ASTM E185-2015 red 2146 Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels《轻水慢化核电反应堆容器监督程序设计的标准实施规程》.pdf

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1、Designation: E185 10E185 15Standard Practice forDesign of Surveillance Programs for Light-Water ModeratedNuclear Power Reactor Vessels1This standard is issued under the fixed designation E185; 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 procedures for designing a surveillance program for monitoring the radiatio

3、n-induced changes in themechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. New advanced light-water smallmolecular reactor designs with a nominal design output of 300 MWe or less have not been specifically considered in this practice.This practice inc

4、ludes the minimum requirements for the design of a surveillance program, selection of vessel material to beincluded, and the initial schedule for evaluation of materials.1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximumfast ne

5、utron fluence (E 1 MeV) at the end of license (EOL) exceeds 1 1021 neutrons/m2 (1 1017 n/cm2) at the inside surfaceof the ferritic steel reactor vessel.1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built afterthe effective date of

6、 this practice. Previous versions of Practice E185 apply to earlier reactor vessels.1.3 This practice does not provide specific procedures for monitoring the radiation induced changes in properties beyond thedesign life, butlife. Practice E2215the procedure described may provide guidance for develop

7、ing such a surveillance program.addresses changes to the withdrawal schedule during and beyond the design life.1.4 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.NOTE 1The increased complexity of the requirements for a light

8、-water moderated nuclear power reactor vessel surveillance program has necessitatedthe separation of the requirements into three related standards. Practice E185 describes the minimum requirements for design of a surveillance program.Practice E2215 describes the procedures for testing and evaluation

9、 of surveillance capsules removed from a surveillance program as defined in the currentor previous editions of Practice reactor vessel. E185. Guide E636 provides guidance for conducting additional mechanical tests. A summary of the manymajor revisions to Practice E185 since its original issuance is

10、contained in Appendix X1.NOTE 2This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective dateof this practice. Previous versions of Practice E185 apply to earlier reactor vessels. See Appendix X1.2. Referenced Documents2

11、.1 ASTM Standards:2A370 Test Methods and Definitions for Mechanical Testing of Steel ProductsA751 Test Methods, Practices, and Terminology for Chemical Analysis of Steel ProductsE8/E8M Test Methods for Tension Testing of Metallic MaterialsE21 Test Methods for Elevated Temperature Tension Tests of Me

12、tallic MaterialsE23 Test Methods for Notched Bar Impact Testing of Metallic MaterialsE170 Terminology Relating to Radiation Measurements and DosimetryE208 Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic SteelsE482 Guide for Application of Neu

13、tron Transport Methods for Reactor Vessel Surveillance, E706 (IID)E636 Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E 706 (IH)E844 Guide for Sensor Set Design and Irradiation for Reactor Surveillance, E 706 (IIC)E853 Practice for Analysis and Interpretation

14、 of Light-Water Reactor Surveillance ResultsE900 Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials1 This practice is under the jurisdiction of ASTM Committee E10 on Nuclear Technology and Applications and is the direct responsibility of Subcommittee E10.

15、02 onBehavior and Use of Nuclear Structural Materials.Current edition approved March 1, 2010June 1, 2015. Published April 2010July 2015. Originally approved in 1961 as E185 61 T. Last previous edition approved in20022010 as E185 02.E185 10. DOI: 10.1520/E0185-10.10.1520/E0185-15.2 For referencedASTM

16、 standards, visit theASTM 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

17、 an ASTM standard an indication 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 pub

18、lished by ASTM is to be considered the official document.Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States1E1214 Guide for Use of Melt Wire Temperature Monitors for Reactor Vessel Surveillance, E 706 (IIIE)E1253 Guide for Reconstitution

19、 of Irradiated Charpy-Sized SpecimensE1820 Test Method for Measurement of Fracture ToughnessE1921 Test Method for Determination of Reference Temperature, To, for Ferritic Steels in the Transition RangeE2215 Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Rea

20、ctor VesselsE2298 Test Method for Instrumented Impact Testing of Metallic MaterialsE2956 Guide for Monitoring the Neutron Exposure of LWR Reactor Pressure Vessels2.2 ASME Standards:3American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, SectionsSection III and XI Subsection NB-20

21、00ASME Boiler and Pressure Vessel Code Case N-629,Code, Section XI Use of Fracture Toughness Test Data to EstablishReference Temperature for Pressure Retaining Materials, Section XI, Division 1Nonmandatory Appendix A, Analysis ofFlaws, and Nonmandatory Appendix G, Fracture Toughness Criteria for Pro

22、tection Against FailureASME Boiler and Pressure Vessel Code Case N-631, Use of Fracture Toughness Test Data to Establish Reference Temperaturefor Pressure Retaining Materials Other Than Bolting for Class 1 Vessels, Section III, Division 13. Terminology3.1 Definitions:3.1.1 base metalas-fabricated pl

23、ate material or forging material 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 and plates or forgings) thatdirectly surrounds the effective height of the active core. Note that materials i

24、n regions adjacent to the beltline may sustainsufficient neutron damage to warrant consideration in the selection of surveillance materials.3.1.3 Charpy transition temperature curvea graphic or a curve-fitted presentation, or both, of absorbed energy, lateralexpansion, andor fracture appearance as f

25、unctions of test temperature, extending over a range including the lower shelf (5 % orless shear fracture appearance), transition region, and the upper shelf (95 % or greater shear fracture appearance).3.1.4 Charpy transition temperature shiftthe difference in the 40.741 J (30 ft-lbf)ftlbf) index te

26、mperatures for the best fit(average) Charpy absorbed energy curve measured before and after irradiation. Similar measures of temperature shift can bedefined based on other indices in 3.1.3, but the current industry practice is to use 41 J (30 ftlbf) and is consistent with Guide E900.3.1.5 Charpy upp

27、er-shelf energy levelthe average energy value for all Charpy specimen tests (preferably three or more) whosetest temperature is at or above the Charpy upper-shelf onset; specimens tested at temperatures greater than 83C (150F) abovethe Charpy upper-shelf onset shall not be included, unless no data a

28、re available between the onset temperature and onset +83C(+150F).3.1.6 Charpy upper-shelf onsetthe test temperature aboveat which the fracture appearance of all Charpy specimens tested isat or above 95 % shear.3.1.7 end-of-license (EOL)the design lifetime in terms of years corresponding to the opera

29、ting license period.3.1.7 heat-affected-zone (HAZ)plate material or forging material extending outward from, but not including, the weld fusionline in which the microstructure of the base metal has been altered by the heat of the welding process.3.1.8 index temperaturethatthe temperature correspondi

30、ng to a predetermined level of absorbed energy, lateral expansion, orfracture appearance obtained from the best-fit (average) Charpy transition curve.3.1.9 lead factorthe ratio of the peakaverage neutron fluence(E 1 MeV) of the specimens in a surveillance capsule to the peak neutron fluence (E 1 MeV

31、) of the corresponding materialat the ferritic steel reactor pressure vessel inside surface.surface calculated over the same time period.3.1.9.1 DiscussionChanges in the reactor operating parameters andor fuel management may cause the lead factor to change.3.1.10 limiting materialstypically the weld

32、 and base material with the highest predicted transition temperature at EOL usingthe projected fluence at the end of design life of each material determined by adding the appropriate transition temperature shiftto the unirradiated RTNDT. The referenceMaterials that are projected to most closely appr

33、oach a regulatory limit at the end of thedesign life should be considered in selecting the limiting material. The transition temperature shift can be determined from therelationship found in Guide E900. or other sources, including regulations. The basis for selecting the limiting material weld andba

34、se materials shall be documented.3 Available from the American Society of Mechanical Engineers, Third Park Avenue, New York, NY 10016.E185 1523.1.11 maximum design fluence (MDF)the maximum projected fluence at the inside surface of the ferritic pressure vessel atthe end of design life (if clad, MDF

35、is defined at the interface of the cladding to the ferritic steel). Changes during operation willaffect the projected fluence and are addressed in Practice E2215.3.1.12 reference materialany steel that has been characterized as to the sensitivity of its tensile, impact and fracture toughnessproperti

36、es to neutron radiationradiation-induced embrittlement.3.1.13 reference temperature (RTNDT) see subarticle NB-2300 of the ASME Boiler and Pressure Vessel Code, Section III,“Nuclear Power Plant Components” for the definition of RTNDT for unirradiated material based on Charpy (Test Method E23A370)and

37、drop weight tests (Test Method E208). ASME Code Cases N-629 and N-631Section XI, Appendices A and G provide analternative definition for the reference temperature (RTTo) based on fracture toughness properties (Test Method E1921)3.1.14 standby capsulea surveillance capsule meeting the recommendations

38、 of this practice that is in the reactor vesselirradiation location, but whose withdrawal location as defined by Practice E185, but the testing of which is not required by thispractice.3.2 Neutron Exposure Terminology:3.2.1 Definitions of terms related to neutron dosimetry and exposure are provided

39、in Terminology E170.4. Significance and Use4.1 Predictions of neutron radiation effects on pressure vessel steels are considered in the design of light-water moderatednuclear power reactors. Changes in system operating parameters often are made throughout the service life of the reactor vesselto acc

40、ount for radiation effects. Due to the variability in the behavior of reactor vessel steels, a surveillance program is warrantedto monitor changes in the properties of actual vessel materials caused by long-term exposure to the neutron radiation andtemperature environment of the reactor vessel. This

41、 practice describes the criteria that should be considered in planning andimplementing surveillance test programs and points out precautions that should be taken to ensure that: (1) capsule exposures canbe related to beltline exposures, (2) materials selected for the surveillance program are samples

42、 of those materials most likely tolimit the operation of the reactor vessel, and (3) the test specimen types are appropriate for the evaluation of radiation effects onthe reactor vessel.4.2 The methodology to be used in estimation of neutron exposure obtained for reactor vessel surveillance programs

43、 is definedin GuideGuides E482 and E853.4.3 The design of a surveillance program for a given reactor vessel must consider the existing body of data on similar materialsin addition to the specific materials used for that reactor vessel. The amount of such data and the similarity of exposure condition

44、sand material characteristics will determine their applicability for predicting radiation effects.5. Surveillance Program Design5.1 This section describes the minimum requirements for the design of a surveillance program for monitoring theradiation-induced changes in the mechanical properties of the

45、 ferritic materials in that compose the reactor vessel.5.2 TestSurveillance Materials:5.2.1 Materials SelectionThe surveillance test materials shall include, at minimum, the limiting base metal heat and thelimiting weld. If a limiting material is outside the beltline, the limiting beltline base and

46、weld materials shall also be included. Ifthere is no beltline weld, capsules whose target fluence (Table 1) is greater than two times the design fluence of the limiting weldare not required to contain weld metal, except that the first capsule must contain the limiting weld material.NOTE 3The predict

47、ed limiting material may change during operation. operation due to changes that may occur in the transition temperature shiftTABLE 1 Suggested Withdrawal ScheduleSequence Target Fluence NotesFirst 14 EOL ID Testing RequiredSecond 12 EOL ID Testing Required ifprojectedRTNDT 111C (200F)Third 34 EOL ID

48、 Testing RequiredFourth EOL ID Testing RequiredStandby 2 EOL ID Testing Not RequiredTABLE 1 Recommended Withdrawal ScheduleSequence Target Fluence NotesFirst 14 MDF Testing RequiredSecond 12 MDF Testing RequiredThird 34 MDF Testing RequiredFourth MDF Testing RequiredStandby 2 MDF Testing Not Require

49、dE185 153prediction formulation, or other factors. Therefore, it is prudent to include additional potentially limiting beltline materials in the surveillance programas capsule space permits.5.2.2 Material SamplingA minimum testsurveillance program shall consist of the material selected in 5.2.1, taken from thefollowing: (1) base metal from the actual plate(s) or forging(s) used in the reactor vessel, and (2) weld metal(s) made with the sameheat of weld wire and lot of flux and by the same welding procedure as that used for the reacto