Nuclear Techniques ›› 2020, Vol. 43 ›› Issue (4): 40004-040004.

• SPECIAL SECTION ON THE 11TH NATIONAL CONFERENCE ON NEW AND RESEARCH REACTORS (PART I) •

### Impact analysis of basic burnup cross section on coupled calculation of neutron transport burnup

Wankui YANG,Baoxin YUAN,Huan HUANG,Guanbo WANG,Songbao ZHANG

1. Institute of Nuclear Physics & Chemistry, China Academy of Engineering Physics, Mianyang 621900, China
• Received:2020-01-16 Revised:2020-02-22 Online:2020-04-15 Published:2020-04-20
• About author:YANG Wankui, male, born in 1988, graduated from China Academy of Engineering Physics with a master's degree in 2013, focusing on reactor physics and high performance computing

Abstract: Background

Monte Carlo depletion method is based on Monte Carlo neutron transport code and ORIGEN2 point depletion code, it is widely used for neutron transport burnup. However, the continuous energy neutron cross sections of the nuclides in the current evaluation library are less than the cross section used in the depletion calculation. Therefore, the depletion cross sections generated by Monte Carlo transport calculation cannot replace all the cross sections in the basic depletion library.

Purpose

This study aims to analyse the influence of different basic depletion libraries on coupled calculation of neutron transport burnup.

Methods

The Monte Carlo transport depletion code MCBMPI (Monte Carlo Burnup code in MPI version) was employed to calculate the new VERA (Virtual Environment for Reactor Applications) depletion benchmark. The influence of different basic burnup cross section on the burnup calculation in transportation was compared and analyzed.

Results

Calculation results show that relative error of keff is less than 8‰ in all the depletion stages, and the mass variation of 235U and 135Xe are less than 4‰ and 5‰ respectively, in the last depletion stage. The variation between thermal and fast neutron spectrum cross section library is small in this benchmark.

Conclusions

It is recommended to select a basic depletion cross section library containing similar neutron spectrum instead of a typical thermal neutron cross section library in real application.

CLC Number:

• TL99