1.1 本规程涵盖了通常根据规程设计的轻水慢化核反应堆压力容器监视舱的试样和剂量测定的评估 E185 . 1.2 此外，本规程还为重新评估设计寿命和超过设计寿命的操作的撤出时间表提供了指导。 1.3 该规程是概述核反应堆压力容器所需监督程序的一系列标准规程之一。监测程序监测构成轻水慢化核反应堆压力容器带线的铁素体钢的辐照引起的变化。 1.4 这一实践及其配套的监督计划实践 E185 ，旨在用于监测任何轻水慢化核反应堆中带状线材料的性能。 2. 1.5 指南中描述了对标准测试程序和补充测试的修改 E636 . 1.6 以国际单位制表示的数值应视为标准。括号中给出的值仅供参考。 1.7 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====== 4.1 轻水慢化核反应堆的设计考虑了中子辐射效应。可以在反应堆的整个使用寿命期间对系统操作参数进行改变，以考虑这些影响。监测程序用于测量实际容器材料因辐照环境而产生的性能变化。本规程描述了在更新撤药时间表、确保及时产生相关数据以及评估监测项目测试胶囊时应考虑的标准。 4.2 在本标准首次发布之前，监测程序的设计和监测胶囊的测试都包含在一个单一的标准《实践》中 E185 。从1961年临时通过到与本标准相关的替代品，实践 E185 多次修订（1966年、1970年、1973年、1979年、1982年、1993年和1998年）。因此，根据早期版本的标准设计和实施的监测项目中的胶囊通常在对标准进行实质性修改后进行测试。为了清楚起见，监督项目的标准实践已划分为新的实践 E185 其中涵盖了新监测项目的设计，以及涵盖监测胶囊测试和评估的标准实践。指南中描述了对标准测试程序和补充测试的修改 E636 . 4.3 本规程旨在涵盖所有轻水慢化反应堆压力容器监视舱的测试和评估。本规程适用于根据所有先前版本的《规程》设计和实施的监督程序中的胶囊测试 E185 . 4.4 反应堆压力容器性能的辐射引起的变化通常通过测量指标温度、上层能量和监测程序胶囊样品的拉伸性能来监测。实践中描述了这些辐射引起的变化的意义 E185 . 4.5 存在用于测试监控胶囊材料的替代方法。如指南所示，可提供一些补充和替代测试方法 E636 。使用 T o 试验方法中定义的参考温度方法 E1921 或 J -试验方法中定义的积分技术 E1820 此外，硬度测试可用于补充标准方法，作为监测材料辐照响应的手段。 4.6 实践 E853 描述了一种可用于中子剂量测定数据的分析和解释以及中子注量的确定的方法。监管机构或其他来源可能会描述不同的方法。 4.7 指导 E900 描述了一种用于预测TTS的方法。调节器或其他来源可以描述用于预测TTS的不同方法。 4.8 指导 E509 为轻水冷却核反应堆容器在役热退火程序的开发提供了方向，并证明了该程序的有效性，包括退火后容器辐射监测程序。

1.1 This practice covers the evaluation of test specimens and dosimetry from light-water moderated nuclear power reactor pressure vessel surveillance capsules normally designed according to Practice E185 . 1.2 Additionally, this practice provides guidance on reassessing the withdrawal schedule for design life and operation beyond design life. 1.3 This practice is one of a series of standard practices that outline the surveillance program required for nuclear reactor pressure vessels. The surveillance program monitors the irradiation-induced changes in the ferritic steels that comprise the beltline of a light-water moderated nuclear reactor pressure vessel. 1.4 This practice along with its companion surveillance program practice, Practice E185 , is intended for application in monitoring the properties of beltline materials in any light-water moderated nuclear reactor. 2 1.5 Modifications to the standard test program and supplemental tests are described in Guide E636 . 1.6 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only. 1.7 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 ====== 4.1 Neutron radiation effects are considered in the design of light-water moderated nuclear power reactors. Changes in system operating parameters may be made throughout the service life of the reactor to account for these effects. A surveillance program is used to measure changes in the properties of actual vessel materials due to the irradiation environment. This practice describes the criteria that should be considered in updating withdrawal schedules, ensuring relevant and timely data is produced, and evaluating surveillance program test capsules. 4.2 Prior to the first issue date of this standard, the design of surveillance programs and the testing of surveillance capsules were both covered in a single standard, Practice E185 . Between its provisional adoption in 1961 and its replacement linked to this standard, Practice E185 was revised many times (1966, 1970, 1973, 1979, 1982, 1993, and 1998). Therefore, capsules from surveillance programs that were designed and implemented under early versions of the standard were often tested after substantial changes to the standard had been adopted. For clarity, the standard practice for surveillance programs has been divided into the new Practice E185 that covers the design of new surveillance programs and this standard practice that covers the testing and evaluation of surveillance capsules. Modifications to the standard test program and supplemental tests are described in Guide E636 . 4.3 This practice is intended to cover testing and evaluation of all light-water moderated reactor pressure vessel surveillance capsules. The practice is applicable to testing of capsules from surveillance programs designed and implemented under all previous versions of Practice E185 . 4.4 The radiation-induced changes in the properties of the reactor pressure vessel are generally monitored by measuring the index temperatures, the upper-shelf energy, and the tensile properties of specimens from the surveillance program capsules. The significance of these radiation-induced changes is described in Practice E185 . 4.5 Alternative methods exist for testing surveillance capsule materials. Some supplemental and alternative testing methods are available as indicated in Guide E636 . Direct measurement of the fracture toughness is also feasible using the T o Reference Temperature method defined in Test Method E1921 or J -integral techniques defined in Test Method E1820 . Additionally, hardness testing can be used to supplement standard methods as a means of monitoring the irradiation response of the materials. 4.6 Practice E853 describes a methodology that may be used in the analysis and interpretation of neutron dosimetry data and the determination of neutron fluence. Regulators or other sources may describe different methods. 4.7 Guide E900 describes a method for predicting the TTS. Regulators or other sources may describe different methods for predicting TTS. 4.8 Guide E509 provides direction for development of a procedure for conducting an in-service thermal anneal of a light-water cooled nuclear reactor vessel and demonstrating the effectiveness of the procedure including a post-annealing vessel radiation surveillance program.