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Uncertainty and sensitivity analysis in reactivity-initiated accident fuel modeling: synthesis of organisation for economic co-operation and development (OECD)/nuclear energy agency (NEA) benchmark on reactivity-initiated accident codes phase-II


O. March anda, J. Zhangb, M. Cherubinic

aInstitut de Radioprotection et de Sûrete Nucleaire (IRSN), PSN-RES, SEMIA, Cadarache, St Paul-Lez-Durance, 13115, France
bTractebel (ENGIE), Avenue Ariane 7, 1200 Brussels, Belgium
Nuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy


Nuclear Engineering and Technology, December 2017


Abstract —  In the framework of OECD/NEA Working Group on Fuel Safety, a RIA fuel-rod-code Benchmark Phase I was organized in 2010e2013. It consisted of four experiments on highly irradiated fuel rodlets tested under different experimental conditions. This benchmark revealed the need to better understand the basic models incorporated in each code for realistic simulation of the complicated integral RIA tests with high burnup fuel rods. A second phase of the benchmark (Phase II) was thus launched early in 2014, which has been organized in two complementary activities: (1) comparison of the results of different simulations on simplified cases in order to provide additional bases for understanding the differences in modelling of the concerned phenomena; (2) assessment of the uncertainty of the results. The present paper provides a summary and conclusions of the second activity of the Benchmark Phase II, which is based on the input uncertainty propagation methodology. The main conclusion is that uncertainties cannot fully explain the difference between the code predictions. Finally, based on the RIA benchmark Phase-I and Phase-II conclusions, some recommendations are made.
© 2018 Korean Nuclear Society, Published by Elsevier Korea LLC.
This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/licenses/by-nc-nd/4.0/).


IAEA CRP Project on Benchmark Analysis of EBR-II Shutdown Heat Removal Tests

This publication presents the results and main achievements of an IAEA coordinated research project to verify and validate system and safety codes used in the analyses of liquid metal thermal hydraulics and neutronics phenomena in sodium cooled fast reactors. The publication will be of use to the researchers and professionals currently working on relevant fast reactors programmes. In addition, it is intended to support the training of the next generation of analysts and designers through international benchmark exercises.





Analysis of a small PWR core with the PARCS/Helios and PARCS/Serpent code systems




G. Baiocco,a A. Petruzzi,a S. Bznunib, T. Kozlowskic

aNuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy

bNuclear and Radiation Safety Center (NRSC), Yerevan, Armenia

cDepartment of Nuclear, Plasma and Radiological Engineering, University of Illinois at Urbana-Champaign, IL, USA

Annals of Nuclear Energy Volume 107, September 2017, Pages 42–48

Abstract — 
Lattice physics codes are primarily used to generate cross-section data for nodal codes. In this work the methodology of homogenized constant generation was applied to a small Pressurized Water Reactor (PWR) core, using the deterministic code Helios and the Monte Carlo code Serpent. Subsequently, a 3D analysis of the PWR core was performed with the nodal diffusion code PARCS using the two-group cross section data sets generated by Helios and Serpent. Moreover, a full 3D model of the PWR core was developed using Serpent in order to obtain a reference solution. Several parameters, such as keff, axial and radial power, fission and capture rates were compared and found to be in good agreement.