200MW核供热堆含钆可燃毒物棒

介绍了轻水堆可燃毒物的发展和钆可燃毒物的各种性能，采用压电水堆核电厂燃料元件稳态分析程序ＦＲＡＰＣＯＮ－２，分析了２００ＭＷ核供热堆采用含钆可燃毒物棒的各种设计考虑，并根据其设计参数，对不同含钆量的可燃毒物棒进行了稳态工况的性能分析。     

铅基快堆氧化物燃料性能分析程序FUTURE开发与初步验证

为了对铅基快堆氧化物燃料元件稳态工况下的服役性能和行为演化进行模拟计算,本文基于串行的半隐式耦合求解方法开发了铅基快堆氧化物燃料性能分析程序FUTURE。程序采用两步分析法实现了铅基快堆氧化物燃料棒全域热力分析与局部行为模型的多物理场耦合计算。通过各计算模块与模型算例、基准公式和现有程序的对比分析,对FUTURE程序进行了各分离效应的初步验证。结果表明,FUTURE程序能准确模拟铅基快堆稳态工况条件下氧化物燃料元件内部的温度演化、结构变形、应力分布和相互作用,并实现对燃料重构、氧和钚元素的迁移、裂变气体释放和服役期内液态铅铋腐蚀等内容的计算模拟。     

铅基快堆氧化物燃料性能分析程序FUTURE开发与初步验证

为了对铅基快堆氧化物燃料元件稳态工况下的服役性能和行为演化进行模拟计算,本文基于串行的半隐式耦合求解方法开发了铅基快堆氧化物燃料性能分析程序FUTURE。程序采用两步分析法实现了铅基快堆氧化物燃料棒全域热力分析与局部行为模型的多物理场耦合计算。通过各计算模块与模型算例、基准公式和现有程序的对比分析,对FUTURE程序进行了各分离效应的初步验证。结果表明,FUTURE程序能准确模拟铅基快堆稳态工况条件下氧化物燃料元件内部的温度演化、结构变形、应力分布和相互作用,并实现对燃料重构、氧和钚元素的迁移、裂变气体释放和服役期内液态铅铋腐蚀等内容的计算模拟。     

FeCrAl包壳燃料棒辐照行为研究

铁素体FeCrAl不锈钢具有成为耐事故燃料包壳材料的潜在价值。通过FeCrAl包壳燃料棒堆内性能的初步分析,评估FeCrAl包壳的堆内性能,并对FeCrAl包壳后续的研发及应用提出建议。使用FUPAC程序对FeCrAl包壳燃料棒的堆内稳态辐照行为进行了初步研究。分析结果表明,FeCrAl包壳燃料棒的温度、内压、应力应变均低于设计限值。     

铀氢化锆元件小型动力堆应用可行性分析

介绍了铀氢化锆燃料元件的主要性能和特点(尤其是热物理性能和堆内辐照性能),以及将铀氢化锆元件应用于动力堆所完成的一些研究概况.在此基础上,对铀氧化锆元件小型动力堆的技术可行性进行了论证和分析.结果表明:将铀氢化锆元件作为小型动力堆元件在燃料元件方面不存在严重的技术问题;使用细棒铀氢化锆元件的小型动力堆仍有较大的瞬发负温度系数,具有一定的固有安全性.     

秦山核电厂燃料元件内辐照考验的分析和评价

为了验证秦山核电厂燃料元件的堆内性能,在重水试验堆开展了3×3-2小元件堆内综合辐照考验。本文就影响考验结果的若干技术问题和考验条件进行了仔细的分析,充分论证了该试验具有的实际意义。考验件在堆内经历了相当电厂堆稳态工况和一般事故工况的考验。 考验棒最大燃耗达34GWd/tU,棒最大表面热负荷达1.39MW/m~2。在整个考验过程中没有发生考验棒的破损。文章最后就考验结果在验证燃料元件性能及其在电厂堆内安全可靠运行方面进行了评价。     

秦山核电厂燃料元件堆内辐照考验的分析和评价

为了验证秦山核电厂燃料元件的堆内性能,在重水试验难开展了3×3—2小元件堆内综合辐照考验。本文就影响考验结果的若干技术问题和考验条件进行了仔细的分析,充分论证了该试验具有的实际意义。考验件在堆内经历了相当电厂堆稳态工况和一般事故工况的考验。考验棒最大燃耗达34GWd/tU,棒最大表面热负荷达1.39MW/m_2。在整个考验过程中没有发生考验棒的破损。文章最后就考验结果在验证燃料元件性能及其在电厂堆内安全可靠运行方面进行了评价。     

研究堆燃料元件讨论会

核材料学会于10月26～31日在浙江普陀县召开了研究堆燃料元件讨论会。这次会议集中了我国研究堆燃料元件设计、研制、生产和运行单位的领导和科技人员。这次会议的中心议题是:关于我国研究堆燃料元件的设计、研制、无损检验和使用方面的经验交流,讨论研究堆今后的发展方向和对燃料元件的改进意见。     

高通量工程试验堆堆芯燃料元件温度-流量测量装置

一、引言为了获得燃料元件在辐照条件下性能的有关数据及保证高通量堆第一、二炉高功率、深燃耗安全运行,在堆芯K11栅格位置安装了一盒仪表燃料元件。利用它在反应堆运行、停堆、元件出堆期间完成了一系列稳态和动态试验的元件热工测量,为校核堆芯热工设计和摸清高通量堆的性能提供了实测数据。     

钠冷快堆燃料元件性能分析程序的开发与验证

为了评估钠冷快堆氧化物燃料元件稳态、瞬态和事故条件下的性能和行为演化,开发了钠冷快堆燃料元件性能分析程序FIBER。程序采用有限体积法实现燃料元件温度的计算,用有限元方法实现力学、裂变气体释放的计算,并通过时间步长控制模块控制程序的稳定运行。为验证程序的准确性,通过调研得到俄罗斯BN600反应堆辐照数据,与FIBER程序的裂变气体释放、柱状晶粒等计算结果进行对比分析。结果表明,FIBER程序对最大燃耗11.8at%、最大辐照损伤78 dpa的快堆燃料元件的辐照变形、柱状晶区、裂变气体释放性能评价是有效的。     

六角形轻水堆组件中子通量密度分布的计算

介绍利用穿透概率法求解二维六角形轻水堆燃料组件中子通量密度分布。子区内中子源及通量密度在空间上采用二次分布 ,子区表面通量密度在空间上采用平通量密度分布 ,在方向上采用简化 6P1近似。根据提出的模型 ,编制了TPHEX D程序 ,并对一些轻水堆六角形组件问题作了计算 ,计算结果与MC结果进行了比较 ,符合良好。本程序可用于六角形轻水堆燃料组件计算。     

A new mechanistic code SFPR for modeling of single fuel rod performance under various regimes of LWR operation

A new mechanistic code SFPR for modeling of single fuel rod behavior under various regimes of LWR reactor operation (normal and off-normal, including severe accidents) is under development at IBRAE. The code is designed by coupling of two stand-alone mechanistic codes MFPR (for modeling of irradiated UO₂ fuel behavior and fission product release) and SVECHA/QUENCH, or S/Q (for modeling of Zr cladding thermo-mechanical and physico-chemical behavior). Both codes were initially designed for accident conditions (and for this reason, are rather mechanistic) and later extended to various normal operation conditions. On the base of thorough validation against various out-of-pile and in-pile experiments, development of an advanced fuel performance code for best estimate code calculations for both normal and off-normal LWR operation regimes is foreseen.     

穿透概率法求解二维六角形轻水堆燃料组件中子通量密度分布

研究利用穿透概率法求解二维六角形轻水堆燃料组件内中子通量密度分布。子区内中子源采用线性分布,子区表面通量密度在方向上采用简化6P1近似。提出了六角形组件周边水隙的处理方法。根据提出的模型,编制了TPHEX－C程序,并对六角形组件进行了计算,结果与蒙特卡罗方法计算的结果符合良好。     

二维六角形轻水堆燃料组件中子通量分布的计算

介绍利用穿透概率法求解二维六解形几何多群中子积分输运方程。子区内中子源及通量采用线性分布,子区表面通量在方向上采用简化６Ｐ１近似。根据提出的模型,编制了ＴＰＨＥＸ－Ｂ程序,并对一些轻水堆六解形组件问题做了计算,计算结果与ＭＣ结果进行了比较,符合良好。本程序可用于六解形轻水堆燃料组件计算。     

On the state of LMFBR structural analysis and existing rules with regard to class 1 components of SNR 2

The design and construction of LMFBR components largely depend on thermal loading, normally coupled with the effects of cyclic operation. The structural behaviour must be assessed in relation to high temperature and time-dependent characteristic material data. This also leads to additional demands on analysis and evaluation procedures. Although a considerable effort has been made to cope with these requirements, the important phenomena of ratchetting, elastic follow-up, creep, creep-fatigue, time-dependent buckling and thermal striping continue to need to be studied. This has become evident during the construction of the recent LMFBRs. Some procedures are still conservative simplifications, others such as plastic strain concentration factors, creep-fatigue and the influence of multi-axial states on creep, relaxation and creep-fatigue need deeper and larger investigation.On the other hand non-elastic methods have been developed and used for final analyses and code assessment of important components and highly stressed regions as well as to assess equivalent linear model evaluation schemes. Thus, the rules have been completed to account for these requirements and supplemented by criteria formulated for each kind of non-linear calculations. These new code parts and the articles introduced for the requirements mentioned above cannot easily be grouped into a clear arrangement. The ways in which the diverse codes achieve this matter differ remarkably. Although the primary goal of improved and more sophisticated design rules is a more realistic code assessment, they should also be a tool for optimizations and, in accordance with constructional experience, lead finally to a code for LMFBR constructions similar to LWR codes. This paper presents comparative considerations of the statements formulated in the international codes and in the German licensing procedure with regard to the aforementioned problems.     

轻水堆乏燃料和钍燃料在ACR-700利用的探索

轻水堆乏燃料和钍燃料的利用是解决乏燃料后处理问题和核燃料短缺的有效途径之一。本工作以ACR-700标准燃料为参考,研究了4种不同混合比例的轻水堆乏燃料及钍燃料在ACR-700中的k_∞和燃耗。研究结果表明,将裂变产物分离后,轻水堆乏燃料的重锕系核素在ACR-700中可作为一很好的燃料;只要加入足够的启动燃料,钍燃料也可作为很好的转换燃料,使反应堆内生成²³³U的速率大于易裂变燃料的消耗速率,²³³U的生成对反应堆运行后期维持临界起重要作用。     

Irradiation swelling, creep and thermal-stress analysis of LWR fuel elements, computer code isune-2

The irradiation swelling, creep, and thermal-stress analysis of light-water reactor (LWR) oxide (UO₂) fuel elements is analysed. The analysis is based on the basic physical and mathematical assumptions and the experimental data of the fuel and cladding (or canning) materials. In the analysis, the nuclear, physical, metallurgical, and thermo-mechanical properties of the fuel and cladding materials under irradiation environment are examined carefully. The objectives of the paper are mainly (1) to formulate and carry out the irradiation swelling, irradiation creep, and thermal-stress analysis of fuel elements for LWR power reactors, and (2) to develop a computer code which will facilitate the computations for fuel element design, safety analysis, and economic optimization of the power reactors. In a general procedure of the analysis, the irradiation swelling, irradiation creep, temperature distribution, etc. in the fuel and cladding of the oxide fuel elements during the reactor in operation are studied. Some theoretical models and empirical relations (on the basis of accepted experimental data) for irradiation swelling and creep in the fuel and irradiation creep in cladding materials are postulated and developed. Some analytical and empirical relations (based on test results) for heat generation and temperature distribution in the fuel during fuel restructuring are derived. The fuel restructure is, in general, divided into the central void, columnar grain, equiaxed grain, and unaffected grain zones (or regions) after a sufficiently long period for the fuel elements to be irradiated (or operated). From these relations derived for irradiation swelling, irradiation creep, and temperature distribution in the fuel and cladding, together with the well-known strain-stress, incompressibility, compatibility, and stress equilibrium equations, the irradiation swelling, creep, and thermal-stress analysis for the LWR fuel elements can be carried out.From the analytical results obtained, a computer code, ISUNE-2 (which is in the sequence of computer code ISUNE-1 and -1A developed and used previously for liquid-metal fast breeder reactor fuel element design and safety and economic analysis), can be developed. With some reliable experimental data (measured during fuel elements in operation) as input, the computer code may predict various cases of LWR (oxide or carbide) fuel elements in operation. The general scope and resulting contribution of this paper is to provide a realistic analysis and a reliable operating LWR fuel element code for use by nuclear power utilities to predict the fuel element behavior in power reactors. The fuel element design, safety analysis, and economic optimization depend largely on the fuel element behavior in the power reactors.     

轻水堆燃料关联核素的特征比值与乏燃料堆型的辨别

应用燃耗分析程序MCCOOR计算压水堆和沸水堆的栅元模型满功率运行时2种不同初始燃料富集度情形下不同燃耗深度的燃料核素成分，分析轻水堆燃料的关联核素比值与燃耗深度的关系，获得了轻水堆燃料的关联核素特征比值，并探索了由乏燃料相应核素特征比值确定其堆型的可行性。      

Research and development related to PFM for aged nuclear components

At the Japan Atomic Energy Research Institute (JAERI), research activities related to probabilistic fracture mechanics (PFM) have been conducted as a part of the research program on aging and structural integrity of LWR components. This paper describes the outline of two activities related to PFM, i.e. the development of a PFM code and a contract research on ‘Application of PFM Methodology to Reliability Assessment of Nuclear Components’ implemented by the Japan Welding Engineering Society (JWES). In the former research, a new PFM code PASCAL (PFM Analysis of Structural Components in Aging LWR) was developed. This code has some new functions in models of semi-elliptical crack extension, elastic–plastic fracture analysis based on R6 method and options for the evaluation of overlay cladding and warm pre-stress (WPS) effect. Besides, the code has the function to evaluate the effect of irradiation embrittlement recovery by thermal annealing of a reactor pressure vessel and re-irradiation embrittlement. Based on the analyses on benchmark problem conducted by USNRC/EPRI, performance and functions introduced in the code were examined. Some case studies were also carried out to investigate the influence of various parameters. On the other hand, JAERI has been sponsoring the PFM related activities in relation to the structural integrity of LWR components. These activities have been conducted at JSME and JWES. The objective of this activity has been to provide for the future need of PFM methodology.