ASTM E185-2016 red 4415 Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels《轻水冷却核反应堆容器的监督程序设计的标准实施规程》.pdf

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1、Designation: E185 151E185 16Standard 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 NOTEParagraph X1.8 was corrected editorially in October 2015.1. Scope1.1 This practice covers procedures

3、for designing a surveillance program for monitoring the radiation-induced changes in themechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. New advanced light-water smallmodular reactor designs with a nominal design output of 300 MWe or less have not b

4、een specifically considered in this practice.This practice includes 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 nuclea

5、r power reactor vessels for which the predicted maximumfast neutron fluence (E 1 MeV) exceeds 1 1021 neutrons/m2 (1 1017 n/cm2) at the inside surface of the ferritic steel reactorvessel.1.3 This practice does not provide specific procedures for monitoring the radiation induced changes in properties

6、beyond thedesign life. Practice E2215 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

7、 light-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 eval

8、uation of surveillance capsules removed from a reactor vessel. Guide E636 provides guidancefor conducting additional mechanical tests.Asummary of the many major revisions to Practice E185 since its original issuance is contained in AppendixX1.NOTE 2This practice applies only to the planning and desi

9、gn 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 Documents2.1 ASTM Standards:2A370 Test Methods and Definitions for Mechanical Testing of Ste

10、el 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 Metallic MaterialsE23 Test Methods for Notched Bar Impact Testing of Metallic Materi

11、alsE170 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 Neutron Transport Methods for Reactor Vessel SurveillanceE636 Guide for Conducting Su

12、pplemental 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 of Light-Water Reactor Surveillance ResultsE900 Guide for Predicting Radiation-Induced Transi

13、tion Temperature Shift in Reactor Vessel MaterialsE1214 Guide for Use of Melt Wire Temperature Monitors for Reactor Vessel Surveillance, E 706 (IIIE)E1253 Guide for Reconstitution of Irradiated Charpy-Sized Specimens1 This practice is under the jurisdiction of ASTM Committee E10 on Nuclear Technolog

14、y and Applications and is the direct responsibility of Subcommittee E10.02 onBehavior and Use of Nuclear Structural Materials.Current edition approved June 1, 2015Dec. 1, 2016. Published July 2015December 2016. Originally approved in 1961 as E185 61 T. Last previous edition approved in20102015 as E1

15、85 10.E185 151. DOI: 10.1520/E0185-15E01.10.1520/E0185-16.2 For referencedASTM 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.Thi

16、s document is not an ASTM standard and is intended only to provide the user of 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 editio

17、ns as appropriate. In all cases only the current versionof the standard as published 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 States1E1820 Test Method for Measurement of Fracture Tough

18、nessE1921 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 Reactor VesselsE2298 Test Method for Instrumented Impact Testing of Metallic MaterialsE2956

19、 Guide for Monitoring the Neutron Exposure of LWR Reactor Pressure Vessels2.2 ASME Standards:3Boiler and Pressure Vessel Code, Section III Subsection NB-2000Boiler and Pressure Vessel Code, Section XI Nonmandatory Appendix A, Analysis of Flaws, and Nonmandatory Appendix G,Fracture Toughness Criteria

20、 for Protection Against Failure3. Terminology3.1 Definitions:3.1.1 base metalas-fabricated plate 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 fo

21、rgings) thatdirectly surrounds the effective height of the active core. Note that materials in 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-fitte

22、d presentation, or both, of absorbed energy, lateralexpansion, or fracture appearance as functions of test temperature, extending over a range including the lower shelf (5 % or lessshear fracture appearance), transition region, and the upper shelf (95 % or greater shear fracture appearance).3.1.4 Ch

23、arpy transition temperature shiftthe difference in the 41 J (30 ftlbf) index temperatures for the best fit (average)Charpy absorbed energy curve measured before and after irradiation. Similar measures of temperature shift can be defined basedon other indices in 3.1.3, but the current industry practi

24、ce is to use 41 J (30 ftlbf) and is consistent with Guide E900.3.1.5 Charpy upper-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 (

25、150F) abovethe Charpy upper-shelf onset shall not be included, unless no data are available between the onset temperature and onset +83C(+150F).3.1.6 Charpy upper-shelf onsetthe temperature at which the fracture appearance of all Charpy specimens tested is at or above95 % shear.3.1.7 heat-affected-z

26、one (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 temperaturethe temperature corresponding to a predetermined level of absorbed energ

27、y, lateral expansion, orfracture appearance obtained from the best-fit (average) Charpy transition curve.3.1.9 lead factorthe ratio of the average neutron fluence (E 1 MeV) of the specimens in a surveillance capsule to the peakneutron fluence (E 1 MeV) of the corresponding material at the ferritic s

28、teel reactor pressure vessel inside surface calculated overthe same time period.3.1.9.1 DiscussionChanges in the reactor operating parameters or fuel management may cause the lead factor to change.3.1.10 limiting materialstypically the weld and base material with the highest predicted transition tem

29、perature using theprojected fluence at the end of design life of each material determined by adding the appropriate transition temperature shift to theunirradiated RTNDT. Materials that are projected to most closely approach a regulatory limit at the end of the design life shouldbe considered in sel

30、ecting the limiting material. The transitionGuide E900 temperature shift can be determined from therelationship found in Guidedescribes a method for predicting the Transition Temperature Shift (TTS). Regulators E900or othersources, including regulations. sources may describe different methods for pr

31、edictingTTS.The basis for selecting the limiting weldand base materials shall be documented.3.1.11 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 is defined at the interface of the cladding to the

32、 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 toughnessproperties to neutron radiation-induced embrittlement.3 Av

33、ailable from the American Society of Mechanical Engineers, Third Park Avenue, New York, NY 10016.E185 1623.1.13 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

34、material based on Charpy (Test Method A370)and drop weight tests (Test Method E208). ASME Code Section XI, Appendices A and G provide an alternative definition for thereference temperature (RTTo) based on fracture toughness properties (Test Method E1921)3.1.14 standby capsulea surveillance capsule m

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

36、erminology 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 account

37、 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 pra

38、ctice 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 of

39、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 methodologyGuides E482 to beand E853 used in describe a methodology for estimation of neutron exposure obtaine

40、dfor reactor vessel surveillance programs is defined in Guidesprograms. Regulators or other sources E482 andmay describeE853.different methods.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 m

41、aterials used for that reactor vessel. The amount of such data and the similarity of exposure conditionsand 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

42、 a surveillance program for monitoring theradiation-induced changes in the mechanical properties of the ferritic materials that compose the reactor vessel.5.2 Surveillance Materials:5.2.1 Materials SelectionThe surveillance materials shall include, at minimum, the limiting base metal and the limitin

43、g weld.If a limiting material is outside the beltline, the limiting beltline base and weld materials shall also be included. If there is nobeltline weld, capsules whose target fluence (Table 1) is greater than two times the design fluence of the limiting weld are notrequired to contain weld metal, e

44、xcept that the first capsule must contain the limiting weld material.NOTE 3The predicted limiting material may change during operation due to changes that may occur in the transition temperature shift predictionformulation, or other factors. Therefore, it is prudent to include additional potentially

45、 limiting materials in the surveillance program as capsule spacepermits.5.2.2 Material SamplingA minimum surveillance 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

46、(s) made with the sameheat of weld wire and lot of flux and by the same welding procedure as that used for the reactor vessel welds. The base metalsused to form the weld shall be from the reactor vessel. If a reactor vessel weld is contained in the beltline, it is recommended thatat least one of the

47、 base metals used to fabricate the weldment(s) shall be a base metal beltline material included in the surveillanceprogram. Surveillance test specimens shall be removed from full reactor vessel thickness samples.5.2.2.1 Any non-actual reactor vessel base metal used to fabricate the surveillance weld

48、 shall have a similar nominal chemicalcomposition and thickness to the reactor vessel base metal. Any archived non-actual reactor vessel base metal(s) used to fabricatethe surveillance weldment(s) shall be clearly marked with “Not actual reactor vessel base metal - Do not use for base metal or HAZsp

49、ecimen” or similar, and shall be clearly identified in the associated documentation as non-actual reactor vessel base metal.NOTE 4Experience has shown that it is no longer necessary to include the HAZ material in the surveillance program. However, it is recommendedTABLE 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 RequiredE185 163that the HAZ material be included with the archive ma