1.1 需要中子学计算- 准确计算多个位置的中子注量和注量率对于分析积分剂量测量和预测压力容器中的辐照损伤暴露参数值至关重要。暴露参数值可以直接从计算中获得，也可以间接从通过剂量测量调整的计算中获得；指导 E944 和实践 E853 定义适当的计算程序。 1.2 方法论- 用于反应堆容器监测的中子学计算包括三个基本领域:( 1. )通过比较基准实验中的计算和剂量测量来验证方法( 2. )确定反应堆堆芯中的中子源分布，以及( 3. )监测位置和压力容器中中子注量率的计算。 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。 1.4 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====== 3.1 概述: 3.1.1 本指南中推荐的方法规定了验证计算方法的标准，并概述了适用于试验反应堆和动力反应堆压力容器相关中子学计算的程序。本文提供的材料有助于验证计算方法和执行伴随反应堆容器监测剂量测量的中子学计算（见主矩阵） E706 和实践 E853 ). 简而言之，总体方法包括:( 1. )方法基于至少一个记录良好的基准问题的验证计算，以及( 2. )感兴趣设施的中子学计算。相关设施和基准问题的中子学计算应一致进行，重要建模参数应保持相同或尽可能相似。特别是，对于这两个问题，应使用相同的能量群结构和常见的宽群微观截面。此外，基准问题的特征应类似于感兴趣的设施。例如，功率反应堆基准应用于功率反应堆计算。非功率反应堆可能具有可能影响压力容器注量的特殊特性，在制定基准时需要考虑这些特性，例如束流管、辐照设施和非功率反应堆- 堆芯中子源。中子学计算涉及两项任务:( 1. )通过利用扩散理论（或传输理论）计算和反应堆功率分布测量，确定反应堆堆芯中的中子源分布，以及( 2. )执行固定裂变率中子源（固定源）传输理论计算，以确定反应堆堆芯、堆内构件和压力容器中的中子注量率分布。基准、测试反应堆或动力反应堆计算的一些中子学建模细节将有所不同；因此，本文描述的程序是通用的，适用于每种情况。 （见NUREG/CR-5049、NUREG/CR-1861、NUREG/CR-3318和NUREG/CR-3319。） 3.1.2 预计在压力容器监测剂量测定数据可用时，将进行传输计算，并按照 3.2.2 。适用于特定设施的所有累积剂量测定数据应包括在比较中。 3.2 验证- 在对特定设施进行传输计算之前，必须通过将结果与基准实验上的测量值进行比较来验证计算方法。为验证中子学方法而建立基准实验的标准应包括指南中规定的标准 E944 和 2006年 以及 3.2.1 参考文献中讨论了LWR监测计划的基准验证离散纵坐标辐射传输程序的极限精度 ( 1. ) . 4. 参考 ( 2. ) 提供了蒙特卡洛辐射传输代码的基准验证的详细信息。 3.2.1 基准要求- 为了将特定实验作为计算基准，建议采用以下标准: 3.2.1.1 必须有足够的信息来准确确定反应堆堆芯中的中子源分布。 3.2.1.2 必须在至少两个ex中报告测量值- 堆芯位置，用钢或冷却剂隔开。 3.2.1.3 应报告剂量学测量和计算通量的不确定度估计，包括计算的暴露参数和计算的剂量学活动。 3.2.1.4 定量标准，与方法验证中规定的标准一致 3.2.2 ，必须出版并证明其可实现。 3.2.1.5 测量和计算之间的差异应与 3.2.1.3 . 3.2.1.6 中子注量大于1 MeV和0.1 MeV[φ]的暴露参数值的结果( E >1 MeV和0.1 MeV）]和铁中每原子位移（dpa）的报告应与实践一致 E693 和 E853 . 3.2.1.7 必须报告反应速率（最好相对于中子注量标准确定） 237 Np（n，f）或 238 U（n，f），和 58 Ni（n，p）或 54 Fe（n，p）；基准测量中还应包括有助于光谱表征的其他反应，如铜、钛和钴铝提供的反应。这个 237 Np（n，f）反应特别重要，因为它对与铁dpa相同的中子能量区敏感。实践 E693 和 E853 和指南 E844 和 E944 讨论该标准。 3.2.2 方法验证- 通过准确预测适当的基准剂量测定结果来验证用于预测反应堆压力容器中中子注量的中子学方法至关重要。 此外，应提交以下文件:( 1. )收敛性研究结果，以及( 2. )源和几何建模中的不确定性引起的注量率和反应率的方差和协方差估计。对于蒙特卡洛计算，收敛性研究结果还应包括( 3. )分析作为粒子历史函数的优值（FOM），如果适用( 4. )描述用于生成权重窗口参数的技术。 3.2.2.1 例如，应进行收敛研究的离散坐标法的模型规范包括: ( 1. )中子截面或能量群结构( 2. )空间网格，以及( 3. )角正交。参考 ( 3. ) 单独或组合评估许多离散纵坐标参数的影响，可能有助于指导分析。对于靠近堆芯的区域，可以进行一维计算，以检查群结构和空间网格的充分性。应使用二维计算来检查角度正交的充分性。A. P 3. 建议使用横截面扩展和 S 8. 最小正交。对于与堆芯不相邻的区域，空间网格和角度求积的收敛研究应采用三种方法- 尺寸计算。 3.2.2.2 分析中应考虑由核数据中的已知不确定性传播的不确定性。离散纵坐标代码的不确定性分析可采用参考文献中讨论的灵敏度分析进行 ( 4. , 5. ) .在蒙特卡洛分析中，不确定性可以通过参考文献中讨论的摄动分析来处理 ( 6. ) .适当的计算机程序和协方差数据可用，灵敏度数据可作为确定不确定性估计的中间步骤获得。 5. 3.2.2.3 应根据以下测试用例评估几何形状和源分布中已知不确定性的影响: ( 1. )具有时间平均源分布和对堆芯和压力容器位置的最佳估计的参考计算( 2. )源分布中具有最大和最小预期偏差的参考案例几何形状，以及( 3. )具有堆芯、压力容器和其他相关位置的最大预期空间扰动的参考情况源分布。 3.2.2.4 对于所有测试用例，应比较测量和计算的积分参数。预计与收敛性研究中包括的参数相比，几何和中子源规格相关的不确定性更大。 可以识别和纠正与空间、能量和角度离散化相关的问题。与几何规格相关的不确定性是结构公差固有的。基于预期极值的计算提供了积分参数对所选变量灵敏度的度量。当上述程序与待验证的方法不一致时，建议的收敛性和不确定性评估中的变化是适当的。竣工数据可用于减少几何尺寸的不确定性。 3.2.2.5 为了说明基于应满足的测量和计算的定量标准，让ψ表示一组计算对数( C 一、 )至测量( E 一、 )比率。明确地 哪里 q 一、 和 N 隐式定义，并且 w 一、 是加权因子。由于某些反应在关注的光谱区域上比其他反应提供更大的响应，当其选择方法得到充分证明并得到充分保护时，可以使用加权因子，例如通过指南中详述的最小二乘调整方法 E944 .在没有使用最小二乘调整方法的情况下，集合的平均值 q 由给出 方差的最佳估计， S 2. 是 3.2.2.6 如果（除定性模型评估外）满足以下所有标准，则验证中子学方法: (1) 偏见| q |，小于ε 1. , (2) 标准差， S ，小于ε 2. , (3) 的自然对数的所有绝对值 C / E 比率(| q |, 一、 = 1. . N )小于ε 3. 和 (4) ε 1. , ε 2. ，和ε 3. 由基准测量文件定义，并证明可用于与计算进行比较的所有项目。 3.2.2.7 注意，非零对数平均值 C 一、 / E 一、 比率表明存在偏差。偏差的可能来源是:( 1. )信源归一化( 2. )中子学数据( 3. )横向泄漏校正（如适用）( 4. )几何建模，以及( 5. )数学近似。反应速率、等效裂变注量率或暴露参数值（例如φ( E >1 MeV）和dpa）可用于验证计算方法，如果合适的标准（即，由 3.2.2.5 和 3.2.2.6 )记录为兴趣基准。指南中讨论了反应堆容器监督特定基准验证程序的精度要求 2006年 参考文献中讨论了通用离散坐标和蒙特卡洛传输方法的验证测试 ( 1. , 2. ) . 3.2.2.8 进行这些比较的一个可接受的程序是: ( 1. )从中子学计算中获得剂量计位置的群注量率( 2. )折叠向导 E1018 将推荐的剂量测量横截面数据转换为与中子能量群注量率一致的多群集，或从计算的群注量速率中获得精细的群谱（与剂量测量横断面数据一致）( 3. )用适当的横截面折叠能量群注量率，以及( 4. )根据规定的定量标准比较计算数据和实验数据。 3.3 固定裂变源的测定- 典型反应堆的功率分布在反应堆寿命期间发生显著变化。 建议使用时间平均功率分布来确定用于损伤预测的中子源分布。伴随过程，如中所述 3.3.2 ，可能更适合于半衰期较短的产品核素的剂量学比较。对于多组方法，固定源可根据以下等式确定: 哪里: r = 空间节点， g = 能源集团， v = 每次裂变的平均中子数， x g = 群中裂变谱的分数 g 和 P r = 节点裂变率 r . 3.3.1 注意，除了裂变速率， v 和 x g 将随燃油燃耗而变化，应使用这些量的适当时间平均值。 对于任何给定的空间节点，裂变率和功率（即裂变/秒/瓦）之间的比率也将随燃耗而变化。 3.3.2 可以使用NUREG/CR-5049中建议的伴随程序，而不是使用时间平均源计算。 3.3.2.1 参考文献中讨论了光源分布变化的影响 ( 8. ) 对于涉及半衰期短的产品核素的剂量学比较，功率分布中的这些变化可能是显著的。在这种情况下，可以通过使用与以下各项成比例的因子对时间相关的功率分布进行加权来获得适当平均的功率分布: 哪里: f = 当时的加权因子， t , λ = 感兴趣核素的衰减常数，以及 t = 从曝光开始的时间。 每个核素的平均值不同，因此使用伴随程序可以避免不必要的输运计算重复，以便使用中所述的剂量测定结果验证计算 3.2.2 . 3.3.2.2 应注意确保伴随计算充分处理冷却剂密度的循环变化和反应堆几何结构的任何变化。 3.4 基于反应堆堆芯固定源的中子注量率计算- 本节中的讨论涉及方法验证计算和常规监测计算。在这两种情况下，中子输运计算必须估计堆芯内、通过堆内构件、反应堆压力容器内和容器外的中子注量率（例如，如果使用容器外剂量测定法）。方法验证程序与预测压力容器或试验设施中中子注量率的程序差别很小；因此，建议采用以下程序: 3.4.1 获取运输计算中涉及的材料配置的详细几何和成分描述。 还应估计数据中的不确定性。 3.4.2 从适当的数据库中获取适用的横截面集，例如: 3.4.2.1 评估的核数据文件（ENDF/B或其等效文件），或 3.4.2.2 通过处理上述文件获得的精细组库（例如，请参阅参考资料 ( 9 ) ). 3.4.3 执行一维、固定源、精细组计算，以便将精细组横截面折叠为用于多维计算的广泛组集。建议至少使用两个广义群集来执行一维群结构收敛性评估。广泛的集团结构应强调高- 能量范围，并应考虑重要材料（例如铁）的横截面最小值。 3.4.4 执行中概述的收敛研究 3.2.2 . 3.4.5 根据中建立的模型执行二维或三维固定源传输计算 3.4.1 – 3.4.4 . 3.4.6 将适当的剂量测定结果与 3.4.5 根据中给出的程序 3.2.2 。建议每次有新数据可用时，将所有有效的寿命累积反应堆剂量测定数据包括在该比较中，但进行剂量计特定比较时除外。 3.4.7 如果不满足验证标准，请重复适当的步骤。 注意，如果相关的 C / E 比率与适用的平均值相差很大 C / E 可能存在比率和测量误差。如果与平均值的偏差超过三个标准偏差的等效值，则可能存在测量误差。此外，反应堆计算的源可以按比例缩放，以最小化由定义的偏差和方差 公式2 和 等式3 前提是数据不会因缩放源而被丢弃。 3.4.8 中子学计算结果可以多种方式使用: 3.4.8.1 确定一个归一化常数，使计算值相对于测量值的偏差最小化，以缩放群通量。 这是一种简单且常用的调整程序替代方法。然而，应根据估计的源不确定性严格检查该常数的大小。 3.4.8.2 使用指南中建议的频谱调整程序 E944 使用具有不确定度估计的计算组注量和剂量测定数据，以获得对计算组注量和暴露参数的调整。然后，预测的压力容器注量可以包含从调整后的注量中获得的光谱和归一化数据。 3.4.8.3 将计算出的通量谱用于实践 E693 用于损伤暴露预测。 3.4.8.4 预计在某些情况下，上述建议的程序将与一些待验证的方法不一致。在这些情况下，程序变更是适当的，但应做好记录。

1.1 Need for Neutronics Calculations— An accurate calculation of the neutron fluence and fluence rate at several locations is essential for the analysis of integral dosimetry measurements and for predicting irradiation damage exposure parameter values in the pressure vessel. Exposure parameter values may be obtained directly from calculations or indirectly from calculations that are adjusted with dosimetry measurements; Guide E944 and Practice E853 define appropriate computational procedures. 1.2 Methodology— Neutronics calculations for application to reactor vessel surveillance encompass three essential areas: ( 1 ) validation of methods by comparison of calculations with dosimetry measurements in a benchmark experiment, ( 2 ) determination of the neutron source distribution in the reactor core, and ( 3 ) calculation of neutron fluence rate at the surveillance position and in the pressure vessel. 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 ====== 3.1 General: 3.1.1 The methodology recommended in this guide specifies criteria for validating computational methods and outlines procedures applicable to pressure vessel related neutronics calculations for test and power reactors. The material presented herein is useful for validating computational methodology and for performing neutronics calculations that accompany reactor vessel surveillance dosimetry measurements (see Master Matrix E706 and Practice E853 ). Briefly, the overall methodology involves: ( 1 ) methods-validation calculations based on at least one well-documented benchmark problem, and ( 2 ) neutronics calculations for the facility of interest. The neutronics calculations of the facility of interest and of the benchmark problem should be performed consistently, with important modeling parameters kept the same or as similar as is feasible. In particular, the same energy group structure and common broad-group microscopic cross sections should be used for both problems. Further, the benchmark problem should be characteristically similar to the facility of interest. For example, a power reactor benchmark should be utilized for power reactor calculations. Non-power reactors may have special features that may affect pressure vessel fluence and require consideration when developing a benchmark, such as beam tubes, irradiation facilities, and non-core neutron sources. The neutronics calculations involve two tasks: ( 1 ) determination of the neutron source distribution in the reactor core by utilizing diffusion theory (or transport theory) calculations in conjunction with reactor power distribution measurements, and ( 2 ) performance of a fixed fission rate neutron source (fixed-source) transport theory calculation to determine the neutron fluence rate distribution in the reactor core, through the internals and in the pressure vessel. Some neutronics modeling details for the benchmark, test reactor, or the power reactor calculation will differ; therefore, the procedures described herein are general and apply to each case. (See NUREG/CR-5049, NUREG/CR-1861, NUREG/CR-3318, and NUREG/CR-3319.) 3.1.2 It is expected that transport calculations will be performed whenever pressure vessel surveillance dosimetry data become available and that quantitative comparisons will be performed as prescribed by 3.2.2 . All dosimetry data accumulated that are applicable to a particular facility should be included in the comparisons. 3.2 Validation— Prior to performing transport calculations for a particular facility, the computational methods must be validated by comparing results with measurements made on a benchmark experiment. Criteria for establishing a benchmark experiment for the purpose of validating neutronics methodology should include those set forth in Guides E944 and E2006 as well as those prescribed in 3.2.1 . A discussion of the limiting accuracy of benchmark validation discrete ordinate radiation transport procedures for the LWR surveillance program is given in Reference ( 1 ) . 4 Reference ( 2 ) provides details on the benchmark validation for a Monte Carlo radiation transport code. 3.2.1 Requirements for Benchmarks— In order for a particular experiment to qualify as a calculational benchmark, the following criteria are recommended: 3.2.1.1 Sufficient information must be available to accurately determine the neutron source distribution in the reactor core. 3.2.1.2 Measurements must be reported in at least two ex-core locations, well separated by steel or coolant. 3.2.1.3 Uncertainty estimates should be reported for dosimetry measurements and calculated fluences including calculated exposure parameters and calculated dosimetry activities. 3.2.1.4 Quantitative criteria, consistent with those specified in the methods validation 3.2.2 , must be published and demonstrated to be achievable. 3.2.1.5 Differences between measurements and calculations should be consistent with the uncertainty estimates in 3.2.1.3 . 3.2.1.6 Results for exposure parameter values of neutron fluence greater than 1 MeV and 0.1 MeV [φ( E > 1 MeV and 0.1 MeV)] and of displacements per atom (dpa) in iron should be reported consistent with Practices E693 and E853 . 3.2.1.7 Reaction rates (preferably established relative to neutron fluence standards) must be reported for 237 Np(n,f) or 238 U(n,f), and 58 Ni(n,p) or 54 Fe(n,p); additional reactions that aid in spectral characterization, such as provided by Cu, Ti, and Co-Al, should also be included in the benchmark measurements. The 237 Np(n,f) reaction is particularly important because it is sensitive to the same neutron energy region as the iron dpa. Practices E693 and E853 and Guides E844 and E944 discuss this criterion. 3.2.2 Methodology Validation— It is essential that the neutronics methodology employed for predicting neutron fluence in a reactor pressure vessel be validated by accurately predicting appropriate benchmark dosimetry results. In addition, the following documentation should be submitted: ( 1 ) convergence study results, and ( 2 ) estimates of variances and covariances for fluence rates and reaction rates arising from uncertainties in both the source and geometric modeling. For Monte Carlo calculations, the convergence study results should also include ( 3 ) an analysis of the figure-of-merit (FOM) as a function of particles history, and if applicable, ( 4 ) the description of the technique utilized to generate the weight window parameters. 3.2.2.1 For example, model specifications for discrete-ordinates method on which convergence studies should be performed include: ( 1 ) neutron cross sections or energy group structure, ( 2 ) spatial mesh, and ( 3 ) angular quadrature. Reference ( 3 ) evaluates the effects of many discrete-ordinates parameters individually and in combination and may help guide the analysis. For regions adjacent to the reactor core, one-dimensional calculations may be performed to check the adequacy of group structure and spatial mesh. Two-dimensional calculations should be employed to check the adequacy of the angular quadrature. A P 3 cross section expansion is recommended along with a S 8 minimum quadrature. For regions that are not adjacent to the reactor core, convergence studies for spatial mesh and angular quadrature should apply three-dimensional calculations. 3.2.2.2 Uncertainties that are propagated from known uncertainties in nuclear data should be considered in the analysis. The uncertainty analysis for discrete ordinates codes may be performed with sensitivity analysis as discussed in References ( 4 , 5 ) . In Monte Carlo analysis the uncertainties can be treated by a perturbation analysis as discussed in Reference ( 6 ) . Appropriate computer programs and covariance data are available and sensitivity data may be obtained as an intermediate step in determining uncertainty estimates. 5 3.2.2.3 Effects of known uncertainties in geometry and source distribution should be evaluated based on the following test cases: ( 1 ) reference calculation with a time-averaged source distribution and with best estimates of the core and pressure vessel locations, ( 2 ) reference case geometry with maximum and minimum expected deviations in the source distribution, and ( 3 ) reference case source distribution with maximum expected spatial perturbations of the core, pressure vessel, and other pertinent locations. 3.2.2.4 Measured and calculated integral parameters should be compared for all test cases. It is expected that larger uncertainties are associated with geometry and neutron source specifications than with parameters included in the convergence study. Problems associated with space, energy, and angle discretizations can be identified and corrected. Uncertainties associated with geometry specifications are inherent in the structure tolerances. Calculations based on the expected extremes provide a measure of the sensitivity of integral parameters to the selected variables. Variations in the proposed convergence and uncertainty evaluations are appropriate when the above procedures are inconsistent with the methodology to be validated. As-built data could be used to reduce the uncertainty in geometrical dimensions. 3.2.2.5 In order to illustrate quantitative criteria based on measurements and calculations that should be satisfied, let ψ denote a set of logarithms of calculation ( C i ) to measurement ( E i ) ratios. Specifically, where q i and N are defined implicitly and the w i are weighting factors. Because some reactions provide a greater response over a spectral region of concern than other reactions, weighting factors may be utilized when their selection method is well documented and adequately defended, such as through a least-squares adjustment method as detailed in Guide E944 . In the absence of the use of a least-squares adjustment methodology, the mean of the set q is given by and the best estimate of the variance, S 2 , is 3.2.2.6 The neutronics methodology is validated if (in addition to qualitative model evaluation) all of the following criteria are satisfied: (1) The bias, | q |, is less than ε 1 , (2) The standard deviation, S , is less than ε 2 , (3) All absolute values of the natural logarithmic of the C / E ratios (| q |, i = 1 . N ) are less than ε 3 , and (4) ε 1 , ε 2 , and ε 3 are defined by the benchmark measurement documentation and demonstrated to be attainable for all items with which calculations are compared. 3.2.2.7 Note that a nonzero log-mean of the C i / E i ratios indicates that a bias exists. Possible sources of a bias are: ( 1 ) source normalization, ( 2 ) neutronics data, ( 3 ) transverse leakage corrections (if applicable), ( 4 ) geometric modeling, and ( 5 ) mathematical approximations. Reaction rates, equivalent fission fluence rates, or exposure parameter values (for example, φ( E > 1 MeV) and dpa) may be used for validating the computational methodology if appropriate criteria (that is, as established by 3.2.2.5 and 3.2.2.6 ) are documented for the benchmark of interest. Accuracy requirements for reactor vessel surveillance specific benchmark validation procedures are discussed in Guide E2006 . The validation testing for the generic discrete ordinates and Monte Carlo transport methods is discussed in References ( 1 , 2 ) . 3.2.2.8 One acceptable procedure for performing these comparisons is: ( 1 ) obtain group fluence rates at dosimeter locations from neutronics calculations, ( 2 ) collapse the Guide E1018 recommended dosimetry cross section data to a multigroup set consistent with the neutron energy group fluence rates or obtain a fine group spectrum (consistent with the dosimetry cross section data) from the calculated group fluence rates, ( 3 ) fold the energy group fluence rates with the appropriate cross sections, and ( 4 ) compare the calculated and experimental data according to the specified quantitative criteria. 3.3 Determination of the Fixed Fission Source— The power distribution in a typical reactor undergoes significant change during the life of the reactor. A time-averaged power distribution is recommended for use in determination of the neutron source distribution utilized for damage predictions. An adjoint procedure, described in 3.3.2 , may be more appropriate for dosimetry comparisons involving product nuclides with short half-lives. For multigroup methods, the fixed source may be determined from the equation: where: r = a spatial node, g = an energy group, v = average number of neutrons per fission, x g = fraction of the fission spectrum in group g , and P r = fission rate in node r . 3.3.1 Note that in addition to the fission rate, v and x g will vary with fuel burnup, and a proper time average of these quantities should be used. The ratio between fission rate and power (that is, fission/s per watt) will also vary with burnup for any given spatial node. 3.3.2 An adjoint procedure may be used as suggested in NUREG/CR-5049 instead of calculation with a time-averaged source calculation. 3.3.2.1 The influence of changing source distribution is discussed in Reference ( 8 ) . For dosimetry comparisons involving product nuclides with short half-lives, these changes in the power distribution may be significant. In this situation, a suitably averaged power distribution can be obtained by weighting the time-dependent power distribution using a factor proportional to: where: f = weighting factor at time, t , λ = decay constant for the nuclide of interest, and t = time from the start of the exposure. This averaging is different for each nuclide, therefore the use of the adjoint procedure avoids unecessary repetitions of the transport calculations in order to validate calculations using dosimetry results as described in 3.2.2 . 3.3.2.2 Care should be exercised to ensure that adjoint calculations adequately address cycle-to-cycle variations in coolant densities and any changes to the geometric configuration of the reactor. 3.4 Calculation of the Neutron Fluence Rate Based on a Fixed Source in the Reactor Core— The discussion in this section relates to methods validation calculations and to routine surveillance calculations. In either case, neutron transport calculations must estimate the neutron fluence rate in the core, through the internals, in the reactor pressure vessel, and outside the vessel, if for example, ex-vessel dosimetry is used. Procedures for methods validation differ very little from procedures for predicting neutron fluence rate in the pressure vessel or test facility; consequently, the following procedure is recommended: 3.4.1 Obtain detailed geometric and composition descriptions of the material configurations involved in the transport calculation. Uncertainty in the data should also be estimated. 3.4.2 Obtain applicable cross section sets from appropriate data bases such as: 3.4.2.1 The evaluated nuclear data file (ENDF/B or its equivalent), or 3.4.2.2 A fine group library obtained by processing the above file (for example, see Reference ( 9 ) ). 3.4.3 Perform a one-dimensional, fixed-source, fine-group calculation in order to collapse the fine-group cross sections to a broad-group set for multidimensional calculations. At least two broad-group sets are recommended for performing the one-dimensional group structure convergence evaluation. The broad-group structure should emphasize the high-energy range and should take cross section minima of important materials (for example, iron) into consideration. 3.4.4 Perform the convergence studies outlined in 3.2.2 . 3.4.5 Perform two- or three-dimensional fixed-source transport calculations based on the model established in 3.4.1 – 3.4.4 . 3.4.6 Compare appropriate dosimetry results with neutronics results from 3.4.5 according to the procedure given in 3.2.2 . It is recommended that all valid lifetime-accumulated reactor dosimetry data be included in this comparison each time new data become available except when dosimeter-specific comparisons are made. 3.4.7 Repeat appropriate steps if validation criteria are not satisfied. Note that a reactor dosimetry datum may be discarded if the associated C / E ratios differ substantially from the average of the applicable C / E ratios and a measurement error can be suspected. A measurement error can be suspected if the deviation from the average exceeds the equivalent of three standard deviations. In addition, the source for reactor calculations may be scaled to minimize the bias and variance defined by Eq 2 and Eq 3 provided that data are not discarded as a consequence of scaling the source. 3.4.8 Results from neutronics calculations may be used in a variety of ways: 3.4.8.1 Determine a single normalization constant that minimizes bias in the calculated values relative to the measurements in order to scale the group fluences. This is a simple and frequently used alternative to adjustment procedures. However, the magnitude of this constant should be critically examined in terms of estimated source uncertainties. 3.4.8.2 Use a spectrum adjustment procedure as recommended in Guide E944 using calculated group fluences and dosimetry data with uncertainty estimates to obtain an adjustment to the calculated group fluences and exposure parameters. Predicted pressure vessel fluences could then incorporate the spectral and normalization data obtained from the adjusted fluences. 3.4.8.3 Use the calculated fluence spectrum with Practice E693 for damage exposure predictions. 3.4.8.4 It is expected that in some cases the procedure recommended above will be inconsistent with some methodologies to be validated. In these cases procedural variations are appropriate but should be well documented.