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The monte carlo code MCNP6
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This paper presents the steady-state analysis results of the OECD/NEA and U.S. NRC PWR MOX/UO2 (MOX: Mixed Oxide) Core Transient Benchmark with the modern MCNP6 Monte Carlo code based on the up-to-date ENDF/B-VII.1 nuclear data library.
9p
vicedric
08-02-2023
8
3
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This paper presents the analysis of sensitivity and uncertainty for the infinite multiplication factor (kinf) for the VVR-M2 typed HEU and LEU fuel assemblies of the Dalat nuclear research reactor (DNRR) using the MCNP6.1-Whisper1.1 code.
14p
vimelindagates
18-07-2022
7
2
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This paper presents the steady-state analysis results of the OECD/NEA and U.S. NRC PWR MOX/UO2 (MOX: Mixed Oxide) Core Transient Benchmark with the modern MCNP6 Monte Carlo code based on the ENDF/B-VII.1 evaluated nuclear data library.
10p
meyerowitz
25-12-2021
8
0
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The purpose of this study is to verify the accuracy of innovative ADS core modeling by using simulation codes. The reactivity calculations of CERMET loaded fuel ADS was conducted using two Monte Carlo codes, Serpent [8] and MCNP6 [9] with ENDF/B-VII.0 library [10]. The comparison of results obtained from the two codes is analyzed and discussed in this study.
6p
nguaconbaynhay11
07-04-2021
12
1
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The present work aims to perform burnup calculation of the OECD VVER-1000 LEU (low enriched uranium) computational benchmark assembly using the Monte Carlo code MCNP6 and the deterministic code SRAC2006.
10p
trinhthamhodang1218
14-03-2021
4
1
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