1、Designation: E2215 18Standard Practice forEvaluation of Surveillance Capsules from Light-WaterModerated Nuclear Power Reactor Vessels1This standard is issued under the fixed designation E2215; 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 evaluation of test specimensand dosimetry from light water moderated nu
3、clear powerreactor pressure vessel surveillance capsules.1.2 Additionally, this practice provides guidance on reas-sessing withdrawal schedule for design life and operationbeyond design life.1.3 This practice is one of a series of standard practices thatoutline the surveillance program required for
4、nuclear reactorpressure vessels. The surveillance program monitors theirradiation-induced changes in the ferritic steels that comprisethe beltline of a light-water moderated nuclear reactor pressurevessel.1.4 This practice along with its companion surveillanceprogram practice, Practice E185, is inte
5、nded for application inmonitoring the properties of beltline materials in any light-water moderated nuclear reactor.21.5 Modifications to the standard test program and supple-mental tests are described in Guide E636.1.6 The values stated in SI units are to be regarded as thestandard. The values give
6、n in parentheses are for informationonly.1.7 This international standard was developed in accor-dance with internationally recognized principles on standard-ization established in the Decision on Principles for theDevelopment of International Standards, Guides and Recom-mendations issued by the Worl
7、d Trade Organization TechnicalBarriers to Trade (TBT) Committee.2. Referenced Documents2.1 ASTM Standards:3A370 Test Methods and Definitions for Mechanical Testingof Steel ProductsE8/E8M Test Methods for Tension Testing of Metallic Ma-terialsE21 Test Methods for Elevated Temperature Tension Tests of
8、Metallic MaterialsE23 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 VesselsE208 Test Method for Conducting Drop-Weight
9、Test toDetermine Nil-Ductility Transition Temperature of Fer-ritic SteelsE509 Guide for In-Service Annealing of Light-Water Mod-erated Nuclear Reactor VesselsE636 Guide for Conducting Supplemental SurveillanceTests for Nuclear Power Reactor VesselsE693 Practice for Characterizing Neutron Exposures i
10、n Ironand Low Alloy Steels in Terms of Displacements PerAtom (DPA)E844 Guide for Sensor Set Design and Irradiation forReactor SurveillanceE853 Practice forAnalysis and Interpretation of Light-WaterReactor Surveillance ResultsE900 Guide for Predicting Radiation-Induced TransitionTemperature Shift in
11、Reactor Vessel MaterialsE1214 Guide for Use of Melt Wire Temperature Monitorsfor Reactor Vessel SurveillanceE1253 Guide for Reconstitution of Irradiated Charpy-SizedSpecimensE1820 Test Method for Measurement of Fracture Toughness1This practice is under the jurisdiction of ASTM Committee E10 on Nucle
12、arTechnology and Applications and is the direct responsibility of SubcommitteeE10.02 on Behavior and Use of Nuclear Structural Materials.Current edition approved Aug. 1, 2018. Published August 2018. Originallyapproved in 2002. Last previous edition approved in 2016 as E221516. DOI:10.1520/E2215-18.2
13、Prior to the adoption of these standard practices, surveillance capsule testingrequirements were only contained in Practice E185.3For referenced ASTM standards, visit the ASTM website, www.astm.org, orcontact ASTM Customer Service at serviceastm.org. For Annual Book of ASTMStandards volume informati
14、on, refer to the standards Document Summary page onthe ASTM website.Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United StatesThis international standard was developed in accordance with internationally recognized principles on standardization e
15、stablished in the Decision on Principles for theDevelopment of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.1E1921 Test Method for Determination of ReferenceTemperature, To, for Ferritic Steels in the Transitio
16、nRange2.2 ASME Standards:4Boiler and Pressure Vessel Code, Section III SubarticleNB-2000, Rules for Construction of Nuclear FacilityComponents, Class 1 Components, MaterialsBoiler and Pressure Vessel Code, Section XI NonmandatoryAppendix A, Analysis of Flaws, and Nonmandatory Ap-pendix G, Fracture T
17、oughness Criteria for Protectionagainst Failure3. Terminology3.1 Definitions:3.1.1 base metalas-fabricated plate material or forgingmaterial other than a weld or its corresponding heat-affected-zone (HAZ).3.1.2 beltlinethe irradiated region of the reactor vessel(shell material including weld seams a
18、nd plates or forgings)that directly surrounds the effective height of the active core.Note that materials in regions adjacent to the beltline maysustain sufficient neutron damage to warrant consideration inthe selection of surveillance materials.3.1.3 Charpy transition temperature curvea graphic orc
19、urve-fitted presentation, or both, of absorbed energy, lateralexpansion, or fracture appearance as a function of testtemperature, extending over a range including the lower shelf(5 % or less shear fracture appearance), transition region, andthe upper shelf (95 % or greater shear fracture appearance)
20、.3.1.4 Charpy transition temperature shiftthe difference inthe 41 J (30 ft-lbf) index temperatures for the best fit (average)Charpy absorbed energy curve measured before and afterirradiation. Similar measures of temperature shift can bedefined based on other indices in 3.1.3, but the current U.S.ind
21、ustry practice is to use 41 J (30 ft-lbf) and is consistent withGuide E900.3.1.5 Charpy upper-shelf energy levelthe average energyvalue for all Charpy specimen tests (preferably three or more)whose test temperature is at or above the Charpy upper-shelfonset; specimens tested at temperatures greater
22、than 83C(150F) above the Charpy upper-shelf onset shall not beincluded, unless no data are available between the onsettemperature and onset +83C (+150F).3.1.6 Charpy upper-shelf onsetthe temperature at whichthe fracture appearance of all Charpy specimens tested is at orabove 95 % shear.3.1.7 end-of-
23、license (EOL) fluencethe maximum pre-dicted fluence at the inside surface of the ferritic pressurevessel (if clad, the interface between cladding and ferritic steel)corresponding to the end of the applicable operating licenseperiod.3.1.8 heat-affected-zone (HAZ)plate material or forgingmaterial exte
24、nding outward from, but not including, the weldfusion line in which the microstructure of the base metal hasbeen altered by the heat of the welding process.3.1.9 index temperaturethe temperature corresponding toa predetermined level of absorbed energy, lateral expansion, orfracture appearance obtain
25、ed from the best-fit (average)Charpy transition curve.3.1.10 lead factorthe ratio of the average neutron fluence(E 1 MeV) of the specimens in a surveillance capsule to thepeak neutron fluence (E 1 MeV) of the correspondingmaterial at the ferritic steel reactor pressure vessel insidesurface calculate
26、d over the same time period.3.1.10.1 DiscussionChanges in the reactor operating pa-rameters and fuel management may cause the lead factor tochange.3.1.11 limiting materialstypically, the weld and base ma-terial with the highest predicted transition temperature usingthe projected fluence at the end o
27、f design life of each material,determined by adding the appropriate transition temperatureshift (TTS) to the unirradiated RTNDT. Guide E900 describes amethod for predicting the TTS. Regulators or other sourcesmay describe different methods for predicting TTS.3.1.12 maximum design fluence (MDF)the ma
28、ximum pro-jected fluence at the inside surface of the ferritic pressurevessel at the end of design life (if clad, MDF is defined at theinterface of the cladding to the ferritic steel).3.1.13 reference materialany steel that has been charac-terized as to the sensitivity of its tensile, impact and fra
29、cturetoughness properties to neutron radiation-induced embrittle-ment and is included in the Practice E185 surveillance pro-gram.3.1.14 reference temperature (RTNDT)see subarticle NB-2300 of the ASME Boiler and Pressure Vessel Code, SectionIII, for the definition of RTNDTfor unirradiated material ba
30、sedon Charpy (Test Methods A370) and drop weight tests (TestMethod E208). ASME Code Section XI, Appendices A and Gprovide an alternative definition for the reference temperature(RTTo) based on fracture toughness properties (Test MethodE1921).3.1.15 standby capsulea surveillance capsule meeting there
31、commendations of this practice that is or has been in thereactor vessel irradiation location as defined by Practice E185,but the testing of which is not required by this practice duringthe applicable operating license period.3.2 Neutron Exposure Terminology:3.2.1 Definitions of terms related to neut
32、ron dosimetry andexposure are provided in Terminology E170.4. Significance and Use4.1 Neutron radiation effects are considered in the design oflight-water moderated nuclear power reactors. Changes insystem operating parameters may be made throughout theservice life of the reactor to account for thes
33、e effects. Asurveillance program is used to measure changes in theproperties of actual vessel materials due to the irradiationenvironment. This practice describes the criteria that should beconsidered in evaluating surveillance program test capsules.4Available fromAmerican Society of Mechanical Engi
34、neers, Third ParkAvenue,New York, NY 10016.E2215 1824.2 Prior to the first issue date of this standard, the design ofsurveillance programs and the testing of surveillance capsuleswere both covered in a single standard, Practice E185. Betweenits provisional adoption in 1961 and its replacement linked
35、 tothis standard, Practice E185 was revised many times (1966,1970, 1973, 1979, 1982, 1993 and 1998). Therefore, capsulesfrom surveillance programs that were designed and imple-mented under early versions of the standard were often testedafter substantial changes to the standard had been adopted. For
36、clarity, the standard practice for surveillance programs hasbeen divided into the new Practice E185 that covers the designof new surveillance programs and this standard practice thatcovers the testing and evaluation of surveillance capsules.Modifications to the standard test program and supplemental
37、tests are described in Guide E636.4.3 This practice is intended to cover testing and evaluationof all light-water moderated reactor pressure vessel surveil-lance capsules. The practice is applicable to testing of capsulesfrom surveillance programs designed and implemented underall previous versions
38、of Practice E185.4.4 The radiation-induced changes in the properties of thereactor pressure vessel are generally monitored by measuringthe index temperatures, the upper-shelf energy and the tensileproperties of specimens from the surveillance program cap-sules. The significance of these radiation-in
39、duced changes isdescribed in Practice E185.4.5 Alternative methods exist for testing surveillance cap-sule materials. Some supplemental and alternative testingmethods are available as indicated in Guide E636. Directmeasurement of the fracture toughness is also feasible usingthe ToReference Temperatu
40、re method defined in Test MethodE1921 or J-integral techniques defined in Test Method E1820.Additionally, hardness testing can be used to supplementstandard methods as a means of monitoring the irradiationresponse of the materials.4.6 Practice E853 describes a methodology that may beused in the anal
41、ysis and interpretation of neutron dosimetrydata and the determination of neutron fluence. Regulators orother sources may describe different methods.4.7 Guide E900 describes a method for predicting the TTS.Regulators or other sources may describe different methods forpredicting TTS.4.8 Guide E509 pr
42、ovides direction for development of aprocedure for conducting an in-service thermal anneal of alight-water cooled nuclear reactor vessel and demonstrating theeffectiveness of the procedure including a post-annealingvessel radiation surveillance program.5. Determination of Capsule Condition5.1 Visual
43、 ExaminationA complete visual exam of thecapsule condition should be completed upon receipt and duringdisassembly at the testing laboratory. External identificationmarks on the capsule shall be verified. Signs of damage ordegradation of the capsule exterior shall be recorded.5.2 Capsule ContentThe s
44、pecimen loading pattern shouldbe compared to the capsule fabrication records and anydeviations shall be noted. Any evidence of corrosion or otherdamage to the specimens shall also be noted. The condition ofany temperature monitors shall be noted and recorded.5.3 Irradiation Temperature HistoryThe av
45、erage capsuletemperature during full power operation shall be estimated foreach reactor fuel cycle experienced by the capsule. The localreactor coolant temperature may be used as a reasonableapproximation, although gamma-ray heating should be consid-ered if it leads to a significant temperature diff
46、erence. In atypical pressurized water reactor, the coolant inlet temperaturemay be used as an estimate of the capsule irradiation tempera-ture using a time-weighted average (see Guide E900). In atypical boiling water reactor, the recirculation temperature maybe used as an estimate of the capsule irr
47、adiation temperature.5.4 Peak TemperatureTemperature monitors shall be ex-amined and any evidence of melting shall be recorded inaccordance with Guide E1214.6. Measurement of Irradiation Exposure6.1 The monthly power history of the reactor for all cyclesprior to capsule removal shall be recorded. Ot
48、her data that areneeded on a fuel-cycle-specific basis include: assembly-wisecore power distributions, including enrichments and burnups,axial core power distributions, axial core void distributions(BWRs only), and core and downcomer water temperatures.Other key changes that need to be recorded incl
49、ude the additionor removal of flux suppression rods or shield rods, uprates orderates of reactor power, and other reactor modifications suchas adding neutron shielding or the removal or addition ofstructures such as a thermal shield. Fuel assembly, reactorinternals, and reactor pressure vessel dimensional informationalso need to be recorded. Surveillance capsule locations andmovements: including storage periods outside the reactor, shallbe provided for the evaluation of irradiation exposure.6.2 Practice E853 describes practices for determining theneutron fluen
