Uncertainty analysis of reactor pressure vessel fast neutron fluence calculation

Bin ZHANG; Junxiao ZHENG; Xiaojing LI; Boyu BAO; Yixue CHEN

doi:10.11889/j.0253-3219.2018.hjs.41.020605

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Uncertainty analysis of reactor pressure vessel fast neutron fluence calculation

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

- Uncertainty analysis of reactor pressure vessel fast neutron fluence calculation
- NUCLEAR TECHNIQUES Vol. 41, Issue 2, Article number: 020605(2018)
- 作者机构：
  
  1.华北电力大学北京 102206
  2.中广核研究院有限公司深圳 518026
  3.中国核电工程有限公司北京 100840
- 作者简介：
- 基金信息：
  
  National Natural Science Foundation of China;National Natural Science Foundation of China
- DOI：10.11889/j.0253-3219.2018.hjs.41.020605
  CLC： TL328
- Received：18 September 2017，
  
  Revised：06 November 2017，
  
  Published：10 February 2018
- 稿件说明：
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Bin ZHANG, Junxiao ZHENG, Xiaojing LI, et al. Uncertainty analysis of reactor pressure vessel fast neutron fluence calculation[J]. Nuclear techniques, 2018, 41(2): 020605
DOI：

Bin ZHANG, Junxiao ZHENG, Xiaojing LI, et al. Uncertainty analysis of reactor pressure vessel fast neutron fluence calculation[J]. Nuclear techniques, 2018, 41(2): 020605 DOI： 10.11889/j.0253-3219.2018.hjs.41.020605.

摘要

反应堆压力容器（Reactor Pressure Vessel，RPV）快中子注量的精确计算，可以有效地评估寿期内压力容器材料辐照损伤程度并为延寿技术提供数据支持。离散纵标法在解决深穿透屏蔽问题上的优势，使其成为计算RPV快中子注量率的主要方法之一。反应堆几何材料、截面数据、中子源强、输运求解等参数的不确定性，显著影响快中子注量率模拟结果的可信度。以典型压水堆模型为测试例题，分析了RPV快中子注量率计算流程中引入不确定性的主要来源，重点研究了中子源强处理方法、各向异性散射截面展开和输运方程空间离散格式三个方面对RPV快中子注量率计算的影响。对不确定性因素特别是计算不确定性进行深入的细化分析，有助于我们正确认识目前计算方法的薄弱环节，指明相关的计算模型和数值离散的改进方向，并提高计算结果的确定性。

Abstract

Background

The precise calculation of reactor pressure vessel (RPV) fast neutron fluence can be used to evaluate the materials irradiation damage during the designed lifetime and support analyses for a potential plant life extension. The uncertainties of reactor materials and geometry

nuclear data

neutron source and transport solving process can affect the reliability of fast neutron fluence simulation results.

Purpose

This paper aims to effectively evaluate the calculation results by thorough analysis of the uncertainties described above.

Methods

This paper analyzed the main origin of uncertainties in the RPV fast neutron fluence calculation and focus on the effect of the neutron source processing method

the Legendre expansion of the anisotropic scattering cross section and spatial discretization method of transport equation.

Results and Conclusion

Modeling of fuel assemblies usually produces a deviation about 20%

the effect of fission nuclide and transport correction methods are about 10%. Through analyses we can understand the weak aspects of the calculation methods and improve the accuracy of calculation results.

关键词

Keywords

references

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