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

2018 ◽  
Vol 4 ◽  
pp. 42 ◽  
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.

scholarly journals Nuclear data sensitivity analysis and uncertainty propagation in the KYADJ whole-core transport code

2020 ◽  
Vol 239 ◽  
pp. 22012
Author(s):  
Qu Wu ◽  
Xingjie Peng ◽  
Guanlin Shi ◽  
Yingrui Yu ◽  
Qing Li ◽  
...  
Keyword(s):  
Sensitivity Analysis ◽  
Uncertainty Analysis ◽  
Neutron Transport ◽  
Uncertainty Propagation ◽  
Reactor Core ◽  
Nuclear Data ◽  
Data File ◽  
Neutron Transport Equation ◽  
Transport Code ◽  
Data Sensitivity

Nuclear data sensitivity analysis and uncertainty propagation have been extensively applied to nuclear data adjustment and uncertainty quantification in the field of nuclear engineering. Sensitivity and Uncertainty (S&U) analysis is developed in the KYADJ whole-core transport code in order to meet the requirement of advanced reactor design. KYADJ aims to use two-dimension Method of Characteristic (MOC) and one-dimension discrete ordinate (SN) coupled method to solve the neutron transport equation and achieve one-step direct transport calculation of the reactor core. Developing sensitivity and uncertainty analysis module in KYADJ can minimize deviations caused by modeling approximation and enhance calculation efficiency. This work describes the application of the classic perturbation theory to the KYADJ transport solver. In order to obtain uncertainty, a technique is proposed for processing a covariance data file in 45-group energy grid instead of 44-group SCALE 6.1 covariance data which is extensively used in various codes. Numerical results for Uncertainty Analysis in Modelling (UAM) benchmarks and the SF96 benchmark are presented. The results agree well with the reference and the capability of S&U analysis in KYADJ is verified.

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

2021 ◽  
Vol 154 ◽  
pp. 108099
Author(s):  
Guanlin Shi ◽  
Yuchuan Guo ◽  
Conglong Jia ◽  
Zhiyuan Feng ◽  
Kan Wang ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Uncertainty Analysis ◽  
Monte Carlo Code ◽  
Sensitivity And Uncertainty Analysis

Eigenvalue sensitivity and uncertainty analysis based on a 2-D/1-D whole-core transport code KYADJ

2018 ◽  
Vol 122 ◽  
pp. 185-192 ◽  
Author(s):  
Qu Wu ◽  
Jiankai Yu ◽  
Guanlin Shi ◽  
Xiao Tang ◽  
Yingrui Yu ◽  
...  
Keyword(s):  
Uncertainty Analysis ◽  
Eigenvalue Sensitivity ◽  
Transport Code ◽  
Sensitivity And Uncertainty Analysis

Sensitivity and uncertainty analysis on keff due to nuclear data in the KRITZ-2:19 – Comparison between JENDL-4.0 and ENDF/B-VII.1

2019 ◽  
Vol 129 ◽  
pp. 308-315
Author(s):  
Abdulaziz Ahmed ◽  
H. Boukhal ◽  
T. El Bardouni ◽  
M. Makhloul ◽  
E. Chakir ◽  
...  
Keyword(s):  
Uncertainty Analysis ◽  
Nuclear Data ◽  
Sensitivity And Uncertainty Analysis

Enhancement of Stochastic Sampling Capability in RMC Code

2018 ◽  
Author(s):  
Guanlin Shi ◽  
Yishu Qiu ◽  
Kan Wang
Keyword(s):  
Uncertainty Analysis ◽  
Sampling Method ◽  
Nuclear Data ◽  
Data Uncertainty ◽  
Monte Carlo Code ◽  
Tabular Data ◽  
Stochastic Sampling ◽  
First Order ◽  
Fission Spectrum ◽  
Sensitivity And Uncertainty Analysis

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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