1.1 This practice covers the exposure of plastics to a specific test environment. The test environment is an externally-heated laboratory-scale reactor that simulates a composting system. Plastic exposure occurs in the presence of a media undergoing aerobic composting. The standard media simulates a municipal solid waste from which inert materials have been removed. This practice allows for the use of other media to represent particular waste streams. This practice provides exposed specimens for further testing and for comparison with controls. This test environment does not necessarily reproduce conditions that could occur in a particular full-scale composting process. 1.2 Changes in the material properties of the plastic and controls should be determined using appropriate ASTM test procedures. Changes could encompass physical and chemical changes such as disintegration and degradation. 1.3 This practice may be used for different purposes. Therefore, the interested parties must select: exposure conditions from those allowed by this practice; criteria for a valid exposure, that is, minimum or maximum change requirements for the compost and controls; and the magnitudes of material properties changes required for the plastic specimens. 1.4 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only. 1.5 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. Specific hazard statements are given in Section 8. Note 1-There is no similar or equivalent ISO standard.

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3.1 Temperature monitors are used in surveillance capsules in accordance with Practice E2215 to estimate the maximum value of the surveillance specimen irradiation temperature. Temperature monitors are needed to give evidence of overheating of surveillance specimens beyond the expected temperature. Because overheating causes a reduction in the amount of neutron radiation damage to the surveillance specimens, this overheating could result in a change in the measured properties of the surveillance specimens that would lead to an unconservative prediction of damage to the reactor vessel material.3.2 The magnitude of the reduction of radiation damage with overheating depends on the composition of the material and time at temperature. Guide E900 provides an accepted method for quantifying the temperature effect. Because the evidence from melt wire monitors gives no indication of the duration of overheating above the expected temperature as indicated by melting of the monitor, the significance of overheating events cannot be quantified on the basis of temperature monitors alone. Indication of overheating does serve to alert the user of the data to further evaluate the irradiation temperature exposure history of the surveillance capsule.3.3 This guide is included in Master Matrix E706 that relates several standards used for irradiation surveillance of light-water reactor vessel materials. It is intended primarily to amplify the requirements of Practice E185 in the design of temperature monitors for the surveillance program. It may also be used in conjunction with Practice E2215 to evaluate the post-irradiation test measurements.1.1 This guide describes the application of melt wire temperature monitors and their use for reactor vessel surveillance of light-water power reactors as called for in Practices E185 and E2215.1.2 The purpose of this guide is to recommend the selection and use of the common melt wire technique where the correspondence between melting temperature and composition of different alloys is used as a passive temperature monitor. Guidelines are provided for the selection and calibration of monitor materials; design, fabrication, and assembly of monitor and container; post-irradiation examinations; interpretation of the results; and estimation of uncertainties.1.3 The values stated in SI units are to be regarded as standard. The values given in parentheses are mathematical conversions to inch-pound units that are provided for information only and are not considered standard.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. (See Note 1.)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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AbstractThese methods cover general procedures for the calibration of radiation detectors and the analysis of radionuclides. For each individual radionuclide, one or more of these methods may apply. These methods are concerned only with specific radionuclide measurements. The chemical and physical properties of the radionuclides are beyond the scope of this standard. Among the measurement standards discussed are: the calibration and usage of germanium detectors, scintillation detector systems, scintillation detectors for simple and complex spectra, and counting methods such as beta particle counting, aluminum absorption curve, alpha particle counting, and liquid scintillation counting. For each of the methods, the scope, apparatus used, summary of methods, preparation of apparatus, calibration procedure, measurement of radionuclide, performance testing, sources of uncertainty, precautions and tests, and calculations are detailed.1.1 This guide covers general procedures for the calibration of radiation detectors and measurement for radiation metrology for reactor dosimetry. For any particular radionuclide, one or more of these methods may apply.1.2 These techniques are concerned only with specific radionuclide measurements. The chemical and physical properties of the radionuclides are not within the scope of this standard.1.3 E3376, Standard Practice for Calibration and Usage of Germanium Detectors in Radiation Metrology for Reactor Dosimetry, was previously in Guide E181 and is now found in Volume 12.02 of the Annual Book of ASTM Standards. The discussion herein is not a sufficient substitute for the full standard. This guide is specifically NOT to be used as a direct reference to Practice E3376. Only the standard listed provides sufficient information to serve as a reference.1.4 Additional information on the setup, calibration, and quality control for radiometric detectors and measurements is given in Guide C1402 and Practice D7282.1.5 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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3.1 General: 3.1.1 The methodology recommended in this guide specifies criteria for validating computational methods and outlines procedures applicable to pressure vessel related neutronics calculations for test and power reactors. The material presented herein is useful for validating computational methodology and for performing neutronics calculations that accompany reactor vessel surveillance dosimetry measurements (see Master Matrix E706 and Practice E853). Briefly, the overall methodology involves: (1) methods-validation calculations based on at least one well-documented benchmark problem, and (2) neutronics calculations for the facility of interest. The neutronics calculations of the facility of interest and of the benchmark problem should be performed consistently, with important modeling parameters kept the same or as similar as is feasible. In particular, the same energy group structure and common broad-group microscopic cross sections should be used for both problems. Further, the benchmark problem should be characteristically similar to the facility of interest. For example, a power reactor benchmark should be utilized for power reactor calculations. Non-power reactors may have special features that may affect pressure vessel fluence and require consideration when developing a benchmark, such as beam tubes, irradiation facilities, and non-core neutron sources. The neutronics calculations involve two tasks: (1) determination of the neutron source distribution in the reactor core by utilizing diffusion theory (or transport theory) calculations in conjunction with reactor power distribution measurements, and (2) performance of a fixed fission rate neutron source (fixed-source) transport theory calculation to determine the neutron fluence rate distribution in the reactor core, through the internals and in the pressure vessel. Some neutronics modeling details for the benchmark, test reactor, or the power reactor calculation will differ; therefore, the procedures described herein are general and apply to each case. (See NUREG/CR-5049, NUREG/CR-1861, NUREG/CR-3318, and NUREG/CR-3319.)3.1.2 It is expected that transport calculations will be performed whenever pressure vessel surveillance dosimetry data become available and that quantitative comparisons will be performed as prescribed by 3.2.2. All dosimetry data accumulated that are applicable to a particular facility should be included in the comparisons.3.2 Validation—Prior to performing transport calculations for a particular facility, the computational methods must be validated by comparing results with measurements made on a benchmark experiment. Criteria for establishing a benchmark experiment for the purpose of validating neutronics methodology should include those set forth in Guides E944 and E2006 as well as those prescribed in 3.2.1. A discussion of the limiting accuracy of benchmark validation discrete ordinate radiation transport procedures for the LWR surveillance program is given in Reference (1).4 Reference (2) provides details on the benchmark validation for a Monte Carlo radiation transport code.3.2.1 Requirements for Benchmarks—In order for a particular experiment to qualify as a calculational benchmark, the following criteria are recommended:3.2.1.1 Sufficient information must be available to accurately determine the neutron source distribution in the reactor core.3.2.1.2 Measurements must be reported in at least two ex-core locations, well separated by steel or coolant.3.2.1.3 Uncertainty estimates should be reported for dosimetry measurements and calculated fluences including calculated exposure parameters and calculated dosimetry activities.3.2.1.4 Quantitative criteria, consistent with those specified in the methods validation 3.2.2, must be published and demonstrated to be achievable.3.2.1.5 Differences between measurements and calculations should be consistent with the uncertainty estimates in 3.2.1.3.3.2.1.6 Results for exposure parameter values of neutron fluence greater than 1 MeV and 0.1 MeV [φ(E > 1 MeV and 0.1 MeV)] and of displacements per atom (dpa) in iron should be reported consistent with Practices E693 and E853.3.2.1.7 Reaction rates (preferably established relative to neutron fluence standards) must be reported for 237Np(n,f) or 238U(n,f), and 58Ni(n,p) or 54Fe(n,p); additional reactions that aid in spectral characterization, such as provided by Cu, Ti, and Co-Al, should also be included in the benchmark measurements. The 237Np(n,f) reaction is particularly important because it is sensitive to the same neutron energy region as the iron dpa. Practices E693 and E853 and Guides E844 and E944 discuss this criterion.3.2.2 Methodology Validation—It is essential that the neutronics methodology employed for predicting neutron fluence in a reactor pressure vessel be validated by accurately predicting appropriate benchmark dosimetry results. In addition, the following documentation should be submitted: (1) convergence study results, and (2) estimates of variances and covariances for fluence rates and reaction rates arising from uncertainties in both the source and geometric modeling. For Monte Carlo calculations, the convergence study results should also include (3) an analysis of the figure-of-merit (FOM) as a function of particles history, and if applicable, (4) the description of the technique utilized to generate the weight window parameters.3.2.2.1 For example, model specifications for discrete-ordinates method on which convergence studies should be performed include: (1) neutron cross sections or energy group structure, (2) spatial mesh, and (3) angular quadrature. Reference (3) evaluates the effects of many discrete-ordinates parameters individually and in combination and may help guide the analysis. For regions adjacent to the reactor core, one-dimensional calculations may be performed to check the adequacy of group structure and spatial mesh. Two-dimensional calculations should be employed to check the adequacy of the angular quadrature. A P3 cross section expansion is recommended along with a S8 minimum quadrature. For regions that are not adjacent to the reactor core, convergence studies for spatial mesh and angular quadrature should apply three-dimensional calculations.3.2.2.2 Uncertainties that are propagated from known uncertainties in nuclear data should be considered in the analysis. The uncertainty analysis for discrete ordinates codes may be performed with sensitivity analysis as discussed in References (4, 5). In Monte Carlo analysis the uncertainties can be treated by a perturbation analysis as discussed in Reference (6). Appropriate computer programs and covariance data are available and sensitivity data may be obtained as an intermediate step in determining uncertainty estimates.53.2.2.3 Effects of known uncertainties in geometry and source distribution should be evaluated based on the following test cases: (1) reference calculation with a time-averaged source distribution and with best estimates of the core and pressure vessel locations, (2) reference case geometry with maximum and minimum expected deviations in the source distribution, and (3) reference case source distribution with maximum expected spatial perturbations of the core, pressure vessel, and other pertinent locations.3.2.2.4 Measured and calculated integral parameters should be compared for all test cases. It is expected that larger uncertainties are associated with geometry and neutron source specifications than with parameters included in the convergence study. Problems associated with space, energy, and angle discretizations can be identified and corrected. Uncertainties associated with geometry specifications are inherent in the structure tolerances. Calculations based on the expected extremes provide a measure of the sensitivity of integral parameters to the selected variables. Variations in the proposed convergence and uncertainty evaluations are appropriate when the above procedures are inconsistent with the methodology to be validated. As-built data could be used to reduce the uncertainty in geometrical dimensions.3.2.2.5 In order to illustrate quantitative criteria based on measurements and calculations that should be satisfied, let ψ denote a set of logarithms of calculation (Ci) to measurement (Ei) ratios. Specifically,where qi and N are defined implicitly and the wi are weighting factors. Because some reactions provide a greater response over a spectral region of concern than other reactions, weighting factors may be utilized when their selection method is well documented and adequately defended, such as through a least-squares adjustment method as detailed in Guide E944. In the absence of the use of a least-squares adjustment methodology, the mean of the set q is given byand the best estimate of the variance, S2, is3.2.2.6 The neutronics methodology is validated if (in addition to qualitative model evaluation) all of the following criteria are satisfied:(1) The bias, |q|, is less than ε1,(2) The standard deviation, S, is less than ε2,(3) All absolute values of the natural logarithmic of the C/E ratios (|q|, i = 1 ... N) are less than ε3, and(4) ε1, ε2, and ε3 are defined by the benchmark measurement documentation and demonstrated to be attainable for all items with which calculations are compared.3.2.2.7 Note that a nonzero log-mean of the Ci/Ei ratios indicates that a bias exists. Possible sources of a bias are: (1) source normalization, (2) neutronics data, (3) transverse leakage corrections (if applicable), (4) geometric modeling, and (5) mathematical approximations. Reaction rates, equivalent fission fluence rates, or exposure parameter values (for example, φ(E > 1 MeV) and dpa) may be used for validating the computational methodology if appropriate criteria (that is, as established by 3.2.2.5 and 3.2.2.6) are documented for the benchmark of interest. Accuracy requirements for reactor vessel surveillance specific benchmark validation procedures are discussed in Guide E2006. The validation testing for the generic discrete ordinates and Monte Carlo transport methods is discussed in References (1, 2).3.2.2.8 One acceptable procedure for performing these comparisons is: (1) obtain group fluence rates at dosimeter locations from neutronics calculations, (2) collapse the Guide E1018 recommended dosimetry cross section data to a multigroup set consistent with the neutron energy group fluence rates or obtain a fine group spectrum (consistent with the dosimetry cross section data) from the calculated group fluence rates, (3) fold the energy group fluence rates with the appropriate cross sections, and (4) compare the calculated and experimental data according to the specified quantitative criteria.3.3 Determination of the Fixed Fission Source—The power distribution in a typical reactor undergoes significant change during the life of the reactor. A time-averaged power distribution is recommended for use in determination of the neutron source distribution utilized for damage predictions. An adjoint procedure, described in 3.3.2, may be more appropriate for dosimetry comparisons involving product nuclides with short half-lives. For multigroup methods, the fixed source may be determined from the equation:where:r   =   a spatial node,g   =   an energy group,v   =   average number of neutrons per fission,xg   =   fraction of the fission spectrum in group g, andPr   =   fission rate in node r.3.3.1 Note that in addition to the fission rate, v and xg will vary with fuel burnup, and a proper time average of these quantities should be used. The ratio between fission rate and power (that is, fission/s per watt) will also vary with burnup for any given spatial node.3.3.2 An adjoint procedure may be used as suggested in NUREG/CR-5049 instead of calculation with a time-averaged source calculation.3.3.2.1 The influence of changing source distribution is discussed in Reference (8). For dosimetry comparisons involving product nuclides with short half-lives, these changes in the power distribution may be significant. In this situation, a suitably averaged power distribution can be obtained by weighting the time-dependent power distribution using a factor proportional to:where:f   =   weighting factor at time, t,λ   =   decay constant for the nuclide of interest, andt   =   time from the start of the exposure.This averaging is different for each nuclide, therefore the use of the adjoint procedure avoids unecessary repetitions of the transport calculations in order to validate calculations using dosimetry results as described in 3.2.2.3.3.2.2 Care should be exercised to ensure that adjoint calculations adequately address cycle-to-cycle variations in coolant densities and any changes to the geometric configuration of the reactor.3.4 Calculation of the Neutron Fluence Rate Based on a Fixed Source in the Reactor Core—The discussion in this section relates to methods validation calculations and to routine surveillance calculations. In either case, neutron transport calculations must estimate the neutron fluence rate in the core, through the internals, in the reactor pressure vessel, and outside the vessel, if for example, ex-vessel dosimetry is used. Procedures for methods validation differ very little from procedures for predicting neutron fluence rate in the pressure vessel or test facility; consequently, the following procedure is recommended:3.4.1 Obtain detailed geometric and composition descriptions of the material configurations involved in the transport calculation. Uncertainty in the data should also be estimated.3.4.2 Obtain applicable cross section sets from appropriate data bases such as:3.4.2.1 The evaluated nuclear data file (ENDF/B or its equivalent), or3.4.2.2 A fine group library obtained by processing the above file (for example, see Reference (9)).3.4.3 Perform a one-dimensional, fixed-source, fine-group calculation in order to collapse the fine-group cross sections to a broad-group set for multidimensional calculations. At least two broad-group sets are recommended for performing the one-dimensional group structure convergence evaluation. The broad-group structure should emphasize the high-energy range and should take cross section minima of important materials (for example, iron) into consideration.3.4.4 Perform the convergence studies outlined in 3.2.2.3.4.5 Perform two- or three-dimensional fixed-source transport calculations based on the model established in 3.4.1 – 3.4.4.3.4.6 Compare appropriate dosimetry results with neutronics results from 3.4.5 according to the procedure given in 3.2.2. It is recommended that all valid lifetime-accumulated reactor dosimetry data be included in this comparison each time new data become available except when dosimeter-specific comparisons are made.3.4.7 Repeat appropriate steps if validation criteria are not satisfied. Note that a reactor dosimetry datum may be discarded if the associated C/E ratios differ substantially from the average of the applicable C/E ratios and a measurement error can be suspected. A measurement error can be suspected if the deviation from the average exceeds the equivalent of three standard deviations. In addition, the source for reactor calculations may be scaled to minimize the bias and variance defined by Eq 2 and Eq 3 provided that data are not discarded as a consequence of scaling the source.3.4.8 Results from neutronics calculations may be used in a variety of ways:3.4.8.1 Determine a single normalization constant that minimizes bias in the calculated values relative to the measurements in order to scale the group fluences. This is a simple and frequently used alternative to adjustment procedures. However, the magnitude of this constant should be critically examined in terms of estimated source uncertainties.3.4.8.2 Use a spectrum adjustment procedure as recommended in Guide E944 using calculated group fluences and dosimetry data with uncertainty estimates to obtain an adjustment to the calculated group fluences and exposure parameters. Predicted pressure vessel fluences could then incorporate the spectral and normalization data obtained from the adjusted fluences.3.4.8.3 Use the calculated fluence spectrum with Practice E693 for damage exposure predictions.3.4.8.4 It is expected that in some cases the procedure recommended above will be inconsistent with some methodologies to be validated. In these cases procedural variations are appropriate but should be well documented.1.1 Need for Neutronics Calculations—An accurate calculation of the neutron fluence and fluence rate at several locations is essential for the analysis of integral dosimetry measurements and for predicting irradiation damage exposure parameter values in the pressure vessel. Exposure parameter values may be obtained directly from calculations or indirectly from calculations that are adjusted with dosimetry measurements; Guide E944 and Practice E853 define appropriate computational procedures.1.2 Methodology—Neutronics calculations for application to reactor vessel surveillance encompass three essential areas: (1) validation of methods by comparison of calculations with dosimetry measurements in a benchmark experiment, (2) determination of the neutron source distribution in the reactor core, and (3) calculation of neutron fluence rate at the surveillance position and in the pressure vessel.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 Vegetative biofilm bacteria are phenotypically different from suspended cells of the same genotype. Biofilm growth reactors are engineered to produce biofilms with specific characteristics. Altering either the engineered system or operating conditions will modify those characteristics.5.2 The purpose of this test method is to direct a user in how to grow, sample, and analyze a P. aeruginosa biofilm under low fluid shear and close to the air/liquid interface using the DFR. The P. aeruginosa biofilm that grows has a smooth appearance that varies across the coupon surface and is loosely attached. Microscopically, the biofilm is sheet-like with few architectural details. This laboratory biofilm could represent those found on produce sprayers, on food processing conveyor belts, on catheters, in lungs with cystic fibrosis, and oral biofilms, for example. The biofilm generated in the DFR is also suitable for efficacy testing. After the 54 h growth phase is complete, the user may add the treatment in situ or harvest the coupons and treat them individually. Research has shown that P. aeruginosa biofilms grown in the DFR were less tolerant to disinfection than biofilms grown under high shear conditions.51.1 This test method specifies the operational parameters required to grow a repeatable2 Pseudomonas aeruginosa biofilm close to the air/liquid interface in a reactor with a continuous flow of nutrients under low fluid shear conditions. The resulting biofilm is representative of generalized situations where biofilm exists at the air/liquid interface under low fluid shear rather than representative of one particular environment.1.2 This test method uses the drip flow biofilm reactor. The drip flow biofilm reactor (DFR) is a plug flow reactor with laminar flow resulting in low fluid shear. The reactor is versatile and may also be used for growing and/or characterizing biofilms of different species, although this will require changing the operational parameters to optimize the method based upon the growth requirements of the new organism.1.3 This test method describes how to sample and analyze biofilm for viable cells. Biofilm population density is recorded as log colony forming units per surface area.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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3.1 Reactor vessels made of ferritic steels are designed with the expectation of progressive changes in material properties resulting from in-service neutron exposure. In the operation of light-water-cooled nuclear power reactors, changes in pressure-temperature (P – T) limits are made periodically during service life to account for the effects of neutron radiation on the ductile-to-brittle transition temperature material properties. If the degree of neutron embrittlement becomes large, the restrictions on operation during normal heat-up and cool down may become severe. Additional consideration should be given to postulated events, such as pressurized thermal shock (PTS). A reduction in the upper shelf toughness also occurs from neutron exposure, and this decrease may reduce the margin of safety against ductile fracture. When it appears that these situations could develop, certain alternatives are available that reduce the problem or postpone the time at which plant restrictions must be considered. One of these alternatives is to thermally anneal the reactor vessel beltline region, that is, to heat the beltline region to a temperature sufficiently above the normal operating temperature to recover a significant portion of the original fracture toughness and other material properties that were degraded as a result of neutron embrittlement.3.2 Preparation and planning for an in-service anneal should begin early so that pertinent information can be obtained to guide the annealing operation. Sufficient time should be allocated to evaluate the expected benefits in operating life to be gained by annealing; to evaluate the annealing method to be employed; to perform the necessary system studies and stress evaluations; to evaluate the expected annealing recovery and reembrittlement behavior; to develop and functionally test such equipment as may be required to do the in-service annealing; and, to train personnel to perform the anneal.3.3 Selection of the annealing temperature requires a balance of opposing conditions. Higher annealing temperatures, and longer annealing times, can produce greater recovery of fracture toughness and other material properties and thereby increase the post-anneal lifetime. The annealing temperature also can have an impact on the reembrittlement trend after the anneal. On the other hand, higher temperatures can create other undesirable property effects such as permanent creep deformation or temper embrittlement. These higher temperatures also can cause engineering difficulties, that is, core and coolant removal and storage, localized heating effects, etc., in preventing the annealing operation from distorting the vessel or damaging vessel supports, primary coolant piping, adjacent concrete, insulation, etc. See ASME Code Case N-557 for further guidance on annealing conditions and thermal-stress evaluations (2).3.3.1 When a reactor vessel approaches a state of embrittlement such that annealing is considered, the major criterion is the number of years of additional service life that annealing of the vessel will provide. Two pieces of information are needed to answer the question: the post-anneal adjusted RTNDT and upper shelf energy level, and their subsequent changes during future irradiation. Furthermore, if a vessel is annealed, the same information is needed as the basis for establishing pressure-temperature limits for the period immediately following the anneal and demonstrating compliance with other design requirements and the PTS screening criteria. The effects on upper shelf toughness similarly must be addressed. This guide primarily addresses RTNDT changes. Handling of the upper shelf is possible using a similar approach as indicated in NRC Regulatory Guide 1.162. Appendix X1 provides a bibliography of existing literature for estimating annealing recovery and reembrittlement trends for these quantities as related to U.S. and other country pressure-vessel steels, with primary emphasis on U.S. steels.3.3.2 A key source of test material for determining the post-anneal RTNDT, upper shelf energy level, and the reembrittlement trend is the original surveillance program, provided it represents the critical materials in the reactor vessel.6 Appendix X2 describes an approach to estimate changes in RTNDT both due to the anneal and reirradiation. The first purpose of Appendix X2 is to suggest ways to use available materials most efficiently to determine the post-anneal RTNDT and to predict the reembrittlement trend, yet leave sufficient material for surveillance of the actual reembrittlement for the remaining service life. The second purpose is to describe alternative analysis approaches to be used to assess test results of archive (or representative) materials to obtain the essential post-anneal and reirradiation RTNDT, upper shelf energy level, or fracture toughness, or a combination thereof.3.3.3 An evaluation must be conducted of the engineering problems posed by annealing at the highest practical temperature. Factors required to be investigated to reduce the risk of distortion and damage caused by mechanical and thermal stresses at elevated temperatures to relevant system components, structures, and control instrumentation are described in 5.1.3 and 5.1.4.3.4 Throughout the annealing operation, accurate measurement of the annealing temperature at key defined locations must be made and recorded for later engineering evaluation.3.5 After the annealing operation has been carried out, several steps should be taken. The predicted improvement in fracture toughness properties must be verified, and it must be demonstrated that there is no damage to key components and structures.3.6 Further action may be required to demonstrate that reactor vessel integrity is maintained within ASME Code requirements such as indicated in the referenced ASME Code Case N-557 (2). Such action is beyond the scope of this guide.AbstractThis guide covers the general procedures to be considered (as well as the demonstration of their effectiveness) for conducting in-service thermal anneals of light-water moderated nuclear reactor vessels. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods, as well as to provide direction for the development of a vessel annealing procedure and a post-annealing vessel radiation surveillance program. Guidelines are provided to determine the post-anneal reference nil-ductility transition temperature (RTNDT), the Charpy V-notch upper shelf energy level, fracture toughness properties, and the predicted reembrittlement trend for these properties for reactor vessel beltline materials.1.1 This guide covers the general procedures for conducting an in-service thermal anneal of a light-water moderated nuclear reactor vessel and demonstrating the effectiveness of the procedure. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. The improvement in mechanical properties generally is assessed using Charpy V-notch impact test results, or alternatively, fracture toughness test results or inferred toughness property changes from tensile, hardness, indentation, or other miniature specimen testing (1).21.2 This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods. Certain inherent limiting factors must be considered in developing an annealing procedure. These factors include system-design limitations; physical constraints resulting from attached piping, support structures, and the primary system shielding; the mechanical and thermal stresses in the components and the system as a whole; and, material condition changes that may limit the annealing temperature.1.3 This guide provides direction for development of the vessel annealing procedure and a post-annealing vessel radiation surveillance program. The development of a surveillance program to monitor the effects of subsequent irradiation of the annealed-vessel beltline materials should be based on the requirements and guidance described in Practices E185 and E2215. The primary factors to be considered in developing an effective annealing program include the determination of the feasibility of annealing the specific reactor vessel; the availability of the required information on vessel mechanical and fracture properties prior to annealing; evaluation of the particular vessel materials, design, and operation to determine the annealing time and temperature; and, the procedure to be used for verification of the degree of recovery and the trend for reembrittlement. Guidelines are provided to determine the post-anneal reference nil-ductility transition temperature (RTNDT), the Charpy V-notch upper shelf energy level, fracture toughness properties, and the predicted reembrittlement trend for these properties for reactor vessel beltline materials. This guide emphasizes the need to plan well ahead in anticipation of annealing if an optimum amount of post-anneal reembrittlement data is to be available for use in assessing the ability of a nuclear reactor vessel to operate for the duration of its present license, or qualify for a license extension, or both.1.4 The values stated in either SI units or inch-pound units are to be regarded separately as standard. The values stated in each system are not necessarily exact equivalents; therefore, to ensure conformance with the standard, each system shall be used independently of the other, and values from the two systems shall not be combined.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 In neutron dosimetry, a fission or non-fission dosimeter, or combination of dosimeters, can be used for determining a fluence rate, fluence, or neutron spectrum in nuclear reactors. Each dosimeter is sensitive to a specific energy range, and, if desired, increased accuracy in a fluence-rate spectrum can be achieved by the use of several dosimeters each covering specific neutron energy ranges.4.2 A wide variety of detector materials is used for various purposes. Many of these substances overlap in the energy of the neutrons which they will detect, but many different materials are used for a variety of reasons. These reasons include available analysis equipment, different cross sections for different fluence-rate levels and spectra, preferred chemical or physical properties, and, in the case of radiometric dosimeters, varying requirements for different half-life isotopes, possible interfering activities, and chemical separation requirements.1.1 This guide covers the selection, design, irradiation, post-irradiation handling, and quality control of neutron dosimeters (sensors), thermal neutron shields, and capsules for reactor surveillance neutron dosimetry.1.2 The values stated in SI units are to be regarded as standard. Values in parentheses are for information only.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 The objectives of a reactor vessel surveillance program are twofold. The first requirement of the program is to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region resulting from exposure to neutron irradiation and the thermal environment. The second requirement is to make use of the data obtained from the surveillance program to determine the conditions under which the vessel can be operated throughout its service life.3.1.1 To satisfy the first requirement of 3.1, the tasks to be carried out are straightforward. Each of the irradiation capsules that comprise the surveillance program may be treated as a separate experiment. The goal is to define and carry to completion a dosimetry program that will, a posteriori, describe the neutron field to which the material test specimens were exposed. The resultant information will then become part of a database applicable in a stricter sense to the specific plant from which the capsule was removed, but also in a broader sense to the industry as a whole.3.1.2 To satisfy the second requirement of 3.1, the tasks to be carried out are somewhat complex. The objective is to describe accurately the neutron field to which the pressure vessel itself will be exposed over its service life. This description of the neutron field must include spatial gradients within the vessel wall. Therefore, heavy emphasis must be placed on the use of neutron transport techniques as well as on the choice of a design basis for the computations. Since a given surveillance capsule measurement, particularly one obtained early in plant life, is not necessarily representative of long-term reactor operation, a simple normalization of neutron transport calculations to dosimetry data from a given capsule may not be appropriate (1-67).3.2 The objectives and requirements of a reactor vessel's support structure's surveillance program are much less stringent, and at present, are limited to physics-dosimetry measurements through ex-vessel cavity monitoring coupled with the use of available test reactor metallurgical data to determine the condition of any support structure steels that might be subject to neutron induced property changes (1, 29, 44-58, 65-70).1.1 This practice covers the methodology, summarized in Annex A1, to be used in the analysis and interpretation of neutron exposure data obtained from LWR pressure vessel surveillance programs and, based on the results of that analysis, establishes a formalism to be used to evaluate present and future condition of the pressure vessel and its support structures2 (1-74).31.2 This practice relies on, and ties together, the application of several supporting ASTM standard practices, guides, and methods (see Master Matrix E706) (1, 5, 13, 48, 49).2 In order to make this practice at least partially self-contained, a moderate amount of discussion is provided in areas relating to ASTM and other documents. Support subject areas that are discussed include reactor physics calculations, dosimeter selection and analysis, and exposure units.1.3 This practice is restricted to direct applications related to surveillance programs that are established in support of the operation, licensing, and regulation of LWR nuclear power plants. Procedures and data related to the analysis, interpretation, and application of test reactor results are addressed in Practice E1006, Guide E900, and Practice E1035.1.4 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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4.1 Regulatory Requirements—The USA Code of Federal Regulations (10CFR Part 50, Appendix H) requires the implementation of a reactor vessel materials surveillance program for all operating LWRs. Other countries have similar regulations. The purpose of the program is to (1) monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline6 resulting from exposure to neutron irradiation and the thermal environment, and (2) make use of the data obtained from surveillance programs to determine the conditions under which the vessel can be operated with adequate margins of safety throughout its service life. Practice E185, derived mechanical property data, and (r, θ, z) physics-dosimetry data (derived from the calculations and reactor cavity and surveillance capsule measurements (1) using physics-dosimetry standards) can be used together with information in Guide E900 and Refs. 4, 11-18 to provide a relation between property degradation and neutron exposure, commonly called a “trend curve.” To obtain this trend curve at all points in the pressure vessel wall requires that the selected trend curve be used together with the appropriate (r, θ, z) neutron field information derived by use of this guide to accomplish the necessary interpolations and extrapolations in space and time.4.2 Neutron Field Characterization—The tasks required to satisfy the second part of the objective of 4.1 are complex and are summarized in Practice E853. In doing this, it is necessary to describe the neutron field at selected (r, θ, z) points within the pressure vessel wall. The description can be either time dependent or time averaged over the reactor service period of interest. This description can best be obtained by combining neutron transport calculations with plant measurements such as reactor cavity (ex-vessel) and surveillance capsule or RPV cladding (in-vessel) measurements, benchmark irradiations of dosimeter sensor materials, and knowledge of the spatial core power distribution, including the time dependence. Because core power distributions change with time, reactor cavity or surveillance capsule measurements obtained early in plant life may not be representative of long-term reactor operation. Therefore, a simple normalization of neutron transport calculations to dosimetry data from a given capsule is unlikely to give a satisfactory solution to the problem over the full reactor lifetime. Guide E482 and Guide E944 provide detailed information related to the characterization of the neutron field for BWR and PWR power plants.4.3 Fracture Mechanics Analysis—Currently, operating limitations for normal heat up and cool down transients imposed on the reactor pressure vessel are based on the fracture mechanics techniques outlined in the ASME Boiler and Pressure Vessel Code. This code requires the assumption of the presence of a surface flaw of depth equal to one fourth of the pressure vessel thickness. In addition, the fracture mechanics analysis of accident-induced transients (Pressurized Thermal Shock, (PTS)) may involve evaluating the effect of flaws of varying depth within the vessel wall (4). Thus, information is required regarding the distribution of neutron exposure and the corresponding radiation damage within the pressure vessel, both in space and time (4). In this regard, Practice E185 provides guidelines for designing a minimum surveillance program, selecting materials, and evaluating metallurgical specimen test results for BWR and PWR power plants. Practice E2215 covers the evaluation of test specimens and dosimetry from LWR surveillance capsules.4.4 Neutron Spectral Effects and DPA—Analysis of the neutron fields of operating power reactors has shown that the neutron spectral shape changes with radial depth into the pressure vessel wall (2, 3). The ratio of dpa/ϕt (where ϕ is the fast (E > 1.0 MeV) neutron fluence rate and t is the time that the material was exposed to an average fluence rate) changes by factors of the order of 2.0/1.0 in traversing from the inner to the outer radius. Although dpa, since it includes a more detailed modeling of the displacement phenomenon, should theoretically provide a better correlation with property degradation than fluence (E > 1.0 MeV) (1, 19), this topic is still controversial and the available experimental data does not provide clear guidance (19, 20). Thus it is recommended to calculate and report both quantities; see Practice E853 and Practice E693.4.5 In-Vessel Surveillance Programs: 4.5.1 The neutron dosimetry monitors used in reactor vessel surveillance capsules provide measurements of the neutron fluence and fluence rate at single points on the core midplane within the reactor, and near the vessel wall; that is, at the surveillance capsule locations (1). In actual practice, the surveillance capsules may be located within the reactor at an azimuthal position that differs from that associated with the maximum neutron exposure (or that differs from the azimuthal and axial location of the assumed flaw); and at a radial position a few centimeters or more from the flaw and the pressure vessel wall (4, 5). Although the surveillance capsule dosimetry does provide points for normalization of the neutron physics transport calculations, it is still necessary to use analytical methods that provide an accurate representation of the spatial variation (axial, radial and azimuthal) of the neutron fluence (refer to Guide E482). It is also necessary to use other measurements to confirm the spatial distribution of RPV neutron exposure.4.5.2 Given that surveillance capsules are located radially closer to the core than the surface of the RPV, they may be shifted azimuthally away from the peak exposure location in order to limit the magnitude of the surveillance capsule lead factor. The lead factor is defined as the ratio of the fast neutron fluence at the center of the surveillance capsule to the peak fast neutron fluence at the clad–base metal interface of the RPV. One adverse effect of this azimuthal shift away from the peak is that the surveillance capsule dosimetry does not “see” the part of the core that produces the peak exposure of the reactor vessel. As a result, the surveillance capsule is unable to monitor the effect of changes in the core power distribution that are made to reduce the peak RPV neutron exposure. Another adverse effect is that with larger lead factors, the capsules are rapidly exposed to a high neutron fluence. For example, with a lead factor of five, a surveillance capsule will receive an exposure in as little as twelve years that is equivalent to what the reactor pressure vessel peak may see in 60 years of operation. Practices E185 and E2215 suggest not exceeding twice the maximum design fluence (MDF) or twice the end-of-license fluence (EOLF). In this example, this would require withdrawing any remaining surveillance capsules after 24 years of operation. Thus, without taking other steps, the reactor would be operated for the remaining 36 years (of a 60 year life) with no dosimetry present.4.5.3 New or replacement surveillance capsules should recognize and correct operating deficiencies by using improved capsule dosimetry. For example, for one class of PWR, the copper wire is cadmium shielded to minimize interference from trace amounts of cobalt. In about one third of the measurements the copper has become incorporated into the cadmium preventing separation and further processing. A simple solution to this problem is to use stainless steel hypodermic tubing to contain and separate the radiometric monitor wire inside the cadmium tubing. Example dimensions include: Typical radiometric monitor wire outside diameter = 0.020 in. (0.5 mm). Typical 19 gauge stainless steel tubing is 0.042 in. outside diameter by 0.027 in. inside diameter, 0.008 in. wall thickness. Typical cadmium tubing is 0.090 in. outside diameter by 0.050 in. inside diameter, 0.020 in. wall thickness.4.5.4 Guide E844 states that radionuclides with half-lives that are short compared to the irradiation duration should not be used. For one class of BWR reactor, the surveillance capsule dosimetry is minimal; consisting of an iron wire and a copper wire (sometimes also a nickel wire). This dosimetry is not suitable for longer irradiations as the “memory” of the activation products is too short to measure the accumulated fluence. For example, for the iron (n,p) activation product, 54Mn, the half-life is 312 d. For the copper (n,α) activation product, 60Co, the half-life is 5.27 a. After three half-lives the remaining activity is on the same order as the counting statistics. The result is that the iron wire has “forgotten” everything that has happened more than two cycles ago and the copper wire has forgotten everything that has happened more than eight cycles ago. This assumes 24-month-long fuel cycles. The copper (n,α) reaction is induced by high energy neutrons and that at a BWR surveillance capsule position only 1 % to 3 % of the fast (E > 1.0 MeV) neutrons are of high enough energy. This limits the value of the copper wire as a neutron fluence monitor. In order to monitor the neutron exposure of the RPV other dosimetry is needed. Installation of ex-vessel neutron dosimetry is the most reasonable and cost-effective option.4.5.5 The neutron fluence calculation on the RPV inner surface can be further verified by means of analyzing small samples of the irradiated stainless steel RPV cladding. Analyzing RPV cladding samples has been a well-established practice for over 30 years (21-36). During the reactor shut down periods, small samples (50 mg to 100 mg) can be machined from the RPV cladding. For retrospective dosimetry purposes the measured 54Mn, 58Co, and 93mNb activities are used. Because of its long half-life, 93mNb is especially useful for integrating fluence over time periods where accurate neutron transport calculations are not available. With sample locations properly selected, the fast neutron fluence distribution and its maximum on the RPV inner surface can be determined. By comparison of these data to the dosimetry data of the surveillance capsules, the lead factor at the time of measurement can also be obtained. This technique works best if the cladding material is one of the niobium-stabilized stainless steels. Type 347 with 0.7 % niobium is one example. Retrospective dosimetry has been successfully demonstrated for ordinary Type 304 stainless steel cladding with only a trace (~50 ppm) of niobium (35). It is important that the cladding surface is first polished to remove radioactive corrosion products before the sample is machined otherwise competing activity may compromise the sample. The tooling used to take these samples needs to be accurately located relative to reactor landmarks in order to know the actual axial and azimuthal locations of the samples. A reasonable accuracy target is ±25 mm axially and azimuthally. The effect of the sampling position error can be estimated by examining the spatial fast neutron fluence rate gradient in the vicinity of the sample point. In general, in the areas where the fast neutron fluence is the greatest, the gradient tends to be very small; approaching flat in the case of the axial distribution opposite the middle of the core. At extreme axial positions, well beyond the ends of the core, the gradient is steep. There the positioning error could lead to an estimated fluence error of ±20 %. A similar discussion applies to the azimuthal fluence rate gradients. The tooling also needs to be designed to completely retain all machined cladding chips and to prevent cross-contamination from one sample to another. Access to the full extent of azimuthal and axial clad samples is generally limited to PWRs due to the extensive structure (jet pumps, etc.) blocking general access to the RPV cladding of many BWRs. It may be possible to take a more limited set of samples from the cladding of a BWR RPV.4.5.6 The design and manufacture of new reactor pressure vessels should consider using one of the stainless steels or Inconel alloys that contains niobium for the purpose of cladding the inner surface of the vessel. This would result in a designed-in retrospective dosimetry system that would capture neutron exposure data from reactor startup.4.6 Ex-Vessel Surveillance Program: 4.6.1 Ex-vessel neutron dosimetry (EVND) has also been in wide scale application in nuclear reactors for over 30 years (28, 29, 31, 33, 35, 37-97). The main advantages of EVND are the relative simplicity and the relatively low cost of the dosimetry system. Removal and replacement of irradiated dosimetry takes little time. Typical installations have dosimetry that spans the active core height and continues to cover the extended beltline region of the RPV. Installation of dosimetry at multiple angles allows full octant coverage (for octant symmetric cores). Some EVND installations include multiple measurements at symmetric azimuthal angles to confirm symmetry in the azimuthal fluence rate distributions. Asymmetries may result from such things as non-symmetric core power distributions, differences in water temperatures from one loop to another, or ovality in the as-built dimensions for the reactor internals or RPV. Dosimetry capsules typically contain a full complement of radiometric monitors (refer to Guide E844) to ensure good spectral coverage and fluence integration. Typically, capsules are connected and supported by stainless steel wires or chains, which are, in turn, segmented and counted to provide axial gradient information.4.6.2 In order to minimize measurement field perturbation, the dosimeter capsules should be made of a neutron-transparent material such as aluminum. This also serves to reduce the radiation dose rates encountered when removing and replacing dosimetry. The gradient chains or wires should be a low mass per linear foot material, again to reduce the dose rates encountered during handling of irradiated dosimetry.4.6.3 An ex-vessel neutron dosimetry system needs to be accurately located with respect to well-known and easily verified reactor features. A reasonable accuracy target is ±25 mm axially and azimuthally. The effect of the dosimetry position error can be estimated by examining the spatial fast neutron fluence rate gradient in the vicinity of the measurement point. In general, in the areas where the fast neutron fluence is the greatest, the gradient tends to be very small; approaching flat in the case of the axial distribution opposite the middle of the core. At extreme axial positions, well beyond the ends of the core, the gradient is steep. There the positioning error could lead to an estimated fluence error of ±20 %. A similar discussion applies to the azimuthal fluence rate gradients.4.6.4 Ideally, the ex-vessel neutron dosimetry is installed before reactor startup so that it can provide data over the operating lifetime of the reactor. It is recommended that the ex-vessel neutron dosimetry be analyzed before and after significant plant modifications that would alter the neutron exposure of the reactor vessel. Some examples include switching from low-leakage core loading patterns back to out-in loading patterns (or vice versa), performing a significant (>10 %) uprating of the plant power, adding (or removing) core flux suppression absorbers or dummy fuel rods, or modifying the reactor internals geometry. The typical dosimetry replacement interval is between one and five 18-month-long fuel cycles (or equivalent intervals for other fuel cycle lengths).4.6.5 Periodic measurements (either RPV cladding samples or EVND) serve to confirm neutron fluence projections and help to avoid problems that result from errors in reactor-specific calculational models (98).4.6.6 Calculations of neutron fields in commercial reactors show that the neutron exposure (dpa) at the inner diameter of the pressure vessel can vary by a factor of three or more as a function of azimuthal position (2, 3). Dosimetry monitors in the reactor cavity outside the reactor pressure vessel are a useful tool, therefore, in determining the accuracy of the neutron field calculations at points inside the pressure vessel wall. Practice E853 recommends the use of ex-vessel reactor cavity neutron dosimetry measurements for verification of the physics transport calculations. The status of benchmark field and power reactor applications as well as studies of this approach are discussed in Refs. 1, 18, 19, 37-40, 99-112.1.1 This guide establishes the means and frequency of monitoring the neutron exposure of the LWR reactor pressure vessel throughout its operating life.1.2 The physics-dosimetry relationships determined from this guide may be used to estimate reactor pressure vessel damage through the application of Practice E693 and Guide E900, using fast neutron fluence (E > 1.0 MeV and E > 0.1 MeV), displacements per atom (dpa), or damage-function-correlated exposure parameters as independent exposure variables. Supporting the application of these standards are the E853, E944, E1005, and E1018 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, 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 The SSTR method provides for the measurement of absolute-fission density per unit mass. Absolute-neutron fluence can then be inferred from these SSTR-based absolute fission rate observations if an appropriate neutron spectrum average fission cross section is known. This method is highly discriminatory against other components of the in-core radiation field. Gamma rays, beta rays, and other lightly ionizing particles do not produce observable tracks in appropriate LWR SSTR candidate materials. However, photofission can contribute to the observed fission track density and should therefore be accounted for when nonnegligible. For a more detailed discussion of photofission effects, see 14.4.4.2 In this test method, SSTRs are placed in surface contact with fissionable deposits and record neutron-induced fission fragments. By variation of the surface mass density (μg/cm 2) of the fissionable deposit as well as employing the allowable range of track densities (from roughly 1 event/cm2 up to 105 events/cm2 for manual scanning), a range of total fluence sensitivity covering at least 16 orders of magnitude is possible, from roughly 102 n/cm 2 up to 5 × 10 18 n/cm2. The allowable range of fission track densities is broader than the track density range for high accuracy manual scanning work with optical microscopy cited in 1.2. In particular, automated and semi-automated methods exist that broaden the customary track density range available with manual optical microscopy. In this broader track density region, effects of reduced counting statistics at very low track densities and track pile-up corrections at very high track densities can present inherent limitations for work of high accuracy. Automated scanning techniques are described in Section 11.4.3 For dosimetry applications, different energy regions of the neutron spectrum can be selectively emphasized by changing the nuclide used for the fission deposit.4.4 It is possible to use SSTRs directly for neutron dosimetry as described in 4.1 or to obtain a composite neutron detection efficiency by exposure in a benchmark neutron field. The fluence and spectrum-averaged cross section in this benchmark field must be known. Furthermore, application in other neutron fields may require adjustments due to spectral deviation from the benchmark field spectrum used for calibration. In any event, it must be stressed that the SSTR-fission density measurements can be carried out completely independent of any cross-section standards (6). Therefore, for certain applications, the independent nature of this test method should not be compromised. On the other hand, many practical applications exist wherein this factor is of no consequence so that benchmark field calibration would be entirely appropriate.1.1 This test method describes the use of solid-state track recorders (SSTRs) for neutron dosimetry in light-water reactor (LWR) applications. These applications extend from low neutron fluence to high neutron fluence, including high power pressure vessel surveillance and test reactor irradiations as well as low power benchmark field measurement. (1)2 Special attention is given to the use of state-of-the-art manual and automated track counting methods to attain high absolute accuracies. In-situ dosimetry in actual high fluence-high temperature LWR applications is emphasized.1.2 This test method includes SSTR analysis by both manual and automated methods. To attain a desired accuracy, the track scanning method selected places limits on the allowable track density. Typically, good results are obtained in the range of 5 to 800 000 tracks/cm2 and accurate results at higher track densities have been demonstrated for some cases. (2) Track density and other factors place limits on the applicability of the SSTR method at high fluences. Special care must be exerted when measuring neutron fluences (E>1MeV) above 1016 n/cm2 (3) .1.3 Low fluence and high fluence limitations exist. These limitations are discussed in detail in Sections 13 and 14 and in Refs (3-5).1.4 SSTR observations provide time-integrated reaction rates. Therefore, SSTRs are truly passive-fluence detectors. They provide permanent records of dosimetry experiments without the need for time-dependent corrections, such as decay factors that arise with radiometric monitors.1.5 Since SSTRs provide a spatial record of the time-integrated reaction rate at a microscopic level, they can be used for “fine-structure” measurements. For example, spatial distributions of isotopic fission rates can be obtained at very high resolution with SSTRs.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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4.1 Master Matrix—This matrix document is written as a reference and guide to the use of existing standards and to help manage the development and application of new standards, as needed for LWR-PV surveillance programs. Paragraphs 4.2 – 4.5 are provided to assist the authors and users involved in the preparation, revision, and application of these standards (see Section 6).4.2 Approach and Primary Objectives: 4.2.1 Standardized procedures and reference data are recommended in regard to (1) neutron and gamma dosimetry, (2) physics (neutronics and gamma effects), and (3) metallurgical damage correlation methods and data, associated with the analysis, interpretation, and use of nuclear reactor test and surveillance results.4.2.2 Existing state-of-the-art practices associated with (1), (2), and (3), if uniformly and consistently applied, can provide reliable (10 to 30 %, 1σ) estimates of changes in LWR-PV steel fracture toughness during a reactor’s service life (36).4.2.3 Reg. Guide 1.99 and Section III of the ASME Boiler and Pressure Vessel Code, Part NF2121 require that the materials used in reactor pressure vessels support “...shall be made of materials that are not injuriously affected by ... irradiation conditions to which the item will be subjected.”4.2.4 By the use of this series of standards and the uniform and consistent documentation and reporting of estimated changes in LWR-PV steel fracture toughness with uncertainties of 10 to 30 % (1σ), the nuclear industry and licensing and regulatory agencies can meet realistic LWR power plant operating conditions and limits, such as those defined in Appendices G and H of 10 CFR Part 50, Reg. Guide 1.99, and the ASME Boiler and Pressure Vessel Code.4.2.5 The uniform and consistent application of this series of standards allows the nuclear industry and licensing and regulatory agencies to properly administer their responsibilities in regard to the toughness of LWR power reactor materials to meet requirements of Appendices G and H of 10 CFR Part 50, Reg. Guide 1.99, and the ASME Boiler and Pressure Vessel Code.4.3 Dosimetry Analysis and Interpretation (1, 3-5, 21, 28, 29, 35, 37, 38)—When properly implemented, validated, and calibrated by vendor/utility groups, state-of-the-art dosimetry practices exist that are adequate for existing and future LWR power plant surveillance programs. The uncertainties and errors associated with the individual and combined effects of the different variables (items 1.4.1 – 1.4.10 of 1.4) are considered in this section and in 4.4 and 4.5. In these sections, the accuracy (uncertainty and error) statements that are made are quantitative and representative of state-of-the-art technology. Their correctness and use for making EOL predictions for any given LWR power plant, however, are dependent on such factors as (1) the existing plant surveillance program, (2) the plant geometrical configuration, and (3) available surveillance results from similar plants. As emphasized in Section III-A of Ref (7), however, these effects are not unique and are dependent on (1) the surveillance capsule design, (2) the configuration of the reactor core and internals, and (3) the location of the surveillance capsule within the reactor geometry. Further, the statement that a result could be in error is dependent on how the neutron and gamma ray fields are estimated for a given reactor power plant (1, 11, 28, 36, 39, 40). For most of the error statements in 4.3 – 4.5, it is assumed that these estimates are based on reactor transport theory calculations that have been normalized to the core power history (see 4.4.1.2) and not to surveillance capsule dosimetry results. The 4.3 – 4.5 accuracy statements, consequently, are intended for use in helping the standards writer and user to determine the relative importance of the different variables in regard to the application of the set of ASTM standards, Fig. 1, for (1) LWR-PV surveillance program, (2) as instruments of licensing and regulation, and (3) for establishing improved metallurgical databases.4.3.1 Required Accuracies and Benchmark Field Referencing: 4.3.1.1 The accuracies (uncertainties and errors) (Note 1) desirable for LWR-PV steel exposure definition are of the order of ±10 to 15 % (1σ) while exposure accuracies in establishing trend curves should preferably not exceed ±10 % (1σ) (1, 11, 21, 36, 40-46). In order to achieve such goals, the response of neutron dosimeters should therefore also be interpretable to accuracies within ±10 to 15 % (1σ) in terms of exposure units and be measurable to within ±5 % (1σ).NOTE 1: Uncertainty in the sense treated here is a scientific characterization of the reliability of a measurement result and its statement is the necessary premise for using these results for applied investigations claiming high or at least stated accuracy. The term error will be reserved to denote a known deviation of the result from the quantity to be measured. Errors are usually taken into account by corrections.4.3.1.2 Dosimetry “inventories” should be established in support of the above for use by vendor/utility groups and research and development organizations.4.3.1.3 Benchmark field referencing of research and utilities’ vendor/service laboratories has been completed that is:(1) Needed for quality control and certification of current and improved dosimetry practices; and(2) Extensively applied in standard and reference neutron fields, PCA, PSF, SDMF, VENUS, NESDIP, PWRs, BWRs (1), and a number of test reactors to quantify and minimize uncertainties and errors.4.3.2 Status of Benchmark Field Referencing Work for Dosimetry Detectors—PCA, VENUS, NESDIP experiments with and without simulated surveillance capsules and power reactor tests have provided data for studying the effect of deficiencies in analysis and interpretations; the PCA/PSF/SDMF perturbation experiments have provided data for more realistic PWR and BWR power plant surveillance capsule configurations and have permitted utilities’ vendor/service laboratories to test, validate, calibrate, and update their practices (1, 4, 5, 47). The PSF surveillance capsule test provided data, but of a more one-dimensional nature. PCA, VENUS, and NESDIP experimentation together with some test reactor work augmented the benchmark field quantification of these effects (1, 3, 4, 28, 36, 48-51).4.3.3 Additional Validation Work for Dosimetry Detectors: 4.3.3.1 Establishment of nuclear data, photo-reaction cross sections, and neutron damage reference files.4.3.3.2 Establishment of proper quality assurance procedures for sensor set designs and individual detectors.4.3.3.3 Interlaboratory comparisons using standard and reference neutron fields and other test reactors that provide adequate validations and calibrations, see Guide E2005.4.4 Reactor Physics Analysis and Interpretation (1, 3, 5, 11, 28, 35, 36, 39, 52)—When properly implemented, validated, and calibrated by vendor/utility groups, state-of-the-art reactor physics practices exist that are adequate for in- and ex-vessel estimates of PV-steel changes in fracture toughness for existing and future power plant surveillance programs.4.4.1 Required Accuracies and Benchmark Field Referencing: 4.4.1.1 The accuracies desirable for LWR-PV steel (surveillance capsule specimens and vessels) exposure definition are of the order of ±10 to 15 % (1σ). Under ideal conditions benchmarking computational techniques are capable of predicting absolute in- and ex-vessel neutron exposures and reaction rates per unit reactor core power to within ±15 % (but generally not to within ±5 %). The accuracy will be worse, however, in applications to actual power plants because of geometrical and other complexities (1, 3, 4, 11, 21, 36-39, 52).4.4.1.2 Calculated in-and ex-vessel neutron and gamma ray fields can be normalized to the core power history or to experimental measurements. The latter may include dosimetry from surveillance capsules, other in-vessel locations, or ex-vessel measurements in the cavity outside the vessel. In each case, the uncertainty arising from the calculation needs to be considered.4.4.2 Power Plant Reactor Physics Analysis and Interpretation: 4.4.2.1 Result of Neglect of Benchmarking—One quarter thickness location (1/4T) vessel wall estimates of damage exposure are not easily compared with experimental results. “Lead factors,” based on the different ways they can be calculated (fluence >0.1 or >1.0 MeV and dpa) may not always be conservative; that is, some surveillance capsules have been positioned in-vessel such that the actual lead factor is very near unity—no lead at all. Also the differences between lead factors based on fluence E > 0.1 or > 1 MeV and dpa can be significant, perhaps 50 % or more (1, 11, 21, 28, 36-38, 52).4.4.3 PCA, VENUS, and NESDIP Experiments and PCA Blind Test: 4.4.3.1 Test of transport theory methods under clean geometry and clean core source conditions shall be made (1, 4, 11, 52).4.4.3.2 This is a necessary but not sufficient benchmark test of the adequacy of a vendor/utility group’s power reactor physics computational tools.4.4.3.3 The standard recommendation should be that the vendor/utility group’s observed differences between their own calculated and the PCA, VENUS, and NESDIP measured integral and differential exposure and reaction rate parameters be used to validate and improve their calculational tools (if the differences fall outside the PCA, VENUS, and NESDIP experimental accuracy limits).4.4.4 PWR and BWR Generic Power Reactor Tests: 4.4.4.1 Test of transport theory methods under actual geometry and variable core source conditions (1, 3, 4, 28, 35, 36, 53).4.4.4.2 This is a necessary and partly sufficient benchmark test of the adequacy of a vendor/utility group’s power reactor physics computational tools.4.4.4.3 The standard recommendation should be that the vendor/utility group’s observed differences between their own calculated and the selected PWR or BWR measured integral and differential exposure and reaction rate parameters be used to validate and improve their calculation tools (if the differences fall outside of the selected PWR or BWR experimental accuracy limits).4.4.4.4 The power reactor “benchmarks” that have been established for this purpose are identified and discussed in Refs (1, 3, 4, 35, 53) and their references and in Guide E2006.4.4.5 Operating Power Reactor Tests: 4.4.5.1 This is a necessary test of transport theory methods under actual geometry and variable core source conditions, but for a particular type or class of vendor/utility group power reactors. Here, actual in-vessel surveillance capsule and any required ex-vessel measured dosimetry information will be utilized as in 4.4.4 (1, 3, 4, 28, 35, 36, 53). Note, however, that operating power reactor tests are not sufficient by themselves (Reg. Guide 1.190, Section 4.4.5.1).4.4.5.2 Accuracies associated with surveillance program reported values of exposures and reaction rates are expected to be in the 10 to 30 % (1σ) range (36).4.5 Metallurgical Damage Correlation Analysis and Interpretation (1-8, 10, 11, 13, 15-29, 36-38)—When properly implemented, validated, and calibrated by vendor/utility groups, state-of-the-art metallurgical damage correlation practices exist that are adequate for in- and ex-vessel estimates of PV-steel changes in fracture toughness for existing and future power plant surveillance programs.4.5.1 Required Accuracies and Benchmark Field Referencing: 4.5.1.1 The accuracies desirable and achievable for LWR-PV steel (test reactor specimens, surveillance capsule specimens, and vessels and support structure) data correlation and data extrapolation (to predict fracture toughness changes both in space and time) are of the order of ±10 to 30 % (1σ). In order to achieve such a goal, however, the metallurgical parameters (ΔNDTT, upper shelf, yield strength, etc.) must be interpretable to well within ±20 to 30 % (1σ). This mandates that in addition to the dosimetry and physics variables already discussed that the individual uncertainties and errors associated with a number of other variables (neutron dose rate, neutron spectrum, irradiation temperature, steel chemical composition, and microstructure) must be minimized and results must be interpretable to within the same ±10 to 30 % (1σ) range.4.5.1.2 Advanced sensor sets (including dosimetry, temperature and damage correlation sensors) and practices have been established in support of the above for use by vendor/utility groups (1, 4, 5, 11, 39, 50, 54, 55).4.5.1.3 Benchmark field referencing of utilities' vendor/service laboratories, as well as advanced practices, is in progress or being planned that is (1, 3-6, 28, 50, 54-56):(1) Needed for validation of data correlation procedures and time and space extrapolations (to PV positions: surface, 1/4 T, etc.) of test reactor and power reactor surveillance capsule metallurgical and neutron exposure data.(2) Being or will be tested in test reactor neutron fields to quantify and minimize uncertainties and errors (included here is the use of damage correlation materials—steel, sapphire, etc.).4.5.2 Benchmark Field Referencing—The PSF (all positions: surveillance, surface, 1/4T, 1/2T, and the void box) together with the Melusine PV-simulator and other tests, such as for thermal neutron effects, provide needed validation data on all variables—dosimetry, physics, and metallurgy (1, 4, 10, 19, 21, 22, 37, 38). Other test reactor data comes from surveillance capsule results that have been benchmarked by vendor/service laboratory/utility groups (1, 3, 4, 6, 11, 18, 27, 28, 36, 40-44, 47).4.5.3 Reg. Guide 1.99, NRC, EPRI Databases—NRC and Electric Power Research Institute (EPRI) databases have been studied on an ongoing basis by ASTM Subcommittees E10.02 and E10.05, vendors, utilities, EPRI, and NRC contractors to establish improved databases for existing test and power reactor measured property change data. ASTM task groups recommend the use of updated and new exposure units (fluence total >0.1, >1.0 MeV, and dpa) for the NRC and EPRI databases (1, 2, 6, 7, 13, 18, 27, 36, 40-44, 47), and incorporate these recommendations in the appropriate standards. ASTM subcommittee E10.02 has updated the embrittlement database and the prediction model in E900–15. The exposure unit used is total fluence for E > 1 MeV. The basis of the prediction model is documented in an adjunct associated with E900, available from ASTM.4 The success of this effort depends on good cooperation between research and individual service laboratories and vendor/utility groups. An ASTM dosimetry cross section file based on the latest evaluations, as detailed in Guide E1018, and incorporating corrections for all known variables (perturbations, photo-reactions, spectrum, burn-in, yields, fluence time history, etc.) will be used as required and justified. Test reactor data will be addressed in a similar manner, as appropriate.1.1 This master matrix standard describes a series of standard practices, guides, and methods for the prediction of neutron-induced changes in light-water reactor (LWR) pressure vessel (PV) and support structure steels throughout a pressure vessel’s service life (Fig. 1). Referenced documents are listed in Section 2. The summary information that is provided in Sections 3 and 4 is essential for establishing proper understanding and communications between the writers and users of this set of matrix standards. It was extracted from the referenced standards (Section 2) and references for use by individual writers and users. More detailed writers’ and users’ information, justification, and specific requirements for the individual practices, guides, and methods are provided in Sections 3 – 5. General requirements of content and consistency are discussed in Section 6.FIG. 1 Organization and Use of ASTM Standards in the E706 Master Matrix1.2 This master matrix is intended as a reference and guide to the preparation, revision, and use of standards in the series.1.3 To account for neutron radiation damage in setting pressure-temperature limits and making fracture analyses ((1-12)2 and Guide E509), neutron-induced changes in reactor pressure vessel steel fracture toughness must be predicted, then checked by extrapolation of surveillance program data during a vessel’s service life. Uncertainties in the predicting methodology can be significant. Techniques, variables, and uncertainties associated with the physical measurements of PV and support structure steel property changes are not considered in this master matrix, but elsewhere ((2, 6, 7, 11-26) and Guide E509).1.4 The techniques, variables, and uncertainties related to (1) neutron and gamma dosimetry, (2) physics (neutronics and gamma effects), and (3) metallurgical damage correlation procedures and data are addressed in separate standards belonging to this master matrix (1, 17). The main variables of concern to (1), (2), and (3) are as follows:1.4.1 Steel chemical composition and microstructure,1.4.2 Steel irradiation temperature,1.4.3 Power plant configurations and dimensions, from the core periphery to surveillance positions and into the vessel and cavity walls,1.4.4 Core power distribution,1.4.5 Reactor operating history,1.4.6 Reactor physics computations,1.4.7 Selection of neutron exposure units,1.4.8 Dosimetry measurements,1.4.9 Neutron special effects, and1.4.10 Neutron dose rate effects.1.5 A number of methods and standards exist for ensuring the adequacy of fracture control of reactor pressure vessel belt lines under normal and accident loads ((1, 7, 8, 11, 12, 14, 16, 17, 23-27), Referenced Documents: ASTM Standards (2.1), Nuclear Regulatory Documents (2.3) and ASME Standards (2.4)). As older LWR pressure vessels become more highly irradiated, the predictive capability for changes in toughness must improve. Since during a vessel's service life an increasing amount of information will be available from test reactor and power reactor surveillance programs, procedures to evaluate and use this information must be used (1, 2, 4-9, 11, 12, 23-26, 28). This master matrix defines the current (1) scope, (2) areas of application, and (3) general grouping for the series of ASTM standards, as shown in Fig. 1.1.6 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.1.7 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.1.8 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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5.1 This test method provides a description of the design of the Stirred Reactor Coupon Analysis (SRCA) apparatus and identifies aspects of the performance of the SRCA tests and interpretation of the test results that must be addressed by the experimenter to provide confidence in the measured dissolution rate.5.2 The SRCA methods described in this test method can be used to characterize several aspects of glass corrosion that can be included in mechanistic models of long-term durability of glasses, including nuclear waste glasses.5.3 Depending on the test parameters investigated, the SRCA results can be used to measure the intrinsic dilute glass dissolution rate, as well as the effects of conditions such as temperature, pH, and solution chemistry on the dissolution rate.5.4 Due to the scalable nature of the method, it is particularly applicable to studies of the impact of glass composition on dilute-condition corrosion. Models of glass behavior can be parameterized by testing glass composition matrices and establishing quantitative structure-property relationships.5.5 The step heights present on the corroded sample can be measured by a variety of techniques including profilometry (optical or stylus), atomic force microscopy, interferometry or other techniques capable of determining relative depths on a sample surface. The sample can also be interrogated with other techniques such as scanning electron microscopy to characterize the corrosion behavior. These further analyses can determine if the sample corroded homogenously and possible formation of secondary phases or leached layers. Occurrence of these features may impact the accuracy of glass dissolution. This test method does not address these solid-state characterizations.1.1 This test method describes a test method in which the dissolution rate of a homogenous silicate glass is measured through corrosion of monolithic samples in stirred dilute conditions.1.2 Although the test method was designed for simulated nuclear waste glass compositions per Guide C1174, the method is applicable to glass compositions for other applications including, but not limited to, display glass, pharmaceutical glass, bioglass, and container glass compositions.1.3 Various test solutions can be used at temperatures less than 100 °C. While the durability of the glass can be impacted by dissolving species from the glass, and thus the test can be conducted in dilute conditions or concentrated condition to determine the impact of such species, care must be taken to avoid, acknowledge, or account for the production of alteration layers which may confound the step height measurements.1.4 The dissolution rate measured by this test is, by design, an average of all corrosion that occurs during the test. In dilute conditions, glass is assumed to dissolve congruently and the dissolution rate is assumed to be constant.1.5 Tests are carried out via the placement of the monolithic samples in a large well-mixed volume of solution, achieving a high volume to surface area ratio resulting in dilute conditions with agitation of the solution.1.6 This test method excludes test methods using powdered glass samples, or in which the reactor solution saturates with time. Glass fibers may be used without a mask if the diameter is known to high accuracy before the test.1.7 Tests may be conducted with ASTM Type I water (see Specification D1193 and Terminology D1129), buffered water or other chemical solutions, simulated or actual groundwaters, biofluids, or other dissolving solutions.1.8 Tests are conducted with monolithic glass samples with at least a single flat face. Although having two plane-parallel faces is helpful for certain step height measurements, it is not required. The geometric dimensions of the monolith are not required to be known. The reacted monolithic sample is to be analyzed following the reaction to measure a corroded depth to determine dissolution rate.1.9 Tests may be performed with radioactive samples. However, safety concerns working with radionuclides are not addressed in this test method.1.10 Data from these tests can be used to determine the value of kinetic rate model parameters needed to predict glass corrosion behavior over long periods of time. For an example, see Practice C1662, section 9.5.1.11 This test method must be performed in accordance with all quality assurance requirements for acceptance of the data.1.12 Units—The values stated in SI units are regarded as the standard. Any values given in parentheses are for information only.1.13 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.1.14 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 (5, 7). 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. For example, research has shown that biofilm grown under high shear is more difficult to kill than biofilm grown under low shear (5, 8). The purpose of this test method is to direct a user in the laboratory study of a Pseudomonas aeruginosa biofilm by clearly defining each system parameter. This test method will enable an investigator to grow, sample, and analyze a Pseudomonas aeruginosa biofilm grown under high shear. The biofilm generated in the CDC Biofilm Reactor is also suitable for efficacy testing. After the 48 h growth phase is complete, the user may add the treatment in situ or remove the coupons and treat them individually.1.1 This test method specifies the operational parameters required to grow a reproducible (1)2 Pseudomonas aeruginosa ATCC 700888 biofilm under high shear. The resulting biofilm is representative of generalized situations where biofilm exists under high shear rather than being representative of one particular environment.1.2 This test method uses the Centers for Disease Control and Prevention (CDC) Biofilm Reactor. The CDC Biofilm Reactor is a continuously stirred tank reactor (CSTR) with high wall shear. Although it was originally designed to model a potable water system for the evaluation of Legionella pneumophila (2), the reactor is versatile and may also be used for growing and/or characterizing biofilm of varying species (3-5).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.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 and health 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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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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