C26.10 - Non Destructive Assay

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 an...
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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.  
1.1.1 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...

SIGNIFICANCE AND USE
4.1 The determination of uranium isotopic composition by gamma-ray spectrometry is a nondestructive technique and when used with other nondestructive techniques that quantify a single isotope, such as Test Methods C1133 (Segmented Gamma Scanning), C1221 (Solution Assay), C1455 (Holdup),and C1718 (Tomographic Gamma Scanning), can provide a wholly nondestructive assay of uranium mass necessary for material accountancy and safeguards needs. This method can be used with calorimetry (Test Method C1458) for kilogram quantities of high-enriched uranium and is also used to convert an Active-Well Coincidence Counter (4) measurement of 235U mass to total uranium mass.  
4.2 Because gamma-ray spectrometry systems are typically automated, the routine use of the test method is fast, reliable, and is not labor intensive. The test method is nondestructive, requires no sample preparation, and does not create waste disposal problems.  
4.3 The test method does not require that the system be calibrated to a specific geometry.  
4.4 The test method assumes that all uranium in the measured item has the same isotopic distribution. This is often termed isotopic homogeneity.  
4.5 The application of the test method does not depend upon the physical or chemical form of the material being analyzed.  
4.6 The 236U abundance is not measured by this test method and must be estimated from isotopic correlation techniques, stream averages, historical information, or other measurement techniques.  
4.7 The isotopic composition of a given item of uranium is an attribute of that item and, once determined, can be used in subsequent inventory measurements to verify the identity of an item within the measurement uncertainties.  
4.8 The method can also measure the ratio of other gamma-emitting isotopes in the measured item to uranium assuming they have the same spatial distribution as the uranium in the item. Some of these “other” gamma-emitting isotopes include daughter isotopes of uranium, ces...
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1.1 This test method applies to the nondestructive determination of the isotopic abundances of uranium, typically 234U, 235U, 236U, and 238U, in isotopically homogeneous uranium-bearing materials using gamma spectrometry. The material is commonly inside a container and is measured without specimen preparation.  
1.2 This test method is applicable to items containing sub-gram quantities of uranium to the maximum uranium mass allowed by criticality considerations.  
1.3 Measurable gamma ray emissions from uranium cover the energy range from below 80 keV to above 1000 keV. K-X-ray emissions from the isotopes of uranium and their daughters are found in the energy region around 100 keV. This test method has been applied to all portions of this energy range.  
1.4 The isotopic abundance of 236U is usually not directly determined because its low-energy gamma rays are too weak (1)2 to be detected under normal measurement conditions. Isotopic correlation techniques have been used to estimate its relative abundance (2).  
1.5 This test method has been demonstrated in routine use for isotopic amount fraction (atom %) of 235U from 0.2 % to 97 %.  
1.6 This test method requires decay equilibrium (160 days for 99 %) between 238U and its 24.1 d half-life 234Th daughter. Corrections can be made if the date of chemical separation of the 234Th daughter is known.  
1.7 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.8 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.  
1.9 This international standard was developed in accordance with internationally recognized principles on standardization established in the Dec...

SIGNIFICANCE AND USE
4.1 The process of selection, training and qualification of personnel involved with NDA measurements is one of the quality assurance elements for an overall quality NDA measurement program.  
4.2 This guide describes an approach to selection, qualification, and training of personnel that is to be used in conjunction with other NDA Quality Assurance (QA) program elements. The selection, qualification and training processes can vary and this guide provides one such approach.  
4.3 The qualification activities described in this guide assume that NDA personnel are already proficient in general facility operations and safety procedures. The training and activities that developed this proficiency are not covered in this guide.  
4.4 This guide describes a basic approach and principles for the qualification of NDA professionals and technical specialists and operators. A different approach may be adopted by the management organization based on its particular organization and facility specifics. However, if a variation of the approach of this guide is applied, the resulting selection, training, and qualification programs must meet the requirements of the facility quality assurance program and should provide all the applicable functions of Section 5.  
4.5 This guide may be used as an aid in the preparation of a Training Implementation Plan (TIP) for the Transuranic Waste Characterization Program (TWCP).  
4.6 This guide describes education and expertise guidance for NDA auditors due to the importance and complexity of proper oversight of NDA activities.
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1.1 This guide contains good practices for the selection, training, qualification, and professional development of personnel performing analysis, calibration, physical measurements, or data review using nondestructive assay equipment, methods, results, or techniques. The guide also covers NDA personnel involved with NDA equipment setup, selection, diagnosis, troubleshooting, or repair. General guidelines for the selection, training, and qualification of NDA auditors are included as well, but at a lower level of detail due to the variability of the personnel’s responsibilities performing this functions. Selection, training, and qualification programs based on this guide are intended to provide assurance that NDA personnel are suitably qualified and experienced personnel (SQEP) to perform their jobs competently. This guide presents a series of options but does not recommend a specific course of action.
This standard guide does not address the qualifications per se of an NDA Manager. However, it is expected that the NDA Manager is familiar with NDA techniques, and can make informed decisions on the acceptability of the assay results. If an NDA Manager does not have adequate technical qualifications in the NDA field, they are recommended to undergo training to gain familiarity in this area.
An NDA Manager with no relevant NDA experience should have access to a Senior NDA Professional who will give guidance for all technical decisions such as applicability and limitation of methods, reasonableness of results, needed upgrades and advantageous development investments.  
1.2 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 Measurement results from this test method assists in demonstrating regulatory compliance in such areas as safeguards SNM inventory control, criticality control, waste disposal, and decontamination and decommissioning (D&D). This test method can apply to the measurement of holdup in process equipment or discrete items whose gamma-ray absorption properties may be measured or estimated. This method may be adequate to accurately measure items with complex distributions of radioactive and attenuating material, however, the results are subject to larger measurement uncertainties than measurements of less complex distributions of radioactive material.  
5.2 Scan—A scan is used to provide a qualitative indication of the extent, location, and the relative quantity of holdup. It can be used to plan or supplement the quantitative measurements.  
5.3 Nuclide Mapping—Nuclide mapping measures the relative isotopic composition of the holdup at specific locations. It can also be used to detect the presence of radionuclides that emit radiation which could interfere with the assay. Nuclide mapping is best performed using a high resolution detector (such as HPGe) for best nuclide and interference detection. If the holdup is not isotopically homogeneous at the measurement location, that measured isotopic composition will not be a reliable estimate of the bulk isotopic composition.  
5.4 Quantitative Measurements—These measurements result in quantification of the mass of the measured nuclides in the holdup. They include all the corrections, such as attenuation, and descriptive information, such as isotopic composition, that are available  
5.4.1 High quality results require detailed knowledge of radiation sources and detectors, transmission of radiation, calibration, facility operations and error analysis. Judicious use of subject matter experts is required (Guide C1490).  
5.5 Holdup Monitoring—Periodic re-measurement of holdup at a defined point using the same technique and a...
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1.1 This test method describes gamma-ray methods used to nondestructively measure the quantity of 235U or  239Pu present as holdup in nuclear facilities. Holdup may occur in any facility where nuclear material is processed, in process equipment, in exhaust ventilation systems and in building walls and floors.  
1.2 This test method includes information useful for management, planning, selection of equipment, consideration of interferences, measurement program definition, and the utilization of resources  (1, 2, 3, 4) .2  
1.3 The measurement of nuclear material hold up in process equipment requires a scientific knowledge of radiation sources and detectors, transmission of radiation, calibration, facility operations and uncertainty analysis. It is subject to the constraints of the facility, management, budget, and schedule; plus health and safety requirements. The measurement process includes defining measurement uncertainties and is sensitive to the form and 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, to identify the form of the material, and to measure it without interference from adjacent sources of radiation.  
1.4 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.  
1.5 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles...

SIGNIFICANCE AND USE
4.1 NDA measurement practices aimed at achieving quality results are described in this guide. The application of the material provided in this guide should be determined on a case by case basis. Not all elements are required for all applications.  
4.2 Nondestructive assay measurements are typically performed when the items measured or goals of the measurement program favor or require NDA over destructive analysis. NDA is typically favored when collecting a representative sample of the item is difficult or impractical (for example, scrap and waste items), personnel exposure would be significant, spread of contamination from sampling would occur, generation of secondary waste must be minimized, the weight and/or tare weight of the item cannot easily be determined (for example, in place process equipment), rapid turn-around of the measurement results is needed, or the NDA measurement is significantly less expensive than the equivalent destructive analysis.  
4.3 The principles provided in this guide should be used to determine which type of measurement is best suited to the measurement application. This determination involves consideration of the characteristics of the items to be measured, as well as the goals of the measurement program.  
4.4 This guide applies to the suite of NDA instruments and measurement methods, many of which are described in detail in Refs (1) and (2) .7 A partial listing of measurement methods and applicable use references is provided in 5.5.1. It is incumbent upon the user to seek additional guidance within ASTM method-specific standards, as this guide does not take precedence. Additional information on specific methods is best found in technical meeting transactions, journals, commercial application notes, and NRC/DOE publications.  
4.5 This guide may be applied to many situations spanning the range of nuclear materials from product through waste. Typical applications include: the measurement and characterization of transuranic wastes,...
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1.1 This guide is a compendium of Quality Measurement Practices for performing measurements of radioactive material using nondestructive assay (NDA) instruments. The primary purpose of the guide is to assist users in arriving at quality NDA results, that is, results that satisfy the end user’s needs. This is accomplished by providing an acceptable and uniform basis for the collection, analysis, comparison, and application of data. The recommendations are guidelines to achieving quality NDA measurements in most areas.  
1.2 This guide applies to the use of NDA instrumentation for the measurement of nuclear materials by the observation of spontaneous or stimulated nuclear or atomic radiations, including photons, neutrons, or heat. Recommended calibration, operating, and assurance methods represent guiding principles based on current NDA technology. The diversity of industry-wide nuclear materials measurement applications and instrumentation precludes discussion of specific measurement situations. As a result, compliance with practices recommended in this guide must be based on a thorough understanding of contributing variables and performance requirements of the specific measurement application.  
1.3 Selection of the best instrument for a given measurement application and advice on the use of this instrument must be provided by a qualified NDA professional following guidance provided in Guide C1490. This guide is to be used as a reference, and to supplement the critical thinking, professional skill, expert judgment, and experimental test and verification needed to ensure that the instrumentation and methods have been properly implemented.  
1.4 The intended audience for this guide includes but is not limited to Management, Auditor Support, NDA Qualified Instrument Operators, NDA Technical Specialists, and NDA Professionals.  
1.5 The values stated in either SI units or inch-pound units are to be regarded separately as...

SIGNIFICANCE AND USE
5.1 This test method is useful for quantifying fissile (for example, 233U, 235U,  239Pu and 241Pu) and spontaneously-fissioning nuclei (for example, 238Pu, 240Pu,  242Pu, 244Cm,
248Cm, and  252Cf) in waste and scrap drums. Total elemental mass of the radioactive materials can be calculated if the relative abundances of each radionuclide are known.  
5.1.1 Typically, this test method is used to measure one fissile isotope (for example, 235U or  239Pu).  
5.2 This test method can be used to segregate low level and transuranic waste at the 100 nCi/g concentration level currently required to meet the DOE Waste Isolation Pilot Plant (WIPP) waste acceptance criterion (5, 8, 9).  
5.3 This test method can be used for waste characterization to demonstrate compliance with the radioactivity levels specified in waste, disposal, and environmental regulations (See NRC regulatory guides, DOE Order 435.1, 10 CFR Part 71, 40 CFR Part 191, and DOE /WIPP-069).  
5.3.1 In the active mode, the DDT system can measure the 235U content in the range from 100 g and the 239Pu content, nominally between 20 g.  
5.3.2 In the passive mode, the DDT system is capable of assaying spontaneously-fissioning nuclei, over a nominal range from 0.05 to 15 g of 240Pu, or equivalent (5, 10, 11, 12, 13).  
5.4 This test method should be used in conjunction with a waste management plan that segregates the contents of assay items into material categories according to some or all of the following criteria: bulk density of the waste, chemical forms of the plutonium or uranium and matrix, (α, n) neutron intensity, hydrogen (moderator) and absorber content, thickness of fissile mass(es), and the assay item container size and composition. Each matrix may require a different set of calibration standards and may have different mass calibration limits. The effect on the quality of the assay (that is, minimizing precision and bias) can significantly depend on the degree of adherence to this waste management pl...
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1.1 This test method covers a system that performs nondestructive assay (NDA) of uranium or plutonium, or both, using the active, differential die-away technique (DDT), and passive neutron coincidence counting. Results from the active and passive measurements are combined to determine the total amount of fissile and spontaneously-fissioning material in drums of scrap or waste. Corrections are made to the measurements for the effects of neutron moderation and absorption, assuming that the effects are averaged over the volume of the drum and that no significant lumps of nuclear material are present. These systems are most widely used to assay low-level and transuranic waste, but may also be used for the measurement of scrap materials. The examples given within this test method are specific to the second-generation Los Alamos National Laboratory (LANL) passive-active neutron assay system.  
1.1.1 In the active mode, the system measures fissile isotopes such as 235U and 239Pu. The neutrons from a pulsed, 14-MeV neutron generator are thermalized to induce fission in the assay item. Between generator pulses, the system detects prompt-fission neutrons emitted from the fissile material. The number of detected neutrons between pulses is proportional to the mass of fissile material. This method is called the differential die-away technique.  
1.1.2 In the passive mode, the system detects time-coincident neutrons emitted from spontaneously fissioning isotopes. The primary isotopes measured are 238Pu, 240 Pu, and 242Pu; however, the system may be adapted for use on other spontaneously-fissioning isotopes as well, such as kilogram quantities of 238U. The number of coincident neutrons detected is proportional to the mass of spontaneously-fissioning material.  
1.2 The active mode is used to assay fissile material in the following ranges.  
1.2.1 For uranium-only bearing items, the DDT can measure the 235U content in the range from...

SIGNIFICANCE AND USE
5.1 The TGS provides a nondestructive means of mapping the attenuation characteristics and the distribution of the radionuclide content of items on a voxel by voxel basis. Typically in a TGS analysis a vertical layer (or segment) of an item will be divided into a number of voxels. By comparison, a segmented gamma scanner (SGS) can determine matrix attenuation and radionuclide concentrations only on a segment by segment basis.  
5.2 It has been successfully used to quantify  238Pu, 239Pu, and 235U. SNM loadings from 0.5 g to 200 g of 239Pu (5, 6), from 1 g to 25 g of 235U (7), and from 0.1 to 1 g of 238Pu have been successfully measured. The TGS technique has also been applied to assaying radioactive waste generated by nuclear power plants (NPP). Radioactive waste from NPP is dominated by activation products (for example, 54Mn, 58Co, 60Co,  110mAg) and fission products (for example, 137Cs,  134Cs). The radionuclide activities measured in NPP waste is in the range from 3.7E+04 Bq to 1.0E+07 Bq. Some results of TGS application to non-SNM radionuclides can be found in the literature  (8).  
5.3 The TGS technique is well suited for assaying items that have heterogeneous matrices and that contain a non-uniform radionuclide distribution.  
5.4 Since the analysis results are obtained on a voxel by voxel basis, the TGS technique can in many situations yield more accurate results when compared to other gamma ray techniques such as SGS.  
5.5 In determining the radionuclide distribution inside an item, the TGS analysis explicitly takes into account the cross talk between various vertical layers of the item.  
5.6 The TGS analysis technique uses a material basis set method that does not require the user to select a mass attenuation curve apriori, provided the transmission source has at least 2 gamma lines that span the energy range of interest.  
5.7 A commercially available TGS system consists of building blocks that can easily be configured to operate the system in the ...
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1.1 This test method describes the nondestructive assay (NDA) of gamma ray emitting radionuclides inside containers using tomographic gamma scanning (TGS). High resolution gamma ray spectroscopy is used to detect and quantify the radionuclides of interest. The attenuation of an external gamma ray transmission source is used to correct the measurement of the emission gamma rays from radionuclides to arrive at a quantitative determination of the radionuclides present in the item.  
1.2 The TGS technique covered by the test method may be used to assay scrap or waste material in cans or drums in the 1 to 500 litre volume range. Other items may be assayed as well.  
1.3 The test method will cover two implementations of the TGS procedure: (1) Isotope Specific Calibration that uses standards of known radionuclide masses (or activities) to determine system response in a mass (or activity) versus corrected count rate calibration, that applies to only those specific radionuclides for which it is calibrated, and (2) Response Curve Calibration that uses gamma ray standards to determine system response as a function of gamma ray energy and thereby establishes calibration for all gamma emitting radionuclides of interest.  
1.4 This test method will also include a technique to extend the range of calibration above and below the extremes of the measured calibration data.  
1.5 The assay technique covered by the test method is applicable to a wide range of item sizes, and for a wide range of matrix attenuation. The matrix attenuation is a function of the matrix composition, photon energy, and the matrix density. The matrix types that can be assayed range from light combustibles to cemented sludge or concrete. It is particularly well suited for items that have heterogeneous matrix material and non-uniform radioisotope distributions. Measured transmission values should be available to permit valid attenuation corrections, but are not n...

SIGNIFICANCE AND USE
5.1 SNM monitors are an effective means to search pedestrians for concealed SNM. Maintaining monitor effectiveness rests on appropriate calibration and adjustment being part of a continuing maintenance program.  
5.2 The significance of this guide for monitor users who must detect SNM is to describe calibration and adjustment procedures for the purpose.  
5.3 The significance of this guide for monitor manufacturers is to describe calibration procedures, particularly for detecting forms of SNM that may not be readily available to them.
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1.1 This guide covers calibrating the energy response of the radiation detectors and setting the discriminator and alarm thresholds used in automatic pedestrian special nuclear material (SNM) monitors.  
1.2 Automatic pedestrian SNM Monitors and their application are described in Guide C1112, which suggests that the monitors be calibrated and tested when installed and that, thereafter, the calibration should be checked and the monitor tested with SNM at three-month intervals.  
1.3 Dependable operation of SNM monitors rests, in part, on an effective program to test, calibrate, and maintain them. The procedures and methods described in this guide may help both to achieve dependable operation and obtain timely warning of misoperation.  
1.4 This guide can be used in conjunction with other ASTM standards. Fig. 1 illustrates the relationship between calibration and other procedures described in standard guides, and it also shows how the guides relate to an SNM monitor user. The guides below the user in the figure deal with routine procedures for operational monitors. Note that Guide C993 is an in-plant performance evaluation that is used to verify acceptable detection of SNM after a monitor is calibrated. The guides shown above the user in Fig. 1 give information on applying SNM monitors (C1112) and on evaluating SNM monitors (C1169) to provide comparative information on monitor performance.  
FIG. 1 The Relationship of Calibration to Other Procedures Described in Standard Guides for SNM Monitors  
1.5 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 The HKED technique is highly element specific and depends upon a well-known controlled geometry.  
5.2 The HKED technique can provide concentration measurements of actinides in solutions with precision typically better than 0.3 % for uranium concentrations >50 g/L and 1 % for plutonium in typical U-Pu solutions for a typical measurement time of 3 × 1000 s (3 replicates, 1000 s live time each) (1).  
5.3 For pure plutonium only product solutions, the KED technique can achieve measurement precisions better than 0.3 % for plutonium concentrations >50 g/L for a typical measurement time of 3 × 1000 s.  
5.4 For pure uranium only solutions, precisions of better than 0.3 % can be achieved using the KED technique for uranium concentrations >50 g/L, for a typical measurement time of 3 × 3600 s.  
5.5 For uranium only or plutonium only solutions of concentrations approximately 1 g/L, assayed using XRF, a measurement precision of 1.0 % has been achieved (1). For solutions of concentration approximately 50 g/L, assayed using XRF, measurement precisions of 0.2 % or better have been achieved. The typical measurement time for stand-alone XRF assay is 3 × 3000 s.  
5.6 Quality Control (QC) samples are assayed for a typical measurement time of 3 × 3000 s.  
5.7 It is applicable when solutions to be measured are homogeneous with respect to chemical composition.  
5.8 Results are typically used for fuel fabrication, process control, quality control, material control and accountancy, and safeguards in nuclear fuel reprocessing plants. Each application can have its own data quality objectives (Guide C1068).  
5.9 The HKED instrument may use a single cylindrical vial for both the KED and XRF measurements, or separate sample containers for KED and XRF. The typical values for the path length of the rectangular cuvette and the inner diameter of the cylindrical vial are given in 7.8.  
5.10 The transfer of the sample into the HKED system can be accomplished either horizontally b...
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1.1 This test method specifies the determination of the volumetric uranium and plutonium concentrations, typically, in nitric acid solutions through the combination of K-Edge absorption Densitometry (KED) and K X-Ray fluorescence (XRF) using an X-Ray generator. It is known as the “Hybrid K-Edge” (HKED) technique whose original implementation is described in Ref (1).2 The method is applicable to dissolver (input) solutions and product solutions. The test method also specifies the determination of low concentrations (  
1.2 This test method is applicable to the following common-use conditions:  
1.2.1 Spent nuclear fuel reprocessing and fuel production.  
1.2.2 Homogeneous aqueous solutions contained in cylindrical vials or cuvettes. HKED systems may use two separate sample containers, namely a rectangular cuvette for KED and a cylindrical vial for XRF. Alternatively, there are HKED systems that use a sample contained in a single cylindrical vial, for both K-Edge and XRF.  
1.2.3 The results produced by the two sample configuration (a rectangular cuvette for K-Edge densitometry and a cylindrical vial for XRF) are compliant with the International Target Values (ITV) (1).  
1.2.4 The precision results produced by the single cylindrical vial configuration are degraded in comparison to the two container system.  
1.2.5 This test method is applicable to facilities that do not adopt the ITVs, but have their own Data Quality Objectives (DQO).  
1.2.6 Solutions which contain uranium and plutonium with uranium concentration of 150 to 250 g/L and a U:Pu ratio of 100:1 typically, in the presence of fission products with β, γ, activity of up to 10 TBq/L.
1.2.6.1 This test method is not applicable to samples where a minor element such as U needs to be quantified in which Pu is the major element.
1.2.6.2 This test method is applicable for common use process control applications for quantifying Pu in the 5 g/L to 30 g/L range ...

SIGNIFICANCE AND USE
5.1 This test method is a nondestructive means of determining the nuclide concentration of a solution for special nuclear material accountancy, nuclear safety, and process control.  
5.2 It is assumed that the nuclide to be analyzed is in a homogeneous solution (Practice C1168).  
5.3 The transmission correction makes the test method independent of matrix (solution elemental composition and density) and useful over several orders of magnitude of nuclide concentrations. However, a typical configuration will normally span only two to three orders of magnitude because of detector dynamic range.  
5.4 The test method assumes that the solution-detector geometry is the same for all measured items. This can be accomplished by requiring that the liquid height in the sidelooking geometry exceeds the detector field of view defined by the collimator. For the upward-looking geometry, a fixed solution fill height must be maintained and vials of identical radii must be used unless the vial radius exceeds the field of view defined by the collimator.  
5.5 Since gamma-ray systems can be automated, the test method can be rapid, reliable, and not labor intensive.  
5.6 This test method may be applicable to in-line or off-line situations.
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1.1 This test method covers the determination of the concentration of gamma-ray emitting special nuclear materials dissolved in homogeneous solutions. The test method corrects for gamma-ray attenuation by the solution and its container by measurement of the transmission of a beam of gamma rays from an external source (Refs. (1), (2), and (3)).2  
1.2 Two solution geometries, slab and cylinder, are considered. The solution container that determines the geometry may be either a removable or a fixed geometry container. This test method is limited to solution containers having walls or a top and bottom of equal transmission through which the gamma rays from the external transmission correction source must pass.  
1.3 This test method is typically applied to radionuclide concentrations ranging from a few milligrams per litre to several hundred grams per litre. The assay range will be a function of the specific activity of the nuclide of interest, the physical characteristics of the solution container, counting equipment considerations, assay gamma-ray energies, solution matrix, gamma-ray branching ratios, and interferences.  
1.4 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. For specific hazards, see Section 9.  
1.5 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 test method is useful for determining the plutonium content of scrap and waste in containers ranging from small cans with volumes of the order of a mL to crates and boxes of several thousand liters in volume. A common application would be to 208-L (55-gal) drums. Total Pu content ranges from 10 mg to 6 kg (1). The upper limit may be restricted depending on specific matrix, calibration material, criticality safety, or counting equipment considerations.  
5.2 This test method is applicable for U.S. Department of Energy shipper/receiver confirmatory measurements (9), nuclear material diversion detection, and International Atomic Energy Agency attributes measurements (10).  
5.3 This test method should be used in conjunction with a scrap and waste management plan that segregates scrap and waste assay items into material categories according to some or all of the following criteria: bulk density, the chemical forms of the plutonium and the matrix, americium to plutonium isotopic ratio, and hydrogen content. Packaging for each category should be uniform with respect to size, shape, and composition of the container. Each material category might require calibration standards and may have different Pu mass limits.  
5.4 Bias in passive neutron coincidence measurements is related to item size and density, the homogeneity and composition of the matrix, and the quantity and distribution of the nuclear material. The precision of the measurement results is related to the quantity of nuclear material, the (α,n) reaction rate, and the count time of the measurement.  
5.4.1 For both benign matrix and matrix specific measurements, the method assumes the calibration reference materials match the items to be measured with respect to the homogeneity and composition of the matrix, the neutron moderator and absorber content, and the quantity of nuclear material, to the extent they affect the measurement.  
5.4.2 Measurements of smaller containers containing scrap and waste a...
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1.1 This test method describes the nondestructive assay of scrap or waste for plutonium content using passive thermal-neutron coincidence counting. This test method provides rapid results and can be applied to a variety of carefully sorted materials in containers as large as several thousand liters in volume. The test method applies to measurements of  238Pu, 240Pu, and  242Pu and has been used to assay items whose total plutonium content ranges from 10 mg to 6 kg (1) .2  
1.2 This test method requires knowledge of the relative abundances of the Pu isotopes to determine the total Pu mass (Test Method C1030).  
1.3 This test method may not be applicable to the assay of scrap or waste containing other spontaneously fissioning nuclides.  
1.3.1 This test method may give biased results for measurements of containers that include large amounts of hydrogenous materials.  
1.3.2 The techniques described in this test method have been applied to materials other than scrap and waste (2, 3).  
1.4 This test method assumes the use of shift-register-based coincidence technology (4).  
1.5 Several other techniques that are often encountered in association with passive neutron coincidence counting exist. These include neutron multiplicity counting (5, 6, Test Method C1500), add-a-source analysis for matrix correction (7), flux probes also for matrix compensation, cosmic-ray rejection (8)  to improve precision close to the detection limit, and alternative data collection electronics such as list mode data acquisition. Passive neutron coincidence counting may also be combined with certain active interrogation schemes as in Test Methods C1316 and C1493. Discussions of these established techniques are not included in this 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, an...

SIGNIFICANCE AND USE
5.1 Segmented gamma-ray scanning provides a nondestructive means of measuring the nuclide content of scrap and waste where the specific nature of the matrix and the chemical form and relationship between the nuclide and matrix may be unknown.  
5.2 The procedure can serve as a diagnostic tool that provides a vertical profile of transmission and nuclide concentration within the item.  
5.3 Item preparation is generally limited to good waste/scrap segregation practices that produce relatively homogeneous items that are required for any successful waste/inventory management and assay scheme, regardless of the measurement method used. Also, process knowledge should be used, when available, as part of a waste management program to complement information on item parameters, container properties, and the appropriateness of calibration factors.  
5.4 To obtain the lowest detection levels, a two-pass assay should be used. The two-pass assay also reduces problems related to potential interferences between transmission peaks and assay peaks. For items with higher activities, a single-pass assay may be used to increase throughput.
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1.1 This test method covers the transmission-corrected nondestructive assay (NDA) of gamma-ray emitting special nuclear materials (SNMs), most commonly 235U, 239Pu, and 241Am, in low-density scrap or waste, packaged in cylindrical containers. The method can also be applied to NDA of other gamma-emitting nuclides including fission products. High-resolution gamma-ray spectroscopy is used to detect and measure the nuclides of interest and to measure and correct for gamma-ray attenuation in a series of horizontal segments (collimated gamma detector views) of the container. Corrections are also made for counting losses occasioned by signal processing limitations  (1-3).2  
1.2 There are currently several systems in use or under development for determining the attenuation corrections for NDA of radioisotopic materials (4-8). A related technique, tomographic gamma-ray scanning (TGS), is not included in this test method (9, 10, 11).  
1.2.1 This test method will cover two implementations of the Segmented Gamma Scanning (SGS) procedure: (1) Isotope Specific (Mass) Calibration, the original SGS procedure, uses standards of known radionuclide masses to determine detector response in a mass versus corrected count rate calibration that applies only to those specific radionuclides for which it is calibrated, and (2) Efficiency Curve Calibration, an alternative method, typically uses non-SNM radionuclide sources to determine system detection efficiency vs. gamma energy and thereby calibrate for all gamma-emitting radionuclides of interest (12).  
1.2.1.1 Efficiency Curve Calibration, over the energy range for which the efficiency is defined, has the advantage of providing calibration for many gamma-emitting nuclides for which half-life and gamma emission intensity data are available.  
1.3 The assay technique may be applicable to loadings up to several hundred grams of nuclide in a 208-L [55-gal] drum, with more restricted ranges to be applicable depending on specific packaging and counting equipment considerations.  
1.4 Measured transmission values must be available for use in calculation of segment-specific attenuation corrections at the energies of analysis.  
1.5 A related method, SGS with calculated correction factors based on item content and density, is not included in this standard.  
1.6 The values stated in either SI units or inch-pound units are to be regarded separately as standard. The values stated in each system may not be exact equivalents; therefore, each system shall be used independently of the other. Combining values from the two systems may result in non-conformance with the 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 esta...

SIGNIFICANCE AND USE
4.1 The determination of plutonium isotopic composition by gamma-ray spectrometry is a nondestructive technique and when used with other nondestructive techniques, such as calorimetry (Test Method C1458) or neutron counting (Test Methods C1207, C1316, C1493, and C1500), can provide a wholly nondestructive plutonium assay necessary for material accountancy and safeguards needs.  
4.2 Because gamma-ray spectrometry systems are typically automated, the routine use of the test method is fast, reliable, and is not labor intensive. The test method is nondestructive, requires no sample preparation, and does not create waste disposal problems.  
4.3 This test method assumes that all plutonium in the measured item has the same isotopic distribution, often called isotopic homogeneity (see 7.2.4 and 7.2.5).  
4.4 The 242Pu abundance is not measured by this test method and must be estimated from isotopic correlation techniques, stream averages, historical information, or other measurement techniques.  
4.5 Americium-241 is a daughter product of  241Pu. The 241Am/239Pu atom ratio can also be determined by means of this test method (assuming a homogeneous isotopic distribution of plutonium and 241Am). The determination of the 241Am/239Pu atom ratio is necessary for the correct interpretation of a calorimetric heat measurement.  
4.6 The isotopic composition of a given batch or item of plutonium is an attribute of that item and, once determined, can be used in subsequent inventory measurements to verify the identity of an item within the measurement uncertainties.  
4.7 The method can also measure the ratio of other gamma-emitting isotopes to plutonium assuming they have the same spatial distribution as the plutonium in the item. Some of these “other” gamma-emitting isotopes include isotopes of uranium, neptunium, curium, cesium, and other fission products. The same methods of this standard can be used to measure the isotopic composition of uranium in items containing only u...
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1.1 This test method is applicable to the determination of isotopic abundances in isotopically homogeneous plutonium-bearing materials. This test method may be applicable to other plutonium-bearing materials, some of which may require modifications to the described test method.  
1.2 The procedure is applicable to items containing plutonium masses ranging from a few tens of milligrams up to the maximum plutonium mass allowed by criticality limits.  
1.3 Measurable gamma ray emissions from plutonium cover the energy range from approximately 30 keV to above 800 keV. K-X-ray emissions from plutonium and its daughters are found in the region around 100 keV. This test method has been applied to all portions of this broad spectrum of emissions.  
1.4 The isotopic abundance of the 242Pu isotope is not directly determined because it has no useful gamma-ray signature. Isotopic correlation techniques may be used to estimate its relative abundance Refs (1) and (2).2  
1.5 This test method has been demonstrated in routine use for isotopic abundances ranging from 99 to  239Pu. This test method has also been employed for isotopic abundances outside this range.  
1.6 The values stated in SI units are to be regarded as 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.

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1.1 The terminology defined in this document is associated with nondestructive assay of nuclear material.  
1.2 All of the definitions are associated with measurement techniques that measure nuclear emissions (that is, neutrons, gamma-rays, or heat) directly or indirectly.  
1.3 Definitions are relevant to any standards and guides written by subcommittee C26.10.  
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
5.1 The following methods assist in demonstrating regulatory compliance in such areas as safeguards (Special Nuclear Material), inventory control, criticality control, decontamination and decommissioning, waste disposal, holdup and shipping.  
5.2 This guide can apply to the assay of radionuclides in containers, whose gamma-ray absorption properties can be measured or estimated, for which representative certified standards are not available. It can be applied to in situ measurements, measurement stations, or to laboratory measurements.  
5.3 Some of the modeling techniques described in the guide are suitable for the measurement of fall-out or natural radioactivity homogenously distributed in soil.  
5.4 Source-based efficiency calibrations for laboratory geometries may suffer from inaccuracies due to gamma rays being detected in true coincidence. Modeling can be an advantage since it is unaffected by true coincidence summing effects.
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1.1 This guide addresses the use of models with passive gamma-ray measurement systems. Mathematical models based on physical principles can be used to assist in calibration of gamma-ray measurement systems and in analysis of measurement data. Some nondestructive assay (NDA) measurement programs involve the assay of a wide variety of item geometries and matrix combinations for which the development of physical standards are not practical. In these situations, modeling may provide a cost-effective means of meeting user’s data quality objectives.  
1.2 A scientific knowledge of radiation sources and detectors, calibration procedures, geometry and error analysis is needed for users of this standard. This guide assumes that the user has, at a minimum, a basic understanding of these principles and good NDA practices (see Guide C1592/C1592M), as defined for an NDA professional in Guide C1490. The user of this standard must have at least a basic understanding of the software used for modeling. Instructions or further training on the use of such software is beyond the scope of this standard.  
1.3 The focus of this guide is the use of response models for high-purity germanium (HPGe) detector systems for the passive gamma-ray assay of items. Many of the models described in this guide may also be applied to the use of detectors with different resolutions, such as sodium iodide or lanthanum halide. In such cases, an NDA professional should determine the applicability of sections of this guide to the specific application.  
1.4 Techniques discussed in this guide are applicable to modeling a variety of radioactive material including contaminated fields, walls, containers and process equipment.  
1.5 This guide does not purport to discuss modeling for “infinite plane” in situ measurements. This discussion is best covered in ANSI N42.28.  
1.6 This guide does not purport to address the physical concerns of how to make or set up equipment for in situ measurements but only how to select the model for which the in situ measurement data is analyzed.  
1.7 The values stated in either SI units or inch-pound units are to be regarded separately as standard. The values stated in each system may not be exact equivalents; therefore, each system shall be used independently of the other. Combining values from the two systems may result in non-conformance with the standard.  
1.8 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered standard.  
1.9 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 test method is used to determine the U and Pu content of scrap and waste in containers. Active measurement times have typically been 100 to 1000 s. Passive measurement times have typically been 400 s to several hours. The following limits may be further restricted depending upon specific matrix, calibration material, criticality safety, or counting equipment considerations.  
5.1.1 The passive measurement has been applied to benign matrices in 208 L drums with Pu content ranging from 30 mg to 1 kg.  
5.1.2 The active measurement has been applied to waste drums with 235U content ranging from about 100 mg to 1 kg.  
5.2 This test method can be used to demonstrate compliance with the radioactivity levels specified in safeguards, waste, disposal, and environmental regulations (for example, see NRC regulatory guides 5.11, 5.53, DOE Order 5820.2a, and 10CFR61 sections 61.55 and sections 61.56, 40CFR191, and DOE/WIPP-069).  
5.3 This test method could be used to detect diversion attempts that use shielding to encapsulate nuclear material.  
5.4 The bias of the measurement results is related to the item size and density, the homogeneity and composition of the matrix, and the quantity and distribution of the nuclear material. The precision of the measurement results is related to the quantity of nuclear material and the count time of the measurement.  
5.4.1 For both the matrix-specific and the matrix-correction approaches, the method assumes the calibration materials match the items to be measured with respect to the homogeneity and composition of the matrix, the neutron moderator and absorber content, and the quantity of nuclear material, to the extent they affect the measurement.  
5.4.2 It is recommended that measurements be made on small containers of scrap and waste before they are combined in large containers. Special arrangement may be required to assay small containers to best effect in a large cavity general purpose shuffer.  
5.4.3 It is recommend...
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1.1 This test method covers the nondestructive assay of scrap and waste items for U, Pu, or both, using a 252 Cf shuffler. Shuffler measurements have been applied to a variety of matrix materials in containers of up to several 100 L. Corrections are made for the effects of matrix material. Applications of this test method include measurements for safeguards, accountability, TRU, and U waste segregation, disposal, and process control purposes (1, 2, 3).2  
1.1.1 This test method uses passive neutron coincidence counting (4) to measure the 240Pu-effective mass. It has been used to assay items with total Pu contents between 0.03 g and 1000 g. It could be used to measure other spontaneously fissioning isotopes such as Cm and Cf. It specifically describes the approach used with shift register electronics; however, it can be adapted to other electronics.  
1.1.2 This test method uses neutron irradiation with a moveable Cf source and counting of the delayed neutrons from the induced fissions to measure the 235U equivalent fissile mass. It has been used to assay items with 235U contents between 0.1 g and 1000 g. It could be used to assay other fissile and fissionable isotopes.  
1.2 This test method requires knowledge of the relative isotopic composition (See Test Method C1030) of the special nuclear material to determine the mass of the different elements from the measurable quantities.  
1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.4 The techniques described in this test method have been applied to materials other than scrap and waste. These other applications are not addressed in this test method.  
1.5 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 applicab...

SIGNIFICANCE AND USE
5.1 This test method is useful for determining the plutonium content of items such as impure Pu oxide, mixed Pu/U oxide, oxidized Pu metal, Pu scrap and waste, Pu process residues, and weapons components.  
5.2 Measurements made with this test method may be suitable for safeguards or waste characterization requirements such as:  
5.2.1 Nuclear materials accountability,  
5.2.2 Inventory verification (7),  
5.2.3 Confirmation of nuclear materials content (8),  
5.2.4 Resolution of shipper/receiver differences (9),  
5.2.5 Excess weapons materials inspections  (10, 11),  
5.2.6 Safeguards termination on waste  (12, 13),  
5.2.7 Determination of fissile equivalent content (14).  
5.3 A significant feature of neutron multiplicity counting is its ability to capture more information than neutron coincidence counting because of the availability of a third measured parameter, leading to reduced measurement bias for most material categories for which suitable precision can be attained. This feature also makes it possible to assay some in-plant materials that are not amenable to conventional coincidence counting, including moist or impure plutonium oxide, oxidized metal, and some categories of scrap, waste, and residues (10).  
5.4 Calibration for many material types does not require representative standards. Thus, the technique can be used for inventory verification without calibration standards (7), although measurement bias may be lower if representative standards were available.  
5.4.1 The repeatability of the measurement results due to counting statistics is related to the quantity of nuclear material, interfering neutrons, and the count time of the measurement (15) .  
5.4.2 For certain materials such as small Pu, items of less than 1 g, some Pu-bearing waste, or very impure Pu process residues where the (α,n) reaction rate overwhelms the triples signal, multiplicity information may not be useful because of the poor counting statistics of the triple coinci...
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1.1 This test method describes the nondestructive assay of plutonium in forms such as metal, oxide, scrap, residue, or waste using passive neutron multiplicity counting. This test method provides results that are usually more accurate than conventional neutron coincidence counting. The method can be applied to a large variety of plutonium items in various containers including cans, 208-L drums, or 1900-L Standard Waste Boxes. It has been used to assay items whose plutonium content ranges from 1 g to 1000s of g.  
1.2 There are several electronics or mathematical approaches available for multiplicity analysis, including the multiplicity shift register, the Euratom Time Correlation Analyzer, and the List Mode Module, as described briefly in Ref. (1).2  
1.3 This test method is primarily intended to address the assay of 240Pu-effective by moments-based multiplicity analysis using shift register electronics (1, 2, 3) and high efficiency neutron counters specifically designed for multiplicity analysis.  
1.4 This test method requires knowledge of the relative abundances of the plutonium isotopes to determine the total plutonium mass (See Test Method C1030).  
1.5 This test method may also be applied to modified neutron coincidence counters (4) which were not specifically designed as multiplicity counters (that is, HLNCC, AWCC, etc), with a corresponding degradation of results.  
1.6 The values stated in SI units are to be regarded as 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.

SIGNIFICANCE AND USE
5.1 The enrichment meter principle provides a nondestructive measurement of the  235U fraction of uranium-bearing items. Sampling is not required and no waste is generated, minimizing exposure to hazardous materials and resulting in reduced sampling error.  
5.2 This method relies on a fixed and controlled geometry. The uranium-bearing materials in the measured items and calibration reference materials used for calibration must fill the detector field of view.  
5.3 Use of a low resolution detector (for example, NaI detector) to measure uranium with  235U fraction approximately 10 % which is contained in a thin-walled container can provide a rapid (typically 100 s), easily portable measurement system with precision of 0.6 % and bias of less than 1 %.  
5.4 Use of a high resolution detector (for example, high-purity germanium) can provide measurement with a precision better than 0.2 % and a bias less than 1 % within a 300-s measurement time when measuring uranium with  235U fraction in the range of 0.711 % or above which is contained in thin-walled containers.  
5.5 In order to obtain optimum results using this method, the chemical composition of the item must be well known, the container wall must permit transmission of the 185.7 keV gamma-ray, and the uranium-bearing material within the item must be infinitely thick with respect to the 185.7 keV gamma-ray. All items must be in identical containers or must have a known container wall thickness and composition.  
5.6 Items to be measured must be homogeneous with respect to both  235U fraction and chemical composition.  
5.7 When measuring items, using low-resolution detectors, in thin-walled containers that have not reached secular equilibrium (more than about 120 days after processing), either the method should not be used, additional corrections should be made to account for the age of the uranium, or high-resolution measurements should be performed.  
5.8 The method is often used as a enrichment verificatio...
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1.1 This test method covers the quantitative determination of the fraction of  235U in uranium using measurement of the 185.7 keV gamma-ray produced during the decay of  235U.  
1.2 This test method is applicable to items containing homogeneous uranium-bearing materials of known chemical composition in which the compound is considered infinitely thick with respect to 185.7 keV gamma-rays.  
1.3 This test method can be used for the entire range of  235U fraction as a weight percent, from depleted (0.2 % 235U) to highly enriched (97.5 % 235U).  
1.4 Measurement of items that have not reached secular equilibrium between  238U and  234Th may not produce the stated bias when low-resolution detectors are used with the computational method listed in Annex A2.  
1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.6 This standard may involve hazardous materials, operations, and equipment. 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.

SIGNIFICANCE AND USE
5.1 This test method is considered to be the most accurate NDA technique for the assay of many physical forms of Pu. Isotopic measurements by gamma-ray spectroscopy or destructive analysis techniques are part of this test method when it is applied to the assay of Pu.  
5.1.1 Calorimetry has been applied to a wide variety of Pu-bearing solids including metals, alloys, oxides, fluorides, mixed Pu-U oxides, mixed oxide fuel pins, waste, and scrap, for example, ash, ash heels, salts, crucibles, and graphite scarfings) (2, 3). This test method has been routinely used at U.S. and European facilities for Pu process measurements and nuclear material accountability since the mid 1960’s (2-9).  
5.1.2 Pu-bearing materials have been measured in calorimeter containers ranging in size from about 0.025 m to about 0.63 m in diameter and from about 0.076 m to about 1.38 m in height.  
5.1.3 Gamma-ray spectroscopy typically is used to determine the Pu isotopic composition and 241Am to Pu ratio (see Test Method C1030). However, isotopic information from mass spectrometry and alpha counting measurements may be used instead (see Test Method C697).  
5.2 This test method is considered to be the most accurate NDA method for the measurement of tritium. For many physical forms of tritium compounds calorimetry is currently the only practical measurement technique available.  
5.3 Physical standards representative of the materials being assayed are not required for the test method.  
5.3.1 This test method is largely independent of the elemental distribution of the nuclear materials in the matrix.  
5.3.2 The accuracy of the method can be degraded for materials with inhomogeneous isotopic composition.  
5.4 The thermal power measurement is traceable to national measurement systems through electrical standards used to directly calibrate the calorimeters or to calibrate secondary 238Pu heat standards.  
5.5 Heat-flow calorimetry has been used to prepare secondary standards for neutron...
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1.1 This test method describes the nondestructive assay (NDA) of plutonium, tritium, and 241Am using heat flow calorimetry. For plutonium the typical range of applicability, depending on the isotopic composition, corresponds to ~0.1 g to ~5 g quantities while for tritium the typical range extends from ~0.001 g to ~400 g. This test method can be applied to materials in a wide range of container sizes up to 380 L. It has been used routinely to assay items whose thermal power ranges from 0.001 W to 135 W.  
1.2 This test method requires knowledge of the relative abundances of the plutonium isotopes and the  241Am/Pu mass ratio to determine the total plutonium mass.  
1.3 This test method provides a direct measure of tritium content.  
1.4 This test method provides a measure of  241Am either as a single isotope or mixed with plutonium.  
1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
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 and health practices and determine the applicability of regulatory limitations prior to use.

SIGNIFICANCE AND USE
This test method is useful for quantifying fissile (for example, 233U, 235U, 239Pu and 241Pu) and spontaneously-fissioning nuclei (for example, 238Pu, 240Pu, 242Pu, 244Cm, 248Cm, and 252Cf) in waste and scrap drums. Total elemental mass of the radioactive materials can be calculated if the relative abundances of each radionuclide are known.
Typically, this test method is used to measure one fissile isotope (for example, 235U or 239Pu).
This test method can be used to segregate low level and transuranic waste at the 100 nCi/g concentration level currently required to meet the DOE Waste Isolation Pilot Plant (WIPP) waste acceptance criterion (5, 8, 9).
This test method can be used for waste characterization to demonstrate compliance with the radioactivity levels specified in waste, disposal, and environmental regulations (See NRC regulatory guides, DOE Order 435.1, 10 CFR Part 71, 40 CFR Part 191, and DOE /WIPP-069).
In the active mode, the DDT system can measure the 235U content in the range from 0.02 to >100 g and the 239Pu content, nominally between 0.01 and >20 g.
In the passive mode, the DDT system is capable of assaying spontaneously-fissioning nuclei, over a nominal range from 0.05 to 15 g of 240Pu, or equivalent (5, 10, 11, 12, 13).
This test method should be used in conjunction with a waste management plan that segregates the contents of assay items into material categories according to some or all of the following criteria: bulk density of the waste, chemical forms of the plutonium or uranium and matrix, (α, n) neutron intensity, hydrogen (moderator) and absorber content, thickness of fissile mass(es), and the assay item container size and composition. Each matrix may require a different set of calibration standards and may have different mass calibration limits. The effect on the quality of the assay (that is, minimizing precision and bias) can significantly depend on the degree of adherence to this waste management plan.
The bias of the measu...
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1.1 This test method covers a system that performs nondestructive assay (NDA) of uranium or plutonium, or both, using the active, differential die-away technique (DDT), and passive neutron coincidence counting. Results from the active and passive measurements are combined to determine the total amount of fissile and spontaneously-fissioning material in drums of scrap or waste. Corrections are made to the measurements for the effects of neutron moderation and absorption, assuming that the effects are averaged over the volume of the drum and that no significant lumps of nuclear material are present. These systems are most widely used to assay low-level and transuranic waste, but may also be used for the measurement of scrap materials. The examples given within this test method are specific to the second-generation Los Alamos National Laboratory (LANL) passive-active neutron assay system.
1.1.1 In the active mode, the system measures fissile isotopes such as 235U and 239Pu. The neutrons from a pulsed, 14-MeV neutron generator are thermalized to induce fission in the assay item. Between generator pulses, the system detects prompt-fission neutrons emitted from the fissile material. The number of detected neutrons between pulses is proportional to the mass of fissile material. This method is called the differential die-away technique.  
1.1.2 In the passive mode, the system detects time-coincident neutrons emitted from spontaneously fissioning isotopes. The primary isotopes measured are 238Pu, 240Pu, and 242Pu; however, the system may be adapted for use on other spontaneously-fissioning isotopes as well, such as kilogram quantities of 238U. The number of coincident neutrons detected is proportional to the mass of spontaneously-fissioning material.
1.2 The active mode is used to assay fissile material in the following ranges.
1.2.1 For uranium-only bearing items, the DDT can measure the 235U content in the range from abo...

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 an...
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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.  
1.1.1 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...

SIGNIFICANCE AND USE
4.1 The process of selection, training and qualification of personnel involved with NDA measurements is one of the quality assurance elements for an overall quality NDA measurement program.  
4.2 This guide describes an approach to selection, qualification, and training of personnel that is to be used in conjunction with other NDA Quality Assurance (QA) program elements. The selection, qualification and training processes can vary and this guide provides one such approach.  
4.3 The qualification activities described in this guide assume that NDA personnel are already proficient in general facility operations and safety procedures. The training and activities that developed this proficiency are not covered in this guide.  
4.4 This guide describes a basic approach and principles for the qualification of NDA professionals and technical specialists and operators. A different approach may be adopted by the management organization based on its particular organization and facility specifics. However, if a variation of the approach of this guide is applied, the resulting selection, training, and qualification programs must meet the requirements of the facility quality assurance program and should provide all the applicable functions of Section 5.  
4.5 This guide may be used as an aid in the preparation of a Training Implementation Plan (TIP) for the Transuranic Waste Characterization Program (TWCP).  
4.6 This guide describes education and expertise guidance for NDA auditors due to the importance and complexity of proper oversight of NDA activities.
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1.1 This guide contains good practices for the selection, training, qualification, and professional development of personnel performing analysis, calibration, physical measurements, or data review using nondestructive assay equipment, methods, results, or techniques. The guide also covers NDA personnel involved with NDA equipment setup, selection, diagnosis, troubleshooting, or repair. General guidelines for the selection, training, and qualification of NDA auditors are included as well, but at a lower level of detail due to the variability of the personnel’s responsibilities performing this functions. Selection, training, and qualification programs based on this guide are intended to provide assurance that NDA personnel are suitably qualified and experienced personnel (SQEP) to perform their jobs competently. This guide presents a series of options but does not recommend a specific course of action.
This standard guide does not address the qualifications per se of an NDA Manager. However, it is expected that the NDA Manager is familiar with NDA techniques, and can make informed decisions on the acceptability of the assay results. If an NDA Manager does not have adequate technical qualifications in the NDA field, they are recommended to undergo training to gain familiarity in this area.  
An NDA Manager with no relevant NDA experience should have access to a Senior NDA Professional who will give guidance for all technical decisions such as applicability and limitation of methods, reasonableness of results, needed upgrades and advantageous development investments.

SIGNIFICANCE AND USE
5.1 Measurement results from this test method assists in demonstrating regulatory compliance in such areas as safeguards SNM inventory control, criticality control, waste disposal, and decontamination and decommissioning (D&D). This test method can apply to the measurement of holdup in process equipment or discrete items whose gamma-ray absorption properties may be measured or estimated. This method may be adequate to accurately measure items with complex distributions of radioactive and attenuating material, however, the results are subject to larger measurement uncertainties than measurements of less complex distributions of radioactive material.  
5.2 Scan—A scan is used to provide a qualitative indication of the extent, location, and the relative quantity of holdup. It can be used to plan or supplement the quantitative measurements.  
5.3 Nuclide Mapping—Nuclide mapping measures the relative isotopic composition of the holdup at specific locations. It can also be used to detect the presence of radionuclides that emit radiation which could interfere with the assay. Nuclide mapping is best performed using a high resolution detector (such as HPGe) for best nuclide and interference detection. If the holdup is not isotopically homogeneous at the measurement location, that measured isotopic composition will not be a reliable estimate of the bulk isotopic composition.  
5.4 Quantitative Measurements—These measurements result in quantification of the mass of the measured nuclides in the holdup. They include all the corrections, such as attenuation, and descriptive information, such as isotopic composition, that are available  
5.4.1 High quality results require detailed knowledge of radiation sources and detectors, transmission of radiation, calibration, facility operations and error analysis. Judicious use of subject matter experts is required (Guide C1490).  
5.5 Holdup Monitoring—Periodic re-measurement of holdup at a defined point using the same technique and a...
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1.1 This test method describes gamma-ray methods used to nondestructively measure the quantity of 235U or  239Pu present as holdup in nuclear facilities. Holdup may occur in any facility where nuclear material is processed, in process equipment, in exhaust ventilation systems and in building walls and floors.  
1.2 This test method includes information useful for management, planning, selection of equipment, consideration of interferences, measurement program definition, and the utilization of resources  (1, 2, 3, 4) .2  
1.3 The measurement of nuclear material hold up in process equipment requires a scientific knowledge of radiation sources and detectors, transmission of radiation, calibration, facility operations and uncertainty analysis. It is subject to the constraints of the facility, management, budget, and schedule; plus health and safety requirements. The measurement process includes defining measurement uncertainties and is sensitive to the form and 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, to identify the form of the material, and to measure it without interference from adjacent sources of radiation.  
1.4 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.

SIGNIFICANCE AND USE
5.1 Measurement results from this test method assists in demonstrating regulatory compliance in such areas as safeguards SNM inventory control, criticality control, waste disposal, and decontamination and decommissioning (D&D). This test method can apply to the measurement of holdup in process equipment or discrete items whose gamma-ray absorption properties may be measured or estimated. This method may be adequate to accurately measure items with complex distributions of radioactive and attenuating material, however, the results are subject to larger measurement uncertainties than measurements of less complex distributions of radioactive material.  
5.2 Scan—A scan is used to provide a qualitative indication of the extent, location, and the relative quantity of holdup. It can be used to plan or supplement the quantitative measurements.  
5.3 Nuclide Mapping—Nuclide mapping measures the relative isotopic composition of the holdup at specific locations. It can also be used to detect the presence of radionuclides that emit radiation which could interfere with the assay. Nuclide mapping is best performed using a high resolution detector (such as HPGe) for best nuclide and interference detection. If the holdup is not isotopically homogeneous at the measurement location, that measured isotopic composition will not be a reliable estimate of the bulk isotopic composition.  
5.4 Quantitative Measurements—These measurements result in quantification of the mass of the measured nuclides in the holdup. They include all the corrections, such as attenuation, and descriptive information, such as isotopic composition, that are available  
5.4.1 High quality results require detailed knowledge of radiation sources and detectors, transmission of radiation, calibration, facility operations and error analysis. Judicious use of subject matter experts is required (Guide C1490).  
5.5 Holdup Monitoring—Periodic re-measurement of holdup at a defined point using the same technique and...
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1.1 This test method describes gamma-ray methods used to nondestructively measure the quantity of 235U or  239Pu present as holdup in nuclear facilities. Holdup may occur in any facility where nuclear material is processed, in process equipment, in exhaust ventilation systems and in building walls and floors.  
1.2 This test method includes information useful for management, planning, selection of equipment, consideration of interferences, measurement program definition, and the utilization of resources  (1, 2, 3, 4) .2  
1.3 The measurement of nuclear material hold up in process equipment requires a scientific knowledge of radiation sources and detectors, transmission of radiation, calibration, facility operations and uncertainty analysis. It is subject to the constraints of the facility, management, budget, and schedule; plus health and safety requirements. The measurement process includes defining measurement uncertainties and is sensitive to the form and 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, to identify the form of the material, and to measure it without interference from adjacent sources of radiation.  
1.4 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.

SCOPE
1.1 The terminology defined in this document is associated with nondestructive assay of nuclear material.
1.2 All of the definitions are associated with measurement techniques that measure nuclear emissions (that is, neutrons, gamma-rays, or heat) directly or indirectly.
1.3 definitions are relevant to any standards and guides written by subcommittee C26.10.

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1.1 The terminology defined in this document is associated with nondestructive assay of nuclear material.
1.2 All of the definitions are associated with measurement techniques that measure nuclear emissions (that is, neutrons, gamma-rays, or heat) directly or indirectly.
1.3 definitions are relevant to any standards and guides written by subcommittee C26.10.

SIGNIFICANCE AND USE
The process of selection, training and qualification of personnel involved with NDA measurements is one of the quality assurance elements for an overall quality NDA measurement program.
This guide describes an approach to selection, qualification, and training of personnel that is to be used in conjunction with other NDA QA program elements. The selection, qualification and training processes can vary and this guide provides one such approach.
The qualification activities described in this guide assume that NDA personnel are already proficient in general facility operations and safety procedures. The training and activities that developed this proficiency are not covered in this guide.
This guide describes a basic approach and principles for the qualification of NDA professionals and technical specialists and operators. A different approach may be adopted by the management organization based on its particular organization and facility specifics. However, if a variation of the approach of this guide is applied, the resulting selection, training, and qualification programs must meet the requirements of the facility quality assurance program and should provide all the applicable functions of Section 4.
This guide may be used as an aid in the preparation of a Training Implementation Plan (TIP) for the Transuranic Waste Characterization Program (TWCP). Requirements for TWCP training include but are not limited to: 10 CFR 830.120, DOE Order 5480.20A, DOE QAPD CAO-94-1012, TWCP-QAPP CAO-94-1010, TWCP-QAPjP and the TUM.
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1.1 This guide contains good practices for the selection, training, qualification, and professional development of personnel performing analysis, calibration, physical measurements, or data review using nondestructive assay equipment, methods, results, or techniques. The guide also covers NDA personnel involved with NDA equipment setup, selection, diagnosis, troubleshooting, or repair. Selection, training, and qualification programs based on this guide are intended to provide assurance that NDA personnel are qualified to perform their jobs competently. This guide presents a series of options but does not recommend a specific course of action.

SIGNIFICANCE AND USE
The determination of plutonium isotopic composition by gamma-ray spectrometry is a nondestructive technique and when used with other nondestructive techniques, such as calorimetry (Test Method C1458) or neutron counting (Test Methods C1207, C1316, C1493, and C1500), can provide a wholly nondestructive plutonium assay necessary for material accountancy and safeguards needs.
Because gamma-ray spectrometry systems are typically automated, the routine use of the test method is fast, reliable, and is not labor intensive. The test method is nondestructive, requires no sample preparation, and does not create waste disposal problems.
This test method assumes that all plutonium in the measured item has the same isotopic distribution, often called isotopic homogeneity (see 7.2.4 and 7.2.5).
The 242Pu abundance is not measured by this test method and must be estimated from isotopic correlation techniques, stream averages, historical information, or other measurement techniques.
Americium-241 is a daughter product of 241Pu. The 241Am/239Pu atom ratio can also be determined by means of this test method (assuming a homogeneous isotopic distribution of plutonium and 241Am). The determination of the 241Am/239Pu atom ratio is necessary for the correct interpretation of a calorimetric heat measurement.
The isotopic composition of a given batch or item of plutonium is an attribute of that item and, once determined, can be used in subsequent inventory measurements to verify the identity of an item within the measurement uncertainties.
The method can also measure the ratio of other gamma-emitting isotopes to plutonium assuming they have the same spatial distribution as the plutonium in the item. Some of these “other” gamma-emitting isotopes include isotopes of uranium, neptunium, curium, cesium, and other fission products. The same methods of this standard can be used to measure the isotopic composition of uranium in items containing only uranium (3, 4, 5, 6).
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1.1 This test method is applicable to the determination of isotopic abundances in isotopically homogeneous plutonium-bearing materials. This test method may be applicable to other plutonium-bearing materials, some of which may require modifications to the described test method.
1.2 The procedure is applicable to items containing plutonium masses ranging from a few tens of milligrams up to the maximum plutonium mass allowed by criticality limits.
1.3 Measurable gamma ray emissions from plutonium cover the energy range from approximately 30 keV to above 800 keV. K-X-ray emissions from plutonium and its daughters are found in the region around 100 keV. This test method has been applied to all portions of this broad spectrum of emissions.  
1.4 The isotopic abundance of the 242Pu isotope is not directly determined because it has no useful gamma-ray signature. Isotopic correlation techniques may be used to estimate its relative abundance Refs (1) and (2).  
1.5 This test method has been demonstrated in routine use for isotopic abundances ranging from 99 to 50 % 239Pu. This test method has also been employed for isotopic abundances outside this range.
1.6 The values stated in SI units are to be regarded as 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.

SCOPE
1.1 The terminology defined in this document is associated with nondestructive assay of nuclear material.
1.2 All of the definitions are associated with measurement techniques that measure nuclear emissions (that is, neutrons, gamma-rays, or heat) directly or indirectly.
1.3 definitions are relevant to any standards and guides written by subcommittee C26.10.

SIGNIFICANCE AND USE
This test method is useful for determining the plutonium content of scrap and waste in containers ranging from small cans with volumes of the order of a mL to crates and boxes of several thousand liters in volume. A common application would be to 208-L (55-gal) drums. Total Pu content ranges from 10 mg to 6 kg (1). The upper limit may be restricted depending on specific matrix, calibration material, criticality safety, or counting equipment considerations.
This test method is applicable for U.S. Department of Energy shipper/receiver confirmatory measurements (9), nuclear material diversion detection, and International Atomic Energy Agency attributes measurements (10).
This test method should be used in conjunction with a scrap and waste management plan that segregates scrap and waste assay items into material categories according to some or all of the following criteria: bulk density, the chemical forms of the plutonium and the matrix, americium to plutonium isotopic ratio, and hydrogen content. Packaging for each category should be uniform with respect to size, shape, and composition of the container. Each material category might require calibration standards and may have different Pu mass limits.
Bias in passive neutron coincidence measurements is related to item size and density, the homogeneity and composition of the matrix, and the quantity and distribution of the nuclear material. The precision of the measurement results is related to the quantity of nuclear material, the (α,n) reaction rate, and the count time of the measurement.
For both benign matrix and matrix specific measurements, the method assumes the calibration reference materials match the items to be measured with respect to the homogeneity and composition of the matrix, the neutron moderator and absorber content, and the quantity of nuclear material, to the extent they affect the measurement.
Measurements of smaller containers containing scrap and waste are generally more accurate than m...
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1.1 This test method describes the nondestructive assay of scrap or waste for plutonium content using passive thermal-neutron coincidence counting. This test method provides rapid results and can be applied to a variety of carefully sorted materials in containers as large as several thousand liters in volume. The test method applies to measurements of  238Pu, 240Pu, and  242Pu and has been used to assay items whose total plutonium content ranges from 10 mg to 6 kg (1).  
1.2 This test method requires knowledge of the relative abundances of the Pu isotopes to determine the total Pu mass (Test Method C1030).
1.3 This test method may not be applicable to the assay of scrap or waste containing other spontaneously fissioning nuclides.
1.3.1 This test method may give biased results for measurements of containers that include large amounts of hydrogenous materials.
1.3.2 The techniques described in this test method have been applied to materials other than scrap and waste (2, 3).
1.4 This test method assumes the use of shift-register-based coincidence technology (4).
1.5 Several other techniques that are often encountered in association with passive neutron coincidence counting exist These include neutron multiplicity counting (5, 6, Test Method C1500), add-a-source analysis for matrix correction (7), flux probes also for matrix compensation, cosmic-ray rejection (8) to improve precision close to the detection limit, and alternative data collection electronics such as list mode data acquisition. Passive neutron coincidence counting may also be combined with certain active interrogation schemes as in Test Methods C1316 and C1493. Discussions of these established techniques are not included in this 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 and health practices ...

SIGNIFICANCE AND USE
This test method is a nondestructive means of determining the nuclide concentration of a solution for special nuclear material accountancy, nuclear safety, and process control.
It is assumed that the nuclide to be analyzed is in a homogeneous solution (Practice C1168).
The transmission correction makes the test method independent of matrix (solution elemental composition and density) and useful over several orders of magnitude of nuclide concentrations. However, a typical configuration will normally span only two to three orders of magnitude because of detector dynamic range.
The test method assumes that the solution-detector geometry is the same for all measured items. This can be accomplished by requiring that the liquid height in the sidelooking geometry exceeds the detector field of view defined by the collimator. For the upward-looking geometry, a fixed solution fill height must be maintained and vials of identical radii must be used unless the vial radius exceeds the field of view defined by the collimator.
Since gamma-ray systems can be automated, the test method can be rapid, reliable, and not labor intensive.
This test method may be applicable to in-line or off-line situations.
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1.1 This test method covers the determination of the concentration of gamma-ray emitting special nuclear materials dissolved in homogeneous solutions. The test method corrects for gamma-ray attenuation by the solution and its container by measurement of the transmission of a beam of gamma rays from an external source (Refs. (1), (2), and (3)).  
1.2 Two solution geometries, slab and cylinder, are considered. The solution container that determines the geometry may be either a removable or a fixed geometry container. This test method is limited to solution containers having walls or a top and bottom of equal transmission through which the gamma rays from the external transmission correction source must pass.
1.3 This test method is typically applied to radionuclide concentrations ranging from a few milligrams per litre to several hundred grams per litre. The assay range will be a function of the specific activity of the nuclide of interest, the physical characteristics of the solution container, counting equipment considerations, assay gamma-ray energies, solution matrix, gamma-ray branching ratios, and interferences.
1.4 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. For specific hazards, see Section 9.

SIGNIFICANCE AND USE
The following methods assist in demonstrating regulatory compliance in such areas as safeguards (Special Nuclear Material), inventory control, criticality control, decontamination and decommissioning, waste disposal, holdup and shipping.
This guide can apply to the assay of radionuclides in containers, whose gamma-ray absorption properties can be measured or estimated, for which representative certified standards are not available. It can be applied to in situ measurements, measurement stations, or to laboratory measurements.
Some of the modeling techniques described in the guide are suitable for the measurement of fall-out or natural radioactivity homogenously distributed in soil.
Source-based efficiency calibrations for laboratory geometries may suffer from inaccuracies due to gamma rays being detected in true coincidence. Modeling can be an advantage since it is unaffected by true coincidence summing effects.
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1.1 This guide addresses the use of models with passive gamma-ray measurement systems. Mathematical models based on physical principles can be used to assist in calibration of gamma-ray measurement systems and in analysis of measurement data. Some nondestructive assay (NDA) measurement programs involve the assay of a wide variety of item geometries and matrix combinations for which the development of physical standards are not practical. In these situations, modeling may provide a cost-effective means of meeting user’s data quality objectives.
1.2 A scientific knowledge of radiation sources and detectors, calibration procedures, geometry and error analysis is needed for users of this standard. This guide assumes that the user has, at a minimum, a basic understanding of these principles and good NDA practices (see Guide C1592), as defined for an NDA professional in Guide C1490. The user of this standard must have at least a basic understanding of the software used for modeling. Instructions or further training on the use of such software is beyond the scope of this standard.
1.3 The focus of this guide is the use of response models for high-purity germanium (HPGe) detector systems for the passive gamma-ray assay of items. Many of the models described in this guide may also be applied to the use of detectors with different resolutions, such as sodium iodide or lanthanum halide. In such cases, an NDA professional should determine the applicability of sections of this guide to the specific application.
1.4 Techniques discussed in this guide are applicable to modeling a variety of radioactive material including contaminated fields, walls, containers and process equipment.
1.5 This guide does not purport to discuss modeling for “infinite plane” in situ measurements. This discussion is best covered in ANSI N42.28.
1.6 This guide does not purport to address the physical concerns of how to make or set up equipment for in situ measurements but only how to select the model for which the in situ measurement data is analyzed.
1.7 The values stated in either SI units or inch-pound units are to be regarded separately as standard. The values stated in each system may not be exact equivalents; therefore, each system shall be used independently of the other. Combining values from the two systems may result in non-conformance with the standard.
1.8 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered standard.

SIGNIFICANCE AND USE
Segmented gamma-ray scanning provides a nondestructive means of measuring the nuclide content of scrap and waste where the specific nature of the matrix and the chemical form and relationship between the nuclide and matrix may be unknown.
The procedure can serve as a diagnostic tool that provides a vertical profile of transmission and nuclide concentration within the item.
Item preparation is generally limited to good waste/scrap segregation practices that produce relatively homogeneous items that are required for any successful waste/inventory management and assay scheme, regardless of the measurement method used. Also, process knowledge should be used, when available, as part of a waste management program to complement information on item parameters, container properties, and the appropriateness of calibration factors.
To obtain the lowest detection levels, a two-pass assay should be used. The two-pass assay also reduces problems related to potential interferences between transmission peaks and assay peaks. For items with higher activities, a single-pass assay may be used to increase throughput.
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1.1 This test method covers the transmission-corrected nondestructive assay (NDA) of gamma-ray emitting special nuclear materials (SNMs), most commonly 235U, 239Pu, and 241Am, in low-density scrap or waste, packaged in cylindrical containers. The method can also be applied to NDA of other gamma-emitting nuclides including fission products. High-resolution gamma-ray spectroscopy is used to detect and measure the nuclides of interest and to measure and correct for gamma-ray attenuation in a series of horizontal segments (collimated gamma detector views) of the container. Corrections are also made for counting losses occasioned by signal processing limitations (1-3).  
1.2 There are currently several systems in use or under development for determining the attenuation corrections for NDA of radioisotopic materials (4-8). A related technique, tomographic gamma-ray scanning (TGS), is not included in this test method (9, 10, 11).
1.2.1 This test method will cover two implementations of the Segmented Gamma Scanning (SGS) procedure: (1) Isotope Specific (Mass) Calibration, the original SGS procedure, uses standards of known radionuclide masses to determine detector response in a mass versus corrected count rate calibration that applies only to those specific radionuclides for which it is calibrated, and (2) Efficiency Curve Calibration, an alternative method, typically uses non-SNM radionuclide sources to determine system detection efficiency vs. gamma energy and thereby calibrate for all gamma-emitting radionuclides of interest (12).  
1.2.1.1 Efficiency Curve Calibration, over the energy range for which the efficiency is defined, has the advantage of providing calibration for many gamma-emitting nuclides for which half-life and gamma emission intensity data are available.
1.3 The assay technique may be applicable to loadings up to several hundred grams of nuclide in a 208-L [55-gal] drum, with more restricted ranges to be applicable depending on specific packaging and counting equipment considerations.
1.4 Measured transmission values must be available for use in calculation of segment-specific attenuation corrections at the energies of analysis.
1.5 A related method, SGS with calculated correction factors based on item content and density, is not included in this standard.
1.6 The values stated in either SI units or inch-pound units are to be regarded separately as standard. The values stated in each system may not be exact equivalents; therefore, each system shall be used independently of the other. Combining values from the two systems may result in non-conformance with the 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...

SIGNIFICANCE AND USE
The TGS provides a nondestructive means of mapping the attenuation characteristics and the distribution of the radionuclide content of items on a voxel by voxel basis. Typically in a TGS analysis a vertical layer (or segment) of an item will be divided into a number of voxels. By comparison, a segmented gamma scanner (SGS) can determine matrix attenuation and radionuclide concentrations only on a segment by segment basis.
It has been successfully used to quantify 238Pu, 239Pu, and 235U. SNM loadings from 0.5 g to 200 g of 239Pu (5, 6), from 1 g to 25 g of 235U (7), and from 0.1 to 1 g of 238Pu have been successfully measured. The TGS technique has also been applied to assaying radioactive waste generated by nuclear power plants (NPP). Radioactive waste from NPP is dominated by activation products (for example, 54Mn, 58Co, 60Co, 110mAg) and fission products (for example, 137Cs, 134Cs). The radionuclide activities measured in NPP waste is in the range from 3.7E+04 Bq to 1.0E+07 Bq. Some results of TGS application to non-SNM radionuclides can be found in the literature (8).
The TGS technique is well suited for assaying items that have heterogeneous matrices and that contain a non-uniform radionuclide distribution.
Since the analysis results are obtained on a voxel by voxel basis, the TGS technique can in many situations yield more accurate results when compared to other gamma ray techniques such as SGS.
In determining the radionuclide distribution inside an item, the TGS analysis explicitly takes into account the cross talk between various vertical layers of the item.
The TGS analysis technique uses a material basis set method that does not require the user to select a mass attenuation curve apriori, provided the transmission source has at least 2 gamma lines that span the energy range of interest.
A commercially available TGS system consists of building blocks that can easily be configured to operate the system in the SGS mode or in a far-field geometry.
...
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1.1 This test method describes the nondestructive assay (NDA) of gamma ray emitting radionuclides inside containers using tomographic gamma scanning (TGS). High resolution gamma ray spectroscopy is used to detect and quantify the radionuclides of interest. The attenuation of an external gamma ray transmission source is used to correct the measurement of the emission gamma rays from radionuclides to arrive at a quantitative determination of the radionuclides present in the item.
1.2 The TGS technique covered by the test method may be used to assay scrap or waste material in cans or drums in the 1 to 500 litre volume range. Other items may be assayed as well.
1.3 The test method will cover two implementations of the TGS procedure: (1) Isotope Specific Calibration that uses standards of known radionuclide masses (or activities) to determine system response in a mass (or activity) versus corrected count rate calibration, that applies to only those specific radionuclides for which it is calibrated, and (2) Response Curve Calibration that uses gamma ray standards to determine system response as a function of gamma ray energy and thereby establishes calibration for all gamma emitting radionuclides of interest.
1.4 This test method will also include a technique to extend the range of calibration above and below the extremes of the measured calibration data.
1.5 The assay technique covered by the test method is applicable to a wide range of item sizes, and for a wide range of matrix attenuation. The matrix attenuation is a function of the matrix composition, photon energy, and the matrix density. The matrix types that can be assayed range from light combustibles to cemented sludge or concrete. It is particularly well suited for items that have heterogeneous matrix material and non-uniform radioisotope distributions. Measured transmission values should be available to permit valid attenuation corrections, but are not neede...

SIGNIFICANCE AND USE
NDA measurement practices aimed at achieving quality results are described in this guide. The application of the material provided in this guide should be determined on a case by case basis. Not all elements are required for all applications.
Nondestructive assay measurements are typically performed when the items measured or goals of the measurement program favor NDA over destructive analysis. NDA is typically favored when collecting a representative sample of the item is difficult or impractical (for example, scrap and waste items), personnel exposure would be significant, spread of contamination from sampling would occur, generation of secondary waste must be minimized, the weight and/or tare weight of the item cannot easily be determined (for example, in place process equipment), rapid turn around of the measurement results is needed, or the NDA measurement is significantly less expensive than the equivalent destructive analysis.
The principles provided in this guide should be used to determine which type of measurement is best suited to the measurement application. This determination involves consideration of the characteristics of the items to be measured, as well as the goals of the measurement program.
This guide applies to the suite of NDA instruments and measurement methods, many of which are described in detail in Refs (1) and (2). A partial listing of measurement methods and applicable use references is provided in 5.6.1. It is incumbent upon the user to seek additional guidance within ASTM method-specific standards, as this guide does not take precedence. Additional information on specific methods is best found in technical meeting transactions, journals, commercial application notes, and NRC/DOE publications.
This guide may be applied to many situations spanning the range of nuclear materials from product through waste. Typical applications include: the measurement and characterization of transuranic wastes, low-level wastes, and mixed wastes; ...
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1.1 This guide is a compendium of Quality Measurement Practices for performing measurements of radioactive material using nondestructive assay (NDA) instruments. The primary purpose of the guide is to assist users in arriving at quality NDA results, that is, results that satisfy the end user’s needs. This is accomplished by providing an acceptable and uniform basis for the collection, analysis, comparison, and application of data. The recommendations are not compulsory or prerequisites to achieving quality NDA measurements, but are considered contributory in most areas.
1.2 This guide applies to the use of NDA instrumentation for the measurement of nuclear materials by the observation of spontaneous or stimulated nuclear radiations, including photons, neutrons, or the flow of heat. Recommended calibration, operating, and assurance methods represent guiding principles based on current NDA technology. The diversity of industry-wide nuclear materials measurement applications and instrumentation precludes discussion of specific measurement situations. As a result, compliance with practices recommended in this guide must be based on a thorough understanding of contributing variables and performance requirements of the specific measurement application.
1.3 Selection of the best instrument for a given measurement application and advice on the use of this instrument must be provided by a qualified NDA professional following guidance provided in Guide C 1490. This guide is to be used as a reference, and to supplement the critical thinking, professional skill, expert judgment, and experimental test and verification needed to ensure that the instrumentation and methods have been properly implemented.
1.4 The intended audience for this guide includes but is not limited to Management, Auditor Support, NDA Qualified Instrument Operators, NDA Technical Specialists, and NDA Professionals.
1.5 The values stated in either SI units...

SIGNIFICANCE AND USE
This test method is the most accurate NDA technique for the assay of many physical forms of Pu. Isotopic measurements by gamma-ray spectroscopy or destructive analysis techniques are part of the test method when it is applied to the assay of Pu.
Calorimetry has been applied to a wide variety of Pu-bearing solids including metals, alloys, oxides, fluorides, mixed Pu-U oxides, mixed oxide fuel pins, waste, and scrap, for example, ash, ash heels, salts, crucibles, and graphite scarfings) (2,3). This test method has been routinely used at U.S. and European facilities for Pu process measurements and nuclear material accountability for the last 40 years (2-9).
Pu-bearing materials have been measured in calorimeter containers ranging in size from about 0.025 m to about 0.60 m in diameter and from about 0.076 m to about 0.9 m in height.
Gamma-ray spectroscopy typically is used to determine the Pu-relative isotopic composition and 241Am to Pu ratio (see Test Method C 1030). Isotopic information from mass spectrometry and alpha counting measurements may be used (see Test Method C 697).
This test method is the most accurate NDA method for the measurement of tritium. For many physical forms of tritium compounds calorimetry is the only practical measurement technique available.
Physical standards representative of the materials being assayed are not required for the test method.
This test method is largely independent of the elemental distribution of the nuclear materials in the matrix.
The accuracy of the method can be degraded for materials with inhomogeneous isotopic composition.
The thermal power measurement is traceable to national measurement systems through electrical standards used to directly calibrate the calorimeters or to calibrate secondary 238Pu heat standards.
Heat-flow calorimetry has been used to prepare secondary standards for neutron and gamma-ray assay systems (7-12).
Calorimetry measurement times are typically longer than other NDA techni...
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1.1 This test method describes the nondestructive assay (NDA) of plutonium, tritium, and 241Am using heat flow calorimetry. For plutonium the typical range of applicability corresponds to ~1 g to ~2000 g quantities while for tritium the typical range extends from ~0.001 g to~ 10 g. This test method can be applied to materials in a wide range of container sizes up to 50 L. It has been used routinely to assay items whose thermal power ranges from 0.001 W to 135 W.
1.2 This test method requires knowledge of the relative abundances of the plutonium isotopes and the  241Am/Pu mass ratio to determine the total plutonium mass.
1.3 This test method provides a direct measure of tritium content.
1.4 This test method provides a measure of 241Am either as a single isotope or mixed with plutonium.
1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.
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 and health practices and determine the applicability of regulatory limitations prior to use.

SIGNIFICANCE AND USE
This test method is the most accurate NDA technique for the assay of many physical forms of Pu. Isotopic measurements by gamma-ray spectroscopy or destructive analysis techniques are part of the test method when it is applied to the assay of Pu.
Calorimetry has been applied to a wide variety of Pu-bearing solids including metals, alloys, oxides, fluorides, mixed Pu-U oxides, mixed oxide fuel pins, waste, and scrap, for example, ash, ash heels, salts, crucibles, and graphite scarfings) (2,3). This test method has been routinely used at U.S. and European facilities for Pu process measurements and nuclear material accountability for the last 40 years (2-9).
Pu-bearing materials have been measured in calorimeter containers ranging in size from about 0.025 m to about 0.60 m in diameter and from about 0.076 m to about 0.9 m in height.
Gamma-ray spectroscopy typically is used to determine the Pu-relative isotopic composition and 241Am to Pu ratio (see Test Method C 1030). Isotopic information from mass spectrometry and alpha counting measurements may be used (see Test Method C 697).
This test method is the most accurate NDA method for the measurement of tritium. For many physical forms of tritium compounds calorimetry is the only practical measurement technique available.
Physical standards representative of the materials being assayed are not required for the test method.
This test method is largely independent of the elemental distribution of the nuclear materials in the matrix.
The accuracy of the method can be degraded for materials with inhomogeneous isotopic composition.
The thermal power measurement is traceable to national measurement systems through electrical standards used to directly calibrate the calorimeters or to calibrate secondary 238Pu heat standards.
Heat-flow calorimetry has been used to prepare secondary standards for neutron and gamma-ray assay systems (7-12).
Calorimetry measurement times are typically longer than other NDA techni...
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1.1 This test method describes the nondestructive assay (NDA) of plutonium, tritium, and 241Am using heat flow calorimetry. For plutonium the typical range of applicability corresponds to ~1 g to ~2000 g quantities while for tritium the typical range extends from ~0.001 g to~ 10 g. This test method can be applied to materials in a wide range of container sizes up to 50 L. It has been used routinely to assay items whose thermal power ranges from 0.001 W to 135 W.
1.2 This test method requires knowledge of the relative abundances of the plutonium isotopes and the  241Am/Pu mass ratio to determine the total plutonium mass.
1.3 This test method provides a direct measure of tritium content.
1.4 This test method provides a measure of 241Am either as a single isotope or mixed with plutonium.
1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.
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 and health practices and determine the applicability of regulatory limitations prior to use.

SIGNIFICANCE AND USE
The enrichment meter principle provides a nondestructive measurement of the 235U fraction of uranium-bearing items. Sampling is not required and no waste is generated, minimizing exposure to hazardous materials and resulting in reduced sampling error.
This method relies on a fixed and controlled geometry. The uranium-bearing materials in the measured items and calibration reference materials used for calibration must fill the detector field of view.
Use of a low resolution detector (for example, NaI detector) to measure uranium with 235U fraction approximately 10 % which is contained in a thin-walled container can provide a rapid (typically 100 s), easily portable measurement system with precision of 0.6 % and bias of less than 1 %.
Use of a high resolution detector (for example, high-purity germanium) can provide measurement with a precision better than 0.2 % and a bias less than 1 % within a 300-s measurement time when measuring uranium with 235U fraction in the range of 0.711 % or above which is contained in thin-walled containers.
In order to obtain optimum results using this method, the chemical composition of the item must be well known, the container wall must permit transmission of the 185.7 keV gamma-ray, and the uranium-bearing material within the item must be infinitely thick with respect to the 185.7 keV gamma-ray. All items must be in identical containers or must have a known container wall thickness and composition.
Items to be measured must be homogeneous with respect to both 235U fraction and chemical composition.
When measuring items, using low-resolution detectors, in thin-walled containers that have not reached secular equilibrium (more than about 120 days after processing), either the method should not be used, additional corrections should be made to account for the age of the uranium, or high-resolution measurements should be performed.
The method is often used as a enrichment verification technique.
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1.1 This test method covers the quantitative determination of the fraction of 235U in uranium using measurement of the 185.7 keV gamma-ray produced during the decay of 235U.
1.2 This test method is applicable to items containing homogeneous uranium-bearing materials of known chemical composition in which the compound is considered infinitely thick with respect to 185.7 keV gamma-rays.
1.3 This test method can be used for the entire range of 235U fraction as a weight percent, from depleted (0.2 % 235U) to highly enriched (97.5 % 235U).
1.4 Measurement of items that have not reached secular equilibrium between 238U and 234Th may not produce the stated bias when low-resolution detectors are used with the computational method listed in Annex A2.
1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.
1.6 This standard may involve hazardous materials, operations, and equipment. 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.

SIGNIFICANCE AND USE
This test method is useful for determining the plutonium content of items such as impure Pu oxide, mixed Pu/U oxide, oxidized Pu metal, Pu scrap and waste, Pu process residues, and weapons components.
Measurements made with this test method may be suitable for safeguards or waste characterization requirements such as:
Nuclear materials accountability,
Inventory verification (7),
Confirmation of nuclear materials content (8),
Resolution of shipper/receiver differences (9),
Excess weapons materials inspections (10, 11),
Safeguards termination on waste (12, 13),
Determination of fissile equivalent content (14).
A significant feature of neutron multiplicity counting is its ability to capture more information than neutron coincidence counting because of the availability of a third measured parameter, leading to reduced measurement bias for most material categories for which suitable precision can be attained. This feature also makes it possible to assay some in-plant materials that are not amenable to conventional coincidence counting, including moist or impure plutonium oxide, oxidized metal, and some categories of scrap, waste, and residues (10).
Calibration for many material types does not require representative standards. Thus, the technique can be used for inventory verification without calibration standards (7), although measurement bias may be lower if representative standards were available.
The repeatability of the measurement results due to counting statistics is related to the quantity of nuclear material, interfering neutrons, and the count time of the measurement (15).
For certain materials such as small Pu, items of less than 1 g, some Pu-bearing waste, or very impure Pu process residues where the (α,n) reaction rate overwhelms the triples signal, multiplicity information may not be useful because of the poor counting statistics of the triple coincidences within practical counting times (12).
For pure Pu metal, pure oxide, or other...
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1.1 This test method describes the nondestructive assay of plutonium in forms such as metal, oxide, scrap, residue, or waste using passive neutron multiplicity counting. This test method provides results that are usually more accurate than conventional neutron coincidence counting. The method can be applied to a large variety of plutonium items in various containers including cans, 208-L drums, or 1900-L Standard Waste Boxes. It has been used to assay items whose plutonium content ranges from 1 g to 1000s of g.
1.2 There are several electronics or mathematical approaches available for multiplicity analysis, including the multiplicity shift register, the Euratom Time Correlation Analyzer, and the List Mode Module, as described briefly in Ref. (1).  
1.3 This test method is primarily intended to address the assay of 240Pu-effective by moments-based multiplicity analysis using shift register electronics (1, 2, 3) and high efficiency neutron counters specifically designed for multiplicity analysis.
1.4 This test method requires knowledge of the relative abundances of the plutonium isotopes to determine the total plutonium mass (See Test Method C 1030).
1.5 This test method may also be applied to modified neutron coincidence counters (4) which were not specifically designed as multiplicity counters (that is, HLNCC, AWCC, etc), with a corresponding degradation of results.
1.6 The values stated in SI units are to be regarded as 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.

SIGNIFICANCE AND USE
This test method is used to determine the U and Pu content of scrap and waste in containers. Active measurement times have typically been 100 to 1000 s. Passive measurement times have typically been 400 s to several hours. The following limits may be further restricted depending upon specific matrix, calibration material, criticality safety, or counting equipment considerations.
The passive measurement has been applied to benign matrices in 208 L drums with Pu content ranging from 30 mg to 1 kg.
The active measurement has been applied to waste drums with 235U content ranging from about 100 mg to 1 kg.
This test method can be used to demonstrate compliance with the radioactivity levels specified in safeguards, waste, disposal, and environmental regulations (for example, see NRC regulatory guides 5.11, 5.53, DOE Order 5820.2a, and 10CFR61 sections 61.55 and sections 61.56, 40CFR191, and DOE/WIPP-069).  
This test method could be used to detect diversion attempts that use shielding to encapsulate nuclear material.
The bias of the measurement results is related to the item size and density, the homogeneity and composition of the matrix, and the quantity and distribution of the nuclear material. The precision of the measurement results is related to the quantity of nuclear material and the count time of the measurement.
For both the matrix-specific and the matrix-correction approaches, the method assumes the calibration materials match the items to be measured with respect to the homogeneity and composition of the matrix, the neutron moderator and absorber content, and the quantity of nuclear material, to the extent they affect the measurement.
It is recommended that measurements be made on small containers of scrap and waste before they are combined in large containers. Special arrangement may be required to assay small containers to best effect in a large cavity general purpose shuffer.
It is recommended that measurements be made on containers with homog...
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1.1 This test method covers the nondestructive assay of scrap and waste items for U, Pu, or both, using a 252Cf shuffler. Shuffler measurements have been applied to a variety of matrix materials in containers of up to several 100 L. Corrections are made for the effects of matrix material. Applications of this test method include measurements for safeguards, accountability, TRU, and U waste segregation, disposal, and process control purposes (1, 2, 3).  
1.1.1 This test method uses passive neutron coincidence counting (4) to measure the 240Pu-effective mass. It has been used to assay items with total Pu contents between 0.03 g and 1000 g. It could be used to measure other spontaneously fissioning isotopes such as Cm and Cf. It specifically describes the approach used with shift register electronics; however, it can be adapted to other electronics.  
1.1.2 This test method uses neutron irradiation with a moveable Cf source and counting of the delayed neutrons from the induced fissions to measure the 235U equivalent fissile mass. It has been used to assay items with 235U contents between 0.1 g and 1000 g. It could be used to assay other fissile and fissionable isotopes.
1.2 This test method requires knowledge of the relative isotopic composition (See Test Method C 1030) of the special nuclear material to determine the mass of the different elements from the measurable quantities.
1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.
1.4 The techniques described in this test method have been applied to materials other than scrap and waste. These other applications are not addressed in this test method.
1.5 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...

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1.1 The terminology defined in this document is associated with nondestructive assay of nuclear material.
1.2 All of the definitions are associated with measurement techniques that measure nuclear emissions (that is, neutrons, gamma-rays, or heat) directly or indirectly.
1.3 definitions are relevant to any standards and guides written by subcommittee C26.10.

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1.1 This test method describes gamma-ray methods used to nondestructively measure the quantity of 235U, or 239Pu remaining as holdup in nuclear facilities. Holdup occurs in all facilities where nuclear material is processed, in process equipment, in exhaust ventilation systems and in building walls and floors.
1.2 This test method includes information useful for management, planning, selection of equipment, consideration of interferences, measurement program definition, and the utilization of resources (1, 2, 3, 4).  
1.3 The measurement of nuclear material hold up in process equipment requires a scientific knowledge of radiation sources and detectors, transmission of radiation, calibration, facility operations and error 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. The measurement process includes defining measurement uncertainties and is sensitive to the form and 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, to identify the form of the material, and to measure it without interference from adjacent sources of radiation.
1.4 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.

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1.1 The terminology defined in this document is associated with nondestructive assay of nuclear material.
1.2 All of the definitions are associated with measurement techniques that measure nuclear emissions (that is, neutrons, gamma-rays, or heat) directly or indirectly.
1.3 definitions are relevant to any standards and guides written by subcommittee C26.10.

SIGNIFICANCE AND USE
This test method is a nondestructive means of determining the nuclide concentration of a solution for special nuclear material accountancy, nuclear safety, and process control.
It is assumed that the nuclide to be analyzed is in a homogeneous solution (Practice C 1168).
The transmission correction makes the test method independent of matrix and useful over several orders of magnitude of nuclide concentrations. However, a typical configuration will normally span only two to three orders of magnitude.
The test method assumes that the sample-detector geometry is the same for all measured items. This can be accomplished by requiring that the liquid height in the sidelooking geometry exceeds the detector field of view defined by the collimator. For the uplooking geometry, a fixed sample fill height must be maintained and vials of identical radii must be used unless the vial radius exceeds the field of view defined by the collimator.
Since gamma-ray systems can be automated, the test method can be rapid, reliable, and not labor intensive.
This test method may be applicable to in-line or off-line situations.
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1.1 This test method covers the determination of the concentration of gamma-ray emitting special nuclear materials dissolved in homogeneous solutions. The test method corrects for gamma-ray attenuation by the sample and its container by measurement of the transmission of a beam of gamma rays from an external source (Refs. (1), (2), and  (3)).
1.2 Two sample geometries, slab and cylinder, are considered. The sample container that determines the geometry may be either a removable or a fixed geometry container. This test method is limited to sample containers having walls or a top and bottom of equal transmission through which the gamma rays from the external transmission correction source must pass.
1.3 This test method is typically applied to radionuclide concentrations ranging from a few milligrams per litre to several hundred grams per litre. The assay range will be a function of the specific activity of the nuclide of interest, the physical characteristics of the sample container, counting equipment considerations, assay gamma-ray energies, sample matrix, gamma-ray branching ratios, and interferences.
1.4 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. For specific hazards, see Section 9.

SIGNIFICANCE AND USE
The process of selection, training and qualification of personnel involved with NDA measurements is one of the quality assurance elements for an overall quality NDA measurement program.
This guide describes an approach to selection, qualification, and training of personnel that is to be used in conjunction with other NDA QA program elements. The selection, qualification and training processes can vary and this guide provides one such approach.
The qualification activities described in this guide assume that NDA personnel are already proficient in general facility operations and safety procedures. The training and activities that developed this proficiency are not covered in this guide.
This guide describes a basic approach and principles for the qualification of NDA professionals and technical specialists and operators. A different approach may be adopted by the management organization based on its particular organization and facility specifics. However, if a variation of the approach of this guide is applied, the resulting selection, training, and qualification programs must meet the requirements of the facility quality assurance program and should provide all the applicable functions of Section 4.
This guide may be used as an aid in the preparation of a Training Implementation Plan (TIP) for the Transuranic Waste Characterization Program (TWCP). Requirements for TWCP training include but are not limited to: 10 CFR 830.120, DOE Order 5480.20A, DOE QAPD CAO-94-1012, TWCP-QAPP CAO-94-1010, TWCP-QAPjP and the TUM.
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1.1 This guide contains good practices for the selection, training, qualification, and professional development of personnel performing analysis, calibration, physical measurements, or data review using nondestructive assay equipment, methods, results, or techniques. The guide also covers NDA personnel involved with NDA equipment setup, selection, diagnosis, troubleshooting, or repair. Selection, training, and qualification programs based on this guide are intended to provide assurance that NDA personnel are qualified to perform their jobs competently. This guide presents a series of options but does not recommend a specific course of action.

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1.1 This guide is a compendium of Good Practices for performing measurements of radioactive material using nondestructive assay (NDA) instruments. The primary purpose of the guide is to assist its users in arriving at quality NDA results, that is, results that satisfy the end users needs. This is accomplished by providing an acceptable and uniform basis for the collection, analysis, comparison, and application of data. The recommendations are not compulsory or pre requisites to achieving quality NDA measurements, but are considered contributory in most areas.
1.2 This guide applies to the use of NDA instrumentation for the measurement of nuclear materials by the observation of spontaneous or stimulated nuclear radiations, including photons, neutrons, or the flow of heat. Recommended calibration, operating, and assurance methods represent guiding principles based on current NDA technology. The diversity of industry-wide nuclear materials measurement applications and instrumentation precludes discussion of specific measurement situations. As a result, compliance with practices recommended in this guide must be based on a thorough understanding of contributing variables and performance requirements of the specific measurement application.
1.3 Selection of the best instrument for a given measurement application and advice on the use of this instrument must be provided by a qualified NDA professional following guidance provided in Guide C 1490. This guide is to be used as a reference, and to supplement the critical thinking, professional skill, expert judgement, and experimental test and verification needed to ensure that the instrumentation and methods have been properly implemented.
1.4 The intended audience for this guide includes but is not limited to Management, Auditor Support, NDA Qualified Instrument Operators, NDA Technical Specialists, and NDA Professionals.

SIGNIFICANCE AND USE
Segmented gamma-ray scanning provides a nondestructive means of measuring the nuclide content of scrap and waste where the specific nature of the matrix and the chemical form and relationship between the nuclide and matrix may be unknown.
The procedure can serve as a diagnostic tool that provides a vertical profile of transmission and nuclide concentration within the item.
Sample preparation is generally limited to good waste/scrap segregation practices that produce relatively homogeneous items that are required for any successful waste/inventory management and assay scheme, regardless of the measurement method used. Also, process knowledge should be used, when available, as part of a waste management program to complement information on sample parameters, container properties, and the appropriateness of calibration factors.
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1.1 This test method covers the transmission-corrected nondestructive assay (NDA) of gamma-ray emitting special nuclear materials (SNMs), most commonly 235U, 239Pu, and  241Am, in low-density scrap or waste, packaged in cylindrical containers. The method can also be applied to NDA of other gamma-emitting nuclides including fission products. High-resolution gamma-ray spectroscopy is used to detect and measure the nuclides of interest and to measure and correct for gamma-ray attenuation in a series of horizontal segments (collimated gamma detector views) of the container. Corrections are also made for counting losses occasioned by signal processing limitations (1-3).
1.2 There are currently several systems in use or under development for determining the attenuation corrections for NDA of radioisotopic materials (4-8). A related technique, tomographic gamma-ray scanning (TGS), is not included in this test method (9,10).
1.2.1 This test method will cover two implementations of the Segmented Gamma Scanning (SGS) procedure: (1) Isotope Specific (Mass) Calibration, the original SGS procedure, uses standards of known radionuclide masses to determine detector response in a mass versus corrected count rate calibration that applies only to those specific radionuclides for which it is calibrated, and (2) Efficiency Curve Calibration, an alternative method, typically uses non-SNM radionuclide sources to determine system detection efficiency vs. gamma energy and thereby calibrate for all gamma-emitting radionuclides of interest (11). These two methods will be covered in detail in the remainder of the main body of this test method and Annex A1.
1.2.1.1 Efficiency Curve Calibration, over the energy range for which the efficiency is defined, has the advantage of providing calibration for many gamma-emitting nuclide for which half-life and gamma emission intensity data are available.
1.3 The assay technique may be applicable to loadings up to several hundred grams of nuclide in a 208-L (55-gal) drum, with more restricted ranges to be applicable depending on specific packaging and counting equipment considerations.
1.4 Measured transmission values must be available for use in calculation of segment-specific attenuation corrections at the energies of analysis.
1.5 A related method, SGS with calculated correction factors based on sample content and density, is not included in this standard.
1.6 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.
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. Specific precautionary statements are given in Section 8.

SIGNIFICANCE AND USE
The determination of isotopic composition by gamma-ray spectrometry is a nondestructive technique and when used with other nondestructive techniques, such as calorimetry (Test Method C 1458) or neutron counting (Test Methods C 1207C 1493C 1500), can provide a totally nondestructive plutonium assay necessary for material accountancy and safeguards needs.
Since gamma-ray spectrometry systems are typically automated, the routine use of the test method is fast, reliable, and is not labor intensive. Since the test method is nondestructive, requiring no sample preparation, it does not create waste disposal problems.
This test method assumes that the isotopic composition of plutonium in the sample being measured is homogeneous (see see 7.2.4 and (5)).
The 242Pu abundance is not measured by this test method and must be estimated from isotopic correlation techniques, stream averages, historical information, or other measurement techniques.
A daughter product of 241Pu is 241Am. The 241Am/239Pu atom ratio can also be determined by means of this test method (assuming a homogeneous isotopic distribution of plutonium and 241Am) and is necessary for the correct interpretation of a calorimetric heat measurement.
The isotopic composition of a given batch or sample of plutonium is an attribute of that sample and, once determined, can be used in subsequent inventory measurements to verify the identity of a sample within the measurement uncertainties.
The method can also measure the ratio of other gamma emiting isotopes to plutonium assuming they have the same spatial distribution as the plutonium in the sample. Some of these “other” gamma-emitting isotopes include isotopes of uranium, neptium, curium, cesium, and other fission products. (The same methods of this standard can be used to measure the isotopic composition of uranium in samples containing only uranium (4–6)).
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1.1 This test method is applicable to the determination of isotopic abundances in isotopically homogeneous Pu-bearing materials. This test method may be applicable to other plutonium-bearing materials, some of which may require modifications to the described test method.
1.2 The procedure is applicable to sample sizes ranging from a few tenths of a gram up to the maximum sample weight allowed by criticality limits.
1.3 Because  242 Pu has no useful gamma-ray signature, its isotopic abundance is not determined. Isotopic correlation techniques may be used to estimate its relative abundance (Refs 1, 2).
1.4 This test method has been demonstrated in routine use for isotopic abundances ranging from 94 to 70%  239 Pu. This test method has also been employed for isotopic abundances outside this range.
1.5 The values stated in SI units are to be regarded as the standard.
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 and health practices and determine the applicability of regulatory limitations prior to use.

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1.1 This test method describes the nondestructive assay of scrap or waste for plutonium content using passive thermal-neutron coincidence counting. This test method provides rapid results and can be applied to a variety of carefully sorted materials in containers as large as 208-L drums. The test method applies to measurements of 238Pu, 240Pu, and 242Pu and has been used to assay items whose total plutonium content ranges from 0.01 to 6000 g (1).
1.2 This test method requires knowledge of the relative abundances of the plutonium isotopes to determine the total plutonium mass.
1.3 This test method may not be applicable to the assay of scrap or waste containing other spontaneously fissioning nuclides.
1.3.1 This test method may give biased results for measurements of containers that include large amounts of hydrogenous materials.
1.3.2 The techniques described in this test method have been applied to materials other than scrap and waste (2, 3).
1.4 This test method assumes the use of shift-register-based coincidence technology (4).
1.5 Several other techniques that are related to passive neutron coincidence counting exist These include neutron multiplicity counting (5,6), add-a-source analysis (7), and cosmic-ray rejection (8). Discussions of these techniques are not included in this 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 and health practices and determine the applicability of regulatory limitations prior to use.

SIGNIFICANCE AND USE
The enrichment meter principle provides a nondestructive measurement of the 235U fraction of uranium-bearing items. Sampling is not required and no waste is generated, minimizing exposure to hazardous materials and resulting in reduced sampling error.
Use of a low resolution detector (e.g., NaI detector) to measure uranium with 235U fraction approximately 10 % which is contained in a thin-walled container can provide a rapid (typically 100 s), easily portable measurement system with precision of 0.6 % and bias of less than 1 %.
Use of a high resolution detector (e.g., high-purity germanium) can provide measurement with a precision better than 0.2 % and a bias less than 1 % within a 300-sec measurement time when measuring uranium with 235U fraction in the range of 0.711 % to 4.46 % which is contained in thin-walled containers.
In order to obtain optimum results using this method, the chemical composition of the item must be well known, the container wall must permit transmission of the 185.7 keV gamma ray, and the uranium-bearing material within the item must be infinitely thick with respect to the 185.7 keV gamma ray.
Items must be homogeneous with respect to both 235U fraction and chemical composition.
The uranium-bearing materials in the measured items and calibration reference materials used for calibration must fill the detector field of view.
When measuring items, using low-resolution detectors, in thin-walled containers that have not reached secular equilibrium, either the method should not be used, additional corrections should be made to account for the age of the uranium, or high-resolution measurements should be performed.
The method is typically used as a verification technique, not to establish enrichment.
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1.1 This test method covers the quantitative determination of the fraction of 235U in uranium using measurement of the 185.7 keV gamma ray produced during the decay of 235U.
1.2 This test method is applicable to items containing homogeneous uranium-bearing materials of known chemical composition in which the compound is considered infinitely thick with respect to 185.7 keV gamma rays.
1.3 This test method can be used for the entire range of 235U fraction, from depleted (0.2 %  235U) to very highly enriched (97.5 % 235U).
1.4 Measurement of items that have not reached secular equilibrium between 238U and 234Th, may not produce the stated bias when low-resolution detectors are used with the computational method listed in Appendix B.
1.5 This standard may involve hazardous materials, operations, and equipment. 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.