1.1本主矩阵标准描述了一系列标准实践、指南和方法，用于预测轻水反应堆（LWR）压力容器（PV）和支撑结构钢在整个压力容器使用寿命内的中子诱发变化（图1）。其中一些是现有的ASTM标准，一些是经过修改的ASTM标准，一些是拟议的ASTM标准。第6节讨论了内容和一致性的一般要求。第3-5节提供了关于九种实践、十种指南和三种方法的更详细的作者和用户信息、理由和具体要求。第2节讨论了参考文件。第3节和第4节中提供的摘要类型信息对于在这组矩阵标准的编写者和用户之间建立正确的理解和沟通至关重要。 它摘自参考文件第2节和参考文献 (1-106) 供个人作者和用户使用。 1.2本主矩阵旨在作为编制、修订和使用系列标准的参考和指南，并用于规划和调度目的。该指数旨在确保实现目标，而不考虑所需时间、相关ASTM委员会的数量或所涉及问题的复杂性。 1.3本主矩阵标准提供了与将在指南E584-77第10节裂变反应堆开发类别下开发的能量关键区域相关的ASTM标准指南，并在实践E583-97中进行了讨论。 1.4解释设定压力下的中子辐射损伤- 温度极限和断裂分析（见参考文献 2-7, 9-14, 21-57, 63, 69-71, 77, 78, 83-104 和推荐指南E509），必须预测反应堆压力容器钢断裂韧性的中子诱导变化，然后在容器使用寿命期间通过外推监测程序数据进行检查。预测方法中的不确定性可能很大。与PV和支撑结构钢性能变化的物理测量相关的技术、变量和不确定性不在本主矩阵中考虑，而是在其他地方考虑( 1, 3, 4, 10-13, 17, 21, 22-27, 32-39, 42, 43, 45, 49-57, 71, 77, 78, 83, 91-104, 和推荐指南E509）。与以下方面相关的技术、变量和不确定性:( 1. )中子和γ剂量测定( 2. )物理学（中子学和伽马效应），以及( 3. )冶金损伤相关程序和数据在该主矩阵中进行了说明 (2,34 ) . 关注的主要变量( 1. ), ( 2. )，和( 3. )具体如下: 1.4.1钢的化学成分和微观结构， 1.4.2钢辐照温度， 1.4.3从堆芯边缘到监视位置以及进入容器和空腔壁的电站配置和尺寸， 1.4.4核心配电， 1.4.5反应堆运行历史， 1.4.6反应堆物理计算， 1.4.7中子暴露单元的选择， 1.4.8剂量测量， 1.4.9中子光谱效应，以及 1.4.10中子剂量率效应。 1.5存在许多潜在的方法和标准，以确保反应堆压力容器带线在正常和事故载荷下的断裂控制充分性( 1-4, 6, 7, 13, 14, 21-28, 29-34, 52-57, 71, 77, 78, 91, 93, 推荐指南E509和2.3 ASME标准）。随着较旧的轻水反应堆压力容器受到更高的辐照，必须提高韧性变化的预测能力。由于在容器的使用寿命期间，测试反应堆和动力反应堆监督计划将提供越来越多的信息，因此可以而且必须制定更好的程序来评估和使用这些信息 (1-4, 6, 7, 9-15, 17, 21-34, 52-57, 69, 71-73, 77, 78, 91-104 和推荐指南E509）。因此，该主矩阵定义了电流( 1. )范围( 2. )应用领域，以及( 3. )22个ASTM标准系列的一般分组，如图所示。1-3. 1.6以国际单位制表示的数值应视为标准。 1.7 本标准可能涉及危险材料、操作和设备。本标准并非旨在解决与其使用相关的所有安全问题（如有）。本标准的用户有责任在使用前制定适当的安全和健康实践，并确定监管限制的适用性。

1.1 This master matrix standard describes a series of standard practices, guides, and methods for the prediction of neutron-induced changes in light-water reactor (LWR) pressure vessel (PV) and support structure steels throughout a pressure vessel's service life (Fig. 1). Some of these are existing ASTM standards, some are ASTM standards that have been modified, and some are proposed ASTM standards. General requirements of content and consistency are discussed in Section 6. More detailed writers' and users' information, justification, and specific requirements for the nine practices, ten guides, and three methods are provided in Sections 3-5. Referenced documents are discussed in Section 2. The summary-type information that is provided in Sections 3 and 4 is essential for establishing proper understanding and communications between the writers and users of this set of matrix standards. It was extracted from the referenced documents, Section 2 and references (1-106) for use by individual writers and users. 1.2 This master matrix is intended as a reference and guide to the preparation, revision, and use of standards in the series and for planning and scheduling purposes. This index is to ensure the accomplishment of an objective irrespective of the time required, the number of ASTM committees concerned, or the complexity of the issues involved. 1.3 This master matrix standard provides a guide to ASTM standards related to the energy-critical areas that are to be developed under the category of Fission Reactor Development, Section 10, of Guide E584-77 and as discussed in Practice E583-97. 1.4 To account for neutron radiation damage in setting pressure-temperature limits and making fracture analyses (see Refs 2-7, 9-14, 21-57, 63, 69-71, 77, 78, 83-104 and Recommended Guide E509), neutron-induced changes in reactor pressure vessel steel fracture toughness must be predicted, then checked by extrapolation of surveillance program data during a vessel's service life. Uncertainties in the predicting methodology can be significant. Techniques, variables, and uncertainties associated with the physical measurements of PV and support structure steel property changes are not considered in this master matrix, but elsewhere ( 1, 3, 4, 10-13, 17, 21, 22-27, 32-39, 42, 43, 45, 49-57, 71, 77, 78, 83, 91-104, and Recommended Guide E509). The techniques, variables and uncertainties related to ( 1 ) neutron and gamma dosimetry, ( 2 ) physics (neutronics and gamma effects), and ( 3 ) metallurgical damage correlation procedures and data are addressed in this master matrix (2,34 ) . The main variables of concern to ( 1 ), ( 2 ), and ( 3 ) are as follows: 1.4.1 Steel chemical composition and microstructure, 1.4.2 Steel irradiation temperature, 1.4.3 Power plant configurations and dimensions, from the core edge to surveillance positions and into the vessel and cavity walls, 1.4.4 Core power distribution, 1.4.5 Reactor operating history, 1.4.6 Reactor physics computations, 1.4.7 Selection of neutron exposure units, 1.4.8 Dosimetry measurements, 1.4.9 Neutron spectral effects, and 1.4.10 Neutron dose rate effects. 1.5 A number of potential methods and standards exist for ensuring the adequacy of fracture control of reactor pressure vessel belt lines under normal and accident loads ( 1-4, 6, 7, 13, 14, 21-28, 29-34, 52-57, 71, 77, 78, 91, 93, Recommended Guide E509, and 2.3 ASME Standards). As older LWR pressure vessels become more highly irradiated, the predictive capability for changes in toughness must improve. Since during a vessel's service life an increasing amount of information will be available from test reactor and power reactor surveillance programs, better procedures to evaluate and use this information can and must be developed (1-4, 6, 7, 9-15, 17, 21-34, 52-57, 69, 71-73, 77, 78, 91-104 and Recommended Guide E509). This master matrix, therefore, defines the current ( 1 ) scope, ( 2 ) areas of application, and ( 3 ) general grouping for the series of 22 ASTM standards, as shown in Figs. 1-3. 1.6 The values stated in SI units are to be regarded as the standard. 1.7 This standard may involve hazardous materials, operations, and equipment. 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.