3.1 This guide describes approaches for using neutron fields with well known characteristics to perform calibrations of neutron sensors, to intercompare different methods of dosimetry, and to corroborate procedures used to derive neutron field information from measurements of neutron sensor response.3.2 This guide discusses only selected standard and reference neutron fields which are appropriate for benchmark testing of light-water reactor dosimetry. The Standard Fields considered here include neutron source environments that closely approximate: a) the unscattered neutron spectra from 252Cf spontaneous fission; and b) the 235U thermal neutron induced fission. These standard fields were chosen for their spectral similarity to the high energy region (E > 2 MeV) of reactor spectra. The various categories of benchmark fields are defined in Terminology E170.3.3 There are other well known neutron fields that have been designed to mockup special environments, such as pressure vessel mockups in which it is possible to make dosimetry measurements inside of the steel volume of the “vessel.” When such mockups are suitably characterized, they are also referred to as benchmark fields. A variety of these engineering benchmark fields have been developed, or pressed into service, to improve the accuracy of neutron dosimetry measurement techniques. These special benchmark experiments are discussed in Guide E2006, and in Refs (1)4 and (2).1.1 This guide covers facilities and procedures for benchmarking neutron measurements and calculations. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to calibrate integral neutron sensors; the use of certified-neutron-fluence standards to calibrate radiometric counting equipment or to determine interlaboratory measurement consistency; development of special benchmark fields to test neutron transport calculations; use of well-known fission spectra to benchmark spectrum-averaged cross sections; and the use of benchmarked data and calculations to determine the uncertainties in derived neutron dosimetry results.1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this 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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4.1 This guide deals with the difficult problem of benchmarking neutron transport calculations carried out to determine fluences for plant specific reactor geometries. The calculations are necessary for fluence determination in locations important for material radiation damage estimation and which are not accessible to measurement. Typically, the most important application of such calculations is the estimation of fluence within the reactor vessel of operating light water reactors (LWR) to provide accurate estimates of the irradiation embrittlement of the base and weld metal in the vessel. The benchmark procedure must not only prove that calculations give reasonable results but that their uncertainties are propagated with due regard to the sensitivities of the different input parameters used in the transport calculations. Benchmarking is achieved by building up data bases of benchmark experiments that have different influences on uncertainty propagation. For example, in simple vessel wall mockups where measurements are made within a simulated reactor vessel wall, the integral effect of uncertainties in iron cross sections (absorption and elastic and inelastic scattering) are dominant and have been bounded by the agreement between calculation and measurement. For more complicated integral benchmarks, other factors such as: uncertainties in the distribution of fission sources, geometry, the energy-dependent cross sections, and the angular scattering distribution for elemental components of major materials in the neutron field (such as water and iron) may all be important uncertainty contributors. This guide describes general procedures for using neutron fields with known characteristics to corroborate the calculational methodology and nuclear data used to derive neutron field information from measurements of neutron sensor response.4.2 The bases for benchmark field referencing are usually irradiations performed in standard neutron fields with well-known energy spectra and intensities. There are, however, less well known neutron fields that have been designed to mockup special environments, such as pressure vessel mockups in which it is possible to make dosimetry measurements inside of the steel volume of the “vessel”. When such mockups are suitably characterized, they are also referred to as benchmark fields. A benchmark is that against which other things are referenced, hence the terminology “to benchmark reference” or “benchmark referencing”. A variety of benchmark neutron fields, other than standard neutron fields, have been developed, or pressed into service, to improve the accuracy of neutron dosimetry measurement techniques. Some of these special benchmark experiments are discussed in this standard because they have identified needs for additional benchmarking or because they have been sufficiently documented to serve as benchmarks.4.3 One dedicated effort to provide benchmarks whose radiation environments closely resemble those found outside the core of an operating reactor was the Nuclear Regulatory Commission's Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program (LWR-PV-SDIP) (1)3. This program promoted better monitoring of the radiation exposure of reactor vessels and, thereby, provided for better assessment of vessel end-of-life conditions. An objective of the LWR-PV-SDIP was to develop improved procedures for reactor surveillance and document them in a series of ASTM standards (see Matrix E706). The primary means chosen for validating LWR-PV-SDIP procedures was by benchmarking a series of experimental and analytical studies in a variety of fields (see Guide E2005).1.1 This guide covers general approaches for benchmarking neutron transport calculations for pressure vessel surveillance programs in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor pressure vessel surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactor fluences with a higher degree of confidence.1.2 Although this guide and the companion guide, Guide E2005, are focused on power reactors, the principle of this guide is also applicable to non-power light water reactor pressure vessel surveillance programs.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 The Single Tube Method is designed to evaluate the efficacy of disinfectants against biofilm grown in the CDC biofilm reactor following the procedures outlined in Practice E3161. Biofilm grown in the CDC reactor is representative of biofilm that forms under high fluid shear on surfaces conducive to biofilm formation.5.1.1 Vegetative biofilm bacteria are phenotypically different from suspended planktonic cells of the same genotype. Biofilm growth reactors are engineered to produce biofilm with specific characteristics (2). Altering either the engineered system or operating conditions will modify those characteristics as well as the physicochemical environment. The goal in biofilm research and testing is to choose the growth reactor and operating conditions that generate the most relevant biofilm for the particular study.5.2 The test method was designed to determine the log10 reduction in bacteria after exposure to a disinfectant in a closed system.5.3 The test method uses 50 mL conical tubes. The conical geometry allows for disinfectant exposure to biofilm on all surfaces of the coupon. For foaming disinfectants or for disinfectants requiring a larger volume of neutralizer, 250 mL conical tubes are used which preserve the required geometry and allow for greater neutralization capacity.5.4 Each test includes three untreated control coupons (exposed to buffered dilution water) and five treated coupons (per disinfectant/concentration/contact time combination).1.1 This test method specifies the operational parameters required to perform a quantitative liquid disinfectant efficacy test against bacterial biofilm.1.2 The test method was optimized and validated for a Pseudomonas aeruginosa or Staphylococcus aureus biofilm grown in the CDC Biofilm Reactor (E3161). The method is suitable for evaluating additional bacteria grown using the procedures outlined in methods with comparable coupon dimensions such as Practice E3161, Test Method E2562, or Test Method E2196.1.3 Disinfectant preparation and contact time are used in the assessment according to the manufacturer’s instructions for use.1.4 The test method uses a closed system to treat biofilm. A coupon is placed in a single tube for the treatment, neutralization, and harvesting steps to prevent the loss of cells.1.5 This test method describes a harvesting and analysis procedure which includes vortexing and sonicating treated and untreated control biofilm, and recovery of culturable cells using filtration to lower the limit of detection. Biofilm population density is recorded as log10 colony-forming units per coupon. Efficacy is reported as a log10 reduction of culturable cells.1.6 Basic microbiology training is required to perform this assay.1.7 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.1.8 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.1.9 This international standard was developed in accordance with internationally recognized principles on standardization established in the 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 biofilms are phenotypically different from suspended cells of the same genotype. Research has shown that biofilm bacteria are more difficult to kill than suspended bacteria (4, 5). 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 practice is to direct a user in the growth of a P. aeruginosa or S. aureus biofilm by clearly defining the operational parameters to grow a biofilm that can be assessed for efficacy using the Standard Test Method for Evaluating Disinfectant Efficacy Against Pseudomonas aeruginosa Biofilm Grown in CDC Biofilm Reactor Using Single Tube Method (E2871).5.2 Operating the CDC Biofilm Reactor at the conditions specified in this method generates biofilm at log densities (log10 CFU per coupon) ranging from 8.0 to 9.5 for P. aeruginosa and 7.5 to 9.0 for S. aureus. These levels of biofilm are anticipated on surfaces conducive to biofilm formation such as the conditions outlined in this method.5.2.1 To achieve an S. aureus biofilm with a population comparable to that for P. aeruginosa using the bacterial liquid growth medium conditions specified here, the S. aureus biofilm must be grown at 36 °C ±2 °C rather than at room temperature (21 °C ±2 °C).1.1 This practice specifies the parameters for growing a Pseudomonas aeruginosa (ATCC 15442) or Staphylococcus aureus (ATCC 6538) biofilm that can be used for disinfectant efficacy testing using the Test Method for Evaluating Disinfectant Efficacy Against Pseudomonas aeruginosa Biofilm Grown in CDC Biofilm Reactor Using Single Tube Method (E2871) or in an alternate method capable of accommodating the coupons used in the CDC Biofilm Reactor. The resulting biofilm is representative of generalized situations where biofilm exist on hard, non-porous surfaces under shear rather than being representative of one particular environment. Additional bacteria may be grown using the basic procedure outlined in this document, however, alternative preparation procedures for frozen stock cultures and biofilm generation (for example, medium concentrations, baffle speed, temperature, incubation times, coupon types, etc.) may be necessary.1.2 This practice uses the CDC Biofilm Reactor created by the Centers for Disease Control and Prevention (1).2 The CDC Biofilm Reactor is a continuously stirred tank reactor (CSTR) with high wall shear. The reactor is versatile and may also be used for growing or characterizing various species of biofilm, or both (2-4) provided appropriate adjustments are made to the growth media and operational parameters of the reactor.1.3 Basic microbiology training is required to perform this practice.1.4 Units—The values stated in SI units are to be regarded as standard. No other units of measurement are included in this practice.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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4.1 Predictions of neutron radiation effects on pressure vessel steels are considered in the design of light-water moderated nuclear power reactors. Changes in system operating parameters often are made throughout the service life of the reactor vessel to account for radiation effects. Due to the variability in the behavior of reactor vessel steels, a surveillance program is warranted to monitor changes in the properties of actual vessel materials caused by long-term exposure to the neutron radiation and temperature environment of the reactor vessel. This practice describes the criteria that should be considered in planning and implementing surveillance test programs and points out precautions that should be taken to ensure that: (1) capsule exposures can be related to beltline exposures, (2) materials selected for the surveillance program are samples of those materials most likely to limit the operation of the reactor vessel, and (3) the test specimen types are appropriate for the evaluation of radiation effects on the reactor vessel.4.2 Guides E482 and E853 describe a methodology for estimation of neutron exposure obtained for reactor vessel surveillance programs. Regulators or other sources may describe different methods.4.3 The design of a surveillance program for a given reactor vessel must consider the existing body of data on similar materials in addition to the specific materials used for that reactor vessel. The amount of such data and the similarity of exposure conditions and material characteristics will determine their applicability for predicting radiation effects.1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. New advanced light-water small modular reactor designs with a nominal design output of 300 MWe or less have not been specifically considered in this practice. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials.1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) exceeds 1 × 1021 neutrons/m 2 (1 × 1017 n/cm2) at the inside surface of the ferritic steel reactor vessel.1.3 This practice does not provide specific procedures for monitoring the radiation induced changes in properties beyond the design life. Practice E2215 addresses changes to the withdrawal schedule during and beyond the design life.1.4 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.NOTE 1: The increased complexity of the requirements for a light-water moderated nuclear power reactor vessel surveillance program has necessitated the separation of the requirements into three related standards. Practice E185 describes the minimum requirements for design of a surveillance program. Practice E2215 describes the procedures for testing and evaluation of surveillance capsules removed from a reactor vessel. Guide E636 provides guidance for conducting additional mechanical tests. A summary of the many major revisions to Practice E185 since its original issuance is contained in Appendix X2.NOTE 2: This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. See Appendix X2.1.5 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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1.1 This practice covers analytical and analytical-experimental approaches that can be used to determine the variation in neutron exposure (fluence E > 1.0 MeV, dpa, etc.) and exposure rate and energy spectrum between surveillance locations and points in the pressure vessel wall. Procedures for reporting the results of these analyses with assigned uncertainties are also suggested. This practice deals with the physics-dosimetry aspects of surveillance programs and must be used in conjunction with other Matrix E 706 standards to provide extrapolations based on metallurgical damage correlations.1.2 The physics-dosimetry relationships determined from this practice may be used to estimate pressure vessel damage through application of Practice E 693 and Guide E 900 standards, using fluence (E > 1.0 MeV), dpa, or damage function derived exposure parameters as independent exposure variables. Supporting the application of these standards is E 944, E 1018, E 1005, and E 854 standards, identified in 2.1.1.3 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 and health practices and determine the applicability of regulatory limitations prior to use.

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4.1 Operation of commercial power reactors must conform to pressure-temperature limits during heatup and cooldown to prevent over-pressurization at temperatures that might cause non-ductile behavior in the presence of a flaw. Radiation damage to the reactor vessel is compensated for by adjusting the pressure-temperature limits to higher temperatures as the neutron damage accumulates. The present practice is to base that adjustment on the TTS produced by neutron irradiation as measured at the Charpy V-notch 41-J (30-ft·lbf) energy level. To establish pressure temperature operating limits during the operating life of the plant, a prediction of TTS must be made.4.1.1 In the absence of surveillance data for a given reactor material (see Practice E185 and E2215), the use of calculative procedures are necessary to make the prediction. Even when credible surveillance data are available, it will usually be necessary to interpolate or extrapolate the data to obtain a TTS for a specific time in the plant operating life. The embrittlement correlation presented herein has been developed for those purposes.4.2 Research has established that certain elements, notably copper (Cu), nickel (Ni), phosphorus (P), and manganese (Mn), cause a variation in radiation sensitivity of reactor pressure vessel steels. The importance of other elements, such as silicon (Si), and carbon (C), remains a subject of additional research. Copper, nickel, phosphorus, and manganese are the key chemistry parameters used in developing the calculative procedures described here.4.3 Only power reactor (PWR and BWR) surveillance data were used in the derivation of these procedures. The measure of fast neutron fluence used in the procedure is n/m2 (E > 1 MeV). Differences in fluence rate and neutron energy spectra experienced in power reactors and test reactors have not been accounted for in these procedures.1.1 This guide presents a method for predicting values of reference transition temperature shift (TTS) for irradiated pressure vessel materials. The method is based on the TTS exhibited by Charpy V-notch data at 41-J (30-ft·lbf) obtained from surveillance programs conducted in several countries for commercial pressurized (PWR) and boiling (BWR) light-water cooled (LWR) power reactors. An embrittlement correlation has been developed from a statistical analysis of the large surveillance database consisting of radiation-induced TTS and related information compiled and analyzed by Subcommittee E10.02. The details of the database and analysis are described in a separate report (ADJE090015-EA).2,3 This embrittlement correlation was developed using the variables copper, nickel, phosphorus, manganese, irradiation temperature, neutron fluence, and product form. Data ranges and conditions for these variables are listed in 1.1.1. Section 1.1.2 lists the materials included in the database and the domains of exposure variables that may influence TTS but are not used in the embrittlement correlation.1.1.1 The range of material and irradiation conditions in the database for variables used in the embrittlement correlation: 1.1.1.1 Copper content up to 0.4 %.1.1.1.2 Nickel content up to 1.7 %.1.1.1.3 Phosphorus content up to 0.03 %.1.1.1.4 Manganese content within the range from 0.55 to 2.0 %.1.1.1.5 Irradiation temperature within the range from 255 to 300°C (491 to 572°F).1.1.1.6 Neutron fluence within the range from 1 × 1021 n/m2 to 2 × 1024 n/m2 (E> 1 MeV).1.1.1.7 A categorical variable describing the product form (that is, weld, plate, forging).1.1.2 The range of material and irradiation conditions in the database for variables not included in the embrittlement correlation: 1.1.2.1 A533 Type B Class 1 and 2, A302 Grade B, A302 Grade B (modified), and A508 Class 2 and 3. Also, European and Japanese steel grades that are equivalent to these ASTM Grades.1.1.2.2 Submerged arc welds, shielded arc welds, and electroslag welds having compositions consistent with those of the welds used to join the base materials described in 1.1.2.1.1.1.2.3 Neutron fluence rate within the range from 3 × 1012 n/m2/s to 5 × 1016 n/m2/s (E > 1 MeV).1.1.2.4 Neutron energy spectra within the range expected at the reactor vessel region adjacent to the core of commercial PWRs and BWRs (greater than approximately 500MW electric).1.1.2.5 Irradiation exposure times of up to 25 years in boiling water reactors and 31 years in pressurized water reactors.1.2 It is the responsibility of the user to show that the conditions of interest in their application of this guide are addressed adequately by the technical information on which the guide is based. It should be noted that the conditions quantified by the database are not distributed evenly over the range of materials and irradiation conditions described in 1.1, and that some combination of variables, particularly at the extremes of the data range are under-represented. Particular attention is warranted when the guide is applied to conditions near the extremes of the data range used to develop the TTS equation and when the application involves a region of the data space where data is sparse. Although the embrittlement correlation developed for this guide was based on statistical analysis of a large database, prudence is required for applications that involve variable values beyond the ranges specified in 1.1. Due to strong correlations with other exposure variables within the database (that is, fluence), and due to the uneven distribution of data within the database (for example, the irradiation temperature and flux range of PWR and BWR data show almost no overlap) neither neutron fluence rate nor irradiation time sufficiently improved the accuracy of the predictions to merit their use in the embrittlement correlation in this guide. Future versions of this guide may incorporate the effect of neutron fluence rate or irradiation time, or both, on TTS , as such effects are described in (1).4 The irradiated material database, the technical basis for developing the embrittlement correlation, and issues involved in its application, are discussed in a separate report (ADJE090015-EA). That report describes the nine different TTS equations considered in the development of this guide, some of which were developed using more limited datasets (for example, national program data (2, 3)). If the material variables or exposure conditions of a particular application fall within the range of one of these alternate correlations, it may provide more suitable guidance.1.3 This guide is expected to be used in coordination with several standards addressing irradiation surveillance of light-water reactor vessel materials. Method of determining the applicable fluence for use in this guide are addressed in Guides E482, E944, and Test Method E1005. The overall application of these separate guides and practices is described in Practice E853.1.4 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 guide does not define how the TTS should be used to determine the final adjusted reference temperature, which would typically include consideration of the transition temperature before irradiation, the predicted TTS, and the uncertainties in the shift estimation method.1.6 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, 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 Each power reactor has a specific DEX value that is their technical requirement limit. These values may vary from about 200 to about 900 μCi/g based upon the height of their plant vent, the location of the site boundary, the calculated reactor coolant activity for a condition of 1 % fuel defects, and general atmospheric modeling that is ascribed to that particular plant site. Should the DEX measured activity exceed the technical requirement limit, the plant enters an LCO requiring action on plant operation by the operators.5.2 The determination of DEX is performed in a similar manner to that used in determining DEI, except that the calculation of DEX is based on the acute dose to the whole body and considers the noble gases 85mKr, 85Kr, 87Kr, 88Kr, 131mXe, 133mXe, 133Xe, 135mXe, 135Xe, and 138Xe which are significant in terms of contribution to whole body dose.5.3 It is important to note that only fission gases are included in this calculation, and only the ones noted in Table 1. For example 83mKr is not included even though its half-life is 1.86 hours. The reason for this is that this radionuclide cannot be easily determined by gamma spectrometry (low energy X-rays at 32 and 9 keV) and its dose consequence is vanishingly small compared to the other, more prevalent krypton radionuclides.5.4 Activity from 41Ar, 19F, 16N, and 11C, all of which predominantly will be in gaseous forms in the RCS, are not included in this calculation.5.5 If a specific noble-gas radionuclide is not detected, it should be assumed to be present at the minimum-detectable activity. The determination of dose-equivalent Xe-133 shall be performed using effective dose-conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12,3 or the average gamma-disintegration energies as provided in ICRP Publication 38 (“Radionuclide Transformations”) or similar source.1.1 This practice applies to the calculation of the dose equivalent to 133Xe in the reactor coolant of nuclear power reactors resulting from the radioactivity of all noble gas fission products.1.2 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered standard.1.3 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 Radiometric monitors shall provide a proven passive dosimetry technique for the determination of neutron fluence rate (flux density), fluence, and spectrum in a diverse variety of neutron fields. These data are required to evaluate and estimate probable long-term radiation-induced damage to nuclear reactor structural materials such as the steel used in reactor pressure vessels and their support structures.5.2 A number of radiometric monitors, their corresponding neutron activation reactions, and radioactive reaction products and some of the pertinent nuclear parameters of these RMs and products are listed in Table 1. Table 2 provides data (37) on the cumulative and independent fission yields of the important fission monitors. Not included in these tables are contributions to the yields from photo-fission, which can be especially significant for non-fissile nuclides (2-5, 27-29, 38-41).(A) All yield data are given as a percentage with associated uncertainties given as percentages of the percentage at the 1σ level.(B) For this fission yield evaluation (37), “Fast” indicates that the data was extracted from a wide range of reactor-based fission neutron spectra that can be characterized as having an average energy of ~0.4 MeV. Almost all of the fission reactions for U-238 and Th-232 occur above an effective threshold energy of ~1 MeV and, for Np-237, above ~0.2 MeV.1.1 This test method describes procedures for measuring the specific activities of radioactive nuclides produced in radiometric monitors (RMs) by nuclear reactions induced during surveillance exposures for reactor vessels and support structures. More detailed procedures for individual RMs are provided in separate standards identified in 2.1 and in Refs (1-5).2 The measurement results can be used to define corresponding neutron induced reaction rates that can in turn be used to characterize the irradiation environment of the reactor vessel and support structure. The principal measurement technique is high resolution gamma-ray spectrometry, although X-ray photon spectrometry and Beta particle counting are used to a lesser degree for specific RMs (1-29).1.1.1 The measurement procedures include corrections for detector background radiation, random and true coincidence summing losses, differences in geometry between calibration source standards and the RMs, self absorption of radiation by the RM, other absorption effects, radioactive decay corrections, and burn out of the nuclide of interest (6-26).1.1.2 Specific activities are calculated by taking into account the time duration of the count, the elapsed time between start of count and the end of the irradiation, the half life, the mass of the target nuclide in the RM, and the branching intensities of the radiation of interest. Using the appropriate half life and known conditions of the irradiation, the specific activities may be converted into corresponding reaction rates (2-5, 28-30).1.1.3 Procedures for calculation of reaction rates from the radioactivity measurements and the irradiation power time history are included. A reaction rate can be converted to neutron fluence rate and fluence using the appropriate integral cross section and effective irradiation time values, and, with other reaction rates can be used to define the neutron spectrum through the use of suitable computer programs (2-5, 28-30).1.1.4 The use of benchmark neutron fields for calibration of RMs can reduce significantly or eliminate systematic errors since many parameters, and their respective uncertainties, required for calculation of absolute reaction rates are common to both the benchmark and test measurements and therefore are self canceling. The benchmark equivalent fluence rates, for the environment tested, can be calculated from a direct ratio of the measured saturated activities in the two environments and the certified benchmark fluence rate (2-5, 28-30).1.2 This test method is intended to be used in conjunction with ASTM Guide E844 and existing or proposed ASTM practices, guides, and test methods that are also directly involved in the physics-dosimetry evaluation of reactor vessel and support structure surveillance measurements.1.3 The procedures in this test method are applicable to the measurement of radioactivity in RMs that satisfy the specific constraints and conditions imposed for their analysis. More detailed procedures for individual RM monitors are identified in 2.1 and in Refs 1-5 (see Table 1).1.4 This test method, along with the individual RM monitor standard methods, are intended for use by knowledgeable persons who are intimately familiar with the procedures, equipment, and techniques necessary to achieve high precision and accuracy in radioactivity measurements.1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard, except for the energy units based on the electron volt, keV and MeV, and the time units: minute (min), hour (h), day (d), and year (a).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 The HAFM test method is one of several available passive neutron dosimetry techniques (see, for example, Test Methods E854 and E1005). This test method can be used in combination with other dosimetry methods, or, if sufficient data are available from different HAFM sensor materials, as an alternative dosimetry test method. The HAFM method yields a direct measurement of total helium production in an irradiated sample. Absolute neutron fluence can then be inferred from this, assuming the appropriate spectrum integrated total helium production cross section. Alternatively, a calibration of the composite neutron detection efficiency for the HAFM method may be obtained by exposure in a benchmark neutron field where the fluence and spectrum averaged cross section are both known (see Guide E2005).5.2 HAFMs have the advantage of producing an end product, helium, which is stable, making the HAFM method very attractive for both short-term and long-term fluence measurements without requiring time-dependent corrections for decay. HAFMs are therefore ideal passive, time-integrating fluence monitors. Additionally, the burnout of the daughter product, helium, is negligible.5.2.1 Many of the HAFM materials can be irradiated in the form of unencapsulated wire segments (see 1.1.2). These segments can easily be fabricated by cutting from a standard inventoried material lot. The advantage is that encapsulation, with its associated costs, is not necessary. In several cases, unencapsulated wires such as Fe, Ni, Al/Co, and Cu, which are already included in the standard radiometric (RM) dosimetry sets (Table 1) can be used for both radiometric and helium accumulation dosimetry. After radiometric counting, the samples are later vaporized for helium measurement.(A) Evaluated 235U fission neutron spectrum averaged helium production cross section and energy range in which 90 % of the reactions occur. All values are obtained from ENDF/B-V Gas Production Dosimetry File data. Bracketed terms indicate cross section is for naturally occurring element.(B) Often included in dosimetry sets as a radiometric monitor, either as a pure element foil or wire or, in the case of aluminum, as an allaying material for other elements.5.3 The HAFM method is complementary to RM and solid state track recorder (SSTR) foils, and has been used as an integral part of the multiple foil method. The HAFM method follows essentially the same principle as the RM foil technique, which has been used successfully for accurate neutron dosimetry. Various HAFM sensor materials exist which have significantly different neutron energy sensitivities from each other. HAFMs containing 10B and 6Li have been used routinely in LMFBR applications in conjunction with RM foils. The resulting data are entirely compatible with existing adjustment methods for radiometric foil neutron dosimetry (refer to Guide E944 ).5.4 An application for the HAFM method lies in the direct analysis of pressure vessel wall scrapings or Charpy block surveillance samples. Measurements of the helium production in these materials can provide in situ integral information on the neutron fluence spectrum. This application can provide dosimetry information at critical positions where conventional dosimeter placement is difficult if not impossible. Analyses must first be conducted to determine the boron, lithium, and other component concentrations, and their homogeneities, so that their possible contributions to the total helium production can be determined. Boron (and lithium) can be determined by converting a fraction of the boron to helium with a known thermal neutron exposure. Measurements of the helium in the material before and after the exposure will enable a determination of the boron content (7). Boron level down to less than 1 wt. ppm can be obtained in this manner.5.5 By careful selection of the appropriate HAFM sensor material and its mass, helium concentrations ranging from ∼10−14 to 10−1 atom fraction can be generated and measured. In terms of fluence, this represents a range of roughly 1012 to 1027 n/cm2. Fluence (>1 MeV) values that may be encountered during routine surveillance testing are expected to range from ∼3 × 1014 to 2 × 10 20 n/cm2, which is well within the range of the HAFM technique.5.6 The analysis of HAFMs requires an absolute determination of the helium content. The analysis system specified in this test method incorporates a specialized mass spectrometer in conjunction with an accurately calibrated helium spiking system. Helium determination is by isotope dilution with subsequent isotope ratio measurement. The fact that the helium is stable makes the monitors permanent with the helium analysis able to be conducted at a later time, often without the inconvenience in handling caused by induced radioactivity. Such systems for analysis exist, and additional analysis facilities could be reproduced, should that be required. In this respect, therefore, the analytical requirements are similar to other ASTM test methods.1.1 This test method describes the concept and use of helium accumulation for neutron fluence dosimetry for reactor vessel surveillance. Although this test method is directed toward applications in vessel surveillance, the concepts and techniques are equally applicable to the general field of neutron dosimetry. The various applications of this test method for reactor vessel surveillance are as follows:1.1.1 Helium accumulation fluence monitor (HAFM) capsules,1.1.2 Unencapsulated, or cadmium or gadolinium covered, radiometric monitors (RM) and HAFM wires for helium analysis,1.1.3 Charpy test block samples for helium accumulation, and1.1.4 Reactor vessel (RV) wall samples for helium accumulation.1.2 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.1.3 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.

定价： 646元 / 折扣价： 550 元 加购物车

3.1 The mechanical properties of steels and other metals are altered by exposure to neutron radiation. These property changes are assumed to be a function of chemical composition, metallurgical condition, temperature, fluence (perhaps also fluence rate), and neutron spectrum. The influence of these variables is not completely understood. The functional dependency between property changes and neutron radiation is summarized in the form of damage exposure parameters that are weighted integrals over the neutron fluence spectrum.3.2 The evaluation of neutron radiation effects on pressure vessel steels and the determination of safety limits requires the knowledge of uncertainties in the prediction of radiation exposure parameters (for example, dpa (Practice E693), neutron fluence greater than 1.0 MeV, neutron fluence greater than 0.1 MeV, thermal neutron fluence, etc.). This practice describes recommended procedures and data for determining these exposure parameters (and the associated uncertainties) for test reactor experiments.3.3 The nuclear industry draws much of its information from databases that come from test reactor experiments. Therefore, it is essential that reliable databases are obtained from test reactors to assess safety issues in Light Water Reactor (LWR) nuclear power plants.1.1 This practice covers the methodology summarized in Annex A1 to be used in the analysis and interpretation of physics-dosimetry results from test reactors.1.2 This practice relies on, and ties together, the application of several supporting ASTM standard practices, guides, and methods.1.3 Support subject areas that are discussed include reactor physics calculations, dosimeter selection and analysis, exposure units, and neutron spectrum adjustment methods.1.4 This practice is directed towards the development and application of physics-dosimetry-metallurgical data obtained from test reactor irradiation experiments that are performed in support of the operation, licensing, and regulation of LWR nuclear power plants. It specifically addresses the physics-dosimetry aspects of the problem. Procedures related to the analysis, interpretation, and application of both test and power reactor physics-dosimetry-metallurgy results are addressed in Practices E185, E853, and E1035, Guides E900, E2005, E2006 and Test Method E646. See also E706.1.5 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.

定价： 590元 / 折扣价： 502 元 加购物车

This PDF includes GI #2. 1. Scope 1.1 This Standard applies to outdoor bushings that have lightning impulse insulation levels 110-1950kV. They are for use as components of liquid-filled transformers and reactors on systems with a nominal voltage of

定价： 455元 / 折扣价： 387 元

1.1 This specification covers seamless annealed or cold-worked, austenitic or martensitic stainless steel tubing of 0.100 to 1.0 in. [2.5 to 25 mm] outside diameter with wall thickness of 0.050 in. [1.3 mm] or less for use at high temperature in liquid metal-cooled reactor plants.1.2 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 must be used independently of the other. Combining values from the two systems may result in nonconformance with the specification.1.3 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.

定价： 0元 / 折扣价： 0 元

1.1 This specification covers seamless, annealed or cold worked, austenitic or martensitic stainless steel duct tubes of 2 to 7-in. [51 to 178 mm] outside dimensions with wall thickness of 0.250 in. [6.35 mm] or less for use at high temperature in liquid metal-cooled reactor plants.1.2 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 must be used independently of the other. Combining values from the two systems may result in nonconformance with the specification.1.3 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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