A calculation method for parameters of secondary neutron source in nuclear reactor

Junyun YANG; Lan LI; Huanwen LYU; Yi TAN; Feng XIAO; Futing JING; Songqian TANG; Bin LIU

doi:10.11889/j.0253-3219.2018.hjs.41.060605

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A calculation method for parameters of secondary neutron source in nuclear reactor

NUCLEAR ENERGY SCIENCE AND ENGINEERING | 更新时间：2021-05-19

- A calculation method for parameters of secondary neutron source in nuclear reactor
- NUCLEAR TECHNIQUES Vol. 41, Issue 6, Article number: 060605(2018)
- 作者机构：
  
  中国核动力研究设计院核反应堆系统设计技术重点实验室成都 610213
- 作者简介：
- 基金信息：
- DOI：10.11889/j.0253-3219.2018.hjs.41.060605
  CLC： TL4
- Received：13 November 2017，
  
  Revised：06 December 2017，
  
  Published：10 June 2018
- 稿件说明：
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Junyun YANG, Lan LI, Huanwen LYU, et al. A calculation method for parameters of secondary neutron source in nuclear reactor[J]. Nuclear techniques, 2018, 41(6): 060605
DOI：

Junyun YANG, Lan LI, Huanwen LYU, et al. A calculation method for parameters of secondary neutron source in nuclear reactor[J]. Nuclear techniques, 2018, 41(6): 060605 DOI： 10.11889/j.0253-3219.2018.hjs.41.060605.

摘要

基于详细的燃耗计算和粒子输运计算，获得二次中子源的源强、能谱和空间分布等关键参数的数值模拟方法。以某商用反应堆的换料实践为例，首先，通过对二次中子源的辐照和衰变过程的模拟，计算出换料时二次中子源棒内

124

Sb的含量；然后，使用MCNP（Monte Carlo N Particle Transport Code）程序及相应的连续能量核反应截面数据，对核反应堆换料结束时的二次中子源实施中子-光子耦合输运模拟，获得相应的二次中子源参数；最后，根据所得中子源参数对某反应堆的堆外探测器响应进行计算，得到了与实测值吻合较好的理论值，验证了该方法的可行性。

Abstract

Background

In the design of nuclear reactor

it is significant to determine the parameters of secondary neutron source.

Purpose

The aim of this paper is to elaborate a calculation method for parameters of secondary neutron source systematically.

Methods

Firstly

compositions of the fuel and secondary neutron source were generated by depletion calculation. Then

gamma production and photonuclear reaction in secondary neutron source were simulated. Specific parameters of secondary neutron source

such as source strength

energy spectrum and space distribution

have been obtained.

Results

The parameters of secondary neutron source have been used to calculate the response of detectors outside the core. The results are in agreement with the measurements.

Conclusion

The calculation method suggested in this paper is available.

关键词

Keywords

references

景福庭 , 肖锋 , 刘嘉嘉 , . 压水堆二次中子源源强计算研究 . 核动力工程 , 2014 . ( S2 ): 64 - 66 . http://mall.cnki.net/magazine/Article/HDLG2014S2018.htm .

Futing JING , Feng XIAO , Jiajia LIU , . Calculation of neutron source strength of PWR two times . Nuclear Power Engineering , 2014 . ( S2 ): 64 - 66 . http://mall.cnki.net/magazine/Article/HDLG2014S2018.htm .

田东风 , 孙伟力 . 中子核反应激发函数 , : 北京国防工业出版社 , 2006 .

Dongfeng TIAN , Weili SUN . Neutron nuclear reaction excitation function , : Beijing National Defense Industry Press , 2006 .

Hermann O W, Westfall R M. ORIGEN-S: SCALE system module to calculate fuel depletion, actinide transmutation, fission product buildup and decay, and associated radiation source terms[R]. Sect. F7 of SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation, NUREG/CR-0200, Rev, 1995, Vol. Ⅱ: 6.

Briesmeister J F. MCNP-a general Monte Carlo N-particle transport code (LA-12625)[R]. USA: Los Alamos National Laboratory, 1993.

徐治龙 , 刘勇进 , 万海霞 , . 反应堆中子源强计算程序开发 . 原子能科学技术 , 2013 . 47 ( B12 ): 438 - 442 . http://mall.cnki.net/magazine/article/YZJS2013S2013.htm .

Zhilong XU , Yongjin LIU , Haixia WAN , . Reactor neutron source strength calculation program development . Atomic Energy Science and Technology , 2013 . 47 ( B12 ): 438 - 442 . http://mall.cnki.net/magazine/article/YZJS2013S2013.htm .

苏耿华 , 包鹏飞 , 韩嵩 . 反应堆二次中子源源强计算及验证 . 强激光与粒子束 , 2017 . 29 ( 3 ): 123 - 126 . http://mall.cnki.net/magazine/Article/HDLG2014S2018.htm .

Genghua SU , Pengfei BAO , Song HAN . Calculation and verification of the two neutron source strength of the reactor . Strong Laser and Particle Beam , 2017 . 29 ( 3 ): 123 - 126 . http://mall.cnki.net/magazine/Article/HDLG2014S2018.htm .

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