1.1 This standard presents a matrix to identify existing and potentially needed standards for light water reactor (LWR) fuel reprocessing. 1.2 This matrix pertains to facilities for the reprocessing of LWR spent fuel including its dissolution and separation of the reusable nuclear materials from the waste byproducts and conversion of these products and byproducts to suitable forms for shipment off-site. 1.3 The matrix is defined as an array of fuel reprocessing systems and components as the horizontal axis and the functional activities as the vertical axis. The matrix also has multiple overlays for generic issues. This might also be considered as a third orthogonal axis. 1.4 The terms used for the systems and components, functional, and overlay activities apply specifically to this matrix and are not intended to be universal. See Section 3. 1.5 Matrix Standards on Decommissioning of Nuclear Facilities (in preparation), Fuel Fabrication (in preparation), LWR Spent Fuel Receiving and Storage, and Nuclear Safeguards deal in detail with their respective subjects. Therefore, this standard will only refer to these standards and not develop the subjects. 1.6 This standard will be developed in two steps: (1) identifying a matrix of systems and component/functional/ overlay intersection where standards exist or are potentially needed and (2) completing the matrix by listing the existing standards and those potentially needed, assigning priorities to those needed and identifying potential secretariats. 1.7 While it is recognized that federal directives and guidelines are not national consensus standards, they may be included.

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4.1 Property data obtained with the recommended test methods identified herein may be used for research and development, design, manufacturing control, specifications, performance evaluation, and regulatory statutes pertaining to nuclear reactors that utilize graphite.4.2 The referenced test methods are applicable primarily to specimens in the non-irradiated and non-oxidized state. Testing irradiated specimens often requires specimen geometries that do not meet the requirements of the standard. Specific instructions or recommendations with respect to testing non-conforming geometries can be found in STP 15784 and/or Guide D7775. When testing irradiated specimens at elevated temperatures, the effects of annealing should be considered (see Note 1).NOTE 1: Exposure to fast neutron radiation will result in atomic and microstructural changes to graphite. This radiation damage occurs when energetic particles, such as fast neutrons, impinge on the crystal lattice and displace carbon atoms from their equilibrium positions, creating a lattice vacancy and an interstitial carbon atom. The lattice strain that results from displacement damage causes significant structural and property changes in the graphite and is a function of the irradiation temperature and dose. When the temperature of the graphite is brought above the temperature at which it was irradiated, enough energy is provided that the structure of the graphite will anneal back to its original condition. Therefore, measurement techniques that bring the specimen temperature above the irradiation temperature can result in property values that change during the measurement process. For this reason, measurements made on irradiated test specimens below the irradiation temperature will produce results that are representative of the irradiation damage. However, measurements made at temperatures above the irradiation temperature could include the effects of annealing.4.3 Additional test methods are in preparation and will be incorporated. The user is cautioned to employ the latest revision.1.1 This practice covers the application and limitations of test methods for measuring the properties of graphite materials. These properties may be used for the design and evaluation of gas-cooled reactor components.1.2 The test methods referenced herein are applicable to materials used for replaceable and permanent components as defined in Section 7 and includes fuel elements; removable reflector elements and blocks; permanent side reflector elements and blocks; core support pedestals and elements; control rod, reserve shutdown, and burnable poison compacts; and neutron shield material. Specific aspects with respect to testing of irradiated materials are addressed.1.3 This practice includes test methods that have been selected from ASTM standards and guides that are specific to the testing of materials listed in 1.2. Comments on individual test methods for graphite components are given in Section 8. The test methods are summarized in Table 1.1.4 The values stated in SI units are to be regarded as standard. The values given in parentheses after SI units are provided for information only and are not considered standard.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, health, and environmental practices and determine the applicability of regulatory limitations prior to use.1.6 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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3.1 Adjustment methods provide a means for combining the results of neutron transport calculations with neutron dosimetry measurements (see Test Method E1005 and NUREG/CR-5049) in order to obtain optimal estimates for neutron damage exposure parameters with assigned uncertainties. The inclusion of measurements reduces the uncertainties for these parameter values and provides a test for the consistency between measurements and calculations and between different measurements (see 3.3.3). This does not, however, imply that the standards for measurements and calculations of the input data can be lowered; the results of any adjustment procedure can be only as reliable as are the input data.3.2 Input Data and Definitions: 3.2.1 The symbols introduced in this section will be used throughout the guide.3.2.2 Dosimetry measurements are given as a set of reaction rates (or equivalent) denoted by the following symbols:These data are, at present, obtained primarily from radiometric dosimeters, but other types of sensors may be included (see 4.1).3.2.3 The neutron spectrum (see Terminology E170) at the dosimeter location, fluence or fluence rate Φ(E) as a function of neutron energy E , is obtained by appropriate neutronics calculations (neutron transport using the methods of discrete ordinates or Monte Carlo, see Guide E482). The results of the calculation are customarily given in the form of multigroup fluences or fluence rates.where:Ej and Ej+1 are the lower and upper bounds for the j-th energy group, respectively, and k is the total number of groups.3.2.4 The reaction cross sections of the dosimetry sensors are obtained from an evaluated cross section file. The cross section for the i-th reaction as a function of energy E will be denoted by the following:Used in connection with the group fluences, Eq 2, are the calculated group-averaged cross sections σij. These values are defined through the following equation:3.2.5 Uncertainty information in the form of variances and covariances must be provided for all input data. Appropriate corrections must be made if the uncertainties are due to bias producing effects (for example, effects of photo reactions).3.3 Summary of the Procedures: 3.3.1 An adjustment algorithm modifies the set of input data as defined in 3.2 in the following manner (adjusted quantities are indicated by a tilde, for example, ãi):or for group fluence ratesor for group-averaged cross sectionsThe adjusted quantities must satisfy the following conditions:or in the form of group fluence ratesSince the number of equations in Eq 11 is much smaller than the number of adjustments, there exists no unique solution to the problem unless it is further restricted. The mathematical algorithms in current adjustment codes are intended to make the adjustments as small as possible relative to the uncertainties of the corresponding input data. Codes like STAY'SL, FERRET, LEPRICON, and LSL-M2 (see Table 1) are based explicitly on the statistical principles such as “Maximum Likelihood Principle” or “Bayes Theorem,” which are generalizations of the well-known least squares principle, and are taking into account variances and correlations of the input fluence, dosimetry, and cross section data (see 4.1.1, 4.2.2, and 4.3.3). A detailed discussion of the mathematical derivations can be found in NUREG/CR-2222 and EPRI NP-2188. Even the older codes, notably SAND-II and CRYSTAL BALL, apply a minimization algorithm although the statistical assumptions are not spelled out explicitly in the supporting documentation. Table 1 lists some of the available unfolding codes; however, the first four codes listed: SAND-II, SPECTRA, IUNFLD/UNFOLD, and WINDOWS have severe limitations in that they do not typically provide uncertainty characterization of the resulting unfolded spectrum and the adjusted damage exposure parameters.(A) The boldface numbers in parentheses refer to the list of references appended to this guide.3.3.1.1 An important problem in reactor surveillance is the determination of neutron fluence inside the pressure vessel wall at locations which are not accessible to dosimetry. Estimates for exposure parameter values at these locations can be obtained from adjustment codes which adjust fluences simultaneously at more than one location when the cross correlations between fluences at different locations are given. LEPRICON has provisions for the estimation of cross correlations for fluences and simultaneous adjustment. LSL-M2 also allows simultaneous adjustment, but cross correlations must be given.3.3.2 The adjusted data ãi, etc., are, for any specific algorithm, unique functions of the input variables. Thus, uncertainties (variances and covariances) for the adjusted parameters can, in principle, be calculated by propagation the uncertainties for the input data. Linearization may be used before calculating the uncertainties of the output data if the adjusted data are nonlinear functions of the input data.3.3.2.1 The algorithms of the adjustment codes tend to decrease the variances of the adjusted data compared to the corresponding input values. The linear least squares adjustment codes yield estimates for the output data with minimum variances, that is, the “best” unbiased estimates. This is the primary reason for using these adjustment procedures.3.3.3 Properly designed adjustment methods provide means to detect inconsistencies in the input data which manifest themselves through adjustments that are larger than the corresponding uncertainties or through large values of chi-square, or both. (See NUREG/CR-3318 and NUREG/CR-3319.) Any detection of inconsistencies should be documented, and output data obtained from inconsistent input should not be used. All input data should be carefully reviewed whenever inconsistencies are found, and efforts should be made to resolve the inconsistencies as stated below.3.3.3.1 Input data should be carefully investigated for evidence of gross errors or biases if large adjustments are required. Note that the erroneous data may not be the ones that required the largest adjustment; thus, it is necessary to review all input data. Data of dubious validity may be eliminated if proper corrections cannot be determined. Any elimination of data must be documented and reasons stated which are independent of the adjustment procedure. Inconsistent data may also be omitted if they contribute little to the output under investigation.3.3.3.2 Inconsistencies may also be caused by input variances which are too small. The assignment of uncertainties to the input data should, therefore, be reviewed to determine whether the assumed precision and bias for the experimental and calculational data may be unrealistic. If so, variances may be increased, but reasons for doing so should be documented. Note that in statistically based adjustment methods, listed in Table 1 the output uncertainties are determined only by the input uncertainties and are not affected by inconsistencies in the input data (see NUREG/CR-2222). Note also that too large adjustments may yield unreliable data because the limits of the linearization are exceeded even if these adjustments are consistent with the input uncertainties.3.3.4 Using the adjusted fluence spectrum, estimates of damage exposure parameter values can be calculated. These parameters are weighted integrals over the neutron fluenceor for group fluenceswith given weight (response) functions w(E) or w j, respectively. The response function for dpa of iron is listed in Practice E693. Fluence greater than 1.0 MeV or fluence greater than 0.1 MeV is represented as w(E) = 1 for E above the limit and w(E) = 0 for E below.3.3.4.1 Finding best estimates of damage exposure parameters and their uncertainties is the primary objective in the use of adjustment procedures for reactor surveillance. If calculated according to Eq 12 or Eq 13, unbiased minimum variance estimates for the parameter p result, provided the adjusted fluence Φ ˜ is an unbiased minimum variance estimate. The variance of p can be calculated in a straightforward manner from the variances and covariances of the adjusted fluence spectrum. Uncertainties of the response functions, wj, if any, should not be considered in the calculation of the output variances when a standard response function, such as the dpa for iron in Practice E693, is used. The calculation of damage exposure parameters and their variances should ideally be part of the adjustment code.1.1 This guide covers the analysis and interpretation of the physics dosimetry for Light Water Reactor (LWR) surveillance programs. The main purpose is the application of adjustment methods to determine best estimates of neutron damage exposure parameters and their uncertainties.1.2 This guide is also applicable to irradiation damage studies in research reactors.1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.1.4 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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3.1 Practices E185 and E2215 describe a minimum program for the surveillance of reactor vessel materials, specifically mechanical property changes that occur in service. This guide may be applied to generate additional information on radiation-induced property changes to better assist the determination of the optimum reactor vessel operation schemes.1.1 This guide discusses test procedures that can be used in conjunction with, but not as alternatives to, those required by Practices E185 and E2215 for the surveillance of nuclear reactor vessels. The supplemental mechanical property tests outlined permit the acquisition of additional information on radiation-induced changes in mechanical properties of the reactor vessel steels.1.2 This guide provides recommendations for the preparation of test specimens for irradiation, and identifies special precautions and requirements for reactor surveillance operations and post-irradiation test planning. Guidance on data reduction and computational procedures is also given. Reference is made to other ASTM test methods for the physical conduct of specimen tests and for raw data acquisition.1.3 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.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.

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3.1 Prediction of neutron radiation effects to pressure vessel steels has long been a part of the design and operation of light water reactor power plants. Both the federal regulatory agencies (see 2.3) and national standards groups (see 2.1 and 2.2) have promulgated regulations and standards to ensure safe operation of these vessels. The support structures for pressurized water reactor vessels may also be subject to similar neutron radiation effects (1, 3-6).2 The objective of this practice is to provide guidelines for determining the neutron radiation exposures experienced by individual vessel supports.3.2 It is known that high-energy photons can also produce displacement damage effects that may be similar to those produced by neutrons. These effects are known to be much less at the belt line of a light water reactor pressure vessel than those induced by neutrons. The same has not been proven for all locations within vessel support structures. Therefore, it may be prudent to apply coupled neutron-photon transport methods and photon-induced displacement cross sections to determine whether gamma-induced dpa exceeds the screening level of 3.0 × 10–4 used in this practice for neutron exposures. (See 1.3.)1.1 This practice covers procedures for monitoring the neutron radiation exposures experienced by ferritic materials in nuclear reactor vessel support structures located in the vicinity of the active core. This practice includes guidelines for:1.1.1 Selecting appropriate dosimetric sensor sets and their proper installation in reactor cavities.1.1.2 Making appropriate neutronics calculations to predict neutron radiation exposures.1.2 The values stated in SI units are to be regarded as standard; units that are not SI can be found in Terminology E170 and are to be regarded as standard. Any values in parentheses are for information only.1.3 This practice is applicable to all pressurized water reactors whose vessel supports will experience a lifetime neutron fluence (E > 1 MeV) that exceeds 1 × 1017 neutrons/cm2 or exceeds 3.0 × 10−4 dpa (1).2 (See Terminology E170.)1.4 Exposure of vessel support structures by gamma radiation is not included in the scope of this practice, but see the brief discussion of this issue in 3.2.1.5 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, health, and environmental practices and determine the applicability of regulatory limitations prior to use. (For example, (2).)1.6 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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5.1 Bacteria that exist in a biofilm are phenotypically different from suspended cells of the same genotype. The study of biofilm in the laboratory requires protocols that account for this difference. Laboratory biofilms are engineered in growth reactors designed to produce a specific biofilm type. Altering system parameters will correspondingly result in a change in the biofilm. The purpose of this method is to direct a user in the laboratory study of biofilms by clearly defining each system parameter. This method will enable a person to grow, sample, and analyze a laboratory biofilm. The method was originally developed to study toilet bowl biofilms, but may also be utilized for research that requires a biofilm grown under moderate fluid shear.1.1 This test method is used for growing a reproducible (1)2 Pseudomonas aeruginosa biofilm in a continuously stirred tank reactor (CSTR) under medium shear conditions. In addition, the test method describes how to sample and analyze biofilm for viable cells.1.2 Although this test method was created to mimic conditions within a toilet bowl, it can be adapted for the growth and characterization of varying species of biofilm (rotating disk reactor—repeatability and relevance (2)).1.3 This test method describes how to sample and analyze biofilm for viable cells. Biofilm population density is recorded as log10 colony forming units per surface area (rotating disk reactor—efficacy test method (3)).1.4 Basic microbiology training is required to perform this test method.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, health, and environmental practices and determine the applicability of regulatory limitations prior to use.1.7 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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5.1 High-purity germanium detectors are used for precise gamma-ray spectroscopy for the purpose of determining radioactivity in materials. Typical applications include monitoring, mapping, and characterization of neutron energy spectra in nuclear reactors or isotopic fission sources.1.1 This standard establishes techniques for calibration, usage, and performance testing of germanium detectors for the measurement of gamma-ray emission rates of radionuclides in radiation metrology for reactor dosimetry. The practice is applicable only to samples of small size, approximating to point sources. It covers the energy and full-energy peak efficiency calibration as well as the determination of gamma-ray energies in the 0.06 MeV to 2 MeV energy region and is designed to yield gamma-ray emission rates with an uncertainty of ±3 % (see Note 1). This technique applies to measurements that do not involve overlapping peaks, and in which peak-to-continuum considerations are not important.NOTE 1: Uncertainty U is given at the 68 % confidence level; that is,where δi are the estimated maximum systematic uncertainties, and σi are the random uncertainties at the 68 % confidence level. Other techniques of error analysis are in use (1, 2).21.2 Additional information on the setup, calibration, and quality control for radiometric detectors and measurements is given in IEEE/ANSI N42.14 and in Guide C1402 and Practice D7282.1.3 The values stated in SI units are generally to be regarded as standard. The rad is an exception.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 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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This specification covers seamless wrought zirconium-alloy tubes for nuclear reactor fuel cladding application. Two grades of reactor grade zirconium alloys are described. Tubes covered by this specification shall be made from ingots produced by multiple vacuum arc or electron beam melting in furnaces of a type conventionally used for reactive materials. The tubes shall conform to the requirements for chemical composition prescribed. Recrsytallisation annealed tubes shall conform to the requirements for mechanical properties at room temperature prescribed. The tension test shall be conducted. Yield strength and tension properties shall be determined. Burst testing, when specified, shall be performed at room temperature on finished tubing.1.1 This specification covers seamless wrought zirconium-alloy tubes for nuclear fuel cladding application, in the outside diameter (OD) size range of 0.200 in. (5.1 mm) to 0.650 in. (16.5 mm) and wall thickness range of 0.010 in. (0.25 mm) to 0.035 in. (0.89 mm).1.2 Two grades of reactor grade zirconium alloys are described.1.2.1 The present UNS numbers designated for the two grades are given in Table 1.1.3 Unless a single unit is used, for example corrosion mass gain in mg/dm2, the values stated in either inch-pound or SI units are to be regarded separately as standard. The values stated in each system are not exact equivalents; therefore each system must be used independently of the other. SI values cannot be mixed with inch-pound values.1.4 The following precautionary caveat pertains only to the test method portions of this specification: 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 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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