RELAP5缓发中子先驱核输运模型扩展及验证  

Improvement and validation of the delayed neutron precursor transport model in RELAP5 code for liquid fuel molten salt reactor

作　　者：李锐 程懋松[2] 戴志敏[1,2] LI Rui;CHENG Maosong;DAI Zhimin(ShanghaiTech University,Shanghai 201210,China;Shanghai Institute of Applied Physics,Chinese Academy of Sciences,Shanghai 201800,China)

机构地区：[1]上海科技大学,上海201210 [2]中国科学院上海应用物理研究所,上海201800

出　　处：《核技术》2021年第6期89-97,共9页Nuclear Techniques

基　　金：中国科学院战略先导科技专项(No.XD02001005)资助。

摘　　要：液态燃料熔盐堆作为第四代核反应堆概念之一,在安全、经济、防核扩散方面都具有独特的优势。液态燃料熔盐堆特有的中子动力学和热工水力学特性,致使传统固态燃料堆系统分析程序不再适用于液态燃料熔盐堆的瞬态分析和安全评估。为了提高反应堆系统安全分析程序RELAP5/Mod4.0(Reactor Excursion and Leak Analysis Program)在液态燃料熔盐堆安全分析中的适用性和精确度,基于一维缓发中子先驱核输运模型和二阶Godunov数值方法扩展了RELAP5/Mod4.0点堆模型。采用美国橡树岭国家实验室(Oak Ridge National Laboratory,ORNL)熔盐实验堆(Molten Salt Reactor Experiment,MSRE)实验数据,对采用新点堆模型的RELAP5/Mod4.0程序进行了验证。结果显示:改进后的RELAP5/Mod4.0程序数值结果与实验结果吻合较好,表明该模型和数值方法能够较好地模拟缓发中子先驱核输运过程,满足液态燃料熔盐堆安全分析要求,能够应用于液态燃料熔盐堆安全分析。[Background]The liquid-fueled molten salt reactor(MSR)is one of the Generation IV advanced reactor concepts,which has unique advantages in aspects of safety,economy and nonproliferation.However,in view of the special neutron dynamics and hydrodynamics of the liquid-fueled MSR,traditional system analysis codes for solid fuel reactors cannot be applied to liquid-fueled MSR directly without improvement.[Purpose]This study aims to improve the applicability and accuracy of the RELAP5/Mod4.0 code in liquid-fueled MSR simulation.[Methods]First of all,the RELAP5/Mod4.0 code was expanded by adding a 1-D delayed neutron precursor(DNP)transport model to the code,hence the DNP concentration was calculated by an algorithm using a second-order Godunov method.The DNP model was coupled with built-in point kinetics model and hydrodynamics model.Then the code was validated by experimental data of the molten salt reactor experiment(MSRE)at Oak Ridge National Laboratory(ORNL),USA.Four transient scenarios studied in the MSRE were reproduced,including pump startup,pump coast down,natural-convection circulation and reactivity perturbation transients.[Results]The results show that the numerical results are in good agreement with the experimental results,and the model with second-order Godunov method can simulate the delayed neutron precursor transport well.[Conclusions]The modified RELAP5/Mod4.0 code is suitable for liquid-fueled MSR transient and safety analysis.It provides a better and more powerful tool for design and optimization of the liquid-fueled MSR.

关 键 词：熔盐堆 一维缓发中子先驱核 输运模型 二阶Godunov 

分 类 号：TL426[核科学技术—核技术及应用]