基于先驱核迁移与蒙特卡罗粒子输运耦合的熔盐堆缓发中子共轭通量计算

唐嘉诚; 朱贵凤; 贾舒阳; 曹金通; 吕善杰; 戴叶; 严睿; 邹杨

doi:10.3724/j.0253-3219.2026.hjs.49.250418

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基于先驱核迁移与蒙特卡罗粒子输运耦合的熔盐堆缓发中子共轭通量计算

中国核学会研究堆与新堆分会第二届学术年会优秀论文专刊 | 更新时间：2026-07-29

- 基于先驱核迁移与蒙特卡罗粒子输运耦合的熔盐堆缓发中子共轭通量计算
- Calculation of delayed neutron adjoint flux in molten salt reactor based on the coupling of precursor transport and Monte Carlo particle transport
- 核技术 2026年49卷第7期文章编号：071120
- 作者机构：
  
  1.中国科学院上海应用物理研究所钍基核裂变能全国重点实验室上海 201800
  2.中国科学院大学北京 100049
- 作者简介：
  
  唐嘉诚，男，2001年出生，2024年毕业于成都理工大学，现为硕士研究生，研究领域为反应堆物理
  朱贵凤，E-mail：zhuguifeng@sinap.ac.cn
- 基金信息：
  
  中国科学院战略性先导科技专项(XDA02010000);上海市自然科学基金(25ZR1401383)
- DOI：10.3724/j.0253-3219.2026.hjs.49.250418
  中图分类号： TL3
- 收稿：2025-11-09，
  
  修回：2026-01-13，
  
  网络首发：2026-07-29，
  
  纸质出版：2026-07-15
- 稿件说明：
移动端阅览
唐嘉诚,朱贵凤,贾舒阳等.基于先驱核迁移与蒙特卡罗粒子输运耦合的熔盐堆缓发中子共轭通量计算[J].核技术,2026,49(07):071120.

TANG Jiacheng,ZHU Guifeng,JIA Shuyang,et al.Calculation of delayed neutron adjoint flux in molten salt reactor based on the coupling of precursor transport and Monte Carlo particle transport[J].NUCLEAR TECHNIQUES,2026,49(07):071120.
唐嘉诚,朱贵凤,贾舒阳等.基于先驱核迁移与蒙特卡罗粒子输运耦合的熔盐堆缓发中子共轭通量计算[J].核技术,2026,49(07):071120. DOI： 10.3724/j.0253-3219.2026.hjs.49.250418. CSTR： 32193.14.hjs.CN31-1342/TL.2026.49.250418.

TANG Jiacheng,ZHU Guifeng,JIA Shuyang,et al.Calculation of delayed neutron adjoint flux in molten salt reactor based on the coupling of precursor transport and Monte Carlo particle transport[J].NUCLEAR TECHNIQUES,2026,49(07):071120. DOI： 10.3724/j.0253-3219.2026.hjs.49.250418. CSTR： 32193.14.hjs.CN31-1342/TL.2026.49.250418.

摘要

熔盐堆的熔盐燃料流动会引发缓发中子有效份额变化，将对熔盐堆的运行控制产生重要影响。本文基于蒙特卡罗方法获取缓发中子共轭通量分布，用于流动下的中子点堆模型分析。通过植入迭代裂变概率法（Iterated Fission Probability Method）代码来计算流动模型中缓发中子共轭通量，获取不同组缓发中子的有效份额，并使用熔盐实验堆（Molten Salt Reactor Experiment，MSRE）模型进行验证。验证表明，流动反应性变化计算与实验结果较为一致。进一步计算不同流速下缓发中子的轴向共轭通量分布。研究发现，流动会导致若干组缓发中子的一维轴向共轭通量降低，主要由堆外返回的缓发中子先驱核在堆内径向分布差异导致，而计算得到的缓发中子二维共轭通量分布则受流速干扰影响较小。该研究表明，更准确的先驱核流动中子点堆模型需在先驱核输运方面扩展为二维，同时还需考虑各组缓发中子的共轭通量差异。

Abstract

Background

The flow of molten salt fuel in molten salt reactors (MSRs) induces changes in the effective fraction of delayed neutrons

which exerts a significant impact on the operational control of MSRs.

Purpose

This study aims to use the Monte Carlo method coupled with precursors transport to obtain the adjoint flux distribution of delayed neutrons for the analysis of the flow-based point reactor model.

Methods

Firstly

based on the Monte Carlo method

the MCNP (Monte Carlo N Particle Transport Code) was employed in conjunction with a simplified flow field model. The adjoint flux of delayed neutron precursors in the flow model was calculated by incorporating the Iterated Fission Probability Method code

and the effective fractions of delayed neutrons in different groups were thereby obtained. Then

the MSRE (Molten Salt Reactor Experiment) model was utilized to verify the accuracy of the proposed method. Finally

by calculating the two-dimensional axial adjoint flux distribution of each group of delayed neutrons

the effect of flow on the adjoint flux was investigated

and the ratio of delayed neutrons effective fraction and delayed neutrons fraction under static conditions in each group were applied to eliminating the statistical fluctuations in the calculation.

Results

Calculation results show that the computed values of flow-induced reactivity changes are in good agreement with the experimental results

and certain differences exist between the adjoint flux of delayed neutrons and that of prompt neutrons. Additionally

the flow causes a reduction in the axial adjoint flux of several groups of delayed neutrons

is primarily attributed to the radial mixing inside the core of delayed neutron precursors returning from outside the core. The deviation of the two-dimensional adjoint flux distribution computation of delayed neutrons

is within 2.9%

which proves the feasibility of the scheme.

Conclusion

Results of this study suggest that the flow point reactor model needs to be extended to two dimensions in terms of precursor transport within the core

and the adjoint flux differences of delayed neutrons in each group should also be considered. It lays the foundation for the systematic transient analysis of molten salt reactors.

关键词

Keywords

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