ASTM E706-2016 5361 Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards《轻水反应堆压力容器监视标准的标准主模型》.pdf

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1、Designation: E706 16Standard Master Matrix forLight-Water Reactor Pressure Vessel SurveillanceStandards1This standard is issued under the fixed designation E706; the number immediately following the designation indicates the year oforiginal adoption or, in the case of revision, the year of last revi

2、sion. 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 master matrix standard describes a series of stan-dard practices, guides, and methods for the prediction ofneutron-induced

3、 changes in light-water reactor (LWR) pressurevessel (PV) and support structure steels throughout a pressurevessels service life (Fig. 1). Referenced documents are listedin Section 2. The summary information that is provided inSections 3 and 4 is essential for establishing proper understand-ing and

4、communications between the writers and users of thisset of matrix standards. It was extracted from the referencedstandards (Section 2) and references for use by individualwriters and users. More detailed writers and usersinformation, justification, and specific requirements for theindividual practic

5、es, guides, and methods are provided inSections 35. General requirements of content and consis-tency are discussed in Section 6.1.2 This master matrix is intended as a reference and guideto the preparation, revision, and use of standards in the series.1.3 To account for neutron radiation damage in s

6、ettingpressure-temperature limits and making fracture analyses (1-12)2and Guide E509), neutron-induced changes in reactorpressure vessel steel fracture toughness must be predicted, thenchecked by extrapolation of surveillance program data duringa vessels service life. Uncertainties in the predicting

7、 method-ology can be significant. Techniques, variables, and uncertain-ties associated with the physical measurements of PV andsupport structure steel property changes are not considered inthis master matrix, but elsewhere (2, 6, 7), (11-26), and GuideE509).1.4 The techniques, variables and uncertai

8、nties related to (1)neutron and gamma dosimetry, (2) physics (neutronics andgamma effects), and (3) metallurgical damage correlationprocedures and data are addressed in separate standards be-longing to this master matrix (1, 17). The main variables ofconcern to (1), (2), and (3) are as follows:1.4.1

9、 Steel chemical composition and microstructure,1.4.2 Steel irradiation temperature,1.4.3 Power plant configurations and dimensions, from thecore periphery to surveillance positions and into the vessel andcavity walls.1.4.4 Core power distribution,1.4.5 Reactor operating history,1.4.6 Reactor physics

10、 computations,1.4.7 Selection of neutron exposure units,1.4.8 Dosimetry measurements,1.4.9 Neutron special effects, and1.4.10 Neutron dose rate effects.1.5 A number of methods and standards exist for ensuringthe adequacy of fracture control of reactor pressure vessel beltlines under normal and accid

11、ent loads (1, 7, 8, 11, 12, 14, 16,17, 23-27), Referenced Documents: ASTM Standards (2.1),Nuclear Regulatory Documents (2.3) and ASME Standards(2.4). As older LWR pressure vessels become more highlyirradiated, the predictive capability for changes in toughnessmust improve. Since during a vessels ser

12、vice life an increasingamount of information will be available from test reactor andpower reactor surveillance programs, procedures to evaluateand use this information must be used (1, 2, 4-9, 11, 12, 23-26,28). This master matrix defines the current (1) scope, (2) areasof application, and (3) gener

13、al grouping for the series ofASTMstandards, as shown in Fig. 1.1.6 The values stated in SI units are to be regarded asstandard. No other units of measurement are included in thisstandard.1.7 This standard does not purport to address all of thesafety concerns, if any, associated with its use. It is t

14、he1This 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 Dec. 1, 2016. Published January 2017. Originallyapproved in 1979. Last previous editio

15、n approved in 2002 as E0706 -2002 whichwas withdrawn July 2011 and reinstated in December 2016. DOI: 10.1520/E0706-16.2The boldface numbers in parentheses refer to a list of references at the end ofthis standard.Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA

16、19428-2959. United StatesThis international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for theDevelopment of International Standards, Guides and Recommendations issued by the World Trade Organization Te

17、chnical Barriers to Trade (TBT) Committee.1responsibility 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. Referenced Documents2.1 ASTM Standards:3C859 Terminology Relating to Nuclear Materia

18、lsE23 Test Methods for Notched Bar Impact Testing of Me-tallic MaterialsE170 Terminology Relating to Radiation Measurements andDosimetryE185 Practice for Design of Surveillance Programs forLight-Water Moderated Nuclear Power Reactor VesselsE482 Guide for Application of Neutron Transport Methodsfor R

19、eactor Vessel SurveillanceE509 Guide for In-Service Annealing of Light-Water Mod-erated Nuclear Reactor VesselsE636 Guide for Conducting Supplemental SurveillanceTests for Nuclear Power Reactor Vessels, E 706 (IH)E646 Test Method for Tensile Strain-Hardening Exponents(n -Values) of Metallic Sheet Ma

20、terialsE693 Practice for Characterizing Neutron Exposures in Ironand Low Alloy Steels in Terms of Displacements PerAtom (DPA), E 706(ID)E844 Guide for Sensor Set Design and Irradiation forReactor Surveillance, E 706 (IIC)E853 Practice forAnalysis and Interpretation of Light-WaterReactor Surveillance

21、 ResultsE854 Test Method for Application and Analysis of SolidState Track Recorder (SSTR) Monitors for ReactorSurveillance, E706(IIIB)3For referenced ASTM standards, visit the ASTM website, www.astm.org, orcontact ASTM Customer Service at serviceastm.org. For Annual Book of ASTMStandards volume info

22、rmation, refer to the standards Document Summary page onthe ASTM website.FIG. 1 Organization and Use of ASTM Standards in the E706 Master MatrixE706 162E900 Guide for Predicting Radiation-Induced TransitionTemperature Shift in Reactor Vessel MaterialsE910 Test Method for Application and Analysis of

23、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 Monitors for Reactor Vessel SurveillanceE1006 Practice

24、for Analysis and Interpretation of PhysicsDosimetry Results from Test Reactor ExperimentsE1018 Guide for Application of ASTM Evaluated CrossSection Data File, Matrix E706 (IIB)E1035 Practice for Determining Neutron Exposures forNuclear Reactor Vessel Support StructuresE1214 Guide for Use of Melt Wir

25、e Temperature Monitorsfor Reactor Vessel Surveillance, E 706 (IIIE)E1253 Guide for Reconstitution of Irradiated Charpy-SizedSpecimensE2005 Guide for Benchmark Testing of Reactor Dosimetryin Standard and Reference Neutron FieldsE2006 Guide for Benchmark Testing of Light Water ReactorCalculationsE2215

26、 Practice for Evaluation of Surveillance Capsulesfrom Light-Water Moderated Nuclear Power Reactor Ves-selsE2956 Guide for Monitoring the Neutron Exposure of LWRReactor Pressure Vessels2.2 ASTM Adjunct:4ADJE090015-EA Adjunct for E900-15 Technical Basis forthe Equation Used to Predict Radiation-Induce

27、d Transi-tion Temperature Shift in Reactor Vessel Materials2.3 Nuclear Regulatory Documents:5Code of Federal Regulations, Chapter 10, Part 50 Appendi-ces G and HCode of Federal Regulations, Chapter 10, Part 21 Reportingof Defects and NoncomplianceRegulatory Guide 1.99 Radiation Embrittlement of Reac

28、torVessel MaterialsRegulatory Guide 1.150 Ultrasonic Testing of Reactor Ves-sel Welds During Preservice and Inservice ExaminationsRegulatory Guide 1.190 Calculational and Dosimetry Meth-ods for Determining Pressure Vessel Neutron Fluence2.4 American Society of Mechanical Engineers Standard:6Boiler a

29、nd Pressure Vessel Code Sections III and XI2.5 Bureau International de Poids et Mesures Documents:7The SI Brochure: The International System of Units (SI)3. LWR Pressure Vessel SurveillanceJustification,Requirements, and Status of Work3.1 Aging light water reactor pressure vessels (LWR-PV)accumulate

30、 significant neutron fluence exposures, with conse-quent changes in their state of steel embrittlement. Recogniz-ing that accurate and validated measurement and predictivemethods are needed to periodically evaluate the metallurgicalcondition of these reactor vessels, and in some instancesreactor ves

31、sel support structures (16, 17), international multi-laboratory work directed towards the improvement ofLWR-PV surveillance has been conducted (1, 2, 4, 29-34).3.2 The assessment of the radiation-induced degradation ofmaterial properties in a power reactor pressure vessel requirescharacterization of

32、 the neutron field from the edge of thereactor core to boundaries outside the pressure vessel. Mea-surements of neutron fluence, fluence rate, and spectrum forthis characterization are associated with two distinct compo-nents of LWR-PV radiation surveillance procedures: (1) propercalculational estim

33、ates of the neutron fluence delivered toin-vessel surveillance positions, various locations in the vesselwall, and ex-vessel support structures and surveillancepositions, and (2) understanding the interrelationship betweenmaterial property changes in reactor vessels, in vessel supportstructures, and

34、 in metallurgical test specimens irradiated in testreactors and at accelerated neutron fluence positions near thepressure vessel in operating power reactions (see Sections 4and 5).3.3 The first component referred to above requires valida-tion and calibration in a variety of neutron irradiation testf

35、acilities, including LWR-PV mock-ups, power reactor surveil-lance positions, and related benchmark neutron fields. Thebenchmarks also serve as a permanent measurement referencefor neutron fluence and fluence rate detection techniques.3.4 In order to meet the LWR-PV radiation monitoringrequirements,

36、a variety of neutron fluence, fluence rate, anddamage detectors are employed. Each detector must be vali-dated for application to the higher fluence rate and harderneutron spectrum of the test reactor test regions and to thelower fluence rate and softer neutron spectrum of the surveil-lance position

37、s. Required detectors must respond to neutrons ofvarious energies, so that multigroup spectra can be determinedwith accuracy sufficient for adequate damage response esti-mates for PV and support structure steels at end of life (EOL).3.5 The necessity for well-established and documented testreactor a

38、nd pressure vessel mock-up facilities for dosimetryand physics investigations and for irradiation of metallurgicalspecimens is recognized. These facilities provide well-characterized neutron environments where active and passiveneutron dosimetry, various types of LWR-PV neutron fieldphysics calculat

39、ions, and temperature-controlled metallurgicaldamage exposures are brought together for validation andcalibration. The neutron radiation field characteristics forsurveillance capsule in- and ex-vessel power reactor positionsare simulated in these mock-up facilities (1, 35).3.6 A few operating PWR an

40、d BWR power reactor bench-mark facilities have been selected for testing, validation, and4Available from ASTM International Headquarters. Order Adjunct No.ADJE090015-EA. Original adjunct produced in 2015.5Available from Superintendent of Documents, U.S. Government PrintingOffice, Washington, DC 2040

41、2.6Available from American Society of Mechanical Engineers (ASME), ASMEInternational Headquarters, Two Park Ave., New York, NY 10016-5990, http:/www.asme.org.7Available from Bureau International de Poids et Mesures, http:/www.bipm.org/en/publications/si-brochure/.E706 163calibration of physics compu

42、tational methods, processing andadjustment codes, nuclear data, and dosimetry techniques (1, 3,35).3.7 Federal Regulation 10 CFR 50 calls for adherence toseveral ASTM standards that require establishment of a sur-veillance program for each power reactor and incorporation offluence monitors for post-

43、irradiation neutron field evaluation.Revised and new standards must be structured to be up-to-date,flexible, and, above all consistent (see Section 6).4. Significance and Use4.1 Master MatrixThis matrix document is written as areference and guide to the use of existing standards and to helpmanage th

44、e development and application of new standards, asneeded for LWR-PV surveillance programs. Paragraphs 4.2 4.5 are provided to assist the authors and users involved in thepreparation, revision, and application of these standards (seeSection 6).4.2 Approach and Primary Objectives:4.2.1 Standardized pr

45、ocedures and reference data are rec-ommended in regard to (1) neutron and gamma dosimetry, (2)physics (neutronics and gamma effects), and (3) metallurgicaldamage correlation methods and data, associated with theanalysis, interpretation, and use of nuclear reactor test andsurveillance results.4.2.2 E

46、xisting state-of-the-art practices associated with (1),(2), and (3), if uniformly and consistently applied, can providereliable (10 to 30 %, 1) estimates of changes in LWR-PV steelfracture toughness during a reactors service life (36).4.2.3 Reg. Guide 1.99 and Section III of the ASME Boilerand Press

47、ure Vessel Code, Part NF2121 require that thematerials used in reactor pressure vessels support “.shall bemade of materials that are not injuriously affected by .irra-diation conditions to which the item will be subjected.”4.2.4 By the use of this series of standards and the uniformand consistent do

48、cumentation and reporting of estimatedchanges in LWR-PV steel fracture toughness with uncertaintiesof 10 to 30 % (1), the nuclear industry and licensing andregulatory agencies can meet realistic LWR power plantoperating conditions and limits, such as those defined inAppendices G and H of 10 CFR Part

49、 50, Reg. Guide 1.99, andthe ASME Boiler and Pressure Vessel Code.4.2.5 The uniform and consistent application of this seriesof standards allows the nuclear industry and licensing andregulatory agencies to properly administer their responsibilitiesin regard to the toughness of LWR power reactor materials tomeet requirements of Appendices G and H of 10 CFR Part 50,Reg. Guide 1.99, and the ASME Boiler and Pressure VesselCode.4.3 Dosimetry Analysis and Interpretation(1, 3-5, 21, 28,29, 35, 37 and 38) When properly implemented, validated, andcalibrated by vendor/utility