TRU燃料热谱MSR定期换料及嬗变特性数值研究

熔盐反应堆（MSR）燃料制备方便、中子经济性好、燃料管理灵活,具有直接利用轻水堆乏燃料中超铀核素（TRU）的潜力。本文通过优化燃料选取、栅格参数及燃料/石墨体积分数和去除裂变气体和惰性金属等方法,对TRU燃料热谱MSR堆芯寿期、TRU核素积存量、次锕系核素MA嬗变支持比和TRU焚毁率等进行计算分析,证明TRU燃料热谱MSR可实现长周期定期换料,减少在线换料的难度,同时对MA和TRU核素具有一定的嬗变能力,可降低乏燃料放射性毒性。      

三维六角形组件压水堆堆芯燃料管理计算及程序系统研究

介绍所研制的WWER型压水堆堆芯燃料管理计算程序系统TPFAP-H／CSIM-H，六角形组件均匀化计算程序TPFAP-H是在压水堆正方形组件程序TPFAP的基础上，采用穿透概率法与响应矩阵方法相结合计算六角形组件内中子能谱分布，并考虑六角形栅元特点改造开发而成的CSIM-H是以先进六角形节块扩散程序为基础．参照SIMULATE程序功能而研制的物理-热工水力耦合的三维六角形节块PWR堆芯燃料管理程序两者通过接口程序LINK连接起来，可以考虑燃耗，功率、慢化剂密度变化．控制棒、氙等参数的多种反馈效应对IAEA的WWER-1000型Kalinin核电厂基准问题的校算的结果表明，临界硼浓度、功率和燃耗分布等结果与国际各研究机构的结果吻合良好，偏差均在工程要求之内。     

六角形组件均匀化参数计算程序TPFAP-HEX的验证及其对WWER反应堆的应用

利用IAEA及文献提供的俄罗斯WWER-1000反应堆燃料组件计算的基准问题,对我国研制的压水堆六角形组件均匀化参数计算软件TPFAP-HEX进行了校核计算。通过和CASMO-HEX以及俄罗斯库尔恰托夫研究院等国外研究机构在组件k_∞、栅元的功率分布以及各种反应性效应等方面的比较,可以初步得出结论:TPFAP-HEX软件的研制是成功的,它基本达到了工程计算的功能和精度要求。     

基于CMS程序包的环形燃料堆芯物理计算分析

为研究在大型商用压水堆中采用环形燃料元件的可能性,需分析环形燃料的堆芯物理性能。本文研究了CASMO5程序计算环形燃料组件物理参数出现偏差的原因及其处理方法,分析了4组环形燃料先导组件加入秦山二期核电站平衡循环堆芯之后的堆芯物理参数。计算结果表明,装载的环形燃料先导组件对堆芯物理性能影响较小,基于CMS程序包开展环形燃料堆芯物理性能分析计算是可行的。     

角度相关不连续因子在JRR3M堆芯输运计算中的应用

针对含有强吸收体控制组件的日本研究堆JRR3M,在进行堆芯输运方程计算时,给出了角通量不连续因子(AFDF)的定义,并指出使用角通量不连续因子的必要性,提出使用迭代求解的方法来提高计算精度,并使之满足不连续因子自洽性。针对堆芯设计计算量大的特点,使用了超栅元近似方法。该方法能有效缩短计算时间,且灵活性强。利用组件形状函数,能重构出非均匀模型堆芯通量分布。最后讨论了扩散计算时不连续因子的选取问题,指出根据参考的不同,应选择不同的不连续因子。     

堆芯测量传感器发展现状及其在燃料元件考验中的应用

简单介绍了发展堆芯测量传感器的重要意义,我国近年来各种堆芯测量传感器的研制和发展现状以及某些堆芯传感器在秦山核电站燃料元件考验中的初步应用。这些传感器主要包括测量燃料中心温度的套管式高温W/Re热电偶组件,测量燃料包壳伸长的差动变压器型位移传感器,测量裂变气体内压的膜片式压力传感器,测量燃料棒相对功率分布的γ温度计,测量辐照然料元件中子通量和通量分布的自给能探测器和测量燃料包壳温度和考验元件出入口冷却剂温度的铠装热电偶等等。     

堆芯测量传感器发展现状及其在燃料元件考验中的应用

简单介绍了发展堆芯测量传感器的重要意义,我国近年来各种堆芯测量传感器的研制和发展现状以及某些堆芯传感器在秦山核电站燃料元件考验中的初步应用。这些传感器主要包括测量燃料中心温度的套管式高温W/Re热电偶组件,测量燃料包壳伸长的差动变压器型位移传感器,测量裂变气体内压的膜片式压力传感器,测量燃料棒相对功率分布的γ温度计,测量辐照燃料元件中子通量和通量分布的自给能探测器和测量燃料包壳温度和考验元件出入口冷却剂温度的铠装热电偶等等。     

压水堆六角形燃料组件均匀化 计算软件包TPFAP-HEX

介绍了所研制的具有工程实用价值的压水堆六角形燃料组件均匀化计算软件包。该组件中子空间能谱的计算采用穿透概率法与响应矩阵法相结合的方法,在六角形几何内求解中子积分输运方程。在此方法中,栅元内中子源采用空间线性或二次近似,栅元表面中子通量密度角分布采用简化6P     

环形燃料超临界水冷堆中子学计算方法研究

基于先进组件程序HELIOS和堆芯节块法程序SIXTUS,研发了超临界水冷堆（SCWR）的中子学计算程序FENNEL-N,并通过与蒙特卡罗程序对比分析了其用于环形燃料超临界水冷堆计算的精度。组件验证结果表明：制作多群数据库的压水堆能谱与超临界水冷堆能谱的差异是导致计算误差的主要原因。堆芯验证结果表明：传统的组件均匀化方法在计算超临界水冷堆时会引入较大误差。应用FENNEL-N程序对组件均匀化方法进行了研究,结果表明,采用优化的组件参数少群结构能减少堆芯能谱变化对精度的影响,采用超组件模型计算组件参数可考虑反射层对组件参数的影响。采用新的组件均匀化方法后,FENNEL-N的计算精度满足了预概念设计需求。     

基于查表方式的组件均匀化参数表达方式研究

为高精度地由事先算好的组件均匀化参数表产生堆芯计算所需的随组件当地工况变化的参数,发展了一套基于查表方式的处理方法。5个算例的检验,以及在秦山核电站堆芯计算中的应用表明,建立的这一套参数计算方案以及相应的作表表达方法是一种较高精度的方法,能较真实地反应组件参数随各独立变量的变化关系。     

基于SCALE/TRITON的单流双区熔盐堆燃耗计算方法

美国橡树岭国家实验室开发的SCALE/TRITON程序广泛用于反应堆临界安全、中子物理、辐射屏蔽和灵敏度与不确定度等方面的计算分析。基于SCALE/TRITON程序,采用等效体积、均匀混合和平均截面等三种外部耦合方法,处理单流双区熔盐堆的燃耗计算,解决了SCALE/TRITON程序在计算中不能精确反映流动燃料周期性均匀混合的问题。研究表明平均截面法与均匀混合法的计算结果几乎完全一致,与橡树岭文献结果也能很好符合,等效体积法因为没有考虑堆芯分区结构的差异而导致计算结果与其他两种方法偏离较大。基于SCALE/TRITON发展的平均截面法,放宽了对步长的要求,具有准确性好、计算效率高的优点,适用于熔盐堆两区(或多区)的堆芯设计与燃耗性能分析,具有重要的应用意义。     

Feasibility of a critical molten salt reactor for waste transmutation

This study assesses the feasibility of designing a Molten Salt Reactor (MSR) using the salt mixture of LiF (15 mol%), NaF (58 mol%) and BeF₂ (27 mol%) to be critical when fuelled with TRU from LWR spent fuel without exceeding the actinides solubility limit and while extracting fission products at realistic rates. The first part of the study investigated the graphite-to-MS volume ratio on the neutron balance, transmutation characteristics and graphite lifetime. It is found that a core without graphite moderator is the preferred design option; it offers the best neutron balance, most compact design and alleviated graphite lifetime problem. The second part of the study investigated sensitivity of the epithermal spectrum core to the feed composition, power density, fission products residence time and actinides loss fraction. It is found that the transmutation effectiveness improves with increasing power density and that the shorter the LWR spent fuel cooling time is, the better becomes the MSR neutron balance. The optimal MSR design offers a remarkably high transmutation capability – fissioning of as high as 99.8% of the TRU fed. The transmutation capability of the MSR is also rated in terms of final waste radiotoxicity, decay heat, spontaneous fission neutrons emission, fissile and ²³⁷Np inventory.     

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.     

熔盐堆物理热工耦合程序开发及验证分析

熔盐堆(Molten Salt Reactor,MSR)是第四代反应堆6种堆型中唯一的液态燃料反应堆,与固态燃料-液体冷却剂反应堆相比,原理上有较大不同。在熔盐堆中,流动的熔盐既是燃料又是冷却剂与慢化剂,中子物理学与热工水力学相互耦合;由于熔盐的流动性,缓发中子先驱核会随燃料流至堆芯外衰变,造成缓发中子的丢失,导致堆芯反应性降低。正是由于熔盐堆的这些新特性,造成熔盐堆内缓发中子先驱核、温度等参数变化与固态燃料反应堆有所不同,需要研究熔盐堆在各种工况下的相关物理参数变化。本文主要工作是考虑缓发中子先驱核的流动性对熔盐堆的影响,研究适用于熔盐堆的二维圆柱几何时空中子动力学程序及与之耦合的热工水力学程序;利用该程序对熔盐堆中子物理学和热工水力学进行耦合计算,验证熔盐堆相关实验数据;并且计算了熔盐堆无保护启停泵及堆芯入口温度过冷过热工况,用于分析熔盐堆的安全特性。计算结果表明,程序能够对熔盐反应堆实验(Molten Salt Reactor Experiment,MSRE)的相关实验数据进行较好的模拟计算,并且验证了熔盐堆的固有安全性。     

基于空气冷却的熔盐堆非能动余热排出系统优化设计

上海应用物理研究所基于TRISO包覆球形颗粒燃料与液态氟盐提出了基于钍基熔盐固态试验堆（TMSR-SF1）技术方案,其中一个重要的工作是非能动余热排出系统（PRHRS）设计。由于熔盐与水的不兼容特性,以及其高运行温度,采用空气作为最终热阱来设计PRHRS成为必然。为实现系统最简化、体积最小化以及排热与保温兼顾的设计目标,本文从MSR堆芯活性区到外界空气热阱传热过程的模型入手,建立了PRHRS优化设计模型,获得了优化设计方案,并基于改进的RELAP5/MOD4.0程序（针对TMSR-SF1的专门改进程序）开展了PRHRS容量论证评价,经计算分析,PRHRS容量设计合理,可确保反应堆全厂断电（SBO）后排热安全。      

SHTV,as a technique for core calculation using spatial homogenization and temperature variation

The accuracy of static neutronic parameters in the nuclear reactors depends upon the determination of group constants of the diffusion equation in the desired geometry. Although several methods have been proposed for calculating these parameters, there is still the need for more reliable methods. In this paper a powerful and innovative method based on Spatial Homogenization and Temperature Variation (SHTV) of physical properties of a WWER-1000 nuclear reactor core for calculating the relative power distribution of Fuel Assemblies (FA) and the hot fuel rod, is presented. The method is based on replacing the heterogeneous lattices of materials with different properties by an equivalent homogeneous mixture of these material for determining the few group constants, while the effect of temperature variation in the fuel and coolant density along the axial core direction is considered. All calculations are performed using WIMS and CITATION codes. The obtained results are compared with the results of Final Safety Analysis Report (FSAR) prepared by the designer, and good agreement between the two results is shown.     

Evaluation of implementation of thorium fuel cycle with LWR and MSR

In order to construct a sustainable society, it is necessary to consider fairness beyond generations and between countries. It is expected that Asian countries continue growing their economy and will result consuming more energy. More CO₂ emission is not acceptable.Nuclear power has many advantages for reducing CO₂ emission. However, it still has concerns of nuclear proliferation, radioactive waste and safety. It is necessary to overcome these concerns if nuclear power is expanded to Asian countries. Thorium utilization as nuclear fuel will be an opening key of these difficulties because thorium produces less plutonium, less radioactive waste. Safety will also be enhanced. The use of molten-salt reactor (MSR) triggered by plutonium supply from ordinary light water reactor (LWR) with uranium fuel will allow implementation of thorium fuel cycle with electricity capacity of about 446 GWe around at 2050.The other important sector in a view of sustainability is transportation. Transportation is essential for economy growth. Therefore it is inevitable to reduce CO₂ emission from transportation sector. Electric vehicle (EV) will be used as a major mobility instead of gasoline engine cars. Rare-earth materials such as neodymium and dysprosium are necessary for producing EV. These materials are expected to be mined from Asian countries. It is often obtained with thorium as by-product. Thorium has not been used as nuclear fuel because it is not good for nuclear weapon and it does not have fissionable isotopes. Recent global trend of nuclear disarmament and accumulation of plutonium from uranium fuel cycle can support starting the use of thorium.Thorium utilization will help both to provide clean energy and to produce rare-earth for clean vehicle. These will create new industries in developing Asian countries. An international collaborative framework can be established by supplying resource from developing countries and supplying technology from developed countries. “THE Bank (THorium Energy Bank)” is proposed here as one part of such a framework.     

Speed-up of cross-section collapsing with weight-window for subcritical burnup calculation

In the author’s group, a fusion–fission (FF) hybrid energy system has been analyzed using our own burnup calculation system consisting of Monte Carlo transport code MCNP-4C and point burnup code ORIGEN2.1. Since the neutron energy spectrum changes along with progress of burnup in a subcritical system, it is necessary to update one-group cross-section library in each burnup step. The one-group cross-sections are normally updated by collapsing the evaluated nuclear data such as JENDL and ENDF using a neutron flux calculated by an appropriate transport code such as MCNP. The collapsed cross-sections are handed over to ORIGEN, and the reaction rates for burnup of elements are thereafter estimated accurately.As well known, MCNP generates track-length (TL) data in the neutron transport calculation, which are base data to estimate the neutron flux. We thus use the track-length data directly instead of the calculated neutron flux, in order to evaluate the reaction rate as accurately as possible. However, the number of TLs becomes extremely large and thus it takes a longer computation time. We therefore reduce the number of TLs used in the cross-section collapsing process as far as the accuracy is conserved. However, in some energy region the number of TLs is inversely too small to conserve the original cross-section accuracy of the evaluated nuclear data files, because the number of TL data per unit energy is smaller than that of the nuclear data.In the present study, the weight-window (WW) technique of MCNP was applied to our burnup calculation system in order to control the number of TLs in such an energy region artificially and to complete the collapsing process accurately in the whole energy region. As a result, the variance of the calculated neutron flux thus deteriorates slightly, but the number of TLs could be successfully adjusted to conserve the accuracy of the nuclear data file in the whole energy region. And the accurate reaction rate estimation for burnup with MCNP was finally realized and simultaneously the computation time could be saved reasonably.     

Performance of Woodcock delta-tracking in lattice physics applications using the Serpent Monte Carlo reactor physics burnup calculation code

This paper presents the delta-tracking based geometry routine used in the Serpent Monte Carlo reactor physics burnup calculation code. The method is considered a fast and efficient alternative to the conventional surface-to-surface ray-tracing, and well suited to the lattice physics applications for which the code is mainly intended. The advantages and limitations of the routine are discussed and the applicability put to test in four example cases. It is concluded that the method performs well in LWR lattice applications, but really shows its efficiency when modeling HTGR particle fuels.     

燃耗信任制临界计算中保守性因素研究

在运用燃耗信任制技术进行乏燃料储存、运输等环节的临界安全分析时,临界计算所采用的条件是否具有足够的包络性十分关键。本文借助于OECD/NEA发布的若干燃耗信任制临界安全基准题,使用SCALE5.1软件中的STARBUCS模块进行分析,对信任核素选取、乏燃料冷却时间以及端末效应等因素对乏燃料系统临界安全性的影响进行了研究,得出了各参数保守性的有关结论。