1、Designation: E 853 01Standard 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 case of revision, t
2、he year of last revision. A 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 practice covers the methodology, summarized inAnnex A1, to be used in the analysis and interpretation
3、 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, and ties togeth
4、er, 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 documents. Support subje
5、ct 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 practice is restricted
6、 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 inPractice E 560, Practic
7、e 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-bility of regulator
8、y limitations prior to use.2. Referenced Documents2.1 ASTM Standards:E 170 Terminology Relating to Radiation Measurementsand Dosimetry4E 184 Practice for Effects of High-Energy Neutron Radia-tion on the Mechanical Properties of Metallic Materials,E706 (IB)4E 185 Practice for Conducting Surveillance
9、Tests for Light-Water Cooled Nuclear Power Reactor Vessels, E706 (IF)4E 482 Guide for Application of Neutron Transport Methodsfor Reactor Vessel Surveillance, E706 (IID)4E 560 Practice for Extrapolating Reactor Vessel Surveil-lance Dosimetry Results, E706 (IC)4E 636 Guide for Conducting Supplemental
10、 SurveillanceTests for Nuclear Power Reactor Vessels, E706 (IH)4E 693 Practice for Characterizing Neutron Exposures inIron and Low Alloy Steels in Terms of Displacements PerAtom (DPA), E706 (ID)4E 706 Master Matrix for Light-Water Reactor PressureVessel Surveillance Standards4E 844 Guide for Sensor
11、Set Design and Irradiation forReactor Surveillance, E706 (IIC)4E 854 Test Method for Application and Analysis of SolidState Track Recorder (SSTR) Monitors for Reactor Sur-veillance, E706 (IIIB)4E 900 Guide for Predicting Neutron Radiation Damage toReactor Vessel Materials, E706 (IIF)4E 910 Specifica
12、tion for Application and Analysis of HeliumAccumulation Fluence Monitors for Reactor Vessel Sur-veillance, E706 (IIIC)4E 944 Guide for Application of Neutron Spectrum Adjust-ment Methods in Reactor Surveillance, E706 (IIA)4E 1005 Test Method for Application and Analysis of Radio-metric Monitors for
13、Reactor Vessel Surveillance, E706(IIIA)4E 1006 Practice for Analysis and Interpretation of PhysicsDosimetry Results for Test Reactors, E706 (II)4E 1018 Guide for Application of ASTM Evaluated CrossSection Data File, E706 (IIB)4E 1035 Practice for Determining Radiation Exposures forNuclear Reactor Ve
14、ssel Support Structures, E706 (IG)4E 1214 Guide for Use of Melt Wire Temperature Monitorsfor Reactor Vessel Surveillance, E706 (IIIE)41This practice is under the jurisdiction of ASTM Committee E10 on NuclearTechnology and Applications and is the direct responsibility of SubcommitteeE10.05 on Nuclear
15、 Radiation Metrology.Current edition approved June 10, 2001. Published September 2001 Originallypublished as E 853 81. Last previous edition E 853 95e1.2ASTM Practice E 185 gives reference to other standards and references thataddress the variables and uncertainties associated with property change m
16、easure-ments. The reference standards are A370, E8, E21, E23, and 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.4Annual Book of ASTM Standards, Vol 12.02.1Copyright ASTM, 100 Barr Harbo
17、r Drive, West Conshohocken, PA 19428-2959, United States.E 2005 Guide for the Benchmark Testing of Reactor Do-simetry in Standard and Reference Neutron Fields, E706(IIE-1)4E 2006 Guide for the Benchmark Testing of Light WaterReactor Calculation42.2 Other Documents:NUREG/CR-1861 HEDL-TME 80-87 LWR Pr
18、essure 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 vessel surveillance programare tw
19、ofold. 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 use of the data obtained from
20、 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 surveillance program may be treated
21、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 stricter sense to the specif
22、ic 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 pressurevessel itself will be expos
23、ed 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 a given surveil-lance capsul
24、e 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 and requirements of a reacto
25、r 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 support structure steels tha
26、t 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 that should be taken toensure t
27、hat: (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 radiation effects on the reactor ve
28、ssel.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 capsule relative to the pressure
29、 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 in which the end of life shi
30、ft 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 adequately benchmark neutron tr
31、ansport 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 program.However, the fourth item,
32、 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 describing the surveil-lance programs
33、 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 documented (1) metallurgi-cal an
34、d (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 MatrixE 706: Guide E 844, Se
35、nsor Set Design; Test Method E 1005,Radiometric Monitors; Test Method E 854, Solid State TrackRecorder Monitors; Specification E 910, Helium AccumulationFluence Monitors; and Damage Monitors. Dosimeter materialscurrently in common usage and acceptable for use in surveil-lance programs include Cu, Ti
36、, 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 information pertinent to the sen
37、sor 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 are5Available from NRC Public Document Room, 1717 H St., NW, Washington,DC 20555.6Available from American Society of Mechanical Engineers, Three Par
38、k Ave.,New York, NY 10016-5990.7Available from Superintendent of Documents, U. S. Government PrintingOffice, Washington, DC 20402.E 8532used both in the design of the surveillance program and in theanalysis and interpretation of capsule measurements. Duringthe design phase, neutron transport calcula
39、tions are used todefine the neutron field within the pressure vessel wall and, inconjunction 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 p
40、lant 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 (Experiment andBlind Test) data documente
41、d 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 NESDIP(1, 46, 50), may also be used to pro
42、vide 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 corepower distributions (1, 65-67). This
43、 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 for a neutron transport meth-odology mus
44、t 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 region of the pressurevessel (1, 12, 19-2
45、1, 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 basispower distributions, the necessary d
46、etail 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 ofneutron transport methods to LWRs are provid
47、ed 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 specimens mustbe documented (see Guide E 121
48、4).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 Individual Surveillance Capsules5.1 It
49、 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 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 results of gamma-ray transp
