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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.

1.2 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.

1.4 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 and health practices and determine the applicability of regulatory limitations prior to 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.

5.3 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.