ASTM E509 E509M-2014 red 3047 Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels《轻水慢化核反应堆容器运转中退火的标准指南》.pdf

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1、Designation: E509 03 (Reapproved 2008)E509/E509M 14Standard Guide forIn-Service Annealing of Light-Water Moderated NuclearReactor Vessels1This standard is issued under the fixed designation E509;E509/E509M; the number immediately following the designation indicates theyear of original adoption or, i

2、n the case of revision, the year of last revision. A number in parentheses indicates the year of lastreapproval. A superscript epsilon () indicates an editorial change since the last revision or reapproval.1. Scope1.1 This guide covers the general procedures to be considered for conducting an in-ser

3、vice thermal anneal of a light-watermoderated nuclear reactor vessel and demonstrating the effectiveness of the procedure. The purpose of this in-service annealing(heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previouslydegrad

4、ed by neutron embrittlement. The improvement in mechanical properties generally is assessed using Charpy V-notch impacttest results, or alternatively, fracture toughness test results or inferred toughness property changes from tensile, hardness,indentation, or other miniature specimen testing (1).21

5、.2 This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing atvarious temperatures and different time periods. Certain inherent limiting factors must be considered in developing an annealingprocedure. These factors include system-design li

6、mitations; physical constraints resulting from attached piping, support structures,and the primary system shielding; the mechanical and thermal stresses in the components and the system as a whole; and, materialcondition changes that may limit the annealing temperature.1.3 This guide provides direct

7、ion for development of the vessel annealing procedure and a post-annealing vessel radiationsurveillance program. The development of a surveillance program to monitor the effects of subsequent irradiation of theannealed-vessel beltline materials should be based on the requirements and guidance descri

8、bed in Practices E185 and E2215. Theprimary factors to be considered in developing an effective annealing program include the determination of the feasibility ofannealing the specific reactor vessel; the availability of the required information on vessel mechanical and fracture properties priorto an

9、nealing; evaluation of the particular vessel materials, design, and operation to determine the annealing time and temperature;and, the procedure to be used for verification of the degree of recovery and the trend for reembrittlement. Guidelines are providedto determine the post-anneal reference nil-

10、ductility transition temperature (RTNDT), the Charpy V-notch upper shelf energy level,fracture toughness properties, and the predicted reembrittlement trend for these properties for reactor vessel beltline materials. Thisguide emphasizes the need to plan well ahead in anticipation of annealing if an

11、 optimum amount of post-anneal reembrittlementdata is to be available for use in assessing the ability of a nuclear reactor vessel to operate for the duration of its present license,or qualify for a license extension, or both.1.4 The values stated in inch-pound or either SI units or inch-pound units

12、 are to be regarded separately as the standard. Thevalues stated in each system may not be exact equivalents; therefore, each system shall be used independently of the other.Combining values from the two systems may result in non-conformance with the standard.1.5 This standard does not purport to ad

13、dress all of the safety concerns, if any, associated with its use. It is the responsibilityof the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatorylimitations prior to use.2. Referenced Documents2.1 ASTM Standards:3E185 Practice

14、for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor VesselsE636 Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E 706 (IH)1 This guide is under the jurisdiction of ASTM Committee E10 on Nuclear Technology and Applications and is

15、 the direct responsibility of Subcommittee E10.02 onBehavior and Use of Nuclear Structural Materials.Current edition approved July 1, 2008Jan. 1, 2014. Published September 2008February 2014. Originally approved in 1997. Last previous edition approved in 20032008as E50903. 03 (2008). DOI: 10.1520/E05

16、09-03R08.10.1520/E0509_E0509M-14.2 The boldface numbers in parentheses refer to the list of references at the end of this standard.3 For referencedASTM standards, visit theASTM website, www.astm.org, or contactASTM Customer Service at serviceastm.org. For Annual Book of ASTM Standardsvolume informat

17、ion, refer to the standards Document Summary page on the ASTM website.This document is not an ASTM standard and is intended only to provide the user of an ASTM standard an indication of what changes have been made to the previous version. Becauseit may not be technically possible to adequately depic

18、t all changes accurately, ASTM recommends that users consult prior editions as appropriate. In all cases only the current versionof the standard as published by ASTM is to be considered the official document.Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 1942

19、8-2959. United States1E900 Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, E706 (IIF)E1253 Guide for Reconstitution of Irradiated Charpy-Sized SpecimensE2215 Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Rea

20、ctor Vessels2.2 ASME Standards:Boiler and Pressure Vessel Code, Section III, Rules for Construction of Nuclear Power Plant Components4Code Case N-557, In-Place Dry Annealing of a PWR Nuclear Reactor Vessel (Section XI, Division 1)42.3 Nuclear Regulatory Commission Documents:NRC Regulatory Guide 1.99

21、, Revision 2, Effects of Residual Elements on Predicted Radiation Damage on Reactor VesselMaterials5NRC Regulatory Guide 1.162, Format and Content of Report for Thermal Annealing of Reactor Pressure Vessels53. Significance and Use3.1 Reactor vessels made of ferritic steels are designed with the expe

22、ctation of progressive changes in material propertiesresulting from in-service neutron exposure. In the operation of light-water-cooled nuclear power reactors, changes inpressure-temperature (P T) limits are made periodically during service life to account for the effects of neutron radiation on the

23、ductile-to-brittle transition temperature material properties. If the degree of neutron embrittlement becomes large, the restrictionson operation during normal heat-up and cool down may become severe. Additional consideration should be given to postulatedevents, such as pressurized thermal shock (PT

24、S).Areduction in the upper shelf toughness also occurs from neutron exposure, andthis decrease may reduce the margin of safety against ductile fracture. When it appears that these situations could develop, certainalternatives are available that reduce the problem or postpone the time at which plant

25、restrictions must be considered. One of thesealternatives is to thermally anneal the reactor vessel beltline region, that is, to heat the beltline region to a temperature sufficientlyabove the normal operating temperature to recover a significant portion of the original fracture toughness and other

26、materialproperties that were lostdegraded as a result of neutron embrittlement.3.2 Preparation and planning for an in-service anneal should begin early so that pertinent information can be obtained to guidethe annealing operation. Sufficient time should be allocated to evaluate the expected benefits

27、 in operating life to be gained byannealing; to evaluate the annealing method to be employed; to perform the necessary system studies and stress evaluations; toevaluate the expected annealing recovery and reembrittlement behavior; to develop and functionally test such equipment as maybe required to

28、do the in-service annealing; and, to train personnel to perform the anneal.3.3 Selection of the annealing temperature requires a balance of opposing conditions. Higher annealing temperatures, and longerannealing times, can produce greater recovery of fracture toughness and other material properties

29、and thereby increase thepost-anneal lifetime. The annealing temperature also can have an impact on the reembrittlement trend after the anneal. On the otherhand, higher temperatures can create other undesirable property effects such as permanent creep deformation or temperembrittlement. These higher

30、temperatures also can cause engineering difficulties, that is, core and coolant removal and storage,localized heating effects, etc., in preventing the annealing operation from distorting the vessel or damaging vessel supports,primary coolant piping, adjacent concrete, insulation, etc. See ASME Code

31、Case N-557 for further guidance on annealingconditions and thermal-stress evaluations (2).3.3.1 When a reactor vessel approaches a state of embrittlement such that annealing is considered, the major criterion is thenumber of years of additional service life that annealing of the vessel will provide.

32、 Two pieces of information are needed to answerthe question: the post-anneal adjusted RTNDT and upper shelf energy level, and their subsequent changes during future irradiation.Furthermore, if a vessel is annealed, the same information is needed as the basis for establishing pressure-temperature lim

33、its forthe period immediately following the anneal and demonstrating compliance with other design requirements and the PTS screeningcriteria.The effects on upper shelf toughness similarly must be addressed.This guide primarily addresses RTNDT changes. Handlingof the upper shelf is possible using a s

34、imilar approach as indicated in NRC Regulatory Guide 1.162.Appendix X1 provides abibliography of existing literature for estimating annealing recovery and reembrittlement trends for these quantities as related toU.S. and other country pressure-vessel steels, with primary emphasis on U.S. steels.3.3.

35、2 A key source of test material for determining the post-anneal RTNDT, upper shelf energy level, and the reembrittlementtrend is the original surveillance program, provided it represents the critical materials in the reactor vessel.6Appendix X2 describesan approach to estimate changes in RTNDT both

36、due to the anneal and after reirradiation. The first purpose of Appendix X2 is tosuggest ways to use available materials most efficiently to determine the post-anneal RTNDT and to predict the reembrittlementtrend, yet leave sufficient material for surveillance of the actual reembrittlement for the r

37、emaining service life. The second purposeis to describe alternative analysis approaches to be used to assess test results of archive (or representative) materials to obtain theessential post-anneal and reirradiation RTNDT, upper shelf energy level, or fracture toughness, or a combination thereof.4 A

38、vailable from the American Society of Mechanical Engineers, 345 E. 47th Street, New York, NY 10017.5 Available from Superintendent of Documents, U.S. Government Printing Office, Washington, DC 20402.6 Consideration can be given to the reevaluation of broken Charpy specimens from capsules withdrawn e

39、arlier which can be reconstituted using Guide E1253 or frommaterial obtained (sampled) from the actual pressure-vessel wall.E509/E509M 1423.3.3 An evaluation must be conducted of the engineering problems posed by annealing at the highest practical temperature.Factors required to be investigated to r

40、educe the risk of distortion and damage caused by mechanical and thermal stresses atelevated temperatures to relevant system components, structures, and control instrumentation are described in 5.1.3 and 5.1.4.3.4 Throughout the annealing operation, accurate measurement of the annealing temperature

41、at key defined locations must bemade and recorded for later engineering evaluation.3.5 After the annealing operation has been carried out, several steps should be taken. The predicted improvement in fracturetoughness properties must be verified, and it must be demonstrated that there is no damage to

42、 key components and structures.3.6 Further action may be required to demonstrate that reactor vessel integrity is maintained within ASME Code requirementssuch as indicated in the referenced ASME Code Case N-557 (2). Such action is beyond the scope of this guide.4. General Considerations4.1 Successfu

43、l use of in-service annealing requires a thorough knowledge of the irradiation behavior of the specificreactor-vessel materials, their annealing response and reirradiation embrittlement trend, the vessel design, fabrication history, andoperating history. Some of these items may not be available for

44、specific older vessels, and documented engineering judgment maybe required to conservatively estimate the missing information.4.1.1 To ascertain the design operating life-knowledge life, knowledge of the following items is needed: reactor vessel materialcomposition, mechanical properties, fabricatio

45、n techniques, nondestructive test results, anticipated stress levels in the vessel,neutron fluence, neutron energy spectrum, operating temperature, and power history.4.1.1.1 The initial RTNDT as specified in subarticle NB-2300 of the ASME Boiler and Pressure Vessel Code, Section III, shouldbe determ

46、ined or estimated for those materials of concern in the high fluence regions of the reactor pressure vessel. Alternativemethods for the determination of RTNDT also may be used. Consideration should be given to the technical justification for alternatemethodologies and the data, which form the basis

47、for the RTNDT determination. Initial RTNDT values should be available orestimated for all materials located in these areas.4.1.1.2 The initial Charpy upper shelf energy as defined by Practices E185 and E2215 should be determined for materials ofconcern in the beltline region of the reactor pressure

48、vessel. Initial upper shelf energy levels should be available or estimated forall materials located in this area.4.1.1.3 Unirradiated archive heats of reactor vessel beltline materials7 should be maintained for preparation of additionalsurveillance samples as required by Practices E185 and E2215. Pr

49、eviously tested specimens should be retained as an additionalsource of material.4.1.1.4 Arecord of the actual fabrication history, including heat treatment and welding procedure, of the materials in the beltlineregion of the vessel should be maintained.4.1.1.5 The chemical composition should be determined for base metal(s) and deposited weld metal(s) and should include allelements potentially relevant to irradiation, annealing, and reirradiation behavior, for example, copper, nickel, phosphorus,manganese and sulfur. The variability in chemical composition should be determined wh