F82 堆用核仪器 标准查询与下载

IEC 60988:2009 核电厂.重要安全性仪器仪表.检测松动件的声学检测系统:特性,设计标准和操作规程

This International Standard is applicable to on-site systems used for continuous monitoring of structure-borne sound measured at the reactor coolant pressure boundary of light water reactors for the purpose of detecting loose parts.

Nuclear power plants - Instrumentation important to safety - Acoustic monitoring systems for detection og loose parts: characteristics, design criteria and operational procedures

IEC 61226:2009 核电站.重要安全性仪器仪表.仪器仪表和控制功能分级

This International Standard establishes a method of classification of the information and command functions for nuclear power plants@ and the IC systems and equipment that provide those functions@ into categories that designate the importance to safety of the function. The resulting classification then determines relevant design criteria. The design criteria are the measures of quality by which the adequacy of each function in relation to its importance to plant safety is ensured. In this standard@ the criteria are those of functionality@ reliability@ performance@ environmental durability (including seismic) and quality assurance (QA). This standard is applicable to all the information and command functions and the instrumentation and control (IC) systems and equipment that provide those functions. The functions@ systems and equipment under consideration provide automated protection@ closed or open loop control and information to the operating staff. They keep the NPP conditions inside the safe operating envelope and provide automatic actions@ or enable manual actions@ that prevent or mitigate accidents@ or that prevent or minimize radioactive releases to the site or wider environment. The IC functions that fulfil these roles safeguard the health and safety of the NPP operators and the public. This standard follows the general principles given in IAEA safety code NS-R-1 and safety guide NS-G-1.3@ and it defines a structured method of applying the guidance contained in those codes and standards to the IC systems that perform functions important to safety in a NPP. This standard should be read in association with the IAEA guides and IEC 61513 in implementing the requirements of IEC 61508 series.

Nuclear power plants - Instrumentation and control important to safety - Classification of instrumentation and control functions

IEC 60951-4:2009 核电站.重要安全性仪器仪表.事故及事故后环境的辐射监测.第4部分:流程中放射能的线内或线上监控设备

This part of IEC 60951 provides general guidance on the design principles and performance criteria for equipment for continuous in-line or on-line monitoring of radioactivity in process stream in nuclear power plants for accident and post-accident conditions. General requirements for technical characteristics@ test procedures@ radiation characteristics@ electrical@ mechanical@ and environmental characteristics are given in IEC 60951-1. These requirements are applicable in this part unless otherwise stated. IEC 60951-4 is only applicable to continuous in-line or on-line measurement@ i.e. monitors for which the detector measures radioactivity by being positioned in the process stream (i.e. immerged in) or adjacent to the process stream (i.e. viewing straight through a pipe or tank). It does not apply to monitors for which the detector measures a representative proportion of the stream at some remote location (sampling assembly)@ which are within the scope of IEC 60951-2. IEC 60951-4 is only applicable to monitors for accident and post-accident conditions. Process stream radiation monitoring equipment for normal and incident conditions are within the scope of IEC 60768.

Nuclear power plants - Instrumentation important to safety - Radiation monitoring for accident and post accident conditions - Part 4: Equipment for continuous in-line or on-line monitoring of radioactivity in process streams

IEC 60951-2:2009 核电站.重要安全性仪器仪表.核电厂事故及事故后辐射监测设备.第2部分:废气和通风气体放射能的连续离线监控设备

This part of IEC 60951 provides general guidance on the design principles and performance criteria for equipment for continuous off-line monitoring of radioactivity in gaseous effluents and ventilation air used in nuclear power plants for accident and post-accident conditions. General requirements for technical characteristics@ test procedures@ radiation characteristics@ electrical@ mechanical@ and environmental characteristics are given in IEC 60951-1. These requirements are applicable in this part unless otherwise stated. This standard is applicable to: ? noble gas activity monitors intended to measure the volumetric activity of radioactive noble gases in gaseous effluents at the discharge point and the variation of volumetric activity with time during accident and post-accident conditions. The monitor may also be used for the determination of the total discharge of noble gas activity over a given period; ? noble gas@ aerosol and specific nuclide (commonly iodine@ in its different forms: inorganic iodine@ organic iodine and iodine sticking on dust) monitors intended to measure the volumetric activity in air or gas systems (control room ventilation@ reactor leakage collection@ drywell ventilation exhaust@ fuel handling building ventilation exhaust@ reactor building ventilation purge exhaust) and detect any significant increase of radioactivity during or after an accident. This standard is only applicable to continuous off-line measurement@ i.e. monitors whose detector measures a representative proportion of the main effluent or ventilation stream at some remote location (sampling assembly). It does not apply to monitors with the detector positioned in or adjacent to the effluent or ventilation stream@ which are within the scope of IEC 60951-4. Sample extraction and laboratory analysis@ which are essential to a complete programme of effluent monitoring@ are not within the scope of this standard.

Nuclear power plants - Instrumentation important to safety - Radiation monitoring system for accident and post-accident conditions - Part 2: Equipment for continuous off-line monitoring of radioactivity in gaseous effluents and ventilation air

IEC 60951-1:2009 核电站.重要安全性仪器仪表.事故和事故后环境辐射监控系统.第1部分:一般要求

This part of IEC 60951 provides general guidance on the design principles and performance criteria for equipment to measure radiation and fluid (gaseous effluents or liquids) radioactivity levels at nuclear power plants during and after an accident. This standard is limited to equipment for continuous monitoring of radioactivity in accident and post-accident conditions. The object of this standard is to lay down mandatory general requirements and give examples of acceptable methods for equipment for continuous monitoring of radioactivity within the plant during and after accident conditions in nuclear power plants using light water reactors. It specifies@ for the equipment described above@ the general characteristics@ general test procedures@ radiation@ electrical@ safety and environmental characteristics and the identification and certification of the equipment. If this equipment is part of a centralized system for continuous radiation monitoring in a nuclear facility@ there may be additional requirements from other standards related to this system. Sample extraction and laboratory analysis@ which are essential to a complete programme of effluent monitoring@ are not within the scope of this standard.

Nuclear power plants - Instrumentation important to safety - Radiation monitoring system for accident and post-accident conditions - Part 1: General requirements

ASTM E844-09 反应堆监视用传感器装置设计和辐照的标准指南E 706(ⅡC)

In neutron dosimetry, a fission or non-fission dosimeter, or combination of dosimeters, can be used for determining a fluence-rate, fluence, or neutron spectrum, or both, in nuclear reactors. Each dosimeter is sensitive to a specific energy range, and, if desired, increased accuracy in a fluence-rate spectrum can be achieved by the use of several dosimeters each covering specific neutron energy ranges. A wide variety of detector materials is used for various purposes. Many of these substances overlap in the energy of the neutrons which they will detect, but many different materials are used for a variety of reasons. These reasons include available analysis equipment, different cross sections for different fluence-rate levels and spectra, preferred chemical or physical properties, and, in the case of radiometric dosimeters, varying requirements for different half-life isotopes, possible interfering activities, and chemical separation requirements.1.1 This guide covers the selection, design, irradiation, post-irradiation handling, and quality control of neutron dosimeters (sensors), thermal neutron shields, and capsules for reactor surveillance neutron dosimetry. 1.2 The values stated in SI units are to be regarded as standard. Values in parentheses are for information only. 1.3 This standard does not purport to address all of the safety problems, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Standard Guide for Sensor Set Design and Irradiation for Reactor Surveillance, E 706(IIC)

BS EN 62340:2010 核电站.重要安全性仪器仪表和控制系统.应对共因失效的要求(CCF)

I&C systems important to safety may be designed using conventional hard-wired equipment, computer-based equipment or by using a combination of both types of equipment. This International Standard provides requirements and recommendations1 for the overall architecture of I&C systems, which may contain either or both technologies. The scope of this standard is: a) to give requirements related to the avoidance of CCF of I&C systems that perform category A functions; b) to additionally require the implementation of independent I&C systems to overcome CCF, while the likelihood of CCF is reduced by strictly applying the overall safety principles of IEC SC 45A (notably IEC 61226, IEC 61513, IEC 60880 and IEC 60709); c) to give an overview of the complete scope of requirements relevant to CCF, but not to overlap with fields already addressed in other standards. These are referenced. This standard emphasises the need for the complete and precise specification of the safety functions, based on the analysis of design basis accidents and consideration of the main plant safety goals. This specification is the pre-requisite for generating a comprehensive set of detailed requirements for the design of I&C systems to overcome CCF. This standard provides principles and requirements to overcome CCF by means which ensure independence2: a) between I&C systems performing diverse safety functions within category A which contribute to the same safety target; b) between I&C systems performing different functions from different categories if e.g. a category B function is claimed as back-up of a category A function and; c) between redundant channels of the same I&C system. The implementation of these requirements leads to various types of defence against initiating CCF events.

Nuclear power plants - Instrumentation and control systems important to safety - Requirements for coping with common cause failure (CCF)

IEEE 497/COR 1-2007(R2008) 核发电站用意外事故监测仪表的IEEE标准.勘误表1:合并自2005年用户反馈

IEEE Standard Criteria for Accident Monitoring Instrumentation for Nuclear Power Generating Stations Corrigendum 1: Incorporation of User Feedback through 2005

IEEE Standard Criteria for Accident Monitoring Instrumentation for Nuclear Power Generating Stations Corrigendum 1: Incorporation of User Feedback through 2005

BS IEC 62385:2007 核电站.对安全性重要的仪器和控制.安全系统测量途径的性能评定方法

The purpose of this International Standard is to define the requirements for demonstrating acceptable performance of safety system instrument channels through response time testing, calibration verification, and other means. The same requirements may be adopted for demonstrating the acceptable performance of non-safety systems and other instrument channels. This Standard contains the main topics in its body and includes annexes to provide further information. The annexes are for information only and contain a selection of the available methods. The methods described in this Standard are used to check instrument calibration for accuracy and time response. It covers direct methods used to set calibration within required tolerances and indirect methods to indicate a need for a direct calibration. The use of the indirect methods allows for longer periods between the routine direct calibrations.

Nuclear power plants - Instrumentation and control important to safety - Methods for assessing the performance of safety system instrument channels

BS EN 60671:2011 核电站.与安全密切相关的仪表与控制装置.检查性试验

Nuclear power plants. Instrumentation and control systems important to safety. Surveillance testing

KS C IEC 61513:2006 核电站.重视安全系统的计测控制.通用要求

1.1 일 반 안전에 중요한 계측 제어 계통은 재래적인 실배선 장비, 컴퓨터 기반 장비 또

Nuclear power plants-Instrumentation and control for systems important to safety-General requirements for systems

ASTM E1214-06 使用反应堆堆芯压力容器监督用熔丝温度监视器的标准指南.E 706 (IIIE)

Temperature monitors are used in surveillance capsules in accordance with Practice E 2215 to verify the estimated value of the surveillance specimen irradiation temperature. Temperature monitors are needed to give evidence of overheating of surveillance specimens beyond the expected temperature. Because overheating causes a reduction in the amount of neutron radiation damage to the surveillance specimens, this overheating could result in a change in the measured properties of the surveillance specimens that would lead to an unconservative prediction of damage to the reactor vessel material. The magnitude of the reduction of radiation damage with overheating depends on the composition of the material and time at temperature. Guide E 900 provides an accepted method for quantifying the temperature effect. Because the evidence from melt wire monitors gives no indication of the duration of overheating above the expected temperature as indicated by melting of the monitor, the significance of overheating events cannot be quantified on the basis of thermal monitors alone. Indication of overheating does serve to alert the user of the data to further evaluate the irradiation temperature exposure history of the surveillance capsule. This guide is IIIE of Master Matrix E 706 that relates several standards used for irradiation surveillance of light water reactor vessel materials. It is intended primarily to amplify the requirements of Practice E 185 in the design of temperature monitors for the surveillance program. It may also be used in conjunction with Practice E 2215 to evaluate the post-irradiation test measurements..1.1 This guide describes the application of melt wire temperature monitors and their use for reactor vessel surveillance of light-water power reactors as called for in Practice E 185.1.2 The purpose of this guide is to recommend the selection and use of the common melt wire technique where the correspondence between melting temperature and composition of different alloys is used as a passive temperature monitor. Guidelines are provided for the selection and calibration of monitor materials; design, fabrication, and assembly of monitor and container; post-irradiation examinations; interpretation of the results; and estimation of uncertainties.1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. (See Note 1.)

Standard Guide for Use of Melt Wire Temperature Monitors for Reactor Vessel Surveillance, E 706 (IIIE)

EJ/T 1058.2-2005 核电厂安全系统计算机软件.第2部分:预防软件导致的共因故障、软件工具和预开发软件的使用

本部分规定了在核电厂安全重要计算机软件生产过程中的每个阶段(包括开发、设计、验证、确认与运行)以及每个阶段中的文件档案对于预防软件导致的共因故障、软件工具使用和预开发软件使用的基本要求以及应遵循的原则。 本部分适用于核电厂安全系统中用于执行安全功能的计算机软件生产过程中的对软件导致共因故障的预防、软件工具的使用和预开发软件的使用及其涉及到的数据准备和确认。

Software for computers impotant to safety for nuclear power plants - Part 2: Software aspects of defence against common cause failure, use of software tools and of pre-developed software

BS EN 60709:2010 核电站.重要安全性仪器仪表和控制系统.隔离

This standard is applicable to nuclear power plant instrumentation and control (I&C) systems, and their cables, that are important to safety, as defined in IAEA Safety Guide NS-G-1.3. It is also applicable to temporary installations which are part of those I&C systems important to safety (for example, auxiliary equipment for commissioning tests and experiments). Clause 6 is intended particularly for the cabling of the I&C systems important to safety. This standard applies to the I&C of new nuclear power plants as well as to I&C upgrading or back-fitting of existing plants. For existing plants, only a subset of the requirements is applicable; this subset is to be identified at the beginning of any project. Where independence is required by general safety standards such as IAEA safety guides or IEC 61513, one aspect of achieving this independence is physical separation between the systems and their equipment that perform functions important to safety. This standard defines the assessments needed and the technical requirements to be met for I&C systems important to safety and their cables, in order to achieve adequate physical separation between redundant sections of a system and between a system and another system. This separation is needed to prevent or minimise the impact on safety that could result from faults and failures which could be propagated or affect several sections of a system or several systems.

Nuclear power plants - Instrumentation and control systems important to safety - Separation

IEC 61468:2000/AMD1:2003 核电厂.堆芯仪器仪表.自给能中子探测器的特性和试验方法.修改1

This is Amendment 1 to IEC 61468-2000 (Nuclear power plants - In-core instrumentation - Characteristics and test methods of self-powered neutron detectors)

Nuclear power plants - In-core instrumentation - Characteristics and test methods of self-powered neutron detectors; Amendment 1

ASTM E854-03 反应堆监测用固态径迹记录仪(SSTR)监视器的应用和分析的标准试验方法,E 706(IIIB)

1.1 This test method describes the use of solid-state track recorders (SSTR) for neutron dosimetry in light-water reactor (LWR) applications. These applications extend from low neutron fluence to high neutron fluence, including high power pressure vessel surveillance and test reactor irradiations as well as low power benchmark field measurement. (1) This test method replaces Method E418. This test method is more detailed and special attention is given to the use of state-of-the-art manual and automated track counting methods to attain high absolute accuracies. In-situ dosimetry in actual high fluence-high temperature LWR environs is emphasized. 1.2 This test method includes SSTR analysis by both manual and automated methods. To attain a desired accuracy, the track scanning method selected places limits on the allowable track density. Typically good results are obtained in the range of 5 to 500 000 tracks/cm and accurate results at higher track densities have been demonstrated for some cases. (2) Track density and other factors place limits on the applicability of the SSTR method at high fluences. Special care must be exerted when measuring neutron fluences (E>1MeV) above 10 n/cm. 1.3 High fluence limitations exist. These limitations are discussed in detail in Section 13 and in references (3-5). 1.4 SSTR observations provide time-integrated reaction rates. Therefore, SSTR are truly passive-fluence detectors. They provide permanent records of dosimetry experiments without the need for time-dependent corrections, such as decay factors that arise with radiometric monitors. 1.5 Since SSTR provide a spatial record of the time-integrated reaction rate at a microscopic level, they can be used for "fine-structure" measurements. For example, spatial distributions of isotopic fission rates can be obtained at very high resolution with SSTR. 1.6 This standard does not purport to address the safety problems associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Standard Test Method for Application and Analysis of Solid State Track Recorder (SSTR) Monitors for Reactor Surveillance, E706(IIIB)

ASTM E844-03 反应堆监视用传感器装置设计和辐照的标准指南E-706(ⅡC)

In neutron dosimetry, a fission or non-fission dosimeter, or combination of dosimeters, can be used for determining a fluence-rate, fluence, or neutron spectrum, or both, in nuclear reactors. Each dosimeter is sensitive to a specific energy range, and, if desired, increased accuracy in a flux-spectrum can be achieved by the use of several dosimeters each covering specific neutron energy ranges. A wide variety of detector materials is used for various purposes. Many of these substances overlap in the energy of the neutrons which they will detect, but many different materials are used for a variety of reasons. These reasons include available analysis equipment, different cross sections for different flux levels and spectra, preferred chemical or physical properties, and, in the case of radiometric dosimeters, varying requirements for different half-life isotopes, possible interfering activities, and chemical separation requirements.1.1 This guide covers the selection, design, irradiation, post-irradiation handling, and quality control of neutron dosimeters (sensors), thermal neutron shields, and capsules for reactor surveillance neutron dosimetry. 1.2 The values stated in inch-pound units are to be regarded as the standard. The values given in parentheses are for information only. 1.3 This standard does not purport to address all of the safety problems, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Standard Guide for Sensor Set Design and Irradiation for Reactor Surveillance, E 706(IIC)

ASTM E854-03(2009) 反应堆监测用固态径迹记录仪(SSTR)监视器的应用和分析的标准试验方法,E 706(IIIB)

The SSTR method provides for the measurement of absolute-fission density per unit mass. Absolute-neutron fluence can then be inferred from these SSTR-based absolute fission rate observations if an appropriate neutron spectrum average fission cross section is known. This method is highly discriminatory against other components of the in-core radiation field. Gamma rays, beta rays, and other lightly ionizing particles do not produce observable tracks in appropriate LWR SSTR candidate materials. However, photofission can contribute to the observed fission track density and should therefore be accounted for when nonnegligible. For a more detailed discussion of photofission effects, see 13.4. In this test method, SSTR are placed in surface contact with fissionable deposits and record neutron-induced fission fragments. By variation of the surface mass density (μg/cm2) of the fissionable deposit as well as employing the allowable range of track densities (from roughly 1 event/cm2 up to 105 events/cm2 for manual scanning), a range of total fluence sensitivity covering at least 16 orders of magnitude is possible, from roughly 102 n/cm 2 up to 5 × 1018 n/cm2. The allowable range of fission track densities is broader than the track density range for high accuracy manual scanning work with optical microscopy cited in 1.2. In particular, automated and semi-automated methods exist that broaden the customary track density range available with manual optical microscopy. In this broader track density region, effects of reduced counting statistics at very low track densities and track pile-up corrections at very high track densities can present inherent limitations for work of high accuracy. Automated scanning techniques are described in Section 11. For dosimetry applications, different energy regions of the neutron spectrum can be selectively emphasized by changing the nuclide used for the fission deposit. It is possible to use SSTR directly for neutron dosimetry as described in 4.1 or to obtain a composite neutron detection efficiency by exposure in a benchmark neutron field. The fluence and spectrum-averaged cross section in this benchmark field must be known. Furthermore, application in other neutron fields may require adjustments due to spectral deviation from the benchmark field spectrum used for calibration. In any event, it must be stressed that the SSTR-fission density measurements can be carried out completely independent of any cross-section standards (6). Therefore, for certain applications, the independent nature of this test method should not be compromised. On the other hand, many practical applications exist wherein this factor is of no consequence so that benchmark field calibration would be entirely appropriate. 1.1 This test method describes the use of solid-state track recorders (SSTRs) for neutron dosimetry in light-water reactor (LWR) applications. These applications extend from low neutron fluence to high neutron fluence, including high power pressure vessel surveillance and test reactor irradiations as well as low power benchmark field measurement. (1) This test method replaces Method E 418. This test method is more detailed and special attention is given to the use of state-of-the-art manual and automated track counting methods to attain high absolute accuracies. In-situ dosimetry in actual high fluence-high temperature LWR applications is emphasized. 1.2 This test method includes SSTR analysis by both manual and automated methods. To attain a desired accuracy, the track scanning method selected places limits on the allowable track density. Typically good results are obtained in the range of 5 to ......

Standard Test Method for Application and Analysis of Solid State Track Recorder (SSTR) Monitors for Reactor Surveillance, E706(IIIB)

IEEE 497-2002(R2008) 核发电站意外事故监测仪器装置的IEEE标准

The criteria contain the functional and design requirements for accident monitoring instrumentation for nuclear power generating stations. This standard is intended for new plant designs. The guidance provided in this standard may also prove useful for operating nuclear power stations desiring to perform design modifications or design basis evaluations. This standard contains guidance for the selection of variables and establishes design and performance requirements. Guidance on the use of portable instrumentation and defining various display alternatives for accident monitoring instrumentation is also included.

IEEE Standard Criteria for Accident Monitoring Instrumentation for Nuclear Power Generating Stations

IEEE 420-2001(R2008) 核发电站用1E级控制板、操纵台及操纵仪表架的设计和鉴定用IEEE标准

This standard applies to the design and qualification of Class 1E control boards, panels, and racks. It does not apply to individual components, modules, and external field-run cables except as they may affect the design and qualification of Class 1E control boards, panels, and racks.

IEEE Standard for the Design and Qualification of Class 1E Control Boards, Panels, and Racks Used in Nuclear Power Generating Stations