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.

Development of a MCNP5 and ORIGEN2 based burnup code for molten salt reactor

The Molten Salt Reactor(MSR) is one of the six advanced reactor nuclear energy systems for further research and development selected by Generation IV International Forum(GIF),which is distinguished by its core in which the fuel is dissolved in molten fluoride salt.Because fuel flow in the primary loop,the depletion of MSR is different from that of solid-fuel reactors.In this paper,an MCNP5 and ORIGEN2 Coupled Burnup(MOCBurn) code for MSR is developed under the MATLAB platform.Some new methods and novel arrangements are used to make it suitable for fuel flow in the MSR.To consider the fuel convection and diffusion in the primary loop of MSR,fuel mixing calculation is carried out after each burnup time step.Modeling function for geometry with repeat structures is implicated for reactor analysis with complex structures.Calculation for a high-burnup reactor pin cell benchmark is performed using the MOCBurn code.Results of depletion study show that the MOCBurn code is suitable for the traditional solid-fuel reactors.A preliminary study of the fuel mixture effect in MSR is also carried out.

Determination of fuel burnup distribution of a research reactor based on measurements at subcritical conditions

This paper presents the determination of the fuel burnup distribution of the Dalat nuclear research reactor(DNRR) using a method of measurements at subcritical conditions. The method is based on the assumption of linear dependence of the reactivity on the burnup of fuel bundles and the measurements at subcritical conditions.The measurements were taken for seven selected fuel bundles in two different measuring sequences. The measured burnup values have also been compared with the calculations for verifying the method and the measurement procedure. The results obtained with the three detectors have a good agreement with each other with a discrepancy less than 1.0%. The errors of the measured burnup values are within 6%. Comparison between the calculated and measured burnup values shows that the discrepancy of the C/E ratio is within 9% compared to unity. The results indicate that the method of measurements at subcritical conditions could be well applied to determine the relative burnup distribution of the DNRR.

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.

基于蒙特卡罗方法的三维燃耗计算研究

采用通过编写连接MCNP程序和ORIGEN2程序的接口处理程序的方法进行快中子系统的燃耗计算。由MCNP、ORIGEN2、接口处理程序和截面文件组成的软件系统可用于燃料或堆芯非均匀布置快中子系统的燃料同位素成分和燃耗反应性损失计算,在燃耗反应性损失计算中采用了伪裂变产物的方法。介绍程序系统的研制情况,并给出用该软件系统计算中国实验快堆首炉堆芯和OECD/NEAMOX燃料快堆基准题的燃耗计算结果。

Data decomposition method for full-core Monte Carlo transport–burnup calculation

Monte Carlo transport simulations of a full-core reactor with a high-fidelity structure have been made possible by modern-day computing capabilities. Performing transport–burnup calculations of a full-core model typically includes millions of burnup areas requiring hundreds of gigabytes of memory for burnup-related tallies. This paper presents the study of a parallel computing method for full-core Monte Carlo transport–burnup calculations and the development of a thread-level data decomposition method. The proposed method decomposes tally accumulators into different threads and improves the parallel communication pattern and memory access efficiency. A typical pressurized water reactor burnup assembly along with the benchmark for evaluation and validation of reactor simulations model was used to test the proposed method.The result indicates that the method effectively reduces memory consumption and maintains high parallel efficiency.

改性色谱柱分离辐照后UO_(2)-Zr燃料中燃耗监测体钕的方法研究

钕监测体法整体精度高、适用范围广,是目前最常用的绝对燃耗测量方法之一。新型UO_(2)-Zr燃料中锆基体为燃料的溶解及燃耗监测体钕的分离带来了极大挑战,钕分离效率的提高对提升绝对燃耗测量可靠性、降低操作人员受照剂量具有重要意义。选择三种具有较强镧系元素萃取能力的酸性磷类萃取剂作为改性试剂,通过浸渍法完成商业化色谱柱的改性制备。比较研究表明,(2-乙基己基)膦酸单-2-乙基己基酯改性后的色谱柱分离性能最优,在0.05 mol·L^(-1)硝酸溶液的等度洗脱下,15 min内可实现钕的完全分离,Pr/Nd分离因子为1.73,而Nd/Sm分离因子远远大于1.5。60次重复分离实验证实改性色谱柱分离稳定性良好,在模拟辐照后UO_(2)-Zr燃料溶液中能实现大量铀、锆基体中微量钕的富集与分离。

地心甚高轨道星座构形协同捕获控制策略

针对地心甚高轨道星座构形协同捕获控制问题,基于虚拟编队方法设计了协同捕获控制策略,采用三脉冲燃耗最优轨迹规划算法对构形捕获轨迹进行协同规划;并且结合自适应全程积分滑模控制器对卫星各自转移轨迹进行跟踪控制。以10万km轨道高度的三星星座构形捕获为例进行仿真验证,仿真结果表明:该策略可以有效应用于地心甚高轨道星座构形捕获控制,能够在燃耗较少的情况下使星座中卫星同时到达各自的标称位置,同时具有较高的精度。

燃耗信任制技术在后处理厂中的应用

大通量后处理厂是动力堆乏燃料后处理经济性发展的必由之路,所处理乏燃料具有235U初始富集度高、燃耗深的特点,且单台设备处理量也很大,给工厂的临界安全设计和经济性带来了巨大挑战。文章借鉴国际先进后处理厂设计理念,结合连续溶解器和处理目标乏燃料的特点,提出了采用10个核素的锕系置信燃耗信任制等级的应用,并通过计算给出了溶解器处理燃料的燃耗限值、临界安全分析的接受准则等关键数据。成果表明,燃耗信任制技术在后处理厂的应用不但能从根本上解决工厂的重要核安全问题,还可以为工厂的经济运行带来诸多益处。

秦山一期核电站燃料组件加深燃耗后的燃料棒包壳应力、应变分析

秦山一期核电站第九循环燃料组件加深燃耗后考验组件卸料燃耗预计达到40000MWD／TU，其中热棒的平均燃耗将达到46083MWD/TU，热棒中热点燃耗将达到54175．9MWD／TU。为了确保核电站的安全运行，首先要保证燃料棒包壳的完整性，因此包壳的应力、应变分析十分重要。计算结果表明加深燃耗后的燃料棒应力、应变仍满足设计准则的要求。

Simulating the Production of Medical Radioisotopes in a Fast Thorium-Cycle ADS with SERPENT

Radiopharmaceuticals are used in nuclear medicine for diagnostic or therapeutic acts. The short decay half-lives of medical radioisotopes, especially those used for diagnostics, imply that they should be produced continuously and transported as quickly as possible to the medical units where they are used. Neutron-rich medical radioisotopes are generally produced in research reactors, like technetium-99m, lutetium-177, holmium-166 and iodine-131. On the other hand, proton-rich radioisotopes are produced via reactions with charged particles from accelerators like fluorine-18, gallium-67, iodine-123 and thallium-201. Beside this, innovative nuclear reactors are advocated as solutions to the issues of nuclear waste production and proliferation threats. Fast neutron, thorium-cycle and accelerator-driven subcritical (ADS) reactors are some of the most promising of them, proposed as safer fuel breeders and “waste burners”. This article examines the use of a fast thorium-cycle ADS with liquid lead-bismuth eutectic coolant for the production of molybdenum-99/technetium-99m and lutetium-177. Burnup simulation has been made with the Monte-Carlo (MC) code SERPENT. It is demonstrated that MC codes can advantageously be used to determine the optimal irradiation time for a given radioisotope in a realistic reactor core. It is also shown that fast thorium-cycle ADS is an economical option for the production of medical radioisotopes.

秦山核电厂燃耗深度对DNBR限值的影响分析

本文研究了在秦山核电厂批平均卸料燃耗深度超过原定限值的情况下，为保证电厂的安全运行对DNBR限值进行重新确认的问题。研究过程中，综合考虑了由燃耗加深所引起的核焓升因子的变化和棒弯曲这两种效应对DNBR限值的影响。结果表明，对秦山核电厂而言，在批平均卸料燃耗超过原定限值的情况下，维持原DNBR限值不变仍是安全可行的。

商用乏燃料后处理大厂设计燃耗浅析

燃耗参数是后处理厂设计的重要基础,涉及工厂处理乏燃料的类型、燃耗深度,影响工厂工艺、临界和辐射防护等专业设计,特别是工厂建设的经济性。中国核电堆型多样化发展,新燃料初始富集度略有不同,出堆乏燃料燃耗差异明显,商用后处理大厂的设计需考虑处理乏燃料的包络性、工厂的经济性和产品可应用性等问题,因此设计燃耗成为行业关注和讨论的重点。文章对中国商用后处理大厂拟处理机组的堆芯燃料管理方案进行研究,对出堆乏燃料的燃耗进行计算分析,就设计燃耗对中国商用后处理大厂的影响实施评价,给出商用后处理大厂设计燃耗建议。

压水堆不同燃耗完整和破损棒燃料芯块氧化特征拉曼光谱研究

为研究压水堆不同燃耗完整和破损燃料棒燃料芯块氧化过程和物相变化,采用拉曼光谱分析技术对燃耗为14 GW·d·t^(−1)和45 GW·d·t^(−1)的完整燃料棒及燃耗为14 GW·d·t^(−1)和41 GW·d·t^(−1)的破损燃料棒燃料芯块的氧化特征进行了分析。结果表明:14 GW·d·t^(−1)和45 GW·d·t^(−1)的完整燃料棒燃料芯块由UO2、U4O9和U3O8组成,相比于燃料芯块的内部区域,芯块边缘显示出更强的氧化性;14 GW·d·t^(−1)和41 GW·d·t^(−1)破损燃料棒燃料芯块发生了重结构,形成柱状晶粒,主要物相为UO2和U3O8。燃耗的加深和燃料棒的破损均促进了燃料芯块的氧化过程,但并不会改变燃料芯块的主要相结构。

基于NECP-Bamboo程序的商用压水堆乏燃料组件核素成分分析

乏燃料组件核素成分的精确计算是乏燃料临界安全分析等工作的输入条件,放射性源项计算是乏燃料组件核素成分分析的典型应用。国内现有程序由于存在数据库中核素种类不全、辐照过程无法完全模拟等弊端,限制了乏燃料后处理安全分析的可靠性和经济性。本文基于完全自主化的压水堆堆芯分析软件NECP-Bamboo,研发了商用压水堆乏燃料组件核素成分计算程序Bamboo-SFuel,利用辐照后实验(PIE)实测数据对核素成分进行了定量验证与分析,通过与Scale程序包计算结果进行对比验证了程序源项计算的精度,还探究了不同燃耗数据库对核素成分和源项计算结果的影响。数值结果表明,Bamboo-SFuel能精确分析不同辐照条件下商用压水堆乏燃料组件的核素成分和放射性源项,使用NECP-Bamboo程序中不同核素数目的燃耗数据库对重要核素成分计算结果影响不大,但对总的放射性源项计算结果影响较大;基于内置的包含1547种核素的燃耗数据库,该程序可同时给出可靠的乏燃料临界安全分析和辐射安全分析关注的重要核素成分。

多群输运-燃耗耦合蒙特卡罗计算(英文)

介绍三维多群中子输运 燃耗耦合P3近似蒙特卡罗程序MCMG BURN .该程序是在连续截面蒙特卡罗程序MCNP和反应堆栅元均匀化程序WIMS基础上发展而来的 .使用多群截面库模拟临界实验堆和高通量工程实验堆 ,取得了与MCNP和实验一致的结果 .MCMG BURN具有与MCNP相同的精度 ,但计算时间较MCNP要少得多 .

高通量工程试验堆临界蒙特卡罗计算

用新研制的三维多群P3 中子输运蒙特卡罗程序MCMG ,通过与栅元均匀化WIMS程序耦合 ,计算反应堆临界 燃耗问题。高通量工程试验堆 (HFETR)临界计算取得了与MCNP程序和实验一致的结果 ,且在相同计算精度下 ,MCGM计算时间较MCNP计算时间少数倍。

基于本征正交分解降阶的燃耗计算方法研究

随着堆芯物理计算方法向高保真发展,燃耗区的划分由传统的分米量级降低到厘米量级,全堆燃耗区数量将增加上千倍,燃耗计算精度提升的同时计算代价和复杂度亦陡增。本文研究了基于本征正交分解(POD)降阶的燃耗计算方法,通过建立参数化降阶模型,以1组完备正交基函数实现对各燃耗区燃耗矩阵的快速构造,并进一步研究了离线-在线计算策略以实现高效数值求解。基于JAEA发布的MOX燃料栅元基准题进行了计算分析,数值结果验证了本方法的正确性和可行性。本方法可为有效减少高保真燃耗计算成本、提高计算效率提供参考。

不同燃耗计算模型对商用压水堆乏燃料组件核素成分的影响分析

燃耗计算精度对提高乏燃料贮存效率有着重要影响,在应用燃耗信用制时,燃耗计算得到的核素成分偏差决定了乏燃料贮存的临界安全裕量。不同燃耗计算模型所得到的核素成分偏差各不相同,为提高燃耗计算精度,提出了一种装载不同燃料富集度的多组件燃耗计算模型,并使用不同燃耗计算模型分别对TMI-1反应堆NJ07OG组件中的6个样本进行了计算、对比和分析。结果表明,相比其他模型,考虑不同燃料富集度的多组件模型得到的235U、238U和239Pu等核素平均相对偏差更接近于零且6个样本的相对偏差分布更为平均。

应用于加速器驱动次临界系统的程序开发

开发了应用输运理论方法的两套程序系统 :蒙特卡罗 燃耗程序系统和输运 燃耗程序系统 .针对蒙特卡罗程序不能计算燃料的同位素成分变化 ,提出将它和燃耗计算程序耦合以完成燃耗计算 .燃耗程序所使用的单群截面由蒙特卡罗程序计算 .将用于常规反应堆的计算方法和策略推广到输运 燃耗程序系统 ,这使得计算步序大大简化 ,并改进了精度 .对加速器驱动次临界系统的基准题进行了校核计算 ,其结果和其他国家的计算结果符合良好 。