标准编号:	HAD 102/02-2019
中文名称:	核动力厂抗震设计与鉴定
英文名称:	Seismic Design and Verification of Nuclear Power Plants
行业分类:	HAD    核电行业标准
发布机构:	国家核安全局
发布日期:	2019-12-31
实施日期:	2019-12-31
标准状态:	现行
替代旧标准:	HAD 102/02-1996 核电厂的抗震设计与鉴定 HAD 102/02-1996 核电厂的抗震设计与鉴定
翻译语言:	英文
文件格式:	PDF
中文字符数:	26000 字
翻译价格(元):	4000.00 元
交付时间:	付款后 1 个工作日

1 Introduction



1.1 Purpose



1.1.1 This guide is to explain and refine the relevant provisions of the Regulations on Safety of nuclear power plants: Design (HAF 102, hereinafter referred to as the Regulations) for the purpose of providing acceptable general methods for nuclear safety supervision and management departments, nuclear power plant designers and operating units on design and qualification of nuclear power plants, so that the ground motion at the site will not endanger the safety of nuclear power plants, and giving guidance on the consistency of methods and procedures used for analysis and test qualification of structures and equipment, making them meet the requirements of the Regulations.



1.1.2 Annex I has the equal force as the text.



1.2 Scope



1.2.1 This guide is applicable to the design of land-based fixed water-cooled reactor nuclear power plants that meet the exclusion criteria of relevant guides for seismic risk assessment of nuclear power plants, so as to resist site-specific earthquakes. This guide does not involve the intensity of ground motion or the risk degree of SSCs in nuclear power plants.



1.2.2 When simplified procedures for design and verification are adopted, it is necessary to prove the suitability of these procedures for achieving safety objectives, and properly assess them from the perspective of safety.



1.2.3 This guide is applicable to the design and construction of newly built nuclear power plants, and is usually not used for re-assessment of built nuclear power plants. This guide is not applicable to the assessment of seismic design margin of built nuclear power plants.



1.2.4 This guide may also be used for the design of other types of nuclear power plants, but its applicability shall be assessed by engineering judgment according to the type of reactor and its special safety requirements.



1.2.5 The technical suggestions on modeling and qualification of SSCs in this guide may be applied to the design of vibration caused by other reasons than earthquakes, such as explosion of industrial facilities, aircraft impact and quarry explosion or accidents of high-speed rotating machinery. However, this kind of extended application shall be treated with caution, especially in terms of frequency range, duration, direction of induced vibration and influence mechanism on nuclear power plants, which shall be subjected to engineering judgment. It shall also be noted that the design for resisting such loads may be of different forms (such as anti-collision walls), or may include other different failure forms (such as scab or breakage caused by impact loads). Such special engineering measures are not considered in this guide.



 

2 General



2.1 Overview



2.1.1 In this clause, suggestions on seismic category are put forward according to the requirements in the Regulations, based on the safety importance of structure, system and component (SSC) in the design basis seismic events. In order to ensure proper safety margin in the design, some suggestions on the application of design standard are also given.



2.1.2 For SSC, services and processes that affect the safety of nuclear power plants covered by the applicable scope of this guide, quality assurance measures shall be formulated and effectively implemented.



2.2 Design basis earthquake



2.2.1 The seismic risk of each site shall be assessed, and two levels of design basis ground motions, namely, operational safety ground motion (SL-1) and ultimate safety ground motion (SL-2), shall be given according to relevant procedures and target probability levels or principles determined by the design of nuclear power plants.



2.2.2 In the design of nuclear power plant, SL-2 is related to the most stringent safety requirements, while SL-1 has different safety meanings, which is more likely and less serious, and can be determined by the operating unit through comprehensive assessment. Generally, SL-1 is used for load combination (for reasons related to probability, other events are combined with earthquakes with lower intensity), post-accident inspection and permit requirements. As a low-level ground motion, SL-1 is usually not related to safety requirements, but only related to operational requirements. When the actual ground motion at the site during the operation of the nuclear power plant exceeds SL-1, measures shall be taken to shut down the reactor; the safety-related SSCs of the nuclear power plant shall be assessed according to relevant requirements, and the operation of the nuclear power plant shall not be resumed until being examined and approved by the nuclear safety supervision and administration department.



2.2.3 SL-2 shall be considered for each safe-class SSC of nuclear power plant. The lowest level shall be considered as 0.15g (the value of zero period acceleration in design response spectrum), which is equivalent to the peak value of ground acceleration in free field.



 

2.2.4 The frequency spectrum and duration of potential ground motion are generally considered in the determination of design basis ground motion. When it is judged that there are multiple sources that have made major contributions to the hazard, special attention shall be paid to the influence of spectrum effect and duration of different sources. In this case, more caution shall be given when enveloping the ground motions (or response spectrum) originating from different seismic sources (such as far-field and near-field ones). Considering the different seismic requirements of SSCs, the bearing capacity of different ground motions should be assessed.



2.2.5 It is generally defined that the input ground motion happens in the free field at the surface of the earth or bedrock. When seismic input is needed at the foundation elevation, inversion-forward method may be used to assign values.



2.3 Seismic category for SSCs



2.3.1 Any major site expected effect caused by earthquake is related to vibration transmitted to SSCs through nuclear power plant structures. Vibration may affect the safety function of nuclear power plant through direct or indirect interaction mechanism (such as mechanical interaction between SSCs caused by earthquake, release of dangerous substances, fire or flooding, destruction of operator passage and unavailability of evacuation road or approach road).



2.3.2 All SSCs shall stand up to any possible seismic action, and the required performance during seismic events may be different from the safety functions considered in safety classification. These safety functions are based on the most demanding safety functions under all design basis conditions (postulated initiating event). Therefore, for safety-oriented design methods, besides safety classification, SSCs shall be classified according to their safety importance during and after earthquakes. Seismic resistance of SSCs can be divided into seismic category I, seismic category II and non-nuclear seismic category, or more other categories according to the design characteristics of nuclear power plant units. The purpose of classification is to help protect the public and the environment from the release of radioactive substances and ensure nuclear safety.



2.3.3 Seismic category I SSCs of nuclear power plants shall be specified, which shall be such designed that SL-2 can be borne. Seismic category I SSCs usually correspond to the highest safety category and include all SSCs important to safety. Specifically, seismic category I SSCs shall include the following ones and their supporting structures:



(1) SSCs that directly or indirectly cause an accident condition in case of failure as a result of SL-2;



(2) SSCs necessary for shutting down the reactor, keeping the reactor in a shutdown state, discharging waste heat within the required period, and those necessary for monitoring the parameters of the above functions;



(3) SSCs necessary to prevent or mitigate the radioactive release excessing the limit caused by any postulated initiating event (regardless of its occurrence probability) considered in the design;



(4) SSCs needed to prevent or mitigate unacceptable radioactive release consequences of spent fuel pool.



2.3.4 In 2.3.3 (3), the selection of SSCs is related to defense in depth: in an seismic event of SL-2, all levels of defense shall always be available . Physical barriers designed to defend against external events other than earthquakes shall maintain integrity and functionality during earthquakes.



2.3.5 Although the main pressure boundary of the primary circuit of light-water reactor is designed according to the borne seismic loads, as a conservative measure, it is still assumed that some design basis accidents will occur at the primary circuit pressure boundary, and SSCs to mitigate the consequences will be set up, which also belongs to the seismic category I SSCs.



2.3.6 The design, installation and maintenance of seismic category I SSCs in nuclear power plant shall meet the strictest practice, that is, higher safety margin is adopted for facilities with conventional risks. For any seismic category I SSC, appropriate acceptance criteria (such as design parameters indicating functionality, tightness or maximum deformation) shall be determined according to safety function requirements. However, in some cases, in case of assessing its impact on the safety function of nuclear power plants in detail, the acceptance criteria of physical barriers may be appropriately lowered for load combinations applicable to SL-2.



2.3.7 Seismic category II SSCs of nuclear power plant can be determined. Seismic category II SSCs shall include:



(1) all SSCs with radioactive risk but unrelated to the reactor (such as spent fuel plant building and radioactive waste plant building). These SSCs are required to have safety margins consistent with their potential radiological consequences. As these SSCs are generally related to different release mechanisms (such as waste leakage and damage of spent fuel cask), the expected consequences are different from the potential consequences of the reactor;



(2) SSCs that do not belong to seismic category I, especially those in 2.3.3 (2) and (3), but are needed to prevent or mitigate accident conditions (caused by postulated initiating events other than earthquakes) of nuclear power plants within a long enough time (there is a possibility of SL-2 or SL-1 occurring reasonably during this period);



(3) SSCs related to site accessibility and those required for implementing emergency evacuation plan.



 

2.3.8 The design earthquake level of seismic category II SSCs shall be determined on the following basis: the additional work done to protect the SSCs against this earthquake level must be commensurate with the risk that may be reduced for the nuclear power plant personnel or the public from the earthquake. The acceptable limits for the release of radioactive substances prescribed by the state must be followed.



2.3.9 Non-nuclear seismic category SSCs that do not belong to seismic categories I and II shall be designed according to the national non-nuclear facility specifications, i.e. facilities with conventional risks. For some of the SSCs that are important to the operation of nuclear power plants, stricter acceptance criteria may be selected according to the operation objectives. This approach can reduce the need to shut down, inspect and re-apply for a license, thus keeping the nuclear power plant operating continuously.



2.3.10 Among all SSCs of the nuclear power plant (including those that are not important to safety), for those that may have spatial interaction (such as due to collapse, fall or displacement) or other interactions (such as interaction caused by release of hazardous substances, fire, flooding or earthquake) with seismic categories I and II SSCs, the potential impact and damage caused by these SSCs shall be demonstrated that they won’t either affect the safety function of any seismic category I and II SSC, nor operator’s actions important to safety.



2.3.11 As a consequence of earthquake, according to analysis, test or experience, if it is expected that some interactions will occur and endanger the functions (including operation actions) of seismic category I or II SSCs, one of the following measures shall be taken:



(1) this kind of SSCs shall be reclassified as seismic category I or II, and redesigned;



(2) in order to avoid adverse effects on seismic category I or II SSCs, this kind of SSCs shall be subject to qualification according to SL-2;



(3) the endangered seismic category I or II SSCs shall be properly protected so as to prevent their functions from being damaged due to interaction with this kind of SSCs.



2.3.12 The SSCs mentioned in 2.3.10 shall be designed, installed and maintained according to the nuclear application practice.



However, in 2.3.11 (2), when it is considered that the frequency of interaction with seismic category I or II SSCs is very low, the safety margin may be appropriately lowered.



2.3.13 The seismic category of SSCs shall be based on a clear understanding of the functional requirements for ensuring safety during or after earthquakes. According to different safety functions, different components in the same system may belong to different seismic categories, for example, the aspects shall be considered on tightness, damage degree (such as fatigue, wear and cracking), mechanical or electrical functions, maximum displacement, permanent deformation degree and maintenance of geometric dimensions, etc.



2.3.14 Seismic loads shall be considered for all possible operation modes of nuclear power plants. In the seismic design, the seismic category of the designed SSCs shall be considered.



2.3.15 Seismic category shall be carried out according to reactor type, nuclear safety regulations and standards, and special boundary conditions of the site (such as availability of cooling water source), etc.



2.3.16 As part of the design process, a detailed list of all SSCs with relevant acceptance criteria shall be listed. See Annex I for the example list.

1 Introduction
2 General
3 Seismic design
4 Equipment seismic qualification
5 Seismic instruments
Annex I Examples of seismic category

1 引言
1.1 目的
1.1.1 本导则是对《核动力厂设计安全规定》（HAF102， 以下简称《规定》）有关条款的说明和细化，其目的是给核安 全监督管理部门、核动力厂设计人员和营运单位就核动力厂 设计与鉴定提供可接受的通用方法，使场址地震动不致危及 核动力厂安全，并在构筑物和设备的分析、试验鉴定所用方 法和程序的一致性方面给予指导，使其满足《规定》的安全 要求。
1.1.2 附件Ⅰ与正文具有同等效力。
1.2 范围
1.2.1 本导则适用于符合核动力厂地震危险性评价相关 导则排除准则的陆上固定式水冷反应堆核动力厂的设计，以 抵御场址特定地震。本导则不涉及地震动的强度或核动力厂 各物项的风险度。
1.2.2 当采用简化程序进行设计和验证时，应证明这些 程序对于实现安全目标的适宜性，并从安全的角度进行恰当 的评价。
1.2.3 本导则适用于新建核动力厂的设计与建造，通常 不用于对已建核动力厂的重新评价。本导则不适用于已建核 动力厂的抗震设计裕度评价。
1.2.4 本导则也可用于其他类型核动力厂的设计，但应 根据反应堆类型及其特殊的安全要求，采用工程判断的方法 评价其适用性。
1.2.5 本导则中关于模型化与物项鉴定方面的技术建议 可应用于地震以外其他原因引发振动的设计，如工业设施的 爆炸、飞机撞击、采石场爆炸或高速旋转机械的事故等。但 是，对于此类扩展应慎用，尤其是关于诱发振动的频率范围、 持续时间、方向和对核动力厂的影响机理等方面，应进行工 程判断。还应注意到，抵御此类荷载的设计可采用不同的形 式（如防撞墙），或可能包括其他不同的破坏形式（如冲击荷 载引起的结痂或破碎）。本导则不考虑这些特殊的工程措施。

2 总则
2.1 概述
2.1.1 本章依据《规定》中的要求，按构筑物、系统和 部件在设计基准地震事件中的安全重要性，提出抗震分类的 建议。为保证在设计中有适当的安全裕度1，还给出了关于设 计标准应用的建议。
2.1.2 对于本导则适用范围内所涵盖的影响核动力厂安 全的物项、服务和过程，应制定质量保证措施并有效实施。

1 本文中，安全裕度是指在设计、材料选择、建造、维修和质量保证中的特殊条 款的结果。

2.2 设计基准地震
2.2.1 对每个场址应评定其地震危险性，并根据相关程 序及核动力厂设计确定的目标概率水平或原则，给出两个级 别的设计基准地震动：运行安全地震动（SL-1）和极限安全 地震动（SL-2）。
2.2.2 在核动力厂的设计中，SL-2 与最严格的安全要求 相关，而 SL-1 则具有不同的安全意义，其可能性较大且严重 性较低，可由营运单位经综合评估确定。通常，SL-1 用于荷 载组合（由于与概率相关的原因，其他事件与较低强度的地 震组合）、事故后的检查及许可证要求。作为较低水平的地震 动，SL-1 通常不与安全要求相关，只与运行要求相关。当核 动力厂运行中场址实际发生的地震动超越 SL-1 时，应采取措 施停堆，并应依据相关要求对核动力厂安全相关物项进行评 估，经过核安全监督管理部门的审查认可后，方可恢复核动 力厂运行。
2.2.3 对核动力厂每个安全级物项均应考虑 SL-2。最低 水平应考虑相当于自由场地面加速度峰值 0.15g（设计反应谱 中零周期加速度的值）。
2.2.4 设计基准地震动的确定一般考虑潜在地震动的频 谱及持续时间。当判定有多个震源对危险性具有主要贡献时， 尤其应注意不同震源的频谱效应与持续时间的影响。在此情 况下，对源于不同震源（如远场和近场）的地震动（或反应谱）进行包络时应更加谨慎。考虑到构筑物、系统和部件的 抗震要求不同，宜对不同的地震动分别进行承载力评价。
2.2.5 输入地震动一般定义在地表或基岩表面处的自由 场。当需要在基础标高处进行地震输入时，可采用反演-正演 的方法来赋值。
2.3 构筑物、 系统和部件的抗震分类
2.3.1 由地震引起的任何主要的场址预期效应，与通过 核动力厂构筑物传至构筑物、系统和部件的震动相关。震动 可通过直接或间接相互作用机制（如由地震引发的物项间的 机械相互作用、危险物质的释放、火灾或水淹、操作人员通 道的破坏以及撤离道路或进场道路的不可用等）影响核动力 厂的安全功能。
2.3.2 所有构筑物、系统和部件都要经受任何可能发生 的地震作用，而地震事件发生时所要求的性能可以不同于在 安全分级中考虑的安全功能。这些安全功能是基于在所有设 计基准工况下（假设始发事件）要求最高的安全功能。因此 对于从安全出发的设计方法，除了安全分级以外，还要根据 其在地震期间和地震后的安全重要性将构筑物、系统和部件 进行分类。构筑物、系统和部件的抗震可分为抗震Ⅰ类、抗 震Ⅱ类和非核抗震类，或根据核动力厂机组的设计特性分为 更多类。分类的目的是为了有利于公众和环境对放射性物质 释放的防护和保障核安全。

2.3.3 应规定核动力厂的抗震Ⅰ类物项。此类物项应设计 为可承受 SL-2。抗震Ⅰ类物项通常相应于安全上的最高类别， 并包括所有安全重要物项。具体来说，抗震Ⅰ类物项应包括下 列物项及其支承结构：
（1）作为 SL-2 的后果，其失效会直接或间接导致事故 工况的物项；
（2）使反应堆停堆，保持反应堆处于停堆状态，在要求 期间内排出余热所需的物项，以及对上述功能的参数进行监 测所必需的物项；
（3）预防或缓解设计中考虑的任何假设始发事件（不论 其发生的概率如何）引起的放射性释放超过限值所必需的物 项；
（4）预防或缓解乏燃料池不可接受的放射性释放后果所 需的物项。
2.3.4 在 2.3.3 节（3）中物项的选取与纵深防御有关： 在 SL-2 水平的地震事件中，所有层次的防御应总是处于可用 状态2。为防御地震以外的外部事件所设计的实体屏障，在地 震期间应保持完整性和功能性。
2.3.5 尽管轻水堆一回路主要压力边界是按承受地震荷 载进行设计的，但作为一种保守措施，仍假设在一回路压力 边界会发生某些设计基准事故而设臵了减轻其后果的物项，

2 在纵深防御的框架中，对所有外部事件的防御是第一层次纵深防御的一部分。

这类物项也要包括在抗震Ⅰ类物项中。
2.3.6 核动力厂抗震Ⅰ类物项的设计、安装与维修应符 合严格的实践，即应高于常规风险的设施所采用的安全裕度。 对于任何抗震Ⅰ类的物项，应按照安全功能要求确定适当的 验收准则3（如表明功能性、密封性或最大变形的设计参数）。 但是在某些情况下，如果详细评价其对核动力厂安全功能的 影响，对于包含 SL-2 的荷载组合，实体屏障的验收准则可以 适当降低。
2.3.7 可确定核动力厂的抗震Ⅱ类物项。抗震Ⅱ类物项 应包括：
（1）所有具有放射性风险但与反应堆无关的物项（如乏 燃料厂房和放射性废物厂房）。要求这些物项具有的安全裕度 与其潜在放射性后果相一致。由于这些物项一般来说与不同 的释放机理有关（如废物泄漏、乏燃料筒损坏），其预期后果 与反应堆的潜在后果不同；
（2）不属于抗震Ⅰ类„特别是 2.3.3 节（2）和（3）中 的物项‟，但在足够长的时间内（在该时间段内具有合理地发 生 SL-2 或 SL-1 的可能性）预防或缓解核动力厂事故工况（由 地震以外的假设始发事件引起的）所需要的物项；
3 验收准则是对评价构筑物、系统和部件执行其设计功能的能力所用的功能性的 或状态性指标而规定的边界值。此处所用的验收准则是指在所定义的假想初始事 件下，对构筑物、系统和部件功能性的或状态性的指标规定的边界值（如：与功 能性、密封性或无相互作用相关的指示）。
（3）与场址可达性相关的物项及实施应急撤离计划所需 的物项。
2.3.8 抗震Ⅱ类物项的设计地震水平应在以下基础上确 定：为保护物项防御这一地震水平所做的附加工作必须与可 能减轻核动力厂人员或公众遭受地震引起的风险相称。必须 遵守国家规定的放射性物质释放可接受的限值。
2.3.9 不属于抗震Ⅰ类、抗震Ⅱ类的非核抗震类物项应 依据国家非核设施的规范进行设计，即按常规风险的设施进 行设计。其中的一些对核动力厂运行重要的物项，可根据运 行目标选择较严格的验收准则。这种做法可减少核动力厂停 堆、检查和重新申请许可证的需要，从而使核动力厂持续运 行。
2.3.10 在核动力厂所有物项中（包括那些安全上不重要 的物项），那些可能与抗震Ⅰ类和抗震Ⅱ类物项发生空间相互 作用（如由于倒塌、坠落或移位）或其他相互作用（如通过 危险物质释放、火灾、水淹或地震引起的相互作用）的物项， 应论证这些物项引起的潜在影响和造成的损害，既不影响任 何抗震Ⅰ类及抗震Ⅱ类物项的安全功能，也不影响任何与安 全相关的操纵员行动。
2.3.11 作为地震后果，根据分析、试验或经验，预计会 发生某些相互作用，并且会危及抗震Ⅰ类或抗震Ⅱ类物项的 功能（包括操作行动）时，应采取下述措施之一：
（1）这类物项应重新分类为抗震Ⅰ类或抗震Ⅱ类，并重 新进行设计；
（2）为了避免对抗震Ⅰ类或抗震Ⅱ类物项产生不利影 响，这类物项应按 SL-2 进行鉴定；
（3）应适当地保护被危及的抗震Ⅰ类或抗震Ⅱ类物项， 以免其功能受到与此类物项相互作用的危害。
2.3.12 第 2.3.10 节所述物项应按照核应用实践进行设计、安装和维修。但是，在第 2.3.11 节（2）中，当认为其 与抗震Ⅰ类或抗震Ⅱ类物项发生相互作用的频率非常低时， 可以适当降低安全裕度。
2.3.13 对物项的抗震分类，应以清楚地了解为保证安全 在地震期间或地震后对其功能的要求为基础。根据不同的安 全功能，同一系统中的不同部件可能属于不同的抗震类别， 例如应考虑密封性、损坏（如疲劳、磨损及开裂等）程度、 机械或电气功能、最大位移、永久变形的程度和几何尺寸的 保持等方面。
2.3.14 对核动力厂所有可能的运行模式都应考虑地震 荷载。在抗震设计中，对所设计的物项应考虑其抗震分类。 2.3.15 应依据反应堆类型、核安全法规和标准以及场址 的特殊边界条件（如冷却水源的可用性）等进行抗震分类。 2.3.16 作为设计过程的一部分，应列出具有相关验收准则的所有物项的详细清单。示范的清单见附件Ⅰ。