Development of the multigroup cross section library for the CASL neutronics simulator MPACT: Verification

Kim, Kang Seog; Gentry, Cole; Godfrey, Andrew; Liu, Yuxuan; Palmtag, Scott

doi:10.1016/j.anucene.2019.03.041

Cited by 18 publications

(15 citation statements)

References 4 publications

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“…MPACT is an advanced pin-resolved whole-core multi-group deterministic neutron transport capability based on the 2-D/1-D synthesis method, on the frame of a 3-D coarse mesh finite difference method, for which radial and axial correction factors are obtained from 2-D method-ofcharacteristics and 1-D PN, respectively [8]. The transport is performed using 51 energy-group cross sections, based on the subgroup method of on-the-fly resonance self-shielding [9]. The discretization of the core is typically three radial and eight azimuthal flat source regions per fuel pellet (See Fig.…”

Section: Mpactmentioning

confidence: 99%

Watts Bar Unit 2 Startup Results with VERA

Godfrey

Collins

Gentry

et al. 2017

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CASL members TVA, Westinghouse, and Oak Ridge National Laboratory have successfully completed a detailed simulation of the initial startup of Watts Bar Nuclear Unit 2 (WBN2) using the advanced reactor simulation tools known as VERA. WBN2 is the first commercial power reactor to join the nation's electrical grid in over two decades, and the modern core design and availability of data make it an excellent benchmark for CASL. Calculations were performed three months prior to the startup, and in the first blind application of VERA to a new reactor, predicted criticality and physics parameters very close to those later measured by TVA. Subsequent calculations with the latest version of VERA and using exact measurement conditions improved the results even further. The escalation to full power required approximately five months, including several intermediate testing power plateaus. The entire power, temperature, and control rod history was simulated with VERA by the hour, requiring 4,130 time steps, and included isotopic depletion and decay through ten additional shutdown intervals. TVA provided the startup data, as well as measured boron concentrations, reactor temperatures, ex-core axial flux difference (AFD), and twelve measured incore neutron flux distributions. The entire simulation required 892,837 core-hours, or 13.5 days on 2,784 cores. This included 16,605 neutronics/thermal-hydraulic iterations.

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Section: Mpactmentioning

confidence: 99%

Watts Bar Unit 2 Startup Results with VERA

Godfrey

Collins

Gentry

et al. 2017

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“…MPACT is an advanced pin-resolved, whole-core, multigroup deterministic neutron transport capability based on the 2D/1D synthesis method, on the frame of a 3D coarse mesh finite difference (CMFD) method, for which radial and axial correction factors are obtained from 2D MOC and 1D P N , respectively [2]. The transport is performed using 51 energy group XSs, based on the subgroup method of on-the-fly resonance self-shielding [13]. The discretization of the core is typically three radial and eight azimuthal flat source regions per fuel pellet at each of approximately 60 axial planes, explicitly treating such features as spacer grids, fuel and absorber plena, and end plugs.…”

Section: System Requirementsmentioning

confidence: 99%

VERA 4.3 Release Notes

Baird

Collins²,

Godfrey

et al. 2022

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“…The PWR benchmark problems were taken from Kim et al [7], which verified the ENDF/B-VII.1 MPACT 51-group library for the VERA MPACT. Only a brief overview of the problems is included herein; detailed descriptions can be found in the aforementioned paper [7]. These benchmark problems were used in the comparison of neutronic characteristics between ENDF/B-VII.1 and -VIII.0.…”

Section: Pwr Benchmark Problemsmentioning

confidence: 99%

“…The VERA depletion benchmark problems [7,18] were developed based on the VERA progression problems. The depletion benchmark suite includes 8 single-pin problems and 16 fuel-assembly problems with various fuel temperatures, 235 U enrichments, control rods and burnable poisons.…”

Section: Pwr Benchmark Problemsmentioning

confidence: 99%

“…VERA development continues through the Nuclear Energy Advanced Modeling and Simulation (NEAMS) project [3] for the boiling water reactor (BWR) analysis. The PWR simulation results using VERA with the Evaluated Nuclear Data File (ENDF)/B-VII.1 nuclear evaluated data [4] were very consistent with the PWR measured data [5], which were processed to generate the 51-group neutron cross-section library [6,7] for the primary VERA neutronics simulator MPACT [8]. The ENDF/B-VII.1 MPACT 51-group library supports both non-epithermal and epithermal upscattering for 238 U, which has been known to improve the Doppler temperature coefficient [9].…”

Section: Introductionmentioning

confidence: 99%

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Neutronic Characteristics of ENDF/B-VIII.0 Compared to ENDF/B-VII.1 for Light-Water Reactor Analysis

Kim

Wieselquist

2021

JNE

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The Evaluated Nuclear Data File (ENDF)/B-VIII.0 data library was released in 2018. To assess the new data during development and shortly after release, many validation calculations were performed with static, room-temperature benchmarks. Recently, when performing validation of ENDF/B-VIII.0 for pressurized water reactor depletion calculations, a regression in performance compared to ENDF/B-VII.1 was observed. This paper documents extensive benchmark calculations for light-water reactors performed using continuous energy Monte Carlo code with ENDF/B-VII.1 and -VIII.0 and neutronic characteristics of ENDF/B-VIII.0 are discussed and compared to those of ENDF/B-VII.1. It is our recommendation that ENDF/B data library assessment should include reactor-specific benchmark assessments, including depletion cases, such that these types of regressions may be caught earlier in the data development cycle.

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Development of the multigroup cross section library for the CASL neutronics simulator MPACT: Verification

Cited by 18 publications

References 4 publications

Watts Bar Unit 2 Startup Results with VERA

Watts Bar Unit 2 Startup Results with VERA

VERA 4.3 Release Notes

Neutronic Characteristics of ENDF/B-VIII.0 Compared to ENDF/B-VII.1 for Light-Water Reactor Analysis

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