球形燃料元件温度分布对包覆燃料颗粒失效概率的影响

固态熔盐堆采用TRISO(Tristructural isotropic)包覆颗粒球形燃料元件。在运行工况下,燃料元件内部存在一定的温度分布,填充在燃料元件内部不同位置的TRISO颗粒的失效概率会因此受到影响。利用体积微元的方法分析了温度分布对包覆颗粒失效概率的影响,并进一步研究了球形燃料元件尺寸对TRISO颗粒平均失效概率的影响。结果表明,在一定的功率密度下,如果利用球心温度或者平均温度计算燃料元件内部TRISO颗粒的平均失效概率,结果相比实际值会有至少一个数量级的差别;在相同功率密度和相同燃耗条件下,燃料元件直径每减小1 cm,其包覆颗粒平均失效概率降低两个数量级左右。     

辐照后HTR-10球形燃料元件电化学解体及裂变产物分布研究

辐照后检验是开展燃料性能评价的重要手段。在10 MW高温气冷堆(HTR-10)球形燃料元件的辐照后检验中,为研究元件中TRISO包覆燃料颗粒的破损机制,本文利用基于电化学氧化原理的两步解体法,对所选元件(燃耗约35 GW·d/tU)进行了包覆燃料颗粒与基体石墨的分离,获得了元件中不同位置区域的包覆燃料颗粒、解体石墨粉和电解液,通过γ能谱定量分析了解体石墨粉和电解液中的放射性核素成分及含量,并基于此明确了放射性核素在辐照后球形燃料元件基体石墨中的分布。结果表明：部分电解液中裂变产物¹³⁷Cs与¹⁴⁴Ce活度显著高于其他电解液样品,表明其对应区域中可能存在破损包覆燃料颗粒;元件表层位置对应的电解液中活化产物⁶⁰Co高于内部区域,主要来源于HTR-10一回路的放射性粉尘沾污。本工作初步建立了高温气冷堆辐照后球形燃料元件电化学解体和解体样品分析测试的平台及方法,为TRISO型包覆燃料颗粒破损机制分析及其堆内行为研究提供了重要基础。     

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。      

基于有限元方法的惰性基弥散燃料传热性能研究

惰性基弥散燃料芯块(Inert Matrix Dispersion Pellet,IMDP)以高温气冷堆燃料技术为基础,采用惰性材料作为三重各向同性型(Tristructural Isotropic,TRISO)燃料颗粒的弥散基体,相比传统的UO2燃料,其最典型特征是具备高热导率。采用通用有限元软件ABAQUS,结合其二次开发功能,建立有限元计算模型,研究了温度、燃耗以及燃料颗粒与惰性基体间热阻对IMDP燃料有效热导率影响规律,并与UO2陶瓷燃料进行对比。结果表明:IMDP有效热导率随燃耗及温度的增加而减小,且在不同寿期及不同温度下,IMDP有效热导率均明显高于UO2热导率;反应堆正常运行工况下,相比UO2芯块,IMDP较高的热导率会使芯块中心温度显著降低;此外,燃料颗粒与惰性基体间热阻在0~4×10~(-4) m~2·oC·W~(-1)范围内对IMDP的有效热导率影响程度最为敏感。     

高体积份额下包覆颗粒弥散燃料等效热学模型

准确预测核燃料的内部温度场分布，对于多层包覆颗粒弥散核燃料元件的设计及筛选具有重要的指导意义。在多层包覆颗粒及其弥散块体的等效热导率模型基础上，本文针对高体积份额情况分析建立了等效传热计算方法及其数值模型，并研究了燃料颗粒的空间分布、颗粒尺寸、团聚等因素对核燃料元件温度场的影响。本研究有助于理解多层包覆颗粒弥散核燃料元件的微观-宏观传热机制，为核燃料的设计、优化和安全分析提供了分析工具。      

锆基弥散微封装燃料等效传热系数数值模拟研究

为研究锆基弥散微封装燃料(M3燃料)的等效传热系数,假定TRISO(三层各向同性)颗粒球在锆基体内呈现体心立方排布,通过ABAQUS软件,基于均匀化理论,建立了M3燃料等效传热系数的模拟方法。依据建立的模拟方法,对不同相体积M3燃料的等效传热系数进行了研究分析。模拟结果显示:等效传热系数会随着温度的升高而升高、随燃耗和相体积的增加而降低。     

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。      

基于热-流-固耦合的燃料元件性能分析方法研究

中空棱柱形燃料元件形式和运行工况特殊,没有现成的燃料性能分析软件能够满足计算要求,需要建立新的分析方法。本研究基于COMSOL软件二次开发,采用颗粒增强复合材料的等效物性模型和共轭传热技术实现中空六棱柱形燃料的三维热-流-固耦合计算,通过与美国通用电气公司数据的对比证明了该分析方法的有效性。采用该方法计算了多种燃料元件尺寸和不同轴向功率分布下的热应力和温度,结果表明侧棱处温度最高而内壁面壁厚最薄处热应力最大,壁厚越薄、长度越长,燃料元件的最大热应力和温度越小,展平入口段的轴向功率分布也能够略微降低最大热应力和温度。以上分析方法可以用于新型中空棱柱形燃料元件的优化设计。      

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。     

TRISO燃料颗粒三维多物理场耦合计算模型开发

三向同性燃料(TRISO)颗粒是高温气冷堆弥散型燃料和全陶瓷微密封(FCM)耐事故燃料芯块的裂变区。为研究TRISO燃料颗粒在辐照环境中的复杂行为,基于COMSOL有限元软件开发了TRISO燃料颗粒的三维多物理场耦合性能分析模型。通过采用随辐照条件变化的材料物性参数和行为模型,可模拟燃料颗粒在稳态运行和事故工况下复杂的堆内热-力学行为,以及CO气体产生和裂变气体释放、裂变产物扩散等重要物理过程,还可以计算燃料颗粒的失效概率。基于COMSOL开发三维分析模型的计算结果与美国BISON程序对TRISO燃料颗粒的计算结果相比同样符合较好,说明了所开发模型的合理性。     

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.     

燃料热导率降级对CAP1000大破口失水事故的影响分析

在高燃耗情况下,燃料芯块的热导率随燃耗降低,该现象被称之为热导率降级(TCD)现象。TCD现象影响失水事故(LOCA)前稳态工况的燃料平均温度和燃料储能,进而影响大破口LOCA过程中的包壳峰值温度(PCT)。本研究采用大破口LOCA分析程序WCOBRA/TRAC对CAP1000冷段双端剪切断裂事故进行了不同燃耗的敏感性分析,并获得了不同工况下的PCT。分析中采用美国核燃料研究所(NFI)修正的TCD模型对降级后的燃料热导率进行模拟,同时考虑了燃耗大于30GW·d/tU后FQ和FΔh峰值因子的降低。敏感性分析表明,考虑TCD和峰值因子降低的影响,PCT极限工况不再出现在低燃耗区间,而出现在燃耗为29GW·d/tU附近。与其他燃耗水平相比,该燃耗点的PCT第1峰值和第2峰值均处于最高水平。本研究结果可为高燃耗情况下非能动电厂大破口LOCA的分析评估提供参考。     

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。     

Fission product monitoring of TRISO coated fuel for the advanced gas reactor-1 experiment

The US Department of Energy has embarked on a series of tests of TRISO coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burnup of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B's) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.     

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。     

Two-temperature homogenized model for steady-state and transient thermal analyses of a pebble with distributed fuel particles

In a pebble-bed type very high temperature gas-cooled reactor (VHTGR), a typical fuel pebble consists of over ten thousand five-layer TRISO particles in a graphite-matrix. The high heterogeneity in composition leads to difficulty in explicit thermal calculation of pebble fuels. Thus, a homogenization model becomes essential. Currently, a simple volumetric-average thermal conductivity approach is used. However, this approach is non-conservative and underestimates the fuel temperature.     

Model development for the estimation of fission product release under normal and accident conditions in a HTGR

This article has attempted to estimate the radioactivity release from fuel materials during normal and transient conditions by coupling the TRISO fracture and the fission product (FP) diffusion. Two calculation models, named TRISO Fracture Analyzer (TRIFA) and DIFfusion Analyzer (DIFA), are developed. TRIFA is initially used to calculate the fraction of fractured fuel particles, thus determining the amount of fission gas release. The obtained particle fracture function is then used as input for the diffusion rate calculation. DIFA simulates with a single spherical fuel element, a pebble, irradiated under normal and accident conditions. It describes the diffusive transport of fission products by numerically solving the diffusion equation. The finite difference method is applied to obtain fission product release rates from a pebble to coolant. The model comparisons show that the new developed models are reliable, fast, and correspond with previous results of other models. As for HTR-10, the coupled models, TRIFA and DIFA, are applied to calculate the level of fission product release after accidents. The following conclusions can be drawn. First, the mitigation should be carried out until the maximum fuel temperature reaches under transient. Second, the mitigation should be intensively considered if the burn-up exceeds 5%FIMA (∼48 GWd/MTU) when transient happens. Additionally, it is found that there is the threshold burn-up where the rapid FP release occurs due to the numerous TRISOs fractured. Further investigations are needed to extend the use of the method developed in this work to the safety assessments for high-temperature gas-cooled reactors (HTGRs). This article will hopefully serve as a platform for designing the advanced TRISO that can minimize the activity release, and providing the rationale of development of the intensive accident mitigation system in future.