Publication Date: 4/1/84
    Pages: 12
    Date Entered: 12/15/86
    Title: NONDESTRUCTIVE ASSAY OF SPECIAL NUCLEAR MATERIAL CONTAINED IN SCRAP AND WASTE (11/73)
    Revision 1(*)
    April 1984
    U.S. NUCLEAR REGULATORY COMMISSION
    REGULATORY GUIDE
    OFFICE OF NUCLEAR REGULATORY RESEARCH
    REGULATORY GUIDE 5.11
    (Task SG 043-4) NONDESTRUCTIVE ASSAY OF SPECIAL NUCLEAR MATERIAL
    CONTAINED IN SCRAP AND WASTE
A. INTRODUCTION
    Section 70.51, "Material Balance, Inventory, and Records
    Requirements," 10 CFR Part 70, "Domestic Licensing of Special Nuclear
    Material," requires licensees authorized to possess at any one time more
    than one effective kilogram of special nuclear material (SNM) to
    establish and maintain a system of control and accountability to ensure
    that the standard error (estimator) of any inventory difference (ID)
    ascertained as a result of a measured material balance meets established
    minimum standards. The selection and proper application of an adequate
    measurement method for each of the material forms in the fuel cycle is
    essential for the maintenance of these standards.
    For some material categories, particularly scrap and waste,
    nondestructive assay (NDA) is the only practical, and sometimes the most
    accurate, means for measuring SNM content. This guide details
    procedures acceptable to the NRC staff to provide a framework for the
    use of NDA in the measurement of scrap and waste components generated in
    conjunction with the processing of SNM. Other guides detail procedures
    specific to the application of a selected technique to a particular
    problem.
    Any guidance in this document related to information collection
    activities has been cleared under OMB Clearance No. 3150-0009.
B. DISCUSSION
1. APPLICABLE NDA PRINCIPLES
    The NDA of the SNM content of heterogeneous material forms is
    usually achieved through observing either stimulated or spontaneously
    occurring radiations emitted from the isotopes of either plutonium or
    uranium, from their radioactive decay products, or from some combination
    thereof. Some NDA techniques such as absorption-edge densitometry and
    X-ray resonance fluorescence determine the elemental SNM concentration
    rather than the presence of specific isotopes. If isotopic radiation is
    measured, the isotopic composition of the material must be known or
    determined to permit a conversion of the amount of isotope measured to
    the amount of element present in the container. Assays are performed by
    isolating the container of interest to permit a measurement of its
    contents through a comparison with the response observed from known
    calibration standards. This technology permits quantitative assays of
    the SNM content of heterogeneous materials in short measurement times
    without sample preparation and without affecting the form of the
    material to be assayed. The proper application of this technology
    requires the understanding and control of factors influencing NDA
    measurements.
    1.1 Passive NDA Techniques
    Passive NDA is based on observing spontaneously emitted radiations
    created through the radioactive decay of plutonium or uranium isotopes
    or of their radioactive daughters. Radiations attributable to alpha
    (alpha) particle decay, to gamma ray transitions following alpha and
    beta (beta) particle decay, and to spontaneous fission have served as
    the basis for practical passive NDA measurements.
    1.1.1NDA Techniques Based on Alpha Particle Decay
    Alpha particle decay is indirectly detected using calorimetry
    measurements. (Note that additional contributions are attributable to
    the alpha decay of (241)Am and the beta decay of (241)Pu in plutonium
    calorimetry applications.) The kinetic energy of the emitted alpha
    particle and the recoiling daughter nucleus is transformed into heat,
    together with some fraction of the gamma ray energies that may be
    emitted by the excited daughter nucleus in lowering its energy to a more
    stable nuclear configuration. The calorimetric measurement of the heat
    produced by a sample can be converted to the amount of
    alpha-particle-emitting nuclides present through the use of the isotopic
    abundance and the specific power (W/g-s) of each nuclide (Refs. 1-3).
    Plutonium, because of the relatively high specific powers of (238)Pu and
    (240)Pu, is amenable to assay by calorimetry, with possible complication
    from the presence of alpha-active (241)Am.
    ----------
    (*)The substantial number of changes in this revision has made it
    impractical to indicate the changes with lines in the margin.
    ----------
    Another technique based on alpha decay involves the interaction of
    high-energy alpha particles with some light nuclides (e.g., (7)Li,
    (9)Be, (10)B, (18)O, and (19)F) that may produce a neutron through an
    (alpha, n) reaction (Ref. 4). When the isotopic composition of the
    alpha-particle-emitting nuclides is known and the content of high-yield
    (alpha, n) targets is fixed, the observation of the neutron yield from a
    sample can be converted to the amount of SNM present.
    1.1.2NDA Techniques Based on Gamma Ray Analysis
    The gamma ray transitions that reduce the excitation of a daughter
    nucleus following either alpha- or beta-particle emission from an
    isotope of SNM occur at discrete energies (Refs. 5, 6). The known
    alpha- or beta-particle-decay activity of the SNM parent isotope and the
    probability that a specific gamma ray will be emitted following the
    alpha- or beta-particle decay can be used to convert the measurement of
    that gamma ray to a measurement of the amount of the SNM parent isotope
    present in the container being measured. High-resolution gamma ray
    spectroscopy is required when the gamma rays being measured are observed
    in the presence of other gamma rays or X-rays that, without being
    resolved, would interfere with the measurement of the desired gamma ray
    (Ref. 5).
    1.1.3NDA Techniques Based on Spontaneous Fission
    A fission event is accompanied by the emission of an average of 2
    to 3.5 neutrons (depending on the parent nucleus) and an average of
    about 7.5 gamma rays. A total of about 200 MeV of energy is released,
    distributed among the fission fragments, neutrons, gamma rays, beta
    particles, and neutrinos. Spontaneous fission occurs with sufficient
    frequency in (238)Pu, (240)Pu, (242)Pu, and marginally in (238)U to
    facilitate assay measurements through the observation of this reaction.
    Systems requiring the coincident observation of two or more of the
    prompt radiations associated with the spontaneous fission event provide
    the basis for available measurement systems (Ref. 7).
    1.2 Active NDA Techniques
    Most active NDA is based on the observation of radiations (gamma
    rays or neutrons) that are emitted from the isotope under investigation
    when that isotope undergoes a transformation resulting from an
    interaction with stimulating radiation provided by an appropriate
    external source. Isotopic (Refs. 8, 9) and accelerator (Ref. 7) sources
    of stimulating radiation have been investigated. For a thorough
    discussion of active NDA techniques, see Reference 10.
    Stimulation with accelerator-generated high-energy neutrons or
    gamma rays is normally considered only after all other NDA methods have
    been evaluated and found to be inadequate. Operational requirements,
    including operator qualifications, maintenance, radiation shielding, and
    calibration considerations, normally require an inordinate level of
    support in comparison to the benefits of in-plant application.
    Neutron bombardment readily induces fissions of (233)U, (235)U,
    (239)Pu, and (241)Pu. Active NDA systems have been developed using
    spontaneous fission ((252)Cf) neutron sources, as well as (gamma, n)
    (Sb-Be) sources and a variety of (alpha, n) (Am-Li, Pu-Li, Pu-Be)
    sources (Refs. 8, 9). Active techniques rely on one of the following
    three properties of the induced fission radiation to distinguish the
    induced radiation from the background and the stimulating radiation:
    * High-energy radiation (neutrons with about 2 MeV energy and
    gamma rays with 1-2 MeV energy) * Coincident radiation (simultaneous emission of two or more
    neutrons and about seven to eight gamma rays) * Delayed radiation (neutrons emitted from certain fission
    products with half-lives ranging from 0.2 to 50 seconds and
    gamma rays emitted from fission products with half-lives
    ranging from submicro-seconds to years. The usable delayed
    gamma rays are emitted from fission products with half-lives
    similar to those of delayed-neutron-emitting fission
    products.) Examples of the use of these properties with the types of isotopic
    neutron sources listed above are (1) fissions are induced by low-energy
    neutrons from a (124)Sb-Be source, and energetic fission neutrons are
    counted (Refs. 9, 11); (2) fissions are induced by an intense (252)Cf
    source, and delayed neutrons are counted after the source has been
    withdrawn (Refs. 9, 12-14); and (3) fissions are induced by single
    emitted neutrons from an (alpha, n) source (Refs. 9, 15). Coincident
    gamma rays and neutrons resulting from the induced fission are counted
    by means of electronic timing gates (of less than 100 microseconds
    duration) that discriminate against noncoincident events (Refs. 9, 13).
2. FACTORS AFFECTING THE RESPONSE OF NDA SYSTEMS
    Regardless of the technique selected, the observed NDA response
    depends on (1) the operational characteristics of the system, (2) the
    isotopic composition of the SNM, (3) the amount and distribution of SNM,
    (4) the amount and distribution of other materials within the container,
    and (5) the composition and dimensions of the container itself. Each of
    these variables increases the overall uncertainty associated with an NDA
    measurement.
    The observed NDA response represents contributions from the
    different SNM isotopes present in the container. To determine the amount
    of SNM present, the isotopic composition of the SNM must be known
    (except for cases in which the NDA system measures the isotopic
    composition) and the variation in the observed response as a function of
    varying isotopic composition must be understood. The effects due to
    items (3), (4), and (5) on the observed response can be reduced through
    appropriate selection of containers, compatible segregation of scrap and
    waste categories, and consistent use of packaging procedures designed to
    improve the uniformity of container loadings.
    2.1 Operational Characteristics
    The operational characteristics of the NDA system, together with
    the ability of the system to resolve the desired response from a
    composite signal, determine the ultimate usefulness of the system.
    These operational characteristics include (1) operational stability, (2)
    uniform detection efficiency, (3) stimulating radiation uniformity (for
    active systems), and (4) energy of the stimulating radiation.
    The impact of these operational characteristics on the uncertainty
    of the measured response can be reduced through the design of the
    system, the use of radiation shielding (where required), and
    standardized packaging and handling (as discussed below and in Reference
    16).
    2.1.1Operational Stability
    The ability of an NDA system to reproduce a given measurement may
    be sensitive to fluctuations in the operational environment.
    Temperature, humidity, line voltage variations, electromagnetic fields,
    and microphonics affect NDA systems to some extent. These effects may
    be manifested through the introduction of spurious electronic noise or
    changes in the high voltage applied to detectors or amplifiers, thereby
    changing the detection efficiency. To the extent that it is possible, a
    measurement technique and the hardware to implement that technique are
    selected to be insensitive to changes routinely expected in the
    operational environment. Accordingly, the instrument is designed to
    minimize environmental effects by placing components that operate at
    high voltages in hermetically sealed enclosures and shielding sensitive
    components from spurious noise pickup. In addition, electronic gain
    stabilization of the pulse-processing electronics may be advisable. As
    a final measure, the instrument environment can be controlled (e.g.,
    through the use of a dedicated environmental enclosure for the
    instrument hardware) if expected environmental fluctuations result in
    severe NDA response variations that cannot be eliminated through
    calibration and operational procedures.
    The sensitivity to background radiations can be monitored and
    controlled through proper location of the system and the utilization of
    radiation shielding, if required.
    2.1.2Uniform Detection Efficiency
    For those NDA systems for which the sample or item to be counted
    is placed within a detection chamber, if the response throughout the
    detection chamber is not uniform, positioning guides or holders may be
    utilized to ensure consistent (reproducible) sample or item positioning.
    The residual geometric response dependence can be measured using an
    appropriate source that emits radiation of the type being measured. If
    the source is small with respect to the dimensions of the detection
    chamber, the system response can be measured with the source positioned
    in different locations to determine the volume of the detection chamber
    that can be reliably used.
    An encapsulated plutonium source can be used to test gamma ray
    spectroscopic systems, active or passive NDA systems detecting neutrons
    or gamma rays, or calorimetry systems. Active NDA systems can be
    operated in a passive mode (stimulating source removed) to evaluate the
    magnitude of this effect. Rotating and scanning containers during assay
    is a recommended means of reducing the response uncertainties
    attributable to residual nonuniform geometric detection sensitivity.
    2.1.3Uniformity of Stimulating Radiation
    The stimulating radiation field (i.e., interrogating neutron or
    gamma ray flux) in active NDA systems is designed to be uniform in
    intensity and energy spectrum throughout the volume of the irradiation
    chamber. The residual effect can be measured using an SNM sample that
    is small with respect to the dimensions of the irradiation chamber. The
    response can then be measured with the SNM sample positioned in
    different locations within the irradiation chamber. If the same chamber
    is employed for irradiation and detection, a single test for the
    combined geometric nonuniformity is recommended.
    Having both a uniform detection efficiency and a uniform
    stimulating radiation field is the most direct approach and the
    recommended approach to obtaining a uniform response for the combined
    effects. However, it is possible in some cases either to tailor the
    stimulating radiation field to compensate for deficiencies in the
    detection uniformity or, conversely, to tailor the detection efficiency
    to compensate for deficiencies in the stimulating radiation field. An
    example of this combined approach is the use of interrogating sources on
    one side of the sample and placement of detectors on the other. A
    combined uniform response in this example relies both on material closer
    to the stimulating radiation source having a higher fission probability
    but a lower induced-radiation detection probability and on material
    closer to the detector having a lower stimulated fission probability but
    a higher induced-fission radiation detection probability. This type of
    approach may be necessary when there are spatial constraints. When the
    measurement system is optimized for these combined effects, a passive
    measurement with such a system will have a greater uncertainty than
    would be obtained with a system having a uniform detection efficiency.
    Various methods have been used to reduce the response uncertainty
    attributable to a nonuniform stimulating radiation field, including
    rotating and scanning the container, source scanning, distributed
    sources, and combinations of these methods.
    2.1.4Energy of Stimulating Radiation
    If the energy of the stimulating radiation is as high as
    practicable but below the threshold of any interfering reactions such as
    the neutron-induced fission in (238)U, the penetration of the
    stimulating radiation will be enhanced throughout the volume of the
    irradiation chamber. A high-energy source providing neutrons above the
    energy of the fission threshold for a fertile constituent such as (238)U
    or (232)Th can be employed to assay the fertile content of a container.
    The presence of extraneous materials, particularly those of low
    atomic number, lowers the energy spectrum of the interrogating neutron
    flux in active neutron NDA systems. Incorporating a thermal neutron
    detector to monitor this effect and thereby provide a basis for a
    correction to reduce the response uncertainty caused by this variable
    effect is recommended.
    Active neutron NDA systems with the capability to moderate the
    interrogating neutron spectrum can provide increased assay sensitivity
    for samples containing small amounts of fissile material (less than 100
    grams). This moderation capability should be removable to enhance the
    range of usefulness of the system.
    2.2 Response Dependence on SNM Isotopic Composition
    The observed NDA response may be a composite of contributions from
    more than a single isotope of uranium or plutonium. Observed effects
    are generally attributable to one of the three sources described below.
    2.2.1Multiple Gamma Ray Sources
    Plutonium contains the isotopes (238)Pu through (242)Pu in varying
    quantities. With the exception of (242)Pu, these isotopes emit many
    gamma rays (Refs. 5, 6). The observed plutonium gamma ray spectrum
    represents the contribution of all gamma rays from each isotope,
    together with the gamma rays emitted in the decay of (241)Am, which may
    also be present.
    Gamma rays from (233)U and (235)U are generally lower in energy
    than those from (239)Pu. However, (232)U, which occurs in combination
    with (233)U, has a series of daughter products that emit prolific and
    energetic gamma rays. It should be noted that one of these daughter
    products is (228)Th, and therefore the daughter products of (232)U and
    (232)Th are identical beyond (228)Th.
    2.2.2Multiple Spontaneously Fissioning Plutonium Isotopes
    In addition to the spontaneous fission observed from (240)Pu, the
    minor isotopes (238)Pu and (242)Pu typically contribute a few percent to
    the total neutron rate observed (Refs. 17-19). In mixtures of uranium
    and plutonium blended for reactor fuel applications, the spontaneous
    fission yield from (238)U may approach one percent of the (240)Pu yield.
    2.2.3Multiple Fissile Isotopes
    In active systems, the observed fission response may consist of
    contributions from more than one isotope. For uranium, if the energy
    spectrum of the stimulating radiation extends above the threshold for
    (238)U fission, that response contribution will be in addition to the
    induced (235)U fission response.
    In plutonium, the observed response will be the sum of
    contributions from the variable content of (239)Pu and (241)Pu, with
    small contributions from the even plutonium isotopes.
    When elements (e.g., plutonium and uranium) are mixed for reactor
    utilization, the uncertainty in the response is compounded by
    introducing additional fissile components in variable combinations.
    2.3 Response Dependence on Amount and Distribution of SNM in a
    Container
    If a system has a geometrically uniform detection sensitivity and
    a uniform field of stimulating radiation (where applicable), a variation
    in the response per gram of the isotope or isotopes being measured is
    generally attributable to one of the three causes described below.
    2.3.1Self-Absorption of the Emitted Radiation Within the SNM
    For a fixed amount of SNM in a container, the probability that
    radiation emitted by the SNM nuclei will interact with other SNM atoms
    increases as the localized density of the SNM increases within the
    container. This is a primary source of uncertainty in gamma ray
    spectroscopy applications. It becomes increasingly important as the SNM
    aggregates into lumps and is more pronounced for low-energy gamma rays.
    2.3.2Multiplication of the Detected Radiation
    The neutrons given off in either a spontaneous or an induced
    fission reaction can be absorbed in a fissile nucleus and subsequently
    induce that nucleus to fission, resulting in the emission of two or more
    neutrons. Multiplication affects the response of active NDA systems,
    passive coincidence neutron or gamma ray detection systems (used to
    detect spontaneous fission), and passive neutron systems used to count
    (alpha, n) neutrons. Multiplication becomes increasingly pronounced as
    the energy of the neutrons traversing the container becomes lower or as
    the density of SNM increases within the container. For further details
    on multiplication effects, see References 20 and 21.
    2.3.3Self-Shielding of the Stimulating Radiation
    Attenuation of incident radiation by the SNM, or self-shielding,
    is particularly pronounced in active systems incorporating a neutron
    source to stimulate the fissile isotopes of the SNM to fission. More of
    the incident low-energy neutrons will be absorbed near the surface of a
    high-density lump of SNM, and fewer will penetrate deeper into the lump.
    Thus, the fissile nuclei located deep in the lump will not be stimulated
    to fission at the same rate as the fissile nuclei located near the
    surface, and a low assay content will be indicated. This effect is
    dependent on the energy spectrum of the incident neutrons and the
    density of fissile nuclei. It becomes increasingly pronounced as the
    energy of the incident neutrons is decreased or as the density of the
    SNM fissile content is increased. The density of fissile nuclei is
    increased when the SNM is lumped in aggregates or when the fissile
    enrichment of the SNM is increased.
    2.4 Response Dependence on Amount and Distribution of Extraneous
    Materials Within the Container
    The presence of materials other than SNM within a container can
    affect the emitted radiations in passive and active NDA systems and can
    also affect the stimulating radiation in active assay systems. The
    presence of extraneous materials can result in either an increase or a
    decrease in the observed response.
    Effects on the observed NDA response are generally attributable to
    one of the four causes described below.
    2.4.1Interfering Radiations
    Interference arises when the material being assayed emits
    radiation that cannot be separated easily from the signal of interest.
    This problem is generally encountered in gamma ray spectroscopy and
    calorimetry applications. In gamma ray assays, the problem is manifest
    in the form of additional gamma rays that must be separated from the
    desired radiations, often with high-resolution detection systems. In
    calorimetry, the decay daughters of (241)Pu, (238)U, and (232)U
    contribute additional heat that cannot be corrected for without detailed
    knowledge of the isotopic composition of the sample.
    2.4.2Interference to Stimulating Radiation
    Material lowers the energy of neutrons through collision
    processes. This lowering of the neutron energy is called moderation.
    Low-atomic-weight elements have greater moderating power than
    high-atomic-weight elements and therefore reduce energetic neutrons to
    thermal energies with fewer collisions. Hydrogen has the greatest
    moderating power. Hydrogenous materials such as water or plastics have
    a strong moderating power because of their hydrogen content.
    Low-energy neutrons have interaction characteristics different
    from high-energy neutrons. If moderation of the stimulating neutron
    radiation occurs, the response will be altered and the assay value could
    be in error. Three effects arise from moderated neutrons. First, the
    fission probability for fissile isotopes increases with decreasing
    neutron energy. In this case, the response increases and,
    correspondingly, so does self-shielding. Second, absorption by materials
    other than SNM also increases. This absorption decreases the response
    of the system. Third, if isotopes with a fission threshold such as
    (232)Th or (238)U are being assayed with high-energy neutrons,
    moderation can lower the energy of the stimulating neutrons below the
    fission threshold. In this case, the response by these isotopes can be
    sharply reduced.
    Efforts to minimize moderation effects are particularly important
    if energetic neutrons are employed for the stimulating radiation.
    Segregation of waste categories according to their moderating
    characteristics and use of separate calibrations for each category are
    direct steps to minimize moderation effects. Another step that can be
    used with segregation, and sometimes independently, is to monitor the
    stimulating neutron radiation and then correct the assay result.
    Because several effects are associated with moderation, this latter step
    may be difficult to implement. In some cases, it may be impossible.
    2.4.3Attenuation of the Emitted Radiation
    Attenuation may range from partial energy loss of the emitted
    radiation (through scattering processes) to complete absorption of the
    radiation by the sample material. This effect can be particularly
    severe for gamma ray assay systems; unless gamma ray attenuation is
    fully accounted for by measurement or calculation, the assay value
    assigned to an unknown sample may be underestimated (Refs. 4, 22). The
    attenuation of gamma radiation increases with atomic number and material
    density within the container. Also, systems that detect emitted
    neutrons above a given energy (threshold) will observe fewer neutrons
    above the detection threshold when low-atomic-number (i.e., highly
    moderating) material is added to the container and will thus produce a
    low assay.
    The attenuation of the emitted radiation may be complete, as in
    the case of the absorption of neutrons in the nuclei of extraneous
    material. The probability for this absorption generally increases as
    the energy of the incident neutron decreases. Hence, this effect is
    further aggravated when low-atomic-number materials are present to
    reduce the energy of the emitted neutrons.
    2.4.4Attenuation of the Stimulating Radiation
    This phenomenon is similar to the phenomenon of the preceding
    section. In this instance, some portion of the stimulating radiation
    does not penetrate to the SNM within the container and thus does not
    have the opportunity to induce fission. The presence of neutron poisons
    (e.g., lithium, boron, cadmium, gadolinium) may attenuate the
    stimulating radiation to the extent that the response is independent of
    the SNM fissile content. Most materials absorb neutrons. The severity
    of this absorption effect is dependent on the type of material, its
    distribution, the energy of the stimulating neutrons, and the relative
    neutron absorbing strength of the SNM compared to the combined effect of
    the remaining material.
    The presence of extraneous material can thus alter the observed
    response, providing either a high or a low SNM content indication. This
    effect is further aggravated by nonuniformity within the container of
    either the SNM or the matrix in which it is contained. This dependence
    of response on material distributions and matrix variations is severe.
    Failure to attend to its ramifications through the segregation of scrap
    and waste categories and the utilization of representative(1)
    calibration standards may produce gross inaccuracies in NDA
    measurements.
    2.5 Response Dependence on Container Dimensions and Composition
    The items identified as potential sources of uncertainty in the
    observed response of an NDA system in Sections 2.1, 2.3, and 2.4 can be
    minimized or aggravated through the selection of containers to be
    employed when assaying SNM contained in scrap or waste.
    2.5.1Container Dimensions
    The practical limitation on container size for scrap and waste to
    be nondestructively assayed represents a compromise of throughput
    requirements and the increasing uncertainties in the observed NDA
    response incurred as a penalty for assaying large containers.
    Radiations emitted deep within the container must travel a greater
    distance to escape the confines of the container. Therefore, with
    increasing container size, the probability that radiations emitted near
    the center of the container will escape the container to the detectors
    decreases with respect to the radiations emitted near the surface of the
    container. This will result in large attenuation corrections that can
    cause added uncertainty in the assay result.
    In active neutron NDA systems, a relatively uniform field of
    stimulating radiation must be provided throughout the volume of the
    container that is observed by the detection system. This criterion is
    required to obtain a uniform response from a lump of SNM positioned
    anywhere within a container. With increasing container size, it becomes
    increasingly difficult to satisfy this criterion and maintain a compact
    geometrically efficient system. For this reason, the assay of small-size
    containers is recommended for the highest accuracy.
    ----------
    (1) The term "representative" is taken to mean representative with
    respect to attenuation, moderation, multiplication, density, and any
    other properties to which the measurement technique is sensitive.
    ----------
    If small containers are to be loaded into larger containers for
    storage or offsite shipment following assay, the size and shape of the
    inner and outer containers should be chosen to be compatible.
    Packaging in small containers will produce more containers to be
    assayed for the same scrap and waste generation rates. An offsetting
    benefit, however, is that the assay accuracy of an individual container
    should be significantly improved over that of large containers.
    2.5.2Container Structural Composition
    The structural composition of containers will affect the
    penetration of the incident or the emerging radiation. Provided all
    containers are uniform, their effect on the observed response can be
    factored into the calibration of the system. The attainable assay
    accuracy will be reduced when containers with poor penetrability or
    varying composition or dimensions are selected.
    Uniform containers of the same composition, dimensions, and wall
    thickness provide improved or best accuracy (for a given material
    category). Variability in the wall thickness of nonhydrogenous
    containers usually is not critical for neutron assays, but it can be a
    significant factor for gamma spectroscopy applications when the
    container is constructed of relatively high-density material or when
    low-energy (less than approximately 200-keV) gamma rays are being
    measured. However, when hydrogenous materials (such as polyethylene)
    are used in containers, wall thickness variability can have a
    significant effect on neutron assay results.
3. NDA FOR SNM CONTAINED IN SCRAP AND WASTE
    3.1 NDA Performance Objectives
    The measurement accuracy objectives for any material balance
    component can be estimated by considering the amount of material
    typically contained in that component. The measurement performance
    required is such that, when the uncertainty corresponding to the scrap
    and waste material balance component is combined with the uncertainties
    corresponding to the other material components, the constraints on the
    total standard error of the inventory difference (SEID) will be
    satisfied.
    3.2 NDA Technique Selection
    Factors that influence NDA technique selection are the accuracy
    requirements for the assay and the type and range of scrap and waste
    categories to be encountered. No single technique appears capable of
    meeting all requirements. When more than one type of information is
    required to separate a composite response, more than one NDA technique
    may be required to provide that information.
    3.2.1Plutonium Applications
    Calorimetry determinations are the least sensitive to matrix
    effects but rely on a detailed knowledge of the (241)Am content and the
    plutonium isotopic composition to calculate grams of plutonium from the
    measured heat flux (Ref. 1). In addition, a calorimetry measurement
    usually requires several hours in order to achieve or to carefully
    predict thermal equilibrium.
    Gamma ray spectroscopy systems complement the potential of other
    assay methods by providing the capability to verify or determine
    nondestructively the (241)Am content and the plutonium isotopic
    composition (except (242)Pu). High-resolution gamma ray systems are
    capable of extracting the maximum amount of information (elemental
    content, isotopic distributions, presence of extraneous gamma ray
    sources) from an assay, but content density severely affects the
    accuracy of quantitative predictions based on that assay method in large
    samples.
    Passive coincidence detection of the spontaneous fission yield of
    plutonium-bearing systems provides an indication of the combined
    (238)Pu, (240)Pu, and (242)Pu sample content. With known isotopic
    composition, the plutonium content can be computed (Ref. 17 and
    Regulatory Guide 5.34(2)). Neutron multiplication effects become severe
    at high plutonium sample loadings (Refs. 20, 21).
    Combining passive and active measurements in a single system is a
    valuable approach for plutonium assay. Plastic scintillation
    coincidence detection systems have been designed in conjunction with
    active neutron interrogation source systems (Ref. 23). Delayed neutron
    counting systems have an inherent active-passive counting capability
    (Refs. 9, 13, 14). Operated in passive and active modes, such systems
    are able to provide an assay of both the spontaneously fissioning
    content and the fissile content of the sample. The spontaneous fission
    and (alpha, n) backgrounds can be subtracted from an active NDA response
    to provide a yield attributable to the fissile SNM content of the
    container.
    3.2.2Uranium Applications
    Active neutron systems can provide both high-energy and moderated
    interrogation spectra. Operation with the high-energy neutron source
    will decrease the density dependence and neutron self-shielding effects,
    significantly enhancing the uniqueness of the observed response. To
    extend the applicability of such a system to small fissile loadings, a
    well-moderated interrogating spectrum can be used to take advantage of
    the increased (235)U fission probability for neutrons of low energy. In
    highly enriched uranium scrap and waste (greater than 20% (235)U),
    active NDA featuring a high-energy stimulating neutron flux is
    recommended.
    ----------
    (2) Regulatory Guide 5.34, "Nondestructive Assay for Plutonium in
    Scrap Material by Spontaneous Fission Detection." A proposed revision
    to this guide has been issued for comment as Task SG 046-4.
    ----------
    The 185-keV transition observed in the decay of (235)U is
    frequently employed in uranium applications. The penetration of this
    (235)U primary gamma ray is so poor that the gamma ray NDA technique is
    not applicable with high-density nonhomogeneous materials in large
    containers.
    Occasions arise when a passive enrichment determination is
    practical through the measurement of the 185-keV gamma ray. Enrichment
    assay applications for uranium are the subject of Regulatory Guide 5.21,
    "Nondestructive Uranium-235 Enrichment Assay by Gamma Ray Spectrometry."
    Calorimetry is not applicable to the assay of uranium because of
    the low specific alpha activity. In (233)U applications, the intense
    activity of the daughter products of (232)U imposes a severe
    complication on the use of calorimetry.
    3.3 Categorization and Segregation of Scrap and Waste for NDA
    The range of variations in the observed response of an NDA system
    attributable to the effects noted in Sections 2.3 and 2.4 can be reduced
    or controlled. Following an analysis of the types of scrap and waste
    generated in conjunction with SNM processing, a plan to segregate scrap
    and waste at the generation points can be formulated. Recovery or
    disposal compatibility is important in determining the limits of each
    category. Limiting the variability of those extraneous NDA interference
    parameters discussed in Sections 2.3 and 2.4 is a primary means of
    improving the accuracy of the scrap and waste assay. Once the
    categories are established, it is important that steps be taken to
    ensure that segregation into separate uniquely identified containers
    occurs at the generation point.
    Category limits can be established on the basis of measured
    variations observed in the NDA response of a container loaded with a
    known amount of SNM. The variation in extraneous parameters can then be
    mocked up and the resultant effect measured. In establishing
    categories, the following specific items are significant sources of
    error.
    3.3.1Calorimetry
    The presence of extraneous materials capable of absorbing heat
    (endothermic) or emitting heat (exothermic) will cause the observed
    response to be different from the correct response for the plutonium in
    the sample.
    3.3.2Neutron Measurements
    The presence of high-yield (alpha, n) target material will
    increase the number of neutrons present in the sample. A fraction of
    these neutrons will induce fission in the fissile SNM isotopes and add
    another source of error to the measurement. These multiplication and
    self-multiplication effects are discussed thoroughly in References 4,
    20, and 21.
    3.3.3Gamma Ray Measurements
    Gamma rays are severely attenuated in interactions with heavy
    materials. Mixing contaminated combustibles with heavy, dense materials
    complicates the attenuation problem. Mixing of isotopic batches, mixing
    with radioactive materials other than SNM, or lumps of SNM can also add
    to the complexity of the response.
    3.3.4Fission Measurements
    Scrap or waste having low-atomic-number materials will reduce the
    energy of the neutrons present in the container, which will
    significantly affect the probability of stimulating fission reactions.
    Neutron-absorbing materials present in SNM scrap or waste may
    significantly affect the operation of NDA systems. Table 1 identifies
    neutron absorbers in the order of decreasing probability of absorption
    of thermal neutrons. An estimate of the significance of the presence of
    one of these materials may be obtained from the ratio of its absorption
    cross section to the absorption cross section of the SNM present in the
    container:
    (Due to database constraints, this equation is not included. Please
    contact LIS to obtain a copy.)(Due to database constraints, Tables 1-4 are not included. Please
    contact LIS to obtain a copy.) The magnitude of this effect is dependent on the distribution of
    the materials and the energy of the neutrons present within the
    container. The relationship above is a gross approximation. For
    convenience in calculation, including only the primary fissile isotope
    is sufficient to determine which materials may constitute a problem
    requiring separate categorization for assay. In extreme cases, it will
    be necessary either to seek methods for measuring the content of the
    neutron absorber to provide a correction for the NDA response or to seek
    a different method for assay of that category.
    3.4 Packaging for NDA
    NDA provides optimal accuracy when the packages to be assayed are
    essentially identical and when the calibration standards represent those
    packages in content and form. Containers for most scrap and waste can
    be loaded using procedures that will enhance the uniformity of the
    loading within each container and from container to container. For
    further discussion and recommendations on container standardization, see
    Reference 16.
    3.5 Calibration of NDA Systems for Scrap and Waste
    To obtain an assay value on SNM in a container of scrap or waste
    with an associated standard error, the observed NDA response or the
    predicted content must be corrected for background and for significant
    effects attributable to the factors described in the preceding parts of
    this discussion. Several approaches are available to correct an assay
    for effects that significantly perturb the assay result. The first
    approach is to use a separate calibration for each material category
    that results in a different assay response. The second approach is to
    make auxiliary measurements as part of the assay. The assay is then
    corrected according to a procedure developed for interpreting each
    auxiliary measurement. A third possible calibration technique is one in
    which a random number of containers are assayed (by the NDA method to be
    used) a sufficient number of times (to minimize random error) and then
    destructively measured (in such a way that the entire container contents
    are measured). A calibration curve depicting the relationship between
    destructive assay values and NDA response can then be derived. This
    approach may give rise to relatively large errors for individual items,
    but it can minimize the error associated with the total SNM quantity
    measured by the particular NDA method. This calibration procedure can
    also be used to confirm a calibration curve derived from calibration
    standards.
    Each approach has its advantages and limitations. Separate
    calibrations are appropriate when (1) the perturbing effects are well
    characterized for each category, (2) there are relatively few
    categories, and (3) the instrument design will not allow collection of
    data suitable for making corrections. A calibration with auxiliary
    measurements for correction factors is appropriate when (1) the
    perturbing effects are variable within a material category, (2) the
    various categories are not reliably segregated, and (3) the measurement
    method facilitates the use of suitable auxiliary measurements.
    Calibration by comparison of NDA and destructive analyses on randomly
    selected actual samples may be useful in cases when well-characterized
    standards are not available or are not practical for the measurements
    involved. However, in view of the potential for greater errors with
    this calibration method, measurements based on this technique should be
    regarded as verifications rather than as careful quantitative assays.
    The relative difficulty in implementing one calibration scheme
    over the other depends on the type of facility and available personnel.
    A steady operation with perhaps some initial set-up assistance might
    favor the correction factor approach because only one calibration is
    used. Often additional material categories can be assayed without
    preparing additional calibration standards. The separate calibration
    scheme might be favored by facilities that have well-characterized
    categories. A separate calibration is made for each category without
    the need for establishing relationships among the categories.
    The calibration of radiometric NDA systems is the subject of
    Regulatory Guide 5.53, "Qualification, Calibration, and Error Estimation
    Methods for Nondestructive Assay," which endorses ANSI N15.20-1975,
    "Guide to Calibrating Nondestructive Assay Systems."(3)C. REGULATORY POSITION
    In the development of an acceptable framework for the
    incorporation of NDA for the measurement of SNM-bearing scrap and waste,
    strong consideration should be given to technique selection,
    calibration, and operational procedures; to the segregation of scrap and
    waste categories; and to the selection and packaging of containers. The
    guidelines presented below are generally acceptable to the NRC staff for
    use in developing such a framework that can serve to improve materials
    accountability.
1. ORIGIN OF SCRAP AND WASTE
    The origin of scrap and waste generated in conjunction with SNM
    processing activities should be determined as follows:
    a. Identify those operations that generate SNM-bearing scrap or
    waste as a normal adjunct of a process.
    b. Identify those operations that occasionally generate
    SNM-bearing scrap or waste as the result of an abnormal operation that
    renders the product unacceptable for further processing or use without
    treatment.
    c. Identify those scrap and waste items generated in
    conjunction with equipment cleanup, maintenance, or replacement.
    ----------
    (3) Copies may be obtained from the American National Standards
    Institute, 1430 Broadway, New York, New York 10018.
    ----------
    The quantities of scrap and waste generated during normal
    operations in each category in terms of the total volume and SNM content
    should be estimated. Bulk measurement throughput requirements should be
    determined to ensure that such assay will not constitute an operational
    bottleneck.
2. NDA SELECTION
    2.1 Technique
    The performance objectives for the NDA system should be such that,
    when the uncertainty corresponding to the scrap and waste material
    balance component is combined with the uncertainties corresponding to
    the other material components, the quality constraints on the total
    standard error of the inventory difference will be satisfied.
    Techniques should be considered for implementation in the order of
    precedence established in Table 2 of this guide. Often, techniques
    within a given instrument category in Table 2 will have different
    accuracies, lower-limit sensitivities, costs, availabilities, and sizes.
    Selection should be based on attainable accuracy with due consideration
    of the characteristics of the scrap and waste categories as well as
    cost, availability, and size.
    2.2 System Specifications
    NDA systems for SNM accountability should be designed and
    shielding should be provided to meet the following objectives:
    a. Performance characteristics should be essentially
    independent of fluctuations in the ambient operational environment,
    including:
    (1) External background radiations,
    (2) Temperature,
    (3) Humidity, and
    (4) Electric power.
    b. Response should be essentially independent of positioning of
    SNM within the scrap or waste container, including effects attributable
    to:
    (1) Detector geometrical efficiency and
    (2) Stimulating source intensity and energy.
    Techniques to achieve these objectives are discussed in Section B
    of this guide.
3. CATEGORIZATION AND SEGREGATION
    Scrap and waste categories should be developed on the basis of NDA
    interference control, recovery or disposal compatibility (Ref. 3), and
    relevant safety considerations. Categorization for NDA interference
    control should be directed to limiting the range of variability in an
    interference. Items to be considered depend on the sensitivity of the
    specific NDA technique, as shown in Table 3.
    The means through which these interferences are manifested are
    detailed in Section B. When such effects or contents are noted,
    separate categories should be established to isolate the materials.
4. CONTAINERS
    4.1 Size Constraints
    Scrap and waste should be packaged for assay in containers as
    small as practicable consistent with the capability and sensitivity of
    the NDA system. Discussion of container standardization and
    recommendations for NDA measurements can be found in Reference 16.
    To enhance the penetration of stimulating or emitted radiations,
    containers should be cylindrical. If possible, the diameter should be
    less than 5 inches (12.7 cm) to provide for significant loading
    capability, ease in loading, reasonable penetrability characteristics,
    and where applicable, compatibility with criticality-safe geometry
    requirements for individual containers.
    Containers having an outside diameter of 4-3/8 inches (11.1 cm)
    will permit 19 such containers to be arranged in a cross section of a
    55-gallon drum, even when that drum contains a plastic liner.
    Containers having an overall length equal to some integral fraction of
    the length of a 55-gallon drum are further recommended when shipment or
    storage within such containers is to be considered. For normal
    operations, an overall length of either 16-1/2 inches (41.9 cm) (two
    layers or 38 containers per drum) or 11 inches (27.9 cm) (three layers
    or 57 containers per drum) is recommended.
    Certain objectives may be inconsistent with the above size
    recommendations, such as the objective to limit handling, reduce cost,
    and keep waste volume to a minimum. It may therefore be necessary to
    package scrap and waste materials in containers of sizes that exceed
    these recommendations, and this may result in a significant impairment
    in the accuracy of NDA techniques on such samples. The relative merits
    of various NDA techniques with samples of different sizes are addressed
    in Table 2. With small containers (about 2 liters), an accuracy of 2 to
    5 percent is routinely obtainable; with a 55-gallon drum a lower
    accuracy of 15 to 30 percent is to be expected. In cases of uniformly
    mixed well-characterized material, a better accuracy may be possible.
    On the other hand, certain combinations of adverse circumstances can
    lead to a considerably worse accuracy. The potential for an adverse
    measurement situation is greater with a larger container than with a
    smaller container, and the consequences of that situation can lead to a
    greater error with larger containers. Conditions leading to measurement
    errors are discussed in Section B.2, and they are listed as
    interferences in the column headings of Table 3.
    If unusual container sizes are necessary, it is often useful to
    employ a second measurement method in a comparative analysis to obtain a
    comparison of results. The other measurement method should be more
    accurate and one that is not sensitive to the interferences affecting
    the first measurement method. For example, if the first measurement is
    one that measures neutrons and is affected by the amount of
    low-atomic-weight moderating material present (which is difficult to
    duplicate in the standards), the second method should be one insensitive
    to the amount of moderator present. Or, if uncertainty in the
    calibration of the first method is due to geometry effects, the second
    method should be one that is insensitive to those effects, e.g., through
    subdivision of the containers. Complete ashing, dissolution, sampling,
    and chemical and mass spectrometric analysis of waste containers
    constitutes a useful second measurement method in some cases.
    The second, more accurate measurement method should be traceable
    to national standards(4) and should be employed to verify the
    calibration relationship of the primary method. Process items should be
    selected at random from the population of items being measured. A
    sufficient number of items analyzed by the first method should be
    selected to ensure, as a minimum, that a stable estimate of the
    population variance is obtained. If simple linear regression is
    applicable, the minimum number of items selected per material balance
    period should be 17 in order to provide 15 degrees of freedom for the
    standard error of estimate and test for a proportional bias (Ref. 25).
    If a second NDA method is employed for comparative analysis, the
    container size for the second method analyses should be consistent with
    the recommendations in this guide.
    4.2 Structural Features
    Containers should be selected in accordance with normal safety
    considerations and should be:
    a. Structurally identical for all samples to be assayed within
    each category,
    b. Structurally identical for as many categories as practicable
    to facilitate loading into larger containers or storage facilities,
    c. Uniform in wall thickness and material composition,
    d. Fabricated of materials that do not significantly interfere
    with the radiations entering or leaving the sample,
    e. Capable of being sealed to verify postassay integrity, and
    ----------
    (4) See Regulatory Guide 5.58, "Considerations for Establishing
    Traceability of Special Nuclear Material Accounting Measurements."
    ----------
    f. Compatible with subsequent recovery, storage, and disposal
    requirements, as applicable.
    In most NDA applications, uniformity of composition is more
    important than the specification of a particular material. Table 4
    gives general recommendations in order of preference for container
    structural materials.
    4.3 Container Identification
    To facilitate loading and assay within the segregation categories,
    containers should either be color-coded or carry color-coded
    identification labels. Identification of categories should be
    documented, and operating personnel should be instructed to ensure
    compliance with established segregation objectives.
5. PACKAGING
    Containers, where practicable, should be packaged with a quantity
    of material containing sufficient SNM to ensure that the measurement is
    not being made at the extremes of the performance bounds for that
    system. Packaging procedures should be consistent with relevant safety
    practices.
    Containers should be packaged in as reproducible a manner as
    possible, with special attention to the maintenance of uniform fill
    heights. Low-density items should be compacted to reduce bulk volume
    and to increase the container SNM loading. Lowering the bulk volume
    reduces the number of containers to be assayed and generally improves
    the assay precision.
    The sample containers should be loaded with SNM as uniformly as
    possible. If significant variability in the distribution of container
    contents is suspected, rotating or scanning the container during assay
    will aid in improving the accuracy of many NDA methods. An example of
    this approach is described in Reference 26.
6. CALIBRATION
    The calibration should be verified for each material category.
    Within each category, the variation of interference effects should be
    measured within the boundaries defining the limits of that category.
    Calibration standards should employ containers identical to those to be
    employed for the scrap or waste. Their contents should be mocked up to
    represent the range of variations in the interferences to be
    encountered. To minimize the number of standards required, the
    calibration standards should permit the range of interference variations
    to be simulated over a range of SNM loadings.
    Verification of the calibration should be made at the start of
    each assay section. If different calibrations are to be used, each
    calibration should be independently verified with material appropriate
    for that calibration. A record should be kept of the verification
    measurements for quality assurance and to identify long-term instrument
    drifts. Verification measurements should be used to periodically update
    the calibration data when the comparison with predicted quantities is
    satisfactory. Calibration of the system is not acceptable when the NDA
    predicted value does not agree with the measured value to within the
    value of the combined standard error.
    Calibration data and hypotheses should be reinvestigated when this
    criterion is not satisfied. For a detailed discussion of calibration
    and measurement control procedures, see Regulatory Guide 5.53.
    Assay values should be periodically checked through an independent
    measurement using a technique sufficiently accurate to resolve the assay
    uncertainty. Periodically, a container of scrap or waste should be
    randomly selected for verification. Once selected, the NDA analysis
    should be repeated a minimum of five times to determine the precision
    characteristics of the system. The contents of that container should
    then be independently measured using a technique sufficiently accurate
    to check the NDA.
    REFERENCES
1. F. A. O'Hare et al., "Calorimetry for Safeguards Purposes," Mound
    Facility, Miamisburg, Ohio, MLM-1798, January 1972.
2. R. Sher and S. Untermeyer, The Detection of Fissionable Material
    by Nondestructive Means, American Nuclear Society Monograph, 1980,
    and references cited therein; also, C. T. Roche et al., "A
    Portable Calorimeter System for Nondestructive Assay of
    Mixed-Oxide Fuels," in Nuclear Safeguards Analysis, E. A. Hakkila,
    ed., ACS Symposium No. 79, p. 158, 1978, and references cited
    therein.
3. U.S. Nuclear Regulatory Commission, "Calorimetric Assay for
    Plutonium," NUREG-0228, 1977.
4. R. H. Augustson and T. D. Reilly, "Fundamentals of Passive
    Nondestructive Assay of Fissionable Material," Los Alamos
    Scientific Laboratory, LA-5651-M, 1974.
5. R. Gunnink et al., "A Re-evaluation of the Gamma-Ray Energies and
    Absolute Branching Intensities of (237)U, (238,239,240,241)Pu, and
    (241)Am," Lawrence Livermore Laboratories, UCRL-52139, 1976.
6. J. E. Cline, R. J. Gehrke, and L. D. McIsaac, "Gamma Rays Emitted
    by the Fissionable Nuclides and Associated Isotopes," Aerojet
    Nuclear Co., Idaho Falls, Idaho, ANCR-1069, July 1972.
7. L. A. Kull, "Catalogue of Nuclear Material Safeguards
    Instruments," Battelle National Laboratories, BNL-17165, August
    1972.
8. J. R. Beyster and L. A. Kull, "Safeguards Applications for
    Isotopic Neutrons Sources," Battelle National Laboratories,
    BNL-50267 (T-596), June 1970.
9. T. W. Crane, "Measurement of Uranium and Plutonium in Solid Waste
    by Passive Photon or Neutron Counting and Isotopic Neutron Source
    Interrogation," Los Alamos Scientific Laboratory, LA-8294-MS,
    1980.
10. T. Gozani, "Active Nondestructive Assay of Nuclear Materials,"
    Nuclear Regulatory Commission, NUREG/CR-0602, 1981.
    11. H. P. Filss, "Direct Determination of the Total Fissile Content in
    Irradiated Fuel Elements, Water Containers and Other Samples of
    the Nuclear Fuel Cycle," Nuclear Materials Management, Vol. VIII,
    pp. 74-79, 1979.
    12. H. O. Menlove and T. W. Crane, "A (252)Cf Based Nondestructive
    Assay System for Fissile Material," Nuclear Instruments and
    Methods, Vol. 152, pp. 549-557, 1978.
    13. T. W. Crane, "Test and Evaluation Results of the (252)Cf Shuffler
    at the Savannah River Plant," Los Alamos National Laboratory,
    LA-8755-MS, March 1981.
    14. T. W. Crane, "Measurement of Pu Contamination at the 10-nCi/g
    Level in 55-Gallon Barrels of Solid Waste with a (252)Cf Assay
    System," Proceedings of the International Meeting of
    Pu-Contamination, Ispra, Italy, J. Ley, Ed., JRC-1, pp. 217-226,
    September 25-28, 1979.
    15. D. Langner et al., "The CMB-8 Material Balance System," Los Alamos
    Scientific Laboratory, LA-8194-M, pp. 4-14, 1980.
    16. K. R. Alvar et al., "Standard Containers for SNM Storage,
    Transfer, and Measurement," Nuclear Regulatory Commission,
    NUREG/CR-1847, 1980.
    17. R. Sher, "Operating Characteristics of Neutron Well Coincidence
    Counters," Battelle National Laboratories, BNL-50332, January
    1972.
    18. N. Ensslin et al., "Neutron Coincidence Counters for Plutonium
    Measurements," Nuclear Materials Management, Vol. VII, No. 2, p.
    43, 1978.
    19. M. S. Krick and H. O. Menlove, "The High-Level Neutron Coincidence
    Counter (HLNCC): Users' Manual," Los Alamos Scientific
    Laboratory, LA-7779-MS (ISPO-53), 1979.
    20. R. B. Perry, R. W. Brandenburg, N. S. Beyer, "The Effect of
    Induced Fission on Plutonium Assay with a Neutron Coincidence Well
    Counter," Transactions of the American Nuclear Society, Vol. 15,
    p. 674, 1972.
    21. N. Ensslin, J. Stewart, and J. Sapir, "Self-Multiplication
    Correction Factors for Neutron Coincidence Counting," Nuclear
    Materials Management, Vol. VIII, No. 2, p. 60, 1979.
    22. J. L. Parker and T. D. Reilly, "Bulk Sample Self-Attenuation
    Correction by Transmission Measurement," Proceedings of the ERDA
    X- and Gamma-Ray Symposium, Ann Arbor, Michigan, Conf. 760639, p.
    219, May 1976.
    23. N. Ensslin et al., "Description and Operating Manual for the Fast
    Neutron Coincidence Counter," Los Alamos National Laboratory,
    LA-8858-M, 1982.
    24. "Reactor Physics Constants," Argonne National Laboratories,
    ANL-5800, pp. 30-31, 1963.
    25. U.S. Nuclear Regulatory Commission, "Methods of Determining and
    Controlling Bias in Nuclear Materials Accounting Measurements,"
    NUREG/CR-1284, 1980.
    26. E. R. Martin, D. F. Jones, and J. L. Parker, "Gamma-Ray
    Measurements with the Segmented Gamma Scan," Los Alamos Scientific
    Laboratory, LA-7059-M, 1977.
    SUGGESTED READING
    American National Standards Institute and American Society for Testing
    and Materials, "Standard Test Methods for Nondestructive Assay of
    Special Nuclear Materials Contained in Scrap and Waste," ANSI/ASTM C
    853-79.
    This document provides further details on procedures for assaying
    scrap and waste.
D. R. Rogers, "Handbook of Nuclear Safeguards Measurement Methods,"
    Nuclear Regulatory Commission, NUREG/CR-2078, 1983.
    This book provides extensive procedures, with references, for
    assaying scrap and waste.
    VALUE/IMPACT STATEMENT
1. PROPOSED ACTION
    1.1 Description
    Licensees authorized to posses at any one time more than one
    effective kilogram of special nuclear material (SNM) are required in
    paragraph 70.58(f) of 10 CFR Part 70 to establish and maintain a system
    of control and accountability to ensure that the standard error of any
    inventory difference (ID) ascertained as a result of a measured material
    balance meets established minimum standards. The selection and proper
    application of an adequate measurement method for each of the material
    forms in the fuel cycle are essential for the maintenance of these
    standards.
    For some material categories, particularly scrap and waste,
    nondestructive assay (NDA) is the only practical, and sometimes the most
    accurate, means for measuring SNM content. This guide details
    procedures acceptable to the NRC staff to provide a framework for the
    use of NDA in the measurement of scrap and waste components generated in
    conjunction with the processing of SNM.
    The proposed action is to revise Regulatory Guide 5.11, originally
    issued in October 1973, which is still basically sound.
    1.2 Need for Proposed Action
    Regulatory Guide 5.11 was published in 1973. The proposed action
    is needed to bring the guide up to date with respect to advances in
    measurement methods as well as changes in terminology.
    1.3 Value/Impact of Proposed Action
    1.3.1NRC Operations
    The experience and improvements in technology that have occurred
    since the guide was issued will be made available for the regulatory
    procedure. Using these updated techniques should have no adverse
    impact.
    1.3.2Other Government Agencies
    Not applicable.
    1.3.3Industry
    Since industry is already applying the methods and procedures
    discussed in the guide, updating the guide should have no adverse
    impact.
    1.3.4Public
    No impact on the public can be foreseen.
    1.4 Decision on Proposed Action
    The guide should be revised.
2. TECHNICAL APPROACH
    Not applicable.
3. PROCEDURAL APPROACH
    3.1 Procedural Alternatives
    Of the alternative procedures considered, revision of the existing
    regulatory guide was selected as the most advantageous and cost
    effective.
4. STATUTORY CONSIDERATIONS
    4.1 NRC Authority
    Authority for the proposed action is derived from the Atomic
    Energy Act of 1954, as amended, and the Energy Reorganization Act of
    1974, as amended, and implemented through the Commission's regulations.
    4.2 Need for NEPA Assessment
    The proposed action is not a major action that may significantly
    affect the quality of the human environment and does not require an
    environmental impact statement.
5. RELATIONSHIP TO OTHER EXISTING OR PROPOSED REGULATIONS OR POLICIES
    The proposed action is one of a series of revisions of existing
    regulatory guides on nondestructive assay techniques.
6. SUMMARY AND CONCLUSION
    Regulatory Guide 5.11 should be revised to bring it up to date.
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