先进压水堆一回路地震反应谱分析及参数敏感性研究

压水堆一回路系统包含压力容器、蒸汽发生器、主泵、稳压器、主管道和波动管等重要部件,各部件在地震激励下的动态响应与整个系统的结构形式密切相关。本文从系统的角度,以非能动先进压水堆一回路为研究对象,运用ANSYS建立了其三维有限元模型,在模态分析的基础上,进行了三正交方向输入下的反应谱分析,得到了系统在地震载荷下的响应。并对反应谱输入角度和支撑刚度进行了敏感性研究,给出了这些特性参数对结构设计和分析的指导性意见。此外,通过直接积分法得到系统的地震时程响应,并与谱分析的模拟结果进行了对比,同时也为主泵等单个部件的详细地震分析提供位移、加速度输入。最后通过三维实体模型与集中质量模型抗震计算结果的比较,说明建立三维实体模型的必要性。本文为核电站一回路重要设备的结构分析提供了技术支持。     

JSFR key technology evaluation on fuel handling system

A simplified fuel handling system design for the demonstration Japan sodium-cooled fast reactor (JSFR) has been proposed. Fast Reactor Cycle Technology Development project phase I results of key technology evaluations on a pantograph fuel handling machine (FHM), a fuel transfer pot with two core component positions, dry spent fuel cleaning and minor actinide-bearing fresh fuel shipping cask are summarized. A full-scale FHM mockup has been fabricated and tested in the air accumulating performance and seismic tolerance data. A mockup fuel transfer pot with fins and chromium carbide coating has been fabricated and tested with sodium accumulating heat transfer performance data. Several sodium cleaning tests using a dummy subassembly has been conducted accumulating cleaning performance data. For fresh fuel shipping cask, a design tool for evaluation of heat transfer capability has been developed and a helium gas cask shows cooling capability of minor actinide-bearing fresh fuel. Those experimental and analytical efforts have shown that key technologies to develop simplified fuel handling system are matured enough to proceed large-scale sodium experiments and conceptual design study for the demonstration JSFR.     

Alexisismon isolation engineering for nuclear power plants

This paper reviews the special requirements regarding efficiency, licensibility (reliability) and cost which should be met to achieve an optimum base isolated nuclear power plant design. It then describes the Alexisismon-2, patented isolation system developed by the author, underlines its original properties (linearity and separation of functions) and presents a conceptual design of its application to a nuclear power plant. The great reliability of the system components is demonstrated. The efficiency of the A-2 is found to be very high: a reduction factor of the base shear induced in the plant higher than 25 is achieved for all examined real accelerograms scaled to 1 g GPA. So the isolation components, the structural system of the plant, its equipment and systems can be easily designed to remain in the elastic range of stresses and strain even for seismic input with GPA higher than 2 g.     

Evaluation of recriticality behavior in the material-relocation phase for Japan sodium-cooled fast reactor

As the most promising concept of sodium-cooled fast reactors, the Japan Atomic Energy Agency has selected the advanced loop-type fast reactor, so-called Japan sodium-cooled fast reactor (JSFR). Through the evaluation of event progressions during hypothetical core-disruptive accident (CDA) under the design extension condition, a CDA scenario for JSFR has been evaluated. It has already been demonstrated that in-vessel retention (IVR) against CDA could be achieved by taking adequate design measures under best estimate conditions.

The whole sequence of CDA scenario for JSFR was categorized into four phases according to the progress of core-disruption status. In the third phase, so-called material-relocation phase, the accident events would progress in the subcritical state. However, if the uncertainties about the molten state of core remnant and their discharge behavior outward from core are conservatively superposed, the disrupted core may lead up to recriticality.

In the present study, the factors leading to recriticality in the material-relocation phase were investigated using the phenomenological diagrams, and the recriticality behaviors were evaluated through parametric analyses using SIMMER-III/IV codes. The results of parametric analyses suggested that a significant mechanical energy leading to the boundary failure of reactor vessel would not be released even assuming recriticality due to the uncertainties about molten state and discharge behavior. Through the present evaluation of the hypothetical recriticality event, the CDA scenario for JSFR could obtain further robustness from the viewpoint of achieving IVR.     

Safety Design Requirements for Safety Systems and Components of JSFR

Safety design requirements for JSFR were summarized taking the development targets of the FaCT project and design feature of JSFR into account. The related safety principle and requirements for Monju, CRBRP, PRISM, SPX, LWRs, IAEA standards, goals of GIF, basic principle of INPRO etc. were also taken into account so that the safety design requirements can be a next-generation global standard. The development targets for safety and reliability are set based on those of FaCT, namely, ensuring safety and reliability equal to future LWR and related fuel cycle facilities. In order to achieve these targets, the defence-in-depth concept is used as the basic safety design principle. General features of the safety design requirements are 1) Achievement of higher reliability, 2) Achievement of higher inspectability and maintainability, 3) Introduction of passive safety features, 4) Reduction of operator action needs, 5) Design consideration against Beyond Design Basis Events, 6) In-Vessel Retention of degraded core materials, 7) Prevention and mitigation against sodium chemical reactions, and 8) Design against external events. The current specific requirements for each system and component are summarized taking the basic design concept of JSFR into account, which is an advanced loop-type large-output power plant with a mixed-oxide-fuelled core.     

Gas entrainment allowance level at free surface and gas dynamic behavior of sodium-cooled fast reactor

In a sodium-cooled fast reactor (SFR), inert gases exist in the primary coolant system either in a state of dissolved gas or free gas bubbles. The sources of the gas bubbles are entrainment and dissolution of the reactor cover gas (argon) at the vessel free surface and emission of the helium gas that is produced as a result of disintegration of B₄C control rod material. The gas in the primary system may cause disturbance in reactivity, nucleation site for boiling, etc. Therefore, it is a key issue from the design and safety viewpoint and the allowance level is necessary regarding the gas entrainment at the free surface and the gas bubble concentration in the primary system. In the present study, a gas entrainment allowance level at the free surface is discussed and rationalized for the Japanese SFR (JSFR) design. The influence of the gas entrainment is evaluated using the void fraction at the core inlet. Design criteria for the acceptable level of the gas entrainment and gas concentration are proposed in consideration of the background level of gasses in the coolant. For the purpose, a plant dynamics code VIBUL has been developed to apply to the JSFR design to evaluate the concentration distribution of the dissolved gas and the free gas bubble in the JSFR system. Using the plant dynamics code for the bubble behavior, the background level of the free gas (void fraction at the core inlet) has been obtained. Assuming that the total void fraction should be kept below 105% of the background level, a preliminary design allowance level of gas entrainment is proposed as the map in terms of the entrainment rate and the entrained bubble radius. Furthermore, the possibility of bubble removal and design requirement of the device is investigated to satisfy the allowance level. It is noted that the background level is already very low in comparison with the induced void reactivity by the void passing the reactor core.     

PRA-Based SMA: the First Tool toward a Risk-Informed Approach to the Seismic Design of the IRIS

International Reactor Innovative and Secure (IRIS) is an advanced, modular, medium-power PWR with an integral primary system layout. As part of the “safety-by-design_” philosophy that inspired the project from the very beginning, a risk-informed approach to its design phase is being adopted and a probabilistic risk assessment (PRA) is being used as an active tool in pursuing an advanced level of safety. Within this framework, a preliminary PRA-based seismic margin assessment (SMA) has been conducted to assess the ability of the IRIS standard design to respond to seismic events. A high confidence of low probability of failure at the core damage sequence level and then at the entire plant level is the primary result of the SMA model; in the end, it will have to ensure that IRIS can withstand the review-level earthquake of 0.5 g which is consistent with the upper bin level of the NUREG/CR-4334.1) In this preliminary phase of its development, in which the core of the quantitative data is critically extracted from the SMA of other PWR designs, the IRIS SMA model can be seen as a first step toward the development of an extensive seismic PRA model.     

Evaluation of nuclear system requirements for accommodating seismic effects

An integrated approach is presented for the design of nuclear facilities to accommodate seismic effects. Site evaluation and soil-structure interaction are discussed briefly since they determine the magnitude and characteristics of the input forcing function to be used in the analysis. The evaluation of the requirements of the nuclear system to accommodate the effects of specified seismic input forces involves the classification of components of the system, development of a mathematical model which adequately defines and relates the components of the system, consideration of coupling and resonance effects on the interactions among the components, selection of the methods of analysis and choice of the methods of solving the resulting equations. Specific design considerations and criteria for judging the acceptability of the design are discussed.     

多模块高温气冷堆地震概率安全分析关键技术研究

高温气冷堆核电厂采取多个反应堆模块匹配1个汽轮机的设计方式,即1台高温气冷堆机组会包含多个反应堆模块,这使多个高温气冷堆模块在地震外部事件下存在明显的相关性,因此在利用概率风险分析方法来全面地识别和评价高温气冷堆的地震风险时,需要从机组的角度充分考虑和模化机组内多个反应堆模块间的相关性。高温气冷堆示范电站已完成了较为完整的单模块地震概率安全分析,本文将以该分析结果为基础梳理出高温气冷堆多模块地震概率安全分析的关键技术要素并进行研究,研究内容包括多模块事件序列建模和地震相关性失效评价等关键技术,并针对多模块高温气冷堆提出了应用策略。然后以双模块设计的高温气冷堆示范电站为对象,以地震导致丧失厂外电始发事件为代表,对多模块高温气冷堆地震概率安全分析进行了实例分析获得远低于概率安全目标的释放类频率,且分析得到了高温气冷堆多模块事件序列建模策略与地震相关性失效的评价路线可行这一重要结论。     

Seismic response analysis of full-scale boiling water reactor using three-dimensional finite element method

This paper presents the three-dimensional finite element seismic response analysis of full-scale boiling water reactor BWR5 at Kashiwazaki-Kariwa Nuclear Power Station subjected to the Niigata-ken Chuetsu-Oki earthquake that occurred on 16 July 2007. During the earthquake, the automatic shutdown system of the reactors was activated successfully. Although the monitored seismic acceleration significantly exceeded the design level, it was found that there were no significant damages of the reactor cores or other important systems, structures and components through in-depth investigation. In the seismic design commonly used in Japan, a lumped mass model is employed to evaluate the seismic response of structures and components. Although the lumped mass model has worked well so far for a seismic proof design, it is still needed to develop more precise methods for the visual understanding of response behaviors. In the present study, we propose the three-dimensional finite element seismic response analysis of the full-scale and precise BWR model in order to directly visualize its dynamic behaviors. Through the comparison between both analysis results, we discuss the characteristics of both models. The stress values were also found to be generally under the design value.     

Progress on the Plant Design Concept of Sodium-Cooled Fast Reactor

An innovative concept of sodium-cooled fast reactor, named JAEA Sodium Cooled FR (JSFR) has been created and modified through the Feasibility Study on Commercialized FR Cycle System, aiming at full satisfaction of the development targets. A modified concept of JSFR applied double-wall straight tube type steam generator (SG) which is excelling in safety for sodium-water reaction has been developed. In addition, decay heat removal system suitable for the straight tube SG has been selected and in-service inspection and repair capabilities have been improved. As the result of this study, the potential to realize this plant concept has been obtained through evaluation concerning safety and economics.     

关于我国核电厂抗震设计基准的下限值

结合我国核电厂选址地震安全评价以及地震活动背景,对我国核电厂抗震设计基准的下限值进行了讨论。     

A scenario of core disruptive accident for Japan sodium-cooled fast reactor to achieve in-vessel retention

As the most promising concept of sodium-cooled fast reactors, the Japan Atomic Energy Agency (JAEA) has selected the advanced loop-type fast reactor, so-called JSFR. The safety design requirements of JSFR for Design Extension Condition (DEC) are the prevention of severe accidents and the mitigation of severe-accident consequences. For the mitigation of severe-accident consequences, in particular, the In-Vessel Retention (IVR) against postulated Core Disruptive Accidents (CDAs) is required. In order to investigate the sufficiency of these safety requirements, a CDA scenario should be constructed, in which the elimination of power excursion and the in-vessel cooling of degraded core materials are evaluated so as to achieve IVR. In the present study, the factors leading to IVR failure were identified by creating phenomenological diagrams, and the effectiveness of design measures against them were evaluated based on experimental data and computer simulations. This is an unprecedented approach to the construction of a CDA scenario, and is an effective method to objectively investigate the factors leading to IVR failure and the design measures against them. It was concluded that mechanical/thermal failures of the reactor vessel due to power-excursion/thermal-load could be avoided by adequate design measures, and a clear vision for achieving IVR was obtained.     

Status of Integrated IHX/Pump Development for JSFR

This paper illuminates the status of research and development on the integrated IHX/Pump concept. The integrated IHX/Pump is the incorporated component of the intermediate heat exchanger (IHX) and the primary pump. Among the innovative technologies of the Japan Sodium-Cooled Fast Reactor (JSFR) in the Fast Reactor Cycle Technology Development (FaCT) project, the integrated IHX/Pump concept is one of the major innovative ideas for plant economy by reducing the amount of material in the primary cooling system and the building volume. This report summarizes the view of the integrated IHX/Pump, a development plan, evaluation methods, and the present test results with the 1/4-scale IHX/Pump test device.     

Seismic protection technology for nuclear power plants: a systematic review

Seismic protection systems (SPS) have been developed and used successfully in conventional structures, but their applications in nuclear power plants (NPPs) are scarce. However, valuable research has been conducted worldwide to include SPS in nuclear engineering design. This study aims to provide a state-of-the-art review of SPS in nuclear engineering and to answer four significant research questions: (1) why are SPS not adopted in the nuclear industry and what issues have prevented their deployment? (2) what types of SPS are being considered in nuclear engineering research? (3) what are the strategies for location of SPS within NPPs? and (4) how may SPS provide improved structural performance and safety of NPPs under seismic actions? This review is conducted following the procedures of systematic reviews, where possible.

The issues concerning the use of SPS in NPPs are identified: cost, safety, licensing and scarcity of applications. NPPs demand full structural integrity and reactor's safe shutdown during earthquake actions. Therefore, horizontal isolation may be insufficient in active seismic zones and isolation in the vertical direction may be required. Based on the results in this review, it is likely that next generation reactors in seismic zones will include state-of-the-art SPS to achieve full standardised design.     

Design development of the HTGR core and its support structure seismic considerations

This paper discusses the development of the core support structure design from that employed on Fort St Vrain to recently announced contracts by Philadelphia Electric, Delmarva Power and Light and Southern California Edison for the large HTGR. Particular emphasis is given to the seismic considerations in the design of the structure for the large HTGR. The overall configuration of each reactor type is critically compared. Although similar components are employed, the basic difference in layout configuration results in significant conceptual differences in the structural and mechanical requirements of the core support components. The configuration and major components for the large reactor are described in some detail. The essential features and function of components are discussed. The graphite components in the core support floor and permanent reflector are designed to form a tight array during reactor normal operating conditions. This composite structure resists compressive loading due to differential gas forces and concrete pressure vessel movement. This tight array concept has important advantageous effects on primary coolant flow distribution and seismic capability.The paper discusses the inherent requirements and methodology in developing a standard plant design for high seismic sites. A design suitable for 0.15 g operating basis earthquake and 0.25 g safe shutdown earthquake has been developed which is applicable for over 80% of the expected sites in the USA. The HTGR core and support structure consists of many thousand graphite elements. It behaves as an inelastic body having random response when subjected to seismic excitation. The paper describes simplified analytical models which have been developed to investigate this phenomenon. An overview of a test program to substantiate and correlate with the analytical models is provided. The program addresses the interelement collision forces and frequencies of elements within the core and the load/deflection at the boundary. Various one, two and three-dimensional scale models have been tested. A summary of the objectives of the program is provided.     

Safety Strategy of JSFR Eliminating Severe Recriticality Events and Establishing In-Vessel Retention in the Core Disruptive Accident

In the Japan Sodium Cooled Fast Reactor (JSFR) design, elimination of severe power burst events in the Core Disruptive Accident (CDA) is intended as an effective measure to ensure retention of the core materials within the reactor vessel. The design strategy is to control the potential of excessive void reactivity insertion in the initiating phase by selecting appropriate design parameters such as maximum void reactivity on one hand, and to exclude core-wide molten-fuel-pool formation, which has been the main issue of CDA, by introducing an inner duct on the other hand. The effectiveness of these measures is evaluated based on existing experimental data and computer simulation with validated analytical tools. It is judged that the present JSFR design can exclude severe power burst events. Phenomenological consideration of general characteristics and preliminary evaluations for the long-term material relocation and cooling phases gave the perspective that in-vessel retention would be attained with appropriate design measures.     

塑性变形对管系地震动力响应的影响

在评述线弹性分析方法的基础上,阐明了在管系特别是核管系动力响应分析中考虑塑性变形影响的重要性,介绍了现有考虑塑性影响的方法及其存在的问题.指出要降低现行规范的保守性,提出合理的管系抗震设计方法,     

用于快堆主容器支撑的电磁阻尼隔震技术的研究

快堆核电站设计特点是高温薄壁,与压水堆的低温厚壁比起来,更需要考虑抗震设计。区别于传统的叠层橡胶和铅芯橡胶等隔震技术,本文提出了基于电磁阻尼原理的隔震技术。为了验证其有效性,制作了一个简易的电磁阻尼隔震支座,用作一个贮水容器的支撑。针对该水容器支撑系统在振动台上进行了抗震实验,验证了电磁阻尼隔震技术的有效性。