Influence of multigroup nuclear data uncertainties on the reactor core physics calculation

Xinyi PAN; Bing LAN; Chunming ZHANG; Jianping JING; Guoshun YOU

doi:10.11889/j.0253-3219.2016.hjs.39.010602

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Influence of multigroup nuclear data uncertainties on the reactor core physics calculation

更新时间：2021-05-24

- Influence of multigroup nuclear data uncertainties on the reactor core physics calculation
- NUCLEAR TECHNIQUES Vol. 39, Issue 1, Article number: 010602(2016)
- 作者机构：
  
  环境保护部核与辐射安全中心北京 100082
  Nuclear and Radiation Safety Center, Ministry of Environmental Protection, Beijing 100082, China
- 作者简介：
- 基金信息：
- DOI：10.11889/j.0253-3219.2016.hjs.39.010602
  CLC： TL329
- Received：31 August 2015，
  
  Accepted：2015-10-23，
  
  Published：25 January 2016
- 稿件说明：
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Xinyi PAN, Bing LAN, Chunming ZHANG, et al. Influence of multigroup nuclear data uncertainties on the reactor core physics calculation[J]. Nuclear techniques, 2016, 39(1): 010602
DOI：

Xinyi PAN, Bing LAN, Chunming ZHANG, et al. Influence of multigroup nuclear data uncertainties on the reactor core physics calculation[J]. Nuclear techniques, 2016, 39(1): 010602 DOI： 10.11889/j.0253-3219.2016.hjs.39.010602.

摘要

核数据不确定性是造成反应堆物理计算结果不确定性的重要因素之一。基于所需抽样核数据的协方差矩阵开发了随机抽样模块(Stochastic Sampling

SAMP)

在此基础上利用SCALE (Standardized Computer Analyses for Licensing Evaluation)软件包实现了混合法和随机抽样法两种不确定性分析方法

以研究多群核数据不确定性对堆芯物理计算的影响。以3×3假想堆芯为对象

对两种方法进行了验证

然后应用于国际原子能机构(International Atomic Energy Agency

IAEA)燃料管理基准题中的Almaraz核电厂首循环堆芯。分析结果表明

两种方法结果符合良好

Almaraz核电厂堆芯

eff

不确定性约为0.5%

堆芯径向和轴向功率的最大不确定性分别为1.9%和0.45%。

Abstract

Background:

The uncertainty of nuclear data is one of the key factors resulting in the uncertainty of reactor physics calculation.

Purpose:

The influence of multigroup nuclear data uncertainties

on the reactor core physics calculation was studied in this paper.

Methods:

The stochastic sampling modular SAMP based on covariance matrix of nuclear data was developed

and the hybrid method and stochastic sampling method were realized using SCALE (Standardized Computer Analyses for Licensing Evaluation) software package. The two methods were validated using 3×3 hypothetical core and then applied to the first cycle of Almaraz pressurized-water reactor (PWR) in the IAEA (International Atomic Energy Agency) fuel management benchmark.

Results:

Results of the two methods are in good agreement. The uncertainty of core effective multiplication factor is about 0.5%

and the maximum uncertainties of the radial and axial power are about 1.9% and 0.45% respectively in Almaraz PWR.

Conclusion:

The two-step method and stochastic sampling method can both be used for the uncertainty analysis of reactor core calculation.

关键词

Keywords

references

司胜义 . 堆芯核设计程序系统的不确定性综述 . 第十一届反应堆数值计算和粒子输运学术会议暨2006年反应堆物理会议, 哈尔滨, 中国 , 2006 .

SI Shengyi . Review on the uncertainty of software system for reactor core nuclear design . 11 st Academic Conference on the Reactor Numerical Calculation and Particle Transport and Reactor Physics Conference 2006, Harbin, China , 2006 .

Ivanov K , Avramova M , Kamerow S . Overview and discussion of phase I of the OECD LWR UAM benchmark activity . International Conference on Mathematics, Computational Methods & Reactor Physics (M&C 2009), New York, USA , 2009 .

Zhou X M , Liu G M , Li D , . Using activation method to measure neutron spectrum in an irradiation chamber of a research reactor . Nuclear Science and Techniques , 2014 . 25 ( 1 ): 010603 .

Williams M L , Wiarda D , Arbanas G , . SCALE nuclear data covariance library . ORNL/TM-2005/39, Oak Ridge National Laboratory , 2011 .

Rearden B T , Jesssee M A , Williams M L . TSUNAMI-1D: control module for one dimensional cross-section sensitivity and uncertainty analysis for criticality . ORNL/TM-2005/39, Oak Ridge National Laboratory , 2005 .

Kodeli I . Multidimensional deterministic nuclear data sensitivity and uncertainty code system: method and application . Nuclear Science and Engineering , 2001 . 138 ( 1 ): 45 - 66 .

Zwermann W , Krzykacz-Hausmann B , Gallner L , . Uncertainty analysis with nuclear covariance data in reactor core calculations . Journal of the Korean Physical Society , 2011 . 59 ( 2 ): 1256 - 1259 .

Ahnert C , Aragones J M , Merino F , . In-core fuel management code package validation for PWRs . IAEA-TECDOC-815, International Atomic Energy Agency , 1995 .

Bevington P R , Robinson K . Data reduction and error analysis for the physical sciences . Boston: McGraw-Hill , 2003 .

Wilks S . Determination of sample sizes for setting tolerance limits . The Annals of Mathematical Statistics , 1941 . 12 ( 1 ): 91 - 96 .

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