First Conclusions of the WPEC/Subgroup-22 Nuclear Data for Improved LEU-LWR Reactivity Predictions

Courcelle, Arnaud; Wpec,; Subgroup,

doi:10.1063/1.1945048

Cited by 11 publications

(11 citation statements)

References 16 publications

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“…12. These trends are explained by the previous observations 28) that the continuous energy Monte Carlo calculations with major nuclear data libraries give k eff 's of the LWR MOX cores that are slightly larger than those of the LWR UO 2 cores in the EPICURE, MISTRAL and BASALA experimental programs, where long-timeelapsed MOX rods were used for the MOX cores and contained larger amounts of 241 Am originating from the decay of 241 Pu. Lance et al 6) obtained À370 and À340 pcm in the comparison of their calculated and experimental critical heights with WIMS8A-THREEDANT and -KENO-Va, respectively, which correspond to k eff ¼ 1:00370 and 1.00340 for the core with the critical core height, and their study showed the same trend of slightly larger k eff 's for the fresh BR3 MOX core than for the reference core as in this study.…”

Section: D) Burnup Calculations Of Br3 Mox Fuel Rodssupporting

confidence: 48%

Analysis of Core Physics Experiments on Fresh and Irradiated BR3 MOX Fuel in REBUS Program

Yamamoto

Ando

et al. 2009

J. Nucl. Sci. Technol.

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As part of an international experimental program REBUS, core physics experiments have been implemented on a UO 2 core, which consists of 3.3 and 4.0 wt% UO 2 fuel rods in a square pitch of 1.26 cm, and two partial MOX cores, which replace 7 Â 7 UO 2 fuel rods in the center of the UO 2 core by fuel bundles made of fresh BR3 MOX fuel or irradiated BR3 MOX fuel with an average burnup of 20 GWd/t. Burnup calculations of the BR3 MOX fuel were performed using a general-purpose neutronic calculation code SRAC, and core calculations of the three critical cores were carried out using SRAC, a transport calculation code THREEDANT, and a continuous-energy Monte Carlo code MVP. The measured inventories of major U and Pu isotopes on a sample taken from the BR3 MOX fuel agree with the results of the burnup calculations within 3% deviation. The k eff 's of the three cores are from 0.985 to 1.002. The measured burnup reactivity of the irradiated BR3 MOX fuel was well reproduced by the three types of core calculations. The influence of the accuracy of the inventory calculations on burnup reactivity was studied by comparing between the calculated and measured inventories. The result indicates that the biases in the inventory and reactivity calculations compensate each other, and it makes the total biases of the burnup reactivity small.

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Section: D) Burnup Calculations Of Br3 Mox Fuel Rodssupporting

confidence: 48%

Analysis of Core Physics Experiments on Fresh and Irradiated BR3 MOX Fuel in REBUS Program

Yamamoto

Ando

et al. 2009

J. Nucl. Sci. Technol.

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“…2). WPEC SG22 [2] has contributed significantly to these improvements. Very extensive automated benchmarking tests have been set up, which probe the current quality of the new data file.…”

Section: Status Of Jeff-30mentioning

confidence: 99%

Status of the JEFF Nuclear Data Library

Koning¹,

Bersillon²,

Forrest³

et al. 2005

AIP Conference Proceedings

108

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The status of the Joint Evaluated Fission and Fusion file (JEFF) is described. The next version of the library, JEFF-3.1, comprises a significant update of actinide evaluations, evaluations emerging from European nuclear data projects, the activation library JEFF-3/A, the decay data and fission yield library, and fusion-related data files from the EFF project. The revisions were motivated by the availability of new measurements, modelling capabilities, or trends from integral experiments. Various pre-release validation efforts are underway, mainly for criticality and shielding of thermal and fast systems. This JEFF-3.1 library is expected to provide improved performances with respect to previous releases for a variety of scientific and industrial applications.

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“…In recent years, there has been a renewed focus on the issue of under-predicted reactivity in LEU systems, culminating in an NEA sponsored international collaboration, ref. [7], that has successfully resolved this issue. This is illustrated by the data in the ENDB/B-VII.0 column of table 2, where significant increases in calculated eigenvalues are observed.…”

Section: Results For Low-enriched Uo 2 Lattice Systemsmentioning

confidence: 96%

ENDF/B-VII data testing with ICSBEP benchmarks

Kahler

MacFarlane

2007

Nd2007

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Abstract. Continuous energy Monte Carlo eigenvalue calculations have been performed for several hundred critical benchmarks described in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. These calculations were performed with MCNP5 using either ENDF/B-VI.8 or ENDF/B-VII.0 cross sections. ENDF/B-VII cross section data files yield significantly more accurate calculated eigenvalues than those obtained with ENDF/B-VI.8 cross sections for moderated, low-enriched uranium fuel rod lattice configurations and for unmoderated, bare or reflected, critical benchmark assemblies. Accurate calculated eigenvalues were previously obtained with ENDF/B-VI.8 cross sections for both highly-enriched and low-enriched uranium solution assemblies. These accurate eigenvalues are retained with ENDF/B-VII.0 cross sections.

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First Conclusions of the WPEC/Subgroup-22 Nuclear Data for Improved LEU-LWR Reactivity Predictions

Cited by 11 publications

References 16 publications

Analysis of Core Physics Experiments on Fresh and Irradiated BR3 MOX Fuel in REBUS Program

Analysis of Core Physics Experiments on Fresh and Irradiated BR3 MOX Fuel in REBUS Program

Status of the JEFF Nuclear Data Library

ENDF/B-VII data testing with ICSBEP benchmarks

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