标准编号:	GB/T 41090-2021
中文名称:	能动安全系统压水堆核电厂总设计要求
英文名称:	General design requirements of pressurized water reactor nuclear power plants with active safety systems
中标分类:	F65    核电厂核岛
行业分类:	GB    国家标准
ICS分类:	27.120.20 27.120.20    核电站、安全 27.120.20
发布机构:	国家市场监督管理总局 国家标准化管理委员会
发布日期:	2021-12-31
实施日期:	2022-7-1
标准状态:	现行
翻译语言:	英文
文件格式:	PDF
中文字符数:	25000 字
翻译价格(元):	1750.00 元
交付时间:	付款后 1~5 个工作日

1 Scope

This document specifies the basic requirements for the overall design of an energetic safety system pressurised water reactor nuclear power plant (hereinafter referred to as "nuclear power plant") to ensure that it can operate safely and reliably.

This document is applicable to newly built pressurised water reactor nuclear power plants with energy safety systems, and in-service pressurised water reactor nuclear power plants with energy safety systems may refer to the implementation.

2 Normative references

The contents of the following documents constitute essential provisions of this document through the normative references in the text. Among them, the date of the cited documents, only the date of the corresponding version applicable to this document; do not note the date of the cited documents, the latest version (including all the revision of the list) applicable to this document.

GB 6249 Environmental radiation protection regulations for nuclear power plants

GB 11806 Regulations for the safe transport of radioactive substances

GB 18871 Basic standards for ionising radiation protection and safety of radiation sources

3 Terminology and definitions

The following terms and definitions apply to this document.

3.1

safety system

A safety critical system for ensuring the safe shutdown of a reactor, the discharge of waste heat from the core and the consequences of anticipated operational events and design basis accidents.

[Source: HAF102-2016, Glossary].

3.2

active safety system

An energetic system used to ensure the safe shutdown of a reactor, to discharge waste heat from the core or to limit the consequences of anticipated operational events and design basis accidents.

[Source: GB/T35730-2017,3.3, with modifications]

3.3

pressurized water reactor nuclear power plants with active safety systems

Pressurized water reactor nuclear power plants with active safety systems for all safety functions within the design basis accident.

Note: Abbreviated as "energetic nuclear power plant".

[Source: GB/T35730-2017, 3.4, with modifications]

3.4

Safety function

To ensure that the facility or activity is capable of preventing and mitigating the radiological consequences of the normal operation of a nuclear power plant, anticipated operational transients and accidental operating conditions.

4 Design objectives for nuclear power plants

4.1 Nuclear safety design objectives

4.1.1 Basic safety objectives

Basic safety objective:To establish and maintain effective defences against radiological hazards in nuclear power plants in order to protect people and the environment from radiological hazards.

In order to achieve the basic safety objective, the following measures shall be taken:

a) Control of radiation exposure to personnel and the release of radioactive material to the environment during operation;

4.2 General economic objectives

Nuclear power plants need to be designed in such a way that the construction costs and whole life generation costs of nuclear power plants are sufficiently competitive in the market.

The main factors that need to be considered in the design of a nuclear power plant that have an impact on the economic objectives include the design life of the plant, the overall construction cycle of the plant, the average plant availability, the level of unplanned plant outages, the replacement cycle, and the optimisation of energy utilisation taking into account the environmental conditions of the site.

5 General design requirements for nuclear power plants

5.1 Longitudinal defence design

5.1.1 General requirements for the application of defence-in-depth design

Nuclear power plants should be designed with defence-in-depth measures to improve multi-level defence (inherent characteristics, equipment and protocols). In order to prevent possible harmful effects on people and the environment, a balanced safety concept of prevention and mitigation should be implemented to ensure that the consequences of an accident can be mitigated by taking appropriate mitigation measures to protect people and the environment in the event of a protection failure. Each separate effective level of defence is a fundamental part of the defence-in-depth of a nuclear power plant and should ensure that safety-related activities can be incorporated into a separate level of defence-in-depth.

The defence-in-depth concept and the principles of defence-in-depth application in design are in accordance with the requirements of the five levels of defence-in-depth specified in HAF 102.

5.1.2 Design for independence of defence in depth

The various levels of defence in depth should be as independent as possible from each other, so that the failure of one level of defence does not reduce the effectiveness of the other levels. In particular:

a) Safety facilities designed to mitigate the consequences of a core meltdown are, as far as practicable, independent of those designed to mitigate a design basis accident.

a) Safety facilities designed to mitigate the consequences of a core meltdown are, as far as practicable, independent of those used to mitigate the design basis accident.

5.2 Nuclear power plant operating conditions and safety analysis

5.2.1 Nuclear power plant operating conditions

Nuclear power plant operating conditions are classified as normal operation, expected operating events, design basis accidents and design extensions.

The specific principles and list of conditions for each category can be implemented in accordance with NB/T20035. At the same time, a set of design expansion conditions (including design expansion conditions that do not cause significant core damage and core meltdown design expansion conditions) should be derived based on engineering judgement, deterministic and probabilistic evaluation.

5.2.2 Deterministic safety analysis

A deterministic safety analysis shall be performed during design and shall cover all nuclear power plant operating conditions as specified in 5.2.1. The purpose of the deterministic safety analysis is to confirm that

a) the safety functions can be performed reliably;

b) the necessary structures, systems and components, in combination with operator actions, are sufficient to ensure that the release of radioactive material from the nuclear power plant is below acceptable limits and has a suitable safety margin.

Deterministic safety analyses are required to demonstrate that the radioactive barriers of a nuclear power plant maintain their integrity within the required limits. Deterministic safety analysis, when supplemented by probabilistic safety analysis, should also help to demonstrate that:

a) the source term and potential radiological consequences are acceptable under different nuclear power plant operating conditions;

b) that specific conditions leading to early releases or significant releases can be considered as "practically eliminated".

The specific analytical objectives, analytical methods, steep-edge effects and non-discretionary effects of different nuclear power plant conditions for definitive safety analysis. Analytical assumptions, treatment of steep-sided effects and uncertainties, and acceptance guidelines can be implemented in conjunction with the requirements for deterministic safety analyses in the guidelines.

The radioactivity acceptance criteria for different nuclear power plant conditions should meet the requirements of GB 6249 and realistic models and best estimation methods can be used to evaluate the design expansion conditions.

5.2.3 Probabilistic safety analysis

In the design of a nuclear power plant, a probabilistic safety assessment of the nuclear power plant should be completed in order to

a) provide a systematic analysis of the need for the design to properly consider all operating modes and all states of the nuclear power plant (including shutdown conditions) and to compare the results of the analysis with the risk guidelines that have been specified;

b) demonstrate that the overall design is balanced, without any - facility or hypothetical initiating events (hypothetical initiating events, which are based on

deterministic or probabilistic approach or a combination of both) would make an excessive or clearly uncertain contribution to the total risk, and to ensure that the depth of

The contribution of the first and second levels of defence in depth to nuclear safety should be ensured;

Confirmation that small deviations in plant parameters do not cause serious abnormalities in plant performance (steep edge effects);

d) provide a safety assessment of the probability of a severe core damage state and a significant release of radioactive material outside the plant that requires an early response from outside the plant (especially associated with early containment failure);

e) provide an assessment of the frequency and consequences of external hazards (particularly those external events specific to the nuclear power plant site);

f) Identify systems that may reduce the probability of core meltdown accidents or mitigate their consequences through design improvements or modifications to operating procedures;

g) Evaluation of the adequacy of emergency procedures for nuclear power plants.

h) Nuclear power plant designs may refer to NB/T 20037 (all parts) for probabilistic safety evaluation.

5.4 Reliability design requirements

Safety critical structures, systems and components shall be designed in accordance with the latest or currently applicable codes and standards; their design shall have been previously verified under equivalent conditions of use.

Single-failure criteria shall be applied to each safety combination included in the design of a nuclear power plant. The principles for the application of the Single Failure Criteria are as specified in GB/T 13626 and the requirements for the application of the Single Failure Criteria to safety critical fluid systems are as specified in NB/T 20402.

The design needs to consider the possibility of common cause failures in safety critical items to determine how the required reliability should be achieved with the principles of diversity, multiplicity and independence.

6 General design requirements for specialist areas

6.1 General layout requirements

A nuclear power plant building includes the nuclear island building. The turbine plant and the nuclear power plant supporting facilities plant. The nuclear island plant shall be divided according to function, including the reactor plant, fuel plant, nuclear auxiliary plant and electrical plant, etc.

The layout of the nuclear island should follow the following guidelines:

a) Highly radioactive areas should be as compact as possible;

b) The reactor plant should be located in the centre of the nuclear power unit;

c) Safety systems should be located as close as possible to the reactor building;

d) the area between the other nuclear island buildings and the reactor building needs to be as spacious as possible to allow for sufficient containment penetrations;

6.2 Reactor core design

6.2.1 Design margins

The reactor core and associated coolant systems, control and protection systems should be designed with appropriate margins to ensure that the specified acceptable fuel design limits are not exceeded and that radiation safety standards are met under any operational and accident conditions.

6.2.2 Reactor nuclear design

The nuclear design of the reactor shall provide for fuel assemblies within the core. It should provide sufficient residual reactivity and control means to determine the power distribution, fuel consumption distribution and reactivity coefficient of the core to meet safety requirements.

Adequate means should be available to detect the distribution of neutron injection rates in the reactor core and their variation.

The reactor core and associated cooling system should be designed to have a negative power reactivity coefficient under all operating and accident conditions, with the net effect of the instantaneous nuclear feedback inherent in the reactor compensating for the rapid increase in reactivity.

The minimum and maximum limits of the reactivity coefficient are a function of various parameters (e.g. power level, boron concentration. The reasonableness of the reactivity factor envelopes used for the analysis of various operating and accident conditions should be confirmed by appropriate studies.

The reactor core and associated coolant, control and protection systems should be designed in such a way that power oscillation conditions exceeding the specified fuel design limits are unlikely to occur or, if they do occur, can be reliably and rapidly monitored and suppressed.

The basic requirements for the nuclear design of the reactor core can be found in NB/T 20057.1.

6.3 Reactor coolant system pressure boundary design

The reactor coolant system and associated auxiliary, control and protection systems should be designed with sufficient margins to ensure that the design conditions of the reactor cooling system pressure boundary are not exceeded during any normal operation including anticipated operational events.

The reactor coolant system should be protected against overpressure in all modes of operation, with particular attention to overpressure protection in cryogenic watertight conditions.

The design, manufacture, installation and testing of the reactor coolant system pressure boundary should ensure that the probability of abnormal leakage, rapid crack expansion and overall rupture is extremely low.

The design needs to take into account all conditions of the reactor coolant system pressure boundary material in operating conditions including maintenance, test conditions and accident conditions, as well as any uncertainties in determining the initial condition and possible rate of deterioration of the component after the expected effects of many factors such as erosion, creep, fatigue, chemical environment, radiation environment and ageing.

The design, manufacture and arrangement of the components of the reactor coolant system pressure boundary should facilitate adequate inspection and testing of the pressure boundary at regular intervals throughout the life of the nuclear power plant.

Specific reactor coolant system designs can be found in NB/T20187.

6.4 Design of containment systems

6.4.1 Functional design requirements for containment systems

The containment system shall be designed to ensure or contribute to the following safety functions:

a) containment of radioactive material during operational and accident conditions;

b) radiation shielding during operational and accident conditions;

c) defence against external natural and external man-made events.

The reactor containment and associated systems provide an inherent containment barrier to

Foreword
2 Normative references
3 Terminology and definitions
4 Design objectives for nuclear power plants
5 General design requirements for nuclear power plants
6 General design requirements for specialist areas

1范围
本文件规定了能动安全系统压水堆核电厂(以下简称“核电厂”)的总体设计基本要求,以确保其可以安全可靠地运行。
本文件适用于新建的能动安全系统压水堆核电厂,在役的能动安全系统压水堆核电厂可参考执行。
2规范性引用文件
下列文件中的内容通过文中的规范性引用而构成本文件必不可少的条款。其中,注日期的引用文件,仅该日期对应的版本适用于本文件;不注日期的引用文件,其最新版本(包括所有的修改单)适用于本文件。
GB 6249核动力厂环境辐射防护规定
GB 11806放射性物质安全运输规程
GB 18871电离辐射防护与辐射源安全基本标准
3术语和定义
下列术语和定义适用于本文件。
3.1
安全系统safety system
安全上重要的系统,用于保证反应堆安全停堆、从堆芯排出余热制预计运行事件和设计基准事故的后果。
[来源:HAF102-2016,名词解释]
3.2
能动安全系统active safety system
用于保证反应堆安全停堆、从堆芯排出余热或限制预计运行事件和设计基准事故后果的能动系统。
[来源:GB/T35730-2017,3.3,有修改]
3.3
能动安全系统压水堆核电厂
pressurized water reactor nuclear power plants with active safety systems
主要依赖能动安全系统完成设计基准事故内全部安全功能的压水堆核电厂。
注:简称“能动核电厂”。
[来源:GB/T35730-2017,3.4,有修改]
3.4
安全功能safety function
为了保证设施或活动能够预防和缓解核电厂正常运行、预计运行瞬态和事故工况下的放射性后果。
4核电厂设计目标
4.1核安全设计目标
4.1.1基本安全目标
基本安全目标:在核电厂中建立并保持对放射性危害的有效防御，以保护人与环境免受放射性危害。
为了实现基本安全目标，应采取以下措施:
a)控制在运行状态下对人员的辐射照射和放射性物质向环境的释放;
4.2总的经济目标
核电厂在设计中需要充分考忠核电厂的建造成本和全寿期发电成本，使其具有市场竞争力。
设计中主要需要考虑的对经济目标有影响的因素包括核电厂设计寿命、核电厂整体的建造周期、核电厂平均可利用率、核电厂非计划性停堆水平、换料周期、考虑厂址环境的条件使能最利用得到优化等。
5核电厂总体设计要求
5.1纵深防御设计
5.1.1应用纵深防御设计的总体要求
核电厂设计应采用纵深防御措施,以提高多层次防御(固有特性、设备及规程)能力。为预防可能对人员和环境产生的有害影响,应贯彻预防和缓解平衡的安全理念IC保证在防护失效的情况下可以通过采取适当的缓解措施减轻事故后果以保护人员和环境。每-独立有效层次的防御都是核电厂纵深防御的基本组成部分,应确保与安全相关的活动能够被纳人独立的纵深防御层次。
纵深防御概念与纵深防御在设计中的应用原则按照HAF102中规定的纵深防御的五个层次的要求执行。
5.1.2纵深防御独立性设计
纵深防御各个层次之间应尽实际可能地相互独立,避免-一个层次防御的失效降低其他层次的有效性。特别地:
a)设计用于减轻堆芯熔化事故后果的安全设施尽实际可能独立于用于减轻设计基准事故的
设备。
5.2核电厂工况与安全分析
5.2.1核电厂工况
核电厂工况划分为正常运行、预计运行事件、设计基准事故和设计扩展工况。
每类工况的具体划分原则与工况清单可按照NB/T20035执行。同时,应在工程判断、确定论和概率论评价的基础上得出一套设计扩展工况(包括没有造成堆芯明显损伤的设计扩展工况、堆芯熔化设计扩展工况)。
5.2.2确定论安全分析
设计中应进行确定论安全分析,并应涵盖5.2.1中规定的所有核电厂工况。确定论安全分析的目的是确认:
a)安全功能能够可靠地执行;
b)必要的构筑物、系统和部件,结合操纵员动作，足够保证核电厂放射性物质释放低于可接受限值,且具有合适的安全裕度。
确定论安全分析需要证明核电厂放射性屏障在所要求的范围内保持其完整性。确定论安全分析以概率安全分析作为补充后,也应有助于证明:
a)在不同核电厂工况下,源项和潜在的放射性后果是可接受的;
b)导致早期放射性释放或大量放射性释放的特定工况可被认为“实际消除”。
不同核电厂工况的确定论安全分析特定分析目标、分析方法.分析假设、陡边效应与不确定性处理以及验收准则,可结合导则中确定论安全分析的相关要求执行。
不同核电厂工况的放射性验收准则应满足GB6249的规定,可采用现实模型和最佳估算方法来评价设计扩展工况。
5.2.3概率安全分析
在核电厂的设计中,应完成核电厂的概率安全评价,以达到下述目的:
a)提供系统性的分析,设计中需要适当考虑核电厂所有运行模式和所有状态(包括停堆工况),并将分析结果和已规定的风险准则进行比较;
b)证明整个设计是平衡的，没有任何--个设施或假设始发事件(假设始发事件,这些事件是根据
确定论方法或概率论方法或这两者的组合选定的)对于总的风险会有过大的或明显不确定的
贡献,并且保证纵深防御的第一和第二层次承担核安全的主要责任;
确认核电厂参数的小偏离不会引起核电厂性能严重异常(陡边效应);
d)提供发生堆芯严重损伤状态以及要求厂“外早期响应的(特别是与安全壳早期失效相关的)放射性物质向厂外大量释放的概率安全评价;
e)提供外部危险(特别是核电厂厂址特有的那些外部事件)发生频率和后果的评价;
f)鉴别出通过设计改进或运行规程的修改可能降低堆芯熔化事故概率或减轻其后果的系统;
g)评价核电厂应急规程的充分性。
h)核电厂设计可参考NB/T 20037(所有部分)进行概率安全评价。
5.4可靠性设计要求
安全重要构筑物、系统和部件应按照最新的或当前适用的规范和标准进行设计;其设计应是此前在相当使用条件下验证过的。
应对核电厂设计中所包括的每个安全组合都应用单一故障准则。单一故障准则的应用原则参考GB/T 13626的规定,安全重要流体系统应用单一故障准则的要求参考NB/T 20402的规定执行。
设计中需考虑安全重要物项发生共因故障的可能性,以确定应该如何以多样性、多重性、独立性原则来实现所需的可靠性。
6专业领域总体设计要求
6.1总体布置要求
一台核电机组厂房包括核岛厂房.汽轮机厂房和核电厂配套设施厂房。核岛厂房应按照功能划分，包括反应堆厂房、燃料厂房、核辅助厂房和电气厂房等。
核岛布置应遵循如下准则:
a)高放射性区需尽可能紧凑;
b)反应堆厂房需布置在核电机组的中心;
c)安全系统需尽可能设置在靠近反应堆厂房的位置;
d)其他核岛厂房与反应堆厂房连接区需尽可能宽敞些,以布置足够的安全壳贯穿件;
6.2反应堆堆芯设计
6.2.1设计裕度
反应堆的堆芯以及相关的冷却剂系统、控制和保护系统应设计适当的裕度,以确保在任何运行状态和事故工况下不会超过规定可接受的燃料设计限值并符合辐射安全标准。
6.2.2反应堆核设计
反应堆核设计应给出堆芯内燃料组件.固体可燃毒物与控制棒组件的合理布置,提供足够的剩余反应性与控制手段,确定满足安全要求的堆芯功率分布、燃耗分布与反应性系数。
应具备探测反应堆堆芯内中子注量率分布及其变化的充分手段。
反应堆的堆芯以及相关的冷却系统应设计成在任何运行状态和事故工况下,堆芯具有负的功率反应性系数,反应堆固有的瞬时核反馈的净效应可以补偿反应性的快速增长。
反应性系数的最小和最大限值是多种参数(例如功率水平、硼浓度.燃耗等)的函数,应通过适当研究证实用于分析各种运行工况和事故工况所采用的反应性系数包络值的合理性。
反应堆的堆芯以及相关的冷却剂、控制、保护系统设计成应保证不可能发生超过规定的燃料设计限值的功率振荡工况,或者在发生那些工况时,能可靠而迅速地监测并被抑制。
反应堆堆芯核设计的基本要求可参考NB/T 20057.1中的规定。
6.3反应堆冷却剂系统压力边界设计
反应堆冷却剂系统以及相关的辅助、控制和保护系统的设计应有足够的裕度,以保证在任何正常运行包括预计运行事件期间都不会超过反应堆冷却系统压力边界的设计条件。
反应堆冷却剂系统应具备在各种运行模式下的超压保护功能，特别要关注低温水密实工况下的超压保护。
反应堆冷却剂系统压力边界的设计、制造、安装以及试验应保证异常泄漏，裂纹的迅速扩展以及整体破裂发生的概率极低。
设计中需考虑到反应堆冷却剂系统压力边界材料在运行状态包括维修、试验工况以及事故工况下的所有条件,并考虑到预期受到侵蚀、蠕变、疲劳、化学环境、辐射环境和老化等众多因素影响后的寿期末特性以及在确定部件初始状态和可能的劣化速率时的任何不确定因素。
反应堆冷却剂系统压力边界的部件的设计、制造和布置应便于在核电厂整个寿期内对压力边界定期进行充分检查和试验。
具体的反应堆冷却剂系统设计可参考NB/T20187中的规定.
6.4安全壳系统设计
6.4.1安全壳系统功能设计要求
安全壳系统设计应保证或有助于实现下述安全功能:
a)在运行状态和事故工况下包容放射性物质;
b)在运行状态和事故工况下的辐射屏蔽;
c)防御外部自然事件和外部人为事件。
反应堆安全壳以及相关的系统提供了一个固有的密闭屏障以防止放射性物质不可控地释放到环境中,并且在假定的事故工况所要求的时间内,对安全有重要作用的安全壳系统设计条件不会被超出。