Improvement of sensitivity and uncertainty analysis capabilities of generalized response in Monte Carlo code RMC

As people pay more attention to nuclear safety analysis, sensitivity and uncertainty analysis has become a research hotspot. In our previous research, we had developed an integrated, built-in stochastic sampling module in the Reactor Monte Carlo code RMC [1]. Using this module, we can perform nuclear data uncertainty analysis. But at that time the uncertainty of fission spectrum was not considered. So, in this work, the capability of computing the uncertainty of keff induced by the uncertainty of fission spectrum, including tabular data form and formula form, is implemented in RMC code based on the stochastic sampling method. The algorithms and capability of computing keff uncertainty induced by uncertainty of fission spectrum in RMC are verified by comparison with the results calculated by the first order uncertainty quantification method [2].

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Monte Carlo based sensitivity and uncertainty analysis of the HCPB Test Blanket Module in ITER

Fusion Engineering and Design ◽

10.1016/j.fusengdes.2008.06.053 ◽

2008 ◽

Vol 83 (7-9) ◽

pp. 1222-1226 ◽

Cited By ~ 5

Author(s):

U. Fischer ◽

D. Leichtle ◽

R.L. Perel

Keyword(s):

Monte Carlo ◽

Uncertainty Analysis ◽

Sensitivity And Uncertainty Analysis

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UNCERTAINTY ANALYSIS OF SUB-EXERCISES IN UAM-SFR BENCHMARK WITH THE MCS CODE

EPJ Web of Conferences ◽

10.1051/epjconf/202124715017 ◽

2021 ◽

Vol 247 ◽

pp. 15017

Author(s):

Yunki Jo ◽

Vutheam Dos ◽

Nhan Nguyen Trong Mai ◽

Hyunsuk Lee ◽

Deokjung Lee

Keyword(s):

Monte Carlo ◽

Perturbation Theory ◽

Uncertainty Analysis ◽

Covariance Matrix ◽

Nuclear Data ◽

Monte Carlo Code ◽

Sensitivity Coefficients ◽

Classical Perturbation Theory ◽

Generalized Perturbation Theory ◽

Parameters Of Interest

Uncertainty analysis in Modelling (UAM) for Design, Operation and Safety Analysis of Sodium-cooled Fast Reactors (SFRs) has been formed by OECD/NEA to assess the effect of nuclear data uncertainties on parameters of interest in SFR analysis. In this paper, sub-exercises of a medium 1000 MWth metallic core (MET-1000) and a large 3600 MWth oxide core (MOX-3600) are tested by a Monte Carlo code MCS to perform uncertainty analysis. Classical perturbation theory and generalized perturbation theory are used to calculate sensitivity coefficients. Uncertainty is calculated by multiplying the sensitivity coefficients and relative covariance matrix from ENDF/B-VII.1 library.

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Sensitivity and uncertainty analysis of nuclear reactor reactivity coefficients by Monte Carlo second-order perturbation method

Annals of Nuclear Energy ◽

10.1016/j.anucene.2018.06.047 ◽

2018 ◽

Vol 121 ◽

pp. 68-76 ◽

Cited By ~ 1

Author(s):

Seung Yeol Yoo ◽

Hyung Jin Shim

Keyword(s):

Monte Carlo ◽

Uncertainty Analysis ◽

Perturbation Method ◽

Nuclear Reactor ◽

Second Order ◽

Order Perturbation ◽

Reactivity Coefficients ◽

Sensitivity And Uncertainty Analysis

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Sensitivity and uncertainty analysis of βeff for MYRRHA using a Monte Carlo technique

EPJ Nuclear Sciences & Technologies ◽

10.1051/epjn/2018023 ◽

2018 ◽

Vol 4 ◽

pp. 42 ◽

Cited By ~ 2

Author(s):

Hiroki Iwamoto ◽

Alexey Stakovskiy ◽

Luca Fiorito ◽

Gert Van den Eynde

Keyword(s):

Monte Carlo ◽

Uncertainty Analysis ◽

Nuclear Data ◽

Ratio Method ◽

Delayed Neutron ◽

Continuous Energy ◽

A Value ◽

Transport Code ◽

Sensitivity And Uncertainty Analysis ◽

Data Sensitivity

This paper presents a nuclear data sensitivity and uncertainty analysis of the effective delayed neutron fraction βeff for critical and subcritical cores of the MYRRHA reactor using the continuous-energy Monte Carlo N-Particle transport code MCNP. The βeff sensitivities are calculated by the modified k-ratio method proposed by Chiba. Comparing the βeff sensitivities obtained with different scaling factors a introduced by Chiba shows that a value of a = 20 is the most suitable for the uncertainty quantification of βeff. Using the calculated βeff sensitivities and the JENDL-4.0u covariance data, the βeff uncertainties for the critical and subcritical cores are determined to be 2.2 ± 0.2% and 2.0 ± 0.2%, respectively, which are dominated by delayed neutron yield of 239Pu and 238U.

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