TACR热传输系统热工水力瞬态分析

魏诗颖; 王玮; 王成龙; 田文喜; 秋穗正; 苏光辉

doi:10.11889/j.0253-3219.2017.hjs.40.080601

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TACR热传输系统热工水力瞬态分析

核能科学与工程 | 更新时间：2021-05-19

- TACR热传输系统热工水力瞬态分析
- Thermal-hydraulic transient analysis for heat transport system of thorium-based advanced CANDU reactor
- 核技术 2017年40卷第8期文章编号：080601
- 作者机构：
  
  1.西安交通大学能源与动力工程学院西安 710049
  2.中国核动力研究设计院核反应堆系统设计重点实验室成都 610041
- 作者简介：
  
  [ "魏诗颖, 女, 1992年出生, 2014年毕业于西安交通大学, 现为博士研究生, 研究领域为核反应堆热工水力分析", "WEI Shiying, female, born in 1992, graduated in Xi'an Jiaotong University in 2014, doctoral student, focusing on nuclear reactor thermal-hydraulic analysis" ]
  王成龙, E-mail:Wclo29o29@stu.xjtu.edu.cnWANG Chenglong, E-mail:Wclo29o29@stu.xjtu.edu.cn
- 基金信息：
  
  国家国际科技合作专项(2014DFA60650);博士后创新人才计划(BX201600124)
- DOI：10.11889/j.0253-3219.2017.hjs.40.080601
  中图分类号： TL333
- 收稿日期：2017-01-03，
  
  修回日期：2017-04-10，
  
  纸质出版日期：2017-08-10
- 稿件说明：
移动端阅览
魏诗颖, 王玮, 王成龙, 等. TACR热传输系统热工水力瞬态分析[J]. 核技术, 2017,40(8):080601

Shiying WEI, Wei WANG, Chenglong WANG, et al. Thermal-hydraulic transient analysis for heat transport system of thorium-based advanced CANDU reactor[J]. Nuclear techniques, 2017, 40(8): 080601
魏诗颖, 王玮, 王成龙, 等. TACR热传输系统热工水力瞬态分析[J]. 核技术, 2017,40(8):080601 DOI： 10.11889/j.0253-3219.2017.hjs.40.080601.

Shiying WEI, Wei WANG, Chenglong WANG, et al. Thermal-hydraulic transient analysis for heat transport system of thorium-based advanced CANDU reactor[J]. Nuclear techniques, 2017, 40(8): 080601 DOI： 10.11889/j.0253-3219.2017.hjs.40.080601.

摘要

钍燃料的利用对于缓解核燃料资源短缺具有重要意义，坎杜型反应堆（Canadian Deuterium Uranium，CANDU）在堆芯布置、中子利用效率及先进燃料循环方面具有较高的灵活性，使得其在CANDU反应堆中引入钍燃料循环更具现实意义。CANDU型反应堆中钍基燃料应用关键基础技术研究是加拿大与我国正在开展的合作课题，其中开发自主的CANDU堆堆芯热工水力设计和安全分析程序是钍基燃料应用必不可少的设计工作之一。本文针对CANDU型反应堆热传输系统结构特点，采用FORTRAN程序设计语言开发了适用于CANDU型反应堆热传输系统的热工水力瞬态分析程序CANTHAC（CANDU Thermal-Hydraulic Analysis Code）。利用CANTHAC对钍基先进CANDU堆（Thorium-based Advanced CANDU Reactor，TACR）进行了瞬态分析，计算工况包括满功率稳态、无保护蒸汽发生器（Steam Generator，SG）二次侧给水温度降低事故及完全失流事故。其中，满功率稳态计算结果与清华大学设计的钍基先进CANDU堆TACR设计值吻合较好，相对误差不超过2%，在可接受范围内；无保护SG二次侧给水温度降低事故及完全失流事故在计算条件下所得的燃料温度及系统压力等关键热工水力参数均在安全限值内，满足安全准则要求。程序为模块化编程，便于移植和改进，具有一定的通用性，为进一步研究工作奠定了基础。

Abstract

Background

The utilization of thorium in reactors is of great significance to alleviate the shortage of nuclear fuel and Canadian deuterium uranium (CANDU) reactor is one of the best choice to be introduced in thorium fuel cycles. The abundant thorium deposit and rich experience in operating Qinshan phase Ⅲ laid a foundation for researches on thorium-based advanced CANDU reactor (TACR) in China.

Purpose

To support the international cooperative study on TACR between Canada and China and lay a foundation for self-development of the software for CANDU thermal hydraulic analysis in our country

it is necessary to develop analysis code in order to evaluate thermal hydraulic performance of CANDU reactor heat transport system.

Methods

In this study

a CANDU thermal hydraulic transient analysis code (CANTHAC) was developed using FORTRAN language and used to analysis the transient performance of TACR heat transport system. The full power steady state

steam generator (SG) feed water temperature reduction accident and complete loss of flow accident were simulated.

Results

The steady state results by CANTHAC were in good agreement with the rated value designed by Tsinghua University and the maximum calculation error was less than 2%

which is acceptable. The calculation results of SG feed water temperature reduction accident and complete loss of flow accident shown that key parameters including the fuel temperature and system pressure do not exceed specified limits and satisfy the safety criteria.

Conclusion

The steady state results agree well with the rated value and the transient results are reasonable

which preliminarily evaluate the validity of the transient analysis code. Since the code was modularized

it would be convenient for further modification and application.

关键词

Keywords

references

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