MCNP程序计数分段卡存在的问题及解决办法

当应用蒙特卡罗程序(MCNP)的计数分段卡(FS卡)配合其他计数卡进行分段计数时,可能给出错误的计数结果。通过比较FS卡对同一几何体的3种模型的分段计数结果发现,填充形式模型的燃料棒区域的计数结果不正确。其原因在于燃料棒栅元由栅元集合卡(Universe卡)和填充卡(Fill卡)填充得到,而分段计数时MCNP程序认为被填充的栅元仍位于Universe卡定义的几何体处。据此提出了在Universe卡定义的几何体处对被填充栅元进行分段计数的解决方法。对例题的径向中子注量率分布的计算表明,该方法能够解决FS卡对填充形式描述的几何体进行分段计数时出现的错误。     

MCNP源子程序在停机剂量率计算中的应用

在核聚变装置的停堆剂量率的计算中,通常采用MCNP程序来实现光子的输运计算,但由于聚变装置几何和材料的高度复杂性使得栅元数量庞大,同时由于标准MCNP在进行光子输运计算时,SDEF通用源卡只支持1 000个以下的栅元描述,因此直接采用SDEF通用源卡的方法无法实现聚变堆的停堆剂量率精确计算与分析。本论文采用MCNP内置源子程序方法直接对衰变光子源进行抽样,解决了SDEF通用源卡受限的问题。以国际热核聚变实验堆ITER最新发布的停机剂量率基准例题以及ITER-T426基准实验例题对源子程序进行了校验,结果表明了该方法的可用性与正确性。     

函数展开计数在CLUTCH方法中的初步应用

反复裂变几率(IFP)方法广泛应用于求解k特征值对连续能量核数据的灵敏度系数,然而IFP方法存在内存占用大的问题,因此CLUTCH方法被提出以解决该问题。但对于大规模问题,如压水堆全堆问题,基于网格的CLUTCH(CLUTCH-Mesh)方法存在权重函数不易收敛的问题。本文采用函数展开计数(FET)方法对CLUTCH方法中的权重函数进行统计(CLUTCH-FET)以解决该问题,函数展开计数选取的基函数是勒让德多项式。本文在蒙特卡罗粒子输运计算程序NECP-MCX中实现了IFP、CLUTCH-Mesh和CLUTCH-FET 3种方法,以IFP方法的计算结果作为参考解,对CLUTCH-Mesh和CLUTCH-FET方法的精度和效率进行了验证。数值结果表明：对于小规模问题,如Godiva和Flattop问题,CLUTCH-Mesh和CLUTCH-FET方法具有与IFP方法相当的精度,且计算效率较IFP方法更高;对于大规模问题,如AP1000全堆问题,CLUTCH-Mesh方法的计算精度下降,而CLUTCH-FET方法可保持较高的精度和计算效率,CLUTCH-FET方法的品质因子较IFP方法和...     

300~#研究堆安全棒中子注量率计算中的减方差方法对比及应用

针对300#研究堆安全棒顶端中子注量率计算中的小体积远距离中子输运问题,采用MCNP减方差方法,通过多次试算与验证计算,对比了各减方差方法在本问题中的适用情况,得到了符合误差要求的计数值。首先,根据本研究堆几何建模模型,选取了点探测器及相关的3种减方差方法进行对比计算,计算结果显示:随着粒子数的增加,计算呈现不稳定现象,未起到减方差的作用。之后,将原有几何模型重新分层建模,并分配适当的栅元重要性,将中子引向目标栅元。按此方法,通过检查中子在各栅元中的径迹及碰撞情况,并增加粒子数计算,得到的计数结果表明:计数相对误差在5%以内,品质因子保持稳定,实现了减方差的目的。     

中子输运计算界面流方法的微扰计算

以六角形几何中子积分输运计算界面流算法及其相对应的数学共扼方程计算为基础,利用微扰原理计算了当反应堆六角形组件中栅元核参数发生微量变化时系统反应性的变化。计算结果表明,本文所开发的基于六角形几何中子积分输运算法的微扰计算方法是正确的。     

基于CAD邻居列表和包围盒的蒙特卡罗粒子输运几何跟踪加速方法研究

几何跟踪主要进行蒙特卡罗粒子输运计算中粒子位置和径迹长度的计算,它是蒙特卡罗粒子输运计算的关键技术之一。由于聚变堆几何结构极其复杂,使得几何跟踪在整个蒙特卡罗粒子输运计算中占据30%~80%的计算时间,因此几何跟踪方法的效率是决定聚变堆蒙特卡罗粒子输运计算效率的重要因素之一。本文提出了基于CAD的邻居列表和包围盒加速方法,并基于FDS团队自主研发的超级蒙特卡罗核计算仿真软件系统SuperMC进行实现。利用聚变堆FDS-Ⅱ和ITER模型对本文方法进行了数值验证,测试结果表明本文方法不影响计算结果,并能使程序计算效率提高50%~60%,证明了本文方法的正确性和有效性。     

基于JMCT的大亚湾核电站反应堆精细建模与计算

基于通用型蒙特卡罗中子-光子耦合输运程序JMCT搭建了大亚湾核电站反应堆精细模型,计算了有效增值因子(keff)和部分局部计数量,并和蒙特卡罗粒子输运程序MCNP的结果比对。结果显示,计算量相对误差均小于10-3量级,吻合度较高,验证了JMCT大规模精细几何建模和几何处理的能力。     

基于严格两步法的停堆剂量率计算程序的基准分析

基于蒙卡程序cosRMC的网格计数功能,开发了以严格两步法为核心的停堆剂量率计算程序,通过耦合粒子输运计算和活化分析计算,精确求解停堆剂量场。其中,采用ALARA程序开展活化分析计算,将程序应用于ITER诊断窗口计算基准题上,开展了充分的计算分析,并与其他严格两步法程序计算得到的停堆剂量率结果有较好的一致性。另外,由于聚变装置几何十分复杂,结构网格难以准确描述几何结构,往往一个网格包含多个栅元,此时网格的通量平均对停堆剂量率的精度会带来不好的影响,而非结构网格具有良好的几何适应性,因此,基于非结构网格对停堆剂量率程序作了进一步开发,并在基准题上开展计算分析,验证了程序基于非结构网格计算停堆剂量率的可靠性。     

JMCT蒙特卡罗中子-光子输运程序全堆芯pin-by-pin模型的模拟

几何栅元数超过千万、计数达数十亿、模拟粒子数达数百亿规模的反应堆全堆芯pin-by-pin问题是目前国际公认的挑战计算机和计算方法的难题。由于巨大的数据量已超过单核内存的极限,必须进行空间区域分解和数据分解。本文利用基于JCOGIN实体组合几何框架自主开发研制的三维中子 光子输运蒙特卡罗程序JMCT,通过空间区域分解和嵌套并行,完成了对大亚湾核电站1号机组反应堆全堆芯pin-by-pin模型的建模和模拟,计算给出了每个pin的注量率分布及其误差。     

基于非均匀射线追踪的栅元模块化特征线方法研究

宏带方法是一种基于非均匀射线追踪的特征线方法。当系统几何结构复杂、不连续边界较多时，由该方法产生的特征线密度过大，计算量也急剧增加。为了克服该问题，提出了一种基于栅元模块化的非均匀射线追踪方法，以减小系统中宏带及特征线的总长度，进而减小计算量。对于栅元耦合边界处特征线的错位问题，引入了边界离散方法，避免了复杂的边界角通量插值计算。数值结果表明，基于非均匀射线追踪的栅元模块化特征线方法的计算效率和计算精度较高，能满足压水堆常见栅格中子输运计算的需求。     

An easy to implement global variance reduction procedure for MCNP

In this article, an easy to implement global variance reduction procedure is demonstrated. The procedure is capable of flattening the relative error distribution in a global mesh tally covering the geometry of a complex system. The method is fully based on the I/O capabilities of the MCNP Monte Carlo radiation transport code family and requires little additional support code and almost no user judgment to work properly. Increases, with respect to a standard simulation, in a global figure of merit ranging up to a factor 276 for a PWR core configuration have been observed. Also, the relative error distribution was flattened substantially in all studied systems by the application of the procedure and the number of empty mesh tally cells was reduced.     

Verification of the pulse height tally in MCNP 5

Pulse height tallies are commonly used in Monte Carlo codes to predict detailed measured photon spectra for spectrometry purposes. The pulse height tally is unique among the various tallies in MCNP. Unlike flux or current tallies, which are calculated as soon as the particle exits or collides in the cell, the entire set of tracks for a history must be completed before the pulse height tally can be made. The objective of this work was to verify the pulse height tally and prepare to verify the new MCNP 5 variance reduction features with the pulse height tally.In this paper, we give details to the analytic solution of the pulse height distribution using a modification to Shuttleworth’s fictitious elements, report MCNP 5 results for the pulse height tally, energy deposited and current tallies for the problem.     

Particle Population Diagnosis and Euclidean Minimum Spanning Tree in Monte Carlo Calculation of Power Distribution

A mesh-input-free method is not yet established for particle population diagnosis in the power distribution calculation by the Monte Carlo source iteration method. To approach this issue, the Euclidean minimum spanning tree (EMST) in the graph theory was applied to the source particles. A characteristic volume that one particle covers was defined to be the cubic power of the average edge length of EMST. Thirty and one hundred times of the characteristic volume were proposed as weak and strong requirements, respectively, for a minimum tally cell volume since ten particle characteristic volumes can be accommodated within the tally cells producing one-third and 10% of average power density. These requirements were examined against a three-dimensional full-core model of a 1,100MWe pressurized water reactor. The comparison with the population diagnosis with a mesh in Nucl. Sci. Eng., 158, 15 (2008) shows a lot of promise of the EMST-based approach. Further developmental issues are identified concerning computational time and output fitting. In addition, a practically useful result was obtained as follows. If the three-dimensional uniform tally cells have volume larger than the quarter fuel bundle unit, the EMST-based approach yields a less conservative diagnosis than fissile volume per source particle.     

Fractal dimension analysis for run length diagnosis of Monte Carlo criticality calculation

In Monte Carlo criticality calculation (MCCC), each output quantity of interest is a series of tallies under autocorrelation. As a consequence from the functional central limit theorem, the stepwise interpolation of standardized tallies (SIST) converges in distribution to Brownian bridge (BB). Here, the standardization of tallies is a functional version of the statistic in the central limit theorem with the sample mean at each generation and the true mean replaced by the sample mean at the final generation. Fractional Brownian motion (FBM) is a family of stochastic processes including Brownian motion and assumes a unique value of the box-counting dimension (BCD) in fractal geometry. This work shows that the BCD of SIST is an effective diagnostic measure for the run length of MCCC. The sufficiency of the number of generations run can be judged by relating the BCD of FBM to that of BB. Numerical results are presented for the tallies of representative autocorrelation characteristics in a three-dimensional model of a pressurized water reactor and the effective multiplication factor (k_eff) tallies of the criticality problem by D. Mennerdahl.     

有限元SN中子输运模拟的区域分解并行

确定论中子输运方法具有计算速度快、可获取物理量的精细场分布、可高效多物理耦合等优点，随着有限元方法在中子输运模拟中的应用，复杂几何结构、大尺度下的屏蔽问题和临界问题都能得到高保真建模和分析。离散纵标（S_N）法是求解中子输运方程的有效数值方法，基于OpenMP并行机制对各独立离散方向进行并行求解，可提高S_N输运模拟的计算速度，但并行规模较有限。对几何空间进行区域分解并采用MPI并行机制，可实现大规模并行扩展，进而实现对大型问题的高精度快速求解。本文采用并行自适应非结构网格应用框架JAUMIN进行区域分解和进程间通信，通过并行S_N扫描实现了自主有限元输运程序ENTER的高效并行，完成正确性检验后在天河Ⅱ号超级计算机上使用1 440个CPU核完成了1.43×10⁷网格单元、2.81×10⁹自由度规模问题的测试，计算时间约7.4 h。表明该程序具备了有效模拟大型复杂结构中子输运问题的能力，具有一定工程应用价值。     

JCOGIN粒子追踪算法优化

本文分析了多种先进蒙特卡罗程序的CSG粒子追踪算法，对JCOGIN粒子追踪模块进行了优化。粒子定位算法采取了边界粒子定位和源粒子位置预估，能减少一定量的粒子定位计算；径迹求交算法采取了安全距离优化和布尔二叉树展开，其中安全距离优化可减少电子输运径迹求交次数，布尔二叉树展开能使布尔体求交算法的时间复杂度降为O（n）。应用4个典型算例测试了优化效果，结果表明，粒子定位算法优化对于一般问题具有一定的优化效果，安全距离优化显著提升了电子输运效率，布尔二叉树展开大幅提升了JMCT对于非规则复杂几何的计算速度。     

全局减方差方法的HBR-2基准题应用

对于深穿透类型的屏蔽问题，在合理的时间内计算得到可信的结果对于蒙特卡罗(MC)方法是一个巨大的挑战。基于离散纵标(S_N)方法的局部和全局减方差方法能有效降低MC计算深穿透问题的计数误差。本文基于HBR-2基准题比较了全局减方差方法和局部减方差方法的计算效率。结果表明，对于HBR-2基准题，局部和全局减方差方法均取得了较好的结果。全局减方差方法1次计算即可同时优化辐照监督管和堆外探测器的计数，因此实际应用更加方便和高效。     

Global variance reduction method for global Monte Carlo particle transport simulations of CFETR

It can be difficult to calculate some under-sampled regions in global Monte Carlo radiation transport calculations. The global variance reduction(GVR) method is a useful solution to the problem of variance reduction everywhere in a phase space. In this research, a GVR procedure was developed and applied to the Chinese Fusion Engineering Testing Reactor(CFETR). A cylindrical CFETR model was utilized for comparing various implementations of the GVR method to find the optimum.It was found that the flux-based GVR method could ensure more reliable statistical results, achieving an efficiency being 7.43 times that of the analog case. A mesh tally of the scalar neutron flux was chosen for the GVR method to simulate global neutron transport in the CFETR model.Particles distributed uniformly in the system were sampled adequately through ten iterations of GVR weight window.All voxels were scored, and the average relative error was 2.4% in the ultimate step of the GVR iteration.     

Particle and Heavy Ion Transport code System,PHITS, version 2.52

An upgraded version of the Particle and Heavy Ion Transport code System, PHITS2.52, was developed and released to the public. The new version has been greatly improved from the previously released version, PHITS2.24, in terms of not only the code itself but also the contents of its package, such as the attached data libraries. In the new version, a higher accuracy of simulation was achieved by implementing several latest nuclear reaction models. The reliability of the simulation was improved by modifying both the algorithms for the electron-, positron-, and photon-transport simulations and the procedure for calculating the statistical uncertainties of the tally results. Estimation of the time evolution of radioactivity became feasible by incorporating the activation calculation program DCHAIN-SP into the new package. The efficiency of the simulation was also improved as a result of the implementation of shared-memory parallelization and the optimization of several time-consuming algorithms. Furthermore, a number of new user-support tools and functions that help users to intuitively and effectively perform PHITS simulations were developed and incorporated. Due to these improvements, PHITS is now a more powerful tool for particle transport simulation applicable to various research and development fields, such as nuclear technology, accelerator design, medical physics, and cosmic-ray research.     

Optimization treatment of point-wise nuclear data in Monte Carlo criticality and burnup calculations

The Monte Carlo codes used for neutron transport calculations are always time consuming, a large proportion of which is possessed by the treatment of continuous-energy cross sections. In this paper, two companion methods are developed for the optimization treatment of point-wise nuclear data, the first of which is called Computational-Expense Oriented (CEO) method based on the unionized energy grid approach and reconstructs only the computationally expensive cross sections in neutron transport simulation, and the other of which is called energy bin (EB) method, a companion of CEO method when the reaction rate tallies for MC-coupling burnup calculation are performed. These two methods are implemented in the code RMC, a Monte Carlo (MC) code used for reactor analysis, and tested on fast reactor core and BWR assembly problems. The numerical results show that CEO method, in comparison with reconstructing all cross sections under the unionized grid, requires the sharply decreased computer memory while achieving almost the same computational efficiency, and EB method can optimize the processing of nuclide-specific energy grid search and thus effectively reduce the total search number while requiring very small computer memory.