1、Designation: E2006 16Standard Guide forBenchmark Testing of Light Water Reactor Calculations1This standard is issued under the fixed designation E2006; the number immediately following the designation indicates the year oforiginal adoption or, in the case of revision, the year of last revision. A nu
2、mber in parentheses indicates the year of last reapproval. Asuperscript epsilon () indicates an editorial change since the last revision or reapproval.1. Scope1.1 This guide covers general approaches for benchmarkingneutron transport calculations for pressure vessel surveillanceprograms in light wat
3、er reactor systems. A companion guide(Guide E2005) covers use of benchmark fields for testingneutron transport calculations and cross sections in wellcontrolled environments. This guide covers experimentalbenchmarking of neutron fluence calculations (or calculationsof other exposure parameters such
4、as dpa) in more complexgeometries relevant to reactor pressure vessel surveillance.Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indica-tion of the accuracy of the calculational methods and nucleardata when applied to typical cases; an
5、d the use of plant specificmeasurements to indicate bias in individual plant calculations.Use of these two benchmark techniques will serve to limitplant-specific calculational uncertainty, and, when combinedwith analytical uncertainty estimates for the calculations, willprovide uncertainty estimates
6、 for reactor fluences with a higherdegree of confidence.1.2 This standard does not purport to address all of thesafety concerns, if any, associated with its use. It is theresponsibility of the user of this standard to establish appro-priate safety and health practices and determine the applica-bilit
7、y of regulatory limitations prior to use.2. Referenced Documents2.1 ASTM Standards:2E261 Practice for Determining Neutron Fluence, FluenceRate, and Spectra by Radioactivation TechniquesE262 Test Method for Determining Thermal Neutron Reac-tion Rates and Thermal Neutron Fluence Rates by Radio-activat
8、ion TechniquesE706 Master Matrix for Light-Water Reactor Pressure VesselSurveillance Standards, E 706(0) (Withdrawn 2011)3E844 Guide for Sensor Set Design and Irradiation forReactor Surveillance, E 706 (IIC)E944 Guide for Application of Neutron Spectrum Adjust-ment Methods in Reactor Surveillance, E
9、 706 (IIA)E1018 Guide for Application of ASTM Evaluated CrossSection Data File, Matrix E706 (IIB)E2005 Guide for Benchmark Testing of Reactor Dosimetryin Standard and Reference Neutron Fields3. Significance and Use3.1 This guide deals with the difficult problem of bench-marking neutron transport cal
10、culations carried out to determinefluences for plant specific reactor geometries. The calculationsare necessary for fluence determination in locations importantfor material radiation damage estimation and which are notaccessible to measurement. The most important application ofsuch calculations is t
11、he estimation of fluence within the reactorvessel of operating power plants to provide accurate estimatesof the irradiation embrittlement of the base and weld metal inthe vessel. The benchmark procedure must not only prove thatcalculations give reasonable results but that their uncertaintiesare prop
12、agated with due regard to the sensitivities of thedifferent input parameters used in the transport calculations.Benchmarking is achieved by building up data bases ofbenchmark experiments that have different influences on un-certainty propagation. For example, fission spectra are thefundamental data
13、bases which control propagation of crosssection uncertainties, while such physics-dosimetry experi-ments as vessel wall mockups, where measurements are madewithin a simulated reactor vessel wall, control error propaga-tion associated with geometrical and methods approximationsin the transport calcul
14、ations. This guide describes generalprocedures for using neutron fields with known characteristicsto corroborate the calculational methodology and nuclear dataused to derive neutron field information from measurements ofneutron sensor response.3.2 The bases for benchmark field referencing are usuall
15、yirradiations performed in standard neutron fields with well-known energy spectra and intensities. There are, however, less1This test method is under the jurisdiction of ASTM Committee E10 on NuclearTechnology and Applicationsand is the direct responsibility of SubcommitteeE10.05 on Nuclear Radiatio
16、n Metrology.Current edition approved June 1, 2016. Published July 2016. Originally approvedin 1999. Last previous edition approved in 2010 as E2006 10. DOI: 10.1520/E2006-16.2For referenced ASTM standards, visit the ASTM website, www.astm.org, orcontact ASTM Customer Service at serviceastm.org. For
17、Annual Book of ASTMStandards volume information, refer to the standards Document Summary page onthe ASTM website.3The last approved version of this historical standard is referenced onwww.astm.org.Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. Uni
18、ted States1well known neutron fields that have been designed to mockupspecial environments, such as pressure vessel mockups inwhich it is possible to make dosimetry measurements inside ofthe steel volume of the “vessel”. When such mockups aresuitably characterized they are also referred to as benchm
19、arkfields. A benchmark is that against which other things arereferenced, hence the terminology “to benchmark reference” or“benchmark referencing”. A variety of benchmark neutronfields, other than standard neutron fields, have been developed,or pressed into service, to improve the accuracy of neutron
20、dosimetry measurement techniques. Some of these specialbenchmark experiments are discussed in this standard becausethey have identified needs for additional benchmarking orbecause they have been sufficiently documented to serve asbenchmarks.3.3 One dedicated effort to provide benchmarks whoseradiati
21、on environments closely resemble those found outsidethe core of an operating reactor was the Nuclear RegulatoryCommissions Light Water Reactor Pressure Vessel Surveil-lance Dosimetry Improvement Program (LWR-PV-SDIP) (1)4.This program promoted better monitoring of the radiationexposure of reactor ve
22、ssels and, thereby, provided for betterassessment of vessel end-of-life conditions.An objective of theLWR-PV-SDIP was to develop improved procedures for reac-tor surveillance and document them in a series of ASTMstandards (see Matrix E706). The primary means chosen forvalidating LWR-PV-SDIP procedur
23、es was by benchmarking aseries of experimental and analytical studies in a variety offields (see Guide E2005).4. Particulars of Benchmarking Transport Calculations4.1 Benchmarking of neutron transport calculations in-volves several distinct steps that are detailed below.4.1.1 Nuclear data used for t
24、ransport calculations are evalu-ated using differential data or a combination of integral anddifferential data. This process results in a library of crosssections and other needed nuclear data (including fissionspectra) that, in the opinion of the evaluator, gives the best fitto the available experi
25、mental and theoretical results. Some ofinformation used in evaluating the cross sections may be thesame as that used directly for benchmarking transport calcula-tions for LWR systems (see 4.1.2). The cross section bench-marking itself is not addressed in this standard. It is assumedthat the cross-se
26、ction set is derived in this fashion to beapplicable to a variety of calculational geometries and may notgive the most accurate answer for LWR geometries. Thusfurther benchmarking in LWR geometries is required.4.1.2 Transport calculations in LWR geometries may bebenchmarked using measurements made i
27、n well-defined andwell-characterized facilities that each mock-up part of anLWR-type system. These facilities have the advantage overoperating plants that the dimensions and material compositionscan be more accurately defined, the neutron source can be wellcharacterized, and measurements can be made
28、 in a largenumber of locations that would not be accessible in actualpower systems. In power reactors, one is interested in thetransport of neutrons from the distributed source in the fuel,through the reactor internals and water to the vessel, andthrough the vessel to the reactor cavity. Three mocku
29、ps thattogether encompass this entire transport problem are describedin 5.1. Modeling and calculating of neutron transport in thesevarious geometries can be expected to identify any bias inspecific parts of the calculations. Biases that can be detectedinclude those due to modeling the irregular fuel
30、 geometry anddistributed neutron source, those due to errors in the cross-sections or neutron spectra, and those due to calculationalapproximations.4.1.3 The benchmarking described above does not providechecks on geometries identical to actual plants and does notinclude bias that may exist in the de
31、finition of a specific plantmodel. Identification of these types of bias can only beaccomplished using actual plant measurements. Benchmarkingusing these measurements is described in 5.2 and 5.3.4.1.4 The final aspect of benchmarking is the benchmarkingof the dosimetry results. This aspect is treate
32、d in Guide E2005.It is assumed that the measurements in the benchmarkedfacilities and in the actual operating plants are carried out usingbenchmarked reactions and dosimeters. This involves usingreactions whose cross sections have been shown to be consis-tent with results in these types of neutron e
33、nvironments. Also,the dosimeters and measurement facilities must be of adequatequality and have measurement accuracies that have beenverified (such as through round-robin testing). Periodic recali-bration of laboratory measurement devices is also requiredusing appropriate reference standards.4.1.4.1
34、 The selection and use of dosimeters should beaccording to Guide E844, and evaluation of the dosimetryresults should be in accordance with Practice E261 and TestMethod E262. In particular, to compare measured dosimetryresults with calculated reaction rates or fluences, the followingeffects must be a
35、ccounted for: effects of dosimetryperturbations, position or gradient corrections, gamma attenu-ation in counted foils, differences in counting geometry fromthat of calibration standards, dosimeter or reaction productburnup, effects of competing reactions in impurities andphotofission or photoinduce
36、d reactions, and proper treatmentof the irradiation history.4.1.4.2 The benchmarking of the dosimetry results will alsohave indicated any bias that exists in the dosimetry crosssections. These cross sections are essentially independent ofthe transport cross sections discussed in 4.1.1. Recommendeddo
37、simetry cross sections are given in Guide E1018.4.1.5 The use of the benchmark data to determine bias incalculations and to determine best values for fluence incomplex geometries is not straightforward. It often is not clearhow to weight the impact of the different types of informationwhen inconsist
38、encies exist. Although, most calculations pro-duce results that agree with measurements within acceptabletolerance, the cause of discrepancies within the tolerance maynot be apparent from the available information. In this case,there is not universal agreement on the “best” answer, and thevarious ap
39、proaches to use of the benchmark data can beadopted. Some of these approaches are described in Section 6.4The boldface numbers given in parentheses refer to a list of references at theend of the text.E2006 162Caution should be used if it is necessary to extrapolate beyondthe limits of the benchmarks
40、.5. Summary of Reference Benchmarks for TransportCalculations for Reactor Pressure Vessel SurveillancePrograms5.1 Special Benchmark Irradiation Fields:5.1.1 One dedicated effort to provide benchmarks whoseradiation environments closely resemble those found outsidethe core of an operating reactor was
41、 the Nuclear RegulatoryCommissions LWR-PV-SDIP (1). This program promotedbetter monitoring of the radiation exposure of reactor vesselsand, thereby, provided for better assessment of vessel end-of-life conditions. In cooperation with other organizations nation-ally and internationally this program r
42、esulted in three bench-mark configurations, VENUS (2, 3, 4, 5, 6, 7, 8), PCA/PSF (9,10, 11, 12, 13, 14, 15), and NESDIP (16, 17, 18, 19).5.1.1.1 To serve as benchmarks, these special neutron envi-ronments had to be well characterized both experimentally andtheoretically. This came to mean that diffe
43、rences betweenmeasurements and calculations were reconciled and that un-certainty bounds for exposure parameters were well defined.Target uncertainties were 5 % to 10 % (1). To achieve theseobjectives, benchmarked dosimetry measurements were com-bined with neutron transport calculations, and statist
44、ical uncer-tainty analysis and spectral adjustment techniques were used toestablish the uncertainty bounds.5.1.1.2 Taken together, the three benchmarks provide cov-erage from the fuel region to the vessel cavity. The VENUSfacility was set up to measure spatial fluence distributions andneutron spectr
45、a near the fuel region and core barrel/thermalshield region. The PCA/PSF measurements looked at surveil-lance capsule effects and the fluence variation within the vesselitself. The NESDIP measurements overlap the PCA/PSF mea-surements and extend into the cavity behind the vessel.Investigations of ax
46、ial streaming in the cavity were alsoconducted in NESDIP.5.1.2 The VENUS Benchmark:5.1.2.1 The special benchmark field was developed at theVENUS Critical Facility CEN/SCK Laboratories, Belgium (2,3, 4, 5, 6, 7, 8). The facility could mock up PWR fuelgeometries to investigate the fluence rate distrib
47、utions inregions affected by the deviations from cylindrical symmetry.In addition, measurements on the VENUS fuel investigated theedge effects on power produced by individual pins at theoutside of the fuel region and thus better established theneutron source. These data provided verification of both
48、 theflux magnitude and the azimuthal flux shape. The mock upincluded a simulated core barrel and thermal shield.5.1.2.2 There were several phases to the VENUS program.The first PV mockup configuration studies (VENUS-I) pro-vided a link between the PCA and PSF tests and the actualenvironments of LWR
49、power plants. Indeed for actual powerplants, the azimuthal variation of the power distribution deter-mined largely by complex stair-step-shaped core peripheriesand by the core-boundary fuel power distributions could not beignored, otherwise the calculations could contain undetectedbiases. Such biases could be further exacerbated by the use oflow-leakage fuel-management schemes.5.1.2.3 A second configuration, VENUS-2, contained aplutonium-fueled zone at the periphery of the core (to simulateburned fuel), and its objective was to investigate how much thefast neutron fluence is affe
