1、Designation: E2215 15Standard 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.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 ofMetallic MaterialsE23 T
7、est 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 Test toDetermine Nil-Du
8、ctility 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 Vessels, E 706 (IH)E693 Practice for Characterizing Neutron Exposures in Ironand L
9、ow 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 ResultsE900 Guide for Predicting Radiation-Induced TransitionTemperat
10、ure Shift in Reactor Vessel MaterialsE1214 Guide for Use of Melt Wire Temperature Monitorsfor Reactor Vessel Surveillance, E 706 (IIIE)E1253 Guide for Reconstitution of Irradiated Charpy-SizedSpecimensE1820 Test Method for Measurement of Fracture ToughnessE1921 Test Method for Determination of Refer
11、enceTemperature, To, for Ferritic Steels in the TransitionRange2.2 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, An
12、alysis of Flaws, and Nonmandatory Ap-pendix G, Fracture Toughness Criteria for Protectionagainst Failure1This practice is under the jurisdiction of ASTM Committee E10 on NuclearTechnology and Applications and is the direct responsibility of SubcommitteeE10.02 on Behavior and Use of Nuclear Structura
13、l Materials.Current edition approved June 1, 2015. Published July 2015. Originally approvedin 2002. Last previous edition approved in 2010 as E221510. DOI: 10.1520/E2215-15.2Prior to the adoption of these standard practices, surveillance capsule testingrequirements were only contained in Practice E1
14、85.3For referenced ASTM standards, visit the ASTM website, www.astm.org, orcontact ASTM Customer Service at serviceastm.org. For Annual Book of ASTMStandards volume information, refer to the standards Document Summary page onthe ASTM website.4Available fromAmerican Society of Mechanical Engineers, T
15、hird ParkAvenue,New York, NY 10016.Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States13. Terminology3.1 Definitions:3.1.1 base metalas-fabricated plate material or forgingmaterial other than a weld or its corresponding heat-affected-zone
16、 (HAZ).3.1.2 beltlinethe irradiated region of the reactor vessel(shell material including weld seams and plates or forgings)that directly surrounds the effective height of the active core.Note that materials in regions adjacent to the beltline maysustain suficient neutron damage to warrant considera
17、tion inthe selection of surveillance materials.3.1.3 Charpy transition temperature curvea graphic orcurve-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 fr
18、acture appearance), transition region, andthe upper shelf (95 % or greater shear fracture appearance).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
19、 measures of temperature shift can bedefined based on other indices in 3.1.3, but the current U.S.industry 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 t
20、est temperature is at or above the Charpy upper-shelfonset; specimens tested at temperatures greater 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
21、at whichthe fracture appearance of all Charpy specimens tested is at orabove 95 % shear.3.1.7 end-of-license (EOL) fluencethe maximum pre-dicted fluence at the inside surface of the ferritic pressurevessel (if clad, the interface of the cladding to the ferritic steel)corresponding to the end of the
22、operating license period.3.1.8 heat-affected-zone (HAZ)plate material or forgingmaterial extending 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 correspondin
23、g toa predetermined level of absorbed energy, lateral expansion, orfracture appearance obtained 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 th
24、e correspondingmaterial at the ferritic steel reactor pressure vessel insidesurface calculated 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-ter
25、ial with the highest predicted transition temperature usingthe projected fluence at the end of design life of each material,determined by adding the appropriate transition temperatureshift (TTS) to the unirradiated RTNDT. The TTS can bedetermined from the relationship found in Guide E900 or othersou
26、rces, including regulations.3.1.12 maximum design fluence (MDF)the maximum 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 bee
27、n charac-terized as to the sensitivity of its tensile, impact and fracturetoughness 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
28、, SectionIII, for the definition of RTNDTfor unirradiated material basedon 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 Me
29、thodE1921).3.1.15 standby capsulea surveillance capsule meeting therecommendations of this practice that is in the reactor vesselirradiation location as defined by Practice E185, but the testingof which is not required by this practice.3.2 Neutron Exposure Terminology:3.2.1 Definitions of terms rela
30、ted to neutron 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 accou
31、nt for these 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.4.2 Prior to the first issue date of th
32、is 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 tothis standard, Practice E185 was revised many times (1966,1970, 1973, 1979, 1982, 1993 a
33、nd 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. Forclarity, the standard practice for surveillance programs hasbeen divided into the new Pract
34、ice 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 supplementaltests are described in Guide E636.4.3 This practice is intended to cover testing and evalua
35、tionof all light-water moderated reactor pressure vessel surveil-lance capsules. The practice is applicable to testing of capsulesE2215 152from surveillance programs designed and implemented underall previous versions of Practice E185.4.4 The radiation-induced changes in the properties of thereactor
36、 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-induced changes isdescribed in Practice E185.4.5 Alternative methods exist for testi
37、ng 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 Temperature method defined in Test MethodE1921 or J-integral techniques defined in Test Met
38、hod E1820.Additionally, hardness testing can be used to supplementstandard methods as a means of monitoring the irradiationresponse of the materials.4.6 The methodology to be used in the analysis and inter-pretation of neutron dosimetry data and the determination ofneutron fluence is defined in Prac
39、tice E853.4.7 Guide E900 describes the bases used to evaluate theradiation-induced changes in Charpy transition temperature forreactor vessel materials and provides a methodology forpredicting future values.4.8 Guide E509 provides direction for development of aprocedure for conducting an in-service
40、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 ExaminationA complete visual exam of thecapsule condition should be complet
41、ed 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 specimen loading pattern shouldbe compared to the capsule fabrication records
42、 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 average capsuletemperature during full power operation shall be estimated fore
43、ach 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 difference. In atypical pressurized water reactor, the coolant inlet temperature
44、may 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 irradiation temperature.5.4 Peak TemperatureTemperature monitors shall be ex-am
45、ined 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. Other data that areneeded on a fuel-cycle-specific basis include: assembly-wis
46、ecore 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 include the additionor removal of flux suppression rods or shield rods, uprates
47、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 a
48、ndmovements: including storage periods outside the reactor, shallbe provided for the evaluation of irradiation exposure.6.2 The neutron fluence rate, neutron energy spectrum andneutron fluence of the surveillance specimens and the corre-sponding maximum values for the reactor vessel shall bedetermin
49、ed in accordance with Practices E853.6.3 Neutron fluence rate and fluence values (E 1 MeV)and dpa rate and dpa values per Practice E693 shall bedetermined and recorded using a calculated spectrum adjustedor validated by dosimetry measurements.7. Measurement of Mechanical Properties7.1 Generally, all the materials contained in the capsuleexcept the HAZ specimens (if included) should be tested.Testing of the HAZ specimens is optional.57.2 Tension Tests:7.2.1 MethodTension testing shall be conducted in accor-dance with Test Methods E8/E8M and E21.7.2.2 Test Temp
