ASTM E185-2002 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操作规程》.pdf

1、Designation: E 185 02Standard Practice forDesign of Surveillance Programs for Light-Water ModeratedNuclear Power Reactor Vessels1This standard is issued under the fixed designation E 185; the number immediately following the designation indicates the year oforiginal adoption or, in the case of revis

2、ion, the year of last revision. A number in parentheses indicates the year of last reapproval. Asuperscript epsilon (e) indicates an editorial change since the last revision or reapproval.1. Scope1.1 This practice covers procedures for designing a surveil-lance program for monitoring the radiation-i

3、nduced changes inthe mechanical properties of ferritic materials in the beltline oflight-water moderated nuclear power reactor vessels. Thispractice includes the minimum requirements for the design ofa surveillance program, selection of vessel material to beincluded, and a schedule for evaluation of

4、 materials.1.2 This practice was developed for all light-water moder-ated nuclear power reactor vessels for which the predictedmaximum fast neutron fluence (E 1 MeV) at the end of thedesign lifetime (EOL) exceeds 1 3 1017n/cm2(1 3 1021n/m2)at the inside surface of the reactor vessel.1.3 This practic

5、e applies only to the planning and design ofsurveillance programs for reactor vessels designed and builtafter the effective date of this practice. Previous versions ofPractice E 185 apply to earlier reactor vessels.1.4 This practice does not provide specific procedures formonitoring the radiation in

6、duced changes in properties beyondthe design life, but the procedure described may provideguidance for developing such a surveillance program.NOTE 1The increased complexity of the requirements for a light-water moderated nuclear power reactor vessel surveillance program hasnecessitated the separatio

7、n of the requirements into three related stan-dards. Practice E 185 describes the minimum requirements for a surveil-lance program. Practice E 2215, “Standard Practice for the Evaluation ofSurveillance Capsules from Light-Water Moderated Nuclear Power Re-actor Vessels” describes the procedures for t

8、esting and evaluation ofsurveillance capsules removed from a surveillance program as defined inthe current or previous editions of Practice E 185. Another standard guidefor supplementing existing light-water moderated nuclear power reactorvessel surveillance programs is under preparation. A summary

9、of the manymajor revisions to Practice E 185 since its original issuance is containedin Appendix X1.2. Referenced Documents2.1 ASTM Standards:A 370 Test Methods and Definitions for Mechanical Testingof Steel Products2A 751 Test Methods, Practices and Terminology for Chemi-cal Analysis of Steel Produ

10、cts2E 8 Test Methods for Tension Testing of Metallic Materials3E 21 Test Methods for Elevated Temperature Tension Testsof Metallic Materials3E 23 Test Methods for Notched Bar Impact Testing ofMetallic Materials3E 170 Terminology Relating to Radiation Measurementsand Dosimetry4E 208 Test Method for C

11、onducting Drop-Weight Test toDetermine Nil-Ductility Transition Temperature of FerriticSteels3E 399 Test Method for Plane-Strain Fracture Toughness ofMetallic Materials3E 482 Guide for Application of Neutron Transport Methodsfor Reactor Vessel Surveillance, E 706 (IID)4E 560 Practice for Extrapolati

12、ng Reactor Vessel Surveil-lance Dosimetry Results, E 706 (IC)4E 636 Practice for Conducting Supplemental SurveillanceTests for Nuclear Power Reactor Vessels, E 706 (IH)4E 693 Practice for Characterizing Neutron Exposure inFerritic Steels in Terms of Displacements per Atom (DPA),(ID)4E 844 Guide for

13、Sensor Set Design and Irradiation forReactor Surveillance, E 706 (IIC)4E 853 Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results, E 706 (IA)4E 900 Guide for Predicting Radiation-Induced TransitionTemperature Shift in Reactor Vessel Materials, E 706(IIF)4E 1214 Guide

14、for Use of Melt Wire Temperature Monitorsfor Reactor Vessel Surveillance, E 706 (IIIE)4E 1253 Guide for Reconstitution of Irradiated Charpy SizeSpecimens4E 1820 Test Method for Measurement of Fracture Tough-ness31This practice is under the jurisdiction of ASTM Committee E10 on NuclearTechnology and

15、Applications and is the direct responsibility of SubcommitteeE10.02 on Behavior and Use of Structural Materials.Current edition approved June 10, 2002. Published September 2002. Originallypublished as E 185 61 T. Last previous edition E 185 98.2Annual Book of ASTM Standards, Vol 01.03.3Annual Book o

16、f ASTM Standards, Vol 03.01.4Annual Book of ASTM Standards, Vol 12.02.1Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.E 1921 Test Method for the Determination of ReferenceTemperature, To, for Ferritic Steels in the TransitionRange3E

17、2215 Practice for the Evaluation of Surveillance Capsulesfrom Light-Water Moderated Nuclear Power Reactor Ves-sels2.2 Other Documents:American Society of Mechanical Engineers, Boiler andPressure Vessel Code, Sections III and XI5ASME Boiler and Pressure Vessel Code Case N-629, Use ofFracture Toughnes

18、s Test Data to Establish ReferenceTemperature for Pressure Retaining Materials, Section XI,Division 15ASME Boiler and Pressure Vessel Code Case N-631, Use ofFracture Toughness Test Data to Establish ReferenceTemperature for Pressure Retaining Materials Other ThanBolting for Class 1 Vessels, Section

19、III, Division 153. Terminology3.1 Definitions:3.1.1 adjusted reference temperature (ART)the referencetemperature adjusted for irradiation effects by adding to theinitial RTNDT, the transition temperature shift (for example, seeGuide E 900), and an appropriate margin to account foruncertainties.3.1.2

20、 base metal (parent material)as-fabricated plate ma-terial or forging material other than a weld or its correspondingheat-affected-zone (HAZ).3.1.3 beltlinethe irradiated region of the reactor vessel(shell material including weld seams and plates or forgings)that directly surrounds the effective hei

21、ght of the active core,and adjacent regions that are predicted to sustain sufficientneutron damage to warrant consideration in the selection ofsurveillance material.3.1.4 Charpy transition regionthe region on the Charpytransition curve in which toughness increases rapidly withrising temperature; in

22、terms of fracture appearance, it ischaracterized by a change from a primarily cleavage (crystal-line) fracture mode to a primarily shear (fibrous) fracturemode.3.1.5 Charpy transition temperature curvea graphic pre-sentation of Charpy data, including absorbed energy, lateralexpansion, and fracture a

23、ppearance as functions of test tem-perature, extending over a range including the lower shelfenergy (5 % or less shear fracture appearance), transitionregion, and the upper-shelf energy (95 % or greater shearfracture appearance).3.1.6 Charpy transition temperature shiftthe difference inthe 30 ft-lbf

24、 (41J) index temperatures for the best fit (average)Charpy curve measured before and after irradiation.3.1.7 Charpy upper-shelf energy levelthe average energyvalue for all Charpy specimen tests (normally three) whose testtemperature is above the Charpy upper shelf onset; specimenstested at temperatu

25、res greater than 150F (83C) above theCharpy upper-shelf onset need not be included. The range oftest temperatures for which energy values were averaged mustbe reported as well as the individual energy values. Forspecimens tested in sets of three at each test temperature, theset having the highest av

26、erage may be regarded as defining theupper-shelf energy.3.1.8 Charpy upper shelf onsetthe test temperature abovewhich the fracture appearance of all Charpy specimens testedis nominally 100 % shear. Specimens with 95 % or greatershear may be included in this determination.3.1.9 end-of-life (EOL)the d

27、esign lifetime in terms ofyears corresponding to the operating license period.3.1.10 fracture strengthin a tensile test, the measuredforce at fracture divided by the initial cross-sectional area ofthe test specimen.3.1.11 fracture stressin a tensile test, the measured forceat fracture divided by the

28、 cross-sectional area of the testspecimen at the time of fracture.3.1.12 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.3.1.13 in

29、dex temperaturethat temperature correspondingto a predetermined level of absorbed energy, lateral expansion,or fracture appearance obtained from the best-fit (average)Charpy transition curve.3.1.14 lead factorthe ratio of the neutron fluence rate(E 1 MeV) at the specimens in a surveillance capsule t

30、o theneutron fluence rate (E 1 MeV) at the reactor pressure vesselinside surface peak fluence location.NOTE 2Changes in the reactor operating parameters and fuel man-agement may cause the lead factor to change.3.1.15 nil-ductility transition temperature (TNDT)themaximum temperature at which a standa

31、rd drop weight speci-men breaks when tested in accordance with Test MethodE 208.3.1.16 reference materialany steel that has been charac-terized as to the sensitivity of its mechanical and fracturetoughness properties to neutron radiation embrittlement.3.1.17 reference temperature (RTNDT)see subartic

32、le NB-2300 of the ASME Boiler and Pressure Vessel Code, SectionIII, “Nuclear Power Plant Components” for the definition ofRTNDTfor unirradiated material. ASME Code Cases N-629and N-631 provide an alternative definition for the referencetemperature (RTTo).3.2 Neutron Exposure Terminology:3.2.1 Defini

33、tions of terms related to neutron dosimetry andexposure are provided in Terminology E 170.4. Significance and Use4.1 Predictions of neutron radiation effects on pressurevessel steels are considered in the design of light-watermoderated nuclear power reactors. Changes in system operat-ing parameters

34、often are made throughout the service life of thereactor vessel to account for radiation effects. Due to thevariability in the behavior of reactor vessel steels, a surveil-lance program is warranted to monitor changes in the proper-ties of actual vessel materials caused by long-term exposure tothe n

35、eutron radiation and temperature environment of the5Available from the American Society of Mechanical Engineers, Third ParkAvenue, New York, NY 10016.E185022reactor vessel. This practice describes the criteria that shouldbe considered in planning and implementing surveillance testprograms and points

36、 out precautions that should be taken toensure that: (1) capsule exposures can be related to beltlineexposures, (2) materials selected for the surveillance programare samples of those materials most likely to limit the operationof the reactor vessel, and (3) the tests yield results useful forthe eva

37、luation of radiation effects on the reactor vessel.4.2 The methodology to be used in estimation of neutronexposure obtained for reactor vessel surveillance programs isdefined in Guide E 482, which establishes the bases to be usedto evaluate both the design and future condition of the reactorvessel.4

38、.3 The design of a surveillance program for a given reactorvessel must consider the existing body of data on similarmaterials in addition to the specific materials used for thatreactor vessel. The amount of such data and the similarity ofexposure conditions and material characteristics will determin

39、etheir applicability for predicting radiation effects.5. Surveillance Program Design5.1 This section describes the minimum requirements forthe design of a surveillance program for monitoring theradiation-induced changes in the mechanical properties offerritic materials in the reactor vessel beltline

40、 region.5.2 Test Materials5.2.1 Materials Selection:5.2.1.1 Surveillance test materials shall be full thicknesssamples taken from each of the actual materials used infabricating the beltline of the reactor vessel or from weldmentsfabricated to match the reactor vessel weld(s). These surveil-lance te

41、st materials shall include a minimum of one heat of thebase metal and one weld.NOTE 3If there is no weld in the beltline, then there is no requirementto include weld material in the surveillance program. However, it may beprudent to include the weld with the highest projected EOL DRTNDTvalue.5.2.1.2

42、 The base metal and weld metal materials included inthe program shall be those predicted to be most limiting foroperation of the reactor to compensate for radiation effectsduring its lifetime. The beltline materials shall be evaluated onthe basis of adjusted reference temperature. The predictedchang

43、es in the initial properties as a function of chemicalcomposition and the neutron fluence during reactor operationshall be determined in accordance with Guide E 900. The basemetal and the weld with the highest adjusted reference tem-perature at end-of-life shall be selected for the surveillanceprogr

44、am.5.2.1.3 The adjusted reference temperature of each materialin the reactor vessel beltline shall be determined by adding theappropriate value of transition temperature shift to the refer-ence temperature of the unirradiated material plus an appro-priate margin. The reference temperature shift can

45、be deter-mined from the relationship found in Guide E 900.5.2.2 Material SamplingA minimum test program shallconsist of the material selected in 5.2.1, taken from thefollowing locations: (1) base metal from each plate or forgingused in the beltline, and (2) each weld metal made with thesame heat of

46、weld wire and lot of flux and by the same weldingprocedure as that used for each of the beltline welds. The basemetal used to fabricate the weldment shall be one of the basemetals included in the surveillance program.NOTE 4Experience has shown that it is no longer necessary to includethe heat-affect

47、ed zone material in the surveillance program. However, it isrecommended that the heat-affected-zone material be included with thearchives (see 5.2.5).5.2.3 Fabrication HistoryThe fabrication history (austen-itizing, quench and tempering, and post-weld heat treatment) ofthe test materials shall be fu

48、lly representative of the fabricationhistory of the materials in the beltline of the reactor vessel andshall be recorded.5.2.4 Chemical Analysis RequirementsThe chemicalanalysis required by the appropriate product specifications forthe surveillance test materials (base metal and as-depositedweld met

49、al) shall be recorded and shall include phosphorus (P),sulfur (S), copper (Cu), vanadium (V), silicon (Si), manganese(Mn), and nickel (Ni), as well as all other alloying and residualelements commonly analyzed for in low-alloy steel products.The product analysis shall be as described in Test MethodA 751 and verified by analyzing samples selected from thebase metal and the as-deposited weld metal.5.2.5 Archive MaterialsTest stock to fill up to six addi-tional capsules with test specimens of the base metal and weldmaterials used in the program shall be retained