Accident analyses of station blackout for TMSR-SF2

Bo XU; Yang ZOU; Qiang SUN; Xiaohan YU

doi:10.11889/j.0253-3219.2017.hjs.40.100601

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Accident analyses of station blackout for TMSR-SF2

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

- Accident analyses of station blackout for TMSR-SF2
- NUCLEAR TECHNIQUES Vol. 40, Issue 10, Article number: 100601(2017)
- 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.中国科学院大学北京 100049
- 作者简介：
- 基金信息：
  
  Strategic Priority Program of Chinese Academy of Sciences
- DOI：10.11889/j.0253-3219.2017.hjs.40.100601
  CLC： TL364.4
- Received：24 April 2017，
  
  Revised：03 May 2017，
  
  Published：2017-10
- 稿件说明：
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Bo XU, Yang ZOU, Qiang SUN, et al. Accident analyses of station blackout for TMSR-SF2[J]. Nuclear techniques, 2017, 40(10): 100601
DOI：

Bo XU, Yang ZOU, Qiang SUN, et al. Accident analyses of station blackout for TMSR-SF2[J]. Nuclear techniques, 2017, 40(10): 100601 DOI： 10.11889/j.0253-3219.2017.hjs.40.100601.

摘要

钍基熔盐堆-固态燃料二号堆（Thorium Molten Salt Reactor-Solid Fuel 2，TMSR-SF2）是基于球床熔盐堆SF1（Solid Fuel 1）的小型模块化升级堆型，这种新概念堆结合了两者的诸多优点，目前已经完成了预概念设计，对其进行典型事故的分析与安全特性的评估成为当前重要研究内容。本文基于Relap5/MOD4.0程序，建立了反应堆事故模型，进行了全厂断电事故的模拟，分析了反应性、反应堆功率、冷却剂温度和燃料温度等关键参数的变化规律。结果表明，SF2在全厂断电事故中具备高度安全性，其中固有安全性发挥了重要作用。此外还进行了全厂断电事故伴生不同事件的后果比对以及不同温度反应性系数的敏感性分析，证明了直接反应堆辅助冷却系统（Direct Reactor Auxiliary Cooling System，DRACS）在事故前期余热排出能力的局限性，而依靠主泵可以最大限度利用熔盐堆的热惰性从而显著缓解熔盐堆堆芯过热。

Abstract

Background

Thorium fluoride salt reactor-solid fuel 2 (TMSR-SF2) is a small module reactor (SMR) based on the experimental prototype SF1. Its concept design combines many outstanding characteristics from fluoride high temperature reactor (FHR)

and brings in or enhances many more attractive features inspired from the small module idea.

Purpose

This study aims to evaluate its operation performance and assess the safety limits during postulated transients.

Methods

A preliminary analysis of station blackout (SBO) in SF2 is presented by using RELAP5/MOD4.0 code

relative hydraulic components have been equivalently modeled.

Results

Modification result shows that an immediate safety shutdown can be obtained and guaranteed by strong thermal inertia while passive safety system based on natural circulation can maintain long term cooling of the reactor core and guarantee no server consequences and sufficient redundancy. Several postulated initiating events and sensitivity analyses covering several reactivity coefficients are also performed to reveal their influence degree and tendency on final results.

Conclusion

The conclusion is that the shutdown system is crucially responsible for short-term safety while a long-term reactor cooling has to depend on the direct reactor auxiliary cooling system (DRACS).

关键词

Keywords

references

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