F60 核反应堆综合 标准查询与下载

ANSI/ASME OM-2012 核电站的操作和维护规程

Operation and Maintenance of Nuclear Power Plants

ANSI/ANS 2.21-2012 在极限热阱上评个空气影响的准则

This standard describes atmospheric effects for consideration when designing ultimate heat sinks for safety-related systems at nuclear power units. Required analyses are provided for a meteorological assessment of the ultimate heat sink, to ensure design temperatures and cooling capacity requirements for the facility are met. The standard is intended to apply to new nuclear units or the re-design of the cooling systems at existing nuclear units.

Criteria for Assessing Atmospheric Effects on the Ultimate Heat Sink

ASME NUCLEAR-2012 新型核安全结构的锻造

Forging a New Nuclear Safety Construct

ANSI/ANS 15.21-2012 研究反应堆安全分析报告格式和内容

Format and Content for Safety Analysis Reports for Research Reactors

BS ISO 11320:2011 核临界安全.紧急准备和响应

Nuclear criticality safety. Emergency preparedness and response

EJ/T 878-2011 乏燃料离堆贮存水池安全设计准则

本标准规定了乏燃料离堆贮存水池(以下简称“离堆水池”) 设计中为确保核安全所应遵循的准 則和基本要求。 本标准适用于由存從水堆中卸出、經五年以上冷卻的乏燃料的離堆水池的安全設計。本标准也適 用于贮存其它类型的(例如从气冷推卸出的) 乏人燃料和燃料组件的部件等的离堆水池的安全设计。 本标准不适用于在堆水池安全设计。

Safety design criteria for spent fuel off-reactor storage pools

NB/T 20035-2011 压水堆核电厂工况分类

本标准规定了压水堆核电厂运行及事件工况分类方法。 本标准适用于压水堆核电厂设计、事故分析中涉及的工况划分。

Categorization of conditions of PWR nuclear power plants

ASTM D7727-11e1 用于反应堆冷却剂中氙裂变产物计算剂量当量疝 (DEX) 的标准操作规程

5.1 Each power reactor has a specific DEX value that is their technical requirement limit. These values may vary from about 200 to about 900 μCi/g based upon the height of their plant vent the location of the site boundary, the calculated reactor coolant activity for a condition of 1 % fuel defects, and general atmospheric modeling that is ascribed to that particular plant site. Should the DEX measured activity exceed the technical requirement limit the plant enters an LCO requiring action on plant operation by the operators. 5.2 The determination of DEX is performed in a similar manner to that used in determining DEI, except that the calculation of DEX is based on the acute dose to the whole body and considers the noble gases 85mKr, 85Kr, 87Kr, 88Kr, 131mXe, 133mXe, 133Xe, 135mXe, 135Xe, and 138Xe which are significant in terms of contribution to whole body dose. 5.3 It is important to note that only fission gases are included in this calculation, and only the ones noted in Table 1. For example 83mKr is not included even though its half life is 1.86 hours. The reason for this is that this radionuclide cannot be easily determined by gamma spectrometry (low energy X-rays at 32 and 9 keV) and its dose consequence is vanishingly small compared to the other, more prevalent krypton radionuclides. 5.4 Activity from 41Ar, 19F, 16N, and 11C, all of which predominantly will be in gaseous forms in the RCS, are not included in this calculation. 5.5 If a specific noble-gas radionuclide is not detected, it should be assumed to be present at the minimum-detectable activity. The determination of DOSE-EQUIVALENT XE-133 shall be performed using effective dose-conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12 (1),2 or the average gamma-disintegration energies as provided in ICRP Publication 38 (“Radionuclide Transformations”) or similar source. 1.1 This practice applies to the calculation of the dose equivalent to 133Xe in the reactor coolant of nuclear power reactors resulting from the radioactivity of all noble gas fission products. 1.2 The values given in parentheses are mathematical conversions to SI units, which are provided for information only and are not considered standard. 1.3 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered standard.

Standard Practice for Calculation of Dose Equivalent Xenon 40;DEX41; for Radioactive Xenon Fission Products in Reactor Coolant

ANSI/ASME BPVCSection XI-2011 核电厂元部件在使用中进行检查的规则

Rules for Inservice Inspection of Nuclear Power Plant Components

ANSI/ANS 19.3-2011 动力反应堆分析用稳态中子学方法

This standard provides guidance for performing and validating the sequence of steady state calculations leading to prediction, in all types of nuclear reactors, of: (1) Reaction rate spatial distributions (2) Reactivity (3) Change of isotopic compositions with time. The standard provides: (1) Guidance for the selection of computational methods. (2) Criteria for verification of calculational methods used by reactor core analysts (3) Criteria for evaluation of accuracy and range of applicability of data and methods (4) Requirements for documentation of the preceding. The scope of the standard is shown schematically in Figure 1.

Steady-State Neutronics Methods for Power Reactor Analysis

ANSI/ASME BPVCSection III-2011 核设施元部件制造规则

Rules for Construction of Nuclear Facility Components

ASTM D7727-11 用于反应堆冷却剂中氙裂变产物计算剂量当量疝(DEX)的标准操作规程

Each power reactor has a specific DEX value that is their technical requirement limit. These values may vary from about 200 to about 900 μCi/g based upon the height of their plant vent the location of the site boundary, the calculated reactor coolant activity for a condition of 1% fuel defects, and general atmospheric modeling that is ascribed to that particular plant site. Should the DEX measured activity exceed the technical requirement limit the plant enters an LCO requiring action on plant operation by the operators. The determination of DEX is performed in a similar manner to that used in determining DEI, except that the calculation of DEX is based on the acute dose to the whole body and considers the noble gases 85mKr, 85Kr, 87Kr, 88Kr, 131mXe, 133mXe, 133Xe, 135mXe, 135Xe, and 138Xe which are significant in terms of contribution to whole body dose. It is important to note that only fission gases are included in this calculation, and only the ones noted in 1. For example 83mKr is not included even though its half life is 1.86 hours. The reason for this is that this radionuclide cannot be easily determined by gamma spectrometry (low energy x-rays at 32 and 9 keV) and its dose consequence is vanishingly small compared to the other, more prevalent krypton radionuclides. Activity from 41Ar, 19F, 16N, and 11C, all of which predominantly will be in gaseous forms in the RCS, are not included in this calculation. If a specific noble-gas radionuclide is not detected, it should be assumed to be present at the minimum-detectable activity. The determination of DOSE-EQUIVALENT XE-133 shall be performed using effective dose-conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, or the average gamma-disintegration energies as provided in ICRP Publication 38 (′′Radionuclide Transformations'') or similar source.1.1 This practice applies to the calculation of the dose equivalent to 133Xe in the reactor coolant of nuclear power reactors resulting from the radioactivity of all noble gas fission products. 1.2 The values given in parentheses are mathematical conversions to SI units, which are provided for information only and are not considered standard. 1.3 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered standard.

Standard Practice for Calculation of Dose Equivalent Xenon (DEX) for Radioactive Xenon Fission Products in Reactor Coolant

ANSI/ANS 19.6.1-2011 压水堆重新装料起动的物理试验

This standard specifies the minimum acceptable startup reactor physics test program to determine if the operating characteristics of the core are consistent with the design predictions, which provides assurance that the core can be operated as designed.

Reload Startup Physics Tests for Pressurized Water Reactors

ANSI/ANS 58.14-2011 轻水反应器安全和压力完整性分类标准

This standard provides criteria for classification of structures, systems, and components for light water reactors. The standard addresses criteria for classification from both a safety and pressure retaining perspective for licensing design basis events.

Safety and Pressure Integrity Classification Criteria for Light Water Reactors

ASTM E482-11e1 反应堆压力容器监视用中子输运法应用的标准指南, E706(IID)

1.1 The methodology recommended in this guide specifies criteria for validating computational methods and outlines procedures applicable to pressure vessel related neutronics calculations for test and power reactors. The material presented herein is useful for validating computational methodology and for performing neutronics calculations that accompany reactor vessel surveillance dosimetry measurements (see Master Matrix E706 and Practice E853). Briefly, the overall methodology involves: (1) methods-validation calculations based on at least one well-documented benchmark problem, and (2) neutronics calculations for the facility of interest. The neutronics calculations of the facility of interest and of the benchmark problem should be performed consistently, with important modeling parameters kept the same or as similar as is feasible. In particular, the same energy group structure and common broad-group microscopic cross sections should be used for both problems. The neutronics calculations involve two tasks: (1) determination of the neutron source distribution in the reactor core by utilizing diffusion theory (or transport theory) calculations in conjunction with reactor power distribution measurements, and (2) performance of a fixed fission rate neutron source (fixed-source) transport theory calculation to determine the neutron fluence rate distribution in the reactor core, through the internals and in the pressure vessel. Some neutronics modeling details for the benchmark, test reactor, or the power reactor calculation will differ; therefore, the procedures described herein are general and apply to each case. (See NUREG/CR???5049, NUREG/CR???1861, NUREG/CR???3318, and NUREG/CR???3319.) 3.1.2 It is expected that transport calculations will be performed whenever pressure vessel surveillance dosimetry data become available and that quantitative comparisons will be performed as prescribed by 3.2.2. All dosimetry data accumulated that are applicable to a particular facility should be included in the comparisons. 3.2 Validation???Prior to performing transport calculations for a particular facility, the computational methods must be validated by comparing results with measurements made on a benchmark experiment. Criteria for establishing a benchmark experiment for the purpose of validating neutronics methodology should include those set forth in Guides E944 and

Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance, E706 (IID)

NB/T 20033-2010 核电厂初步可行性研究报告内容深度规定

本标准规定了核电厂初步可行性研究的基本工作内容和报告编制深度的要求,是初步可行性研究报告编制和审查的重要依据。 本标准适用于核电厂工程。

Regulation for content of pre-feasibility study report of nuclear power plants

NB/T 20034-2010 核电厂可行性研究报告内容深度规定

本标准规定了核电厂工程可行性研究报告的基本工作内容和报告编制深度的要求，是报告编制和审查的重要依据。 本标准适用于核电厂新建、扩建或改建工程项目。

Regulation for content of feasibility study report of nuclear power plants

NB/T 20016-2010 人因工程在核电厂基于计算机的监测和控制显示设计中的应用

本标准给出了人因工程原则在核电厂基于计算机的监测和控制显示设计中的应用指南。 本标准适用于核电厂基于计算机的监测和控制显示的设计。

The application of human factors engineering in the design of computer-based monitoring and control displays for nuclear power generating stations

IEEE 336-2010 核装置的1E类电力、仪表和控制设备的安装、检验及试验

Installation, inspection, and testing for class 1E power, instrumentation, and control equipment at nuclear facilities

ASTM E2006-10 轻水反应堆计算水准试验标准指南

This guide deals with the difficult problem of benchmarking neutron transport calculations carried out to determine fluences for plant specific reactor geometries. The calculations are necessary for fluence determination in locations important for material radiation damage estimation and which are not accessible to measurement. The most important application of such calculations is the estimation of fluence within the reactor vessel of operating power plants to provide accurate estimates of the irradiation embrittlement of the base and weld metal in the vessel. The benchmark procedure must not only prove that calculations give reasonable results but that their uncertainties are propagated with due regard to the sensitivities of the different input parameters used in the transport calculations. Benchmarking is achieved by building up data bases of benchmark experiments that have different influences on uncertainty propagation. For example, fission spectra are the fundamental data bases which control propagation of cross section uncertainties, while such physics-dosimetry experiments as vessel wall mockups, where measurements are made within a simulated reactor vessel wall, control error propagation associated with geometrical and methods approximations in the transport calculations. This guide describes general procedures for using neutron fields with known characteristics to corroborate the calculational methodology and nuclear data used to derive neutron field information from measurements of neutron sensor response. The bases for benchmark field referencing are usually irradiations performed in standard neutron fields with well-known energy spectra and intensities. There are, however, less well known neutron fields that have been designed to mockup special environments, such as pressure vessel mockups in which it is possible to make dosimetry measurements inside of the steel volume of the “vessel”. When such mockups are suitably characterized they are also referred to as benchmark fields. A benchmark is that against which other things are referenced, hence the terminology “to benchmark reference” or “benchmark referencing”. A variety of benchmark neutron fields, other than standard neutron fields, have been developed, or pressed into service, to improve the accuracy of neutron dosimetry measurement techniques. Some of these special benchmark experiments are discussed in this standard because they have identified needs for additional benchmarking or because they have been sufficiently documented to serve as benchmarks. One dedicated effort to provide benchmarks whose radiation environments closely resemble those found outside the core of an operating reactor was the Nuclear Regulatory Commission''s Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program (LWR-PV-SDIP) (1) . This program promoted better monitoring of the radiation exposure of reactor vessels and, thereby, provided for better assessment of vessel end-of-life conditions. An objective of the LWR-PV-SDIP was to develop improved procedures for reactor surveillance and document them in a series of ASTM standards (see Matrix E706). The primary means chosen for validating LWR-PV-SDIP procedures was by benchmarking a series of experimental and analytical studies in a variety of fields (see Guide E2005).1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide

Standard Guide for Benchmark Testing of Light Water Reactor Calculations