1、Designation: E2006 10Standard 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 in light water reactor systems. Acompanion guide (Gui
3、de E2005) covers use of benchmarkfields for testing neutron transport calculations and crosssections in well controlled environments. This guide coversexperimental benchmarking of neutron fluence calculations (orcalculations of other exposure parameters such as dpa) in morecomplex geometries relevan
4、t to reactor surveillance. Particularsections of the guide discuss: the use of well-characterizedbenchmark neutron fields to provide an indication of theaccuracy of the calculational methods and nuclear data whenapplied to typical cases; and the use of plant specific measure-ments to indicate bias i
5、n individual plant calculations. Use ofthese two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined withanalytical uncertainty estimates for the calculations, will pro-vide uncertainty estimates for reactor fluences with a higherdegree of confidence
6、.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-bility of regulatory limitations prior to use.2. Referenced
7、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-activation TechniquesE706 Master Matrix for Light-Water Reacto
8、r Pressure Ves-sel Surveillance Standards, E 706(0)E844 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 706 (IIA)E1018 Guide for Application of ASTM Evaluated CrossSection Dat
9、a 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 calculations carried out to determinefluences for plant specific reactor ge
10、ometries. 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 the estimation of fluence within the reactorvessel of operating power pla
11、nts 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 propagated with due regard to the sensitivities of thedifferent input parame
12、ters 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 bases which control propagation of crosssection uncertainties, while suc
13、h 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 calculations. This guide describes generalprocedures for using neutron fields
14、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 usuallyirradiations performed in standard neutron fields with well-known energ
15、y spectra and intensities. There are, however, lesswell 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 character
16、ized they are also referred to as benchmark1This test method is under the jurisdiction of ASTM Committee E10 on NuclearTechnology and Applications and is the direct responsibility of SubcommitteeE10.05 on Nuclear Radiation Metrology.Current edition approved Oct. 1, 2010. Published October 2010. Orig
17、inallyapproved in 1999. Last previous edition approved in 2005 as E2006 - 05. DOI:10.1520/E2006-10.2For referenced ASTM standards, visit the ASTM website, www.astm.org, orcontact ASTM Customer Service at serviceastm.org. For Annual Book of ASTMStandards volume information, refer to the standards Doc
18、ument Summary page onthe ASTM website.1Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.fields. A benchmark is that against which other things arereferenced, hence the terminology “to benchmark reference” or“benchmark referencing”. A v
19、ariety of benchmark neutronfields, other than standard neutron fields, have been developed,or pressed into service, to improve the accuracy of neutrondosimetry measurement techniques. Some of these specialbenchmark experiments are discussed in this standard becausethey have identified needs for addi
20、tional benchmarking orbecause they have been sufficiently documented to serve asbenchmarks.3.3 One dedicated effort to provide benchmarks whoseradiation environments closely resemble those found outsidethe core of an operating reactor was the Nuclear RegulatoryCommissions Light Water Reactor Pressur
21、e Vessel Surveil-lance Dosimetry Improvement Program (LWR-PV-SDIP) (1)3.This program promoted better monitoring of the radiationexposure of reactor vessels and, thereby, provided for betterassessment of vessel end-of-life conditions.An objective of theLWR-PV-SDIP was to develop improved procedures f
22、or reac-tor surveillance and document them in a series of ASTMstandards (see Matrix E706). The primary means chosen forvalidating LWR-PV-SDIP procedures was by benchmarking aseries of experimental and analytical studies in a variety offields (see Guide E2005).4. Particulars of Benchmarking Transport
23、 Calculations4.1 Benchmarking of neutron transport calculations in-volves several distinct steps that are detailed below.4.1.1 Nuclear data used for transport calculations are evalu-ated using differential data or a combination of integral anddifferential data. This process results in a library of c
24、rosssections and other needed nuclear data (including fissionspectra) that, in the opinion of the evaluator, gives the best fitto the available experimental and theoretical results. Some ofinformation used in evaluating the cross sections may be thesame as that used directly for benchmarking transpo
25、rt 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-section set is derived in this fashion to beapplicable to a variety of calculational geometries and may notgive the most accurate answer for LWR geometr
26、ies. Thusfurther benchmarking in LWR geometries is required.4.1.2 Transport calculations in LWR geometries may bebenchmarked using measurements made in well-defined andwell-characterized facilities that each mock-up part of anLWR-type system. These facilities have the advantage overoperating plants
27、that the dimensions and material compositionscan be more accurately defined, the neutron source can be wellcharacterized, and measurements can be made in a largenumber of locations that would not be accessible in actualpower systems. In power reactors, one is interested in thetransport of neutrons f
28、rom the distributed source in the fuel,through the reactor internals and water to the vessel, andthrough the vessel to the reactor cavity. Three mockups thattogether encompass this entire transport problem are describedin 5.1. Modeling and calculating of neutron transport in thesevarious geometries
29、can be expected to identify any bias inspecific parts of the calculations. Biases that can be detectedinclude those due to modeling the irregular fuel geometry anddistributed neutron source, those due to errors in the cross-sections or neutron spectra, and those due to calculationalapproximations.4.
30、1.3 The benchmarking described above does not providechecks on geometries identical to actual plants and does notinclude bias that may exist in the definition of a specific plantmodel. Identification of these types of bias can only beaccomplished using actual plant measurements. Benchmarkingusing th
31、ese 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 treated in GuideE2005(IIE-1) (see Matrix E706). It is assumed that the mea-surements in the benchmarked facilities and in the actualoperating plants are car
32、ried out using benchmarked reactionsand dosimeters. This involves using reactions whose crosssections have been shown to be consistent with results in thesetypes of neutron environments. Also, the dosimeters andmeasurement facilities must be of adequate quality and havemeasurement accuracies that ha
33、ve been verified (such asthrough round-robin testing). Periodic recalibration of labora-tory measurement devices is also required using appropriatereference standards.4.1.4.1 Selection and use of dosimetry should be accordingto Guide E844, and evaluation of the dosimetry results shouldbe in accordan
34、ce with Practice E261 and Test Method E262.Inparticular, to compare measured dosimetry results with calcu-lated reaction rates or fluences, the following effects must beaccounted for: effects of dosimetry perturbations, position orgradient corrections, gamma attenuation in counted foils,differences
35、in counting geometry from that of calibrationstandards, dosimeter or reaction product burnup, effects ofcompeting reactions in impurities and photofission or photoin-duced reactions, and proper treatment of the irradiation history.4.1.4.2 The benchmarking of the dosimetry results will alsohave indic
36、ated any bias that exists in the dosimetry crosssections. These cross sections are essentially independent ofthe transport cross sections discussed in 4.1.1. Recommendeddosimetry cross sections are given in Guide E1018.4.1.5 The use of the benchmark data to determine bias incalculations and to deter
37、mine best values for fluence incomplex geometries is not straightforward. It often is not clearhow to weight the impact of the different types of informationwhen inconsistencies exist. Although, most calculations pro-duce results that agree with measurements within acceptabletolerance, the cause of
38、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 approaches to use of the benchmark data can beadopted. Some of these approaches are described in Section 6.Caution should be used i
39、f it is necessary to extrapolate beyondthe limits of the benchmarks.5. Summary of Reference Benchmarks for ReactorPressure Vessel Surveillance Dosimetry5.1 Special Benchmark Irradiation Fields:3The boldface numbers given in parentheses refer to a list of references at theend of the text.E2006 1025.1
40、.1 One dedicated effort to provide benchmarks whoseradiation environments closely resemble those found outsidethe core of an operating reactor was the Nuclear RegulatoryCommissions LWR-PV-SDIP (1). This program promotedbetter monitoring of the radiation exposure of reactor vesselsand, thereby, provi
41、ded for better assessment of vessel end-of-life conditions. In cooperation with other organizations nation-ally and internationally this program resulted 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
42、as benchmarks, these special neutron envi-ronments had to be well characterized both experimentally andtheoretically. This came to mean that differences betweenmeasurements and calculations were reconciled and that un-certainty bounds for exposure parameters were well defined.Target uncertainties we
43、re 5 % to 10 % (1s). To achieve theseobjectives, benchmarked dosimetry measurements were com-bined with neutron transport calculations, and statistical uncer-tainty analysis and spectral adjustment techniques were used toestablish the uncertainty bounds.5.1.1.2 Taken together, the three benchmarks p
44、rovide cov-erage from the fuel region to the vessel cavity. The VENUSfacility was set up to measure spatial fluence distributions andneutron spectra near the fuel region and core barrel/thermalshield region. The PCA/PSF measurements looked at surveil-lance capsule effects and the fluence fall-off wi
45、thin the vesselitself. The NESDIP measurements overlap the PCA/PSF mea-surements and extend into the cavity behind the vessel.Investigations of axial 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
46、Facility CEN/SCK Laboratories, Belgium (2,3, 4, 5, 6, 7, 8). The facility can mock up PWR fuel geometriesto investigate the fluence rate distributions in regions affectedby the deviations from cylindrical symmetry. In addition,measurements on the VENUS fuel can investigate the edgeeffects on power p
47、roduced by individual pins at the outside ofthe fuel region and thus better establish the neutron source.These data provide verification of both the flux magnitude andthe azimuthal flux shape. The mock up includes a simulatedcore barrel and thermal shield.5.1.2.2 There were several phases to the VEN
48、US program.The first PV mockup configuration studies (VENUS-I) pro-vided a link between the PCA and PSF tests and the actualenvironments of LWR power plants. Indeed for actual powerplants, the azimuthal variation of the power distribution deter-mined largely by complex stair-step-shaped core periphe
49、riesand 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 affected by such a core loading, and ifchanges in calculational modeling are necessary to account forany effects. The VENUS facility can also provide data to beuse
