ASTM E853-2013 6875 Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results《轻水反应堆监测结果的分析和说明的标准实施规程》.pdf

《ASTM E853-2013 6875 Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results《轻水反应堆监测结果的分析和说明的标准实施规程》.pdf》由会员分享，可在线阅读，更多相关《ASTM E853-2013 6875 Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results《轻水反应堆监测结果的分析和说明的标准实施规程》.pdf（9页珍藏版）》请在麦多课文档分享上搜索。

1、Designation: E853 13Standard Practice forAnalysis and Interpretation of Light-Water ReactorSurveillance Results1This standard is issued under the fixed designation E853; the number immediately following the designation indicates the year oforiginal adoption or, in the case of revision, the year of l

2、ast revision. A number in parentheses indicates the year of last reapproval. Asuperscript epsilon () indicates an editorial change since the last revision or reapproval.1. Scope1.1 This practice covers the methodology, summarized inAnnex A1, to be used in the analysis and interpretation ofneutron ex

3、posure data obtained from LWR pressure vesselsurveillance programs; and, based on the results of thatanalysis, establishes a formalism to be used to evaluate presentand future condition of the pressure vessel and its supportstructures2(1-74).31.2 This practice relies on, and ties together, the appli

4、cationof several supporting ASTM standard practices, guides, andmethods (see Master Matrix E706) (1, 5, 13, 48, 49).2In orderto make this practice at least partially self-contained, a mod-erate amount of discussion is provided in areas relating toASTM and other documents. Support subject areas that

5、arediscussed include reactor physics calculations, dosimeter se-lection and analysis, and exposure units.1.3 This practice is restricted to direct applications related tosurveillance programs that are established in support of theoperation, licensing, and regulation of LWR nuclear powerplants. Proce

6、dures and data related to the analysis,interpretation, and application of test reactor results are ad-dressed in Practice E1006, Guide E900, and Practice E1035.1.4 This standard does not purport to address all of thesafety concerns, if any, associated with its use. It is theresponsibility of the use

7、r of this standard to establish appro-priate safety and health practices and determine the applica-bility of regulatory limitations prior to use.2. Referenced Documents2.1 ASTM Standards:4E185 Practice for Design of Surveillance Programs forLight-Water Moderated Nuclear Power Reactor VesselsE482 Gui

8、de for Application of Neutron Transport Methodsfor Reactor Vessel Surveillance, E706 (IID)E509 Guide for In-Service Annealing of Light-Water Mod-erated Nuclear Reactor VesselsE706 Master Matrix for Light-Water Reactor Pressure VesselSurveillance Standards, E 706(0) (Withdrawn 2011)5E844 Guide for Se

9、nsor Set Design and Irradiation forReactor Surveillance, E 706 (IIC)E854 Test Method for Application and Analysis of SolidState Track Recorder (SSTR) Monitors for ReactorSurveillance, E706(IIIB)E900 Guide for Predicting Radiation-Induced TransitionTemperature Shift in Reactor Vessel Materials, E706

10、(IIF)E910 Test Method for Application and Analysis of HeliumAccumulation Fluence Monitors for Reactor VesselSurveillance, E706 (IIIC)E944 Guide for Application of Neutron Spectrum Adjust-ment Methods in Reactor Surveillance, E 706 (IIA)E1005 Test Method for Application and Analysis of Radio-metric M

11、onitors for Reactor Vessel Surveillance, E 706(IIIA)E1006 Practice for Analysis and Interpretation of PhysicsDosimetry Results for Test Reactors, E 706(II)E1018 Guide for Application of ASTM Evaluated CrossSection Data File, Matrix E706 (IIB)E1035 Practice for Determining Neutron Exposures forNuclea

12、r Reactor Vessel Support Structures1This practice 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 June 1, 2013. Published July 2013 Originally approved

13、in 1981. Last previous edition approved in 2008 as E853 01(2008). DOI:10.1520/E0853-13.2ASTM Practice E185 gives reference to other standards and references thataddress the variables and uncertainties associated with property change measure-ments. The reference standards are A370, E8, E21, E23, and

14、E208.3The boldface numbers in parentheses refer to the list of references appended tothis practice. For an updated set of references, see the E706 Master Matrix.4For referenced ASTM standards, visit the ASTM website, www.astm.org, orcontact ASTM Customer Service at serviceastm.org. For Annual Book o

15、f ASTMStandards volume information, refer to the standards Document Summary page onthe ASTM website.5The 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. United States1E1

16、214 Guide for Use of Melt Wire Temperature Monitorsfor Reactor Vessel Surveillance, E 706 (IIIE)E2006 Guide for Benchmark Testing of Light Water ReactorCalculations2.2 Other Documents:NUREG/CR-1861 HEDL-TME 80-87 LWR Pressure VesselSurveillance Dosimetry Improvement Program: PCA Ex-periments and Bli

17、nd Test6ASME Boiler and Pressure Vessel Code, Sections III andIX7Code of Federal Regulations, Title 10, Part 50, AppendixesG and H83. Significance and Use3.1 The objectives of a reactor vessel surveillance programare twofold. The first requirement of the program is to monitorchanges in the fracture

18、toughness properties of ferritic materi-als in the reactor vessel beltline region resulting from exposureto neutron irradiation and the thermal environment. The secondrequirement is to make use of the data obtained from thesurveillance program to determine the conditions under whichthe vessel can be

19、 operated throughout its service life.3.1.1 To satisfy the first requirement of 3.1, the tasks to becarried out are straightforward. Each of the irradiation capsulesthat comprise the surveillance program may be treated as aseparate experiment. The goal is to define and carry tocompletion a dosimetry

20、 program that will, a posteriori, de-scribe the neutron field to which the materials test specimenswere exposed. The resultant information will then become partof a data base applicable in a stricter sense to the specific plantfrom which the capsule was removed, but also in a broadersense to the ind

21、ustry as a whole.3.1.2 To satisfy the second requirement of 3.1, the tasks tobe carried out are somewhat complex. The objective is todescribe accurately the neutron field to which the pressurevessel itself will be exposed over its service life. This descrip-tion of the neutron field must include spa

22、tial gradients withinthe vessel wall. Therefore, heavy emphasis must be placed onthe use of neutron transport techniques as well as on the choiceof a design basis for the computations. Since a given surveil-lance capsule measurement, particularly one obtained early inplant life, is not necessarily r

23、epresentative of long-term reactoroperation, a simple normalization of neutron transport calcu-lations to dosimetry data from a given capsule may not beappropriate (1-67).23.2 The objectives and requirements of a reactor vesselssupport structures surveillance program are much lessstringent, and at p

24、resent, are limited to physics-dosimetrymeasurements through ex-vessel cavity monitoring coupledwith the use of available test reactor metallurgical data todetermine the condition of any support structure steels thatmight be subject to neutron induced property changes (1, 29,44-58, 65-70).4. Establi

25、shment of the Surveillance Program4.1 Practice E185 describes the criteria that should beconsidered in planning and implementing surveillance testprograms and points out precautions that should be taken toensure that: (1) capsule exposures can be related to beltlineexposures, (2) materials selected

26、for the surveillance programare samples of those materials most likely to limit the operationof the reactor vessel, and (3) the tests yield results useful forthe evaluation of radiation effects on the reactor vessel.4.1.1 From the viewpoint of the radiation analyst, thecriteria explicated in Practic

27、e E185 are met by the completionof the following tasks: (1) Determine the locations within thereactor that provide suitable lead factors (see Practice E185)for each irradiation capsule relative to the pressure vessel; (2)Select neutron sensor sets that provide adequate coverage overthe energy range

28、and fluence range of interest; (3) Specifysensor set locations within each irradiation capsule to defineneutron field gradients within the metallurgical specimen array.For reactors in which the end of life shift in RTNDTof thepressure vessel beltline material is predicted to be less than100 F, gradi

29、ent measurements are not required. In that casesensor set locations may be chosen to provide a representativemeasurement for the entire surveillance capsule; and (4)Establish and adequately benchmark neutron transport meth-odology to be used both in the analysis of individual sensor setsand in the p

30、rojection of materials properties changes to thevessel itself.4.1.2 The first three items listed in the preceding paragraphare carried out during the design of the surveillance program.However, the fourth item, which directly addresses the analysisand interpretation of surveillance results, is perfo

31、rmed follow-ing withdrawal of the surveillance capsules from the reactor. Toprovide continuity between the designer and the analyst, it isrecommended that the documentation describing the surveil-lance programs of individual reactors provide details of irra-diation capsule construction, locations of

32、 the capsules relativeto the reactor core and internals, and sensor set design that areadequate to allow accurate evaluations of the surveillancemeasurement by the analyst. Well documented (1) metallurgi-cal and (2) physics-dosimetry data bases now exist for use bythe analyst based on both power rea

33、ctor surveillance capsuleand test reactor results (1, 12, 19-38, 58-64).4.1.3 Information regarding the choice of neutron sensorsets for LWR surveillance applications is provided in MatrixE706: Guide E844, Sensor Set Design; Test Method E1005,Radiometric Monitors; Test Method E854, Solid State Track

34、Recorder Monitors; Specification E910, Helium AccumulationFluence Monitors; and Damage Monitors. Dosimeter materialscurrently in common usage and acceptable for use in surveil-lance programs include Cu, Ti, Fe, Ni, Nb, U238,Np237,U235,and Co-Al. All radionuclide analysis of dosimeters should becalib

35、rated to known sources such as those supplied by theNational Institute of Standards and Terchnology (NIST) or TheInternational Atomic Energy Agency (IAEA). All quality6Available from NRC Public Document Room, 1717 H St., NW, Washington,DC 20555.7Available from American Society of Mechanical Engineer

36、s, Three Park Ave.,New York, NY 10016-5990.8Available from Superintendent of Documents, U. S. Government PrintingOffice, Washington, DC 20402.E853 132assurance information pertinent to the sensor sets must bedocumented with the description of the surveillance program(1, 40-43, 48, 51-58).4.1.4 As in

37、dicated in 4.1.1, neutron transport methods areused both in the design of the surveillance program and in theanalysis and interpretation of capsule measurements. Duringthe design phase, neutron transport calculations are used todefine the neutron field within the pressure vessel wall and, inconjunct

38、ion with damage trend curves, to predict the degree ofembrittlement of the reactor vessel over its service life.Embrittlement gradients are in turn used to determine pressure-temperature limitations for normal plant operation as well as toevaluate the effect of various heat-up/cool-down transients o

39、nvessel condition.4.1.5 The neutron transport methodology used for thesecomputations must be well benchmarked and qualified forapplication to LWR configurations. The PCA (Experiment andBlind Test) data documented in Ref 47 provide one configu-ration for benchmarking basic transport methodology as we

40、llas some of the input data used in power reactor calculations.Other suitably defined and documented benchmarkexperiments, such as those for VENUS (1, 43, 45) and forNESDIP (1, 46, 50), may also be used to provide methodverification. However, further analytical/experimental com-parisons are required

41、 to qualify a method for application toLWRs that have a more complex geometry and that require amore complex treatment of some input parameters, particularlyof reactor core power distributions (1, 65-67). This additionalqualification may be achieved by comparison with measure-ments taken in the reac

42、tor cavity external to the pressure vesselof selected operating reactors (1, 51-57).4.1.6 All experimental/analytical comparisons that com-prise the qualification program for a neutron transport meth-odology must be documented. At a minimum, this documen-tation should provide an assessment of the un

43、certainty or errorinherent in applying the methodology to the evaluation ofsurveillance capsule dosimetry and to the determination ofdamage gradients within the beltline region of the pressurevessel (1, 12, 19-21, 23-29, 36, 38, 43-48, 50-57).4.1.7 In the application of neutron transport methodology

44、 tothe evaluation of surveillance dosimetry as well as to theprediction of damage within the pressure vessel, severaloptions are available regarding the choice of design basispower distributions, the necessary detail in the geometricmockup, and the normalization of the analytical results. Themethodo

45、logy chosen by any analyst should be documentedwith sufficient detail to permit a critical evaluation of theoverall approach. Further discussions of the application ofneutron transport methods to LWRs are provided in GuideE482.4.1.8 To ensure that metallurgical results obtained fromsurveillance caps

46、ule measurements may be applied to thedetermination of the pressure vessel fracture toughness, theirradiation temperature of the surveillance test specimens mustbe documented (see Guide E1214).4.2 As stated in 3.2, the requirements for the establishmentof a surveillance program for reactor vessel su

47、pport structuresare much less stringent than for the reactor vessel, and theanalyst is referred to Practice E1035, for more information.5. Analysis of Individual Surveillance Capsules5.1 It is recognized that for many operating power reactors,the documentation of baseline neutron transport calculati

48、onsand sensor set design information may not be available. In thatevent, to whatever extent possible the required informationshould be provided by the service laboratory in the respectivesurveillance report (1, 29, 58).5.2 Radiometric analysis of capsule sensor sets shouldfollow procedures outlined

49、in Test Method E1005. For sensorssuch as the fission monitors which may be gamma-ray-sensitive, photo reaction corrections should be derived fromthe results of gamma-ray transport calculations performed forthe explicit capsule configuration under examination. Photoreaction corrections in LWR environments have been shown tobe extremely configuration dependent (1, 29, 58). Gamma-raycalculations should be well benchmarked. One such suitablereactor geometry benchmark is VENUS-1 (75, 76).5.3 In calculating spectrum averaged reaction cross sectionsfrom neutron transport calculations, c