ASTM E2006-2005 Standard Guide for Benchmark Testing of Light Water Reactor Calculations《轻水反应堆计算的基准检测的标准指南》.pdf

《ASTM E2006-2005 Standard Guide for Benchmark Testing of Light Water Reactor Calculations《轻水反应堆计算的基准检测的标准指南》.pdf》由会员分享，可在线阅读，更多相关《ASTM E2006-2005 Standard Guide for Benchmark Testing of Light Water Reactor Calculations《轻水反应堆计算的基准检测的标准指南》.pdf（7页珍藏版）》请在麦多课文档分享上搜索。

1、Designation: E 2006 05Standard Guide forBenchmark Testing of Light Water Reactor Calculations1This standard is issued under the fixed designation E 2006; the number immediately following the designation indicates the year oforiginal adoption or, in the case of revision, the year of last revision. A

2、number in parentheses indicates the year of last reapproval. Asuperscript epsilon (e) 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 (

3、Guide E 2005) 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 rel

4、evant 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 bi

5、as in 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 confid

6、ence.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. Referen

7、ced Documents2.1 ASTM Standards:2E 261 Practice for Determining Neutron Fluence Rate, Flu-ence, and Spectra by Radioactivation TechniquesE 262 Test Method for Determining Thermal Neutron Re-action and Fluence Rates by Radioactivation TechniquesE 706 Master Matrix for Light Water Reactor PressureVess

8、el Surveillance Standards, E 706 (O)E 844 Guide for Sensor Set Design and Irritation for Reac-tor SurveillanceE 944 Guide for Application of Neutron Spectrum Adjust-ment Methods in Reactor Surveillance, E 706 (IIA)E 1018 Guide for Application of ASTM Evaluated CrossSection Data File, E 706 (IIB)E 20

9、05 Guide for Benchmark Testing of Reactor Dosimetryin Standard and Reference Fields, E 706(IIE-1)3. 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 geometries. The calculat

10、ionsare 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 plants to provide accurat

11、e 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 parameters used in the trans

12、port calculations.Benchmarking is achieved by building up data bases ofbenchmark experiments which have different influences onuncertainty propagation. For example, fission spectra are thefundamental data bases which control propagation of crosssection uncertainties, while such physics-dosimetry exp

13、eri-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 with known characterist

14、icsto 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 wellknown energy spectra and intensitie

15、s. 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 are1This test method is under the jurisdictio

16、n of ASTM Committee E10 on NuclearTechnology and Applications and is the direct responsibility of SubcommitteeE10.05 on Nuclear Radiation and Metrology.Current edition approved Jan. 1, 2005. Published February 2005. Originallyapproved in 1999. Last previous edition approved in 1999 as E 2006 - 99.2F

17、or 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 Document Summary page onthe ASTM website.1Copyright ASTM International, 100 Barr Harbor Drive, PO Box C7

18、00, West Conshohocken, PA 19428-2959, United States.suitably characterized they are also referred to as benchmarkfields. A benchmark is that against which other things arereferenced, hence the terminology “to benchmark reference” or“benchmark referencing”. A variety of benchmark neutronfields, other

19、 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 additional benchmarking orbecause they have

20、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 Pressure Vessel Surveil-lance Dosimetry Improve

21、ment 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 for reac-tor surveillance and document th

22、em in a series of ASTMstandards (see Matrix E 706). The primary means chosen forvalidating LWR-PV-SDIP procedures was by benchmarking aseries of experimental and analytical studies in a variety offields (see Guide E 2005).4. Particulars of Benchmarking Transport Calculations4.1 Benchmarking of neutr

23、on 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 crosssections and other needed nuclear

24、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 transport calcula-tions for LWR systems (see

25、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 geometries. Thusfurther benchmarking in LWR g

26、eometries 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 that the dimensions and material compo

27、sitionscan 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 from the distributed source in the fuel

28、,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 can be expected to identify any bias i

29、nspecific 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.4.1.3 The benchmarking described above d

30、oes 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 these measurements is described in 5.2 a

31、nd 5.3.4.1.4 The final aspect of benchmarking is the benchmarkingof the dosimetry results. This aspect is treated in GuideE 2005(IIE-1) (see Matrix E 706). It is assumed that themeasurements in the benchmarked facilities and in the actualoperating plants are carried out using benchmarked reactionsan

32、d 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 have been verified (such asthrough round

33、-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 E 844, and evaluation of the dosimetry results shouldbe in accordance with Practice E 261 and Test Metho

34、d E 262.In particular, to compare measured dosimetry results withcalculated reaction rates or fluences, the following effects mustbe accounted for: effects of dosimetry perturbations, positionor gradient corrections, gamma attenuation in counted foils,differences in counting geometry from that of ca

35、librationstandards, 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 indicated any bias that exists in the dos

36、imetry crosssections. These cross sections are essentially independent ofthe transport cross sections discussed in 4.1.1. Recommendeddosimetry cross sections are given in Guide E 1018.4.1.5 The use of the benchmark data to determine bias incalculations and to determine best values for fluence incomp

37、lex 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 discrepancies within the tolerance

38、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 if it is necessary to extrapolate be

39、yondthe limits of the benchmarks.3The boldface numbers given in parentheses refer to a list of references at theend of the text.E20060525. Summary of Reference Benchmarks for ReactorPressure Vessel Surveillance Dosimetry5.1 Special Benchmark Irradiation Fields:5.1.1 One dedicated effort to provide b

40、enchmarks 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, provided for better assessment of vessel

41、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 as benchmarks, these special neutron

42、 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 were 5 % to 10 % (1s). To achieve thes

43、eobjectives, 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 provide cov-erage from the fuel regio

44、n 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 within the vesselitself. The NESDIP me

45、asurements 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 Facility CEN/SCK Laboratories, Belgu

46、im (2,3, 4, 5, 6, 7, 8). The facility can mock up PWR fuel geometriesto investigate the flux distributions in regions affected by thedeviations from cylindrical symmetry. In addition, measure-ments on the VENUS fuel can investigate the edge effects onpower produced by individual pins at the outside

47、of the fuelregion and thus better establish the neutron source. These dataprovide verification of both the flux magnitude and theazimuthal flux shape. The mock up includes a simulated corebarrel and thermal shield.5.1.2.2 There were several phases to the VENUS program.The first PV mockup configurati

48、on 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 peripheriesand by the core-boundary fuel power di

49、stributions 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 beused in validation of other sources asymmetries