高温气冷堆乏燃料分析实验室设计与设备研究

针对我国高温气冷堆乏燃料研究设施的空白,研究设计了专门用于高温气冷堆球形燃料元件辐照后性能研究的乏燃料分析实验室和专用工艺设备。基于球形燃料元件与包覆燃料颗粒的特殊结构,所设计的乏燃料分析实验室包括5间热室、6个手套箱和辅助设施,研究设计了专用的工艺实验设备,能够对辐照后的高温气冷堆燃料元件和包覆燃料颗粒进行宏观检查、燃耗测量、元件解体、模拟事故条件加热、辐照微球γ测量分析破损率,通过金相显微镜和扫描电镜进行微观结构分析,开展燃料元件的辐照失效机理研究。     

"乏燃料管理安全和放射性废物管理安全联合公约"述评

“乏燃料管理安全和放射性废物管理安全联合公约”是迄今为止有关放射性废物管理方面最重要的全球性公约,亦是继1994年“核安全公约”以来核安全国际法领域又一新的突破。本文简要介绍“乏燃料管理安全和放射性废物管理安全联合公约”的主要内容,并讨论了其存在的问题及对策。     

TRISO fuel volume fraction and homogeneity:a nondestructive characterization

A new nondestructive method to estimate the volume fraction and homogeneity of tristructural isotropic(TRISO)-coated fuel particles in fuel compacts designed for high-temperature reactors has been developed using image analysis of conventional X-radiographs. The method is demonstrated on surrogate fuel compacts containing TRISO-coated particles with kernels made of zirconium dioxide. The methodology incorporates a correction for superimposed images of TRISO particles such that a single X-ray image obtained in any one random orientation is sufficient to characterize the fuel compact in terms of volume fraction and homogeneity. The method is based on the virtual segregation of images of each particle inside the compact with the aid of a calibration standard.     

The Third International Conference on Nuclear Fuel Reprocessing and Waste Management (RECOD '91)

The purpose of this study is to assess the organizational types and the job stress factors that affect procedure-based job performances in nuclear power plants. We derived 24 organizational factors affecting job stress level in nuclear power plants from the job stress analysis models developed by NIOSH, JDI, and IOR. Considering the safety characteristics in the operating tasks of nuclear power plants, we identified the job contents and characteristics through the analyses of job assignments that appeared in the organizational chart and the results of an activity-based costing. By using questionnaire surveys and structured interviews with the plant personnel and expert panels, we assessed 70 jobs among the 777 jobs managed officially in accordance with the procedures. They consist of the representative jobs of each department and are directly related to safety. We utilized the organizational personality type indicators to characterize the personality types of each organization in nuclear power plants.     

UO2燃料燃耗对高放废物管理的影响研究

从乏燃料的不同燃耗引起放射性和化学组成的变化出发,分析乏燃料经后处理后的衰变热、Mo及贵金属含量对玻璃固化工艺和玻璃固化体储存的影响,计算得到了不同燃耗乏燃料制得的高放玻璃的数量。计算结果认为：对于冷却8 a的乏燃料,决定玻璃固化体包容量的不是高放主组分的热功率;对于燃耗小于40 GW•d/tU的乏燃料,决定玻璃固化体包容量的是Mo元素含量;当燃耗大于45 GW•d/tU时,贵金属含量成为决定玻璃固化体包容量的主要因素,同时UO₂燃料燃耗与高放玻璃固化体数量上存在线性关系,燃耗增加会导致高放废物玻璃固化体数量增加。随着燃耗的增加,以Mo含量及贵金属含量计算得到的玻璃固化体数量比以衰变热计算得到的玻璃固化体数量多,因此,高放废物玻璃固化前将Mo及贵金属进行分离有利于减少高放废物玻璃固化体数量。对于UO₂燃料,燃耗加深对于高放废物玻璃固化体暂存时间几乎无影响。     

THETRIS: A micro-scale temperature and gas release model for TRISO fuel

The dominating mechanism in the passive safety of gas-cooled, graphite-moderated, high-temperature reactors (HTRs) is the Doppler feedback effect. These reactor designs are fueled with submillimeter-sized kernels formed into tristructural-isotropic (TRISO) particles that are imbedded in a graphite matrix. The best spatial and temporal representation of the feedback effect is obtained from an accurate approximation of the fuel temperature. Micro-scale models of TRISO particles are necessary in order to obtain accurate predictions during fast transients or when parameters internal to the TRISO are needed. Most accident scenarios in HTRs are characterized by large time constants and slow changes in the fuel and moderator temperature fields. In these situations, a meso-scale, or pebble- and compact-scale, solution provides a good approximation of the fuel temperature as the fission thermal energy transports out of the kernel and into the surrounding matrix with a much shorter time constant. Therefore, in most cases, the matrix can be assumed to be in quasi-static equilibrium with the kernels. These models, however, fail to provide accurate information on the state of the various components of the TRISO during the early stages of transients. Since the coated particles constitute one of the fundamental design barriers for the release of fission products, it becomes important to understand the transient behavior inside this containment system. An explicit TRISO fuel temperature model named THETRIS has been developed and incorporated into the CYNOD–THERMIX-KONVEK suite of coupled codes. The code includes gas-release models that provide a simple predictive capability of the internal pressure during transients. The new model yields similar results to those obtained with other micro-scale fuel models of TRISO particles, but with the added capability to analyze gas release, internal pressure buildup, and effects of a gap in the TRISO. Analysis of bounding benchmark transients yield good agreement with other codes in which the TRISO particles are modeled explicitly. In addition, a sensitivity study of the potential effects on the transient behavior of high-temperature reactors due to the presence of an inter-layer gap is included. Although the formation of a gap occurs under special conditions, its consequences on the dynamic behavior of the reactor can yield responses during fast transients that depart significantly from those in which no gap is present in the model. The new model was applied to an extreme (beyond design basis) scenario in order to observe the behavior of the fuel during a large prompt critical reactivity insertion. Although a large amount of fission energy was deposited rapidly into the fuel, the kernel temperature is shown to stay well below the melting point and the silicon carbide layer remained well below the temperature above which failure is expected to occur. The explicit treatment of the TRISO particle geometry leads to much lower estimations of power peaking during the transient and a greater degree of negative Doppler feedback.     

采用TVS-2M组件的堆芯燃料管理及其经济性分析

田湾核电站3、4号机组正在考虑使用TVS-2M组件来提高经济性。本文使用KASKAD程序包,对田湾核电站从首循环起使用TVS-2M组件进行研究设计,给出了改进型的燃料管理方案。对采用和未采用TVS-2M组件的两种燃料管理方案进行了经济性分析。分析结果显示,采用TVS-2M组件可显著提高电站经济性。     

超临界水冷堆燃料管理经济性比较与分析

超临界水冷堆热效率高,其预期的燃料经济性高。本文将超临界水冷堆CSR1000与目前主流的压水堆、沸水堆进行燃料管理经济性比较,给出了超临界水冷堆燃料经济性更低的意外结论。因此超临界水冷堆能否真的成为第4代核能系统还有待商榷。     

废物处理流程中管控信息集成技术研究

废物处理工艺流程和设施设备状态的监控信息与管理信息的集成是废物处理与管理信息化建设的关键技术问题之一。探讨废物管控信息集成的主要内容,研究分别采用用于过程控制的对象连接与嵌入(OPC)通信技术、文件传输协议(FTP)技术以及数据推送技术,实现基于不同开发技术的监控系统与信息管理系统的信息集成方法,并应用于废物处理设施信息化建设。     

Fabrication of Mixed Urianium–Plutonium Fuel on the Paket Setup and Waste Handling Methods

The results of fabrication of fuel elements with mixed uranium–plutonium oxide fuel are presented. The experimental fuel assemblies assembled from the fuel elements were tested in BN-350 and -600 reactors. Postreactor investigations of the fuel elements showed that there was no substantial difference in the behavior of the fuel cores consisting of the mixed fuel as compared with UO₂ fuel. Solid and liquid radioactive wastes are produced during the fuel fabrication process. A classification of the wastes and methods for handling them is given. It is shown that the off-grade sintered pellets should be pulverized and returned to the beginning of the mixed-fuel fabrication process.     

INSIGHT: An Integrated Scoping Analysis Tool for In-core Fuel Management of PWR

An integrated software tool for scoping analysis of in-core fuel management, INSIGHT, has been developed to automate the scoping analysis and to improve the fuel cycle cost using advanced optimization techniques. INSIGHT is an interactive software tool executed on UNIX based workstations that is equipped with an X-window system. INSIGHT incorporates the GALLOP loading pattern (LP) optimization module that utilizes hybrid genetic algorithms, the PATMAKER interactive LP design module, the MCA multicycle analysis module, an integrated database, and other utilities. Two benchmark problems were analyzed to confirm the key capabilities of INSIGHT: LP optimization and multicycle analysis. The first was the single cycle LP optimization problem that included various constraints. The second one was the multicycle LP optimization problem that includes the assembly burnup limitation at rod cluster control (RCC) positions. The results for these problems showed the feasibility of INSIGHT for the practical scoping analysis, whose work almost consists of LP generation and multicycle analysis.     

秦山核电厂燃料管理程序及循环—1计算

本文修改PSUI-LEOPARD/NGMARC燃料管理程序,使其适用于秦山核电厂,形成新的版本——LEOPQS/NGMACQS,并用它计算了循环-1。计算结果与秦山核电厂的测量值符合很好。     

乏燃料水池严重事故现象及管理策略研究

针对我国二代改进型三环路核电厂乏燃料水池冷却管线破口事故(LOCA)引发的严重事故,使用MECLOR1.8.6程序进行了建模计算,分析研究了严重事故进程和乏燃料组件加热、熔化以及氢气的产生等主要现象。结果表明,乏燃料水池严重事故进程相对缓慢,但乏燃料组件的熔化及产生的氢气风险还是可能最终造成放射性向环境的大量释放。此外,本文还对乏燃料水池严重事故管理导则中的应急注水策略和氢气风险管理策略的有效性进行了计算分析,得到了严重事故下执行相关策略的时间窗口,从而为同类型核电厂严重事故管理导则的开发和有效执行提供支持。     

核电站退役废物管理

核电站产生的废物的处理尽可能延用运行废物处理系统,考虑到退役废物的废物类别及产生量与运行废物差异较大,可考虑增设必要的废物处理手段或扩大废物处理能力,是否新建废物处理设施应综合考虑是否有适宜场址、工期是否允许以及是否存在与退役无关且足够大的子项可进行改扩建等操作。文章以M310堆为例,估算了单机组退役可能的退役废物产生量,同时,为实现废物最小化的目标,提出了进行设计优化,严格运行管理、避免事故发生,严格进行废物分类,利用废物处理手段减少废物处置量和废物体积以及对材料再循环再利用等建议。     

Impact of Partitioning and Transmutation on High-Level Waste Disposal for the Fast Breeder Reactor Fuel Cycle

The impact of partitioning and/or transmutation (PT) technology on high-level waste management was investigated for the equilibrium state of several potential fast breeder reactor (FBR) fuel cycles. Three different fuel cycle scenarios involving PT technology were analyzed: 1) partitioning process only (separation of some fission products), 2) transmutation process only (separation and transmutation of minor actinides), and 3) both partitioning and transmutation processes. The conventional light water reactor (LWR) fuel cycle without PT technology, on which the current repository design is based, was also included for comparison. We focused on the thermal constraints in a geological repository and determined the necessary predisposal storage quantities and time periods (by defining a storage capacity index) for several predefined emplacement configurations through transient thermal analysis. The relation between this storage capacity index and the required repository emplacement area was obtained. We found that the introduction of the FBR fuel cycle without PT can yield a 35% smaller repository per unit electricity generation than the LWR fuel cycle, although the predisposal storage period is prolonged from 50 years for the LWR fuel cycle to 65 years for the FBR fuel cycle without PT. The introduction of the partitioning-only process does not result in a significant reduction of the repository emplacement area from that for the FBR fuel cycle without PT, but the introduction of the transmutation-only process can reduce the emplacement area by a factor of 5 when the storage period is extended from 65 to 95 years. When a coupled partitioning and transmutation system is introduced, the repository emplacement area can be reduced by up to two orders of magnitude by assuming a predisposal storage of 60 years for glass waste and 295 years for calcined waste containing the Sr and Cs fraction. The storage period of 295 years for the calcined waste does not require a large storage capacity because the number of waste packages produced is significantly reduced by a factor of 5 from that of the glass waste package in the FBR fuel cycle without PT.     

台山核电厂CEPR堆芯设计及燃料管理

台山核电厂采用了第三代压水堆核电技术的CEPR。堆芯采用较大的堆芯尺寸,降低了堆芯的线功率密度,提高了中子经济性;控制棒使用T模式,提高控制棒的控制效率,减小控制棒磨损;通过引入富集硼,优化了冷却剂的化学控制;从首循环开始进入18个月换料的燃料管理方案最大程度上提高了燃料经济性。     

废物处理与管理信息化建设探讨与实践

探讨核动力系统废物处理与管理信息化建设的目标、内容和要求,以及在系统设计和开发过程中所遵循的一些思路和做法。结合某废物处理设施信息化建设的工程实践及建设过程中的一些关键问题,介绍信息系统的组成和体现结构。     

Modeling spent TRISO fuel for geological disposal: corrosion and failure under oxidizing conditions in the presence of water

High temperature gas reactors (HTGRs) are being considered for near term deployment in the United States under the GNEP program and farther term deployment under the Gen IV reactor design (U.S. DOE Nuclear Energy Research Advisory Committee, 2002). A common factor among current HTGR (prismatic or pebble) designs is the use of TRISO coated particle fuel. TRISO refers to the three types of coating layers (pyrolytic carbon, porous carbon, and silicon carbide) around the fuel kernel, which is both protected and contained by the layers. While there have been a number of reactors operated with coated particle fuel, and extensive amount of research has gone into designing new HTGRs, little work has been done on modeling and analysing the degradation rates of spent TRISO fuel for permanent geological disposal. An integral part of developing a spent fuel degradation modeling was to analyze the waste form without taking any consideration for engineering barriers. A basic model was developed to simulate the time to failure of spent TRISO fuel in a repository environment. Preliminary verification of the model was performed with comparison to output from a proprietary model called GARGOYLE that was also used to model degradation rates of TRISO fuel. A sensitivity study was performed to determine which fuel and repository parameters had the most significant effect on the predicted time to fuel particle failure. Results of the analysis indicate corrosion rates and thicknesses of the outer pyrolytic carbon and silicon carbide layers, along with the time dependent temperature of the spent fuel in the repository environment, have a significant effect on the time to particle failure. The thicknesses of the kernel, buffer, and IPyC layers along with the strength of the SiC layer and the pressure in the TRISO particle did not significantly alter the results from the model. It can be concluded that a better understanding of the corrosion rates of the OPyC and SiC layers, along with increasing the quality control of the OPyC and SiC layer thicknesses, can significantly reduce uncertainty in estimates of the time to failure of spent TRISO fuel in a repository environment.     

Microstructure of TRISO coated particles from the AGR-1 experiment: SiC grain size and grain boundary character

Pre-irradiation SiC microstructures in tristructural-isotropic (TRISO) coated fuel particles from the Advanced Gas Reactor Fuel Development and Qualification program’s first irradiation experiment (AGR-1) were quantitatively characterized using electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). From EBSD, it was determined that only the cubic polymorph of as-deposited SiC was present and the SiC had a high fraction of coincident site lattice (CSL) Σ3 grain boundaries. Additionally, the local area misorientation (LAM), which is a qualitative measurement of strain in the SiC lattice, was mapped for each sample fuel variant. The morphology of the SiC/IPyC interfaces were characterized by TEM following site-specific focused ion beam (FIB) specimen preparation. It was determined that the SiC layer had a heavily faulted microstructure typical of chemical vapor deposition (CVD) SiC and that the average grain diameter increased radially from the SiC/IPyC interface for the samples manufactured with similar CVD conditions, while the last sample showed a nearly constant grain size across the layer.