This specification covers finished pellets composed of sintered uranium-plutonium dioxide for fast reactor fuel. Specimens shall be sampled and tested suitably, and shall conform accordingly to chemical (uranium and plutonium content, impurity content, stoichiometry, moisture, gas content, and americium-241 content), nuclear (isotopic content, and equivalent plutonium at a given date), and physical (dimensions, density, grain size an pore morphology, homogeneity, particle size and distribution, integrity, surface cracks, circumferential chips, pellet ends, cleanliness and workmanship, and identification) requirements.1.1 This specification is for finished sintered (uranium-plutonium) dioxide pellets. It applies to (uranium-plutonium) dioxide pellets containing plutonium additions in the range from 10 to 40 weight % and uranium of any 235U content. The isotopic composition of the plutonium component will be as normally produced by in-reactor neutron irradiation of uranium having less than 5 % 235U or by in-reactor neutron irradiation of recycled plutonium mixed with uranium. 1.2 This specification does not include (1) provisions for preventing criticality accidents or (2) requirements for health and safety. Observance of this specification does not relieve the user of the obligation to be aware of and conform to all applicable international, federal, state, and local regulations pertaining to possessing, processing, shipping, or using source or special nuclear material. Examples of U.S. government documents are Code of Federal Regulations Title 10, Part 50 — Domestic Licensing of Production and Utilization Facilities; Title 10, Part 71 — Packaging and Transportation of Radioactive Material; and Title 49, Part 173 — General Requirements for Shipments and Packaging. 1.3 The following safety hazards caveat pertains only to the technical requirements portion, Section 4, 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 and health practices and determine the applicability of regulatory limitations prior to use.

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1.1 This specification covers hot- and cold-finished austenitic and martensitic stainless steel bars, billets, and forgings intended for use in manufacturing core components used at high temperatures in liquid metal cooled nuclear reactors.1.2 The bars, billets, and forgings are intended for machining, welding, hot- and cold-forming operations.1.3 The values stated in either inch-pound units or SI units are to be regarded separately as standard. Within the text, the SI units are shown in brackets. The values stated in each system are not exact equivalents; therefore, each system shall be used independently of the other. Combining values from the two systems may result in nonconformance with the specification.1.4 This specification and the applicable material specifications are expressed in both inch-pound and SI units. However, unless the order specifies the applicable "M" specification designation (SI units), the material shall be furnished in inch-pound units.

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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 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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4.1 Neutron radiation effects are considered in the design of light-water moderated nuclear power reactors. Changes in system operating parameters may be made throughout the service life of the reactor to account for these effects. A surveillance program is used to measure changes in the properties of actual vessel materials due to the irradiation environment. This practice describes the criteria that should be considered in evaluating surveillance program test capsules.4.2 Prior to the first issue date of this standard, the design of surveillance programs and the testing of surveillance capsules were both covered in a single standard, Practice E185. Between its provisional adoption in 1961 and its replacement linked to this standard, Practice E185 was revised many times (1966, 1970, 1973, 1979, 1982, 1993 and 1998). Therefore, capsules from surveillance programs that were designed and implemented under early versions of the standard were often tested after substantial changes to the standard had been adopted. For clarity, the standard practice for surveillance programs has been divided into the new Practice E185 that covers the design of new surveillance programs and this standard practice that covers the testing and evaluation of surveillance capsules. Modifications to the standard test program and supplemental tests are described in Guide E636.4.3 This practice is intended to cover testing and evaluation of all light-water moderated reactor pressure vessel surveillance capsules. The practice is applicable to testing of capsules from surveillance programs designed and implemented under all previous versions of Practice E185.4.4 The radiation-induced changes in the properties of the reactor pressure vessel are generally monitored by measuring the index temperatures, the upper-shelf energy and the tensile properties of specimens from the surveillance program capsules. The significance of these radiation-induced changes is described in Practice E185.4.5 Alternative methods exist for testing surveillance capsule materials. Some supplemental and alternative testing methods are available as indicated in Guide E636. Direct measurement of the fracture toughness is also feasible using the To Reference Temperature method defined in Test Method E1921 or J-integral techniques defined in Test Method E1820. Additionally, hardness testing can be used to supplement standard methods as a means of monitoring the irradiation response of the materials.4.6 Practice E853 describes a methodology that may be used in the analysis and interpretation of neutron dosimetry data and the determination of neutron fluence. Regulators or other sources may describe different methods.4.7 Guide E900 describes a method for predicting the TTS. Regulators or other sources may describe different methods for predicting TTS.4.8 Guide E509 provides direction for development of a procedure for conducting an in-service thermal anneal of a light-water cooled nuclear reactor vessel and demonstrating the effectiveness of the procedure including a post-annealing vessel radiation surveillance program.1.1 This practice covers the evaluation of test specimens and dosimetry from light water moderated nuclear power reactor pressure vessel surveillance capsules.1.2 Additionally, this practice provides guidance on reassessing withdrawal schedule for design life and operation beyond design life.1.3 This practice is one of a series of standard practices that outline the surveillance program required for nuclear reactor pressure vessels. The surveillance program monitors the irradiation-induced changes in the ferritic steels that comprise the beltline of a light-water moderated nuclear reactor pressure vessel.1.4 This practice along with its companion surveillance program practice, Practice E185, is intended for application in monitoring the properties of beltline materials in any light-water moderated nuclear reactor.21.5 Modifications to the standard test program and supplemental tests are described in Guide E636.1.6 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.1.7 This 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 This practice is useful for the determination of the average energy per disintegration of the isotopic mixture found in the reactor-coolant system of a nuclear reactor (1).5 The value is used to calculate a site-specific activity limit for the reactor coolant system, generally identified as: where: K   =   a power reactor site specific constant (usually in the range of 50 to 200). The activity of the reactor coolant system is routinely measured, then compared to the value of Alimiting. If the reactor coolant activity value is less than Alimiting then the 2-h radiation dose, measured at the plant boundary, will not exceed an appropriately small fraction of the Code of Federal Regulations, Title 10, part 100 dose guidelines. It is important to note that the measurement of the reactor coolant system radioactivity is determined at a set frequency by use of gamma spectrometry only. Thus, the radionuclides that go into the calculation of and subsequently Alimiting are only those that are measured using gamma spectrometry. 5.2 In calculating , the energy dissipated by beta particles (negatrons and positrons) and photons from nuclear decay of beta-gamma emitters includes the energy released in the form of extra-nuclear transitions such as X-rays, Auger electrons, and conversion electrons. However, not all radionuclides present in a sample are included in the calculation of . 5.3 Individual nuclear reactor technical specifications vary and each nuclear operator must be aware of limitations affecting plant operation. Typically, iodine radionuclides with half-lives of less than 10 min (except those in equilibrium with the parent) and those radionuclides identified using gamma spectrometry with less than 95 % confidence level are not included in the calculation. However, technical requirements specify that the reported activity must account for at least 95 % of the activity after excluding radioiodines and short-lived radionuclides. There are individual bases for each exclusion. 5.3.1 Radioiodines are typically excluded from the calculation of because United States commercial nuclear reactors are required to operate under a more conservative restriction of 1 μCi (37 kBq) per gram dose equivalent 131I (DEI) in the reactor coolant. 5.3.2 Beta-only-emitting radio isotopes (for example, 90Sr or 63Ni) and alpha emitting radioisotopes (for example, 241Am or 239Pu) which comprise a small fraction of the activity, are not included in the E-bar calculation. These isotopes are not routinely analyzed for in the reactor coolant and, thus, their inclusion in the E-bar calculation is not representative of what is used to assess the 10 CFR 100 dose limits. Tritium, also a beta-only emitter, should not be included in the calculation. Tritium has the largest activity concentration in the reactor coolant system but the lowest beta particle energy. Thus, its dose contribution is always negligible. However, its inclusion in the E-bar calculation would raise the value of Alimiting, yielding a non-conservative value for dose assessment. 5.3.3 Excluding radionuclides with half-lives less than 10 min, except those in equilibrium with the parent, has several bases. 5.3.3.1 The first basis considers the nuclear characteristics of a typical reactor coolant. The radionuclides in a typical reactor coolant have half-lives of less than 4 min or have half-lives greater than 14 min. This natural separation provides a distinct window for choosing a 10-min half-life cutoff. 5.3.3.2 The second consideration is the predictable time delay, approximately 30 min, which occurs between the release of the radioactivity from the reactor coolant to its release to the environment and transport to the site boundary. In this time, the short-lived radionuclides have undergone the decay associated with several half-lives and are no longer considered a significant contributor to . 5.3.3.3 A final practical basis is the difficulty associated with identifying short-lived radionuclides in a sample that requires some significant time, relative to 10 min, to collect, transport, and analyze. 5.3.4 The value of E-bar is usually calculated once every 6 months. However, any time a significant increase in the activity of the reactor coolant occurs, the value of E-bar should be reassessed to ensure compliance with 10 CFR 100. Such reassessment should be done any time there is a significant fuel defect that would alter the value and affect Alimiting. The two possible causes to reassess the value of would be: (1) A significant fuel defect has occurred where the noble gas activity has increased. (2) A significant corrosion product increase has occurred. For the case of a fuel defect, the plant staff may need to include new radionuclides not normally used in the calculation of such as 239U and 239Np. 1.1 This practice applies to the calculation of the average energy per disintegration ( ) for a mixture of radionuclides in reactor coolant water. 1.2 The microcurie (µCi) is the standard unit of measurement for this standard. The values given in parentheses are mathematical conversions to SI units, which are provided for information only and are not considered standard. 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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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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