New features and improved uncertainty analysis in the NEA nuclear data sensitivity tool (NDaST)

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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Nuclear data sensitivity/uncertainty analysis for XT-ADS

Annals of Nuclear Energy ◽

10.1016/j.anucene.2010.12.018 ◽

2011 ◽

Vol 38 (5) ◽

pp. 1098-1108 ◽

Cited By ~ 7

Author(s):

Takanori Sugawara ◽

Massimo Sarotto ◽

Alexey Stankovskiy ◽

Gert Van den Eynde

Keyword(s):

Uncertainty Analysis ◽

Nuclear Data ◽

Data Sensitivity

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

EPJ Web of Conferences ◽

10.1051/epjconf/202023922012 ◽

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.

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