基于波形逼近的快中子脉冲堆耦合计算

快中子脉冲堆在爆发脉冲过程中的中子输运与热弹性力学相互耦合,该耦合作用过程决定了脉冲特性。基于绝热近似下燃料元件温升始终正比于系统总裂变数的事实,提出了通过调整参数使温升随时间变化的曲线逼近裂变率曲线的耦合计算方法。在迭代逼近过程中,采用了有限元商业软件ANSYS处理力学建模和热弹性力学求解,利用点堆方程描述中子学行为,两者利用基于微扰理论的反应性反馈方程进行耦合。通过调整参数使力学模型的温升加载函数波形逼近通过输运计算得到的裂变率波形,直至两者一致。以Lady Godiva脉冲堆为例的裂变产额计算结果与实验结果一致,该计算方法有望用于快中子脉冲堆的研究和设计。     

点堆中子动力学方程的蒙特卡罗方法

基于广义半马尔科夫过程(GSMP)模拟方法实现点堆中子动力学方程的蒙特卡罗求解。该方法模拟裂变系统内中子数和缓发中子先驱核数目的瞬态演化过程,并计算出任意时刻裂变功率和缓发中子源强等物理量。利用本文提出的方法研究快中子增殖堆(简称"快堆")和热堆参数下的点堆动力学方程,对反应性的阶跃输入、斜坡输入和振荡输入的点堆中子场瞬态过程进行模拟,并与传统数值算法的计算结果进行比较。该方法不存在数值计算的刚性问题,能方便地对复杂反应性输入过程进行计算,并能充分考虑瞬态过程中反应性变化对中子代时间的影响。     

反应性动态加入对脉冲堆中子脉冲波形的影响

在点堆模型基础上开发了脉冲模拟程序,该程序考虑了缓发中子效应、裂变热反馈效应和脉冲棒的行进过程,可以描述脉冲堆爆发脉冲全过程的中子强度、裂变能以及反应性的变化,并经过实验数据验证。用该程序计算CFBR-Ⅱ堆在反应性动态加载过程中全波形、峰功率、反应性等脉冲特征参数,结果表明:脉冲引发的越晚,其峰功率和裂变产额越大,而且其最大裂变产额与静态爆发脉冲情况下的裂变产额相同。     

熔盐快堆稳态核热耦合程序开发及验证

无慢化罐式堆芯结构的熔盐快堆(Molten Salt Fast Reactor,MSFR)中存在中子物理与热工水力的强耦合。应用耦合蒙特卡罗粒子输运程序OpenMC与计算流体力学软件OpenFOAM,建立了一套适用于熔盐快堆的三维稳态核热耦合计算程序。该程序基于python编程语言实现了OpenMC和OpenFOAM二者间的功率、燃料盐温度和密度分布等数据交互,可以获得堆芯内三维功率分布、中子通量密度分布、三维速度场和温度场分布。采用该耦合程序,建立了熔盐快堆的基准模型,研究了中子学区域划分数目和初始条件对keff、燃料盐速度和温度分布的影响。根据研究结果,推荐了一套合理的中子学区域划分方法与数目,表明了耦合程序设定的不同初始条件对keff结果无影响。最后,通过与熔盐快堆基准结果的对比验证了耦合程序的正确性,表明该程序适用于熔盐快堆的稳态核热耦合分析。     

基于离散纵标输运计算方法的三维燃耗程序发展研究

为了精确描述和分析具有强烈各向异性中子注量率空间分布的反应堆燃耗过程,本文实现了三维SN 输运计算与燃耗计算的耦合,发展了相应的三维输运燃耗耦合计算程序.该程序系统采用接口程序自动耦合三维SN输运计算程序和同位素燃耗计算程序的方法实现对三维中子学计算模型的精细燃耗计算,获得燃料同位素成分、燃耗反应性、中子注量率空间分布等参数随燃耗时间的变化量.采用IAEA 基准校核例题对程序系统进行了校核,计算结果初步证明了所开发的三维燃耗程序系统的正确性.     

热管式空间反应堆燃耗计算研究

针对热管式空间反应堆，基于OpenMC程序产生均匀化截面参数，并由确定论快堆分析程序SARAX进行堆芯输运及燃耗计算。以蒙特卡罗程序（MCNP）的输运计算结果以及MVP程序的燃耗计算结果作为参考解，通过对比稳态输运计算和燃耗计算的结果，证明了耦合的OpenMC和SARAX程序系统对于空间堆中子学分析和燃耗分析的适用性和高效性。为热管式空间反应堆的设计分析提供了参考。      

HFETR入单晶硅的反应性扰动分析

单晶硅由辐照孔道进入高通量工程试验堆(HFETR)堆芯时会引入反应性扰动和影响局部的中子注量分布。本文使用蒙特卡洛核粒子输运程序(MCNP5)和蒙特卡洛核粒子输运扩展程序(MCNPX2.6)耦合建立了HFETR数学计算模型,通过临界计算验证了模型的可用性,模拟计算了不同质量单晶硅由8#辐照孔道进入堆芯所引入的阶跃反应性扰动,并分析了8 kg单晶硅由8#辐照孔道入堆对邻近电离室孔道内轴向中子注量分布的扰动情况。研究结果表明,单晶硅入堆所引入的反应性扰动很小,符合安全要求,对邻近电离室孔道内局部的中子注量分布存在一定影响,可能会对相应的中子探测仪表产生干扰。     

铀氢锆瞬态实验装置计算程序的研发

本文针对新型核反应堆电源系统物理特性研究的铀氢锆瞬态实验装置，结合点堆动力学方程、非稳态热传导方程和热弹性动力学方程，研发了专用于铀氢锆瞬态实验装置实验过程模拟的瞬态计算程序。以保健物理研究堆（HPRR）为例，利用研发的程序进行模拟，得到HPRR的功率、反应性、温度和径向位移随时间的变化。程序的计算结果与Fuchs-Hansen模型的结果较为一致，验证了该瞬态计算程序的正确性。     

溶液堆落棒及弹棒事故分析

基于实验给出的溶液堆的气泡模型和温度模型,分别用点堆动力学和三维中子输运理论对溶液堆的瞬态进行了模拟和分析。利用研制的程序,对溶液堆不同工况、引入不同反应性的情况进行了模拟,得到了溶液堆可稳定的功率水平和事故情况下的功率波动。数值计算结果表明,基于点堆动力学和反应性反馈机制建立的模型,计算速度快,适合对溶液堆进行在线模拟和快速分析;而基于三维中子输运理论建立的模型,采用改进的准静态方法进行求解,计算精度较高,计算速度可接受,可用来对溶液堆进行精确的安全分析。     

基于不动点理论的核热耦合计算收敛特性研究

对于一些具有强烈核热耦合行为的新型反应堆(如超临界水堆),一般的迭代方法不再适用。本文基于不动点理论提出核反应堆核热耦合计算的数学模型,结合超临界水堆的计算实例,对核热耦合计算的迭代方法进行收敛性分析,总结出核热耦合计算的收敛判定方法,通过该方法可实时得到迭代的收敛情况,同时提出了自适应松弛因子及其算法,并编写了实现该算法的程序。     

PARCS code multi-group neutron diffusion constants generation using Monte Carlo method

PARCS code is a three-dimensional (3D) reactor core simulator which solves the steady-state and time-dependent multi-group neutron diffusion equations if the multi-group diffusion constants (MGDCs) are provided. The MGDCs are mostly prepared for reactor physics problems using deterministic lattice codes. Beside approximation in the geometry, a lattice code inherently applies estimates to the neutron transport model. On the other hand, the geometric flexibility and use of continuous energy cross sections data library associated with the Monte Carlo (MC) method makes it a good candidate for the generation of highly accurate multi-group cross sections. In this study, a new MC based methodology is applied to generate the MGDCs which can be utilized in the PARCS code input file directly or as PMAXS files for a reactor core simulation. To achieve this, a new tool in MATLAB software is developed to compute the MGDCs from the MCNPX 2.7 MC code outputs. Verification of the proposed method for two-group constants generation is carried out using Tehran research reactor (TRR) core simulation in different steady state conditions. The calculated values of axial and radial power distributions and multiplication factor using the PARCS code are verified against the MCNPX 2.7 code results. The results illustrate that the proposed method has high accuracy in MGDCs generation.     

PWS: an efficient code system for solving space-independent nuclear reactor dynamics

The reactor kinetics equations are reduced to a differential equation in matrix form convenient for explicit power series solution involving no approximations beyond the usual space-independent assumption. The coefficients of the series have been obtained from a straightforward recurrence relation. Numerical evaluation is performed by PWS (power series solution) code, written in Visual FORTRAN for a personal computer. The results are applied to the step reactivity insertion, ramp input, zigzag input, and oscillatory reactivity changes. When the reactivity is given, including the case in which the feedback reactivity is a function of neutron density, the developed method can provide a straightforward procedure for computing reactor dynamics problems. The solution of this method was compared to some other analytical and numerical solutions of the point reactor kinetics equations; the results proved that the approach is both efficient and accurate to several significant figures.     

基于IQS/MC方法的ADS次临界反应堆中子时空动力学模拟分析

采用改进准静态近似与蒙特卡罗中子输运程序相结合(IQS/MC)的方法实现了加速器驱动的次临界系统(ADS)中子时空动力学模拟计算。以加速器驱动嬗变研究装置的靶堆耦合参考方案物理模型为例,通过对束流瞬变引入和燃料组件提升两种工况进行动态模拟,计算得到了堆芯总的相对功率、分能群相对中子注量率及相对功率三维网格分布随时间的变化。将IQS/MC方法计算结果与点堆计算结果进行了对比分析,模拟结果符合物理规律,两种方法对比结果与国外相关文献一致,表明IQS/MC方法适用于ADS次临界反应堆中子时空动力学过程的瞬态安全分析。     

高温气冷堆中子时空动力学实时仿真计算

本文研究开发了三维圆柱几何堆芯多群中子时空动力学改进准静态方法模拟计算程序。对给定的模块式高温气冷堆模型进行了模拟计算。在初始状态下,该程序的计算结果与中子扩散程序CITATION的计算结果吻合很好。在动态情况下,模拟了堆芯反应性、堆内各能群中子平均注量率和堆芯相对功率等物理量随时间的变化。计算结果与理论分析一致,在一定精度下,可达到实时仿真计算的要求。     

基于R2S方法的反应堆结构材料活化剂量计算研究

反应堆结构材料在堆芯中子辐照下由于中子活化反应而产生大量的放射性核素,其衰变光子是反应堆停堆检修、换料、退役过程中工作人员职业照射剂量的重要来源。本文基于严格两步法（R2S）,研究了反应堆结构材料栅元活化计算方法,并基于蒙卡粒子输运程序（MCNP）与点活化计算程序（ORIGEN）建立了反应堆结构材料活化剂量计算软件（MOCA）。通过开发功能接口与数据接口程序实现输运程序与活化计算程序的自动耦合,进而实现“中子输运-活化分析-剂量计算”全自动耦合分析。利用M5包壳活化计算模型、不锈钢活化计算模型和NUREG/CR-6115压水堆模型对MOCA进行基准验证,证明了MOCA的正确性与可靠性。      

An analytical solution for the point reactor kinetics equations with one group of delayed neutrons and the adiabatic feedback model

The point reactor kinetics equations with one group of delayed neutrons and the adiabatic feedback model are solved analytically. The analytical solution is based on an expansion of the neutrons density in powers of the small parameter, the prompt neutrons generation time, into the second order differential equation in the neutron density. The relation between the time and the reactivity for reactor excursions near prompt critical is derived. Also, the neutron density and the average density of delayed neutron precursors as functions of reactivity are presented. The relations of reactivity, neutron density and temperature with time are calculated, drawn, and compared with other analytic method.     

基于FLUENT的多物理场耦合研究

基于FLUENT的多物理场耦合分析是当前核安全分析的热点问题。本文运用6组缓发中子的点堆动力学模型（PKM）编写了反应堆核功率计算程序，利用外部调用耦合和用户自定义函数（UDF）动态链接库耦合方法分别建立了FLUENT-REALP5耦合分析模型和FLUENT-PKM耦合分析模型，并在单相范围内利用水平分支管道的喷放问题和线性反应性引入的超功率瞬变问题验证了耦合模型的正确性和有效性。本研究的耦合分析方法可以为FLUENT的多物理场核安全分析提供支撑。      

Study on kinetics of fast subcritical assembly—Contribution of delayed neutrons to the transition after a reactivity insertion in a subcritical system—

This study is an investigation of the effect of the delay neutron on the kinetics in the subcritical system. And, it proposes a method necessary for the kinetics code development that uses the Monte Carlo (MC) computation.

It is generally difficult to analyze three dimensional space and time dependent kinetics by using a MC method. It is because the sampling of the neutron in a region becomes difficult when conditions of the region changes with time. In this study, we consider about the effect of delayed neutron in the kinetics of ADS. The behavior of neutrons is considered spontaneous in this system. It means a neutron is absorbed or leaks in a short period, while the conditions of region do not change. Therefore they are treated by steady state calculation. On the other hand the densities of delayed neutron precursors changes slowly, and the conditions of region change. In the concept of developed MC method, the neutrons are calculated by using steady state equation at each time point, and the delayed neutron precursors are calculated by using time dependent equation. We tried to inspect the accuracy of this method by using a point equation. We obtained strict solution Φ* as a reference solution, Φ1 as a solution by the present method, and Φ2 as the solution where both neutrons and delayed neutron precursors are treated by using static equations. The obtained results show a good agreement between Φ1 and Φ*, though the Φ2 agrees with Φ* poorly in all cases. Especially, we showed that this technique was effective from the reactivity change by ADS, and the relation of a delayed neutron. Finally, the effect of the delay neutron on the beam trip in the neutron source for the drive was examined by using the technique of Φ2.     

Development and validation of the code COUPLE3.0 for the coupled analysis of neutron transport and burnup in ADS

The analysis of the fuel depletion behavior is critical for maintaining the safety of accelerator-driven subcritical systems(ADSs). The code COUPLE2.0 coupling 3-D neutron transport and point burnup calculation was developed by Tsinghua University. A Monte Carlo method is used for the neutron transport analysis, and the burnup calculation is based on a deterministic method. The code can be used for the analysis of targets coupled with a reactor in ADSs. In response to additional ADS analysis requirements at the Institute of Modern Physics at the Chinese Academy of Sciences, the COUPLE3.0 version was developed to include the new functions of(1) a module for the calculation of proton irradiation for the analysis of cumulative behavior using the residual radionuclide operating history,(2) a fixed-flux radiation module for hazard assessment and analysis of the burnable poison, and(3) a module for multi-kernel parallel calculation, which improves the radionuclide replacement for the burnup analysis to balance the precision level and computational efficiency of the program. This paper introduces thevalidation of the COUPLE3.0 code using a fast reactor benchmark and ADS benchmark calculations. Moreover,the proton irradiation module was verified by a comparison with the analytic method of calculating the210 Po accumulation results. The results demonstrate that COUPLE3.0 is suitable for the analysis of neutron transport and the burnup of nuclides for ADSs.