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Title: ENDF/B-V, ENDF/B-VI, AND ENDF/B-VII.0 RESULTS FOR THE DOPPLER-DEFECT BENCHMARK (U)

Abstract

A set of computational benchmarks for the Doppler reactivity defect has been specified for an infinite array of identical fuel pin cells containing normal or enriched UO{sub 2} fuel, reactor-recycle mixed-oxide (MOX) fuel, or weapons-grade MOX fuel. The Doppler coefficient of reactivity, as well as the Doppler defect, can be computed for each of the cells. The MCNP5 Monte Carlo code was used to perform calculations for these benchmarks using cross sections derived from the ENDF/B-V, ENDF/B-VI, and ENDF/B-VII.0 nuclear data sets. The Doppler coefficients obtained from the three data sets exhibit very similar behavior. The Doppler coefficient for UO{sub 2} fuel becomes less negative with increasing enrichment, with a generally asymptotic shape. The Doppler coefficient for the reactor-recycle MOX becomes less negative with increasing PuO{sub 2} content but exhibits less curvature than that for UO{sub 2} fuel. The Doppler coefficient for weapons-grade MOX shows a pronounced shoulder between 1 wt.% and 2 wt.% PuO{sub 2}, with a nearly constant value thereafter. The Doppler coefficient for heavily loaded MOX fuel, whether reactor-recycle or weapons-grade, is significantly more negative than that for highly enriched UO{sub 2} fuel.

Authors:
 [1]
  1. Los Alamos National Laboratory
Publication Date:
Research Org.:
Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
OSTI Identifier:
986512
Report Number(s):
LA-UR-07-0922
TRN: US1006372
DOE Contract Number:
AC52-06NA25396
Resource Type:
Conference
Resource Relation:
Conference: INT'L TOPICAL MEETING ON MATH & COMPUTATION & SUPERCOMPUTI ; 200704 ; MONTEREY
Country of Publication:
United States
Language:
English
Subject:
22; BENCHMARKS; CROSS SECTIONS; DEFECTS; DOPPLER COEFFICIENT; FUEL PINS; SHAPE; SUPERCOMPUTERS

Citation Formats

MOSTELLER, RUSSELL D. ENDF/B-V, ENDF/B-VI, AND ENDF/B-VII.0 RESULTS FOR THE DOPPLER-DEFECT BENCHMARK (U). United States: N. p., 2007. Web.
MOSTELLER, RUSSELL D. ENDF/B-V, ENDF/B-VI, AND ENDF/B-VII.0 RESULTS FOR THE DOPPLER-DEFECT BENCHMARK (U). United States.
MOSTELLER, RUSSELL D. Fri . "ENDF/B-V, ENDF/B-VI, AND ENDF/B-VII.0 RESULTS FOR THE DOPPLER-DEFECT BENCHMARK (U)". United States. doi:. https://www.osti.gov/servlets/purl/986512.
@article{osti_986512,
title = {ENDF/B-V, ENDF/B-VI, AND ENDF/B-VII.0 RESULTS FOR THE DOPPLER-DEFECT BENCHMARK (U)},
author = {MOSTELLER, RUSSELL D.},
abstractNote = {A set of computational benchmarks for the Doppler reactivity defect has been specified for an infinite array of identical fuel pin cells containing normal or enriched UO{sub 2} fuel, reactor-recycle mixed-oxide (MOX) fuel, or weapons-grade MOX fuel. The Doppler coefficient of reactivity, as well as the Doppler defect, can be computed for each of the cells. The MCNP5 Monte Carlo code was used to perform calculations for these benchmarks using cross sections derived from the ENDF/B-V, ENDF/B-VI, and ENDF/B-VII.0 nuclear data sets. The Doppler coefficients obtained from the three data sets exhibit very similar behavior. The Doppler coefficient for UO{sub 2} fuel becomes less negative with increasing enrichment, with a generally asymptotic shape. The Doppler coefficient for the reactor-recycle MOX becomes less negative with increasing PuO{sub 2} content but exhibits less curvature than that for UO{sub 2} fuel. The Doppler coefficient for weapons-grade MOX shows a pronounced shoulder between 1 wt.% and 2 wt.% PuO{sub 2}, with a nearly constant value thereafter. The Doppler coefficient for heavily loaded MOX fuel, whether reactor-recycle or weapons-grade, is significantly more negative than that for highly enriched UO{sub 2} fuel.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Fri Feb 09 00:00:00 EST 2007},
month = {Fri Feb 09 00:00:00 EST 2007}
}

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  • Previous studies have indicated that ENDF/B-VII preliminary releases {beta}-2 and {beta}-3, predecessors to the recent initial release of ENDF/B-VII.0, produce significantly better overall agreement with criticality benchmarks than does ENDF/B-VI. However, one of those studies also suggests that improvements still may be needed for thermal plutonium cross sections. The current study substantiates that concern by examining criticality benchmarks for unreflected spheres of plutonium-nitrate solutions and for slightly and heavily borated mixed-oxide (MOX) lattices. Results are presented for the JEFF-3.1 and JENDL-3.3 nuclear data libraries as well as ENDF/B-VII.0 and ENDF/B-VI. It is shown that ENDF/B-VII.0 tends to overpredict reactivity formore » thermal plutonium benchmarks over at least a portion of the thermal range. In addition, it is found that additional benchmark data are needed for the deep thermal range.« less
  • The fifth version of the Evaluated Nuclear Data File (ENDF/B-V) was released in 1979, with a significant updating in 1981. This file was limited to users in the US and with Atomic Energy of Canada Limited. The first version of the Joint Evaluated File (JEF-1), created by collaboration among laboratories in Europe and Japan and the Nuclear Energy Agency Data Bank, was released in 1986; this file was not available in the US. The sixth version of ENDF/B is in preparation (ENDF/B-VI). This file will be available to all requestors. From comparisons of various calculated and measured values, it appearsmore » that the ENDF/B-V library constitutes a very good data base for LWR analysis. Thus, there is considerable interest in a comparison of thermal lattice integral parameters calculated with JEF-1 and ENDF/B-V; such comparisons have been made. However, in all such comparisons known to the authors, the calculations with the two data sets were performed at two different US and European laboratories with different codes. Such comparisons have a degree of uncertainty concerning the performance of the two evaluated files. In the joint investigations at the Georgia Institute of Technology and the Paul Scherrer Institute, such uncertainty in the comparison discussed above was eliminated by using the same cross-section processing codes and cell codes of the PSI light water reactor (LWR) code system ELCOS and identical code input at both installations. An additional features of this study is an investigation of the influence on calculated LWR parameters of the ENDF/B-VI modifications to the data for H[sub 2]O constituents.« less
  • Calculations for the ANS UO{sub 2} lattice benchmark have been performed with the MCNP Monte Carlo code and its ENDF/B-V and ENDF/B-VI continuous-energy libraries. The ENDF/B-V library produces significantly better agreement with the benchmark value for k{sub eff} than do the ENDF/B-VI libraries. However, the pin power distributions are essentially the same irrespective of the library.
  • Calculations for the ANS UO{sub 2} lattice benchmark have been performed with the MCNP Monte Carlo code and its ENDF/B-V and EnDF/B-VI continuous-energy libraries. Similar calculations were performed previously for the experiments upon which these benchmarks are based, using continuous-energy libraries derived from EnDF/B-V and from Release 2 of EnDF/B-VI (ENDF/B-VI.2). This study extends those calculations to the infinite-lattice configurations given in the benchmark specifications and also includes results from Release 3 of EnDF/B-VI (ENDF/B-VI.3) for both the core and infinite-lattice configurations. For this set of benchmarks, the only significant difference between the ENDF/B-VI.2 and EnDF/B-VI.3 libraries is the cross-sectionmore » behavior of {sup 235}U. EnDF/B-VI.3 contains revised cross sections for {sup 235}U below 900 eV, although those changes principally affect the range below 110 eV. In particular, relative to EnDF/B-VI.2, EnDF/B-VI.3 increases the epithermal capture-to-fission ratio for {sup 235}U and slightly increases its thermal fission cross section.« less
  • An initial assessment of the impact of preliminary data proposed for ENDF/B-VII has been made using the MCNP criticality validation suite. Relative to ENDF/B-VI, the data changes primarily involve high-energy elastic and inelastic scattering in the uranium isotopes and {sup 239}Pu, as well as resonance parameters for {sup 238}U. The criticality validation suite is a collection of 31 benchmarks taken from the International Handbook of Evaluated Criticality Benchmark Experiments. It contains cases for a variety of fuels, including {sup 233}U, highly enriched uranium (HEU), intermediate-enriched uranium (IEU), low-enriched uranium (LEU), and plutonium. For each fuel type, there are cases withmore » a variety of moderators, reflectors, spectra, and geometries. The cases in the suite are summarized in Table I. Three sets of calculations were performed for the suite using the MCNP5 Monte Carlo code. The first set employed nuclear data from ENDF/B-VI Release 8, the final release for ENDF/B-VI. The second set employed preliminary ENDF/B-VII data generated by group T-16 at Los Alamos National Laboratory for the uranium isotopes and for {sup 239}Pu but retained ENDF/B-VI data for all other nuclides. The third set was the same as the second except that a new set of {sup 238}U resonance parameters generated by researchers at Oak Ridge National Laboratory (ORNL) was added to the T-16 evaluation. The MCNPS calculations were run with 5,000,000 active neutron histories for all but two cases in the suite. Only 3,000,000 active histories were used for those cases, SB-5 and Zebra-8H, because they require substantially more computer time per history than the other cases. Nonetheless, the standard deviation for k{sub eff} from those cases is comparable to those for other cases in the suite. The results from these calculations are presented in Table II. The preliminary ENDF/B-VII data produce marked improvements in k{sub eff} for bare spheres of {sup 233}U (Jezebel-233), HEU (Godiva), and plutonium (Jezebel and Jezebel-240) as well as the other unreflected HEU and plutonium cases (Tinkertoy02 (c-11) and Pu Buttons, respectively). Furthermore, the reactivity swings between those bare spheres and the corresponding Flattop cases (which enclose the sphere inside an annulus of normal uranium) are substantially decreased. The changes also significantly improve k{sub eff} for BIG TEN and for HEU and plutonium spheres immersed in water (Godiver and Pu-MF-11, respectively). In addition, inclusion of the ORNL resonance parameters for {sup 238}U produces a significantly better value for k{sub eff} for B&W XI (2), a lattice of LEU fuel pins in water, and ORNL-II, an unreflected sphere of uranyl nitrate solution enriched in {sup 233}U. At the same time, the preliminary ENDF/B-VII data produce worse results for thermal lattices of {sup 233}U and HEU pins in water (SB-2 1/2 and SB-5, respectively), for the bare IEU sphere (IEU-MF-03) and the IEU sphere reflected by graphite (IEU-MF-04), for a plutonium sphere reflected by thorium (THOR), and for a MOX lattice in water (PNL-33). Furthermore, k{sub eff} for the uranium cases with intermediate spectra remains substantially underpredicted, while k{sub eff} for the plutonium case with an intermediate spectrum (HISS/HPG) continues to be significantly over-predicted. In conclusion, preliminary ENDF/B-VII data for the uranium isotopes and {sup 239}Pu produce improvements for most of the cases with fast spectra, BIG TEN, and the lattice of LEU fuel pins in water. However, improvements still are needed in some areas, particularly those cases with intermediate spectra.« less