1.1 本指南介绍了一种预测参考转变温度偏移值的方法( 语音合成 )对于辐照压力容器材料。该方法基于 语音合成 从几个国家对商用加压（PWR）和沸腾（BWR）轻水冷却（LWR）动力反应堆进行的监测项目中获得的41-J（30 ft·lbf）下的夏比V形缺口数据显示。通过对大型监测数据库的统计分析，建立了脆化相关性，该数据库由辐射诱发的 语音合成 以及小组委员会E10.02汇编和分析的相关信息。数据库和分析的详细信息在单独的报告（ADJE090015-EA）中描述。 2. , 3. 这种脆化相关性是使用变量铜、镍、磷、锰、辐照温度、中子注量和产物形式开发的。 中列出了这些变量的数据范围和条件 1.1.1 . 部分 1.1.2 列出数据库中包含的材料以及可能影响 语音合成 但不用于脆化相关性。 1.1.1 数据库中用于脆化相关性的变量的材料和辐照条件范围: 1.1.1.1 铜含量高达0.4%。 1.1.1.2 镍含量高达1.7%。 1.1.1.3 磷含量高达0.03%。 1.1.1.4 锰含量在0.55%到2.0%之间。 1.1.1.5 辐照温度在255至300°C（491至572°F）范围内。 1.1.1.6 1×10范围内的中子注量 21 牛顿/米 2. 至2×10 24 牛顿/米 2. （E>1 MeV）。 1.1.1.7 描述产品形式（即焊缝、板材、锻件）的分类变量。 1.1. 2. 数据库中未包含在脆化相关性中的变量的材料和辐照条件范围: 1.1.2.1 A533型 B类1级和2级， A302型 B级， A302型 B级（修改），以及 A508型 2级和3级。此外，与这些ASTM等级相当的欧洲和日本钢种。 1.1.2.2 埋弧焊、电弧焊和电渣焊的成分与用于连接中所述基材的焊缝的成分一致 1.1.2.1 . 1.1.2.3 中子注量率在3×10范围内 12 牛顿/米 2. /s至5×10 16 牛顿/米 2. /s（E>1 MeV）。 1.1.2.4 商业压水堆和沸水堆堆芯附近反应堆容器区域预期范围内的中子能谱（大于约500MW）。 1.1.2.5 沸水反应堆的辐照暴露时间长达25年，压水反应堆的辐照暴露时间长达31年。 1.2 用户有责任证明其应用本指南的相关条件已通过本指南所依据的技术信息得到充分解决。应注意的是，数据库量化的条件并不是均匀分布在中描述的材料和辐照条件范围内 1.1 并且变量的某些组合，尤其是在数据范围的极值处，表示不足。当指南应用于接近用于开发数据集的数据范围极限的条件时，需要特别注意 语音合成 当应用程序涉及数据稀疏的数据空间区域时。虽然为本指南开发的脆化相关性是基于对大型数据库的统计分析，但对于涉及变量值超过中规定范围的应用，需要谨慎 1.1 . 由于与数据库内的其他暴露变量（即通量）有很强的相关性，由于数据库内数据分布不均匀（例如，压水堆和沸水堆数据的辐照温度和通量范围几乎没有重叠），中子注量率和辐照时间都没有充分提高预测的准确性，因此值得在本指南的脆化相关性中使用。本指南的未来版本可能会纳入中子注量率或辐照时间或两者的影响 语音合成 ，如中所述 ( 1. ) . 4. 辐照材料数据库、发展脆化相关性的技术基础及其应用中涉及的问题在一份单独的报告（ADJE090015-EA）中进行了讨论。该报告描述了九种不同的 语音合成 本指南制定过程中考虑的方程，其中一些方程是使用更有限的数据集（例如，国家项目数据）制定的 ( 2. , 3. ) ). 如果特定应用的材料变量或暴露条件在这些替代相关性之一的范围内，则可以提供更合适的指导。 1.3 本指南预计将与轻水反应堆容器材料辐照监测的几个标准配合使用。指南中介绍了确定本指南中使用的适用通量的方法 E482 , E944 ，以及试验方法 E1005年 . 实践中描述了这些单独指南和实践的总体应用 E853 . 1.4 以国际单位制表示的数值应视为标准值。国际单位制后括号中给出的值仅供参考，不被视为标准值。 1.5 本标准指南未定义 语音合成 应用于确定最终调整的参考温度，这通常包括考虑辐照前的转变温度，预测的 语音合成 以及移位估计方法中的不确定性。 1.6 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、健康和环境实践，并确定监管限制的适用性。 1.7 本国际标准是根据世界贸易组织技术性贸易壁垒（TBT）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》中确立的国际公认标准化原则制定的。 ====意义和用途====== 4.1 在加热和冷却期间，商业动力反应堆的运行必须符合压力-温度限制，以防止在存在缺陷时可能导致非韧性行为的温度下出现过压。随着中子损伤的累积，通过将压力温度限制调整到更高的温度来补偿反应堆容器的辐射损伤。目前的做法是根据 语音合成 在夏比V型缺口41-J（30 ft·lbf）能级下测量的中子辐照产生。为了确定电站运行寿命期间的压力-温度运行极限，预测 语音合成 必须制作。 4.1.1 在缺乏给定反应堆材料的监测数据的情况下（见实践 E185 和 E2215 )因此，有必要使用计算程序进行预测。 即使有可靠的监测数据可用，通常也有必要对数据进行插值或外推，以获得 语音合成 电厂运行寿命中的特定时间。本文提出的脆化相关性是为了这些目的而开发的。 4.2 研究表明，某些元素，尤其是铜（Cu）、镍（Ni）、磷（P）和锰（Mn），会导致反应堆压力容器钢的辐射敏感性发生变化。其他元素的重要性，如硅（Si）和碳（C），仍有待进一步研究。铜、镍、磷和锰是开发本文所述计算程序时使用的关键化学参数。 4.3 在推导这些程序时，仅使用了动力反应堆（压水堆和沸水堆）监测数据。 程序中使用的快中子注量测量值为n/m 2. （E>1 MeV）。在这些程序中，未考虑动力反应堆和试验反应堆中的注量率和中子能谱的差异。

1.1 This guide presents a method for predicting values of reference transition temperature shift ( TTS ) for irradiated pressure vessel materials. The method is based on the TTS exhibited by Charpy V-notch data at 41-J (30-ft·lbf) obtained from surveillance programs conducted in several countries for commercial pressurized (PWR) and boiling (BWR) light-water cooled (LWR) power reactors. An embrittlement correlation has been developed from a statistical analysis of the large surveillance database consisting of radiation-induced TTS and related information compiled and analyzed by Subcommittee E10.02. The details of the database and analysis are described in a separate report (ADJE090015-EA). 2 , 3 This embrittlement correlation was developed using the variables copper, nickel, phosphorus, manganese, irradiation temperature, neutron fluence, and product form. Data ranges and conditions for these variables are listed in 1.1.1 . Section 1.1.2 lists the materials included in the database and the domains of exposure variables that may influence TTS but are not used in the embrittlement correlation. 1.1.1 The range of material and irradiation conditions in the database for variables used in the embrittlement correlation: 1.1.1.1 Copper content up to 0.4 %. 1.1.1.2 Nickel content up to 1.7 %. 1.1.1.3 Phosphorus content up to 0.03 %. 1.1.1.4 Manganese content within the range from 0.55 to 2.0 %. 1.1.1.5 Irradiation temperature within the range from 255 to 300°C (491 to 572°F). 1.1.1.6 Neutron fluence within the range from 1 × 10 21 n/m 2 to 2 × 10 24 n/m 2 (E> 1 MeV). 1.1.1.7 A categorical variable describing the product form (that is, weld, plate, forging). 1.1.2 The range of material and irradiation conditions in the database for variables not included in the embrittlement correlation: 1.1.2.1 A533 Type B Class 1 and 2, A302 Grade B, A302 Grade B (modified), and A508 Class 2 and 3. Also, European and Japanese steel grades that are equivalent to these ASTM Grades. 1.1.2.2 Submerged arc welds, shielded arc welds, and electroslag welds having compositions consistent with those of the welds used to join the base materials described in 1.1.2.1 . 1.1.2.3 Neutron fluence rate within the range from 3 × 10 12 n/m 2 /s to 5 × 10 16 n/m 2 /s (E > 1 MeV). 1.1.2.4 Neutron energy spectra within the range expected at the reactor vessel region adjacent to the core of commercial PWRs and BWRs (greater than approximately 500MW electric). 1.1.2.5 Irradiation exposure times of up to 25 years in boiling water reactors and 31 years in pressurized water reactors. 1.2 It is the responsibility of the user to show that the conditions of interest in their application of this guide are addressed adequately by the technical information on which the guide is based. It should be noted that the conditions quantified by the database are not distributed evenly over the range of materials and irradiation conditions described in 1.1 , and that some combination of variables, particularly at the extremes of the data range are under-represented. Particular attention is warranted when the guide is applied to conditions near the extremes of the data range used to develop the TTS equation and when the application involves a region of the data space where data is sparse. Although the embrittlement correlation developed for this guide was based on statistical analysis of a large database, prudence is required for applications that involve variable values beyond the ranges specified in 1.1 . Due to strong correlations with other exposure variables within the database (that is, fluence), and due to the uneven distribution of data within the database (for example, the irradiation temperature and flux range of PWR and BWR data show almost no overlap) neither neutron fluence rate nor irradiation time sufficiently improved the accuracy of the predictions to merit their use in the embrittlement correlation in this guide. Future versions of this guide may incorporate the effect of neutron fluence rate or irradiation time, or both, on TTS , as such effects are described in ( 1 ) . 4 The irradiated material database, the technical basis for developing the embrittlement correlation, and issues involved in its application, are discussed in a separate report (ADJE090015-EA). That report describes the nine different TTS equations considered in the development of this guide, some of which were developed using more limited datasets (for example, national program data ( 2 , 3 ) ). If the material variables or exposure conditions of a particular application fall within the range of one of these alternate correlations, it may provide more suitable guidance. 1.3 This guide is expected to be used in coordination with several standards addressing irradiation surveillance of light-water reactor vessel materials. Method of determining the applicable fluence for use in this guide are addressed in Guides E482 , E944 , and Test Method E1005 . The overall application of these separate guides and practices is described in Practice E853 . 1.4 The values stated in SI units are to be regarded as standard. The values given in parentheses after SI units are provided for information only and are not considered standard. 1.5 This standard guide does not define how the TTS should be used to determine the final adjusted reference temperature, which would typically include consideration of the transition temperature before irradiation, the predicted TTS , and the uncertainties in the shift estimation method. 1.6 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. 1.7 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 Operation of commercial power reactors must conform to pressure-temperature limits during heatup and cooldown to prevent over-pressurization at temperatures that might cause non-ductile behavior in the presence of a flaw. Radiation damage to the reactor vessel is compensated for by adjusting the pressure-temperature limits to higher temperatures as the neutron damage accumulates. The present practice is to base that adjustment on the TTS produced by neutron irradiation as measured at the Charpy V-notch 41-J (30-ft·lbf) energy level. To establish pressure temperature operating limits during the operating life of the plant, a prediction of TTS must be made. 4.1.1 In the absence of surveillance data for a given reactor material (see Practice E185 and E2215 ), the use of calculative procedures are necessary to make the prediction. Even when credible surveillance data are available, it will usually be necessary to interpolate or extrapolate the data to obtain a TTS for a specific time in the plant operating life. The embrittlement correlation presented herein has been developed for those purposes. 4.2 Research has established that certain elements, notably copper (Cu), nickel (Ni), phosphorus (P), and manganese (Mn), cause a variation in radiation sensitivity of reactor pressure vessel steels. The importance of other elements, such as silicon (Si), and carbon (C), remains a subject of additional research. Copper, nickel, phosphorus, and manganese are the key chemistry parameters used in developing the calculative procedures described here. 4.3 Only power reactor (PWR and BWR) surveillance data were used in the derivation of these procedures. The measure of fast neutron fluence used in the procedure is n/m 2 (E > 1 MeV). Differences in fluence rate and neutron energy spectra experienced in power reactors and test reactors have not been accounted for in these procedures.