Coupled neutronic and thermal-hydraulic analysis of TMSR-SF1 at steady state

Jiahao CHEN; Haiqing ZHANG; Zhiyong ZHU

doi:10.11889/j.0253-3219.2017.hjs.40.080603

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Coupled neutronic and thermal-hydraulic analysis of TMSR-SF1 at steady state

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

- Coupled neutronic and thermal-hydraulic analysis of TMSR-SF1 at steady state
- NUCLEAR TECHNIQUES Vol. 40, Issue 8, Article number: 080603(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.40.080603
  CLC： TL426
- Received：10 March 2017，
  
  Revised：06 June 2017，
  
  Published：10 August 2017
- 稿件说明：
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Jiahao CHEN, Haiqing ZHANG, Zhiyong ZHU. Coupled neutronic and thermal-hydraulic analysis of TMSR-SF1 at steady state[J]. Nuclear techniques, 2017, 40(8): 080603
DOI：

Jiahao CHEN, Haiqing ZHANG, Zhiyong ZHU. Coupled neutronic and thermal-hydraulic analysis of TMSR-SF1 at steady state[J]. Nuclear techniques, 2017, 40(8): 080603 DOI： 10.11889/j.0253-3219.2017.hjs.40.080603.

摘要

固态燃料钍基熔盐堆（Thorium Molten Salt Reactor-Solid Fuel，TMSR-SF1）作为第四代先进核反应堆堆型之一，继承了熔盐冷却剂和球形燃料元件的许多优点和技术基础，具有良好的经济性、设计上的固有安全性、钍铀燃料的可持续性和防核扩散性。本文以10 MW固态燃料钍基熔盐堆为模型，利用MCNP（Monte Carlo N Particle Transport Code）和ANSYS Fluent等模拟程序对其进行多物理耦合分析，同时利用C++语言编写了堆芯活性区的物理-热工耦合计算程序，实现了MCNP计算结果与Fluent程序的对接，通过对比耦合前后结果，分析了堆芯功率密度分布、有效增殖因子、温度分布等主要参数，为熔盐堆的设计、安全性评估和操作运行提供了参考依据。

Abstract

Background

Neutronic and thermal-hydraulic simulations of advanced reactors can affect each other's results.

Purpose

This study focuses on coupling neutronic and thermal-hydraulic simulations to achieve more accurate results for future developments of 10-MW solid-fueled thorium molten salt experimental reactor (TMSR-SF1).

Methods

A program converting the MCNP (Monte Carlo N particle transport code) results to the spatial distribution of power density within the active region was created using C++ programming language. The spatial distribution data were loaded into the ANSYS Fluent in the form of user-defined function (UDF) to accomplish the coupling of the two simulation processes. In regards of TMSR-SF's original design parameters

the physical and thermal-hydraulic models of the whole core were established by using MCNP and ANSYS Fluent respectively.

Results

The coupling method is feasible and can be used to obtain reliable results. The changes in coolant's temperature and velocity in the active region are dependent on the power density distribution. The changes in multiplication factor

power density and maximum of discrepancy in coolant temperature are 1.08%

3.31% and 7.58 K

respectively.

Conclusion

It is necessary to take the coupling effects of the reactor core into consideration in the design of associated reactor systems. In addition

the results confirm that the design parameters of the TMSR-SF1 are reasonable.

关键词

Keywords

references

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