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.

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.

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.

k_(eff)uncertainty quantification and analysis due to nuclear data during the full lifetime burnup calculation for a small-sized prismatic high temperature gas-cooled reactor

To benefit from recent advances in modeling and computational algorithms,as well as the availability of new covariance data,sensitivity and uncertainty analyses are needed to quantify the impact of uncertain sources on the design parameters of small prismatic high-temperature gascooled reactors(HTGRs).In particular,the contribution of nuclear data to the k_(eff)uncertainty is an important part of the uncertainty analysis of small-sized HTGR physical calculations.In this study,a small-sized HTGR designed by China Nuclear Power Engineering Co.,Ltd.was selected for k_(eff)uncertainty analysis during full lifetime burnup calculations.Models of the cold zero power(CZP)condition and full lifetime burnup process were constructed using the Reactor Monte Carlo Code RMC for neutron transport calculation,depletion calculation,and sensitivity and uncertainty analysis.For the sensitivity analysis,the Contribution-Linked eigenvalue sensitivity/Uncertainty estimation via Track length importance Characterization(CLUTCH)method was applied to obtain sensitive information,and the "sandwich" method was used to quantify the k_(eff)uncertainty.We also compared the k_(eff)uncertainties to other typical reactors.Our results show that ^(235)U is the largest contributor to k_(eff)uncertainty for both the CZP and depletion conditions,while the contribution of ^(239)Pu is not very significant because of the design of low discharge burnup.It is worth noting that the radioactive capture reaction of ^(28)Si significantly contributes to the k_(eff)uncertainty owing to its specific fuel design.However,the k_(eff)uncertainty during the full lifetime depletion process was relatively stable,only increasing by 1.12%owing to the low discharge burnup design of small-sized HTGRs.These numerical results are beneficial for neutronics design and core parameters optimization in further uncertainty propagation and quantification study for small-sized HTGR.

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.

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。

高通量快堆辐照生产^(252)Cf关键因素分析

^(252)Cf放射源是一种高强度同位素中子源,在科学研究、装置研发等领域具有重大需求,但长期以来依赖于进口。基于高通量快堆初步设计方案开展辐照靶件设计,结合模拟计算结果得出影响^(252)Cf核素生产的关键因素。在5×10^(15)n·cm^(-2)·s^(-1)中子通量水平下,对三种采用不同氢化锆和Eu_(2)O_(3)吸收体辐照靶件设计方案进行裂变沉积能、能谱计算,并通过燃耗计算程序开展重锔、轻锔靶件燃耗计算,与美国辐照生产^(252)Cf实验值对比。结果表明:方案3的计算值与基准值吻合较好且最适用于重锔靶辐照生产^(252)Cf,方案2适用于通过短的辐照周期辐照生产重锔核素。本文的计算与分析可为高通量快堆辐照生产^(252)Cf提供理论与技术支撑。

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。

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。

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。

Monte Carlo studies on the burnup measurement for the high temperature gas cooling reactor

Online fuel pebble burnup measurement in a future high temperature gas cooling reactor is proposed for implementation through a high purity germanimn （HPGe） gamma spectrometer. By using KORIGEN software and MCNP Monte Carlo simulations, the single pebble gamma radiations to be recorded in the detector are simulated under different, irradiation histories. A specially developed algorithm is applied to analyze the generated spectra to reconstruct the gamma activity of the ~arCs monitoring nuclide. It is demonstrated that by taking into account the intense interfering peaks, the 137Cs activity in the spent pebbles can be derived with a standard deviation of 3.0% （l（r）. The results support the feasibility of utilizing the HPGe spectrometry in the online determination of the pebble burimp in future modular pebble bed reactors.

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。

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。

弥散颗粒毒物的多尺度耦合燃耗算法

弥散颗粒毒物具有特殊的空间结构和较强的空间自屏效应,在燃耗过程中容易出现微观分层现象。在数值模拟中,直接精细化求解将带来巨大的计算量和网格密度,具有一定的挑战。为此,本文提出了一种全新的基于多尺度耦合的燃耗计算方法:通过微观精细球层模型和宏观均匀栅元模型的耦合,将弥散颗粒介质的精细求解问题简化成对一个简单常规介质的快速求解,解决了弥散颗粒毒物在全局范围内精细燃耗求解过程中所面临的计算量大和网格密度高等问题,准确地表征弥散颗粒毒物的燃耗特征,为弥散颗粒介质的求解提供了一个新思路。经过初步验证,该算法在有效增殖因数、中子通量密度和核素核数密度等中子物理参数中有较好的表现。

不同热管工质对热管冷却反应堆堆芯物理参数的影响与分析

热管冷却反应堆采用非能动传热技术,热响应速度快,可避免堆芯单点失效,具有功率密度大、寿命长、环境适应性强、工作性能稳定等特点,是目前空间核反应堆研究的热点。基于清华大学开发的反应堆蒙特卡洛中子输运程序RMC(reactor Monte Carlo code),以美国爱荷华国家实验室(Idaho National Laboratory,INL)设计的热管冷却反应堆INL Design A为研究对象,选取3种热管工质开展热管冷却反应堆堆芯物理计算。计算结果表明:锂热管工质不仅拥有很好的热物性参数,并且使用锂热管工质的热管冷却反应堆缓发中子有效份额最大、中子能谱较硬、燃耗反应性损失最小、增殖性能最佳,有利于热管冷却反应堆堆芯小型化与长寿命。因此,推荐锂为热管冷却反应堆的热管工质。

蓄热式加热炉节能管理与操作浅析

文章针对包钢长材厂1号线蓄热式加热炉吨钢燃耗明显偏高的现象,根据坯料在加热炉内的传热情况,从加热炉管理与加热炉操作两方面进行了详细分析,得出燃耗偏高的原因。根据蓄热式加热炉的工作原理和操作特点有针对性地提出了细化、强化加热炉热工管理、提高热装率、减少热损失、控制排烟温度、合理控制炉膛压力等一系列改进措施及操作优化,有效地降低了煤气消耗,节约了生产成本,取得了一定经济效益。

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。