Inter-comparison of JEF-2.2 and JEFF-3.1 evaluated nuclear data through Monte Carlo analysis of VVER-1000 MOX Core Computational Benchmark

Thilagam, L.; Karthikeyan, R.; Jagannathan, V.; Subbaiah, K.V.; Lee, S.M.

doi:10.1016/j.anucene.2009.11.009

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Cited by 3 publications

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“…Two of the solutions are based on continuous energy Monte Carlo methods (MCNP4C (RSICC, 2000) and MCU (Kalugin et al, 2015)), while the other solution is based on collision probability (or similar) method (RADAR). Thilagam et al (2009) performed same calculations with different codes and Thilagam et al (2010) used benchmark results to perform inter comparison of evaluated data.…”

Section: Introductionmentioning

confidence: 99%

Application of CNUREAS and MCNP5 codes to VVER-1000 MOX Core Computational Benchmark

Lüle

Özdemır

Erdoğan

2015

Progress in Nuclear Energy

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Section: Introductionmentioning

confidence: 99%

Application of CNUREAS and MCNP5 codes to VVER-1000 MOX Core Computational Benchmark

Lüle

Özdemır

Erdoğan

2015

Progress in Nuclear Energy

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Comparative studies on thorium fuel cycles of BWR with JENDL 3.2 and JEF 2.2 nuclear data libraries

Waris

Pramuditya

Perkasa

et al. 2014

IJNEST

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Neutronic calculations for the VVER-1000 MOX core computational benchmark using the OpenMC code

Hossain,

Mollah,

Akter

et al. 2023

NUCET

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The goal of this study is to perform neutronic calculations of the VVER-1000 MOX core computational benchmarks with an OpenMC code along with ENDF/B-VII.1 nuclear data library. The results of neutronic analysis using the OpenMC Monte Carlo code for the VVER-1000 MOX core, containing 30% mixed oxide fuel with low enriched uranium fuel, are presented in this study. As per the benchmark report, all six states are considered in the present study. The keff values, assembly average fission reaction rates, and pin-by-pin fission rates were calculated as per benchmark criteria. In addition, 2D thermal and fast neutron-flux distribution were also generated. The reactivity results and neutron flux distribution were compared with other results in which benchmark analysis was performed using the same core geometry and it showed great similarity with slight deviation. This shows that the modeling of the VVER-1000 MOX core was done successfully using OpenMC. Because OpenMC was successfully used for neutronics calculation of the VVER-1000 whole core, it may be mentioned here that OpenMC code can also be utilized for neutronics and other reactor core physics analyses of the VVER-1200 reactor which is to be commissioned in Bangladesh in the upcoming year.

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Inter-comparison of JEF-2.2 and JEFF-3.1 evaluated nuclear data through Monte Carlo analysis of VVER-1000 MOX Core Computational Benchmark

Cited by 3 publications

References 5 publications

Application of CNUREAS and MCNP5 codes to VVER-1000 MOX Core Computational Benchmark

Application of CNUREAS and MCNP5 codes to VVER-1000 MOX Core Computational Benchmark

Comparative studies on thorium fuel cycles of BWR with JENDL 3.2 and JEF 2.2 nuclear data libraries

Neutronic calculations for the VVER-1000 MOX core computational benchmark using the OpenMC code

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Inter-comparison of JEF-2.2 and JEFF-3.1 evaluated nuclear data through Monte Carlo analysis of VVER-1000 MOX Core Computational Benchmark

Cited by 3 publications

References 5 publications

Application of CNUREAS and MCNP5 codes to VVER-1000 MOX Core Computational Benchmark

Application of CNUREAS and MCNP5 codes to VVER-1000 MOX Core Computational Benchmark

Comparative studies on thorium fuel cycles of BWR with JENDL 3.2 and JEF 2.2 nuclear data libraries

﻿Neutronic calculations for the VVER-1000 MOX core computational benchmark using the OpenMC code

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Neutronic calculations for the VVER-1000 MOX core computational benchmark using the OpenMC code