Estimation of maximum coated particle fuel compact packing fraction

This work develops an analytic fuel fraction packing model for a high temperature gas cooled reactor fuel compact fabricated from overcoated particles of a single size. The model includes the effects of one dimensional compression and finite matrix grain size. One dimensional compression limits the maximum fuel packing fraction to about 48% for the pressed compact in this single sized particle system. This limit is due to two effects. The first is that the process of die loading limits the pre-compression packing configuration to one that is stable under gravity, which is not the most space efficient one. The second effect is due to the one dimensional compression which reduces only the axial dimension of the particle lattice rather than uniformly compressing the lattice. The die wall can also limit the maximum packing fraction by preventing the nearby particles from moving into a more space efficient configuration.     

Improvements in Quality of As-Manufactured Fuels for High-Temperature Gas-Cooled Reactors

The mechanisms of coating failure of the fuel particles for the high-temperature gas-cooled reactors during coating and compaction processes of the fuel fabrication were studied to determine a way to reduce the defective particle fraction of the as-manufactured fuels. Through the observation of the defective particles, it was found that the coating failure during the coating process was mainly caused by the strong mechanical shocks to the particles given by violent particle fluidization in the coater and by unloading and loading of the particles. The coating failure during the compaction process was probably related to the direct contact with neighboring particles in the fuel compacts. The coating process was improved by optimizing the mode of the particle fluidization and by developing the process without unloading and loading of the particles at intermediate coating process. The compaction process was improved by optimizing the combination of the pressing temperature and the pressing speed of the overcoated particles. Through these modifications of the fabrication process, the quality of the as-manufactured fuel compacts was improved outstandingly.     

气冷微堆燃料设计的中子学特性影响研究

为分析气冷微堆燃料设计的中子学特性影响,基于方形燃料组件模型,利用蒙特卡罗程序RMC研究了TRISO颗粒、燃料芯块在燃料设计中的主要参量对组件中子学特性的影响。研究结果表明,燃料颗粒体积占比和包覆层厚度不变时,组件寿期随燃料核芯直径的增大先显著增大,而后趋于平稳;燃料颗粒体积占比和燃料核芯直径不变时,组件寿期随包覆层厚度的增大而减小;燃料装载量不变时,芯块直径增大,组件寿期显著增大,而芯块高度影响较小;无燃料区厚度的增加对组件中子学特性有明显的负面影响,基体材料密度、基体杂质硼当量对组件中子学特性的影响较小。研究结果将为后续气冷微堆包覆颗粒弥散燃料的设计优化提供指导。     

颗粒尺寸分析仪法测量包覆颗粒疏松层密度的研究

包覆燃料颗粒的质量对于高温气冷堆安全运行起着重要作用。低密度热解炭层作为包覆的第一层非常关键,关系到包覆燃料颗粒和燃料元件的性能质量。本文介绍一种用颗粒尺寸分析仪测量疏松热解炭层密度的方法,该方法采用颗粒尺寸分析仪测量包覆前后颗粒的直径,再结合天平称得包覆前后颗粒的质量,经过计算得到包覆燃料颗粒疏松热解炭层的密度。对该方法测量包覆燃料颗粒疏松层密度的测量精度进行了验证。结果表明,该方法的测量精度满足测试要求,且该方法快速、便捷,适于工程应用。     

改进弦长抽样方法开发及在弥散燃料蒙特卡罗模拟的应用

弥散型燃料广泛应用于高温气冷堆、事故容忍燃料、实验研究堆及核动力舰船等，是重要的燃料类型之一。弦长抽样(CLS)方法可简化弥散燃料几何建模，提高计算效率，然而传统CLS方法只能描述单种颗粒的填充，同时在高体积填充率时误差较大。针对CLS方法的两大问题，本文在自主化堆用蒙特卡罗程序RMC中开发了改进CLS方法，并应用于全陶瓷微胶囊封装燃料棒算例及含毒物颗粒的高温堆燃料球算例。计算结果表明，改进CLS方法可解决多种颗粒混合填充的问题，并且可保证体积填充率的准确性，为弥散燃料的临界及燃耗计算提供了高效、精确的方法。     

Development of Improved Chord-length Sampling Method and Its Application in Monte Carlo Simulation of Dispersion Fuel

Dispersion fuel is widely used in high-temperature gas-cooled reactor (HTGR), accident tolerant fuel, experimental research reactor, naval nuclear power plant and so on. The chord-length sampling (CLS) method can simplify the geometry modeling of dispersion fuel, which can improve the efficiency. However, traditional CLS can only handle the packing of single particle, and has large error when the packing fraction is high. Aiming to solve these two problems, the improve CLS method was developed in reactor Monte Carlo code RMC, and applied to the fully ceramic micro-encapsulated fuel pin case and HTGR fuel pebble with mixed fuel and poison particles. Results show that the proposed method can handle mixed particles with multiple types, and preserve the accuracy of packing fraction, which provide precise and high efficiency for the critical and burnup calculations.     

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.     

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。     

Considerations of Irradiation-Induced Transient Creep in Fuel Particle Modeling*

An analytical assessment is made of the potential effects of irradiation-induced transient creep on the behavior of the TRISO-coated fuel particles of the New Production Modular High Temperature Gas-Cooled Reactor (NP-MHTGR). An analytical solution is presented for the three-layer particles, which includes transient creep in addition to steady-state creep behavior. The solution allows for evaluating the effects that transient creep has on individual particle stresses and for determining failure probabilities for particle batches using the Monte Carlo approach. Because experimental data needed to determine parameters for a transient component in a creep model for the pyrocarbons is not available, a range of possible parameter values were considered in the assessments. It was shown that transient creep measurably affects particle stresses early in the irradiation life of the particle. At that time, the hoop stress in the primary load bearing layer of the particle is in compression and the article is not vulnerable to pressure vessel failure. Later in irradiation, the effects of transient creep were typically shown to be less significant. Thus, transient creep had less than an order of magnitude effect on batch failure probabilities for prototypical NP-MHTGR fuel particles and was much less significant than steady-state creep. Whether the presence of transient creep increased or decreased the particle failure probability was dependent on the specific values used for the transient creep material properties.     

小型模块化超级安全气冷堆中子学特性研究

为分析小型模块化超级安全气冷堆堆芯中子学特性，建立六棱柱燃料组件模型，利用蒙特卡罗程序和ORIGEN程序的耦合计算，研究TRISO颗粒致密度、燃料富集度、TRISO颗粒大小、栅距比、TRISO颗粒包层厚度和燃料棒直径等物理参数对寿期等特性的影响。研究结果表明，寿期长度随着燃料富集度、栅距比的增大而单调增大；燃料棒直径、TRISO颗粒致密度、TRISO颗粒尺寸大小对寿期长度也有一定的影响；TRISO颗粒包层厚度对寿期长度的影响很小。基于该结果，初步设计出小型模块化超级安全气冷堆的堆芯装载方案，其寿期满足20 a不换料的寿期长度要求。      

包覆燃料颗粒应力的有限元分析

高温气冷堆的燃料元件由包覆燃料颗粒弥散在石墨基体中组成。在反应堆运行过程中,辐照及各复杂的物理化学反应产生的应力会使包覆燃料颗粒发生破损,对包覆燃料颗粒进行应力分析是评价燃料元件和反应堆运行安全性能的主要内容之一。本文基于压力壳模式,主要考虑内压作用下的球形壳层应力及包覆燃料颗粒的非球形因素,用有限元法对应力进行了分析。     

高燃耗陶瓷燃料颗粒的参数化建模及强度分析

弥散燃料芯体中的陶瓷燃料颗粒在辐照条件下会形成裂变气孔,燃料颗粒内部气孔间的相互干涉作用及气孔内压的增长致使局部拉应力超过材料强度极限,进而导致燃料颗粒开裂。本文考虑高燃耗燃料颗粒内气孔尺寸和位置分布的非均匀性,实现了颗粒内部的细观结构参数化建模。运用有限元方法计算并分析了气孔尺寸、基体约束压应力、温度和气孔分布方式对颗粒内部最大拉应力的影响,研究了颗粒内开裂危险区的分布规律。结果表明,陶瓷燃料颗粒最大拉应力随气孔尺寸和温度的增加而增大,随基体约束压应力的增加而减小;燃料相的断裂强度减小,开裂危险区面积增大;燃料颗粒从内部多处开裂破坏,而表层处开裂的概率更大。本文为弥散燃料失效研究及优化设计提供了分析方法及数值参考。     

Current status and future development of coated fuel particles for high temperature gas-cooled reactors

The coated particles were first invented by Roy Huddle in Harwell 1957. Through five decades of development, the German UO₂ coated particle and US LEU UCO coated particle represent the highly successful coated particle designs up to now. In this paper, current status as well as the failure mechanisms of coated particle so far is reviewed and discussed. The challenges associated with high temperatures for coated particles applied in future VHTR are evaluated. And future development prospects of advanced coated particle suited for higher temperatures are presented. According to the past coated fuel particle development experience, it is unwise to make multiple simultaneous changes in the coated particle design. Two advanced designs which are modifications of standard German UO₂ coated particle (UO₂^∗ herein) and US UCO coated particle (TRIZO) are promising and feasible under the world-wide cooperations and efforts.     

UN核芯TRISO燃料颗粒破损概率模型研究

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。TRISO燃料颗粒破损概率是表征TRISO燃料事故安全特性的关键参数。本文基于修正的PANAMA破损概率计算方法,在考虑UN核芯裂变气体释放导致的气体内压以及内外致密热解炭层辐照蠕变和收缩作用的基础上,开发了UN核芯TRISO燃料颗粒压力壳式破损概率计算方法,并采用IAEA基准题6和基准题9对模型进行了验证;基于开发的UN核芯TRISO颗粒破损概率计算方法,采用随机抽样统计方法分析了事故工况下UN核芯和包覆层设计参数（包括包覆层尺寸及密度）对UN核芯TRISO燃料颗粒破损概率的影响。研究结果显示,疏松热解炭（Buffer）层设计参数是影响TRISO颗粒破损概率的关键因素,可通过降低Buffer层尺寸及密度分布设计标准偏差的方法降低UN核芯TRISO燃料颗粒的破损概率。     

New Sampling Method in Continuous Energy Monte Carlo Calculation for Pebble Bed Reactors

A pebble bed reactor generally has double heterogeneity consisting of two kinds of spherical fuel element. In the core, there exist many fuel balls piled up randomly in a high packing fraction. And each fuel ball contains a lot of small fuel particles which are also distributed randomly. In this study, to realize precise neutron transport calculation of such reactors with the continuous energy Monte Carlo method, a new sampling method has been developed. The new method has been implemented in the general purpose Monte Carlo code MCNP to develop a modified version MCNP-BALL. This method was validated by calculating inventory of spherical fuel elements arranged successively by sampling during transport calculation and also by performing criticality calculations in ordered packing models. From the results, it was confirmed that the inventory of spherical fuel elements could be reproduced using MCNP-BALL within a sufficient accuracy of 0.2%. And the comparison of criticality calculations in ordered packing models between MCNP-BALL and the reference method shows excellent agreement in neutron spectrum as well as multiplication factor.

MCNP-BALL enables us to analyze pebble bed type cores such as PROTEUS precisely with the continuous energy Monte Carlo method.     

Behavior of Coated Fuel Particle of High-Temperature Gas-Cooled Reactor under Reactivity-Initiated Accident Conditions

In order to clarify the failure mechanism and determine the failure limit of the High-Temperature Gascooled Reactor (HTGR) fuel under reactivity-initiated accident (RIA) conditions, pulse irradiations were performed with unirradiated coated fuel particles at the Nuclear Safety Research Reactor (NSRR). The energy deposition ranged from 0.578 to 1.869 kJ/gUO₂, in the pulse irradiations and the estimated peak temperature at the center of the fuel particle ranged from 1,510 to 3,950 K. Detailed examinations after the pulse irradiations showed that the coated fuel particles failed above 1.40 kJ/gUO₂, where the peak fuel temperature reached over the melting point of UO₂ fuel. It was concluded that the coated fuel particle was failed by the mechanical interaction between the melted and swelled fuel kernel and the coating layer under RIA conditions.     

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.     

长寿期压水堆颗粒弥散可燃毒物中子学设计与分析

为实现长寿期压水堆的低硼运行,对颗粒弥散可燃毒物进行了中子学设计与分析,颗粒弥散可燃毒物的自屏效应可通过颗粒半径进行调节,能实现可燃毒物消耗和燃料燃耗的较优匹配。本文选取目前压水堆常用的快燃耗可燃毒物B、Gd为对象,研究了颗粒弥散可燃毒物不同颗粒半径和填充份额对组件中子学特性的影响。结果表明,颗粒弥散可燃毒物能实现长期稳定的反应性控制,其中BISO含硼弥散颗粒符合长寿期压水堆低硼运行的要求,适合作为长寿期压水堆的候选可燃毒物进行下一步研究。     

Low-temperature synthesis of silicon carbide inert matrix fuel through a polymer precursor route

A low temperature process of mixing different sizes of silicon carbide (SiC) particles with a polymer precursor was utilized to synthesize SiC pellets for potential use as inert matrix fuels (IMF) for light water reactors. The lower temperature process is required to prevent the reactions between SiC and the dispersed PuO₂ fuel material. The effect of the polymer content and the cold pressing pressure on the packing of SiC particles was investigated. The effect of mixing coarse and fine SiC particles on the density and the pore size distribution was also investigated. It was found that the density and pore size distribution can be tailored by controlling the SiC size compositions, polymer content and pressing pressure at room temperature. A possible mechanism has been proposed to explain the forming of the pores with respect to the geometric arrangement between SiC particles and the polymer precursor. SEM images showed that ceria (cerium oxide) which is a PuO₂ surrogate in this study, was well distributed in the pellet.