Standardized seismic design spectra for nuclear plant equipment

Standardized seismic design spectra were developed for the nuclear plant equipment. Such standardized spectra will reduce unnecessary spectrum computation for the individual plants as well as expedite the selection, purchase and qualification of equipment. Most attractive of all, standard equipment such as electric components can be conservatively qualified just once and then used by all plants located in the same seismic zones. Presented here are idealized design spectrum peak envelopes for different types of structure materials and for both the SSE and OBE conditions. Procedures are also provided for the application of the design spectrum peak envelopes to the seismic qualification of equipment either by dynamic analysis or testing.     

基于TMSR-LF1的Ⅱ类研究堆抗震设计方法研究

抗震设计是核设施为满足安全与经济综合要求进行设计时的重要内容,目前研究堆的抗震设计缺乏相应的规范与研究,尚未发现较为完善的方法体系。本文推荐了一个匹配结构与设备的Ⅱ类研究堆抗震设计方法,以50 a超越概率2%地震动作为安全停堆地震（SSE）,并以2 MW液态燃料钍基熔盐实验堆（TMSR-LF1）为例,对比分析了采用该方法与采用其他相关规范方法得到的设计反应谱（DRS）,并将其应用于结构和设备的抗震设计计算中。结果表明：推荐方法在满足结构与设备的抗震设计匹配性的前提下,相比核电规范具有较好的经济性,相比民用规范具有较好的保守性,更加合理。     

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。      

控制棒驱动机构抗震支承装置材料国产化研究

核岛设备控制棒驱动机构(CRDM)抗震支承装置是反应堆堆顶重要的抗震和支承部件,属于安全相关级设备(LS级),其生产制造已经实现国产化,因此实现其结构材料以及焊接材料的国产化至关重要.本文就核岛设备CRDM抗震支承装置母材和焊接材料的国产化设计进行了论述,通过与国外成熟核电厂建设中采用的主体结构材料以及焊接材料在化学成...     

Improvement of the seismic safety of existing nuclear power plants by an increase of the component seismic capacity: A case study

This case study produces the scenario earthquakes for an example nuclear power plant (NPP) site and suggests the effective seismic capacity of safety-related equipment and components which significantly contribute to a core damage to improve the seismic safety of an existing NPP by using a probabilistic safety assessment. The response spectra for the scenario earthquakes show greater spectral accelerations than those for the design response spectrum in the frequency range higher than about 12 Hz. In order to improve the seismic safety of an example NPP, the effects of the seismic capacity of safety-related equipment and components on the core damage frequency (CDF) are investigated, and their effective seismic capacities are determined. The results of the case study show that an increase of the seismic capacity of the equipment reduces the CDF considerably. The effective seismic capacities for the diesel generator, offsite power, condensate storage tank and battery rack are determined as 0.84, 0.35, 0.63 and 0.63 g, respectively.     

Seismic fragilities for nuclear power plant risk studies

Seismic fragilities of critical structures and equipment are developed as families of conditional failure frequency curves plotted against peak ground acceleration. The procedure is based on available data combined with judicious extrapolation of design information on plant structures and equipment. Representative values of fragility parameters for typical modern nuclear power plants are provided. Based on the fragility evaluation for about a dozen nuclear power plants, it is proposed that unnecessary conservatism existing in current seismic design practice could be removed by properly accounting for inelastic energy absorption capabilities of structures. The paper discusses the key contributors to seismic risk and the significance of possible correlation between component failures and potential design and construction errors.     

Regulatory perspective of seismic issues

Some of the current seismic issues facing the nuclear power industry, such as seismic design criteria (USI A-40), seismic qualification of equipment in operating nuclear power plants (USI A-46), eastern United States seismicity, operating basis earthquake (OBE) exceedance criteria, seismic instrumentation, post OBE inspection of nuclear power plants, anchor bolts too close to a free edge, seismic margins of plants, and the potential for external events to cause severe accidents, are presented and the Nuclear Regulatory Commission's perspective on the resolution of these issues are discussed.     

Dynamic fragility concepts for equipment design and qualification

Fragility concepts are explored for use in the design and qualification of nuclear plant equipment and for relating the ultimate capability of equipment to that of the overall plant. In the most general sense, the fragility level of a device may depend on several different types of environmental stress or challenge factors (i.e., heat, nuclear radiation, vibration, etc.) that influence its operation. However, emphasis is concentrated on the dynamic and particularly the seismic fragility levels of equipment. A general definition of dynamic fragility and various methods for its measurement are described. The state of published data on nuclear equipment fragility is discussed, and limitations on its use are noted. From there, the concept of a standardized seismic fragility data base and its potential uses are considered. Various gaps in the methodology are identified, and recommendations for further research are suggested.     

Vertical seismic response of overhead crane

Vertical seismic response behavior is an important issue for the seismic design of equipments. The equipment, which is comparatively soft and unrestrained vertically, may resonate and its response is significantly magnified under vertical seismic excitation. Overhead crane is an example of equipment that is unrestrained vertically. The dynamic behavior of an 150-ton-capacity overhead crane under vertical seismic excitation was investigated by scale model excitation test and nonlinear time history analysis. The excitation tests were performed with several input levels and the vertical response with each input level was obtained. The simulation analysis approximately corresponded to the results of the excitation test.     

Seismic individual plant examination of external events of US nuclear power plants: insights and implications

As part of the implementation of the severe accident policy, nuclear power plants in the US are conducting the individual plant examination of external events (IPEEE). Seismic events are treated in these IPEEEs by either a seismic probabilistic risk assessment (PRA) or a seismic margin assessment. The major elements of a seismic PRA are the seismic hazard analysis, seismic fragility evaluation of structures and equipment and systems analysis using event tree and fault tree analysis techniques to develop accident sequences and calculate their frequencies of occurrence. The seismic margin assessment is a deterministic evaluation of the seismic margin of the plant beyond the design basis earthquake. A review level earthquake is selected and some of the components that are on the success paths are screened out as exceeding the review level earthquake; the remaining ones are evaluated for their seismic capacity using information from the original plant design criteria, test data and plant walkdown. The IPEEEs of over 100 operating nuclear power plants are nearing completion. This paper summarizes the lessons learned in conducting the IPEEEs and their applicability to nuclear power plants outside of the United States.     

Seismic re-evaluation of EDF Bugey 900 PWR nuclear power plant in the frame of the 3rd periodic safety review

As compared to the previous Periodic Safety Review (PSR), main features of the seismic facet of the Bugey NPP (France) 3rd PSR are relating to (i) a 45% increase of the Review Level Earthquake (RLE), (ii) a dramatic increase of computer code capabilities and (iii) a certain evolution of safety standards. This paper presents the method implemented to cope with this situation. Seismic input motion is determined according to an updated Basic Safety Rule published in 2001. Concerning buildings, 3D models are established, accounting for torsion effects, which were not represented by the initial design spring-mass models; analyses are carried out as per the IAEA methodology, accounting for ductile capacity through a F_μ factor; finally the major impact of the seismic re-evaluation is a reinforcement of electrical building, so that it is less torsion sensitive. Regarding equipment, the concept of Safe Shutdown Equipment List (SSEL) is enlarged to all the Safety Related Equipment List (SREL). An equipment hierarchy classification is introduced, based on two complementary approaches: a system approach that identifies components associated to success paths for a safe shutdown, and a ruggedness approach that considers experience feedback about seismic behavior of components. On these bases, all the safety related equipments are classified in two categories: High Hierarchy level (HH) and Medium Hierarchy level (MH). MH equipment is mainly verified by walkdown, whereas HH equipment is by calculation, taking benefit from design margins. The optimization procedures enable to focus on the more relevant components. Bugey NPP seismic re-evaluation stands out by its realistic calculation chain and the completeness of its assessment, and eventually, the 3rd PSR is an opportunity to improve the plant safety and increase design margins.     

Research to provide criteria for more economical nuclear power plant structures and components

The material presented in this paper defines areas for future research which may lead to more economical nuclear power plant structures and components. The major points of discussion are: (1) better definition of loads and load combinations; (2) improved techniques for determining material properties and structural capacities and behavior; (3) areas where standardization and simplified techniques may be used; (4) more realistic equipment seismic input and economical seismic qualification techniques; (5) improved information on equipment support hardware; and (6) aseismic inelastic design criteria for Category I structures and components.     

大亚湾核电站反应堆厂房楼层反应谱分析评估

大亚湾核电站核岛厂房的抗震分析遵循技术输出国-法国M310型机组的土建技术规范RCC-G,采用简化的阻抗函数法计算地基岩土的作用.根据大亚湾厂址的地基岩土特点,拟采用更为精确的三维连续半空间边界子结构法来考虑地基岩土的作用,并与原设计进行对比.另外,在原设计中采用多组时程作为地震输入,取各组计算结果的平均值作为设计值的基础(称为"平均"法).在研究中基于相同的时程,拟分别采用"平均"法和更为常用的"包络"法,处理多组时程的响应.基于上述两方面,通过反应堆厂房的地震响应计算,得到核电站系统设备重要的设计基础数据-楼层反应谱(FRS),并将计算的楼层反应谱同设计谱进行比较,从而对设计方法及其结果进行评估,为电站的抗震设计裕量评估和安全管理提供可资参考的结论.     

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.     

核电厂管线上设备地震加速度响应放大系数评估方法与应用

为了方便快捷地确定核电厂管线上设备的地震载荷,基于弹簧-质量模型、管线与安装于管线上的刚性设备的关系,提出了评估管线对设备的地震加速度响应放大系数的方法,以及该评估方法在确定设备抗震鉴定地震输入方面的应用方法,并通过有限元模型对该评估方法和应用予以了验证。结果表明,该评估方法在确定管线上设备抗震鉴定地震输入方面提供了易于操作又不至于太过保守而降低经济性的可行的应用办法。      

Seismic isolation in New Zealand

Bridges, buildings, and industrial equipment can be given increased protection from earthquake damage by limiting the earthquake attack through seismic isolation. A broad summary of the seismic responses of base-isolated structures is of considerable assistance for their preliminary design. Seismic isolation as already used in New Zealand consists of a flexible base or support combined with some form of energy-dissipating device, usually involving the hysteretic working of steel or lead. This paper presents examples of the New Zealand experience, where seismic isolation has been used for 42 bridges, 3 buildings, a tall chimney, and high-voltage capacitor banks.Additional seismic response factors, which may be important for nuclear power plants, are also discussed briefly.     

Probabilistic cost-benefit seismic design criterion for a dedicated shutdown heat removal system

A probabilistic methodology is developed for assessing the risk reduction potential and cost-benefit tradeoff of a Dedicated Shutdown Heat Removal System (DSHRS) for a PWR as a function of its seismic design capability. The option of coping with a very small LOCA is included. The annual seismic risk of a plant and a similar hypothetical plant having a proposed DSHRS with various seismic strengths are computed. The difference in the annual seismic risks is the annual seismic risk reduction benefit for having the system. The present value of the future risk reduction benefit is then compared to the cost of building a DSHRS and the incremental seismic cost associated with building the system to withstand a stronger earthquake.A reactor like Zion was used for application of the method due to the availability of data. Studies were performed to investigate the sensitivity of the results to the assumed seismic hazard, probability of occurrence of seismic-induced accident initiating events, equipment seismic fragility, accident costs, and discount rate. The incremental seismic risk reduction benefit computed in these studies ranges from $207 million for a DSHRS with a median seismic capacity of 1.70g (i.e. 10 × SSE) in a new plant built at the site. The total cost of a DSHRS is crudely estimated to be $25 million or more, if it were built to withstand the current SSE of the plant (for which the system probably would have a median seismic capacity of 0.85g or more due to various design and construction conservatisms). The cost associated with the seismic design aspect of such a system is estimated to be approximately $2.5 million and it may be doubled if the seismic design capability of the system is tripled. The cost/benefit results and their inherent large uncertainties are not definitive but indicate that probabilistic seismic design of a DSHRS should be examined in further detail.     

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.     

核化工厂房设备室企口形盖板受力分析及设计建议

为满足辐射防护要求，核化工厂房设备室顶板需大量使用企口形盖板。在盖板设计中应考虑竖向荷载、地震荷载、检修荷载等各方面因素。本文通过有限元软件ANSYS建立T形及Z形盖板模型并进行了计算，分析了不同工况下企口形盖板的承载力和位移分布规律，为设备室盖板的设计、施工提出了几点建议。