1.1 本指南规定了在LWR反应堆压力容器的整个使用寿命内监测中子暴露的方法和频率。 1.2 根据本指南确定的物理剂量关系可用于通过实施规程的应用来估计反应堆压力容器的损坏 E693 和指南 E900 ，使用快中子注量（E>1.0MeV和E>0.1MeV）、每原子位移（dpa）或损伤函数相关的暴露参数作为独立的暴露变量。支持这些标准应用的是 E853 ， E944 ， E1005 和 E1018 标准，在 2.1 。 1.3 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。 1.4 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 监管要求-- 《美国联邦法规》（10CFR第50部分，附录H）要求对所有运行的LWR实施反应堆容器材料监督计划。其他国家也有类似的规定。该程序的目的是( 1. )监测反应堆容器带线中铁素体材料断裂韧性的变化 6. 由于暴露于中子辐射和热环境，以及( 2. )利用从监测程序中获得的数据来确定船舶在整个使用寿命内可以在足够的安全边际下运行的条件。 实践 E185 ，导出的机械性能数据，以及( r ，θ， z )物理剂量测定数据（源自计算、反应堆腔和监视舱测量 ( 1. ) 使用物理剂量测定标准）可以与指南中的信息一起使用 E900 以及参考文献。 4. ， 11- 18 提供性能退化和中子暴露之间的关系，通常称为“趋势曲线”。要在压力容器壁的所有点获得该趋势曲线，需要将选定的趋势曲线与适当的( r ，θ， z )利用本指南导出的中子场信息，在空间和时间上完成必要的插值和外推。 4.2 中子场特征-- 满足目标第二部分所需的任务 4.1 是复杂的，在实践中总结 E853 在这样做的过程中，有必要描述选定的中子场( r ，θ， z )压力容器壁内的点。该描述可以是时间相关的，也可以是在感兴趣的反应堆服务期内的时间平均值。通过将中子输运计算与核电站测量（如反应堆腔（容器外）和监视舱或RPV包层（容器内）测量）、剂量计传感器材料的基准辐照以及空间堆芯功率分布知识（包括时间相关性）相结合，可以最好地获得该描述。由于堆芯功率分布随时间变化，在核电站寿命早期获得的反应堆腔或监视舱测量值可能无法代表反应堆的长期运行。因此，将中子输运计算简单规范化为给定太空舱的剂量测定数据，不太可能在整个反应堆寿命内为该问题提供令人满意的解决方案。指导 E482 和指南 E944 提供与沸水堆和压水堆发电厂中子场特性相关的详细信息。 4.3 断裂力学分析-- 目前，施加在反应堆压力容器上的正常加热和冷却瞬态的操作限制是基于ASME锅炉和压力容器规范中概述的断裂力学技术。本规范要求假设存在深度等于压力容器厚度四分之一的表面缺陷。此外，事故瞬态（加压热冲击，PTS）的断裂力学分析可能涉及评估容器壁内不同深度缺陷的影响 ( 4. ) 因此，需要关于中子暴露的分布和压力容器内相应的辐射损伤的信息，无论是在空间上还是在时间上 ( 4. ) 。在这方面，实践 E185 为设计沸水堆和压水堆发电厂的最低监督程序、选择材料和评估冶金试样试验结果提供了指南。 实践 E2215 涵盖了LWR监测胶囊的试样评估和剂量测定。 4.4 中子光谱效应与DPA-- 对运行中的动力反应堆中子场的分析表明，中子光谱形状随着压力容器壁的径向深度而变化 ( 2. ， 3. ) 在从内半径到外半径的横向过程中，dpa/Γt的比率（其中，Γ是快（E>1.0MeV）中子注量率，t是材料暴露于平均注量率的时间）以2.0/1.0的因子变化。尽管dpa，因为它包括位移现象的更详细的建模，理论上应该比通量（E>1.0MeV）提供更好的与性能退化的相关性 ( 1. ， 19 ) ，这个话题仍然存在争议，现有的实验数据没有提供明确的指导 ( 19 ， 20 ) 因此，建议计算并报告这两个数量； 请参阅实践 E853 和实践 E693 。 4.5 船内监督计划: 4.5.1 反应堆容器监视舱中使用的中子剂量监测器提供反应堆内堆芯中平面上和容器壁附近单点的中子注量和注量率测量值；也就是说，在监视舱的位置 ( 1. ) 在实际实践中，监视舱可以位于反应堆内的方位角位置，该方位角位置不同于与最大中子暴露相关的方位角（或者不同于假设缺陷的方位角和轴向位置）；并且在距离缺陷和压力容器壁几厘米或更多的径向位置处 ( 4. ， 5. ) 尽管监视舱剂量测定确实为中子物理输运计算的标准化提供了点，但仍有必要使用分析方法来准确表示中子注量的空间变化（轴向、径向和方位角）（参考指南 E482 )。还需要使用其他测量来确认RPV中子暴露的空间分布。 4.5.2 假设监视舱的位置比RPV的表面径向更靠近核心，则它们可以从峰值暴露位置向方位偏移，以限制监视舱超前因子的大小。超前因子定义为监视舱中心的快中子通量与RPV包层-基底金属界面的峰值快中子通量之比。这种偏离峰值的方位角偏移的一个不利影响是，监视舱剂量测定没有“看到”堆芯中产生反应堆容器峰值暴露的部分。因此，监视舱无法监测堆芯功率分布变化的影响，这些变化是为了减少峰值RPV中子暴露。另一个不利影响是，在铅因子较大的情况下，胶囊会迅速暴露在高中子通量下。 例如，在领先因子为5的情况下，监测舱将在短短12年内暴露，这相当于反应堆压力容器运行60年后的峰值。实践 E185 和 E2215 建议不超过两倍的最大设计通量（MDF）或两倍的许可证结束通量（EOLF）。在这个例子中，这将需要在运行24年后取出任何剩余的监测胶囊。因此，在不采取其他步骤的情况下，反应堆将在没有剂量测定的情况下运行剩余的36年（60年的寿命）。 4.5.3 新的或更换的监测胶囊应通过使用改进的胶囊剂量测定法来识别和纠正操作缺陷。例如，对于一类压水堆，铜线是镉屏蔽的，以最大限度地减少微量钴的干扰。在大约三分之一的测量中，铜已被掺入镉中，防止分离和进一步处理。 这个问题的一个简单解决方案是使用不锈钢皮下注射管来容纳和分离镉管内的辐射监测器导线。示例尺寸包括:典型的辐射监测器导线外径=0.020英寸（0.5毫米）。典型的19号不锈钢卡套管外径为0.042英寸，内径为0.027英寸，壁厚为0.008英寸。典型的镉管外径为0.090英寸，内径为0.050英寸，壁厚为0.020英寸。 4.5.4 指导 E844 声明不应使用半衰期比辐照持续时间短的放射性核素。对于一类沸水堆反应堆，监视舱剂量测定是最小的；由一根铁线和一根铜线（有时也是一根镍线）组成。这种剂量测定法不适用于较长的辐照，因为活化产物的“记忆”太短，无法测量累积通量。 例如对于铁（n， 54 Mn，半衰期为312d。对于铜（n，α）活化产物， 60 Co，半衰期为5.27A。三个半衰期后，剩余的活性与计数统计数据的顺序相同。结果是，铁线“忘记”了两个多周期前发生的一切，铜线忘记了八个多周期后发生的一切。这假设燃料循环时间为24个月。铜（n，α）反应是由高能中子引起的，在沸水堆监视舱位置，只有1%至3%的快（E>1.0MeV）中子具有足够高的能量。这限制了铜线作为中子注量监测器的价值。为了监测RPV的中子暴露，需要其他剂量测定。安装容器外中子剂量测定是最合理和最具成本效益的选择。 4.5.5 RPV内表面的中子注量计算可以通过分析辐照不锈钢RPV包层的小样本来进一步验证。分析RPV包层样品已成为30多年来公认的做法 ( 21- 36 ) 在反应器关闭期间，可以从RPV包壳机加工小样品（50mg至100mg）。为了进行回顾性剂量测定 54 Mn， 58 Co，以及 9300万 使用Nb活动。由于其半衰期长， 9300万 Nb对于在无法获得精确中子输运计算的时间段内积分注量特别有用。在适当选择样品位置的情况下，可以确定RPV内表面上的快中子注量分布及其最大值。通过将这些数据与监测胶囊的剂量测定数据进行比较，还可以获得测量时的超前因子。如果包层材料是铌，这种技术效果最好- 稳定不锈钢。含0.7%铌的347型就是一个例子。回顾性剂量测定已成功证明适用于只有微量（~50ppm）铌的普通304型不锈钢覆层 ( 35 ) 重要的是，在加工样品之前，首先对包层表面进行抛光以去除放射性腐蚀产物，否则竞争活动可能会损害样品。用于采集这些样本的工具需要相对于反应堆地标准确定位，以便了解样本的实际轴向和方位角位置。合理的精度目标为轴向和方位±25 mm。采样位置误差的影响可以通过检查采样点附近的空间快中子注量率梯度来估计。一般来说，在快中子注量最大的区域，梯度往往非常小；在与芯的中间相对的轴向分布的情况下接近平坦。 在极端轴向位置，远远超过芯的末端，梯度是陡峭的。在那里，定位误差可能导致±20%的估计通量误差。类似的讨论适用于方位角注量率梯度。该工具还需要设计为完全保留所有机加工的覆层芯片，并防止从一个样品到另一个样品的交叉污染。由于广泛的结构（喷射泵等）阻碍了许多沸水堆的RPV包层的一般访问，因此对方位角和轴向包层样品的全面访问通常仅限于PWR。可以从沸水堆反应堆反应堆反应堆的包壳中提取一组更有限的样本。 4.5.6 新反应堆压力容器的设计和制造应考虑使用一种含有铌的不锈钢或铬镍铁合金来包覆容器的内表面。这将产生一个设计的回顾性剂量测定系统，该系统将捕获反应堆启动时的中子暴露数据。 4.6 离船监督计划: 4.6.1 30多年来，容器外中子剂量测定法也在核反应堆中得到了广泛应用 ( 28 ， 29 ， 31 ， 33 ， 35 ， 37- 97 ) EVND的主要优点是剂量测定系统的相对简单和相对低的成本。移除和更换辐照剂量测定只需很少的时间。典型装置的剂量测定跨越有效堆芯高度，并继续覆盖RPV的延长带线区域。在多个角度安装剂量测定可以实现全八分体覆盖（对于八分体对称型芯）。一些EVND装置包括在对称方位角进行的多次测量，以确认方位角注量率分布的对称性。不对称可能是由非对称堆芯功率分布、一个回路与另一个回路的水温差异或反应堆堆内构件或RPV的竣工尺寸椭圆度等因素造成的。 剂量测定胶囊通常包含完整的辐射监测器（参考指南 E844 )以确保良好的光谱覆盖和通量积分。通常，胶囊由不锈钢丝或链连接和支撑，不锈钢丝或链条依次被分段和计数，以提供轴向梯度信息。 4.6.2 为了最大限度地减少测量场扰动，剂量计胶囊应由中子透明材料（如铝）制成。这也有助于降低在移除和更换剂量测定时遇到的辐射剂量率。梯度链或线应为每线性英尺材料的低质量，以再次降低在处理辐照剂量测定过程中遇到的剂量率。 4.6.3 容器外中子剂量测定系统需要相对于众所周知且易于验证的反应堆特性进行精确定位。合理的精度目标为轴向和方位±25 mm。 剂量测定位置误差的影响可以通过检查测量点附近的空间快中子注量率梯度来估计。一般来说，在快中子注量最大的区域，梯度往往非常小；在与芯的中间相对的轴向分布的情况下接近平坦。在极端轴向位置，远远超过芯的末端，梯度是陡峭的。在那里，定位误差可能导致±20%的估计通量误差。类似的讨论适用于方位角注量率梯度。 4.6.4 理想情况下，容器外中子剂量测定是在反应堆启动前安装的，这样它就可以提供反应堆运行寿命的数据。建议在将改变反应堆容器中子暴露量的重大核电站改造前后对容器外中子剂量测定进行分析。 一些示例包括从低泄漏堆芯加载模式切换回外加载模式（反之亦然），执行显著（>10 %) 提高核电站功率，增加（或移除）堆芯通量抑制吸收器或假燃料棒，或修改反应堆内部结构。典型的剂量测定更换间隔在一到五个18个月长的燃料循环之间（或其他燃料循环长度的等效间隔）。 4.6.5 定期测量（RPV包壳样品或EVND）有助于确认中子注量预测，并有助于避免因反应堆特定计算模型错误而产生的问题 ( 98 ) 。 4.6.6 对商业反应堆中子场的计算表明，压力容器内径处的中子暴露量（dpa）可以随方位角位置的变化而变化三倍或更多 ( 2. ， 3. ) 反应堆压力容器外反应堆腔内的剂量测定监测器是一种有用的工具，因此，在确定压力容器壁内各点中子场计算的准确性时。 实践 E853 建议使用容器外反应堆腔中子剂量测量来验证物理输运计算。参考文献中讨论了基准场和功率反应堆的应用现状以及对这种方法的研究。 1. ， 18 ， 19 ， 37- 40 ， 99- 112 。

1.1 This guide establishes the means and frequency of monitoring the neutron exposure of the LWR reactor pressure vessel throughout its operating life. 1.2 The physics-dosimetry relationships determined from this guide may be used to estimate reactor pressure vessel damage through the application of Practice E693 and Guide E900 , using fast neutron fluence (E > 1.0 MeV and E > 0.1 MeV), displacements per atom (dpa), or damage-function-correlated exposure parameters as independent exposure variables. Supporting the application of these standards are the E853 , E944 , E1005 , and E1018 standards, identified in 2.1 . 1.3 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. 1.4 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 Regulatory Requirements— The USA Code of Federal Regulations (10CFR Part 50, Appendix H) requires the implementation of a reactor vessel materials surveillance program for all operating LWRs. Other countries have similar regulations. The purpose of the program is to ( 1 ) monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline 6 resulting from exposure to neutron irradiation and the thermal environment, and ( 2 ) make use of the data obtained from surveillance programs to determine the conditions under which the vessel can be operated with adequate margins of safety throughout its service life. Practice E185 , derived mechanical property data, and ( r , θ, z ) physics-dosimetry data (derived from the calculations and reactor cavity and surveillance capsule measurements ( 1 ) using physics-dosimetry standards) can be used together with information in Guide E900 and Refs. 4 , 11- 18 to provide a relation between property degradation and neutron exposure, commonly called a “trend curve.” To obtain this trend curve at all points in the pressure vessel wall requires that the selected trend curve be used together with the appropriate ( r , θ, z ) neutron field information derived by use of this guide to accomplish the necessary interpolations and extrapolations in space and time. 4.2 Neutron Field Characterization— The tasks required to satisfy the second part of the objective of 4.1 are complex and are summarized in Practice E853 . In doing this, it is necessary to describe the neutron field at selected ( r , θ, z ) points within the pressure vessel wall. The description can be either time dependent or time averaged over the reactor service period of interest. This description can best be obtained by combining neutron transport calculations with plant measurements such as reactor cavity (ex-vessel) and surveillance capsule or RPV cladding (in-vessel) measurements, benchmark irradiations of dosimeter sensor materials, and knowledge of the spatial core power distribution, including the time dependence. Because core power distributions change with time, reactor cavity or surveillance capsule measurements obtained early in plant life may not be representative of long-term reactor operation. Therefore, a simple normalization of neutron transport calculations to dosimetry data from a given capsule is unlikely to give a satisfactory solution to the problem over the full reactor lifetime. Guide E482 and Guide E944 provide detailed information related to the characterization of the neutron field for BWR and PWR power plants. 4.3 Fracture Mechanics Analysis— Currently, operating limitations for normal heat up and cool down transients imposed on the reactor pressure vessel are based on the fracture mechanics techniques outlined in the ASME Boiler and Pressure Vessel Code. This code requires the assumption of the presence of a surface flaw of depth equal to one fourth of the pressure vessel thickness. In addition, the fracture mechanics analysis of accident-induced transients (Pressurized Thermal Shock, (PTS)) may involve evaluating the effect of flaws of varying depth within the vessel wall ( 4 ) . Thus, information is required regarding the distribution of neutron exposure and the corresponding radiation damage within the pressure vessel, both in space and time ( 4 ) . In this regard, Practice E185 provides guidelines for designing a minimum surveillance program, selecting materials, and evaluating metallurgical specimen test results for BWR and PWR power plants. Practice E2215 covers the evaluation of test specimens and dosimetry from LWR surveillance capsules. 4.4 Neutron Spectral Effects and DPA— Analysis of the neutron fields of operating power reactors has shown that the neutron spectral shape changes with radial depth into the pressure vessel wall ( 2 , 3 ) . The ratio of dpa/ϕt (where ϕ is the fast (E > 1.0 MeV) neutron fluence rate and t is the time that the material was exposed to an average fluence rate) changes by factors of the order of 2.0/1.0 in traversing from the inner to the outer radius. Although dpa, since it includes a more detailed modeling of the displacement phenomenon, should theoretically provide a better correlation with property degradation than fluence (E > 1.0 MeV) ( 1 , 19 ) , this topic is still controversial and the available experimental data does not provide clear guidance ( 19 , 20 ) . Thus it is recommended to calculate and report both quantities; see Practice E853 and Practice E693 . 4.5 In-Vessel Surveillance Programs: 4.5.1 The neutron dosimetry monitors used in reactor vessel surveillance capsules provide measurements of the neutron fluence and fluence rate at single points on the core midplane within the reactor, and near the vessel wall; that is, at the surveillance capsule locations ( 1 ) . In actual practice, the surveillance capsules may be located within the reactor at an azimuthal position that differs from that associated with the maximum neutron exposure (or that differs from the azimuthal and axial location of the assumed flaw); and at a radial position a few centimeters or more from the flaw and the pressure vessel wall ( 4 , 5 ) . Although the surveillance capsule dosimetry does provide points for normalization of the neutron physics transport calculations, it is still necessary to use analytical methods that provide an accurate representation of the spatial variation (axial, radial and azimuthal) of the neutron fluence (refer to Guide E482 ). It is also necessary to use other measurements to confirm the spatial distribution of RPV neutron exposure. 4.5.2 Given that surveillance capsules are located radially closer to the core than the surface of the RPV, they may be shifted azimuthally away from the peak exposure location in order to limit the magnitude of the surveillance capsule lead factor. The lead factor is defined as the ratio of the fast neutron fluence at the center of the surveillance capsule to the peak fast neutron fluence at the clad–base metal interface of the RPV. One adverse effect of this azimuthal shift away from the peak is that the surveillance capsule dosimetry does not “see” the part of the core that produces the peak exposure of the reactor vessel. As a result, the surveillance capsule is unable to monitor the effect of changes in the core power distribution that are made to reduce the peak RPV neutron exposure. Another adverse effect is that with larger lead factors, the capsules are rapidly exposed to a high neutron fluence. For example, with a lead factor of five, a surveillance capsule will receive an exposure in as little as twelve years that is equivalent to what the reactor pressure vessel peak may see in 60 years of operation. Practices E185 and E2215 suggest not exceeding twice the maximum design fluence (MDF) or twice the end-of-license fluence (EOLF). In this example, this would require withdrawing any remaining surveillance capsules after 24 years of operation. Thus, without taking other steps, the reactor would be operated for the remaining 36 years (of a 60 year life) with no dosimetry present. 4.5.3 New or replacement surveillance capsules should recognize and correct operating deficiencies by using improved capsule dosimetry. For example, for one class of PWR, the copper wire is cadmium shielded to minimize interference from trace amounts of cobalt. In about one third of the measurements the copper has become incorporated into the cadmium preventing separation and further processing. A simple solution to this problem is to use stainless steel hypodermic tubing to contain and separate the radiometric monitor wire inside the cadmium tubing. Example dimensions include: Typical radiometric monitor wire outside diameter = 0.020 in. (0.5 mm). Typical 19 gauge stainless steel tubing is 0.042 in. outside diameter by 0.027 in. inside diameter, 0.008 in. wall thickness. Typical cadmium tubing is 0.090 in. outside diameter by 0.050 in. inside diameter, 0.020 in. wall thickness. 4.5.4 Guide E844 states that radionuclides with half-lives that are short compared to the irradiation duration should not be used. For one class of BWR reactor, the surveillance capsule dosimetry is minimal; consisting of an iron wire and a copper wire (sometimes also a nickel wire). This dosimetry is not suitable for longer irradiations as the “memory” of the activation products is too short to measure the accumulated fluence. For example, for the iron (n,p) activation product, 54 Mn, the half-life is 312 d. For the copper (n,α) activation product, 60 Co, the half-life is 5.27 a. After three half-lives the remaining activity is on the same order as the counting statistics. The result is that the iron wire has “forgotten” everything that has happened more than two cycles ago and the copper wire has forgotten everything that has happened more than eight cycles ago. This assumes 24-month-long fuel cycles. The copper (n,α) reaction is induced by high energy neutrons and that at a BWR surveillance capsule position only 1 % to 3 % of the fast (E > 1.0 MeV) neutrons are of high enough energy. This limits the value of the copper wire as a neutron fluence monitor. In order to monitor the neutron exposure of the RPV other dosimetry is needed. Installation of ex-vessel neutron dosimetry is the most reasonable and cost-effective option. 4.5.5 The neutron fluence calculation on the RPV inner surface can be further verified by means of analyzing small samples of the irradiated stainless steel RPV cladding. Analyzing RPV cladding samples has been a well-established practice for over 30 years ( 21- 36 ) . During the reactor shut down periods, small samples (50 mg to 100 mg) can be machined from the RPV cladding. For retrospective dosimetry purposes the measured 54 Mn, 58 Co, and 93m Nb activities are used. Because of its long half-life, 93m Nb is especially useful for integrating fluence over time periods where accurate neutron transport calculations are not available. With sample locations properly selected, the fast neutron fluence distribution and its maximum on the RPV inner surface can be determined. By comparison of these data to the dosimetry data of the surveillance capsules, the lead factor at the time of measurement can also be obtained. This technique works best if the cladding material is one of the niobium-stabilized stainless steels. Type 347 with 0.7 % niobium is one example. Retrospective dosimetry has been successfully demonstrated for ordinary Type 304 stainless steel cladding with only a trace (~50 ppm) of niobium ( 35 ) . It is important that the cladding surface is first polished to remove radioactive corrosion products before the sample is machined otherwise competing activity may compromise the sample. The tooling used to take these samples needs to be accurately located relative to reactor landmarks in order to know the actual axial and azimuthal locations of the samples. A reasonable accuracy target is ±25 mm axially and azimuthally. The effect of the sampling position error can be estimated by examining the spatial fast neutron fluence rate gradient in the vicinity of the sample point. In general, in the areas where the fast neutron fluence is the greatest, the gradient tends to be very small; approaching flat in the case of the axial distribution opposite the middle of the core. At extreme axial positions, well beyond the ends of the core, the gradient is steep. There the positioning error could lead to an estimated fluence error of ±20 %. A similar discussion applies to the azimuthal fluence rate gradients. The tooling also needs to be designed to completely retain all machined cladding chips and to prevent cross-contamination from one sample to another. Access to the full extent of azimuthal and axial clad samples is generally limited to PWRs due to the extensive structure (jet pumps, etc.) blocking general access to the RPV cladding of many BWRs. It may be possible to take a more limited set of samples from the cladding of a BWR RPV. 4.5.6 The design and manufacture of new reactor pressure vessels should consider using one of the stainless steels or Inconel alloys that contains niobium for the purpose of cladding the inner surface of the vessel. This would result in a designed-in retrospective dosimetry system that would capture neutron exposure data from reactor startup. 4.6 Ex-Vessel Surveillance Program: 4.6.1 Ex-vessel neutron dosimetry (EVND) has also been in wide scale application in nuclear reactors for over 30 years ( 28 , 29 , 31 , 33 , 35 , 37- 97 ) . The main advantages of EVND are the relative simplicity and the relatively low cost of the dosimetry system. Removal and replacement of irradiated dosimetry takes little time. Typical installations have dosimetry that spans the active core height and continues to cover the extended beltline region of the RPV. Installation of dosimetry at multiple angles allows full octant coverage (for octant symmetric cores). Some EVND installations include multiple measurements at symmetric azimuthal angles to confirm symmetry in the azimuthal fluence rate distributions. Asymmetries may result from such things as non-symmetric core power distributions, differences in water temperatures from one loop to another, or ovality in the as-built dimensions for the reactor internals or RPV. Dosimetry capsules typically contain a full complement of radiometric monitors (refer to Guide E844 ) to ensure good spectral coverage and fluence integration. Typically, capsules are connected and supported by stainless steel wires or chains, which are, in turn, segmented and counted to provide axial gradient information. 4.6.2 In order to minimize measurement field perturbation, the dosimeter capsules should be made of a neutron-transparent material such as aluminum. This also serves to reduce the radiation dose rates encountered when removing and replacing dosimetry. The gradient chains or wires should be a low mass per linear foot material, again to reduce the dose rates encountered during handling of irradiated dosimetry. 4.6.3 An ex-vessel neutron dosimetry system needs to be accurately located with respect to well-known and easily verified reactor features. A reasonable accuracy target is ±25 mm axially and azimuthally. The effect of the dosimetry position error can be estimated by examining the spatial fast neutron fluence rate gradient in the vicinity of the measurement point. In general, in the areas where the fast neutron fluence is the greatest, the gradient tends to be very small; approaching flat in the case of the axial distribution opposite the middle of the core. At extreme axial positions, well beyond the ends of the core, the gradient is steep. There the positioning error could lead to an estimated fluence error of ±20 %. A similar discussion applies to the azimuthal fluence rate gradients. 4.6.4 Ideally, the ex-vessel neutron dosimetry is installed before reactor startup so that it can provide data over the operating lifetime of the reactor. It is recommended that the ex-vessel neutron dosimetry be analyzed before and after significant plant modifications that would alter the neutron exposure of the reactor vessel. Some examples include switching from low-leakage core loading patterns back to out-in loading patterns (or vice versa), performing a significant (>10 %) uprating of the plant power, adding (or removing) core flux suppression absorbers or dummy fuel rods, or modifying the reactor internals geometry. The typical dosimetry replacement interval is between one and five 18-month-long fuel cycles (or equivalent intervals for other fuel cycle lengths). 4.6.5 Periodic measurements (either RPV cladding samples or EVND) serve to confirm neutron fluence projections and help to avoid problems that result from errors in reactor-specific calculational models ( 98 ) . 4.6.6 Calculations of neutron fields in commercial reactors show that the neutron exposure (dpa) at the inner diameter of the pressure vessel can vary by a factor of three or more as a function of azimuthal position ( 2 , 3 ) . Dosimetry monitors in the reactor cavity outside the reactor pressure vessel are a useful tool, therefore, in determining the accuracy of the neutron field calculations at points inside the pressure vessel wall. Practice E853 recommends the use of ex-vessel reactor cavity neutron dosimetry measurements for verification of the physics transport calculations. The status of benchmark field and power reactor applications as well as studies of this approach are discussed in Refs. 1 , 18 , 19 , 37- 40 , 99- 112 .