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 ). 第节列出了参考文件 2. 第节中提供的摘要信息 3. 和 4. 对于在这套矩阵标准的编写者和用户之间建立适当的理解和沟通至关重要。它摘自参考标准（第 2. )以及供个人作者和用户使用的参考资料。第节提供了更详细的作者和用户信息、理由以及对个别实践、指南和方法的具体要求 3 – 5. 。第节讨论了内容和一致性的一般要求 6. . 图1 E706主矩阵中ASTM标准的组织和使用 1.2 本主矩阵旨在作为本系列标准的编制、修订和使用的参考和指南。 1.3 考虑设定压力中的中子辐射损伤- 温度极限和断裂分析( ( 1- 12 ) 2. 和指南 2009年5月 )，必须预测反应堆压力容器钢断裂韧性的中子引起的变化，然后通过推断容器使用寿命期间的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，但在其他地方( ( 2. , 6. , 7. , 11- 26 ) 和指南 2009年5月 ). 1.4 与( 1. )中子和伽马剂量测定( 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属于该主矩阵的单独标准中进行了说明( 1. , 17 ). 关注的主要变量( 1. ), ( 2. )，以及( 3. )如下所示: 1.4.1 钢的化学成分和微观结构， 1.4.2 钢材辐照温度， 1.4.3 从堆芯外围到监视位置以及进入容器和腔壁的发电厂配置和尺寸， 1.4.4 核心功率分布， 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 剂量测定测量， 1.4.9 中子特效，以及 1.4.10 中子剂量率效应。 1.5 存在许多方法和标准，以确保反应堆压力容器带束线在正常和事故载荷下的断裂控制的充分性( ( 1. , 7. , 8. , 11 , 12 , 14 , 16 , 17 , 23- 27 ) ，参考文件:ASTM标准( 2.1 )，核监管文件( 2.3 )和ASME标准( 2.4 )). 随着旧的LWR压力容器受到更高的辐射，韧性变化的预测能力必须提高。由于在容器的使用寿命期间，试验反应堆和动力反应堆监督计划将提供越来越多的信息，因此必须使用评估和使用这些信息的程序 ( 1. , 2. , 4- 9 , 11 , 12 , 23- 26 , 28 ). 此主矩阵定义当前( 1. )范围( 2. )应用领域，以及( 3. )ASTM标准系列的一般分组，如 图1 . 1.6 以国际单位制表示的数值应视为标准。本标准不包括其他计量单位。 1.7 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。 1.8 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 主矩阵- 本矩阵文件旨在作为现有标准使用的参考和指南，并帮助管理LWR-PV监测项目所需的新标准的开发和应用。段落 4.2 – 4.5 旨在协助参与这些标准的编制、修订和应用的作者和用户（见第节 6. ). 4.2 方法和主要目标: 4.2.1 建议在以下方面采用标准化程序和参考数据( 1. )中子和伽马剂量测定( 2. )物理学（中子学和伽马效应），以及( 3. )冶金损伤相关方法和数据，与核反应堆试验和监测结果的分析、解释和使用相关。 4.2.2 与( 1. ), ( 2. )，以及( 3. )，如果均匀一致地应用，可以对反应堆使用寿命期间LWR-PV钢断裂韧性的变化提供可靠的（10%至30%，1σ）估计 ( 36 ) . 4.2.3 Reg.Guide 1.99和ASME锅炉和压力容器规范第三节，第NF2121部分要求反应堆压力容器中使用的材料支持“。 应采用不受物品所受辐照条件伤害的材料制成。” 4.2.4 通过使用这一系列标准以及不确定度为10%至30%（1σ）的LWR-PV钢断裂韧性估计变化的统一和一致的文件和报告，核工业、许可证和监管机构可以满足实际的LWR发电厂运行条件和限制，如10 CFR第50部分Reg.Guide 1附录G和H中定义的条件和限制。 99和ASME锅炉和压力容器规范。 4.2.5 该系列标准的统一和一致应用使核工业、许可证和监管机构能够正确管理其关于LWR动力反应堆材料韧性的责任，以满足10 CFR第50部分附录G和H、监管指南1.99以及ASME锅炉和压力容器规范的要求。 4.3 剂量测定分析与解释 ( 1. , 3- 5. , 21 , 28 , 29 , 35 , 37 , 38 ) -当供应商/公用事业集团正确实施、验证和校准时，说明- 现有技术的剂量测定实践足以用于现有和未来的LWR发电厂监测计划。与不同变量（项目 1.4.1 – 1.4.10 属于 1.4 )在本节和 4.4 和 4.5 在这些章节中，所做的准确性（不确定性和误差）陈述是定量的，代表了最先进的技术。然而，它们对任何给定LWR发电厂进行EOL预测的正确性和用途取决于以下因素( 1. )现有的工厂监控程序( 2. )工厂几何配置，以及( 3. )类似植物的可用监测结果。如参考文件第III-A节所强调 ( 7. ) 然而，这些影响并不是唯一的，而是取决于( 1. )监控胶囊设计( 2. )反应堆堆芯和堆内构件的配置，以及( 3. )监视舱在反应堆几何结构内的位置。此外，结果可能有误的说法取决于给定反应堆发电厂的中子场和伽马射线场的估计方式 ( 1. , 11 , 28 , 36 , 39 , 40 ) 。对于中的大多数错误语句 4.3 – 4.5 ，假设这些估计是基于反应堆输运理论计算得出的，这些计算已被标准化为堆芯功率历史（见 4.4.1.2 )而不是监测胶囊剂量测定结果。这个 4.3 – 4.5 因此，准确性陈述旨在帮助标准编写者和用户确定不同变量在应用ASTM标准集方面的相对重要性， 图1 ，用于( 1. )LWR-PV监测程序( 2. )作为许可和监管的工具，以及( 3. )用于建立改进的冶金数据库。 4.3.1 要求的精度和基准字段参考: 4.3.1.1 精确度（不确定性和误差）( 注释1 )LWR-PV钢的曝光定义为±10%至15%（1σ），而建立趋势曲线的曝光精度最好不超过±10%（1σ ( 1. , 11 , 21 , 36 , 40- 46 ) 为了实现这些目标，中子剂量计的响应因此也应可解释为在±10%至15%（1σ）范围内的准确度（就暴露单位而言），并可测量为在±5%（1∑）范围内。 注1: 本文所述的不确定性是对测量结果可靠性的科学表征，其声明是将这些结果用于声称高精度或至少声明精度的应用研究的必要前提。术语误差将被保留，以表示结果与待测量量的已知偏差。更正时通常会考虑到错误。 4.3.1.2 应建立剂量测定“库存”，以支持供应商/公用事业集团和研发组织使用上述内容。 4.3.1.3 研究和公用事业公司供应商/服务实验室的基准现场参考已经完成，即: (1) 需要对现有和改进的剂量测定方法进行质量控制和认证；和 (2) 广泛应用于标准和参考中子场、PCA、PSF、SDMF、VENUS、NESDIP、PWR、BWR ( 1. ) 和一些测试反应堆，以量化和最大限度地减少不确定性和误差。 4.3.2 剂量测定探测器基准场参考工作现状- PCA、VENUS、NESDIP实验，包括和不包括模拟监视舱和动力反应堆测试，为研究分析和解释中的缺陷的影响提供了数据； PCA/PSF/SDMF扰动实验为更真实的压水堆和沸水堆发电厂监视舱配置提供了数据，并允许公用事业公司的供应商/服务实验室测试、验证、校准和更新其实践 ( 1. , 4. , 5. , 47 ) PSF监视舱测试提供了数据，但具有更一维的性质。PCA、VENUS和NESDIP实验以及一些试验反应堆工作增强了这些影响的基准现场量化 ( 1. , 3. , 4. , 28 , 36 , 48- 51 ) . 4.3.3 剂量测定探测器的额外验证工作: 4.3.3.1 建立核数据、光反应截面和中子损伤参考文件。 4.3.3.2 为传感器组设计和单个探测器建立适当的质量保证程序。 4.3.3.3 使用标准和参考中子场以及提供充分验证和校准的其他试验反应堆进行实验室间比较，见指南 2005年2月 . 4.4 反应堆物理分析与解释 ( 1. , 3. , 5. , 11 , 28 , 35 , 36 , 39 , 52 ) -当供应商/公用事业集团正确实施、验证和校准时，现有最先进的反应堆物理实践足以对现有和未来发电厂监控计划中光伏钢断裂韧性变化的容器内和容器外估计。 4.4.1 要求的精度和基准字段参考: 4.4.1.1 LWR-PV钢（监测胶囊样品和容器）暴露定义所需的精度约为±10%至15%（1σ）。在理想条件下，基准计算技术能够预测绝对- 每单位反应堆堆芯功率的堆外中子暴露量和反应率在±15%以内（但通常不在±5%以内）。然而，由于几何和其他复杂性，在实际发电厂的应用中，精度会更差 ( 1. , 3. , 4. , 11 , 21 , 36- 39 , 52 ) . 4.4.1.2 计算的船内和船外中子和伽马射线场可以标准化为堆芯功率历史或实验测量。后者可能包括来自监测胶囊、其他船内位置或- 容器外部空腔中的容器测量。在每种情况下，都需要考虑计算产生的不确定性。 4.4.2 发电厂反应堆物理分析与解释: 4.4.2.1 忽视基准的结果- 四分之一厚度位置（1/4 T )损伤暴露的血管壁估计值不容易与实验结果进行比较。基于不同的计算方式（通量>0.1或>1.0 MeV和dpa）的“超前因子”可能并不总是保守的；也就是说，一些监视舱已经被放置在- 使实际铅系数非常接近一，根本没有铅。此外，基于通量E>0.1或>1MeV和dpa的超前因子之间的差异可能是显著的，可能是50%或更多 ( 1. , 11 , 21 , 28 , 36- 38 , 52 ) . 4.4.3 PCA、VENUS和NESDIP实验以及PCA盲测试: 4.4.3.1 应在清洁几何形状和清洁堆芯源条件下对传输理论方法进行测试 ( 1. , 4. , 11 , 52 ) . 4.4.3.2 这是对供应商/公用事业集团动力反应堆物理计算工具充分性的必要但不充分的基准测试。 4.4.3.3 标准建议是，供应商/公用事业集团在其自己计算的与PCA、VENUS和NESDIP测量的积分和微分暴露和反应速率参数之间观察到的差异应用于验证和改进其计算工具（如果差异超出PCA、VENUS和NESDIP的实验精度限值）。 4.4.4 压水堆和沸水堆通用动力反应堆试验: 4.4.4.1 实际几何形状和可变堆芯源条件下的输运理论方法测试 ( 1. , 3. , 4. , 28 , 35 , 36 , 53 ) . 4.4.4.2 这是对供应商/公用事业集团动力反应堆物理计算工具充分性的必要且部分充分的基准测试。 4.4.4.3 标准建议应是，供应商/公用事业集团在其自己计算的和选定的压水堆或沸水堆测量的积分和微分暴露和反应速率参数之间观察到的差异，用于验证和改进其计算工具（如果差异超出选定的压电堆或沸水反应堆实验精度限制）。 4.4.4.4 参考文献中确定并讨论了为此目的建立的动力反应堆“基准” ( 1. , 3. , 4. , 35 , 53 ) 及其参考文献和指南 2006年2月 . 4.4.5 反应堆运行试验: 4.4.5.1 这是在实际几何形状和可变堆芯源条件下对传输理论方法的必要测试，但适用于特定类型或类别的供应商/公用事业集团动力反应堆。在这里，实际的船内监测胶囊和任何所需的船外测量剂量测定信息将被使用，如 4.4.4 ( 1. , 3. , 4. , 28 , 35 , 36 , 53 ) 然而，请注意，运行功率反应堆试验本身是不够的（监管指南1.190，第4.4.5.1节）。 4.4.5.2 与监测项目报告的暴露值和反应率相关的准确度预计在10%至30%（1σ）范围内 ( 36 ) . 4.5 冶金损伤相关性分析与解释 ( 1- 8. , 10 , 11 , 13 , 15- 29 , 36- 38 ) -当供应商/公用事业集团正确实施、验证和校准时，现有最先进的冶金损伤关联实践足以- 以及现有和未来发电厂监测项目中PV钢断裂韧性变化的船外估计。 4.5.1 要求的精度和基准字段参考: 4.5.1.1 LWR-PV钢（试验反应堆试样、监视舱试样、容器和支撑结构）数据相关性和数据外推（用于预测空间和时间上的断裂韧性变化）所需和可实现的精度约为±10%至30%（1σ）。然而，为了实现这一目标，冶金参数（ΔNDTT、上层、屈服强度等）。 )必须在±20至30%（1σ）范围内进行解释。这就要求，除了已经讨论过的剂量测定和物理变量外，与许多其他变量（中子剂量率、中子光谱、辐照温度、钢的化学成分和微观结构）相关的个体不确定性和误差必须最小化，并且结果必须在相同的±10%至30%（1σ）范围内可解释。 4.5.1.2 已经建立了先进的传感器组（包括剂量测定、温度和损伤相关传感器）和实践，以支持供应商/公用事业集团使用上述传感器 ( 1. , 4. , 5. , 11 , 39 , 50 , 54 , 55 ) . 4.5.1.3 公用事业公司供应商/服务实验室的基准现场参考以及先进实践正在进行或正在计划中 ( 1. , 3- 6. , 28 , 50 , 54- 56 ) : (1) 需要验证数据关联程序和时间和空间外推（到PV位置:表面，1/4 T 等）测试反应堆和动力反应堆监视舱冶金和中子暴露数据。 (2) 正在或将在试验反应堆中子场中进行测试，以量化和最大限度地减少不确定性和误差（此处包括使用损伤相关材料钢、蓝宝石等）。 ). 4.5.2 基准字段引用- PSF（所有位置:监视、水面、1/4 T , 1/2 T 和空隙箱）与Melusine PV模拟器和其他测试（如热中子效应）一起，提供了所有变量剂量测定、物理和冶金方面所需的验证数据 ( 1. , 4. , 10 , 19 , 21 , 22 , 37 , 38 ) 其他试验反应堆数据来自供应商/服务实验室/公用事业集团已基准化的监测胶囊结果 ( 1. , 3. , 4. , 6. , 11 , 18 , 27 , 28 , 36 , 40- 44 , 47 ) . 4.5.3 注册指南1.99，NRC，EPRI数据库- 美国材料试验协会E10.02和E10.05小组委员会、供应商、公用事业公司、美国电力研究院和美国核管理委员会承包商对美国核管理局和美国电力研究所（EPRI）数据库进行了持续研究，以建立现有测试和动力反应堆测量特性变化数据的改进数据库。ASTM任务组建议NRC和EPRI数据库使用更新的和新的暴露单位（通量总量>0.1、>1.0 MeV和dpa） ( 1. , 2. , 6. , 7. , 13 , 18 , 27 , 36 , 40- 44 , 47 ) ，并将这些建议纳入适当的标准中。 ASTM小组委员会E10.02在 E900 –15.所使用的曝光单位是E>1时的总通量 百万电子伏特。预测模型的基础记录在与 E900 ，可从ASTM获得。 4. 这项工作的成功取决于研究和个人服务实验室以及供应商/公用事业集团之间的良好合作。基于最新评估的ASTM剂量测定横截面文件，详见指南 2018年1月 ，并结合对所有已知变量的校正（扰动、照片- 反应、光谱、老化、产率、通量时间历程等）将根据需要和理由使用。试验反应堆数据将以类似的方式处理，视情况而定。

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 ). Referenced documents are listed in Section 2 . The summary 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 standards (Section 2 ) and references for use by individual writers and users. More detailed writers’ and users’ information, justification, and specific requirements for the individual practices, guides, and methods are provided in Sections 3 – 5 . General requirements of content and consistency are discussed in Section 6 . FIG. 1 Organization and Use of ASTM Standards in the E706 Master Matrix 1.2 This master matrix is intended as a reference and guide to the preparation, revision, and use of standards in the series. 1.3 To account for neutron radiation damage in setting pressure-temperature limits and making fracture analyses ( ( 1- 12 ) 2 and 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 ( ( 2 , 6 , 7 , 11- 26 ) and Guide E509 ). 1.4 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 separate standards belonging to this master matrix ( 1 , 17 ). 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 periphery 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 special effects, and 1.4.10 Neutron dose rate effects. 1.5 A number of methods and standards exist for ensuring the adequacy of fracture control of reactor pressure vessel belt lines under normal and accident loads ( ( 1 , 7 , 8 , 11 , 12 , 14 , 16 , 17 , 23- 27 ) , Referenced Documents: ASTM Standards ( 2.1 ), Nuclear Regulatory Documents ( 2.3 ) and ASME Standards ( 2.4 )). 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, procedures to evaluate and use this information must be used ( 1 , 2 , 4- 9 , 11 , 12 , 23- 26 , 28 ). This master matrix defines the current ( 1 ) scope, ( 2 ) areas of application, and ( 3 ) general grouping for the series of ASTM standards, as shown in Fig. 1 . 1.6 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.7 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. 1.8 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 Master Matrix— This matrix document is written as a reference and guide to the use of existing standards and to help manage the development and application of new standards, as needed for LWR-PV surveillance programs. Paragraphs 4.2 – 4.5 are provided to assist the authors and users involved in the preparation, revision, and application of these standards (see Section 6 ). 4.2 Approach and Primary Objectives: 4.2.1 Standardized procedures and reference data are recommended in regard to ( 1 ) neutron and gamma dosimetry, ( 2 ) physics (neutronics and gamma effects), and ( 3 ) metallurgical damage correlation methods and data, associated with the analysis, interpretation, and use of nuclear reactor test and surveillance results. 4.2.2 Existing state-of-the-art practices associated with ( 1 ), ( 2 ), and ( 3 ), if uniformly and consistently applied, can provide reliable (10 to 30 %, 1σ) estimates of changes in LWR-PV steel fracture toughness during a reactor’s service life ( 36 ) . 4.2.3 Reg. Guide 1.99 and Section III of the ASME Boiler and Pressure Vessel Code, Part NF2121 require that the materials used in reactor pressure vessels support “.shall be made of materials that are not injuriously affected by. irradiation conditions to which the item will be subjected.” 4.2.4 By the use of this series of standards and the uniform and consistent documentation and reporting of estimated changes in LWR-PV steel fracture toughness with uncertainties of 10 to 30 % (1σ), the nuclear industry and licensing and regulatory agencies can meet realistic LWR power plant operating conditions and limits, such as those defined in Appendices G and H of 10 CFR Part 50, Reg. Guide 1.99, and the ASME Boiler and Pressure Vessel Code. 4.2.5 The uniform and consistent application of this series of standards allows the nuclear industry and licensing and regulatory agencies to properly administer their responsibilities in regard to the toughness of LWR power reactor materials to meet requirements of Appendices G and H of 10 CFR Part 50, Reg. Guide 1.99, and the ASME Boiler and Pressure Vessel Code. 4.3 Dosimetry Analysis and Interpretation ( 1 , 3- 5 , 21 , 28 , 29 , 35 , 37 , 38 ) —When properly implemented, validated, and calibrated by vendor/utility groups, state-of-the-art dosimetry practices exist that are adequate for existing and future LWR power plant surveillance programs. The uncertainties and errors associated with the individual and combined effects of the different variables (items 1.4.1 – 1.4.10 of 1.4 ) are considered in this section and in 4.4 and 4.5 . In these sections, the accuracy (uncertainty and error) statements that are made are quantitative and representative of state-of-the-art technology. Their correctness and use for making EOL predictions for any given LWR power plant, however, are dependent on such factors as ( 1 ) the existing plant surveillance program, ( 2 ) the plant geometrical configuration, and ( 3 ) available surveillance results from similar plants. As emphasized in Section III-A of Ref ( 7 ) , however, these effects are not unique and are dependent on ( 1 ) the surveillance capsule design, ( 2 ) the configuration of the reactor core and internals, and ( 3 ) the location of the surveillance capsule within the reactor geometry. Further, the statement that a result could be in error is dependent on how the neutron and gamma ray fields are estimated for a given reactor power plant ( 1 , 11 , 28 , 36 , 39 , 40 ) . For most of the error statements in 4.3 – 4.5 , it is assumed that these estimates are based on reactor transport theory calculations that have been normalized to the core power history (see 4.4.1.2 ) and not to surveillance capsule dosimetry results. The 4.3 – 4.5 accuracy statements, consequently, are intended for use in helping the standards writer and user to determine the relative importance of the different variables in regard to the application of the set of ASTM standards, Fig. 1 , for ( 1 ) LWR-PV surveillance program, ( 2 ) as instruments of licensing and regulation, and ( 3 ) for establishing improved metallurgical databases. 4.3.1 Required Accuracies and Benchmark Field Referencing: 4.3.1.1 The accuracies (uncertainties and errors) ( Note 1 ) desirable for LWR-PV steel exposure definition are of the order of ±10 to 15 % (1σ) while exposure accuracies in establishing trend curves should preferably not exceed ±10 % (1σ) ( 1 , 11 , 21 , 36 , 40- 46 ) . In order to achieve such goals, the response of neutron dosimeters should therefore also be interpretable to accuracies within ±10 to 15 % (1σ) in terms of exposure units and be measurable to within ±5 % (1σ). Note 1: Uncertainty in the sense treated here is a scientific characterization of the reliability of a measurement result and its statement is the necessary premise for using these results for applied investigations claiming high or at least stated accuracy. The term error will be reserved to denote a known deviation of the result from the quantity to be measured. Errors are usually taken into account by corrections. 4.3.1.2 Dosimetry “inventories” should be established in support of the above for use by vendor/utility groups and research and development organizations. 4.3.1.3 Benchmark field referencing of research and utilities’ vendor/service laboratories has been completed that is: (1) Needed for quality control and certification of current and improved dosimetry practices; and (2) Extensively applied in standard and reference neutron fields, PCA, PSF, SDMF, VENUS, NESDIP, PWRs, BWRs ( 1 ) , and a number of test reactors to quantify and minimize uncertainties and errors. 4.3.2 Status of Benchmark Field Referencing Work for Dosimetry Detectors— PCA, VENUS, NESDIP experiments with and without simulated surveillance capsules and power reactor tests have provided data for studying the effect of deficiencies in analysis and interpretations; the PCA/PSF/SDMF perturbation experiments have provided data for more realistic PWR and BWR power plant surveillance capsule configurations and have permitted utilities’ vendor/service laboratories to test, validate, calibrate, and update their practices ( 1 , 4 , 5 , 47 ) . The PSF surveillance capsule test provided data, but of a more one-dimensional nature. PCA, VENUS, and NESDIP experimentation together with some test reactor work augmented the benchmark field quantification of these effects ( 1 , 3 , 4 , 28 , 36 , 48- 51 ) . 4.3.3 Additional Validation Work for Dosimetry Detectors: 4.3.3.1 Establishment of nuclear data, photo-reaction cross sections, and neutron damage reference files. 4.3.3.2 Establishment of proper quality assurance procedures for sensor set designs and individual detectors. 4.3.3.3 Interlaboratory comparisons using standard and reference neutron fields and other test reactors that provide adequate validations and calibrations, see Guide E2005 . 4.4 Reactor Physics Analysis and Interpretation ( 1 , 3 , 5 , 11 , 28 , 35 , 36 , 39 , 52 ) —When properly implemented, validated, and calibrated by vendor/utility groups, state-of-the-art reactor physics practices exist that are adequate for in- and ex-vessel estimates of PV-steel changes in fracture toughness for existing and future power plant surveillance programs. 4.4.1 Required Accuracies and Benchmark Field Referencing: 4.4.1.1 The accuracies desirable for LWR-PV steel (surveillance capsule specimens and vessels) exposure definition are of the order of ±10 to 15 % (1σ). Under ideal conditions benchmarking computational techniques are capable of predicting absolute in- and ex-vessel neutron exposures and reaction rates per unit reactor core power to within ±15 % (but generally not to within ±5 %). The accuracy will be worse, however, in applications to actual power plants because of geometrical and other complexities ( 1 , 3 , 4 , 11 , 21 , 36- 39 , 52 ) . 4.4.1.2 Calculated in-and ex-vessel neutron and gamma ray fields can be normalized to the core power history or to experimental measurements. The latter may include dosimetry from surveillance capsules, other in-vessel locations, or ex-vessel measurements in the cavity outside the vessel. In each case, the uncertainty arising from the calculation needs to be considered. 4.4.2 Power Plant Reactor Physics Analysis and Interpretation: 4.4.2.1 Result of Neglect of Benchmarking— One quarter thickness location (1/4 T ) vessel wall estimates of damage exposure are not easily compared with experimental results. “Lead factors,” based on the different ways they can be calculated (fluence >0.1 or >1.0 MeV and dpa) may not always be conservative; that is, some surveillance capsules have been positioned in-vessel such that the actual lead factor is very near unity—no lead at all. Also the differences between lead factors based on fluence E > 0.1 or > 1 MeV and dpa can be significant, perhaps 50 % or more ( 1 , 11 , 21 , 28 , 36- 38 , 52 ) . 4.4.3 PCA, VENUS, and NESDIP Experiments and PCA Blind Test: 4.4.3.1 Test of transport theory methods under clean geometry and clean core source conditions shall be made ( 1 , 4 , 11 , 52 ) . 4.4.3.2 This is a necessary but not sufficient benchmark test of the adequacy of a vendor/utility group’s power reactor physics computational tools. 4.4.3.3 The standard recommendation should be that the vendor/utility group’s observed differences between their own calculated and the PCA, VENUS, and NESDIP measured integral and differential exposure and reaction rate parameters be used to validate and improve their calculational tools (if the differences fall outside the PCA, VENUS, and NESDIP experimental accuracy limits). 4.4.4 PWR and BWR Generic Power Reactor Tests: 4.4.4.1 Test of transport theory methods under actual geometry and variable core source conditions ( 1 , 3 , 4 , 28 , 35 , 36 , 53 ) . 4.4.4.2 This is a necessary and partly sufficient benchmark test of the adequacy of a vendor/utility group’s power reactor physics computational tools. 4.4.4.3 The standard recommendation should be that the vendor/utility group’s observed differences between their own calculated and the selected PWR or BWR measured integral and differential exposure and reaction rate parameters be used to validate and improve their calculation tools (if the differences fall outside of the selected PWR or BWR experimental accuracy limits). 4.4.4.4 The power reactor “benchmarks” that have been established for this purpose are identified and discussed in Refs ( 1 , 3 , 4 , 35 , 53 ) and their references and in Guide E2006 . 4.4.5 Operating Power Reactor Tests: 4.4.5.1 This is a necessary test of transport theory methods under actual geometry and variable core source conditions, but for a particular type or class of vendor/utility group power reactors. Here, actual in-vessel surveillance capsule and any required ex-vessel measured dosimetry information will be utilized as in 4.4.4 ( 1 , 3 , 4 , 28 , 35 , 36 , 53 ) . Note, however, that operating power reactor tests are not sufficient by themselves (Reg. Guide 1.190, Section 4.4.5.1). 4.4.5.2 Accuracies associated with surveillance program reported values of exposures and reaction rates are expected to be in the 10 to 30 % (1σ) range ( 36 ) . 4.5 Metallurgical Damage Correlation Analysis and Interpretation ( 1- 8 , 10 , 11 , 13 , 15- 29 , 36- 38 ) —When properly implemented, validated, and calibrated by vendor/utility groups, state-of-the-art metallurgical damage correlation practices exist that are adequate for in- and ex-vessel estimates of PV-steel changes in fracture toughness for existing and future power plant surveillance programs. 4.5.1 Required Accuracies and Benchmark Field Referencing: 4.5.1.1 The accuracies desirable and achievable for LWR-PV steel (test reactor specimens, surveillance capsule specimens, and vessels and support structure) data correlation and data extrapolation (to predict fracture toughness changes both in space and time) are of the order of ±10 to 30 % (1σ). In order to achieve such a goal, however, the metallurgical parameters (ΔNDTT, upper shelf, yield strength, etc.) must be interpretable to well within ±20 to 30 % (1σ). This mandates that in addition to the dosimetry and physics variables already discussed that the individual uncertainties and errors associated with a number of other variables (neutron dose rate, neutron spectrum, irradiation temperature, steel chemical composition, and microstructure) must be minimized and results must be interpretable to within the same ±10 to 30 % (1σ) range. 4.5.1.2 Advanced sensor sets (including dosimetry, temperature and damage correlation sensors) and practices have been established in support of the above for use by vendor/utility groups ( 1 , 4 , 5 , 11 , 39 , 50 , 54 , 55 ) . 4.5.1.3 Benchmark field referencing of utilities' vendor/service laboratories, as well as advanced practices, is in progress or being planned that is ( 1 , 3- 6 , 28 , 50 , 54- 56 ) : (1) Needed for validation of data correlation procedures and time and space extrapolations (to PV positions: surface, 1/4 T , etc.) of test reactor and power reactor surveillance capsule metallurgical and neutron exposure data. (2) Being or will be tested in test reactor neutron fields to quantify and minimize uncertainties and errors (included here is the use of damage correlation materials—steel, sapphire, etc.). 4.5.2 Benchmark Field Referencing— The PSF (all positions: surveillance, surface, 1/4 T , 1/2 T , and the void box) together with the Melusine PV-simulator and other tests, such as for thermal neutron effects, provide needed validation data on all variables—dosimetry, physics, and metallurgy ( 1 , 4 , 10 , 19 , 21 , 22 , 37 , 38 ) . Other test reactor data comes from surveillance capsule results that have been benchmarked by vendor/service laboratory/utility groups ( 1 , 3 , 4 , 6 , 11 , 18 , 27 , 28 , 36 , 40- 44 , 47 ) . 4.5.3 Reg. Guide 1.99, NRC, EPRI Databases— NRC and Electric Power Research Institute (EPRI) databases have been studied on an ongoing basis by ASTM Subcommittees E10.02 and E10.05, vendors, utilities, EPRI, and NRC contractors to establish improved databases for existing test and power reactor measured property change data. ASTM task groups recommend the use of updated and new exposure units (fluence total >0.1, >1.0 MeV, and dpa) for the NRC and EPRI databases ( 1 , 2 , 6 , 7 , 13 , 18 , 27 , 36 , 40- 44 , 47 ) , and incorporate these recommendations in the appropriate standards. ASTM subcommittee E10.02 has updated the embrittlement database and the prediction model in E900 –15. The exposure unit used is total fluence for E > 1 MeV. The basis of the prediction model is documented in an adjunct associated with E900 , available from ASTM. 4 The success of this effort depends on good cooperation between research and individual service laboratories and vendor/utility groups. An ASTM dosimetry cross section file based on the latest evaluations, as detailed in Guide E1018 , and incorporating corrections for all known variables (perturbations, photo-reactions, spectrum, burn-in, yields, fluence time history, etc.) will be used as required and justified. Test reactor data will be addressed in a similar manner, as appropriate.