Develop and verify coupling program of the neutron physics and thermal hydraulic for MSR

Quan WEI; Wei GUO; Hailing WANG; Jingen CHEN; Xiangzhou CAI

doi:10.11889/j.0253-3219.2017.hjs.40.100605

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Develop and verify coupling program of the neutron physics and thermal hydraulic for MSR

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

- Develop and verify coupling program of the neutron physics and thermal hydraulic for MSR
- NUCLEAR TECHNIQUES Vol. 40, Issue 10, Article number: 100605(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：
- DOI：10.11889/j.0253-3219.2017.hjs.40.100605
  CLC： TL364
- Received：10 April 2017，
  
  Revised：12 May 2017，
  
  Published：2017-10
- 稿件说明：
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Quan WEI, Wei GUO, Hailing WANG, et al. Develop and verify coupling program of the neutron physics and thermal hydraulic for MSR[J]. Nuclear techniques, 2017, 40(10): 100605
DOI：

Quan WEI, Wei GUO, Hailing WANG, et al. Develop and verify coupling program of the neutron physics and thermal hydraulic for MSR[J]. Nuclear techniques, 2017, 40(10): 100605 DOI： 10.11889/j.0253-3219.2017.hjs.40.100605.

摘要

熔盐堆（Molten Salt Reactor，MSR）是第四代反应堆6种堆型中唯一的液态燃料反应堆，与固态燃料-液体冷却剂反应堆相比，原理上有较大不同。在熔盐堆中，流动的熔盐既是燃料又是冷却剂与慢化剂，中子物理学与热工水力学相互耦合；由于熔盐的流动性，缓发中子先驱核会随燃料流至堆芯外衰变，造成缓发中子的丢失，导致堆芯反应性降低。正是由于熔盐堆的这些新特性，造成熔盐堆内缓发中子先驱核、温度等参数变化与固态燃料反应堆有所不同，需要研究熔盐堆在各种工况下的相关物理参数变化。本文主要工作是考虑缓发中子先驱核的流动性对熔盐堆的影响，研究适用于熔盐堆的二维圆柱几何时空中子动力学程序及与之耦合的热工水力学程序；利用该程序对熔盐堆中子物理学和热工水力学进行耦合计算，验证熔盐堆相关实验数据；并且计算了熔盐堆无保护启停泵及堆芯入口温度过冷过热工况，用于分析熔盐堆的安全特性。计算结果表明，程序能够对熔盐反应堆实验（Molten Salt Reactor Experiment，MSRE）的相关实验数据进行较好的模拟计算，并且验证了熔盐堆的固有安全性。

Abstract

Background

As one of the six candidates for the Generation IV reactor types

molten salt reactor (MSR) is characterized by its use of the fluid-fuel. Compared with solid-fuel reactors

there are some differences in physics for liquid fuel reactor. The fluid-fuel not only serves as the fuel

but also serves as both the coolant and the moderator

hence a must for the neutron physics coupling with the thermal-hydraulic. Due to fuel flow in MSR

the delayed neutron precursors (DNP) partly flow out and decay outside of the reactor core

resulting in reactivity losses in the core.

Purpose

This study aims to develop and verify coupling program of neutron physics and thermal-hydraulic for MSR.

Methods

Distribution of DNP and temperature is different from solid-fuel reactor

it is necessary to study the change of relevant physical parameters in MSR under different operating conditions. Considering the flow of DNP impacting on MSR

a new space-time neutron dynamics program was developed to couple with the thermal-hydraulics program on the basis of two-dimensional RZ cylindrical geometry. The unprotected pump driven transient and inlet fuel temperature overcooling/overheating transient was simulated to analyze the intrinsic safety of MSR.

Results

The related experimental data in molten salt reactor experiment (MSRE) were verified by coupling calculation of neutron dynamics and thermal hydraulics

and the results showed that simulated data was in good agreement with the experiment results.

Conclusion

The program is available in simulating related experimental data in MSRE

and MSR has intrinsic safety.

关键词

Keywords

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