ASTM E900-2002(2007) Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials E706 (IIF)《在反应堆容器材料中预测辐射感应变化导致温度的变化的标准指南 E 706(IIF)》.pdf

1、Designation: E 900 02 (Reapproved 2007)Standard Guide forPredicting Radiation-Induced Transition Temperature Shiftin Reactor Vessel Materials, E706 (IIF)1This standard is issued under the fixed designation E 900; the number immediately following the designation indicates the year oforiginal adoption

2、 or, in the case of revision, 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 guide presents a method for predicting referencetransition temperature

3、 adjustments for irradiated light-watercooled power reactor pressure vessel materials based onCharpy V-notch 30-ftlbf (41-J) data. Radiation damage calcu-lative procedures have been developed from a statisticalanalysis of an irradiated material database that was available asof May 2000.2The embrittl

4、ement correlation used in this guidewas developed using the following variables: copper and nickelcontents, irradiation temperature, and neutron fluence. Theform of the model was based on current understanding for twomechanisms of embrittlement: stable matrix damage (SMD)and copper-rich precipitatio

5、n (CRP); saturation of coppereffects (for different weld materials) was included. This guideis applicable for the following specific materials, copper,nickel, and phosphorus contents, range of irradiation tempera-ture, and neutron fluence based on the overall database:1.1.1 Materials:1.1.1.1 A 533 T

6、ype B Class 1 and 2, A302 Grade B, A302Grade B (modified), A508 Class 2 and 3.1.1.1.2 Submerged arc welds, shielded arc welds, and elec-troslag welds for materials in 1.1.1.1.1.1.2 Copper contents within the range from 0 to 0.50 wt %.1.1.3 Nickel content within the range from 0 to 1.3 wt %.1.1.4 Pho

7、sphorus content within the range 0 to 0.025 wt %.1.1.5 Irradiation exposure temperature within the rangefrom 500 to 570F (260 to 299C).1.1.6 Neutron fluence within the range from 1 3 1016to 83 1019n/cm2(E 1 MeV).1.1.7 Neutron energy spectra within the range expected atthe reactor vessel core beltlin

8、e region of light water cooledreactors and fluence rate within the range from 2 3 108to 1 31012n/cm2s (E 1 MeV).1.2 The basis for the method of adjusting the referencetemperature is discussed in a separate report.31.3 This guide is Part IIF of Master Matrix E 706 whichcoordinates several standards u

9、sed for irradiation surveillanceof light-water reactor vessel materials. Methods of determiningthe applicable fluence for use in this guide are addressed inMaster Matrix E 706, Practices E 560 (IC) and Guide E 944(IIA), and Test Method E 1005 (IIIA). The overall applicationof these separate guides a

10、nd practices is described in PracticeE 853 (IA).1.4 The values given in customary U.S. units are to beregarded as the standard. The SI values given in parenthesesare for information only.1.5 This standard guide does not define how the shift intransition temperature should be used to determine the fi

11、naladjusted reference temperature. (That would typically includeconsideration of the initial starting point, the predicted shift,and the uncertainty in the shift estimation method.)1.6 This standard does not purport to address all of thesafety concerns, if any, associated with its use. It is theresp

12、onsibility 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:4E 185 Practice for Design of Surveillance Programs forLight-Water Moderated Nuclear Power R

13、eactor VesselsE 560 Practice for Extrapolating Reactor Vessel Surveil-lance Dosimetry Results, E 706(IC)E 693 Practice for Characterizing Neutron Exposures inIron and Low Alloy Steels in Terms of Displacements PerAtom (DPA), E 706(ID)E 706 Master Matrix for Light-Water Reactor Pressure1This guide is

14、 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 Structural Materials.Current edition approved July 15, 2007. Published August 2007. Originallyapproved in 1983. Last previous edit

15、ion approved in 2002 as E 900 02.2The Charpy surveillance data were originally obtained from the Oak RidgeNational Laboratory Power Reactor-Embrittlement Database (PR-EDB) and subse-quently updated by ASTM Subcommittee E10.02, May 2000.3Charpy Embrittlement CorrelationsStatus of Combined Mechanistic

16、 andStatistical Bases for U.S. Pressure Vessel Steels (MRP-45), PWR MaterialsReliability Program (PWRMRP), EPRI, Palo Alto, CA, 2001, 1000705.4For referenced ASTM standards, visit the ASTM website, www.astm.org, orcontact ASTM Customer Service at serviceastm.org. For Annual Book of ASTMStandards vol

17、ume information, refer to the standards Document Summary page onthe ASTM website.1Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.Vessel Surveillance Standards, E 706(0)E 853 Practice for Analysis and Interpretation of Light-Water Rea

18、ctor Surveillance Results, E706(IA)E 944 Guide for Application of Neutron Spectrum Adjust-ment Methods in Reactor Surveillance, E 706 (IIA)E 1005 Test Method for Application and Analysis of Radio-metric Monitors for Reactor Vessel Surveillance, E706(IIIA)3. Terminology3.1 Definitions of Terms Specif

19、ic to This Standard:3.1.1 A, Bmaterial fitting coefficients that are a functionof material type.3.1.2 best-estimate chemical compositionthe best-estimate chemical composition (copper Cu and nickel Ni, inwt %) may be established using one of the following methods:(1) Use a simple mean for a small set

20、 of uniformly distributeddata; that is, sum the measurements and divide by the numberof measurements; (2) Use a weighting process for a non-uniformly distributed data set, especially when the number ofmeasurements from one source are much greater in terms ofmaterial volume analyzed. For a plate, a u

21、nique sample couldbe a set of test specimens taken from one corner of the plate.For a weldment, a unique sample would be a set of testspecimens taken from a unique weld deposit made with aspecific electrode heat. A simple mean is calculated for testspecimens comprising each unique sample, the sample

22、 meansare then summed, and the sum is divided by the number ofunique samples to get the sample weighted mean; (3) Use analternative weighting scheme when other factors have a sig-nificant influence and a physical model can be established. Forthe preceding, the best estimate for the sample should be

23、usedif evaluating surveillance data from that sample.3.1.2.1 DiscussionFor cases where no chemical analysismeasurements are available for a heat of material, the upperlimiting values given in the material specifications to which thevessel was built may be used. Alternately, generic mean valuesfor th

24、e class of material may be used.3.1.2.2 DiscussionIn all cases where engineering judg-ment was used to select a best estimate copper or nickelcontent, the rationale shall be documented which formed thebasis for the selection.3.1.3 CRPthe copper rich precipitate term of the transi-tion temperature sh

25、ift equation and is based on the knowledgeof copper-enriched clustering that occurs in RPV steels.3.1.4 F(Cu)a copper term in the transition temperatureshift equation that is a function of the measured copper contentand material, and is subject to a saturation level at a highcopper content.3.1.5 flu

26、ence (F)the number of neutrons per square cen-timeter with energy greater than 1.0 MeV at the location ofinterest.3.1.6 G (F)a fluence function term in the transitiontemperature shift equation.3.1.7 SMDthe stable matrix damage term of the transitiontemperature shift equation and is based on an assum

27、ed under-standing of matrix damage mechanisms in RPV steels.3.1.8 Tcirradiation temperature at full power, in F, and isthe estimated time-weighted average (based on the meantemperature over each fuel cycle) cold leg temperature forPWRs and recirculation temperature for BWRs.3.1.9 TTSthe predicted me

28、an value of the transition tem-perature shift from the 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 causenonductile behavior in the presence

29、 of a flaw. Radiationdamage to the reactor vessel beltline region is compensated forby adjusting the pressure-temperature limits to higher tempera-ture as the neutron damage accumulates. The present practiceis to base that adjustment on the increase in transition tempera-ture produced by neutron irr

30、adiation as measured at the CharpyV-notch 30-ftlbf (41-J) energy level. To establish pressuretemperature operating limits during the operating life of theplant, a prediction of adjustment in transition temperature mustbe made.4.1.1 In the absence of surveillance data for a given reactor(see Practice

31、 E 185), the use of calculative procedures will benecessary to make the prediction. Even when credible surveil-lance data are available, it will usually be necessary toextrapolate the data to obtain an adjustment in transitiontemperature for a specific time in the plant operating life. Theembrittlem

32、ent correlation presented herein has been developedfor those purposes.4.2 Research has established that certain elements, notablycopper and nickel, cause a variation in radiation sensitivity ofsteels. The importance of other elements, such as phosphorus(P), remains a subject of additional research.

33、Copper and nickelare the key chemistry parameters used in developing thecalculative procedures described here.4.3 Only power reactor surveillance data were used in thederivation of these procedures. The measure of fast neutronfluence used in the procedure is n/cm2(E1MeV). Differ-ences in the neutron

34、 fluence rate and neutron energy spectraexperienced in power reactors and test reactors have not beenapplied in these procedures. The manner in which these factorswere considered is addressed elsewhere.35. Calculative Procedure for Transition TemperatureShift5.1 This guide presents a calculative pro

35、cedure for estimat-ing the transition temperature shift caused by neutron radiation.The form of the correlation involves two major embrittlementterms. The form of the terms is mechanistically guided, and theterms represent the hardening contribution from small micro-structural defects and clusters c

36、reated during irradiation.5.1.1 Mean Transition Temperature Shift:5.1.1.1 The mean value of TTS, in F, is calculated asfollows:TTS 5 SMD 1 CRP (1)whereSMD 5 A exp20730/Tc1 460!#F!0.5076(2)CRP 5 B1 1 2.106Ni1.173#FCu!GF! (3)andE 900 02 (2007)2A 5 6.70 3 10218B 51234, welds;128, forgings;208, Combusti

37、on Engineering plates;156, other plates2GF! 512112tanhFlogF! 2 18.241.052GFCu! 5 S0, Cu #0.072 wt %Cu 2 0.072!0.577, Cu . 0.072 wt %Dsubject toCumax5 S0.25 wt %, for welds with Linde 80 or Linde 0091 flux;0.305 wt %, for other weldsD6. Attenuation Through the Vessel Wall6.1 The effective fluence sho

38、uld be used to calculate thetransition temperature shift for all locations within the vesselwall, rather than the actual fluence (E 1 MeV) for thoselocations.6.2 To calculate the shift at some location within the vesselwall away from the inside surface, it is necessary to account forthe change in ne

39、utron spectrum. Due to these changes inneutron energy spectrum, the use of neutron fluence (E 1MeV) may give a non-conservative estimate of the neutrondamage attenuation within the vessel wall. The preferredexposure parameter for accommodating this change is dis-placements per atom (dpa). Since dpa

40、can be calculatedthrough the vessel wall thickness following Practice E 693during normal surveillance program evaluations, each plantcould have the calculated dpa as a function of depth into thevessel wall. The calculated dpa can be used to obtain theeffective vessel wall fluence for use with the em

41、brittlementcorrelation in this guide as shown below:F!x5 F!ISdpax/dpaIS# (4)where (F)xand dpaxare the effective vessel wall fluence(units of n/cm2for E 1.0 MeV) and calculated dpa,respectively, at any distance x (in inch units) into the vesselwall from the inside surface (IS); and (F)ISand dpaISare

42、thecalculated fluence and dpa, respectively, at the inside surface ofthe pressure vessel wall.6.3 Alternately, the following exponential attenuation for-mula may be used:5F!x5 F!ISexp 20.24 x! (5)where (F)xis the effective vessel wall fluence (units ofn/cm2for E 1.0 MeV) at any distance x (in inch u

43、nits) intothe vessel wall from the inside surface (IS); and (F)ISis thecalculated fluence at the inside surface of the pressure vesselwall. A recent review has confirmed that this formula providesa reasonable estimation of through wall attenuation.66.4 Other forms of attenuation of embrittlement thr

44、ough thewall of the reactor vessel can be used if they can be technicallyjustified.7. Evaluation of Uncertainty7.1 The following guidance on uncertainty is provided foruse in applying the predicted transition temperature shift todetermine an adjusted reference temperature.7.2 The procedure outlined

45、in 5.1.1.1 provides an estimateof the irradiated Charpy transition temperature shift based onanalysis of reactor pressure vessel surveillance data. When thisprocedure is applied to the original database, the standard errorof the correlation is 22.0F (12.2C). Principal contributors tothis error inclu

46、de the uncertainty in the input parameters (Cu,Ni, fluence and irradiation temperature uncertainties). Theremainder of the error is attributed to uncertainties in theCharpy shift determinations and model errors.7.3 The standard error describes how well the model de-scribes the Charpy transition temp

47、erature shift. A significantportion of this error may be attributed to the actual shiftmeasurement. The statistical regression averages the materialresponse over the entire database. In this case, the model willdescribe mean material behavior more accurately than itdescribes an individual measuremen

48、t of shift. The 22.0F(12.2C) standard error of the original analysis provides anupper limit on the error for the mean behavior of an individualsurveillance material.7.4 In the application of the model, the uncertainty in themodel will depend on the uncertainty in the input parameters.If the uncertai

49、nties in the input parameters for the specificapplication are similar to those in the surveillance capsuledatabase, it is conservative to assume that the standard error ofthe original analysis applies to the application. Adjustmentsmay be required if the uncertainties in the input data signifi-cantly exceed the uncertainties in the surveillance capsuledatabase.7.5 Although phosphorus has been identified as a potentialembrittling agent in reactor pressure vessel steels, it has notbeen included in this procedure. Independent analysis hasdemonst