ASTM E900-2015 1079 Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials《预测在反应堆容器材料中辐射感应变化导致温度漂移的标准指南》.pdf

《ASTM E900-2015 1079 Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials《预测在反应堆容器材料中辐射感应变化导致温度漂移的标准指南》.pdf》由会员分享，可在线阅读，更多相关《ASTM E900-2015 1079 Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials《预测在反应堆容器材料中辐射感应变化导致温度漂移的标准指南》.pdf（4页珍藏版）》请在麦多课文档分享上搜索。

1、Designation: E900 15Standard Guide forPredicting Radiation-Induced Transition Temperature Shiftin Reactor Vessel Materials1This standard is issued under the fixed designation E900; the number immediately following the designation indicates the year oforiginal adoption or, in the case of revision, th

2、e 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 guide presents a method for predicting values ofreference transition temperature shift (TTS) for irradi

3、atedpressure vessel materials. The method is based on the TTSexhibited by Charpy V-notch data at 41-J (30-ftlbf) obtainedfrom surveillance programs conducted in several countries forcommercial pressurized (PWR) and boiling (BWR) light-watercooled (LWR) power reactors. An embrittlement correlationhas

4、 been developed from a statistical analysis of the largesurveillance database consisting of radiation-induced TTS andrelated information compiled and analyzed by SubcommitteeE10.02. The details of the database and analysis are describedin a separate report (1).2,3,This embrittlement correlation wasd

5、eveloped using the variables copper, nickel, phosphorus,manganese, irradiation temperature, neutron fluence, and prod-uct form. Data ranges and conditions for these variables arelisted in 1.1.1. Section 1.1.2 lists the materials included in thedatabase and the domains of exposure variables that mayi

6、nfluence TTS but are not used in the embrittlement correlation.1.1.1 The range of material and irradiation conditions inthe database for variables used in the embrittlement correla-tion:1.1.1.1 Copper content up to 0.4 %.1.1.1.2 Nickel content up to 1.7 %.1.1.1.3 Phosphorus content up to 0.03 %.1.1.

7、1.4 Manganese content within the range from 0.55 to 2.0%.1.1.1.5 Irradiation temperature within the range from 255 to300C (491 to 572F).1.1.1.6 Neutron fluence within the range from11021n/m2to21024n/m2(E 1 MeV).1.1.1.7 A categorical variable describing the product form(that is, weld, plate, forging)

8、.1.1.2 The range of material and irradiation conditions inthe database for variables not included in the embrittlementcorrelation:1.1.2.1 A533 Type B Class 1 and 2, A302 Grade B, A302Grade B (modified), and A508 Class 2 and 3. Also, Europeanand Japanese steel grades that are equivalent to these ASTM

9、Grades.1.1.2.2 Submerged arc welds, shielded arc welds, and elec-troslag welds having compositions consistent with those of thewelds used to join the base materials described in 1.1.2.1.1.1.2.3 Neutron fluence rate within the range from31012n/m2/sto51016n/m2/s (E 1 MeV).1.1.2.4 Neutron energy spectr

10、a within the range expected atthe reactor vessel region adjacent to the core of commercialPWRs and BWRs (greater than approximately 500MW elec-tric).1.1.2.5 Irradiation exposure times of up to 25 years inboiling water reactors and 31 years in pressurized waterreactors.1.2 It is the responsibility of

11、 the user to show that theconditions of interest in their application of this guide areaddressed adequately by the technical information on whichthe guide is based. It should be noted that the conditionsquantified by the database are not distributed evenly over therange of materials and irradiation

12、conditions described in 1.1,and that some combination of variables, particularly at theextremes of the data range are under-represented. Particularattention is warranted when the guide is applied to conditionsnear the extremes of the data range used to develop the TTSequation and when the applicatio

13、n involves a region of the dataspace where data is sparse. Although the embrittlement corre-lation developed for this guide was based on statistical analysisof a large database, prudence is required for applications that1This guide is under the jurisdiction of ASTM Committee E10 on NuclearTechnology

14、 and Applications and is the direct responsibility of SubcommitteeE10.02 on Behavior and Use of Nuclear Structural Materials.Current edition approved Feb. 1, 2015. Published April 2015. Originallyapproved in 1983. Last previous edition approved in 2007 as E900 02(2007). DOI:10.1520/E0900-15.2The bol

15、dface numbers in parentheses refer to a list of references at the end ofthis standard.3To inform the TTS prediction of Section 5 of this guide, the E10.02Subcommittee decided to limit the data considered to Charpy shift values (T41J)measured from irradiations conducted in PWRs and BWRs. A database o

16、f 1,878Charpy TTS measurements was compiled from surveillance reports on operating anddecommissioned light water reactors of Western design from 13 countries (Brazil,Belgium, France, Germany, Italy, Japan, Mexico, The Netherlands, South Korea,Sweden, Switzlerland, Taiwan, and the United States), and

17、 from the technicalliterature. For each data record, the following information had to be available:fluence, fluence rate, irradiation temperature, and % content of Cu, Ni, P, and Mn.Reports and technical papers documenting the results of research programsconducted in material test reactors were also

18、 reviewed. Data from these sources wasincluded in the database for information, but was not used in the development of theTTS prediction of Section 5 of this guide.Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States1involve variable value

19、s beyond the ranges specified in 1.1. Dueto strong correlations with other exposure variables within thedatabase (that is, fluence), and due to the uneven distribution ofdata within the database (for example, the irradiation tempera-ture and flux range of PWR and BWR data show almost nooverlap) neit

20、her neutron fluence rate nor irradiation timesufficiently improved the accuracy of the predictions to merittheir use in the embrittlement correlation in this guide. Futureversions of this guide may incorporate the effect of neutronfluence rate or irradiation time, or both, on TTS, as such effectsare

21、 described in (2). The irradiated material database, thetechnical basis for developing the embrittlement correlation,and issues involved in its application, are discussed in aseparate report (1). That report describes the nine different TTSequations considered in the development of this guide, some

22、ofwhich were developed using more limited datasets (forexample, national program data (3,4). If the material variablesor exposure conditions of a particular application fall withinthe range of one of these alternate correlations, it may providemore suitable guidance.1.3 This guide is expected to be

23、used in coordination withseveral standards addressing irradiation surveillance of light-water reactor vessel materials. Method of determining theapplicable fluence for use in this guide are addressed in GuidesE482, E944, and Test Method E1005. The overall applicationof these separate guides and prac

24、tices is described in PracticeE853.1.4 The values stated in SI units are to be regarded asstandard. The values given in parentheses are mathematicalconversions to U.S. Customary units that are provided forinformation only and are not considered standard.1.5 This standard guide does not define how th

25、e TTS shouldbe used to determine the final adjusted reference temperature,which would typically include consideration of the transitiontemperature before irradiation, the predicted TTS, and theuncertainties in the shift estimation method.1.6 This standard does not purport to address all of thesafety

26、 concerns, if any, associated with its use. It is theresponsibility of the user of this standard to establish appro-priate safety and health practices and determine the applica-bility of regulatory limitations prior to use.2. Referenced Documents2.1 ASTM Standards:4A302 Specification for Pressure Ve

27、ssel Plates, Alloy Steel,Manganese-Molybdenum and Manganese-Molybdenum-NickelA508 Specification for Quenched and Tempered Vacuum-Treated Carbon and Alloy Steel Forgings for PressureVesselsA533 Specification for Pressure Vessel Plates, Alloy Steel,Quenched and Tempered, Manganese-Molybdenum andMangan

28、ese-Molybdenum-NickelE185 Practice for Design of Surveillance Programs forLight-Water Moderated Nuclear Power Reactor VesselsE482 Guide for Application of Neutron Transport Methodsfor Reactor Vessel Surveillance, E706 (IID)E693 Practice for Characterizing Neutron Exposures in Ironand Low Alloy Steel

29、s in Terms of Displacements PerAtom (DPA), E 706(ID)E853 Practice forAnalysis and Interpretation of Light-WaterReactor Surveillance ResultsE944 Guide for Application of Neutron Spectrum Adjust-ment Methods in Reactor Surveillance, E 706 (IIA)E1005 Test Method for Application and Analysis of Radio-me

30、tric Monitors for Reactor Vessel Surveillance, E 706(IIIA)E2215 Practice for Evaluation of Surveillance Capsulesfrom Light-Water Moderated Nuclear Power Reactor Ves-sels3. Terminology3.1 Definitions of Terms Specific to This Standard:3.1.1 best-estimate chemical compositionthe best-estimate chemical

31、 composition (copper Cu, nickel Ni,phosphorus P, and manganese Mn in %) may be establishedusing one of the following methods: (1) Use a simple mean fora small set of uniformly distributed data; that is, sum themeasurements and divide by the number of measurements; (2)Use a weighting process for a no

32、n-uniformly distributed dataset, especially when the number of measurements from onesource are much greater in terms of material volume analyzed.For a plate, a unique sample could be a set of test specimenstaken from one corner of the plate. For a weldment, a uniquesample would be a set of test spec

33、imens taken from a uniqueweld deposit made with a specific electrode heat. A simplemean is calculated for test specimens comprising each uniquesample, the sample means are then added, and the sum isdivided by the number of unique samples to get the sampleweighted mean; (3) Use an alternative weighti

34、ng scheme whenother factors have a significant influence and a physical modelcan be established. For the preceding, the best estimate for thesample should be used if evaluating surveillance data from thatsample.3.1.1.1 DiscussionFor cases where no chemical analysismeasurements are available for a he

35、at of material, the upperlimiting values given in the material specifications to which thevessel was built may be used.Alternately, generic mean valuesfor the class of material may be used.3.1.1.2 DiscussionIn all cases where engineering judg-ment is used to select a best estimate copper, nickel,pho

36、sphorus, or manganese content, the rationale shall bedocumented which formed the basis for the selection.3.1.2 fluence ()in this guide the term “fluence” refers tothe fast (E 1MeV) neutron fluence, that is, the number ofneutrons per square centimeter with energy greater than 1.0MeV at the location o

37、f interest.3.1.3 fluence rate()in this guide the term “fluence rate”refers to the fast (E 1MeV) neutron fluence rate, that is, thenumber of neutrons per square centimeter per unit time with4For referenced ASTM standards, visit the ASTM website, www.astm.org, orcontact ASTM Customer Service at servic

38、eastm.org. For Annual Book of ASTMStandards volume information, refer to the standards Document Summary page onthe ASTM website.E900 152energy greater than 1.0 MeV at the location of interest. This isalso referred to as fast neutron flux.3.1.4 SRMStandard Reference Material. Also known ascorrelation

39、 monitor material.3.1.5 Tirradiation temperature at full power, in C, givenby the estimated time-weighted average (based on the meantemperature over each fuel cycle) cold leg temperature forpressurized water reactors (PWRs) and recirculation tempera-ture for boiling water reactors (BWRs).3.1.6 TTSth

40、e mean value of transition temperature shiftpredicted by the embrittlement correlation.4. Significance and Use4.1 Operation of commercial power reactors must conformto pressure-temperature limits during heatup and cooldown toprevent over-pressurization at temperatures that might causenon-ductile beh

41、avior in the presence of a flaw. Radiationdamage to the reactor vessel is compensated for by adjustingthe pressure-temperature limits to higher temperatures as theneutron damage accumulates. The present practice is to basethat adjustment on the TTS produced by neutron irradiation asmeasured at the C

42、harpy V-notch 41-J (30-ftlbf) energy level.To establish pressure temperature operating limits during theoperating life of the plant, a prediction of TTS must be made.4.1.1 In the absence of surveillance data for a given reactormaterial (see Practice E185 and E2215), the use of calculativeprocedures

43、are necessary to make the prediction. Even whencredible surveillance data are available, it will usually benecessary to interpolate or extrapolate the data to obtain a TTSfor a specific time in the plant operating life.The embrittlementcorrelation presented herein has been developed for thosepurpose

44、s.4.2 Research has established that certain elements, notablycopper (Cu), nickel (Ni), phosphorus (P), and manganese (Mn),cause a variation in radiation sensitivity of reactor pressurevessel steels. The importance of other elements, such as silicon(Si), and carbon (C), remains a subject of additiona

45、l research.Copper, nickel, phosphorus, and manganese are the key chem-istry parameters used in developing the calculative proceduresdescribed here.4.3 Only power reactor (PWR and BWR) surveillance datawere used in the derivation of these procedures. The measureof fast neutron fluence used in the pro

46、cedure is n/m2(E1MeV). Differences in fluence rate and neutron energy spectraexperienced in power reactors and test reactors have not beenaccounted for in these procedures.5. Calculative Procedure for Transition TemperatureShift (TTS )5.1 The mean value of TTS, in C, is calculated as followsusing th

47、e embrittlement correlation developed:2TTS 5 TTS11TTS2(1)where:TTS15 A 59 1.8943 310212 0.5695S1.8 T132550D25.47S0.09 1P0.012D0.216S1.66 1Ni8.540.63D0.39SMn1.36D0.3(2)A 511.011 for forgings1.080 for plates and SRM plates0.919 for welds2(3)and:TTS2559 maxmin Cu, 0.28! 2 0.053,0# M (4)M 5 B max$min 11

48、3.87 ln ! 2 ln4.5 3 1020!, 612.6#,0%S1.8T132550D25.45S0.1 1P0.012D20.098S0.168 1Ni0.580.63D0.73(5)B 510.738 for forgings0.819 for plates and SRM plates0.968 for welds2(6)5.2 In the equations of 5.1, Cu, Ni, P, and Mn are allexpressed in weight percent, is in n/m2(E 1 MeV), and T isin C.6. Attenuatio

49、n Through the Vessel Wall6.1 An attenuated fluence should be used to calculate theTTS for all locations within the vessel wall, rather than thefluence (E 1 MeV) at the inside surface.6.2 To calculate the shift at some location within the vesselwall away from the inside surface, it is necessary to account forthe variation in neutron energy spectrum and fluence rateintensity throughout the vessel wall. Due to these variations,the use of the inside surface fluence (E 1 MeV) may give anon-conservative estimate of the neutron damage a