全陶瓷微密封燃料有效多群截面计算方法研究

全陶瓷微密封(FCM)燃料是一种弥散颗粒燃料。由于弥散颗粒燃料存在双重非均匀性,传统的确定论方法及蒙特卡罗方法皆难以处理这种双重非均匀效应以获得有效多群截面。本文基于超细群方法建立FCM燃料的有效多群截面计算方法。为描述燃料棒内TRISO颗粒的非均匀性,在共振能量段,通过采用超细群方法求解包含TRISO颗粒的一维球模型得到超细群缺陷因子,通过超细群缺陷因子修正所有核素的超细群截面即可将颗粒和基质均匀化。由于TRISO颗粒在热能区也存在较强的自屏效应,在热能区,利用穿透概率及碰撞概率等价得到多群缺陷因子,通过多群缺陷因子修正所有核素的多群截面将燃料和基质均匀化。均匀化后的FCM燃料组件即可视为普通压水堆燃料组件进行共振计算。利用丹可夫修正因子等价得到FCM燃料组件各燃料棒的等效一维棒模型,对一维棒模型求解超细群慢化方程从而得到共振能量段的有效自屏截面。数值结果表明,该方法能有效处理FCM燃料的双重非均匀性,得到精确的有效自屏截面。     

基于改进缺陷因子方法的FCM燃料有效多群截面计算方法研究

全陶瓷微胶囊封装（FCM）燃料是重要的候选事故容错燃料,与传统燃料相比,FCM燃料的双重非均匀性使得其有效多群截面计算面临较大的挑战。本文提出一种改进的缺陷因子方法来处理FCM燃料在共振能区和非共振能区的自屏效应,实现FCM燃料的等效均匀化。通过颗粒丹可夫因子守恒来构建新的等效模型以克服传统的体积权重等效模型无法考虑燃料棒间自屏的影响;在共振能量段,基于新的等效一维球模型求解超细群慢化方程获得共振能量段的超细群缺陷因子;在非共振能量段,利用新等效模型的特征值计算获得快群和热群的多群缺陷因子;在此基础上实现FCM燃料棒的等效均匀化。本方法已在高保真中子学程序NECP-X上实现,并在一系列工况下进行了测试,与蒙特卡罗程序的比较表明,本方法能处理不同情况下的双重非均匀性,并可获得准确的有效自屏截面。     

双重非均匀子群参数制作研究

弥散颗粒燃料元件中燃料颗粒以随机形式弥散在基体中,难以获得确定几何。同时由于共振自屏现象的存在,呈现出一种双重非均匀系统。当前均匀系统产生的共振积分在双重非均匀系统中使用时,会在较低的共振能群产生一定的共振计算误差。为满足现有组件计算程序直接进行双重非均匀性共振计算的需求。基于Sanchez-Pomraning模型下的特征线固定源计算方法,建立一套双重非均匀共振积分表,最后结合子群方法实现随机介质燃料元件的共振计算。数值结果表明,考虑双重非均匀性产生的积分表,在相同的输运条件下和积分表的适用范围内,由子群共振部分对keff计算带来的绝对偏差能保持在200 pcm内。该工作的意义是对于一些不宜改动的传统组件程序,如HELIOS,通过在线修改共振积分表和子群参数,从而使其直接进行弥散颗粒燃料问题的计算成为可能。     

核燃料溶解器临界计算对截面的要求

用群截面对燃料溶解过程中出现的栅格、燃料双重不均匀和溶液３种系统作临界计算时，需要考虑中子的共振自屏效应。标准自屏公式或经过丹可夫因子修正的自屏公式不适用于燃料双重不均匀系统。ＯＥＣＤ／ＮＥＡ临界工作小组的结果表明，必须用碰撞概率（ＰＩＣ）方法，子群方法或精细慢化方法修正才能得到共振自屏效应的准确结果。用点截面作临界计算时，不会观察到自屏效应，可以准确进行包括燃料双重不均匀系统在内的临界计算。     

基于子群方法的双重非均匀性共振计算方法研究

为计算双重非均匀性条件下的共振截面,提出了耦合Sanchez Pomraning方法的改进的子群方法(ISSP)。ISSP采用精细化共振能群结构来规避共振干涉处理,通过求解双重非均匀性条件下的子群固定源方程和慢化方程得到颗粒和基体等各材料的有效共振截面,最后进行双重非均匀性条件下的输运计算。数值结果表明,与连续能量蒙特卡罗程序及超细群计算结果相比,ISSP可精确高效地计算双重非均匀性条件下的共振截面。     

用蒙特卡罗方法计算压水堆燃料组件内的超热中子

超热中子计算在压水堆的物理计算中占有重要地位。本文是用蒙特卡罗方法计算压水堆燃料组件内的超热中子谱及其空间分布。在计算中,由于对燃料组件的非均匀布置和共振截面都没有作简化,因而可以得到准确度较高的计算结果。本方法考虑了燃料棒的自屏及互屏效应,可以精确地计算出丹可夫因子,避免了引进各种近似所带来的误差。     

二维任意几何子群共振计算及加速优化

共振计算是反应堆组件堆芯设计和燃料管理的基础.子群共振计算方法基于共振能群子群截面,调用输运程序作为求解器,对子群中子注量率进行求解并且归并得到有效共振自屏截面,实现任意二维复杂几何的共振计算.由于子群方法在每个共振能群内部需要反复调用输运求解器,因此和等价理论相比速度较慢及本文基于子群方法的理论模型和自主开发的子群共振计算程序,提出并且完成了多群数据库、输运计算源项及多共振核素迭代的优化方案.通过基准题的验证可知,该方案在保持精度的同时提高了子群程序的计算效率,保证了该程序在工程上的实用性.     

基于特征线法计算的超细群慢化方程求解方法

为实现对复杂几何、复杂能谱组件的精细计算,提出了一种基于特征线的超细群慢化方程求解方法。通过耦合特征线法中的固定源计算,在共振能量范围内建立超细群慢化方程,通过精细能谱获得复杂结构下的共振自屏截面。对典型压水堆栅元问题、带有温度分布的栅元问题、燃料内部存在不均匀性的栅元问题以及板状燃料组件问题进行了计算。结果表明,基于特征线的超细群慢化方程求解方法可精确计算复杂几何、复杂能谱问题,为共振计算提供基准。     

基于子群共振自屏方法的特征值隐式敏感性分析

采用经典微扰理论,高效地得到问题相关的多群截面的扰动对特征值的直接影响,即显式敏感性。应用广义微扰理论,推导了在子群共振自屏方法中,多群共振自屏截面对非共振核素截面的灵敏度系数的计算方法。结合前两项内容,得到非共振核素通过共振自屏过程对特征值的间接影响,即隐式敏感性。根据与显式灵敏度系数的比较,分析了隐式敏感性相对于显式敏感性的重要性。     

基于特征线法计算的超细群慢化方程求解方法

为实现对复杂几何、复杂能谱组件的精细计算,提出了一种基于特征线的超细群慢化方程求解方法。通过耦合特征线法中的固定源计算,在共振能量范围内建立超细群慢化方程,通过精细能谱获得复杂结构下的共振自屏截面。对典型压水堆栅元问题、带有温度分布的栅元问题、燃料内部存在不均匀性的栅元问题以及板状燃料组件问题进行了计算。结果表明,基于特征线的超细群慢化方程求解方法可精确计算复杂几何、复杂能谱问题,为共振计算提供基准。     

Study on double heterogeneity effect of Pu-rich agglomerates in mixed oxide fuel using cross section homogenization method for particle-dispersed media

A cross section homogenization method for media containing randomly and uniformly dispersed particles, which was originally developed by Shmakov et al., has been applied to MOX fuels containing Pu-rich agglomerates. This method (Shmakov’s method), which is incorporated into a continuous-energy Monte Carlo code MCNP, has been applied to lattice calculations of an infinite MOX fuel rod array. Shmakov’s method can accurately reproduce the criticality calculation results for an explicit heterogeneous arrangement of Pu-rich agglomerates. A correction factor that Shmakov’s method defines to obtain an effective microscopic cross section provides a proper quantitative indication of the double heterogeneity of MOX fuels containing Pu-rich agglomerates. The correction factors exhibit an obvious double heterogeneity effect of Pu-rich agglomerates dispersed in MOX fuel pellets. The effective microscopic cross sections of plutonium isotopes in MOX fuels containing Pu-rich agglomerates are significantly reduced due to the self-shielding effect as compared to the homogeneous MOX fuel model. However, the double heterogeneity effect of Pu-rich agglomerates on k_eff seems to be unexpectedly minor because the underestimate of the reaction rates in the resonance energy range is offset by the overestimate of the reaction rates in the thermal energy range.     

Utilization of TRISO Fuel with LWR Spent Fuel in Fusion-Fission Hybrid Reactor System

HTRs use a high performance particulate TRISO fuel with ceramic multi-layer coatings due to the high burn up capability and very neutronic performance. TRISO fuel because of capable of high burn up and very neutronic performance is conducted in a D-T fusion driven hybrid reactor. In this study, TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 68%. The neutronic effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on the fuel performance has been investigated for Flibe, Flinabe and Li₂₀Sn₈₀ coolants. The reactor operation time with the different first neutron wall loads is 24 months. Neutron transport calculations are evaluated by using XSDRNPM/SCALE 5 codes with 238 group cross section library. The effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on tritium breeding (TBR), energy multiplication (M), fissile fuel breeding, average burn up values are comparatively investigated. It is shown that the high burn up can be achieved with TRISO fuel in the hybrid reactor.     

TRISO燃料颗粒等效导热系数理论模型研究

TRISO (tri-structural isotropic) fuel particle consists of a fuel kernel in the center coated with four layers, with good fission product retention capability. The effective thermal conductivity of TRISO fuel particle is an important basis for calculating the effective thermal conductivity of dispersed fuels. In the present work, the theoretical model of the effective thermal conductivity of TRISO particle is built based on the theory of the effective thermal conductivity in multiphase solids in the framework of spherical coordinate and then the effective thermal conductivity of metal matrix microencapsulated fuel (M3) is analyzed combined with the Chiew-Glandt model which is the effective thermal conductivity model for solid-solid binary composite. The results show that the present model provides an excellent prediction of the thermal conductivity of TRISO particle. Finally the effective thermal conductivity of fully encapsulated fuel (FCM) is presented.     

TRISO燃料颗粒等效导热系数理论模型研究

三层各向同性碳包覆（TRISO）燃料颗粒由核芯和4层包覆层组成,具有良好的裂变产物包容能力,其等效导热系数是计算弥散微封装燃料等效导热系数的重要基础。本文首先从球坐标下基本导热方程出发,基于多相固体宏观等效导热理论,建立了TRISO燃料颗粒等效导热系数理论计算模型;然后,结合固-固二元复合材料等效导热系数Chiew-Glandt模型分析了锆基微封装燃料（M3）芯体等效导热系数。结果表明,本文开发的模型可有效模拟TRISO燃料等效导热系数。基于开发的TRISO等效导热系数模型计算获得了全陶瓷微封装燃料（FCM）的等效导热系数。      

UN核芯TRISO包覆燃料颗粒性能分析

为分析致密热解碳层、内压等因素对TRISO包覆燃料颗粒热-力学性能的影响,基于多物理场耦合软件COMSOL建立了以UN为核芯的TRISO包覆燃料颗粒三维热-力学耦合模型,并通过IAEA CRP-6基准题进行了验证。利用本文模型对稳态运行及反应性引入事故（RIA）工况下典型TRISO包覆燃料颗粒的性能进行了分析,结果表明,正常运行工况下SiC层能维持结构完整性,但IPyC层存在失效风险,需进一步优化TRISO包覆燃料颗粒的设计方案,而RIA工况下热膨胀是造成TRISO包覆燃料颗粒发生结构失效的主要原因。该模型能对轻水堆运行环境下的TRISO包覆燃料颗粒进行复杂的多物理场耦合性能分析,为进一步优化FCM燃料元件设计打下基础。     

New fuel rod design based on LEU/Th fuel and an assessment on in-core fuel cycle length of the MHR core

The current Modular Helium Reactor (MHR) fuel cycle uses fissile LEU (19.8 wt% U-235) and depleted uranium in separate TRISO particles, in a single fuel rod within a graphite matrix. The TRISO particle volume packing fraction (PF) in the fuel rods is 29%, of which the LEU particle PF is 62%. The lifetime between refuelings is about 476 effective full power days (EFPD). In this paper we assess the possibility of replacing the depleted uranium TRISO particles with thorium TRISO particles, and evaluate the impact of such replacement on fuel cycle length. A preliminary scoping study was performed to determine the most promising fuel rod/zoning configurations. The scoping study indicates that there is advantage to separating the thorium TRISO particles from the LEU particles at the fuel rod level instead of mixing them within a single rod. An axial checkerboard distribution of the fuel rods where all uranium and all thorium rods are interchangeable along the axial direction within the graphite block is the most promising configuration that was identified in this study and can be lead to a fuel cycle length extension of 50-80% relative to the current design, with only a modest increase in the fissile material loading (15-20%). To this advantage can be added the benefit of a significant reduction in nuclear waste and in health risk. This study also lays the foundation for improving the fuel rod arrangement within the graphite block and the graphite blocks within the entire reactor core. The analysis is limited to a once - through fuel cycle based on in situ fissioning of the U-233, without further separation and reprocessing. The preliminary heat transfer analysis indicates that the maximum temperature in the fuel will be raised by about 10-15% over that of current MHR design.     

Performance of TRISO particles fueled with plutonium and minor actinides in a PBMR-400 core design

The TRISO particle design of high temperature reactors fueled with plutonium (Pu) and/or minor actinides (MAs) is investigated by calculating the failure fraction of TRISO particles during irradiation. For this purpose, a fuel depletion, neutronics and thermal-hydraulics code system, which delivers the fuel temperature, fast neutron flux and power density profiles, is coupled to an analytical stress analysis code. The latter is being further developed for the calculation of a reliable and realistic failure fraction. The code system has been applied to a PBMR-400 design containing TRISO particles fueled with 1st and 2nd generation plutonium and with a target burn-up of 700 and 600 MWd/kgHM, respectively. It is shown that the pebble-bed type high temperature reactor under consideration is a promising option for burning Pu and MAs if very high burn-ups can be achieved. The TRISO particle failure fraction is also calculated for both Pu and MA fuels, and compared to U-based fuel. It is shown by the present stress analysis code that the Pu-based fuel particles need a better design and this has been achieved for the MA-based fuel, in which helium gas atoms have a significant contribution to the buffer pressure.     

气冷堆用棱柱型弥散微封装燃料性能分析及优化研究

棱柱型弥散微封装燃料是将三重各向同性包覆（TRISO）燃料颗粒弥散于金属或陶瓷基体形成的颗粒增强复合燃料,具有良好的结构稳定性、裂变产物包容能力和辐照稳定性,是高温气冷堆中较具发展前景的燃料形式之一。本文提出将TRISO燃料颗粒弥散于SiC基体的棱柱型弥散微封装燃料设计方案,并基于有限元分析软件COMSOL建立了该燃料元件三维热流固耦合分析模型,初步实现了该燃料元件性能分析和优化设计。结果表明,棱柱型弥散微封装燃料元件的温度最大值位于燃料元件外侧,应力峰值位于冷却剂通道壁面,边距比为0.76~0.84、孔距比为0.68~0.75时燃料元件热应力最小。本文建立的棱柱型弥散微封装燃料性能分析方法和研究结论,可为后续该型气冷堆燃料元件设计提供指导和参考。