations. For the treatment of the cross sections, the multigroup method is usually used but it exists some problems such as the difficulty to choose the weighting function and the necessity of a great number of energy to represent well the cross section`s fluctuation. In this thesis, we propose a new method called ``Probability Table Method`` to treat the neutron cross sections. For the qualification, a program of the simulation of neutron transport by the Monte Carlo method in one dimension has been written; the comparison of multigroup`s results and probability table`s results shows the advantages of this new method. The probability table has also been introduced in the TRIPOLI program; the calculational results of the iron deep penetration benchmark has been improved by comparing with the experimental results. So it is interest to use this new method in the shielding and neutronic calculation. (author). 42 refs., 109 figs., 36 tabs.

Nuclear data libraries for Tripoli-3.5 code; Bibliotheques de donnees nucleaires pour le code tripoli-3.5

Energy Technology Data Exchange (ETDEWEB)

Vergnaud, Th

2001-07-01

The TRIPOLI-3 code uses multigroup nuclear data libraries generated using the NJOY-THEMIS suite of modules: for neutrons, they are produced from the ENDF/B-VI evaluations and cover the range between 20 MeV and 10{sup -5} eV, either in 315 groups and for one temperature, or in 3209 groups and for five temperatures; for gamma-rays, they are from JEF2 and are processed in groups between 14 MeV and keV. The probability tables used for the neutron transport calculations have been derived from the ENDF/B-VI evaluations using the CALENDF code. Cross sections for gamma production by neutron interaction (fission, capture or inelastic scattering) have been derived from ENDF/B-VI in 315 neutron groups and 75 gamma groups. The code also uses two response function libraries: for neutrons; based on several sources, in particular the dosimetry libraries IRDF/85 and IRDF/90; for gamma-rays it is based on the JEF2 evaluation and contains the kerma factors for all the elements and cross sections for all interactions. (author)

The needs for program and cross-section library improvement in calculation of neutron-induced activity inventories

International Nuclear Information System (INIS)

Yavshitz, S.G.; Rubchenya, V.A.; Rimski-Korsakov, A.A.

1993-01-01

The authors demonstrate the possibility of an approach to evaluate the radioactive inventory - induced activity of structural materials and surface contamination of reactor components, that will fit well into ORIGEN code structure and could be used on a modest PC directly on the decommissioning site. This approach would also require only one well tested set of pre-calculated and adjusted by experiment cross-section libraries (averaged by typical neutron spectra outside the reactor core). 15 refs, 1 fig

Continuous energy Monte Carlo method based homogenization multi-group constants calculation

International Nuclear Information System (INIS)

Li Mancang; Wang Kan; Yao Dong

2012-01-01

The efficiency of the standard two-step reactor physics calculation relies on the accuracy of multi-group constants from the assembly-level homogenization process. In contrast to the traditional deterministic methods, generating the homogenization cross sections via Monte Carlo method overcomes the difficulties in geometry and treats energy in continuum, thus provides more accuracy parameters. Besides, the same code and data bank can be used for a wide range of applications, resulting in the versatility using Monte Carlo codes for homogenization. As the first stage to realize Monte Carlo based lattice homogenization, the track length scheme is used as the foundation of cross section generation, which is straight forward. The scattering matrix and Legendre components, however, require special techniques. The Scattering Event method was proposed to solve the problem. There are no continuous energy counterparts in the Monte Carlo calculation for neutron diffusion coefficients. P 1 cross sections were used to calculate the diffusion coefficients for diffusion reactor simulator codes. B N theory is applied to take the leakage effect into account when the infinite lattice of identical symmetric motives is assumed. The MCMC code was developed and the code was applied in four assembly configurations to assess the accuracy and the applicability. At core-level, A PWR prototype core is examined. The results show that the Monte Carlo based multi-group constants behave well in average. The method could be applied to complicated configuration nuclear reactor core to gain higher accuracy. (authors)

Comparison of CASMO and NESSEL few group cross section libraries and their usage in DYN3D

International Nuclear Information System (INIS)

Kuchin, A.; Ovdiyenko, Y.; Loetsch, T.

2007-01-01

This work presents comparative analysis of two group diffusion cross sections libraries which were generated by NESSEL-4 and CASMO-4 lattice codes. Diffusion parameters were calculated for WWER-1000 fuel assemblies with stainless steel spacing grids and guiding tubes. These cross section sets were introduced into reactor core code DYN3D and tested on the base of real reactor core states. In this case operation data of the first three fuel cycles of sixth unit of Zaporizhzhya NPP were used. The work was performed in the framework of the order BMU SR 2511 - 862 500/09, UA-2575. The report describes the opinion and view of the contractor - TUV ENERGIE CONSULT - and does not necessarily represent the opinion of the ordering party - Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (Authors)

Cross section recondensation method via generalized energy condensation theory

International Nuclear Information System (INIS)

Douglass, Steven; Rahnema, Farzad

2011-01-01

Highlights: → A new method is presented which corrects for core environment error from specular boundaries at the lattice cell level. → Solution obtained with generalized energy condensation provides improved approximation to the core level fine-group flux. → Iterative recondensation of the cross sections and unfolding of the flux provides on-the-fly updating of the core cross sections. → Precomputation of energy integrals and fine-group cross sections allows for easy implementation and efficient solution. → Method has been implemented in 1D and shown to correct the environment error, particularly in strongly heterogeneous cores. - Abstract: The standard multigroup method used in whole-core reactor analysis relies on energy condensed (coarse-group) cross sections generated from single lattice cell calculations, typically with specular reflective boundary conditions. Because these boundary conditions are an approximation and not representative of the core environment for that lattice, an error is introduced in the core solution (both eigenvalue and flux). As current and next generation reactors trend toward increasing assembly and core heterogeneity, this error becomes more significant. The method presented here corrects for this error by generating updated coarse-group cross sections on-the-fly within whole-core reactor calculations without resorting to additional cell calculations. In this paper, the fine-group core flux is unfolded by making use of the recently published Generalized Condensation Theory and the cross sections are recondensed at the whole-core level. By iteratively performing this recondensation, an improved core solution is found in which the core-environment has been fully taken into account. This recondensation method is both easy to implement and computationally very efficient because it requires precomputation and storage of only the energy integrals and fine-group cross sections. In this work, the theoretical basis and development

POLIDENT: A Module for Generating Continuous-Energy Cross Sections from ENDF Resonance Data

Energy Technology Data Exchange (ETDEWEB)

Dunn, M.E.; Greene, N.M.

2000-12-01

POLIDENT (Point Libraries of Data from ENDF/B Tapes) is an AMPX module that accesses the resonance parameters from File 2 of an ENDF/B library and constructs the continuous-energy cross sections in the resonance energy region. The cross sections in the resonance range are subsequently combined with the File 3 background data to construct the cross-section representation over the complete energy range. POLIDENT has the capability to process all resonance reactions that are identified in File 2 of the ENDF/B library. In addition, the code has the capability to process the single- and multi-level Breit-Wigner, Reich-Moore and Adler-Adler resonance formalisms that are identified in File 2. POLIDENT uses a robust energy-mesh-generation scheme that determines the minimum, maximum and points of inflection in the cross-section function in the resolved-resonance region. Furthermore, POLIDENT processes all continuous-energy cross-section reactions that are identified in File 3 of the ENDF/B library and outputs all reactions in an ENDF/B TAB1 format that can be accessed by other AMPX modules.

Neutron Thermal Cross Sections, Westcott Factors, Resonance Integrals, Maxwellian Averaged Cross Sections and Astrophysical Reaction Rates Calculated from the ENDF/B-VII.1, JEFF-3.1.2, JENDL-4.0, ROSFOND-2010, CENDL-3.1 and EAF-2010 Evaluated Data Libraries

Science.gov (United States)

Pritychenko, B.; Mughabghab, S. F.

2012-12-01

We present calculations of neutron thermal cross sections, Westcott factors, resonance integrals, Maxwellian-averaged cross sections and astrophysical reaction rates for 843 ENDF materials using data from the major evaluated nuclear libraries and European activation file. Extensive analysis of newly-evaluated neutron reaction cross sections, neutron covariances, and improvements in data processing techniques motivated us to calculate nuclear industry and neutron physics quantities, produce s-process Maxwellian-averaged cross sections and astrophysical reaction rates, systematically calculate uncertainties, and provide additional insights on currently available neutron-induced reaction data. Nuclear reaction calculations are discussed and new results are presented. Due to space limitations, the present paper contains only calculated Maxwellian-averaged cross sections and their uncertainties. The complete data sets for all results are published in the Brookhaven National Laboratory report.

Development of the adjusted nuclear cross-section library based on JENDL-3.2 for large FBR

International Nuclear Information System (INIS)

Yokoyama, Kenji; Ishikawa, Makoto; Numata, Kazuyuki

1999-04-01

JNC (and PNC) had developed the adjusted nuclear cross-section library in which the results of the JUPITER experiments were reflected. Using this adjusted library, the distinct improvement of the accuracy in nuclear design of FBR cores had been achieved. As a recent research, JNC develops a database of other integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. In 1991, the adjusted library based on JENDL-2, JFS-3-J2 (ADJ91R), was developed, and it has been used on the design research for FBR. As an evaluated nuclear library, however, JENDL-3.2 is recently used. Therefore, the authors developed an adjusted library based on JENDL-3.2 which is called JFS-3-J3.2(ADJ98). It is known that the adjusted library based on JENDL-2 overestimated the sodium void reactivity worth by 10-20%. It is expected that the adjusted library based on JENDL-3.2 solve the problem. The adjusted library JFS-3-J3.2(ADJ98) was produced with the same method as the adjusted library JFS-3-J2(ADJ91R) and used more integral parameters of JUPITER experiments than the adjusted library JFS-3-J2(ADJ91R). This report also describes the design accuracy estimation on a 600 MWe class FBR with the adjusted library JFS-3-J3.2(ADJ98). Its main nuclear design parameters (multiplication factor, burn-up reactivity loss, breeding ratio, etc.) except the sodium void reactivity worth which are calculated with the adjusted library JFS-3-J3.2(ADJ98) are almost the same as those predicted with JFS-3-J2(ADJ91R). As for the sodium void reactivity, the adjusted library JFS-3-J3.2(ADJ98) estimates about 4% smaller than the JFS-3-J2(ADJ91R) because of the change of the basic nuclear library from JENDL-2 to JENDL-3.2. (author)

Skyshine analysis using various nuclear data files

International Nuclear Information System (INIS)

Zharkov, V.P.; Dikareva, O.F.; Kartashev, I.A.; Kiselev, A.N.; Nomura, Y.; Tsubosaka, A.

2000-01-01

The calculations of the spacial distributions of dose rate for neutron and secondary photons, thermal neutron fluxes and space-energy distributions of neutron and photons near the air-ground interface were performed by MCNP and DORT codes. Different nuclear data files were used (ENDF/B-IV, ENDF/B-VI, FENDL-2, JENDL-3.2). Either the standard pointwise libraries (MCNP) or special libraries prepared by NJOY code from ENDF/B and others' files were used. Prepared multigroup coupled neutron and photon cross sections libraries for DORT code had CASK-40 group energy structures. The libraries contain pointwise or multigroup cross sections data for all elements included in the atmosphere and ground composition. The validation of the calculated results was performed with using the experimental data obtained for the series of measurements at RA reactor. (author)

Continuous-energy version of KENO V.a for criticality safety applications

International Nuclear Information System (INIS)

Dunn, Michael E.; Greene, N. Maurice; Petrie, Lester M.

2003-01-01

KENO V.a is a multigroup Monte Carlo code that solves the Boltzmann transport equation and is used extensively in the criticality safety community to calculate the effective multiplication factor of systems with fissionable material. In this work, a continuous-energy or pointwise version of KENO V.a has been developed by first designing a new continuous-energy cross-section format and then by developing the appropriate Monte Carlo transport procedures to sample the new cross-section format. In order to generate pointwise cross sections for a test library, a series of cross-section processing modules were developed and used to process 50 ENDF/B-6 Release 7 nuclides for the test library. Once the cross-section processing procedures were in place, a continuous-energy version of KENO V.a was developed and tested by calculating 21 critical benchmark experiments. The point KENO-calculated results for the 21 benchmarks are in agreement with calculated results obtained with the multigroup version of KENO V.a using the 238-group ENDF/B-5 and 199-group ENDF/B-6 Release 3 libraries. Based on the calculated results with the prototypic cross-section library, a continuous-energy version of the KENO V.a code has been successfully developed and demonstrated for modeling systems with fissionable material. (author)

Neutron and proton transmutation-activation cross section libraries to 150 MeV for application in accelerator-driven systems and radioactive ion beam target-design studies

International Nuclear Information System (INIS)

Koning, A.J.; Chadwick, M.B.; MacFarlane, R.E.; Mashnik, S.; Wilson, W.B.

1998-05-01

New transmutation-activation nuclear data libraries for neutrons and protons up to 150 MeV have been created. These data are important for simulation calculations of radioactivity, and transmutation, in accelerator-driven systems such as the production of tritium (APT) and the transmutation of waste (ATW). They can also be used to obtain cross section predictions for the production of proton-rich isotopes in (p,xn) reactions, for radioactive ion beam (RIB) target-design studies. The nuclear data in these libraries stem from two sources: for neutrons below 20 MeV, we use data from the European activation and transmutation file, EAF97; For neutrons above 20 MeV and for protons at all energies we have isotope production cross sections with the nuclear model code HMS-ALICE. This code applies the Monte Carlo Hybrid Simulation theory, and the Weisskopf-Ewing theory, to calculate cross sections. In a few cases, the HMS-ALICE results were replaced by those calculated using the GNASH code for the Los Alamos LA150 transport library. The resulting two libraries, AF150.N and AF150.P, consist of 766 nuclides each and are represented in the ENDF6-format. An outline is given of the new representation of the data. The libraries have been checked with ENDF6 preprocessing tools and have been processed with NJOY into libraries for the Los Alamos transmutation/radioactivity code CINDER. Numerous benchmark figures are presented for proton-induced excitation functions of various isotopes compared with measurements. Such comparisons are useful for validation purposes, and for assessing the accuracy of the evaluated data. These evaluated libraries are available on the WWW at: http://t2.lanl.gov/. 21 refs

Criticality and safety parameter studies for upgrading 3 MW TRIGA MARK II research reactor and validation of generated cross section library and computational method

International Nuclear Information System (INIS)

Bhuiyan, S.I.; Mondal, M.A.W.; Sarker, M.M.; Rahman, M.; Shahdatullah, M.S.; Huda, M.Q.; Chakrroborty, T.K.; Khan, M.J.H.

2000-01-01

This study deals with the neutronic and thermal hydraulic analysis of the 3MW TRIGA MARK II research reactor to upgrade it to a higher flux. The upgrading will need a major reshuffling and reconfiguration of the current core. To reshuffle the current core configuration, the chain of NJOY94.10 - WIMSD-5A - CITATION - PARET - MCNP4B2 codes has been used for the overall analysis. The computational methods, tools and techniques, customisation of cross section libraries, various models for cells and super cells, and a lot of associated utilities have been standardised and established/validated for the overall core analysis. Analyses using the 4-group and 7-group libraries of macroscopic cross sections generated from the 69-group WIMSD-5 library showed that a 7-group structure is more suitable for TRIGA calculations considering its LEU fuel composition. The MCNP calculations established that the CITATION calculations and the generated cross section library are reasonably good for neutronic analysis of TRIGA reactors. Results obtained from PARET demonstrated that the flux upgrade will not cause the temperature limit on the fuel to be exceeded. Also, the maximum power density remains, by a substantial margin below the level at which the departure from nucleate boiling could occur. A possible core with two additional irradiation channels around the CT is projected where almost identical thermal fluxes as in the CT are obtained. The reconfigured core also shows 7.25% thermal flux increase in the Lazy Susan. (author)

Evaluation of ETOG-3Q/ETOG-3, FLANGE-II, XLACS, NJOY and linear/recent/groupie codes for calculations of resonance and reference cross sections

International Nuclear Information System (INIS)

Anaf, J.; Chalhoub, E.S.

1991-01-01

The NJOY and LINEAR/RECENT/GROUPIE calculational procedures for the resolved and unresolved resonance contributions and background cross sections are evaluated. Elastic scattering, fission and capture multigroup cross sections generated by these codes and the previously validated ETOG-3Q, ETOG-3, FLANGE-II and XLACS are compared. Constant weighting function and zero Kelvin temperature are considered. Discrepancies are presented and analyzed. (author)

Development of fine-group (315n/42γ) cross section library ENDL3.0/FG for fusion-fission hybrid systems

International Nuclear Information System (INIS)

Zeng Qin; Zou Jun; Xu Dezhen; Jiang Jieqiong; Wang Minghuang; Wu Yican; Qiu Yuefeng; Chen Zhong; Chen Yan

2011-01-01

To improve the accuracy of the neutron analyses for subcritical systems with thermal fission blanket, a coupled neutron and photon (315 n + 42γ) fine-group cross section library HENDL3.0/FG based on ENDF/B-Ⅶ. 0 has been produced by FDS team. In order to test the availability and reliability of the HENDL3.0/FG data library, shielding and critical safety benchmarks were performed with VisualBUS code. The testing results indicated that the discrepancy between calculation and experimental values of nuclear parameters fell in a reasonable range. (authors)

Qualification of the calculational methods of the fluence in the pressurised water reactors. Improvement of the cross sections treatment by the probability table method

International Nuclear Information System (INIS)

Zheng, S.H.

1994-01-01

It is indispensable to know the fluence on the nuclear reactor pressure vessel. The cross sections and their treatment have an important rule to this problem. In this study, two ''benchmarks'' have been interpreted by the Monte Carlo transport program TRIPOLI to qualify the calculational method and the cross sections used in the calculations. For the treatment of the cross sections, the multigroup method is usually used but it exists some problems such as the difficulty to choose the weighting function and the necessity of a great number of energy to represent well the cross section's fluctuation. In this thesis, we propose a new method called ''Probability Table Method'' to treat the neutron cross sections. For the qualification, a program of the simulation of neutron transport by the Monte Carlo method in one dimension has been written; the comparison of multigroup's results and probability table's results shows the advantages of this new method. The probability table has also been introduced in the TRIPOLI program; the calculational results of the iron deep penetration benchmark has been improved by comparing with the experimental results. So it is interest to use this new method in the shielding and neutronic calculation. (author). 42 refs., 109 figs., 36 tabs

Verification of the cross-section and depletion chain processing module of DRAGON 3.06

International Nuclear Information System (INIS)

Chambon, R.; Marleau, G.; Zkiek, A.

2008-01-01

In this paper we present a verification of the module of the lattice code DRAGON 3.06 used for processing microscopic cross-section libraries, including their associated depletion chain. This verification is performed by reprogramming the capabilities of DRAGON in another language (MATLAB) and testing them on different problems typical of the CANDU reactor. The verification procedure consists in first programming MATLAB m-files to read the different cross section libraries in ASCII format and to compute the reference cross-sections and depletion chains. The same information is also recovered from the output files of DRAGON (using different m-files) and the resulting cross sections and depletion chain are compared with the reference library, the differences being evaluated and tabulated. The results show that the cross-section calculations and the depletion chains are correctly processed in version 3.06 of DRAGON. (author)

A library for X-ray-matter interaction cross sections for X-ray fluorescence applications

Energy Technology Data Exchange (ETDEWEB)

Brunetti, A. [Istituto di Matematica e Fisica, Universita di Sassari, via Vienna 2, 07100 Sassari (Italy) and INFN, Sezione di Cagliari (Italy)]. E-mail: brunetti@uniss.it; Sanchez del Rio, M. [European Synchrotron Radiation Facility, 6 rue Jules Horowitz, 38043 Grenoble Cedex (France); Golosio, B. [INFN, Sezione di Cagliari (Italy); European Synchrotron Radiation Facility, 6 rue Jules Horowitz, 38043 Grenoble Cedex (France); Simionovici, A. [European Synchrotron Radiation Facility, 6 rue Jules Horowitz, 38043 Grenoble Cedex (France); Laboratoire de Sciences de la Terre, Ecole Normale Superieure, Lyon, F-69364 (France); Somogyi, A. [European Synchrotron Radiation Facility, 6 rue Jules Horowitz, 38043 Grenoble Cedex (France)

2004-10-08

Quantitative estimate of elemental composition by spectroscopic and imaging techniques using X-ray fluorescence requires the availability of accurate data of X-ray interaction with matter. Although a wide number of computer codes and data sets are reported in literature, none of them is presented in the form of freely available library functions which can be easily included in software applications for X-ray fluorescence. This work presents a compilation of data sets from different published works and an xraylib interface in the form of callable functions. Although the target applications are on X-ray fluorescence, cross sections of interactions like photoionization, coherent scattering and Compton scattering, as well as form factors and anomalous scattering functions, are also available.

Positive Scattering Cross Sections using Constrained Least Squares

International Nuclear Information System (INIS)

Dahl, J.A.; Ganapol, B.D.; Morel, J.E.

1999-01-01

A method which creates a positive Legendre expansion from truncated Legendre cross section libraries is presented. The cross section moments of order two and greater are modified by a constrained least squares algorithm, subject to the constraints that the zeroth and first moments remain constant, and that the standard discrete ordinate scattering matrix is positive. A method using the maximum entropy representation of the cross section which reduces the error of these modified moments is also presented. These methods are implemented in PARTISN, and numerical results from a transport calculation using highly anisotropic scattering cross sections with the exponential discontinuous spatial scheme is presented

Two-level MOC calculation scheme in APOLLO2 for cross-section library generation for LWR hexagonal assemblies

International Nuclear Information System (INIS)

Petrov, Nikolay; Todorova, Galina; Kolev, Nikola; Damian, Frederic

2011-01-01

The accurate and efficient MOC calculation scheme in APOLLO2, developed by CEA for generating multi-parameterized cross-section libraries for PWR assemblies, has been adapted to hexagonal assemblies. The neutronic part of this scheme is based on a two-level calculation methodology. At the first level, a multi-cell method is used in 281 energy groups for cross-section definition and self-shielding. At the second level, precise MOC calculations are performed in a collapsed energy mesh (30-40 groups). In this paper, the application and validation of the two-level scheme for hexagonal assemblies is described. Solutions for a VVER assembly are compared with TRIPOLI4® calculations and direct 281g MOC solutions. The results show that the accuracy is close to that of the 281g MOC calculation while the CPU time is substantially reduced. Compared to the multi-cell method, the accuracy is markedly improved. (author)

ZZ SNLRML, Dosimetry Cross-Section Recommendations

International Nuclear Information System (INIS)

1996-01-01

Description of program or function: Format: SAND-II; Number of groups: 640 group SAND-II group structure. Nuclides: Cd, B, Au, S, Ni, Li, F, Na, Mg, Al, Si, P, Sc, Ti, Mn, Fe, Co, Cu, Zn, Zr, Nb, Mo, Rh, Ag, In, I, Th, U, Np, Pu, Am. Origin: ENDF/B-VI, ENDF/B-V, IRDF-90, JENDL-3, JEF 2.2 and GLUCS data with special modifications from private communications. Weighting spectrum: flat. SNLRML is a reactor dosimetry library that draws upon all available evaluated cross section libraries and selects the best evaluation for application to research reactor spectrum determinations. Many of the components of the SNLRML come from the ENDF/B-VI and IRDF-90 (DLC-0161) libraries. The library format was selected for easy interface with spectrum determination codes such as SAND-II (CCC-0112 and LSL-M2 (PSR-233) and the new PSR-0345/SNL/SAND-II has been enhanced to interface with SNLRML. The data is recommended for spectrum determination applications and for the prediction of neutron activation of typical radiation sensor materials. The library has been tested for consistency of the cross section in wide variety of neutron environments. The results and cautions from this testing have been documented. The data has been interfaced with radiation transport codes, such as TWODANT-SYS (CCC-0547) and MCNP (CCC-0200), in order to compare calculated and measured activities for benchmark reactor experiments

Recommended evaluation procedure for photonuclear cross section

Energy Technology Data Exchange (ETDEWEB)

Lee, Young-Ouk; Chang, Jonghwa; Fukahori, Tokio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-03-01

In order to generate photonuclear cross section library for the necessary applications, data evaluation is combined with theoretical evaluation, since photonuclear cross sections measured cannot provide all necessary data. This report recommends a procedure consisting of four steps: (1) analysis of experimental data, (2) data evaluation, (3) theoretical evaluation and, if necessary, (4) modification of results. In the stage of analysis, data obtained by different measurements are reprocessed through the analysis of their discrepancies to a representative data set. In the data evaluation, photonuclear absorption cross sections are evaluated via giant dipole resonance and quasi-deutron mechanism. With photoabsorption cross sections from the data evaluation, theoretical evaluation is applied to determine various decay channel cross sections and emission spectra using equilibrium and preequilibrium mechanism. After this, the calculated results are compared with measured data, and in some cases the results are modified to better describe measurements. (author)

Continuous energy cross section library for MCNP/MCNPX based on JENDL high energy file 2007. FXJH7

International Nuclear Information System (INIS)

Sasa, Toshinobu; Sugawara, Takanori; Fukahori, Tokio; Kosako, Kazuaki

2008-11-01

The latest JENDL High Energy File (JENDL/HE) was released in 2007 to respond the requirements of reaction data in high energy range up to several GeV to design accelerator facilities such as accelerator-driven systems and research complex like J-PARC. To apply the JENDL/HE-2007 file to the design study, the cross section library of FXJH7 series was constructed from the JENDL/HE file for the calculation using MCNP and MCNPX codes which are widely used in the field of nuclear reactors, fusion reactors, accelerator facilities, medical applications, and so on. In this report, the outline of the JENDL/HE-2007 file, modification of nuclear data processing code NJOY99, construction of FXJH7 library and test calculations for shielding and eigenvalue analyses are summarized. (author)

Research on the display of nuclear cross-section library

International Nuclear Information System (INIS)

Huang Shien; Wang Kan; Yu Ganglin

2008-01-01

Minutely parsed the dot cross-section format (ACE format) data of the ENDF/ B-6.8 database, which is the foundation of the program that achieved the reading and related handling of ACE format data. This program achieved the plotting, zooming and comparing display functions of nuclear cross section-energy of ENDF/B-6.8 database. It also provides the standard picture formatting file output and/or standard text formatting file output of interesting nuclear data. It accomplished some appropriate validations of this program via the comparing between program results and reference data. (authors)

Computation of Resonance-Screened Cross Section by the Dorix-Speng System

Energy Technology Data Exchange (ETDEWEB)

Haeggblom, H

1968-09-15

The report describes a scheme for computation of group cross sections for fast reactors in energy regions where the resonance structure of the cross sections may be dense. A combination of the programmes Dorix and Speng is then used. Dorix calculates group cross sections for each resonance absorber separately. The interaction between resolved resonances in the same isotope is treated using a method described in a separate report. The interaction between correlated and non-correlated resonances in the unresolved region is also considered. By a Dorix calculation we obtain effective microscopic cross sections which are then read in on a library tape. This library contains both point-by-point data and group cross sections and is used in the Speng programme for computation of spectrum and/or macroscopic cross sections. The resonance interaction between different isotopes is computed in Speng by the same method as was used in the Dorix programme for non-correlated unresolved resonances. Consideration is also given to the width of the resonances compared to the energy loss by a neutron colliding with some of the scattering elements.

Computation of Resonance-Screened Cross Section by the Dorix-Speng System

International Nuclear Information System (INIS)

Haeggblom, H.

1968-09-01

The report describes a scheme for computation of group cross sections for fast reactors in energy regions where the resonance structure of the cross sections may be dense. A combination of the programmes Dorix and Speng is then used. Dorix calculates group cross sections for each resonance absorber separately. The interaction between resolved resonances in the same isotope is treated using a method described in a separate report. The interaction between correlated and non-correlated resonances in the unresolved region is also considered. By a Dorix calculation we obtain effective microscopic cross sections which are then read in on a library tape. This library contains both point-by-point data and group cross sections and is used in the Speng programme for computation of spectrum and/or macroscopic cross sections. The resonance interaction between different isotopes is computed in Speng by the same method as was used in the Dorix programme for non-correlated unresolved resonances. Consideration is also given to the width of the resonances compared to the energy loss by a neutron colliding with some of the scattering elements

The Multigroup Neutron Diffusion Equations/1 Space Dimension

Energy Technology Data Exchange (ETDEWEB)

Linde, Sven

1960-06-15

A description is given of a program for the Ferranti Mercury computer which solves the one-dimensional multigroup diffusion equations in plane, cylindrical or spherical geometry, and also approximates automatically a two-dimensional solution by separating the space variables. In section A the method of calculation is outlined and the preparation of data for two group problems is described. The spatial separation of two-dimensional equations is considered in section B. Section C covers the multigroup equations. These parts are self contained and include all information required for the use of the program. Details of the numerical methods are given in section D. Three sample problems are solved in section E. Punching and operating instructions are given in an appendix.

The Multigroup Neutron Diffusion Equations/1 Space Dimension

International Nuclear Information System (INIS)

Linde, Sven

1960-06-01

A description is given of a program for the Ferranti Mercury computer which solves the one-dimensional multigroup diffusion equations in plane, cylindrical or spherical geometry, and also approximates automatically a two-dimensional solution by separating the space variables. In section A the method of calculation is outlined and the preparation of data for two group problems is described. The spatial separation of two-dimensional equations is considered in section B. Section C covers the multigroup equations. These parts are self contained and include all information required for the use of the program. Details of the numerical methods are given in section D. Three sample problems are solved in section E. Punching and operating instructions are given in an appendix

Skyshine analysis using various nuclear data files

Energy Technology Data Exchange (ETDEWEB)

Zharkov, V.P.; Dikareva, O.F.; Kartashev, I.A.; Kiselev, A.N. [Research and Development Inst. of Power Engineering, Moscow (Russian Federation); Nomura, Y.; Tsubosaka, A. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

2000-03-01

The calculations of the spacial distributions of dose rate for neutron and secondary photons, thermal neutron fluxes and space-energy distributions of neutron and photons near the air-ground interface were performed by MCNP and DORT codes. Different nuclear data files were used (ENDF/B-IV, ENDF/B-VI, FENDL-2, JENDL-3.2). Either the standard pointwise libraries (MCNP) or special libraries prepared by NJOY code from ENDF/B and others' files were used. Prepared multigroup coupled neutron and photon cross sections libraries for DORT code had CASK-40 group energy structures. The libraries contain pointwise or multigroup cross sections data for all elements included in the atmosphere and ground composition. The validation of the calculated results was performed with using the experimental data obtained for the series of measurements at RA reactor. (author)

NJOY-94, General ENDF/B Processing System for Reactor Design Problems

International Nuclear Information System (INIS)

1997-01-01

1 - Description of program or function: The NJOY nuclear data processing system is a comprehensive computer code system for producing pointwise and multigroup cross sections and related quantities from ENDF/B evaluated nuclear data in the ENDF format, including the latest US library, ENDF/B-VI. The NJOY code works with neutrons, photons, and charged particles and produces libraries for a wide variety of particle transport and reactor analysis codes. It is capable of processing data in ENDF/B-4, ENDF/B-5, and ENDF/B-6 formats for evaluated data. Short descriptions of the different modules follow: RECONR Reconstructs pointwise cross sections from ENDF/B resonance parameters and interpolation schemes. BROADR Doppler broadens and thins pointwise cross sections. UNRESR Computes effective self-shielded pointwise cross sections in the unresolved-resonance region. HEATR Generates pointwise heat production cross sections and radiation-damage-energy production cross sections. THERMR Produces incoherent inelastic energy-to-energy matrices for free or bound scatterers, coherent elastic cross sections for hexagonal materials, and incoherent elastic cross sections. GROUPR Generates self-shielded multigroup cross sections, group- to-group neutron scattering matrices, and photon production matrices from pointwise input. GAMINR Calculates multigroup photon interaction cross sections and KERMA factors and group-to-group photon scattering matrices. ERRORR Produces multigroup covariance matrices from ENDF/B uncertainties. COVR Reads the output of ERRORR and performs covariance plotting and output-formatting operations. DTFR Formats multigroup data for transport codes such as DTF-IV and ANISN. CCCCR Formats multigroup data for the CCCC standard interface files ISOTXS, BRKOXS, and DLAYXS. MATXSR Formats multigroup data for the MATXS cross section interface file. ACER Prepares libraries for the Los Alamos continuous-energy Monte Carlo code MCNP. POWR Prepares libraries for the EPRI

The correction of pebble bed reactor nodal cross sections for the effects of leakage and depletion history

Science.gov (United States)

Hudson, Nathanael Harrison

An accurate and computationally fast method to generate nodal cross sections for the Pebble Bed Reactor (PBR) was presented. In this method, named Spectral History Correction (SHC), a set of fine group microscopic cross section libraries, pre-computed at specified depletion and moderation states, was coupled with the nodal nuclide densities and group bucklings to compute the new fine group spectrum for each node. The relevant fine group cross-section library was then recollapsed to the local broad group cross-section structure with this new fine group spectrum. This library set was tracked in terms of fuel isotopic densities. Fine group modulation factors (to correct the homogeneous flux for heterogeneous effects) and fission spectra were also stored with the cross section library. As the PBR simulation converges to a steady state fuel cycle, the initial nodal cross section library becomes inaccurate due to the burnup of the fuel and the neutron leakage into and out of the node. Because of the recirculation of discharged fuel pebbles with fresh fuel pebbles, a node can consist of a collection of pebbles at various burnup stages. To account for the nodal burnup, the microscopic cross sections were combined with nodal averaged atom densities to approximate the fine group macroscopic cross-sections for that node. These constructed, homogeneous macroscopic cross sections within the node were used to calculate a numerical solution for the fine group spectrum with B1 theory. This new fine spectrum was used to collapse the pre-computed microscopic cross section library to the broad group structure employed by the fuel cycle code. This SHC technique was developed and practically implemented as a subroutine within the PBR fuel cycle code PEBBED. The SHC subroutine was called to recalculate the broad group cross sections during the code convergence. The result was a fast method that compared favorably to the benchmark scheme of cross section calculation with the lattice

ENDF/B-5 fission product cross section evaluations

International Nuclear Information System (INIS)

Schenter, R.E.; England, T.R.

1979-12-01

Cross section evaluations were made for the 196 fission product nuclides on the ENDF/B-5 data files. Most of the evaluations involve updating the capture cross sections of the important absorbers for fast and thermal reactor systems. This included updating thermal values, resonance integrals, resonance parameter sets, and fast capture cross sections. For the fast capture results generalized least-squares calculations were made with the computer code FERRET. Input for these cross section adjustments included nuclear models calculations and both integral and differential experimental data results. The differential cross sections and their uncertainties were obtained from the CSIRS library. Integral measurement results came from CFRMF and STEK Assemblies 500, 1000, 2000, 3000, 4000. Comparisons of these evaluations with recent capture measurements are shown. 15 figures, 10 tables

User manual for version 4.3 of the Tripoli-4 Monte-Carlo method particle transport computer code; Notice d'utilisation du code Tripoli-4, version 4.3: code de transport de particules par la methode de Monte Carlo

Energy Technology Data Exchange (ETDEWEB)

Both, J.P.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B

2003-07-01

This manual relates to Version 4.3 TRIPOLI-4 code. TRIPOLI-4 is a computer code simulating the transport of neutrons, photons, electrons and positrons. It can be used for radiation shielding calculations (long-distance propagation with flux attenuation in non-multiplying media) and neutronic calculations (fissile medium, criticality or sub-criticality basis). This makes it possible to calculate k{sub eff} (for criticality), flux, currents, reaction rates and multi-group cross-sections. TRIPOLI-4 is a three-dimensional code that uses the Monte-Carlo method. It allows for point-wise description in terms of energy of cross-sections and multi-group homogenized cross-sections and features two modes of geometrical representation: surface and combinatorial. The code uses cross-section libraries in ENDF/B format (such as JEF2-2, ENDF/B-VI and JENDL) for point-wise description cross-sections in APOTRIM format (from the APOLLO2 code) or a format specific to TRIPOLI-4 for multi-group description. (authors)

User manual for version 4.3 of the Tripoli-4 Monte-Carlo method particle transport computer code

International Nuclear Information System (INIS)

Both, J.P.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B.

2003-01-01

This manual relates to Version 4.3 TRIPOLI-4 code. TRIPOLI-4 is a computer code simulating the transport of neutrons, photons, electrons and positrons. It can be used for radiation shielding calculations (long-distance propagation with flux attenuation in non-multiplying media) and neutronic calculations (fissile medium, criticality or sub-criticality basis). This makes it possible to calculate k eff (for criticality), flux, currents, reaction rates and multi-group cross-sections. TRIPOLI-4 is a three-dimensional code that uses the Monte-Carlo method. It allows for point-wise description in terms of energy of cross-sections and multi-group homogenized cross-sections and features two modes of geometrical representation: surface and combinatorial. The code uses cross-section libraries in ENDF/B format (such as JEF2-2, ENDF/B-VI and JENDL) for point-wise description cross-sections in APOTRIM format (from the APOLLO2 code) or a format specific to TRIPOLI-4 for multi-group description. (authors)

View-CXS neutron and photon cross-sections viewer

International Nuclear Information System (INIS)

Subbaiah, K.V.; Sunil Sunny, C.

2004-01-01

A graphical user-friendly interface is developed in Visual Basic (VB)-6 to view the variation of neutron and photon interaction cross-sections of different isotopes as a function of energy. VB subroutines developed read the binary data files of cross-sections created in MCNP-ACE (Briesmeister, J.F., 1993. MCNP - a general purpose Monte Carlo N-Particle Transport code. Version 4A. LANL, USA), ANISN-DLC (Engle W.W. Jr., 1967, A User's Manual for ANISN, K-1693; ORNL, 1974. 100 group neutron cross section data based on ENDF/B-III. Oak Ridge National Laboratory, USA) and KENO-AMPX (Petrie, L.M., Landers, N.F., 1984 KENO-Va- An Improved Monte Carlo Criticality Program with Super Grouping. RSICC-CCC-548, USA) formats using LAHEY-77 Fortran Compiler. The information on isotopes present in each library will be displayed with the help of database files prepared using Micro-Soft ACESS. The cross-section data can be viewed in different presentation styles namely, line graphs, bar graphs, histograms etc., with different color and symbol options. The cross-section plots generated can be saved as Bit-Map file to embed in any other text files. This software enables inter comparison of cross-sections from different type of libraries for isotopes as well as mixtures. Provision is made to view the cross-sections for nuclear reactions such as (n,γ), (n,f), (n,α), etc. The software can be obtained from Radiation Safety Information and Computational Centre (RSICC), ORNL, USA with the code package identification number PSR-514. The software package needs a hard disk space of about 80 MB when installed and works in WINDOWS-95/98/2000 operating systems

Two-dimensional shielding benchmarks for iron at YAYOI, (1)

International Nuclear Information System (INIS)

Oka, Yoshiaki; An, Shigehiro; Kasai, Shigeru; Miyasaka, Shun-ichi; Koyama, Kinji.

The aim of this work is to assess the collapsed neutron and gamma multigroup cross sections for two dimensional discrete ordinate transport code. Two dimensional distributions of neutron flux and gamma ray dose through a 70cm thick and 94cm square iron shield were measured at the fast neutron source reactor ''YAYOI''. The iron shield was placed over the lead reflector in the vertical experimental column surrounded by heavy concrete wall. The detectors used in this experiment were threshold detectors In, Ni, Al, Mg, Fe and Zn, sandwitch resonance detectors Au, W and Co, activation foils Au for neutrons and thermoluminescence detectors for gamma ray dose. The experimental results were compared with the calculated ones by the discrete ordinate transport code ANISN and TWOTRAN. The region-wise, coupled neutron-gamma multigroup cross-sections (100n+20gamma, EURLIB structure) were generated from ENDF/B-IV library for neutrons and POPOP4 library for gamma-ray production cross-sections by using the code system RADHEAT. The effective microscopic neutron cross sections were obtained from the infinite dilution values applying ABBN type self-shielding factors. The gamma ray production multigroup cross-sections were calculated from these effective microscopic neutron cross-sections. For two-dimensional calculations the group constants were collapsed into 10 neutron groups and 3 gamma groups by using ANISN. (auth.)

The CNCSN: one, two- and three-dimensional coupled neutral and charged particle discrete ordinates code package

International Nuclear Information System (INIS)

Voloschenko, A.M.; Gukov, S.V.; Kryuchkov, V.P.; Dubinin, A.A.; Sumaneev, O.V.

2005-01-01

The CNCSN package is composed of the following codes: -) KATRIN-2.0: a three-dimensional neutral and charged particle transport code; -) KASKAD-S-2.5: a two-dimensional neutral and charged particle transport code; -) ROZ-6.6: a one-dimensional neutral and charged particle transport code; -) ARVES-2.5: a preprocessor for the working macroscopic cross-section format FMAC-M for transport calculations; -) MIXERM: a utility code for preparing mixtures on the base of multigroup cross-section libraries in ANISN format; -) CEPXS-BFP: a version of the Sandia National Lab. multigroup coupled electron-photon cross-section generating code CEPXS, adapted for solving the charged particles transport in the Boltzmann-Fokker-Planck formulation with the use of discrete ordinate method; -) SADCO-2.4: Institute for High-Energy Physics modular system for generating coupled nuclear data libraries to provide high-energy particles transport calculations by multigroup method; -) KATRIF: the post-processor for the KATRIN code; -) KASF: the post-processor for the KASKAD-S code; and ROZ6F: the post-processor for the ROZ-6 code. The coding language is Fortran-90

Parametric equations for calculation of macroscopic cross sections

International Nuclear Information System (INIS)

Botelho, Mario Hugo; Carvalho, Fernando

2015-01-01

Neutronic calculations of the core of a nuclear reactor is one thing necessary and important for the design and management of a nuclear reactor in order to prevent accidents and control the reactor efficiently as possible. To perform these calculations a library of nuclear data, including cross sections is required. Currently, to obtain a cross section computer codes are used, which require a large amount of processing time and computer memory. This paper proposes the calculation of macroscopic cross section through the development of parametric equations. The paper illustrates the proposal for the case of macroscopic cross sections of absorption (Σa), which was chosen due to its greater complexity among other cross sections. Parametric equations created enable, quick and dynamic way, the determination of absorption cross sections, enabling the use of them in calculations of reactors. The results show efficient when compared with the absorption cross sections obtained by the ALPHA 8.8.1 code. The differences between the cross sections are less than 2% for group 2 and less than 0.60% for group 1. (author)

Measurements of D-T neutron induced radioactivity in plasma-facing materials and their role in qualification of activation cross-section libraries and codes

International Nuclear Information System (INIS)

Kumar, A.; Abdou, M.A.; Kosako, K.; Oyama, Y.; Nakamura, T.; Maekawa, H.

1995-01-01

The D-T neutron-induced radioactivity constitutes one of the foremost issues in fusion reactor design. The validation of activation cross-sections and decay data libraries is one of the important requirements for validating ITER design from safety and waste disposal viewpoints. An elaborate, experimental program was initiated in 1988, under USDOE-JAERI collaborative program, to validate the radioactivity codes/libraries. The measurements of decay-γ spectra from irradiated, high purity samples of Al, Si, Ti, V, Cr, Mn-Cu alloy, Fe, Co, Ni, Cu, stainless steel 316 (AISI 316), Zn, Zr, Nb, Mo, In, Sn, Ta, W, and Pb, among others, were conducted under D-T neutron fluences varying from 1.6 x 10 10 ncm -2 to 6.1 x 10 13 ncm -2 . As many as 14 neutron energy spectra were covered for a number of materials. The analysis of isotopic activities of the irradiated materials using activation cross-section libraries of four leading radioactivity codes, i.e. ACT4/THIDA-2, REAC-3, DKR-ICF, and RACC, has shown large discrepancies among the calculations, on the one hand, and between the calculations and the measurements, on the other. A discussion is also presented on definition and obtention of safety cum quality factors for various activation libraries. (orig.)

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

Simplified polynomial representation of cross sections for reactor calculation

International Nuclear Information System (INIS)

Dias, A.M.; Sakai, M.

1985-01-01

It is shown a simplified representation of a cross section library generated by transport theory using the cell model of Wigner-Seitz for typical PWR fuel elements. The effect of burnup evolution through tables of reference cross sections and the effect of the variation of the reactor operation parameters considered by adjusted polynomials are presented. (M.C.K.) [pt

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)

Index of Nuclear Data Libraries available from the IAEA Nuclear Data Section

International Nuclear Information System (INIS)

Lemmel, H.D.

1994-01-01

This document lists more than 100 nuclear data libraries together with references that give more detailed information about these libraries. The data libraries include neutron cross-sections, resonance parameters, fission-product yields, nuclear structure and decay data, gamma-rays from radionuclides, data of nuclear reactions induced by charged particles or heavy ions, photonuclear data, photoatomic interaction data, and many others, partly with related data processing computer codes. All data and documentation references are available upon request from the IAEA Nuclear Data Section, free of charge on magnetic tape, PC diskettes, or through the online Nuclear Data Information System (NDIS). (author)

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

TIMS-1, Multigroup Cross-Sections of Heavy Isotope Mixture with Resonance from ENDF/B

International Nuclear Information System (INIS)

Takano, Hideki; Ishiguro, Yukio; Matsui, Yasushi

1984-01-01

1 - Description of problem or function: TIMS-1 is a code for calculating the group constants of heavy resonant nuclei by using ENDF/ B-4 format data. This code calculates infinitely dilute cross sections and self-shielding factors as a function of composition sigma-0 temperature T and R-parameter, where R is the ratio of ato- mic number density of two different resonant nuclei. 2 - Method of solution: In the unresolved resonance region, a ladder of resonance parameters and levels is generated with Monte Carlo method. The temperature dependent cross sections are calculated with the Breit-Wigner single-level and multi-level formula. The neutron spectrum is accurately calculated by solving numerically the neutron slowing down equation using a recurrence formula for neutron slowing down source. 3 - Restrictions on the complexity of the problem: The maximum numbers of energy groups, temperatures and compositions are 60, 4 and 10 respectively

Mechanized evaluation of neutron cross-sections

International Nuclear Information System (INIS)

Horsley, A.; Parker, J.B.

1967-01-01

The evaluation work to provide accurate and consistent neutron cross-section data for multigroup neutronics calculations is not fully exploiting the available theoretical and experimental results; this has been so particularly since the introduction of on-line data handling techniques enabled experimenters to turn out vast quantities of numbers. This situation can be radically improved only by mechanizing the evaluation processes. Systems such as the SC1SRS tape will not only largely overcome the task of collecting data but will provide speedy access to it; by using computers and graph-plotting machines to tabulate and display this data, the labour of evaluation can be very greatly reduced. With some types of cross-section there is hope that by using modern curve-fitting techniques the actual evaluation and statistical accounting of the data can be performed automatically. Some areas where automatic evaluation would seem likely to succeed are specified and a discussion of the mathematical difficulties incurred, such as the elimination of anomalous data, is given. Particularly promising is the use of splines in the mechanized evaluation of data. Splines are the mathematical analogues of the draughtsman's spline used in drawing smooth curves. Their principal properties are the excellent approximations they give to the derivatives of a function; in contrast to conventional polynomial fitting, this feature ensures good interpolation and, when required, stable extrapolation. Various methods of using splines in data graduation and the problem of marrying these methods to standard statistical procedures are examined. The results of work done at AWRE with cubic splines on the mechanized evaluation of neutron scattering total cross-section and angular distribution data are presented. (author)

Evaluation of ETOG-3Q, ETOG-3, FLANGE-II, XLACS, NJOY and LINEAR/RECENT/GROUPIE computer codes concerning to the resonance contribution and background cross sections

International Nuclear Information System (INIS)

Anaf, J.; Chalhoub, E.S.

1988-12-01

The NJOY and LINEAR/RECENT/GROUPIE calculational procedures for the resolved and unresolved resonance contributions and background cross sections are evaluated. Elastic scattering, fission and capture multigroup cross sections generated by these codes and the previously validated ETOG-3Q, ETOG-3, FLANGE-II and XLACS are compared. Constant weighting function and zero Kelvin temperature are considered. Discrepancies are presented and analysed. (author) [pt