基于OpenMC的瞬发中子衰减常数计算模块开发与验证

瞬发中子基波衰减常数α可定量描述反应堆内中子随时间的变化,是计算绝对反应性所需的中子动力学参数之一,对次临界（特别是较深次临界）绝对反应性的精确测量具有重要意义。本文在开源程序OpenMC基础上,基于k α迭代方法,以中子径迹长度上的平均时间吸收权重修正作为k α迭代参数因子,在输运过程中对瞬发、缓发中子分别考虑,开发了具有瞬发α本征值问题计算功能的OpenMC PA模块。以Godiva衍生基准题和MUSE 4次临界实验装置为计算对象,对程序计算瞬发α本征值问题能力进行验证。结果表明,该计算模块有优于MCNP4C的计算速度与计算范围,计算值与参考值的相对误差小于05%。OpenMC PA能满足次临界系统瞬发α本征值和中子动力学参数计算需求。     

Comparative analyses of coated and composite UN fuel – Monte Carlo based full core LWR study

Longer continuous operation of a nuclear reactor leads to higher availability of nuclear power plant, entailing economic gain. In this context the option being explored, recently, is the use of higher density fuels as compared to current fuels i.e. UO₂. Uranium mono nitride (UN) fuel is one of the options being explored in this regard. UN fuel has been used in nuclear industry for a long time with fast reactor option. More recently, studies have targeted its use in Light Water Reactor (LWR) environment. The main problem with using UN fuel in LWR is its potential reaction with water which produces hydrogen. Researchers have proposed different approaches to overcome this problem.One option is the use of coatings around UN fuel pellet to avoid the direct contact of water with fuel. The second option is use of a secondary phase (10 vol. percent) like ZrO₂ which can make an oxide layer in case of contact with water and protect the main phase, Uranium mono nitride (80 vol. percent). Remaining 10 vol. percent is considered to be consumed by porosity. This study aims at comparison of neutron physics behavior of both these options in LWR environment. The upshot of the study is to quantify the impact of adding layers or secondary phase with respect to pure/complete UN fuel. To study these effects, two theoretical densities i.e. 95% and 80% for UN fuel are chosen for analyses. To avoid the problem of C-14 production from N-14, fuels studied are considered to be having 100% N-15.A validated model of Benchmark for Evaluation and Validation of Reactor Studies (BEAVRS) is used to perform all the analyses. Integral neutron physics parameters like neutron energy spectra, Radial Peaking Factors, Axial Peaking Factor, Doppler coefficient of reactivity, Isothermal Coefficient of reactivity and Temperature defect, Control Rod worth and excess reactivity for whole core are compared at Beginning of Cycle (BoC). Burnup obtained by different fuel option is also compared. Consistent with pin-cell based earlier findings, this full 3D analyses with UN based fuel shows noticeable spectral hardening leading to decrease in the value of control rod worth and less negative Doppler coefficient of reactivity while power peaking factors remain mostly unchanged. The economic advantage of switching to UN based fuels is expected when UN fuel above 80% TD is used. Approximately 19% increase in Equivalent Full Power Days (EFPDs) is witnessed by using 95% TD UN based fuels.     

Improved cache performance in Monte Carlo transport calculations using energy banding

We present an energy banding algorithm for Monte Carlo (MC) neutral particle transport simulations which depend on large cross section lookup tables. In MC codes, read-only cross section data tables are accessed frequently, exhibit poor locality, and are typically too much large to fit in fast memory. Thus, performance is often limited by long latencies to RAM, or by off-node communication latencies when the data footprint is very large and must be decomposed on a distributed memory machine. The proposed energy banding algorithm allows maximal temporal reuse of data in band sizes that can flexibly accommodate different architectural features. The energy banding algorithm is general and has a number of benefits compared to the traditional approach. In the present analysis we explore its potential to achieve improvements in time-to-solution on modern cache-based architectures.     

基于OpenMC的多群截面库制作及有效性验证

OpenMC是麻省理工大学计算反应堆物理组开发的开源蒙特卡罗程序,能够方便地制作适用于特定堆芯中子能谱分布的多群反应截面及高阶勒让德散射截面以用于离散坐标输运程序ANISN的计算。本文基于ENDF/B-Ⅶ.1和CENDL-3.1评价数据库,利用OpenMC计算制作了ANSIN格式的多群截面并通过基准题的计算验证计算结果的准确性。通过截面转换程序的编写,将OpenMC给出的堆芯各阶勒让德散射分量,堆芯中子能谱分布,散射、吸收反应率以及裂变中子产生速率等信息转换为ANISN程序可读取的截面库格式。采用制作的截面库利用ANINS计算有效中子增殖因子及堆芯中子通量分布。结果表明,ANISN确定论的计算结果与OpenMC给出的蒙特卡罗计算结果相吻合,验证了这种方法可有效地为ANISN提供截面数据,将来可推广应用于二维、三维确定论中子输运计算。     

基于OpenMC执行码的编译检测及临界基准验证

开源蒙特卡罗程序OpenMC(OpenMonte Carlo code)只提供源代码而没有执行码,在编译OpenMC的过程中发现不同版本的辅助程序与之存在兼容性问题。本文通过分析OpenMPI、Mpich及HDF5各版本辅助程序,对0.6.2版本OpenMC源代码的支持情况进行研究,为正确编译OpenMC执行码给出了直接参考。为进一步验证OpenMC执行码计算临界问题的正确性,选择国际临界安全基准评价实验手册(The International Criticality Safety Benchmark Evaluation Project,ICSBEP)中的96道代表性例题进行基准校验,与通用蒙特卡罗程序的计算结果进行对比并以实验值作为参考。结果表明,OpenMC计算值与实验值及其他程序计算值吻合较好,验证了OpenMC临界计算的可行性和正确性,上述结论将为程序以后的实际应用及完善奠定基础。