Development and validation of depletion code system IMPC Burnup for ADS

Depletion calculation is important for studying the transmutation efficiency of minor actinides and longlife fission products in accelerator-driven subcritical reactor system(ADS). Herein the Python language is used to develop a burnup code system called IMPC-Burnup by coupling FLUKA, OpenMC, and ORIGEN2. The program is preliminarily verified by OECD-NEA pin cell and IAEAADS benchmarking by comparison with experimental values and calculated results from other studies. Moreover,the physics design scheme of the CIADS subcritical core is utilized to test the feasibility of IMPC-Burnup program in the burnup calculation of ADS system. Reference results are given by the COUPLE3.0 program. The results of IMPC-Burnup show good agreement with those of COUPLE3.0. In addition, since the upper limit of the neutron transport energy for OpenMC is 20 MeV, neutrons with energies greater than 20 MeV in the CIADS subcritical core cannot be transported; thus, an equivalent flux method has been proposed to consider neutrons above 20 MeV in the OpenMC transport calculation. The results are compared to those that do not include neutrons greater than 20 MeV. The conclusion is that the accuracy of the actinide nuclide mass in the burnup calculation is improved when the equivalent flux method is used. Therefore, the IMPC-Burnup code is suitable for burnup analysis of the ADS system.     

铅基快堆RBEC-M燃耗库制作方法研究及验证

      提出了一套新的方法流程，用来处理和生成燃耗计算所需的数据。利用核数据处理程序NJOY处理评价数据库ENDF-B-Ⅶ.1生成33群的MATXS格式库，再根据具体问题中的材料信息，经截面处理程序MGGC处理得到相关核素的微观、宏观截面，经自编写的处理模块Triso对其进行格式转化、合并，最终得到提供给燃耗计算程序使用的ISOTXS库文件，其中一般核素以微观截面的形式表示，裂变产物以类似宏观截面的伪裂变产物形式表示。对铅冷快堆基准题900 MW RBEC-M进行了计算，采用REBUS-3进行燃耗计算，对比了结果中的有效增殖系数k_eff随燃耗的变化趋势、功率分布以及中子能谱，最终结果与参考报告较为符合，初步验证了这一系列燃耗库制作流程的可行性。      

Validation of IRBURN calculation code system through burnup benchmark analysis

Assessment of the reactor fuel composition during the irradiation time, fuel management and criticality safety analysis require the utilization of a validated burnup calculation code system. In this work a newly developed burnup calculation code system, IRBURN, is introduced for the estimation and analysis of the fuel burnup in LWR reactors. IRBURN provides the full capabilities of the Monte Carlo neutron and photon transport code MCNP4C as well as the versatile code for calculating the buildup and decay of nuclides in nuclear materials, ORIGEN2.1, along with other data processing and linking subroutines. This code has the capability of using different depletion calculation schemes.     

基于最佳一致逼近多项式的燃耗计算方法研究

本文研究了一种基于最佳一致逼近多项式（MMPA）的燃耗计算方法求解燃耗方程。相比于切比雪夫有理近似方法（CRAM）和围道积分有理近似方法（QRAM）,MMPA方法只需一次矩阵求逆计算即可求解燃耗方程,且所有计算都是实数运算,具有数值稳定性好、求解效率高等优点。进一步研制了基于MMPA方法的点燃耗程序AMAC,并耦合蒙特卡罗输运程序OpenMC,采用衰变例题、固定辐照例题、OECD/NEA压水堆栅元燃耗基准题和沸水堆组件燃耗基准题进行验证,程序计算结果与实验值及各参考值吻合良好,初步验证了MMPA方法在理论和数值上的正确性和有效性。     

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。      

IAEA加速器驱动系统(ADS)中子学基准问题的计算

为了验证加速器驱动洁净核能系统研究拟采用的程序系统,对IAEA加速器驱动系统(ADS)中子学第一阶段基准问题进行了校算。其中,散裂中子源的计算采用LAHET程序;中子输运计算采用MCNP程序,核素的燃耗计算采用0RIGEN2。计算结果与IAEA的ADS研究协调项目组(ADSCRP)成员俄罗斯和瑞典的结果吻合较好。     

基于蒙特卡罗方法与指数欧拉法耦合的反应堆三维燃耗计算程序的研发

燃耗计算在反应堆设计、分析研究中起着重要作用.一维、二维耦合燃耗程序因其几何限制难以满足先进反应堆精细设计分析的要求.本文研发了基于蒙特卡罗方法与指数欧拉法耦合的三维燃耗程序.程序采用编写耦合MCNP与FISPACT的接口程序的方法,结合了MCNP处理复杂几何能力强,FISPACT计算核素全面、能谱多样的特点,实现了考...     

基于蒙特卡罗方法的IVR熔融池内热源时序模型构建及分析

堆内熔融物滞留（IVR）作为反应堆严重事故的关键缓解策略,目前已广泛应用于新一代压水堆（PWR）。针对IVR的有效性,如熔融池内对流、下封头传热、壁面临界热流密度（CHF）的估算等研究,是该领域数年来的热点。针对上述问题,国内外先后开展了数起实验,如COPO、BALI、SEMICO、COPRA等,并基于实验结果展开了大量数值模拟,以探索IVR下的传热规律,为其性能及设计提供参照。本文基于中子物理蒙特卡罗程序RMC对压力容器下封头熔融池模型进行了细网格建模及材料填充,并通过燃耗/衰变热计算DEPTH程序构建了熔融池内热源时序模型。研究结果显示,该模型能体现熔融池内热源变化趋势,得到的时序数据对IVR的进一步研究有重要意义。     

HARMONY程序计算中子扩散方程高阶λ本征值问题的基准验证

高阶λ谐波在反应堆堆芯功率重构、换料优化、ADS次临界反应堆物理特性研究等领域有着重要应用价值。为进行高阶λ谐波的计算,本文基于隐式重启动Arnoldi方法(IRAM)编制了可用于一维、二维、三维笛卡尔坐标系中子扩散方程的任意阶λ谐波及本征值计算的HARMONY程序,并进行了基准题的数值验证。结果表明,HARMONY程序能实现高阶λ本征值问题计算,具有较高的精度,为未来基于λ谐波的ADS次临界反应堆物理特性研究奠定了基础。     

聚变-裂变混合能源堆球模型中子学对算研究

利用蒙特卡罗程序和自主开发的蒙特卡罗-燃耗耦合程序MOCouple-s,对北京应用物理与计算数学研究所提出的聚变-裂变混合能源堆球模型进行了对算研究。对初始时刻及各燃耗时刻下的有效增殖因数、能量倍增因子、氚增殖比、中子源强度等堆芯参数进行了比较,结果总体符合较好。对寿期末重要核素的成分进行了详细比较,除个别核素外,偏差很小,表明所采用的计算程序与核参数库一致性良好。对核参数库的选择、铀水体积比等对燃耗计算结果的影响进行敏感性分析,并对外中子源驱动的次临界堆芯的燃耗计算进行详细讨论,提出可行的燃耗计算基准。     

反应堆核-热-燃耗多物理耦合框架研究与应用

近年来随着高性能计算技术的不断发展,依托先进的超级计算机和数学物理计算方法,对核反应堆开展多物理、多尺度计算成为前沿研究热点。根据反应堆堆芯多物理耦合分析需求,研究了多物理耦合算法,构建了基于中子输运、燃耗、热工子通道的堆芯多物理耦合系统,完成耦合程序开发,实现中子物理、燃耗、热工子通道的多物理耦合计算。利用压水堆组件模型与快堆模型开展输运-燃耗耦合计算测试和核-热耦合计算测试,初步验证了耦合系统功能。     

二维输运燃耗耦合程序开发与验证

为准确计算反应堆内燃耗问题,建立了基于二维离散纵标法及BATEMAN燃耗方法的输运燃耗耦合计算方法,并开发相应的计算程序。基于ENDF/B-Ⅶ评价库开发了175群中子和42群光子截面数据库MUSE-F1.0,采用OECD/NEA发布的MOX燃料快堆基准题对耦合计算方法及程序系统进行验证计算。结果表明,耦合计算程序结果与基准题吻合良好,误差在8%以内,初步验证了耦合计算程序在快堆嬗变工程应用中的可行性。     

Calculation of fuel burnup and radionuclide inventory in the Syrian miniature neutron source reactor using the GETERA code

Calculations of the fuel burnup and radionuclide inventory in the Syrian miniature neutron source reactor (MNSR) after 10 years (the reactor core expected life) of the reactor operation time are presented in this paper using the GETERA code. The code is used to calculate the fuel group constants and the infinite multiplication factor versus the reactor operating time for 10, 20, and 30 kW operating power levels. The amounts of uranium burntup and plutonium produced in the reactor core, the concentrations and radionuclides of the most important fission products and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core were calculated using the GETERA code as well. It is found that the GETERA code is better than the WIMSD4 code for the fuel burnup calculation in the MNSR reactor since it is newer, has a bigger library of isotopes, and is more accurate.     

Development and Validation of Burnup Code Based on High-order Chebyshev Rational Approximation Method

Based on high-order Chebyshev rational approximation method (CRAM), a point-burnup code named ICRAM was developed and internally coupled to Monte Carlo code OpenMC, forming a burnup calculation and analysis program OPICE. Compared with the traditional partial fraction decomposition (PFD) form of CRAM, the high-order incomplete partial fractions (IPF) form of CRAM has the characteristics of good numerical stability, high calculation accuracy and better step tolerance, etc., which meets the needs of high-fidelity burnup calculation development. In order to improve the accuracy of coupling calculations, two coupling strategies including prediction-correction method and sub-step method were implemented in OPICE. Three different calculation modes were supported by OPICE to execute the decay, constant flux and constant power calculations. By calculating the OECD/NEA burnup benchmark and fast reactor burnup benchmark, the calculation results of OPICE are in good agreement with the experimental data and each reference value. The correctness and validity of OPICE are verified preliminarily.     

Neutronics and burnup analysis of an accelerator-based tru-nitride fuel transmutation system with the ATRAS code

The neutronics and burnup analyses of an accelerator-based transmutation system with tungsten target and TRU-nitride fuel were performed with a newly developed code system named ATRAS (Accelerator-based Transmutation Reactor Analysis System). The ATRAS code is an integrated code system which can perform the hadronic cascade process above 20 MeV and neutron transport and core burnup process below 20 MeV with the spallation neutron source.

The specifications of the transmutation system are investigated. The core consists of the central spallation target region and the surrounding TRU-mononitride fuel region. The core is driven by protons at an energy of 1.0 GeV. This system was also proposed as a benchmark problem in the “OECD NEA/NSC Benchmark on Physics aspects of Different Transmutation Concepts”.

According to the calculation results by the ATRAS code, higher power density and transmutation rate were achieved with nitride fuel, and the neutron spectrum was slightly harder than that of the metallic fuel system. The burnup calculation for thermal power 800 MW was also performed with the ATRAS code. It is shown that about 300 kg of TRU are transmuted annually.     

基于高阶CRAM的燃耗程序开发与验证

本文基于高阶切比雪夫有理近似方法（CRAM）研制了点燃耗程序ICRAM,并内耦合于蒙特卡罗输运程序OpenMC,形成了一套燃耗计算分析程序OPICE。与传统部分分式分解（PFD）形式的CRAM相比,高阶不完全局部分解（IPF）形式的CRAM具有数值稳定性好、计算精度高和步长包容性更好等特点,满足高保真燃耗计算发展的需求。为提高耦合计算精度,OPICE采用了预估-校正和子步法两种耦合策略,支持纯衰变、定通量和定功率3种计算模式。通过OECD/NEA压水堆栅元燃耗基准题和快堆燃耗基准题的验证,程序计算结果与实验值及各参考值吻合良好,初步验证了OPICE的正确性与有效性。     

Study on kinetics of subcritical system: Contribution of delayed neutrons to the transient after a reactivity insertion

It is generally difficult to employ three dimensional spaces and time dependent kinetics by using a Monte Carlo method for nuclear reactor kinetics, because the random work analysis of a neutron becomes extremely difficult when conditions for neutron transport change with time.

However, in most cases of kinetic analysis, in particular relevant to ADS systems, we meet with many cases, where the lifetime of prompt neutron is much shorter than that of the neutron transport condition. In calculating the kinetic evolution of a subcritical (SC) system, in this paper we propose a new calculation technique where a steady state equation for the neutron flux and a time-dependent one for the delayed neutron precursor density will be treated.

In this paper, the accuracy of this theory is investigated by using a simple point reactor equation for several cases typical for ADS system. We obtained strict solution Φ* as a reference solution, Φ1 as a solution by the present method, and Φ2 as a 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 our cases. Finally, we can employ a conventional Monte Carlo code for neutron transport and analysis many kinetics problems of ADS systems with the help of delayed neutron precursor analysis.     

Calculation of radial burnup and nuclides atom density distributions in a VVER-1000 fuel rod using Monte Carlo method

A fundamental knowledge of fuel behavior in different situations is required for safe and economic nuclear power generation. Due to the importance of a fuel rod behavior modelling in high burnup, in this paper, the radial distribution of burnup, fission products, and actinides atom density and their variations by increasing burnup and other factors such as temperature, enrichment and power density are studied in a fuel pellet of a VVER-1000 reactor in an operational cycle using the MCNPX 2.7 Monte Carlo code. A benchmark including a Uranium-Gadolinium (UGD) fuel assembly is used for verification of the developed model in the MCNPX code for radial burnup calculation. A sensitivity study is carried out to investigate the effect of different parameters such as the number of particles per cycle, the number of geometrical radial nodes in the fuel pellet, the number of burnup steps and the selection of different fission-product contents (i.e. those isotopes that are used for particle transport) on the MCNPX model for speed and accuracy compromising. To calculate the radial temperature profiles and to analyze the effect of temperature on the radial burnup distribution and vice versa, the HEATING 7.2 code, which is a general-purpose conduction heat transfer program, and the MCNPX code are applied together. The results show the accuracy and capability of the proposed model in the MCNPX and HEATING codes for radial burnup calculation.     

Development of a Method for Evaluation of Pin-wise Power Distribution in Fuel Assemblies of Fast Reactors

In the design of fast reactor core with higher burnup and higher linear power, prediction accuracy of burnup history of fuel pin should be upgraded so as to assure fuel integrity without extra design margin under increased neutron fluence and burnup. A method is studied to predict fuel pin-wise power and its burnup history in fast reactors accurately based on an analytic solution of diffusion theory equation on hexagonal geometry with boundary condition from core calculation by finite-differenced diffusion calculation code. The present method is applied to a fast reactor core model, and its accuracy in predicting fuel pin power is tested. The result is compared with the reference solution by the finite difference calculation with very fine mesh. It is found that the present method predicts the power peaking factors in fuel assemblies accurately. The fuel pin-wise nuclide depletion calculation is also done using neutron fluxes for each fuel pin. The result shows that the fuel pin-wise depletion calculation is very important in predicting the burnup history of the fuel assembly in detail.     

加速器驱动的10 MW次临界反应堆物理方案研究

加速器驱动的次临界系统（ADS）是未来最有可能实现工业化嬗变核废料的装置。通过设计1个10 MW的ADS物理方案,研究ADS的嬗变能力。采用MCNPX和ORIGEN的耦合程序,利用基于ENDF6.8处理所得的6个温度（300、600、900、1 200、1 500、1 800 K）下连续能量核数据库,计算得到ADS随燃耗时间变化的有效增殖因数k_eff、功率峰因子和质子束流强度。同时通过计算给出了该设计方案下ADS燃料多普勒系数、冷却剂空泡系数和有效缓发中子份额,利用这些物理量研究了该ADS方案的安全特性,并通过燃耗计算研究了ADS的嬗变能力。结果表明,在1 000 d燃耗时长内,k_eff和质子流强随时间的波动较小,燃料燃耗深度较浅,系统可提升功率运行,在假想事故下系统能保持次临界状态。系统嬗变支持比约为8。