1.1 本实践涵盖方法，总结如下 附件A1 ，用于分析和解释从LWR压力容器监测项目中获得的中子暴露数据，并根据分析结果，建立一种用于评估压力容器及其支撑结构当前和未来状况的形式 2. ( 1- 74 ) 。 3. 1.2 该实践依赖于并结合了几种支持ASTM标准实践、指南和方法的应用（见主矩阵 E706 ) ( 1. ， 5. ， 13 ， 48 ， 49 ) 。 2. 为了使该实践至少部分独立，在与ASTM和其他文件相关的领域进行了适度的讨论。 讨论的支持主题领域包括反应堆物理计算、剂量计的选择和分析以及暴露单元。 1.3 这种做法仅限于与为支持LWR核电站的运营、许可和监管而制定的监督计划相关的直接应用。与试验反应堆结果的分析、解释和应用相关的程序和数据在实践中进行了阐述 2006年 指导 E900 ，和实践 E1035 。 1.4 本标准并非旨在解决与其使用相关的所有安全问题（如有）。本标准的使用者有责任在使用前制定适当的安全、健康和环境实践，并确定监管限制的适用性。 1.5 本国际标准是根据世界贸易组织技术性贸易壁垒委员会发布的《关于制定国际标准、指南和建议的原则的决定》中确立的国际公认的标准化原则制定的。 ===意义和用途====== 3.1 反应堆容器监督计划的目标有两个。该程序的第一个要求是监测反应堆容器带线区域铁素体材料因暴露于中子辐照和热环境而导致的断裂韧性特性的变化。 第二项要求是利用从监测程序中获得的数据来确定船只在整个使用寿命内可以运行的条件。 3.1.1 满足 3.1 ，要执行的任务很简单。包括监测程序的每个辐照胶囊可以被视为单独的实验。目标是定义并完成一个剂量测定程序，该程序将在后验上描述材料试样所暴露的中子场。由此产生的信息将成为数据库的一部分，该数据库在更严格的意义上适用于移除胶囊的特定工厂，但在更广泛的意义上也适用于整个行业。 3.1.2 满足 3.1 要执行的任务有些复杂。目的是准确描述压力容器在其使用寿命内将暴露在其中的中子场。中子场的描述必须包括容器壁内的空间梯度。因此，必须高度重视中子输运技术的使用以及计算设计基础的选择。由于给定的监视舱测量值，特别是在核电站早期获得的测量值，不一定代表反应堆的长期运行，因此将中子输运计算简单归一化为给定舱的剂量测定数据可能不合适 ( 1- 67 ) 。 3.2 反应堆容器的支撑结构的监视程序的目标和要求远没有那么严格，仅限于通过容器外空腔监测进行物理剂量测定，并结合使用可用的试验反应堆冶金数据来确定任何可能受到中子诱导性能变化影响的支撑结构钢的状况 ( 1. ， 29 ， 44- 58 ， 65- 70 ) 。

1.1 This practice covers the methodology, summarized in Annex A1 , to be used in the analysis and interpretation of neutron exposure data obtained from LWR pressure vessel surveillance programs and, based on the results of that analysis, establishes a formalism to be used to evaluate present and future condition of the pressure vessel and its support structures 2 ( 1- 74 ) . 3 1.2 This practice relies on, and ties together, the application of several supporting ASTM standard practices, guides, and methods (see Master Matrix E706 ) ( 1 , 5 , 13 , 48 , 49 ) . 2 In order to make this practice at least partially self-contained, a moderate amount of discussion is provided in areas relating to ASTM and other documents. Support subject areas that are discussed include reactor physics calculations, dosimeter selection and analysis, and exposure units. 1.3 This practice is restricted to direct applications related to surveillance programs that are established in support of the operation, licensing, and regulation of LWR nuclear power plants. Procedures and data related to the analysis, interpretation, and application of test reactor results are addressed in Practice E1006 , Guide E900 , and Practice E1035 . 1.4 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, health, and environmental practices and determine the applicability of regulatory limitations prior to use. 1.5 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee. ====== Significance And Use ====== 3.1 The objectives of a reactor vessel surveillance program are twofold. The first requirement of the program is to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region resulting from exposure to neutron irradiation and the thermal environment. The second requirement is to make use of the data obtained from the surveillance program to determine the conditions under which the vessel can be operated throughout its service life. 3.1.1 To satisfy the first requirement of 3.1 , the tasks to be carried out are straightforward. Each of the irradiation capsules that comprise the surveillance program may be treated as a separate experiment. The goal is to define and carry to completion a dosimetry program that will, a posteriori, describe the neutron field to which the material test specimens were exposed. The resultant information will then become part of a database applicable in a stricter sense to the specific plant from which the capsule was removed, but also in a broader sense to the industry as a whole. 3.1.2 To satisfy the second requirement of 3.1 , the tasks to be carried out are somewhat complex. The objective is to describe accurately the neutron field to which the pressure vessel itself will be exposed over its service life. This description of the neutron field must include spatial gradients within the vessel wall. Therefore, heavy emphasis must be placed on the use of neutron transport techniques as well as on the choice of a design basis for the computations. Since a given surveillance capsule measurement, particularly one obtained early in plant life, is not necessarily representative of long-term reactor operation, a simple normalization of neutron transport calculations to dosimetry data from a given capsule may not be appropriate ( 1- 67 ) . 3.2 The objectives and requirements of a reactor vessel's support structure's surveillance program are much less stringent, and at present, are limited to physics-dosimetry measurements through ex-vessel cavity monitoring coupled with the use of available test reactor metallurgical data to determine the condition of any support structure steels that might be subject to neutron induced property changes ( 1 , 29 , 44- 58 , 65- 70 ) .