快堆群常数连接程序LEK

LEK核数据处理程序是实现从ENDF/B-IV格式的评价核数据库中,用计算机自动取出快堆群常数程序KQCS的输入数据的程序。LEK取数既准确又节省大量的人力,改变了KQCS程序需手工输入上万个核数据的繁烦、落后局面,使中国核数据中心(CNDC)用计算机计算大量核素的快中子反应堆多群常数成为可能,为检验中国评价核数据库(CENDL)创造了条件。     

新一代压水堆燃料组件技术特点及其在华专利申请

以美国西屋电气公司的Next Generation Fuel燃料组件技术特点为线索,收集了美国西屋电气公司在中国燃料组件技术方面的专利申请和专利文献,从中筛选出与NGF燃料组件技术特点符合的专利申请和专利文献,对其技术方案进行了深入剖析,从中了解西屋新一代压水堆燃料组件技术的发展趋势。     

国外核潜艇反应堆系统事故浅析

本文针对国外已发生的核潜艇反应堆系统事故进行了梳理分析研究,发现国外核潜艇反应堆系统事故多发生于早期型号,近年各国在役及新一代核潜艇未出现反应堆系统发生事故的报道。此外,还发现各国已发生的核潜艇反应堆系统事故中,失水事故和反应性事故所占比例最大。本文研究表明,通过先进核安全方法及技术的采用、核安全文化的重视、核安全监管力度的加强,反应堆系统事故可不会给核潜艇带来额外的事故风险,核反应堆及核安全能够不成为制约核潜艇发展的主要因素。     

Systematic methodology for diagnosis of water hammer in LWR power plants

The paper gives the dimensions of the knowledge base that is necessary to carry out a diagnosis of water hammer susceptibility/root cause analyses for Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) nuclear power plant systems. After introducing some fundamentals, water hammer phenomena are described. Situations where each phenomenon is encountered are given and analytical models capable of simulating the phenomena are referenced. Water hammer events in operating plants and their inclusion in the knowledge base is discussed. The diagnostic methodology is presented through an application on a system in a typical light water reactor plant. The methodology presented serves as a possible foundation for the creation of an expert water hammer diagnosis system.     

The inherent safety characteristics of an unconventional LMR design—The Trench Reactor

This paper presents the transient behavior during off-normal operation of an unconventional liquid metal reactor design, called the Trench Reactor. Under the postulated accident conditions, this reactor design responds in an inherently safe manner to loss of heat sink accidents, loss of flow accidents, overcooling accidents and transient overpower accidents with 25 cents of reactivity insertion. The characteristics that cause such inherently save behavior are the properties of the materials and the configuration of the reactor primary system, even without any activated safety devices.     

A cost-effective methodology to internalize nuclear safety in nuclear reactor conceptual design

A new methodology to perform nuclear reactor design, balancing safety and economics at the conceptual engineering stage, is presented in this work. The goal of this integral methodology is to take into account safety aspects in an optimization design process where the design variables are balanced in order to obtain a better figure of merit related with reactor economic performance. Design parameter effects on characteristic or critical safety variables, chosen from reactor behavior during accidents (safety performance indicators), are synthesized on Design Maps. These maps allow one to compare the safety indicator with limits, which are determined by design criteria or regulations, and to transfer these restrictions to the design parameters. In this way, reactor dynamic response and other safety aspects are integrated in a global optimization process, by means of additional rules to the neutronic, thermal-hydraulic, and mechanical calculations.An application of the methodology, implemented in Integrated Reactor Evaluation Program 3 (IREP3) code, to optimize safety systems of CAREM prototype is presented. It consists in balancing the designs of the Emergency Injection System (EIS), the Residual Heat Removal System (RHRS), the primary circuit water inventory and the containment height, to cope with loss of coolant and loss of heat sink (LOHS) accidental sequences, taking into account cost and reactor performance.This methodology turns out to be promising to internalize cost-efficiently safety issues. It also allows one to evaluate the incremental costs of implementing higher safety levels.     

Examination of reactor grade graphite using neutron powder diffraction

Graphite is of principal interest in Generation IV nuclear reactor concepts. In particular, graphite will be the moderator for the Very High Temperature Reactor. In support of experimental and computational investigations that aim at understanding the behavior of reactor grade graphite under operating conditions, neutron powder diffraction experiments have been performed at the North Carolina State University PULSTAR reactor. The collected diffraction patterns exhibit intense broadening of several of the reflections, characteristic of turbostratic stacking. In order to quantify this disorder structurally, a model combined with a Rietveld-like refinement approach was implemented, which includes several refinable parameters that aim at describing this type of structure. Stacking parameters representing the probabilities of a random and registered shift between stacking packages were defined. The results indicate that the studied reactor grade graphite specimens contain a small fraction of layer disorder. The inferred interlayer spacing for the specimens is slightly larger than the theoretical value for graphite of 0.335 nm and the lattice constant is slightly less than 0.246 nm. The developed methodology is found to be successful in fitting the neutron diffraction patterns of reactor grade graphite.     

CANDU technology for Generation III+ and IV reactors

Atomic Energy of Canada Limited (AECL) is the original developer of the CANDU^® reactor, one of the three major commercial power reactor designs now used throughout the world. For over 60 years, AECL has continued to evolve the CANDU design from the CANDU prototypes in the 1950s and 1960s through to the second-generation reactors now in operation, including the Generation II+ CANDU 6. The next phase of this evolution, the Generation III+ Advanced CANDU Reactor™ (ACR™), continues the strategy of basing next generation technology on existing CANDU reactors. Beyond the ACR, AECL is developing the Generation IV CANDU Super Critical Water Reactor.Owing to the evolutionary nature of these advanced reactors, advanced technology from the development programs is also being applied to operating CANDU plants, for both refurbishments and upgrading of existing systems and components. In addition, AECL is developing advanced technology that covers the entire life cycle of the CANDU plant, including waste management and decommissioning. Thus, AECL maintains state-of-the-art expertise and technology to support both operating and future CANDU plants.This paper outlines the scale of the current core knowledge base that is the foundation for advancement and support of CANDU technology. The knowledge base includes advancements in materials, fuel, safety, plant operations, components and systems, environmental technology, waste management, and construction. Our approach in each of these areas is to develop the underlying science, carry out integrated engineering scale tests, and perform large-scale demonstration testing. AECL has comprehensive R&D and engineering development programs to cover all of these elements.The paper will show how the ongoing expansion of the CANDU knowledge base has led to the development of the Advanced CANDU Reactor. The ACR is a Generation III+ reactor with substantially reduced costs, faster construction, and enhanced passive safety and operating margins. The ACR is also the basis for the Generation IV Super Critical Water Reactor, which extends operation to higher temperatures and pressures.     

Radiological aspects of fusion reactor safety: Risk constraints in severe accidents

A methodology is proposed for determination of the constraints on severe accidents in lithium cooled fusion reactors, based on the potential hazards associated with such accidents. The method utilizes a probabilistic approach to risk calculation. The most effective mechanism for activation product release is found to be volatilization of structure as a result of lithium fires. Several factors were found to influence the consequence of lithium fires, most notably the reactor structural material type and total volume. It is concluded that the consequences of estimated maximum possible release from a properly designed fusion reactor are substantially less than the maximum light water reactor accident consequences.     

美国核电厂监督管理简况及其启示

美国核管会新的反应堆监督检查程序将监督管理力度集中在反应堆安全，辐射安全，电厂安全保卫3个领域，具体落实在初始事件，缓解系统，屏障完整性，应急准备，职业辐射安全，公众辐射安全，实体保卫这7项基本点上，并配合一系列的检查活动，达到更有效，客观，及时地评价核电厂运行安全水平的目的，结合我国核安全监督管理的实践及现状，我们应该吸取美国的经验，对有限的资源进行优化，把监管重点放在风险较大的问题上，以预防事故的发生，并适当地引入美国新的反应堆监督检查程序的一些思想及方法，发展和完善国家核安全局的监督管理模式。     

Calculations of reactivity-initiated transients in gas-cooled fast reactors using the code system fast

The FAST code system is a general tool for analyzing advanced reactors from the viewpoint of the static and dynamic behavior of the whole reactor system. It includes an integrated three-dimensional representation of the core neutronics, appropriate modeling of the core thermal-hydraulics and fuel pin behavior, coupled to models of the reactor primary and secondary systems. Use is made largely of well-established individual neutronic, thermal-hydraulic and fuel behavior modules. Clearly, it is important to verify the individual parts of the code, including the links between them. The paper is focused on this detailed verification procedure. Steady-state conditions, as well as the transient behavior of hypothetical reactivity-initiated accidents, are investigated for two specific gas-cooled fast reactors. While the first system, a CO₂-cooled CAPRA-CADRA core, is loaded with Superphénix-like MOX fuel, the second system being analyzed, a He-cooled Generation IV-like core, uses ceramic (U,Pu)C fuel dispersed in a silicon-carbide matrix. In the current study, the TRAC/PARCS elements of FAST are compared with the 3D-kinetics stand-alone ERANOS/KIN-3D code, which is considered state-of-the-art, using as far as possible equivalent options. A new methodology is proposed to improve a diffusion-theory, coarse-group PARCS-solution by scaling the original cross-section derivatives and input kinetic parameters.     

The swedish underground containment studies: State of art

Since 1973 studies of underground siting for nuclear power plants have been going on in Sweden. War protection, being the primary aim in accordance with the instructions, the first containment study has lead to siting in rock or in a pit. Rock siting gives better war protection than pit siting and also has less effect on the landscape, the cost being about equal. The second study was aimed at surveying the advantages and disadvantages of a rock sited 1000 MW BWA nuclear power plant from a reactor safety standpoint, compared to a plant above ground.Based on the instructions and considerations within the study group, the following criteria for the plant design have been established. (1) The plant should be designed to give protection against external acts of war with conventional weapons. (2) The plant should have a safety level equal to that of an above ground plant. It should fulfil the demands set by the authorities for above ground plants with respect to normal operation and accidents. No accidents that can be dealt with above ground may be permitted to result in more serious consequences, nor may they have a higher probability in a plant sited in rock. (3) The design of the plant should moreover utilize the possibilities of improving the safety afforded by rock siting. The criterion about war protection leads to siting in rock or pit, as shown in a previous CDL study. The study group has concentrated its work on rock siting.To clarify the two other criteria, the study group has outlined four alternative designs of a rock sited plant: (1) Reactor with complete pressure suppression (PS) containment placed together with central auxiliary equipment in a closed cavern with 48 m span that is placed about 50 m under the rock surface. (2) Reactor with complete PS containment placed together with auxiliary equipment in similar cavern as alternative (1) but open to the atmosphere. (3) Reactor placed ment. (4) Reactor placed in a containment directly surrounded by the rock. Auxiliary equipment placed in separate caverns.Standardization and quality improvement today are preferred to choosing new systems and more advanced technical solutions. Also, considering the desire from a safety standpoint, a rock sited plant should as much as possible exploit established technology. A consequence of the desire to use established technology is that the reactor cavern should be open to the atmosphere. If the cavern is closed, certain pipe rupture accidents may give overpressures that are difficult to master without large design alterations. The criterion about utilizing the possibilities for increasing the safety leads the interest to extreme improbable accidents, where certain advantages seem to be attainable with rock siting.A tight, strong cavern around the reactor would thereby be an ideal solution. This design appears, however, difficult to combine with the criterion about equal safety level as above ground. This criterion that controls the safety in the circumstances normally considered for nuclear power plants must be satisfied primarily, since extreme accidents have such very low probability. The tight cavern has therefore had to stand back for the open. The study shows, however, that an open reactor cavern can also be designed to significantly increase the protection of the surroundings in extreme improbable accidents compared to above ground plants.The chosen technical design of the reactor plant demands a cavern with a 45–50 m span. Caverns without strengthening efforts with such spans are used in mines, but have not previously been used for industrial plants. Studies of the stability of such caverns show that a safety level is attainable corresponding to the safety required for the other parts of the nuclear power plant. The conditions are that the rock is of high quality, that necessary strengthening measures are taken and that careful studies of the rock are made before and during the blasting, and also during operation of the plant.The third study was delivered to the government in 1977. One part in this study is going deeper in certain questions (safety, operation, maintenance, sabotage, war protection, cost and decommissioning). Another part aims to a broader view of risks and consequences in peace and war and also advantages and disadvantages of nuclear power plant for district heating.     

Analysis of heat source distribution in internal circulating surface heat transfer molten salt reactor

The Molten Salt Reactor (MSR) is one of the Generation IV nuclear reactor concepts that were selected by the Generation IV International Forum in 2000. The concept is based on liquid fuel instead of solid fuel assemblies. Besides the advantages, there are several aspects of operation that can hinder the realization of this reactor concept. In this paper, the authors investigate the neutronics behaviour of a new sub-concept that offers solutions for many of the technical problems. The analysis was performed using the particle transport code MCNPX 2.7. The paper focuses on the short-term and steady state heat source distribution in the fuel salt and in the graphite moderator. Accordingly, neither burn-up effects nor reactivity transients are considered. The sensitivity of the effective multiplication factor on different geometrical and material parameters was studied. The results obtained indicate that the main region of heat deposition is in the internal and external channels of the graphite moderator. Only a few percent of the total heat power is released in the graphite moderator, where the gamma and neutron related heat deposition is on the same scale. The results also prove that the heat source distribution does not change drastically upon the actuation of the control rods.     

An integral CFD approach for the thermal simulation of the PBMR Reactor Unit

A CFD method was developed to conduct integral thermal reactor analysis for the complete Reactor Unit of the Pebble Bed Modular Reactor (Pty) Ltd (PBMR). The requirement was however also to include very detailed aspects such as leakage and bypass flow paths through the reflector blocks and sleeves. The aim was therefore to investigate the influence of leakage and bypass flow on the thermal performance of the Reactor Unit in an integral fashion.The focus of this paper is to discuss the methodology that was developed. The discussion will firstly highlight all the required inputs, elaborate briefly on the underlying theory and how this was implemented into the CFD modeling capability. Results will be discussed briefly, but the focus is on the methodology.     

“玲龙一号”反应堆研发关键技术

模块式小型反应堆（SMR）是一种新型的核能系统。“玲龙一号”反应堆（ACP100）是我国完全自主创新的多用途模块化小型压水反应堆。本文介绍了ACP100的研发过程、堆芯设计和安全设计的主要特点,主要包括堆芯核设计、热工水力设计、安全设计理念、固有安全设计、事故应对策略等关键技术。ACP100反应堆通过基于全非能动的设计理念以及确定论与概率安全评价相结合的设计方法,极大地提高了安全性,超过了三代核电安全标准要求。      

中国先进研究堆严重事故辐射后果研究

针对高功率研究堆建在大城市远郊区的特殊情况，提出了中国先进研究堆（CARR）严重事故辐射后果的验收准则。为进行CARR严重事故排放方案的设计，研究了不同事故排放方案下，CARR发生严重事故时的环境辐射后果。最终推荐提高反应堆大厅密封性并优化事故后密闭与排风组合排放方案，实现了CARR工程无场外应急的安全设计目标。     

基于CFD的蒸汽发生器U型管内回流现象数值分析

反应堆一回路系统在自然循环条件下，蒸汽发生器（SG）部分U型管内可能会出现回流现象，利用计算流体动力学（CFD）方法，对某非能动三代反应堆蒸汽发生器U型管内流体的流动传热特性进行数值模拟分析。选取6组不同管长的U型管，对比分析U型管内单相流体的流动传热特性。基于数值仿真结果，得出6组U型管质量流量-进出口压降曲线，并?T分析了U型管长度和一次侧进口流体温度与二次侧壁面温度温差（?T）对流体回流的影响。研究结果表明，当?T一定时，随着进出口压降的降低，长管内更容易发生回流。当U型管长度一定时，?T越小越容易发生回流。      

Comparison of MCNPX-C90 and TRIPOLI-4-D for fuel depletion calculations of a Gas-cooled Fast Reactor

The Gas-cooled Fast Reactor is one of the reactor concepts selected by the Generation IV International Forum for the next generation of innovative nuclear energy systems. Several fuel design concepts are being investigated. Burnup depletion of mixed fuel of uranium and plutonium, cooled with gas in a fast neutron energy spectrum must be simulated. Various codes are being developed and/or adapted to improve the quality of the results, and also to reduce the computing time required for the simulations.     

在核电厂操纵人员资格核准工作中对其所要求文化程度的一点认识

本文阐述了核安全法规中对操纵人员文化程度的要求,并针对当前在操纵员执照核准工作中出现的新问题,结合对《高等教育法》的理解,提出了一点个人认识,并建议尽快改进和完善核电厂操纵人员执照核准工作,保证核电厂运行安全.