基于SCALE/TRITON的单流双区熔盐堆燃耗计算方法

崔德阳; 夏少鹏; 余呈刚; 蔡翔舟; 陈金根

doi:10.11889/j.0253-3219.2017.hjs.40.080602

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基于SCALE/TRITON的单流双区熔盐堆燃耗计算方法

核能科学与工程 | 更新时间：2021-05-19

- 基于SCALE/TRITON的单流双区熔盐堆燃耗计算方法
- Methodologies for single-fluid, two-zone MSR burnup calculation based on SCALE/TRITON
- 核技术 2017年40卷第8期文章编号：080602
- 作者机构：
  
  1.中国科学院上海应用物理研究所嘉定园区上海 201800
  2.中国科学院先进核能创新研究院上海 201800
  3.中国科学院大学北京 100049
- 作者简介：
  
  [ "崔德阳, 男, 1987年出生, 2012年毕业于河南师范大学, 现为博士研究生, 研究方向为熔盐堆燃料循环", "CUI Deyang, male, born in 1987, graduated from Henan Normal University in 2012, doctoral student, focusing on fuel cycle in molten salt reactor" ]
  蔡翔舟, E-mail:caixz@sinap.ac.cnCAI Xiangzhou, E-mail:caixz@sinap.ac.cn
- 基金信息：
  
  中国科学院战略性先导科技专项(XDA02010000);若干关键问题研究项目(QYZDY-SSW-JSC016);国家自然科学基金(91326201)
- DOI：10.11889/j.0253-3219.2017.hjs.40.080602
  中图分类号： TL99
- 收稿日期：2017-04-19，
  
  修回日期：2017-05-04，
  
  纸质出版日期：2017-08-10
- 稿件说明：
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崔德阳, 夏少鹏, 余呈刚, 等. 基于SCALE/TRITON的单流双区熔盐堆燃耗计算方法[J]. 核技术, 2017,40(8):080602

Deyang CUI, Shaopeng XIA, Chenggang YU, et al. Methodologies for single-fluid, two-zone MSR burnup calculation based on SCALE/TRITON[J]. Nuclear techniques, 2017, 40(8): 080602
崔德阳, 夏少鹏, 余呈刚, 等. 基于SCALE/TRITON的单流双区熔盐堆燃耗计算方法[J]. 核技术, 2017,40(8):080602 DOI： 10.11889/j.0253-3219.2017.hjs.40.080602.

Deyang CUI, Shaopeng XIA, Chenggang YU, et al. Methodologies for single-fluid, two-zone MSR burnup calculation based on SCALE/TRITON[J]. Nuclear techniques, 2017, 40(8): 080602 DOI： 10.11889/j.0253-3219.2017.hjs.40.080602.

摘要

美国橡树岭国家实验室开发的SCALE/TRITON程序广泛用于反应堆临界安全、中子物理、辐射屏蔽和灵敏度与不确定度等方面的计算分析。基于SCALE/TRITON程序，采用等效体积、均匀混合和平均截面等三种外部耦合方法，处理单流双区熔盐堆的燃耗计算，解决了SCALE/TRITON程序在计算中不能精确反映流动燃料周期性均匀混合的问题。研究表明平均截面法与均匀混合法的计算结果几乎完全一致，与橡树岭文献结果也能很好符合，等效体积法因为没有考虑堆芯分区结构的差异而导致计算结果与其他两种方法偏离较大。基于SCALE/TRITON发展的平均截面法，放宽了对步长的要求，具有准确性好、计算效率高的优点，适用于熔盐堆两区（或多区）的堆芯设计与燃耗性能分析，具有重要的应用意义。

Abstract

Background

The standardized computer analysis for licensing evaluation (SCALE) developed in the Oak Ridge National Laboratory (ORNL) of USA has been widely used in criticality safety

neutron physics

radiation shielding

and sensitivity and uncertainty analysis. However

the burnup calculation for single-fluid

two-zone molten salt reactor (MSR) has not been well dealt with in SCALE/TRITON due to the cell information card (Celldata) which is used in unit cell calculations to generate problem-dependent multigroup cross sections.

Purpose

This study aims to develop and evaluate possible solutions to the problem above.

Methods

Based on external program

three methods (i.e.

homogeneous mixing method

equivalent volume method and average cross section method)

are developed without any modification of the existing codes in SCALE6 and they are tested in a MSR with two-zone core. Test results are compared and analyzed.

Results

Comparison of the three methods indicates that the results obtained by average cross section method are almost equal to those obtained by homogeneous mixing method and moreover they accord well with the results given in ORNL's work

whilst the equivalent volume method is not sufficient to describe the difference of unit cells in the core.

Conclusion

The average cross section method with a relatively high computational efficiency and accuracy is recommended for burnup calculation in the MSR with two or more zones when using SCALE/TRITON.

关键词

Keywords

references

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