ASTM E509-2003(2008) 952 Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels《轻水冷却核反应堆容器在运转中逐渐冷却的标准指南》.pdf

《ASTM E509-2003(2008) 952 Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels《轻水冷却核反应堆容器在运转中逐渐冷却的标准指南》.pdf》由会员分享，可在线阅读，更多相关《ASTM E509-2003(2008) 952 Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels《轻水冷却核反应堆容器在运转中逐渐冷却的标准指南》.pdf（11页珍藏版）》请在麦多课文档分享上搜索。

1、Designation: E 509 03 (Reapproved 2008)Standard Guide forIn-Service Annealing of Light-Water Moderated NuclearReactor Vessels1This standard is issued under the fixed designation E 509; the number immediately following the designation indicates the year oforiginal adoption or, in the case of revision

2、, 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 guide covers the general procedures to be consid-ered for conducting an in-service thermal anneal o

3、f a light-water moderated nuclear reactor vessel and demonstrating theeffectiveness of the procedure. The purpose of this in-serviceannealing (heat treatment) is to improve the mechanicalproperties, especially fracture toughness, of the reactor vesselmaterials previously degraded by neutron embrittl

4、ement. Theimprovement in mechanical properties generally is assessedusing Charpy V-notch impact test results, or alternatively,fracture toughness test results or inferred toughness propertychanges from tensile, hardness, indentation, or other miniaturespecimen testing (1).21.2 This guide is designed

5、 to accommodate the variableresponse of reactor-vessel materials in post-irradiation anneal-ing at various temperatures and different time periods. Certaininherent limiting factors must be considered in developing anannealing procedure. These factors include system-designlimitations; physical constr

6、aints resulting from attached piping,support structures, and the primary system shielding; themechanical and thermal stresses in the components and thesystem as a whole; and, material condition changes that maylimit the annealing temperature.1.3 This guide provides direction for development of theve

7、ssel annealing procedure and a post-annealing vessel radia-tion surveillance program. The development of a surveillanceprogram to monitor the effects of subsequent irradiation of theannealed-vessel beltline materials should be based on therequirements and guidance described in Practices E 185 andE 2

8、215. The primary factors to be considered in developing aneffective annealing program include the determination of thefeasibility of annealing the specific reactor vessel; the avail-ability of the required information on vessel mechanical andfracture properties prior to annealing; evaluation of the

9、par-ticular vessel materials, design, and operation to determine theannealing time and temperature; and, the procedure to be usedfor verification of the degree of recovery and the trend forreembrittlement. Guidelines are provided to determine thepost-anneal reference nil-ductility transition tempera

10、ture (RT-NDT), the Charpy V-notch upper shelf energy level, fracturetoughness properties, and the predicted reembrittlement trendfor these properties for reactor vessel beltline materials. Thisguide emphasizes the need to plan well ahead in anticipation ofannealing if an optimum amount of post-annea

11、l reembrittle-ment data is to be available for use in assessing the ability ofa nuclear reactor vessel to operate for the duration of its presentlicense, or qualify for a license extension, or both.1.4 The values stated in inch-pound or SI units are to beregarded separately as the standard.1.5 This

12、standard does not purport to address all of thesafety 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

13、.1 ASTM Standards:3E 185 Practice for Design of Surveillance Programs forLight-Water Moderated Nuclear Power Reactor VesselsE 636 Guide for Conducting Supplemental SurveillanceTests for Nuclear Power Reactor Vessels, E 706 (IH)E 900 Guide for Predicting Radiation-Induced TransitionTemperature Shift

14、in Reactor Vessel Materials, E706 (IIF)E 1253 Guide for Reconstitution of Irradiated Charpy-SizedSpecimensE 2215 Practice for Evaluation of Surveillance Capsulesfrom Light-Water Moderated Nuclear Power Reactor Ves-sels2.2 ASME Standards:Boiler and Pressure Vessel Code, Section III, Rules forConstruc

15、tion of Nuclear Power Plant Components4Code Case N-557, In-Place Dry Annealing of a PWRNuclear Reactor Vessel (Section XI, Division 1)41This guide is under the jurisdiction of ASTM Committee E10 on NuclearTechnology and Applications and is the direct responsibility of SubcommitteeE10.02 on Behavior

16、and Use of Nuclear Structural Materials.Current edition approved July 1, 2008. Published September 2008. Originallyapproved in 1997. Last previous edition approved in 2003 as E 50903.2The boldface numbers in parentheses refer to the list of references at the end ofthis standard.3For referenced ASTM

17、standards, visit the ASTM website, www.astm.org, orcontact ASTM Customer Service at serviceastm.org. For Annual Book of ASTMStandards volume information, refer to the standards Document Summary page onthe ASTM website.4Available from the American Society of Mechanical Engineers, 345 E. 47thStreet, N

18、ew York, NY 10017.1Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.2.3 Nuclear Regulatory Commission Documents:NRC Regulatory Guide 1.99, Revision 2, Effects of Re-sidual Elements on Predicted Radiation Damage on Re-actor Vessel Mater

19、ials5NRC Regulatory Guide 1.162, Format and Content ofReport for Thermal Annealing of Reactor Pressure Ves-sels53. Significance and Use3.1 Reactor vessels made of ferritic steels are designed withthe expectation of progressive changes in material propertiesresulting from in-service neutron exposure.

20、 In the operation oflight-water-cooled nuclear power reactors, changes inpressure-temperature (PT) limits are made periodicallyduring service life to account for the effects of neutronradiation on the ductile-to-brittle transition temperature mate-rial properties. If the degree of neutron embrittlem

21、ent becomeslarge, the restrictions on operation during normal heat-up andcool down may become severe. Additional considerationshould be given to postulated events, such as pressurizedthermal shock (PTS). A reduction in the upper shelf toughnessalso occurs from neutron exposure, and this decrease may

22、reduce the margin of safety against ductile fracture. When itappears that these situations could develop, certain alternativesare available that reduce the problem or postpone the time atwhich plant restrictions must be considered. One of thesealternatives is to thermally anneal the reactor vessel b

23、eltlineregion, that is, to heat the beltline region to a temperaturesufficiently above the normal operating temperature to recovera significant portion of the original fracture toughness andother material properties that were lost as a result of neutronembrittlement.3.2 Preparation and planning for

24、an in-service anneal shouldbegin early so that pertinent information can be obtained toguide the annealing operation. Sufficient time should beallocated to evaluate the expected benefits in operating life tobe gained by annealing; to evaluate the annealing method to beemployed; to perform the necess

25、ary system studies and stressevaluations; to evaluate the expected annealing recovery andreembrittlement behavior; to develop and functionally test suchequipment as may be required to do the in-service annealing;and, to train personnel to perform the anneal.3.3 Selection of the annealing temperature

26、 requires a bal-ance of opposing conditions. Higher annealing temperatures,and longer annealing times, can produce greater recovery offracture toughness and other material properties and therebyincrease the post-anneal lifetime. The annealing temperaturealso can have an impact on the reembrittlement

27、 trend after theanneal. On the other hand, higher temperatures can create otherundesirable property effects such as permanent creep deforma-tion or temper embrittlement. These higher temperatures alsocan cause engineering difficulties, that is, core and coolantremoval and storage, localized heating

28、effects, etc., in prevent-ing the annealing operation from distorting the vessel ordamaging vessel supports, primary coolant piping, adjacentconcrete, insulation, etc. See ASME Code Case N-557 forfurther guidance on annealing conditions and thermal-stressevaluations (2).3.3.1 When a reactor vessel a

29、pproaches a state of embrittle-ment such that annealing is considered, the major criterion isthe number of years of additional service life that annealing ofthe vessel will provide. Two pieces of information are neededto answer the question: the post-anneal adjusted RTNDTandupper shelf energy level,

30、 and their subsequent changes duringfuture irradiation. Furthermore, if a vessel is annealed, thesame information is needed as the basis for establishingpressure-temperature limits for the period immediately follow-ing the anneal and demonstrating compliance with other designrequirements and the PTS

31、 screening criteria. The effects onupper shelf toughness similarly must be addressed. This guideprimarily addresses RTNDTchanges. Handling of the uppershelf is possible using a similar approach as indicated in NRCRegulatory Guide 1.162.Appendix X1 provides a bibliographyof existing literature for es

32、timating annealing recovery andreembrittlement trends for these quantities as related to U.S.and other country pressure-vessel steels, with primary empha-sis on U.S. steels.3.3.2 A key source of test material for determining thepost-anneal RTNDT, upper shelf energy level, and the reem-brittlement tr

33、end is the original surveillance program, providedit represents the critical materials in the reactor vessel.6Appendix X2 describes an approach to estimate changes inRTNDTboth due to the anneal and after reirradiation. The firstpurpose of Appendix X2 is to suggest ways to use availablematerials most

34、 efficiently to determine the post-anneal RTNDTand to predict the reembrittlement trend, yet leave sufficientmaterial for surveillance of the actual reembrittlement for theremaining service life. The second purpose is to describealternative analysis approaches to be used to assess test resultsof arc

35、hive (or representative) materials to obtain the essentialpost-anneal and reirradiation RTNDT, upper shelf energy level,or fracture toughness, or a combination thereof.3.3.3 An evaluation must be conducted of the engineeringproblems posed by annealing at the highest practical tempera-ture. Factors r

36、equired to be investigated to reduce the risk ofdistortion and damage caused by mechanical and thermalstresses at elevated temperatures to relevant system compo-nents, structures, and control instrumentation are described in5.1.3 and 5.1.4.3.4 Throughout the annealing operation, accurate measure-men

37、t of the annealing temperature at key defined locationsmust be made and recorded for later engineering evaluation.3.5 After the annealing operation has been carried out,several steps should be taken. The predicted improvement infracture toughness properties must be verified, and it must bedemonstrat

38、ed that there is no damage to key components andstructures.3.6 Further action may be required to demonstrate thatreactor vessel integrity is maintained within ASME Code5Available from Superintendent of Documents, U.S. Government PrintingOffice, Washington, DC 20402.6Consideration can be given to the

39、 reevaluation of broken Charpy specimensfrom capsules withdrawn earlier which can be reconstituted using Guide E 1253 orfrom material obtained (sampled) from the actual pressure-vessel wall.E 509 03 (2008)2requirements such as indicated in the referenced ASME CodeCase N-557 (2). Such action is beyon

40、d the scope of this guide.4. General Considerations4.1 Successful use of in-service annealing requires a thor-ough knowledge of the irradiation behavior of the specificreactor-vessel materials, their annealing response and reirra-diation embrittlement trend, the vessel design, fabricationhistory, an

41、d operating history. Some of these items may not beavailable for specific older vessels, and documented engineer-ing judgment may be required to conservatively estimate themissing information.4.1.1 To ascertain the design operating life-knowledge ofthe following items is needed: reactor vessel mater

42、ial compo-sition, mechanical properties, fabrication techniques, nonde-structive test results, anticipated stress levels in the vessel,neutron fluence, neutron energy spectrum, operating tempera-ture, and power history.4.1.1.1 The initial RTNDTas specified in subarticle NB-2300of the ASME Boiler and

43、 Pressure Vessel Code, Section III,should be determined or estimated for those materials ofconcern in the high fluence regions of the reactor pressurevessel. Alternative methods for the determination of RTNDTalso may be used. Consideration should be given to thetechnical justification for alternate

44、methodologies and the data,which form the basis for the RTNDTdetermination. InitialRTNDTvalues should be available or estimated for all materialslocated in these areas.4.1.1.2 The initial Charpy upper shelf energy as defined byPractices E 185 and E 2215 should be determined for materialsof concern i

45、n the beltline region of the reactor pressure vessel.Initial upper shelf energy levels should be available or esti-mated for all materials located in this area.4.1.1.3 Unirradiated archive heats of reactor vessel beltlinematerials7should be maintained for preparation of additionalsurveillance sample

46、s as required by Practices E 185 andE 2215. Previously tested specimens should be retained as anadditional source of material.4.1.1.4 A record of the actual fabrication history, includingheat treatment and welding procedure, of the materials in thebeltline region of the vessel should be maintained.4

47、.1.1.5 The chemical composition should be determined forbase metal(s) and deposited weld metal(s) and should includeall elements potentially relevant to irradiation, annealing, andreirradiation behavior, for example, copper, nickel, phospho-rus, and sulfur. The variability in chemical composition sh

48、ouldbe determined when possible.4.1.2 The anticipated remaining operating lifetime of thereactor vessel without annealing should be established usingneutron embrittlement projections for the reactor vessel mate-rials.4.1.2.1 A surveillance program conducted in accordancewith the requirements of Prac

49、tices E 185 and E 2215 willprovide information from which to evaluate vessel condition.Attention should be given to assuring that variations in thefluence-rate, neutron energy spectrum, and irradiation tempera-ture for all different reactor neutron environments utilized aretaken into account.4.1.2.2 Transition temperature and upper-shelf Charpy en-ergy level data have been compiled and used to developcorrelations of DRTNDTand upper shelf drop versus fluence, forexample, Guide E 900 or NRC Regulatory Guide 1.99, Revi-sion 2. These approaches, or other class-specific correlations,shou