ASTM E853-2018 Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Neutron Exposure Results.pdf

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1、Designation: E853 13E853 18Standard Practice forAnalysis and Interpretation of Light-Water ReactorSurveillance Neutron Exposure 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

2、 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 in Annex A1, to be used in the analysis and in

3、terpretation of neutronexposure data obtained from LWR pressure vessel surveillance programs; and, based on the results of that analysis, establishes aformalism to be used to evaluate present and future condition of the pressure vessel and its support structures2 (1-74).31.2 This practice relies on,

4、 and ties together, the application of several supporting ASTM standard practices, guides, andmethods (see Master Matrix E706) (1, 5, 13, 48, 49).2 In order to make this practice at least partially self-contained, a moderateamount of discussion is provided in areas relating toASTM and other document

5、s. Support subject areas that are discussed includereactor physics calculations, dosimeter selection and analysis, and exposure units.1.3 This practice is restricted to direct applications related to surveillance programs that are established in support of theoperation, licensing, and regulation of

6、LWR nuclear power plants. Procedures and data related to the analysis, interpretation, andapplication of test reactor results are addressed in Practice E1006, Guide E900, and Practice E1035.1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It

7、is the responsibilityof the user of this standard to establish appropriate safety safety, health, and healthenvironmental practices and determine theapplicability of regulatory limitations prior to use.1.5 This international standard was developed in accordance with internationally recognized princi

8、ples on standardizationestablished in the Decision on Principles for the Development of International Standards, Guides and Recommendations issuedby the World Trade Organization Technical Barriers to Trade (TBT) Committee.2. Referenced Documents2.1 ASTM Standards:4E185 Practice for Design of Surveil

9、lance Programs for Light-Water Moderated Nuclear Power Reactor VesselsE482 Guide for Application of Neutron Transport Methods for Reactor Vessel SurveillanceE509 Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor VesselsE706 Master Matrix for Light-Water Reactor Pressure Vessel

10、Surveillance StandardsE844 Guide for Sensor Set Design and Irradiation for Reactor SurveillanceE854 Test Method for Application and Analysis of Solid State Track Recorder (SSTR) Monitors for Reactor SurveillanceE900 Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vesse

11、l MaterialsE910 Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel SurveillanceE944 Guide for Application of Neutron Spectrum Adjustment Methods in Reactor SurveillanceE1005 Test Method for Application and Analysis of Radiometric Monitors for Reactor

12、Vessel SurveillanceE1006 Practice for Analysis and Interpretation of Physics Dosimetry Results from Test Reactor ExperimentsE1018 Guide for Application of ASTM Evaluated Cross Section Data FileE1035 Practice for Determining Neutron Exposures for Nuclear Reactor Vessel Support StructuresE1214 Guide f

13、or Use of Melt Wire Temperature Monitors for Reactor Vessel Surveillance1 This practice is under the jurisdiction of ASTM Committee E10 on Nuclear Technology and Applications and is the direct responsibility of Subcommittee E10.05 onNuclear Radiation Metrology.Current edition approved June 1, 2013De

14、c. 1, 2018. Published July 2013December 2018 Originally approved in 1981. Last previous edition approved in 20082013 asE853 01E853 13.(2008). DOI: 10.1520/E0853-13.10.1520/E0853-18.2 ASTM Practice E185 gives reference to other standards and references that address the variables and uncertainties ass

15、ociated with property change measurements. Thereference standards are A370, E8, E21, E23, and E208.3 The boldface numbers in parentheses refer to the list of references appended to this practice. For an updated set of references, see the E706 Master Matrix.4 For referencedASTM standards, visit theAS

16、TM website, www.astm.org, or contactASTM Customer Service at serviceastm.org. For Annual Book of ASTM Standardsvolume information, refer to the standards Document Summary page on the ASTM website.This document is not an ASTM standard and is intended only to provide the user of an ASTM standard an in

17、dication of what changes have been made to the previous version. Becauseit may not be technically possible to adequately depict all changes accurately, ASTM recommends that users consult prior editions as appropriate. In all cases only the current versionof the standard as published by ASTM is to be

18、 considered the official document.Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States1E2006 Guide for Benchmark Testing of Light Water Reactor CalculationsE2215 Practice for Evaluation of Surveillance Capsules from Light-Water Moderated N

19、uclear Power Reactor VesselsE2956 Guide for Monitoring the Neutron Exposure of LWR Reactor Pressure Vessels2.2 Other Documents:NUREG/CR-1861 HEDL-TME 80-87 LWR Pressure Vessel Surveillance Dosimetry Improvement Program: PCA Experimentsand Blind Test5ASME Boiler and Pressure Vessel Code, Sections III

20、 and IX6Code of Federal Regulations, Title 10, Part 50, Appendixes G and H73. Significance and Use3.1 The objectives of a reactor vessel surveillance program are twofold. The first requirement of the program is to monitorchanges in the fracture toughness properties of ferritic materials in the react

21、or vessel beltline region resulting from exposure toneutron irradiation and the thermal environment. The second requirement is to make use of the data obtained from the surveillanceprogram to determine the conditions under which the vessel can be operated throughout its service life.3.1.1 To satisfy

22、 the first requirement of 3.1, the tasks to be carried out are straightforward. Each of the irradiation capsules thatcomprise the surveillance program may be treated as a separate experiment. The goal is to define and carry to completion adosimetry program that will, a posteriori, describe the neutr

23、on field to which the materials test specimens were exposed. Theresultant information will then become part of a data base applicable in a stricter sense to the specific plant from which the capsulewas removed, but also in a broader sense to the industry as a whole.3.1.2 To satisfy the second requir

24、ement of 3.1, the tasks to be carried out are somewhat complex. The objective is to describeaccurately the neutron field to which the pressure vessel itself will be exposed over its service life. This description of the neutronfield must include spatial gradients within the vessel wall. Therefore, h

25、eavy emphasis must be placed on the use of neutrontransport techniques as well as on the choice of a design basis for the computations. Since a given surveillance capsulemeasurement, particularly one obtained early in plant life, is not necessarily representative of long-term reactor operation, a si

26、mplenormalization of neutron transport calculations to dosimetry data from a given capsule may not be appropriate (1-67).23.2 The objectives and requirements of a reactor vessels support structures surveillance program are much less stringent, andat present, are limited to physics-dosimetry measurem

27、ents through ex-vessel cavity monitoring coupled with the use of availabletest reactor metallurgical data to determine the condition of any support structure steels that might be subject to neutron inducedproperty changes (1, 29, 44-58, 65-70).4. Establishment of the Surveillance Program4.1 Practice

28、 E185 describes the criteria that should be considered in planning and implementing surveillance test programs andpoints out precautions that should be taken to ensure that: (1) capsule exposures can be related to beltline exposures, (2) materialsselected for the surveillance program are samples of

29、those materials most likely to limit the operation of the reactor vessel, and(3) the tests yield results useful for the evaluation of radiation effects on the reactor vessel.4.1.1 From the viewpoint of the radiation analyst, the criteria explicated in Practice E185 are met by the completion of thefo

30、llowing tasks: (1) Determine the locations within the reactor that provide suitable lead factors (see Practice E185) for eachirradiation capsule relative to the pressure vessel; (2) Select neutron sensor sets that provide adequate coverage over the energyrange and fluence range of interest; (3) Spec

31、ify sensor set locations within each irradiation capsule to define neutron field gradientswithin the metallurgical specimen array. For reactors in which the end of life shift in RTNDT of the pressure vessel beltline materialis predicted to be less than 100 F, gradient measurements are not required.

32、In that case sensor set locations may be chosen toprovide a representative measurement for the entire surveillance capsule; and (4) Establish and adequately benchmark neutrontransport methodology to be used both in the analysis of individual sensor sets and in the projection of materials properties

33、changesto the vessel itself.4.1.2 The first three items listed in the preceding paragraph are carried out during the design of the surveillance program.However, the fourth item, which directly addresses the analysis and interpretation of surveillance results, is performed followingwithdrawal of the

34、surveillance capsules from the reactor. To provide continuity between the designer and the analyst, it isrecommended that the documentation describing the surveillance programs of individual reactors provide details of irradiationcapsule construction, locations of the capsules relative to the reacto

35、r core and internals, and sensor set design that are adequate toallow accurate evaluations of the surveillance measurement by the analyst. Well documented (1) metallurgical and (2)physics-dosimetry data bases now exist for use by the analyst based on both power reactor surveillance capsule and test

36、reactorresults (1, 12, 19-38, 58-64).4.1.3 Information regarding the choice of neutron sensor sets for LWR surveillance applications is provided in Matrix E706:Guide E844, Sensor Set Design; Test Method E1005, Radiometric Monitors; Test Method E854, Solid State Track Recorder5 Available from NRC Pub

37、lic Document Room, 1717 H St., NW, Washington, DC 20555.6 Available from American Society of Mechanical Engineers, Three Park Ave., New York, NY 10016-5990.7 Available from Superintendent of Documents, U. S. Government Printing Office, Washington, DC 20402.E853 182Monitors; Specification E910, Heliu

38、m Accumulation Fluence Monitors; and Damage Monitors. Dosimeter materials currently incommon usage and acceptable for use in surveillance programs include Cu, Ti, Fe, Ni, Nb, U238, Np237, U235, and Co-Al. Allradionuclide analysis of dosimeters should be calibrated to known sources such as those supp

39、lied by the National Institute ofStandards and Terchnology (NIST) or The International Atomic Energy Agency (IAEA). All quality assurance informationpertinent to the sensor sets must be documented with the description of the surveillance program (1, 40-43, 48, 51-58).4.1.4 As indicated in 4.1.1, neu

40、tron transport methods are used both in the design of the surveillance program and in the analysisand interpretation of capsule measurements. During the design phase, neutron transport calculations are used to define the neutronfield within the pressure vessel wall and, in conjunction with damage tr

41、end curves, to predict the degree of embrittlement of thereactor vessel over its service life. Embrittlement gradients are in turn used to determine pressure-temperature limitations fornormal plant operation as well as to evaluate the effect of various heat-up/cool-down transients on vessel conditio

42、n.4.1.5 The neutron transport methodology used for these computations must be well benchmarked and qualified for applicationto LWR configurations. The PCA (Experiment and Blind Test) data documented in Ref 47 provide one configuration forbenchmarking basic transport methodology as well as some of th

43、e input data used in power reactor calculations. Other suitablydefined and documented benchmark experiments, such as those for VENUS (1, 43, 45) and for NESDIP (1, 46, 50), may also beused to provide method verification. However, further analytical/experimental comparisons are required to qualify a

44、method forapplication to LWRs that have a more complex geometry and that require a more complex treatment of some input parameters,particularly of reactor core power distributions (1, 65-67). This additional qualification may be achieved by comparison withmeasurements taken in the reactor cavity ext

45、ernal to the pressure vessel of selected operating reactors (1, 51-57).4.1.6 All experimental/analytical comparisons that comprise the qualification program for a neutron transport methodologymust be documented. At a minimum, this documentation should provide an assessment of the uncertainty or erro

46、r inherent inapplying the methodology to the evaluation of surveillance capsule dosimetry and to the determination of damage gradients withinthe beltline region of the pressure vessel (1, 12, 19-21, 23-29, 36, 38, 43-48, 50-57).4.1.7 In the application of neutron transport methodology to the evaluat

47、ion of surveillance dosimetry as well as to the predictionof damage within the pressure vessel, several options are available regarding the choice of design basis power distributions, thenecessary detail in the geometric mockup, and the normalization of the analytical results. The methodology chosen

48、 by any analystshould be documented with sufficient detail to permit a critical evaluation of the overall approach. Further discussions of theapplication of neutron transport methods to LWRs are provided in Guide E482.4.1.8 To ensure that metallurgical results obtained from surveillance capsule meas

49、urements may be applied to the determinationof the pressure vessel fracture toughness, the irradiation temperature of the surveillance test specimens must be documented (seeGuide E1214).4.2 As stated in 3.2, the requirements for the establishment of a surveillance program for reactor vessel support structures aremuch less stringent than for the reactor vessel, and the analyst is referred to Practice E1035, for more information.5. Analysis of Individual Surveillance Capsules5.1 For surveillance programs designed according to E185, individual surveillance capsules are periodica