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

Journal Article · · Journal of Nuclear Engineering
DOI:https://doi.org/10.3390/jne2040026· OSTI ID:1822049
 [1];  [1]
  1. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
+ Show Author Affiliations
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.
Research Organization:
Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
Sponsoring Organization:
USDOE Office of Science (SC)
Grant/Contract Number:
AC05-00OR22725
OSTI ID:
1822049
Journal Information:
Journal of Nuclear Engineering, Journal Name: Journal of Nuclear Engineering Journal Issue: 4 Vol. 2; ISSN 2673-4362
Publisher:
MDPICopyright Statement
Country of Publication:
United States
Language:
English

References (11)

The impact of 238U resonance elastic scattering approximations on thermal reactor Doppler reactivity journal April 2009
The Serpent Monte Carlo code: Status, development and applications in 2013 journal August 2015
Development of the multigroup cross section library for the CASL neutronics simulator MPACT: Verification journal October 2019
Comparison of ENDF/B-VIII.0 and ENDF/B-VII.1 in criticality, depletion benchmark, and uncertainty analyses by McCARD journal September 2019
Development of the multigroup cross section library for the CASL neutronics simulator MPACT: Method and procedure journal November 2019
The Virtual Environment for Reactor Applications (VERA): Design and architecture journal December 2016
ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data journal December 2011
ENDF/B-VIII.0: The 8 th Major Release of the Nuclear Reaction Data Library with CIELO-project Cross Sections, New Standards and Thermal Scattering Data journal February 2018
VERA Core Simulator Methodology for Pressurized Water Reactor Cycle Depletion journal January 2017
Initial MCNP6 Release Overview journal December 2012
Impact of the ENDF/B-VIII.0 Library on Advanced Reactor Simulations
  • Bostelmann, F.; Holcomb, A.; Marshall, W.
  • Tranactions - 2019 Winter Meeting, Transactions of the American Nuclear Society - Volume 121 https://doi.org/10.13182/T31078
conference January 2019

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Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
ENDF/B-VII.1 and VIII.0
benchmark calculation
light water reactor
neutronic characteristics