1.1 本指南描述了被动中子测量方法，用于无损地估计核设施中作为滞留物的中子发射特殊核材料化合物的量。所有处理核材料的设施都会发生滞留。例如，材料可能存在于工艺设备、排气通风系统以及建筑墙壁和地板中。 1.1.1 被动中子持率技术最常用于测量加工设施中的铀或钚矿床。 1.2 本指南包括对管理、规划、设备选择、干扰考虑、测量程序定义和资源利用有用的信息。 1.3 计数模式包括单计数（总数）或总计数和中子重合技术。 1.3.1 铀的中子持率测量通常使用单个（总数）或总计数对（α，n）反应和自发裂变过程中发射的中子进行。虽然该方法不排除使用重合或多重计数对铀进行测量，但测量效率通常不足以在合理的计数时间内进行测定。 1.3.2 对于手套箱中钚的测量，安装的测量设备可以在合理的计数时间内为使用中子重合技术进行计数提供足够的效率。 1.4 测量工艺设备中的核材料滞留量需要掌握辐射源和探测器、辐射传输、建模方法、校准、设施操作和不确定性分析的科学知识。它受制于设施、管理、预算和时间表的限制，加上健康和安全要求以及物理定律。本指南并不旨在指导NDA从业者遵守这些原则。 1.5 测量过程包括定义测量不确定性，并对化学成分、同位素成分、材料分布、各种背景和干扰敏感。 这项工作包括调查设施内的材料分布，其中可能包括潜在的大滞留表面积。管道、管道系统、手套箱和重型设备中的核材料通常以分散和不规则的方式分布。很难定义测量几何形状，识别材料的形状并进行测量。 1.6 单位-- 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 本标准并不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====== 5.1 本指南有助于满足保障措施、SNM库存控制、核临界安全、废物处理以及去污和退役（D&D）等领域的要求。本指南可适用于可测量或估计中子产生特性的工艺设备或离散项目中的滞留率测量。 这些方法可以满足在存在调节剂、吸收剂和中子毒物的情况下具有复杂SNM分布的项目的目标精度；然而，与不太复杂项目的测量相比，结果受到更大的测量不确定性的影响。 5.2 定量测量-- 这些测量结果导致滞留物中SNM质量的量化。它们包括所有可用的校正和描述性信息，如同位素组成。 5.2.1 高质量的结果需要详细了解辐射源和探测器、辐射传输、校准、设施操作和误差分析。 建议咨询合格的NDA人员（指南 C1490 )。 5.2.2 单个工艺设备或管道的滞留估计通常包括多个测量值的汇编。滞留率估计必须适当地结合每个单独测量的结果。此外，必须对每个单独测量的不确定度进行估计，并进行适当组合。 5.3 扫描-- 辐射扫描，通常为伽马，可用于提供滞留范围、位置和相对量的定性描述。它可以用于计划或补充定量中子测量。 其他指示器（例如，视觉指示器）也可以指示对滞留量测量的需要。 5.4 核素映射-- 为了适当地解释中子数据，需要特定的中子产额。根据设施的不同，可能需要进行同位素测量，以确定特定位置滞留物的相对同位素组成。 5.5 抽查和验证测量-- 使用相同的技术和假设定期重新测量定义点的滞留量，可用于检测或跟踪该点滞留量随时间的相对变化。可以使用定性或定量方法。 5.6 间接测量-- 中子测量无法确定产生中子信号的放射性核素。应独立于中子测量来确定比中子产额。 5.7 建模-- 建议将建模作为评估复杂测量情况的辅助手段。测量数据与辐射传输模型一起使用，该模型包括设备和材料的物理位置描述。由于中子输运计算的复杂性，通常使用MCNP等输运代码来开发模型。也可以使用几何模型，但通常不考虑散射和中子逃逸分数估计等现象。

1.1 This guide describes passive neutron measurement methods used to nondestructively estimate the amount of neutron-emitting special nuclear material compounds remaining as holdup in nuclear facilities. Holdup occurs in all facilities in which nuclear material is processed. Material may exist, for example, in process equipment, in exhaust ventilation systems, and in building walls and floors. 1.1.1 The most frequent uses of passive neutron holdup techniques are for the measurement of uranium or plutonium deposits in processing facilities. 1.2 This guide includes information useful for management, planning, selection of equipment, consideration of interferences, measurement program definition, and the utilization of resources. 1.3 Counting modes include both singles (totals) or gross counting and neutron coincidence techniques. 1.3.1 Neutron holdup measurements of uranium are typically performed on neutrons emitted during (α, n) reactions and spontaneous fission using singles (totals) or gross counting. While the method does not preclude measurement using coincidence or multiplicity counting for uranium, measurement efficiency is generally not sufficient to permit assays in reasonable counting times. 1.3.2 For measurement of plutonium in gloveboxes, installed measurement equipment may provide sufficient efficiency for performing counting using neutron coincidence techniques in reasonable counting times. 1.4 The measurement of nuclear material holdup in process equipment requires a scientific knowledge of radiation sources and detectors, radiation transport, modeling methods, calibration, facility operations, and uncertainty analysis. It is subject to the constraints of the facility, management, budget, and schedule, plus health and safety requirements, as well as the laws of physics. This guide does not purport to instruct the NDA practitioner on these principles. 1.5 The measurement process includes defining measurement uncertainties and is sensitive to the chemical composition, isotopic composition, distribution of the material, various backgrounds, and interferences. The work includes investigation of material distributions within a facility, which could include potentially large holdup surface areas. Nuclear material held up in pipes, ductwork, gloveboxes, and heavy equipment is usually distributed in a diffuse and irregular manner. It is difficult to define the measurement geometry, identify the form of the material, and measure it. 1.6 Units— The values stated in SI units are to be regarded as the 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 ====== 5.1 This guide assists in satisfying requirements in such areas as safeguards, SNM inventory control, nuclear criticality safety, waste disposal, and decontamination and decommissioning (D&D). This guide can apply to the measurement of holdup in process equipment or discrete items whose neutron production properties may be measured or estimated. These methods may meet target accuracy for items with complex distributions of SNM in the presence of moderators, absorbers, and neutron poisons; however, the results are subject to larger measurement uncertainties than measurements of less complex items. 5.2 Quantitative Measurements— These measurements result in quantification of the mass of SNM in the holdup. They include all the corrections and descriptive information, such as isotopic composition, that are available. 5.2.1 High-quality results require detailed knowledge of radiation sources and detectors, radiation transport, calibration, facility operations, and error analysis. Consultation with qualified NDA personnel is recommended (Guide C1490 ). 5.2.2 Holdup estimates for a single piece of process equipment or piping often include some compilation of multiple measurements. The holdup estimate must appropriately combine the results of each individual measurement. In addition, uncertainty estimates for each individual measurement must be made and appropriately combined. 5.3 Scan— Radiation scanning, typically gamma, may be used to provide a qualitative description of the extent, location, and the relative quantity of holdup. It can be used to plan or supplement the quantitative neutron measurements. Other indicators (for example, visual) may also indicate a need for a holdup measurement. 5.4 Nuclide Mapping— To appropriately interpret the neutron data, the specific neutron yield is needed. Isotopic measurements to determine the relative isotopic composition of the holdup at specific locations may be required, depending on the facility. 5.5 Spot Check and Verification Measurements— Periodic re-measurement of holdup at a defined point using the same technique and assumptions can be used to detect or track relative changes in the holdup quantity at that point over time. Either a qualitative or quantitative method can be used. 5.6 Indirect Measurements— Neutron measurements do not identify the radionuclide that produced the neutron signal. The specific neutron yield shall be determined independently of the neutron measurement. 5.7 Modeling— Modeling is recommended as an aid in the evaluation of complex measurement situations. Measurement data are used with a radiation transport model that includes a description of the physical location of equipment and materials. Because of the complexity of neutron transport calculations, models are often developed using a transport code such as MCNP. Geometric models can also be used but, generally, do not account for phenomena such as scattering and estimation of neutron escape fraction.