1、Designation: E 853 01 (Reapproved 2008)Standard Practice forAnalysis and Interpretation of Light-Water ReactorSurveillance Results, E706(IA)1This standard is issued under the fixed designation E 853; the number immediately following the designation indicates the year oforiginal adoption or, in the c
2、ase of revision, the year of last 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 a
3、nd interpretation ofneutron exposure 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(1-70).31.2 This practice relies on
4、, and ties together, the applicationof several supporting ASTM standard practices, guides, andmethods (see Master Matrix E 706) (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 documen
5、ts. Support subject areas that arediscussed include reactor physics calculations, dosimeter se-lection and analysis, and exposure units.NOTE 1(Figure 1 is deleted in the latest update. The user is refered toMaster Matrix E 706 for the latest figure of the standards interconnectiv-ity).1.3 This pract
6、ice is restricted to direct applications related tosurveillance programs that are established in support of theoperation, licensing, and regulation of LWR nuclear powerplants. Procedures and data related to the analysis, interpreta-tion, and application of test reactor results are addressed inPracti
7、ce E 560, Practice E 1006, Guide E 900, and PracticeE 1035.1.4 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-bi
8、lity of regulatory limitations prior to use.2. Referenced Documents2.1 ASTM Standards:4E 185 Practice for Design of Surveillance Programs forLight-Water Moderated Nuclear Power Reactor VesselsE 482 Guide for Application of Neutron Transport Methodsfor Reactor Vessel Surveillance, E706 (IID)E 560 Pra
9、ctice for Extrapolating Reactor Vessel Surveil-lance Dosimetry Results, E 706(IC)E 706 Master Matrix for Light-Water Reactor PressureVessel Surveillance Standards, E 706(0)E 844 Guide for Sensor Set Design and Irradiation forReactor Surveillance, E 706(IIC)E 854 Test Method for Application and Analy
10、sis of SolidState Track Recorder (SSTR) Monitors for Reactor Sur-veillance, E706(IIIB)E 900 Guide for Predicting Radiation-Induced TransitionTemperature Shift in Reactor Vessel Materials, E706 (IIF)E 910 Test Method for Application and Analysis of HeliumAccumulation Fluence Monitors for Reactor Vess
11、el Sur-veillance, E706 (IIIC)E 944 Guide for Application of Neutron Spectrum Adjust-ment Methods in Reactor Surveillance, E 706 (IIA)E 1005 Test Method for Application and Analysis of Radio-metric Monitors for Reactor Vessel Surveillance, E706(IIIA)E 1006 Practice for Analysis and Interpretation of
12、PhysicsDosimetry Results for Test Reactors, E 706(II)E 1035 Practice for Determining NeutronExposures forNuclear Reactor Vessel Support StructuresE 1214 Guide for Use of Melt Wire Temperature Monitorsfor Reactor Vessel Surveillance, E 706 (IIIE)E 2006 Guide for Benchmark Testing of Light Water Reac-
13、tor Calculations2.2 Other Documents:1This 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 Nov. 1, 2008. Published November 2008 Originallyappr
14、oved in 1981. Last previous edition approved in 2001 E 853 01.2ASTM Practice E 185 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 E208.3The boldface numbe
15、rs in parentheses refer to the list of references appended tothis practice. For an updated set of references, see the E706 Master Matrix.4Annual Book of ASTM Standards, Vol 12.02.1Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.NUREG/
16、CR-1861 HEDL-TME 80-87 LWR Pressure Ves-sel Surveillance Dosimetry Improvement Program: PCAExperiments and Blind Test5ASME Boiler and Pressure Vessel Code, Sections III andIX6Code of Federal Regulations, Title 10, Part 50, AppendixesG and H73. Significance and Use3.1 The objectives of a reactor vess
17、el surveillance programare twofold. The first requirement of the program is to monitorchanges in the fracture 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
18、use of the data obtained from thesurveillance program to determine the conditions under whichthe vessel can be 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 surveil
19、lance program may be treated as aseparate experiment. The goal is to define and carry tocompletion a dosimetry 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
20、 stricter sense to the specific plantfrom which the capsule was removed, but also in a broadersense to the industry 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 pressu
21、revessel itself will be exposed over its service life. This descrip-tion of the neutron field must include spatial 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
22、 a given surveil-lance capsule measurement, particularly one obtained early inplant life, is not necessarily representative 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
23、 and requirements of a reactor vesselssupport structures surveillance program are much less strin-gent, and at present, are limited to physics-dosimetry measure-ments through ex-vessel cavity monitoring coupled with theuse of available test reactor metallurgical data to determine thecondition of any
24、 support structure steels that might be subjectto neutron induced property changes (1, 29, 44-58, 65-70).4. Establishment of the Surveillance Program4.1 Practice E 185 describes the criteria that should beconsidered in planning and implementing surveillance testprograms and points out precautions th
25、at should be taken toensure that: (1) capsule exposures can be related to beltlineexposures, (2) materials selected for the surveillance programare samples of those materials most likely to limit the opera-tion of the reactor vessel, and (3) the tests yield results usefulfor the evaluation of radiat
26、ion effects on the reactor vessel.4.1.1 From the viewpoint of the radiation analyst, thecriteria explicated in Practice E 185 are met by the completionof the following tasks: (1) Determine the locations within thereactor that provide suitable lead factors (see Practice E 185)for each irradiation cap
27、sule relative to the pressure vessel; (2)Select neutron sensor sets that provide adequate coverage overthe energy range and fluence range of interest; (3) Specifysensor set locations within each irradiation capsule to defineneutron field gradients within the metallurgical specimen array.For reactors
28、 in which the end of life shift in RTNDTof thepressure vessel beltline material is predicted to be less than100F, gradient 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 ad
29、equately benchmark neutron transport meth-odology to be used both in the analysis of individual sensor setsand in the projection 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 prog
30、ram.However, the fourth item, which directly addresses the analysisand interpretation of surveillance results, is performed follow-ing withdrawal of the surveillance capsules from the reactor. Toprovide continuity between the designer and the analyst, it isrecommended that the documentation describi
31、ng the surveil-lance programs of individual reactors provide details of irra-diation capsule construction, locations of 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 doc
32、umented (1) metallurgi-cal and (2) physics-dosimetry data bases now exist for use bythe analyst based on both power reactor 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
33、 MatrixE 706: Guide E 844, Sensor Set Design; Test Method E 1005,Radiometric Monitors; Test Method E 854, Solid State TrackRecorder Monitors; Specification E 910, HeliumAccumulationFluence Monitors; and Damage Monitors. Dosimeter materialscurrently in common usage and acceptable for use in surveil-l
34、ance programs include Cu, Ti, Fe, Ni, U238,Np237,U235, andCo-Al. All radionuclide analysis of dosimeters should becalibrated to known sources such as those supplied by theNational Institute of Standards and Terchnology (NIST) or TheInternational Atomic Energy Agency (IAEA). All qualityassurance info
35、rmation pertinent to the sensor sets must bedocumented with the description of the surveillance program(1, 40-43, 48, 51-58).4.1.4 As indicated in 4.1.1, neutron transport methods areused both in the design of the surveillance program and in theanalysis and interpretation of capsule measurements. Du
36、ringthe design phase, neutron transport calculations are used todefine the neutron field within the pressure vessel wall and, in5Available from NRC Public Document Room, 1717 H St., NW, Washington,DC 20555.6Available from American Society of Mechanical Engineers, Three Park Ave.,New York, NY 10016-5
37、990.7Available from Superintendent of Documents, U. S. Government PrintingOffice, Washington, DC 20402.E 853 01 (2008)2conjunction 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 pressur
38、e-temperature limitations for normal plant operation as well as toevaluate the effect of various heat-up/cool-down transients onvessel condition.4.1.5 The neutron transport methodology used for thesecomputations must be well benchmarked and qualified forapplication to LWR configurations. The PCA (Ex
39、periment andBlind Test) data documented in Ref 47 provide one configu-ration for benchmarking basic transport methodology as wellas some of the input data used in power reactor calculations.Other suitably defined and documented benchmark experi-ments, such as those for VENUS (1, 43, 45) and for NESD
40、IP(1, 46, 50), may also be used to provide method verification.However, further analytical/experimental comparisons are re-quired to qualify a method for application to LWRs that havea more complex geometry and that require a more complextreatment of some input parameters, particularly of reactor co
41、repower distributions (1, 65-67). This additional qualificationmay be achieved by comparison with measurements taken inthe reactor cavity external to the pressure vessel of selectedoperating reactors (1, 51-57).4.1.6 All experimental/analytical comparisons that com-prise the qualification program fo
42、r a neutron transport meth-odology must be documented. At a minimum, this documen-tation should provide an assessment of the uncertainty or errorinherent in applying the methodology to the evaluation ofsurveillance capsule dosimetry and to the determination ofdamage gradients within the beltline reg
43、ion 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 tothe evaluation of surveillance dosimetry as well as to theprediction of damage within the pressure vessel, severaloptions are available regarding the choice of design bas
44、ispower distributions, the necessary detail in the geometricmockup, and the normalization of the analytical results. Themethodology chosen by any analyst should be documentedwith sufficient detail to permit a critical evaluation of theoverall approach. Further discussions of the application ofneutro
45、n transport methods to LWRs are provided in PracticeE 560 and Guide E 482.4.1.8 To ensure that metallurgical results obtained fromsurveillance capsule measurements may be applied to thedetermination of the pressure vessel fracture toughness, theirradiation temperature of the surveillance test specim
46、ens mustbe documented (see Guide E 1214).4.2 As stated in 3.2, the requirements for the establishmentof a surveillance program for reactor vessel support structuresare much less stringent than for the reactor vessel, and theanalyst is referred to Practice E 1035, for more information.5. Analysis of
47、Individual Surveillance Capsules5.1 It is recognized that for many operating power reactors,the documentation of baseline neutron transport calculationsand sensor set design information may not be available. In thatevent, to whatever extent possible the required informationshould be provided by the
48、service laboratory in the respectivesurveillance report (1, 29, 58).5.2 Radiometric analysis of capsule sensor sets shouldfollow procedures outlined in Test Method E 1005. For sensorssuch as the fission monitors which may be gamma-ray-sensitive, photo reaction corrections should be derived fromthe r
49、esults 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).5.3 In calculating spectrum averaged reaction cross sectionsfrom neutron transport calculations, care should be taken tomodel the explicit capsule configuration and location underexamination (see Guide E 482.) It will be necessary to deter-mine uncertainties associated with the determination of dam-age exposure parameters. The procedures outlined in GuideE 944,
