1、Designation: E 509 03Standard 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, the year of last
2、 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 covers the general procedures to be consid-ered for conducting an in-service thermal anneal of a light-water m
3、oderated 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 embrittlement. Theimprove
4、ment 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 to accommodate t
5、he 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 constraints resulting f
6、rom 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 thevessel annealing pr
7、ocedure 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 2215. The primary
8、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 par-ticular vesse
9、l 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 temperature (RT-NDT), th
10、e 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-anneal reembrittle-men
11、t 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 standard does not
12、 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 Documents2.1 ASTM Standards
13、:E 185 Practice for Design of Surveillance Programs Testsfor Light-Water Moderated Nuclear Power Reactor Ves-sels3E 636 Practice for Conducting Supplemental SurveillanceTests for Nuclear Power Reactor Vessels E 706 (IH)3E 900 Guide for Predicting Radiation-Induced TransitionTemperature Shift in Reac
14、tor Vessel Materials E 706 (IIF)3E 1253 Guide for Reconstitution of Irradiated CharpySpecimens3E 2215 Practice for the Evaluation of Surveillance Capsulesfrom Light-Water Moderated Nuclear Reactor Vessels32.2 ASME Standards:Boiler and Pressure Vessel Code, Section III, Rules forConstruction of Nucle
15、ar Power Plant Components4Code Case N-557, In-Place Dry Annealing of a PWRNuclear Reactor Vessel (Section XI, Division 1)42.3 Nuclear Regulatory Commission Documents:1This guide is under the jurisdiction of ASTM Committee E10 on NuclearTechnology and Applications and is the direct responsibility of
16、SubcommitteeE10.02 on Behavior and Use of Metallic Materials in Nuclear Systems.Current edition approved March 10, 2003. Published May 2003. Originallypublished as E 50997. Last previous edition E 50997.2The boldface numbers in parentheses refer to the list of references at the end ofthis standard.3
17、Annual Book of ASTM Standards, Vol 12.02.4Available from the American Society of Mechanical Engineers, 345 E. 47thStreet, New York, NY 10017.1Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.NRC Regulatory Guide 1.99, Revision 2, Effec
18、ts of Re-sidual Elements on Predicted Radiation Damage on Re-actor Vessel Materials5NRC 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 progre
19、ssive changes in material propertiesresulting from in-service neutron exposure. 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 t
20、ransition temperature mate-rial properties. If the degree of neutron embrittlement 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 th
21、e upper shelf toughnessalso occurs from neutron exposure, and this decrease mayreduce 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 c
22、onsidered. One of thesealternatives is to thermally anneal the reactor vessel beltlineregion, 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
23、were lost as a result of neutronembrittlement.3.2 Preparation and planning for 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 a
24、nnealing; to evaluate the annealing method to beemployed; to perform the necessary 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 t
25、rain personnel to perform the anneal.3.3 Selection of the annealing temperature 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 li
26、fetime. The annealing temperaturealso can have an impact on the reembrittlement 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
27、 difficulties, that is, core and coolantremoval and storage, localized heating 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 annea
28、ling conditions and thermal-stressevaluations (2).3.3.1 When a reactor vessel approaches 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
29、answer the question: the post-anneal adjusted RTNDTandupper shelf energy level, 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 th
30、e anneal and demonstrating compliance with other designrequirements and the PTS 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 NRCRegulato
31、ry Guide 1.162. Appendix X1 provides a bibliographyof existing literature for estimating 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 determi
32、ning thepost-anneal RTNDT, upper shelf energy level, and the reem-brittlement trend 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. T
33、he firstpurpose of Appendix X2 is to suggest ways to use availablematerials most 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
34、 describealternative analysis approaches to be used to assess test resultsof archive (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 enginee
35、ringproblems posed by annealing at the highest practical tempera-ture. Factors required 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 describe
36、d in5.1.3 and 5.1.4.3.4 Throughout the annealing operation, accurate measure-ment 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 improv
37、ement infracture toughness properties must be verified, and it must bedemonstrated 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. Go
38、vernment PrintingOffice, Washington, DC 20402.6Consideration can be given to the 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.E509032requirements such as in
39、dicated in the referenced ASME CodeCase N-557 (2). Such action is beyond 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
40、-diation embrittlement trend, the vessel design, fabricationhistory, and 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 operati
41、ng life-knowledge ofthe following items is needed: reactor vessel material 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 Th
42、e initial RTNDTas specified in subarticle NB-2300of the ASME Boiler and 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. Con
43、sideration should be given to thetechnical justification for alternate 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 byPr
44、actices E 185 and E 2215 should be determined for materialsof concern in 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 beltlinematerial
45、s7should be maintained for preparation of additionalsurveillance samples 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 t
46、he materials in thebeltline region of the vessel should be maintained.4.1.1.5 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, nick
47、el, phospho-rus, and sulfur. The variability in chemical composition shouldbe 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 su
48、rveillance program conducted in accordancewith the requirements of Practices 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 differ
49、ent 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,should be used to estimate DRTNDTand upper shelf energydrop for the specific heats of materials in the vessel beltline.4.1.2.3 The results of surveillance specimen tests requiredby Practice E 2215 should be compared to the data develope
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