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Title: Nuclear data uncertainty and sensitivity analysis of the VHTRC benchmark using SCALE

Journal Article · · Annals of Nuclear Energy (Oxford)
ORCiD logo [1]; ORCiD logo [2]
  1. Gesellschaft fur Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)
  2. Idaho National Lab. (INL), Idaho Falls, ID (United States)
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The Very High Temperature Critical Assembly (VHTRC) experiment represents one of the few data sets available for the validation of HTGR lattice physics. Within the framework of the IAEA Coordinated Research Project on HTGR Uncertainty Analysis in Modeling, uncertainty and sensitivity analyses of this graphite-moderated facility are performed as the validation reference to the prismatic MHTGR-350 fuel block calculations. Nominal multi-group and continuous-energy criticality calculations with the KENO-VI Monte Carlo code of the SCALE code package are compared with the continuous-energy Monte Carlo Code Serpent. Good agreement with a maximum difference of 250 pcm is obtained. The experimental data set, however, differs by several hundred pcm with the results of both codes when using the ENDF-VII.0 library. When using ENDF/B-VII.1 data, this difference is reduced to a few hundred pcm and the calculations lie within the experimental error bars. Uncertainties of the VHTRC multiplication factors due to uncertainties in nuclear data are determined with SAMPLER/KENO-VI of SCALE 6.2 and with continuous-energy TSUNAMI of SCALE 6.2. For all experimental configurations, the obtained uncertainty is found to be 0.58% when using ENDF/B-VII.0 data. The top contributor to this uncertainty is the average number of neutrons per fission event of U-235. With ENDF/B-VII.1 data, uncertainties of about 0.66% are obtained in particular due to an increased uncertainty of U-235 nu-bar in the latest ENDF release. In conclusion, when considering these nuclear data uncertainties, the obtained 1s uncertainty intervals are overlapping with the experimental error bars.

Research Organization:
Idaho National Laboratory (INL), Idaho Falls, ID (United States)
Sponsoring Organization:
USDOE Office of Nuclear Energy (NE)
Grant/Contract Number:
AC07-05ID14517
OSTI ID:
1474117
Alternate ID(s):
OSTI ID: 1495526
Report Number(s):
INL/JOU-17-42606-Rev000
Journal Information:
Annals of Nuclear Energy (Oxford), Vol. 110, Issue C; ISSN 0306-4549
Publisher:
ElsevierCopyright Statement
Country of Publication:
United States
Language:
English
Citation Metrics:
Cited by: 15 works
Citation information provided by
Web of Science

References (7)

A collision history-based approach to sensitivity/perturbation calculations in the continuous energy Monte Carlo code SERPENT journal November 2015
Criticality calculations of the Very High Temperature Reactor Critical Assembly benchmark with Serpent and SCALE/KENO-VI journal April 2016
Review of the natC(n,γ) cross section and criticality calculations of the graphite moderated reactor BR1 journal October 2013
Impact of Revised Thermal Neutron Capture Cross Section of Carbon Stored in JENDL-4.0 on HTTR Criticality Calculation journal July 2011
The Serpent Monte Carlo code: Status, development and applications in 2013 journal August 2015
SCALE Continuous-Energy Eigenvalue Sensitivity Coefficient Calculations journal March 2016
A Statistical Sampling Method for Uncertainty Analysis with SCALE and XSUSA journal September 2013

Cited By (2)

Thermal-Hydraulic and Neutronic Phenomena Important in Modeling and Simulation of Liquid-Fuel Molten Salt Reactors journal April 2019
Further development of methodology to model TRISO fuel and BISO absorber particles and related uncertainty quantification using SCALE 6 journal May 2019

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11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
HTGR
VHTRC
SCALE