小型氟盐冷却高温堆耦合布雷顿循环系统分析与研究

为满足小型氟盐冷却高温堆（FHR）能量转换需求,开发与之匹配的高效、紧凑、无水冷却动力转换系统,本文对比了超临界二氧化碳（SCO₂）、空气、氩气（Ar）、氮气（N₂）、氙气（Xe）5种气体工质在不同布雷顿循环构型中的热电转换效率、?效率、?损失分布。研究发现,SCO₂布雷顿循环相比其它工质循环具有最高的热电转换效率和?效率,且结构更为紧凑,易于小型化和模块化,与小型氟盐冷却高温堆耦合更具优势;进而对SCO₂布雷顿循环进行构型优化,得出匹配小型氟盐冷却高温堆的最佳循环构型方式,构成固有安全模块化小型氟盐冷却高温堆热电转换系统,为西部能源利用提供新研究思路。      

利用HHT产生与多阻尼楼板谱匹配的地震波

美国原子能管理委员会（USNRC）规范规定了用于核电厂抗震分析和设计的地震波要求。在抗震分析和设计中，采用的地震波可与多阻尼目标反应谱匹配，也可与单阻尼目标反应谱匹配。然而，在对核电设备和部件进行动力时程分析时，则需要与多阻尼目标楼板谱匹配的地震波。基于此问题，提出利用希尔伯特-黄变换（HHT）方法，通过修改种子地震波的频率和振幅信息，使之与多阻尼目标楼板谱匹配，且完全符合USNRC规范的匹配标准，从而为核电设备和部件的地震安全评估提供合适的地震激励。      

匹配设计反应谱的目标功率谱密度的确定方法

美国核管会标准审查大纲（SRP）3.7.1节要求，核电厂结构、系统和部件（SSCs）抗震设计时程需同时满足包络设计反应谱和匹配设计反应谱的目标功率谱密度（PSD）的要求。本文结合2014版SRP 3.7.1，对匹配设计反应谱的目标PSD的确定方法进行介绍，并根据其算法编写相应的计算程序，通过算例分析对程序结果进行验证。结果表明:计算所得RG1.60谱、美国中东部和西部基岩厂址谱的目标PSD与SRP 3.7.1结果具有较好的一致性，且基于本文所得目标PSD和三角级数叠加法所构造的加速度时程反应谱与设计反应谱匹配良好。本文所给出的目标PSD的确定方法可为核电厂抗震设计时程的PSD检验提供重要依据，且采用本文方法生成目标PSD，设计时程的PSD检验仅需包络该目标PSD的70%。      

AP1000核电站主控室盘台抗震鉴定试验研究和工程实践

抗震试验是设备鉴定的一部分,AP1000核电设备的抗震鉴定相比于传统抗震鉴定有了新的要求和方法。为满足AP1000核电设备的抗震鉴定要求,本次试验与传统抗震试验有所不同。本文以完成的主控室盘台抗震试验为例,介绍和分析该试验在反应谱、加速度计布置、功能性测试等方面的特殊要求。试验结果表明主控室盘台满足AP1000抗震鉴定的要求。这些特殊要求不仅可保证很好地鉴定试验件的结构完整性和安全功能性,而且能发掘其设计裕量。     

先进轻水堆安全系统简化方案研究

本文研究了国内外工程经验、法规标准和用户要求，提出了一套简化先进轻水堆安全系统配置方案。这套安全系统采用非能动安全系统应对设计基准工况（DBC），采用能动安全系统应对设计扩展工况（DEC）。工程判断和分析表明，这套安全系统可以应对所有DBC和DEC，与现有“华龙一号”相比，安全性一定程度提升，经济性显著提升。     

核电站关键设备存储装置的抗震性能研究

本文应用ANSYS软件对安全壳设备闸门的存储装置建立了完整的有限元模型,分析了设备闸门存储装置在OBE地震工况下的结构强度。本文在模型中采用壳单元对设备闸门组件和存储装置全部建模,以模拟实际工况下的整体结构,并对设备闸门存储装置抗震性能进行分析。研究结果表明,在OBE地震工况下,存储装置整体钢结构和所有连接螺栓强度均满足RCC-M规范的要求。本文采用有限元分析方法,通过模态分析和响应谱分析,对设备闸门存储装置的抗震性能进行了研究,该方法对设备闸门存储装置的抗震设计和校核具有一定的意义,同时,适用于乏燃料干式存储设备装置的抗震设计。     

试验堆主厂房楼板谱计算与比较

土壤－结构动力相互作用分析是核工程抗震设计与安全分析中的重要环节。核在地震动中的反应主要取决于地震特性、地基土特征与结构特性三个方面。对于非岩石地基上,由于构筑物基础面的运动受到土壤－结构动力相互作用的影响,与地表自由场运动会有显著的差别。在抗震设计规范中要求考虑这一相互作用的影响。本文采用FJUSH、SASSI1000程序进行了试验堆土壤－结构的动力分析,给出堆厂址地表自由场运动与各楼层的反应谱,以供结构与设备抗震设计之用,并进行了比较分析。     

中国实验快堆控制棒驱动线抗震试验研究

本工作依据相关规范,参考当前核电厂控制棒驱动线抗震试验的先进技术,结合中国实验快堆控制棒驱动线的结构特点,对中国实验快堆安全棒驱动线进行了抗震鉴定试验研究.研究结果为其安全评审提供了重要数据.     

核设备抗震鉴定试验中功率谱密度检验方法分析评估

为满足核设备抗震鉴定试验中输入运动的功率谱密度（PSD）要求，基于对规范背景和目标PSD算法的调研以及典型算例的对比分析，对PSD的检验方法进行分析评估。结果表明，检验PSD最为直观的方法即对比输入运动PSD与目标PSD；根据各类目标PSD算法的结果精度、保守性及其规范依据，推荐使用2014版美国核管理委员会标准审查大纲（SRP）3.7.1节附录B中人工合成时程的方式来计算目标PSD:虽然该算法通常适用于核电厂的厂址设计反应谱，但对于设备抗震鉴定反应谱，仅需将人工合成时程的目标反应谱替换为鉴定反应谱即可；采用本文推荐方法计算目标PSD时，设备抗震鉴定输入运动的PSD检验应与SRP 3.7.1保持一致，即在0.3 Hz到目标反应谱的最高截断频率范围内包络目标PSD的70%。      

中国实验快堆堆容器应力强度计算与评定

概述了中国实验快堆（CEFR）堆容器应力强度的计算与评定过程，重点介绍了在计算与评定中遇到的等效热民率、温度场热应力计算、热冲击计算、套管接管力旋加及复杂结构的粗细网格过渡等设计中需解决的问题。计算与评定结果表明：CEFR堆容器的应力强度满足《ASME规范》和《核电厂抗震设计规范》（GB50267－95）要求。     

A proposal for an aseismic design method of equipment and piping for NPPs in a low seismicity area

Recently a regulatory code for an aseismic design of high-pressure gas facilities became effective by the order of the Ministry of International Trade and Industry (MITI) in Japan. This order includes details of the aseismic design of vessels whose “factor of importance” are relatively lower than Class A (Class I) items in nuclear power plants.The author develops his idea on an aseismic design method of equipment and piping of nuclear power plants in a Low Seismicity Area (LSA) based on his experience of the new code for petro-chemical industries and oil refinaries pertaining to high pressure gas facilities mentioned above.The definition of LSA is usually the area whose maximum intensity has never exceeded MMI VI or VII. However, there are two types of LSA, one is really such a low seismicity area, and the other type is the area which has the possibility of stronger earthquake occurrence than those mentioned above, even though it is low. One of the typical examples is the area subjected to “New Madrid Earthquake-1812”. The author develops his concept along these two lines.He briefly describes the new code for high-pressure gas facilities in Japan. This code describes the design methodology of both types of aseismic design analysis, that is, rather sophisticated dynamic methods for facilities whose potential hazard is as high as those in a nuclear power plant, such as liquified chlorine gas storage, and simplified dynamic and static methods for most of the equipment and vessels in those plants. One of the features of this code is the use of design formulae and charts to simplify their design procedure as well as the set of specific computer codes by the MITI. These computer codes are prepared by the MITI or approved by the MITI for providing equivalent capability to the practice designated in the MITI order.The author's philosophy for the code of equipment and pipings in LSA is that they must be as simple as possible, and most of the analytical work for the design should be eliminated, or at least limit the use of simplified methods, such as the static seismic coefficient method or the modified seismic coefficient method with a simplified response spectrum. The use of general design criteria or a guideline of structural details may be better than a sophisticated design analysis as a result.     

Seismic analysis of liquid storage container in nuclear reactors

Seismic analysis of liquid storage containers is always difficult in the seismic design of nuclear reactor equipment. The main reason is that the liquid will generate significant seismic loads under earthquake. These dynamic liquid loads usually form the main source of the stresses in the container. For this kind of structure–fluid coupling problem, some simplified theoretical methods were usually used previously. But this cannot satisfy the requirements of engineering design. The Finite Element Method, which is now full developed and very useful for the structural analysis, is still not mature for the structure–fluid coupling problem. This paper introduces a method suitable for engineering mechanical analysis. Combining theoretical analysis of the dynamic liquid loads and finite element analysis of the structure together, this method can give practical solutions in the seismic design of liquid storage containers.     

Seismic analysis of ZPR-6 Reactor Facility

A safe shutdown earthquake analysis of ZPR 6 Reactor Facility was conducted through seismic risk analysis, soil-structure interaction analysis, reactor building dynamic time history analysis and equipment response spectrum analysis due to an assumed El Centro earthquake. Several ASME, AISC and ANSI design codes were used to demonstrate the adequacy of this facility and to design several equipment and piping supports.     

Development of the site-specific uniform hazard spectra for Korean nuclear power plant sites

A seismic risk analysis has been performed to evaluate the seismic safety of a nuclear power plant for strong earthquakes beyond a design earthquake level. A site-specific median spectrum has generally been used for a seismic fragility analysis of structures and equipment in Korean nuclear power plants as a part of a probabilistic seismic risk assessment. The site-specific response spectrum, however, does not represent the same probability of an exceedance over entire frequency range of interest. The site-specific uniform hazard spectrum (UHS) is more appropriate for use in a seismic probabilistic risk assessment (SPRA) than the site-specific spectrum, since there are only a few strong motion data and seismological information for the nuclear plant sites in Korea. In this study, the uniform hazard spectra were developed using the available seismic hazard data for four Korean NPP sites.     

压水堆重反射层堆芯核热耦合高精度计算分析研究

重反射层的应用可提高反应堆中子经济性,其结构和中子吸收特性均与压水堆常规围板/反射层差异较大,因此对核设计程序的计算分析能力提出了新的要求。为分析重反射层建模方案对堆芯中子学计算结果的影响,使用先进中子学程序SCAP N和确定论堆芯高保真模拟程序NECP X对压水堆重反射层问题进行了高保真模拟,分析了5种反射层建模方案下计算结果的差异,并将高精度计算结果与商用核设计程序系统进行了对比。数值结果表明,重反射层水洞内冷却剂温度变化对计算结果影响较小;相较精确建模方案,重反射层铁水打混建模方案造成的反应性计算偏差在±30 pcm以内、组件相对功率分布计算偏差在±2%以内。     

快堆核电厂热力系统加热器设计方案热经济性分析

核电厂热力系统设计需在保证安全性的前提下尽量提高经济性。经济性的影响因素较多，包括技术成熟度、设备成熟度以及系统热经济性等。为了从热经济性的角度确定最优的快堆核电厂热力系统加热器配置，以俄方800 MW钠冷快堆热力系统为参考，基于热平衡分析法建立了适用于快堆核电厂的加热器、立式汽水分离再热器等设备的热力分析模型，进而开发了快堆核电厂热力系统热经济性分析程序，利用俄方设计计算值进行了程序验证。利用新开发的程序研究了不同加热器布置方案、给水焓升分配方案，确定了等焓降分配法的给水焓升分配方案热经济性最好。     

Seismic proof test of a reinforced concrete containment vessel (RCCV): Part 3. Evaluation of seismic safety margin

In Japan, the Nuclear Power Engineering Corporation (NUPEC), sponsored by the Ministry of Economy, Trade and Industry (METI), had conducted a series of seismic reliability proving tests using full-scale or close to full-scale models to simulate an actual important equipment that is critical for seismic safety of nuclear power plants. The tests are intended to validate the seismic design and reliability with a sufficient margin even under destructive earthquakes. A series of tests was carried out on a reinforced concrete containment vessel (RCCV) for advanced boiling water reactor (ABWR) from 1992 to 1999. A large-scale high-performance shaking table at Tadotsu Engineering Laboratory, was used for this test. The test model and the results of pressure and leak tests are described in Part 1. Test procedures, input waves and the results of verification tests such as changes of stiffness, characteristic frequency and damping ratio, the failure of the model and the load–deformation relationship are described in Part 2. Part 3 reports the seismic design safety margin that was evaluated from the energy input during the failure test to a design basis earthquake. Part 4 will report simulation analysis results by a stick model with lumped masses.     

Philosophy and practice of the aseismic design of nuclear power plants — Summary of the guidelines in Japan

Aseismic design is considered to be one of the most important factors for the safety of the nuclear power plants built in zones of high seismicity such as Japan. All structures, equipment and piping are classified in accordance with the importance of their radioactive safety to the plant, and the dynamic analysis and/or factored seismic coefficient analysis are applied accordingly. Site and ground conditions, as well as seismicity, should be studied thoroughly in order to estimate the intensities of the design earthquake and the safety margin check earthquake. Dynamic analyses of buildings and structures are performed using the multi-lumped-mass-system supported by soil springs with time history analysis conceptions. This idea is also applied to the design of equipment and piping by coupled system to the major structure or by the floor response spectra criteria. Tolerances are applied to damping factors although some experiments show more realistic results. Allowable stresses of ferrous metals for equipment and piping during earthquakes are more scientifically precise.

This report summarizes a guideline for aseismic design of nuclear power plants. The guideline was prepared by the Japan Electric Association in April, 1970, after three years laborious work.

In sect. 1, the philosophy and criteria are described. All components of a plant should be classified into three classes in accordance with their contributions to reactor safeties. Design to earthquake loadings should be based on “design basis earthquake” which is decided in consideration of local seismicity.

In sect 2, site selection and review for ground are described in the sense of seismic aspects.

In sect 3, deciding the earthquake motion for design is discussed. In Japan, semi-statistical approaches are used in normal practice.

In sect. 4, design philosophy and practice of building structures and containment vessels are described. They are designed under statical seismic forces, and the design of the class “A” structures should be checked by a dynamic response technique.

In sect. 5, design philosophy and practice of piping, vesels and equipment are described. Those which belong to class “A” items should be designed in a dynamic sense. Several programs for dynamic analyses of these items are prepared. Allowable stress under earthquake conditions is discussed in relation to other codes, for example, ASME Section III.

The greater part of the philosophy and design criteria have been adopted to all nuclear power plants which have been and are currently being built in Japan.     

储液容器地震易损性参数计算

本文介绍了核电厂设备的易损性分析方法以及易损性模型的参数化计算方法。对核电厂中的典型储液容器应急补水箱（ASG水箱）使用Housner质量-弹簧简化模型进行了分析。对ASG水箱的各项易损性参数进行了计算，绘制出其易损性曲线，并得出高置信度低失效概率（HCLPF）值。结果表明：ASG水箱的HCLPF值低于安全停堆地震（SSE）水平，属于抗震能力较低的设备，需在结构上进行加强。