NRC Generic Letter 1992-01

NRC Generic Letter 1992-001: Reactor Vessel Structural Integrity, 10 CFR 50.54(f)

ML031200626
Person / Time
Issue date:	02/28/1992
From:	Partlow J G Office of Nuclear Reactor Regulation
To:
References
GL-88-011 GL-92-001, NUDOCS 9202260115
Download: ML031200626 (12)

VUNITED STATES* -0uA ul Ad tNUCLEAR REGULATORY COMMISSIONWASHINGTON. 0. C. 205S5February i', 1992TO: ALL HOLDERS OF OPERATING LICENSES OR CONSTRUCTION PERMITS FOR NUCLEARPOWER PLANTS (EXCEPT YANKEE ATOMIC ELECTRIC COMPANY, LICENSEE FOR THEYANKEE NUCLEAR POWER STATION)SUBJECT: REACTOR VESSEL STRUCTUPAL INTEGRITY, 10 CFR 50.54(f)(GENERIC LETTER 92-01)The U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter toobtain information needed to assess compliance with requirements and commitmentsregarding reactor vessel integrity in view of certain concerns raised in thestaff's review of reactor vessel integrity for the Yankee Nuclear Power Station.In Section 50.60(a) of Title 10 of the Code of Federal Regulations (10 CFR50.60(a)). the NRC requires that licensees for all light water nuclear power,reactors meet fracture toughness requirements and have a material surveillanceprogram for the reactor coolant pressure boundary. These requirements are setforth in Appendices G and H to 10 CFR Part 50. In 10 CFR 50.60(b), where therequirements of Appendices G and H to 10 CFR Part 50 cannot be met, an exemptionis necessary pursuant to 10 CFR 50.12. In 10 CFR 50.61 the NRC also providedfracture toughness requirements for protecting pressurized water reactorsagainst pressurized thermal shock events. Licensees and permit holders havealso made commitments in response to Generic Letter (GL) 88-11, INRC Positionon Radiation Embrittlement of Reactor Vessel Materials and its Impact on PlantOperations," to use the methodology in Regulatory Guide 1.99, Revision 2,"Radiation Embrittlement of Reactor Vessel Materials,m to predict the effectsof neutron irradiation as required by Paragraph V.A of 10 CFR Part 50, AppendixG. The 10 CFR 50.60 and 10 CFR 50.61 requirements and GL 88-11 are in theoverall regulatory program to maintain the structural integrity of the reactorvessel. While reviewing the Integrity of the reactor vessel at the YankeeNuclear Power Station, the NRC staff raised concerns regarding the licensee'scompliance with certain requirements and commitments.This generic letter is part of a program to evaluate reactor vessel integrityand take regulatory actions, if needed, to ensure that licensees and permitholders are complying with 10 CFR 50.60 and 10 CFR 50.61, and are fulfillingcommitments made in response to CL 88-11. Enclosure 1 is a discussion of theapplicable regulatory requirements. The NRC is requiring information oncompliance under the provisions of 10 CFR 50.54(f).Assessment of Embrittlement for the Yankee Nuclear Power Station Reactor VesselIn an effort to resolve concerns regarding the neutron embrittlement of theYankee reactor vessel, the staff performed a safety assessment of the Yankee

-2 -reactor vessel. The staff found that tht licensee for the Yankee NuclearPower Station might not be in compliance with 10 CFR 50.60 and had not protier)completed tIe assessment required in 10 CFR 50.61. Further, the licensee forthe Yankee hucicar P.wee Station had incorrectly applied the methodology inRegulatory Guide 1.99, Revision 2.The staff found that the Charpy upper shelf energy of the Yankee reactor vesselmaterial could be as lcw as 35.5 foot-pounds which is less than the 50 foot-poundvalue required in Appendix G to 10 CFR Part 50. However, the licensee for theYankee Nuclear Power Statior hac not performed the actions required in ParagraphsIV.A.1 or V.C of Appendix G to IC CFR Part 50. Since then, the licensee hasperformed an analysis in accordance tith Faragraph IV.A.1 o' Appcncix £ to ItCCFR Part 50 using criteria being developed by the American Society of MechanicalEngineers (ASME) to demonstrate margins of safety equivalent to those in theASME Code.The NRC expressed a concern regarding compliance with the requirements ofAppendix H to IC CFR Part 50. Section E 185 of the American Society forTesting and Materials (ASTM) Coce requires that the licensee take samplespecimens from actual material used in fabricating the beltline of the reactorvessel. These surveillance materials shall include one heat of base metal.one butt weld, and one wela 'heat affected zone." The licensee for the YankeeNuclear Power Station terminated the material surveillance program in 1965.Therefore, the Yankee Nuclear Power Station had no material surveillanceprogram on July 26, 1983, when Appendix H to 10 CFR Part 50 became effect've.Further, the samples irradiated at Yankee Rowe before 1965 were comprised onlyof base meta'.The licensee for the Yankee Nuclear Power Station had used the methodology inRegulatory Guide 1.99, Revision 2, to predict the effects of neutronerLrittlement. However, the staff found that the methodology in RegulatoryGuide 1.99, Revision 2, was incorrectly applied by the licensee. The specificissues were (1) the irradiation temperature, (2) the chemistry composition ofreactor vessel material, and (3) the results of the material surveillanceprogram.The irradiation temperature at the Yankee Nuclear Power Station is between454 VF and 520 OF, which is below the nominal irradiation temperature of 550 OFused in developing Regulatory Guide 1.99, Revision 2. A lower irradiationtemperature increases the effect of neutron embrittlement. The regulatoryguide indicates that for irradiation temperatures less than 525 OF,embrittlement effects should be considered to be greater than predicted by themethods of the guide. Adjustments that were made by the licensee wereinsufficient to account for this effect.The limited results of the surveillance program from the Yankee Nuclear PowerStation indicated that the increase in the reference temperature exceeds themean-plus-two standard deviations as predicted by the procedures in RegulatoryGuide 1.99, Revision 2. The regulatory guide states that the licensee shoulduse credible surveillance data to predict the increase in reference temperatureresulting from neutron irradiation.I --- .-.- .------I -------------.-.r. ..-. -- -- -. -- ----- ---.. .- -- --- --- ----- -- .----- ---- ------I ..-

-3 -The staff implemented RG 1.99, Revision 2, by issuing GL 88-11. In committingto GL 88-11. licersees have committed to calculate radiation embrittlementin accordance with the procedures documented in RG 1.99, Revision 2. To gleetthe limitations in Section 1.3 of the regulatory guide, the licensee shouldconsider the effects on irradiatior, er.irittlement during ccre critical operationwith irradiation temperatures less than 525 IF. Section 2 of the regulatoryguide states that the licensees should consider the effects of the results fromits surveillance capsules.The Summer 1972 Addenda ef the ;,a: Edition of Section III of the ASME Boilerand Pressure Vessel Code are the earliest code requirements for testing materialsto determine their unirradiated reference temperature. Since the YankeeredLotr vessel was constructed to an ASME Code earlier than the Summer 1972, ithtd t.ct been sufficiently tested to determine its unirradiated referencetemperature. The licensee for the Yankee Nuclear Fower Station extrapolatedthe available test results to determine an unirradiated reference temperature.The staff determined that the licensee's extrapolatior. yas rot conservative.The chemical composition of the Yankee reactor vessel welde is unkr~ov:n. The,material's sensitivity to neutron embrittlement depends on its chemical content.The licensee assumed that the chemistry of its weld' was equivalent to that ofthe BP-3 reactor vessel *n Mol, Belgium. However,.,the licensee could notidentify the heat number of the wire used to fabr'cate the Yankee welds. Thelicensee was assuming a chemical composition tha.t was not based on itsplant-specific information, since the chemical composition, in particular, theamount of copper, depends upon the heat number of the weld wire.These factors prompted the staff to find that the licensee for the YankeeNuclear Power Station had not considered plant-specific information in assessingcompliance with 10 CFR 50.61. When plgnt-specific information is considered,the Yankee reactor vessel may have exceeded the screening criteria in 10 CFR50.61. Since then, the licensee his performed a probabilistic fracturemechanics analysis in accordance with 10 CFR 50.61(b)(4) and the staff iscontinuing its review.Upon conducting the Yankee Nuclear Power Station review, the staff becameconcerned that this ray not be an isolated case regarding compliance with10 CFR 50.60 and 10 CFR 50.61 and fulfillment of commitments made in responseto GL 88-11. Thus, the staff is issuing this generic letter to obtain informationto assess compliance with these regulations and fulfillment of commitments.The staff is continuing to pursue this concern with the Yankee Atomic ElectricCompany. Therefore, the Yankee Atomic Electric Company need not respond tothis generic letter.Required InformationPortions of the following information requested are not applicable to alladdressees. The responses provided should, in these cases, indicate that therequested information is not applicable and why it is not applicable.

-4 -1. Certain addresbets are requested to provide the following informationregarding Appendix H to 10 CFR Part 50:

Addressees

who do not have a surveillance program meeting ASTh E;85-73, -79, or -82 and who de not have an irtegrated surveillanceprogram approved by the NRC (see Enclosure 2). are requested todescribe actions taken or to be taken to ensure compliance withAppendix H to 10 CFP Part 50.

Addressees

who plan to revise thesurveillance program; tu rreet Appendix II to IC CFR Part 50 arerequested to indicate when the rfvised program will be submitted tothe NRC staff for review. If the surveillance program is not to berevised to meet Appendix H to 10 CFR Part 50, addressees arerequested to indicate when they plan to request an exemption fromAppendix H to 10 CFR Part 50 under 10 CFR 50.60(b).2. Certain addressees are requested to provide the followinginformation regarding Appendix G to 10 CFR Part 50:a.

Addressees

of plants for which the Charpy upper shelf energy ispredicted to be less than 50 foot-pounds at the end of theirlicenses using the guidance in Paragraphs C.1.2 or C.2.2 inRegulatory Guide 1.99, Revision 2, are requested to provide to theNRC the Charpy upper shelf energy predicted for December 16, 1991,and for the end of their current license for the limiting beltlineweld and the plate or forging and are requested to describe theactions taken pursuant to Paragraphs IY.A.I or V.C of Appendix G to10 CFR Part 50.b.

Addressees

whose reactor vessels were constructed to an ASME Codeearlier than the Summer 1972 Addenda of the 1971 Edition arerequested to describe the consideration given to the followingmaterial properties it titir ealuations performed pursuant to10 CFR 50.61 and Paragraph III.A of 10 CFR Part 50, Appendix G:(1) the results from all Charpy and drop weight tests for allunirradiated beltline materials, the unirradiated referencetemperature for each beltline material, and the method ofdetermining the unirradiated reference temperature from theCharpy and drop weight test;(2) the heat treatment received by all beltline and surveillancematerials;(3) the heat number for each beltline plate or forging and the heatnumber of wire and flux lot number used to fabricate eachbeltline weld;

-5 -(4) the heat number for each surveillance plate or forging and theheat number of wire and flux lot numter used to fabricate thesurveillance weld;(5) the chemical composition, in particular the weight in percent ofcopper, nickel, phosphorous, and sulfur for each beltline andsurveillance material; and(6) the heat number of the wire used for determining the weldmetal chemical composition if different than Item (3) above.3.

Addressees

are requested to provide the following informationregarding tuni.1trxr.ts nPcda to respond to GL 88-11:a. How the embrittlement effects of operating at an irradiationtemperature (cold leg or recirculation suction temperature) below 525'F were considered. In particular licensees are requested to describeconsideration given to determining the effect of lower irradiationtemperature on the reference temperature and on the Charpy uppershelf energy.b. How their surveillance results on the predicted amount ofembrittlement were considered.c. If a measured increase in reference temperature exceeds themean-plus-two standard deviations predicted by Regulatory Guide1.99, Revision 2, or if a measured decrease in Charpy upper shelfenergy exceeds the value predicted using the guidance in ParagraphC.1.2 in Regulatory Guide 1.99, Revision 2, the licensee is requestedto report the information and describe the effect of the surveillanceresults on the adjusted reference temperature and Charpy upper shelfenergy for each beltline material as predicted for December 16,1991, and for the end of its current license.Reporting RequirementsPursuant to Section 182a of the Atomic Energy Act of 1954, as amended, and10 CFR 50.54(f), each addressee shall submit a letter within 120 days of thedate of this generic letter providing the information described under *RequiredInformation.' The letter shall be addressed to the U.S. Nuclear RegulatoryCommission, ATTN: Document Control Desk, Washington, DC 20555, under oath oraffirmation. A copy shall also be submitted to the appropriate RegionalAdministrator. This generic letter requests information that will enable theNRC to verify that the licensee is complying with its current licensing basisregarding reactor vessel fracture toughness and Material surveillance forthe reactor coolant pressure boundary. Accordingly, an evaluation Justifyingthis information request is not necessary under 10 CFR 50.54(f).

-6 -Eackfit DiscussionThis generic letter requests information that will enable the NRC staff todetermine whether licensees are complying with their prior commitments andany license conditions regarding 10 CFR 50.60, 10 CFR 50.61, and GL 88-Il.The staff is not establishing a new position for such compliance in thisgeneric letter. The staff is requesting information to verify that thelicensee is complying with its previously established commitments and is notestablishing any new position. Therefore, this generic letter does notconstitute a backf't and no documented evaluation or backfit analysis need beprepared.Request fcr Voluntary Submittal of Impact DataThis request is covered by Office of Manaoement and Budoet Clearance Number3150-0011, which expires May 31, 1994. The estimated average number of burdenhours is 2?C person hours for each addressee's response, including the timerequired to assess the requirements, search data sources, gather and analyzethe data, and prepare the required letters. This estimated average number ofburden hours pertains only to the identified response-related matters and doesnot include the time to implement the actions required by the regulations.Comments on the accuracy of this estimate and suggestions to reduce the burdenmay be directed to Ronald Minsk, Office of Information and Regulatory Affairs(3150-0011), NEOB-3019, Office of Management and Budget, Washington, DC20503, and to the U.S. Nuclear Regulatory Commission, Information and RecordsManagement Branch, Division of Information Support Services, Office ofInformation and Resources Management, Washington, DC 20555.Although no specific request or requirement is intended, the followinginformation would assist the NRC in evaluating the cost of complyingwith this generic letter:(1) the licensee staff's time and costs to perform requested inspections,corrective actions, and associated testing;(2) the licensee staff's time and costs to prepare the requested reports anddocumentation;(3) the additional short-term costs incurred to address the inspection findingssuch as the costs of the corrective actions or the costs of down time; and(4) an estimate of the additional long-tern costs that will be incurredas a result of implementing commitments such as the estimated costs ofconducting future inspections or increased maintenance.

-7 -If you have any questions about this matter, please contact one of the NRCtechnical contacts or the lead project manager listed below.Sincerely,J s G. PartlowAssociate Director for ProjectsOffice of Nuclear Reactor RegulationEnclosures:1. Applicable Regulatory Requirements2. Plants with Integrated Programs3. List of Recently IssuedGeneric LettersTechnical Contacts:Barry J. Elliot, NRR(301) 504-2709Keith R. Wichman, NRR(301) 504-2757Lead Project Manager:Daniel G. McDonald, NRR(301) 504-1408 Enclosure IRegulatory Requirements_A pi!tjStoReactor Vessel Structural Integrity10 CFR 50.60Pursuant to 10 CFR 50.60. all light water nuclear power reactors must meet thefracture toughness and material surveillance program requirements for thereactor coolant pressure boundary set forth in Appendices G and H to In CFRPart 5C.The fracture toughness of the reactor coolant pressure boundary required by 10CFR 50.60 is necessary to provide adequate margins of safety during anycondition of normal plant operation, including anticipated operationaloccurrences and system hydrostatic tests. The material surveillance programrequired by 10 CFR 50.60 monitors changes in the fracture toughness propertiesof ferritic materials in the reactor vessel beltline region of light waternuclear power reactors resulting from exposure of these materials to neutronirradiation and the thermal environment. Under the program, fracturetoughness test data are obtained from material specimens exposed insurveillance capsules, which are withdrewn periodically from the reactorvessel.Appendix G to 10 CFR Part 50 requires that the reactor vessel beltlinematerials must have Charpy upper shelf energy of no less than 50 ft-lbthroughout the life of the vessel. Otherwise, licensees are required toprovide demonstration of equivalent margins of safety in accordance withParagraph IY.A.1 of Appendix G to 10 CFR Part 50 or perform actions inaccordance with Paragraph V.C of Appendix G to 10 CFR Part 50.Appendix H to 10 CFR Part 50 requires the surveillance program to meet theAmerican Society for Testing and Materials (ASTM) Standard E 185, 'StandardPractice for Conducting Surveillance Tests for Light-Water Cooled NuclearPower Reactor Vessels.' Further, Appendix H to 10 CFR Part 50 specifies theapplicable edition of ASTH E 185. Appendix H to 10 CFR Part 50, as amended onJuly 26, 1983, requires that the part of the surveillance program conductedbefore the first capsule is withdrawn must meet the requirements of the 1973,the 1979, or the 1982 edition of ASTM E 185 that is current on the issue dateof the American Society of Mechanical Engineers (ASME) Boiler and PressureVessel Code under which the reactor vessel was purchased. The licensee mayalso use later editions of ASTM E 185 which have been endorsed by the FRC.The test procedures and reporting requirements for each capsule withdrawalafter July 26, 1983 must meet the requirements of the 1982 edition of ASTME 185 to the extent practical for the configuration of the specimens in thecapsule. The licensee may use either the 1973, the 1979, or the 1982 editionof ASTM E 185 for each capsule withdrawal before July 26, 1983.

Enclosure 1-2 -Licensees, especially o h reactor vessels purchased before ASTMissued the 1973 edition of ASTM E 185, may have surveillance programs that donot meet the requirements of Appendix H to 10 CFR Part 50 but may havealternative surveillance proarams. The licensee may use these alternativesurveillance programs in accordance with 10 CFR 50.60(b) if the licensee hasbeen granted an exemption by the Commission under I) CFR 50.12.The licensee must monitor the test results from the material surveillanceprogram. According to Paragraph 1II.C of Appendix H to 10 CFR Part 50, theresults of the surveillance program may indicate that a technicalspecifications change is required, either in the pressure-temperature limitsor in the operating procedures required to meet the limits.10 CFR 50.61Pursuant to 10 CFR 50.61, there are fracture toughness requirements forprotection against pressurized thermal shock events for pressurized waterreactors. Licensees are required to perform an assessment of the projectedvalues of reference temperature. If the projected reference temperatureexceeds the screening criteria established in 10 CFR 50.61, licensees arerequired to submit an analysis and schedule for such flux reduction programsas are reasonably practicable to avoid exceeding the screening criteria. Ifno reasonably practicable flux reduction program will avoid exceeding thescreening criteria, licensees shall submit a safety analysis to determinewhat actions are necessary to prevent potential failure of the reactor vesselif continued operation beyond the screening criteria is allowed. In 10 CFR50.61(b)(1), as amended effective June 14, 1991 (56 Fed Reg 22300 et. seq.,May 15, 1991), licensees are required to submit their assessment byDecember 16, 1991, if the projected reference temperature will exceed thescreening criteria before the expiration of the operating license.Plant-specific information is required to be considered in assessing the levelof neutron embrittlement as specified in 10 CFR 50.61(b)(3). This informationincludes but is not limited to the reactor vessel operating temperature andsurveillance results.Prediction of Irradiation EmbrittlementParagraph V.A of Appendix G to 10 CFR Part 50 requires tne prediction of theeffects of neutron irradiation on reactor vessel materials. The extent ofneutron embrittlement depends on the material properties, thermal environment,and results of the material surveillance program. In Generic Letter 88-11,NRC Position on Radiation Enbrittlement of Reactor Vessel Materials and itsImpact on Plant Operations , the staff stated that it will use the guidance inRegulatory Guide 1.99, Revslion 2, 'Radiation Embrittlement of Reactor VesselMaterials,' in estimating the embrittlement of the materials in the reactorvessel beltline. All licensees and permittees have responded to GenericLetter 88-11 committing to use the methodology in Regulatory Guide 1.99, Enclosure 1Revision 2. in predicting the effects of neutron irradiation as required byParaqraph V.A of 10 CFR Part 50, Appendix G. The methodology in RegulatoryGuide 1.99, Revision 2, is also the basis in 10 CFR 50.61 in projecting thereference temperature.

Enclosure 2Plants With Intearated Surveillance Proorams Approved By The NRCOconee Units 1, 2, and 3Arkansas Nuclear One Unit IRancho SecoThree Mile Island Unit IDavis-BesseGinnaPoint Feach Units I and 2Surry Units I and 2Turkey Point Units 3 and 4Zion Units I and 2--

All NRR Project Managers-2 -February 25. 1992The Committee to Review Generic Requirements recommended in favor of issuingthis generic letter at its meeting number 211 held on November 26. 1991.CR .: -_J ;;t,James G. PartlowAssociate Director for ProjectsOffice of Nuclear Reactor RegulationEnclosure:Generic Letter 92-01cc w/enclosure:Z. TaylorH. ThompsonJ. SniezekAssociate Directors, NRRDivision Directors, NRRAssistant Directors, NRRProJect Directors, NRRRegional AdministratorsC. BerlingerS. Treby, OGCJ. Conran, CRGRn Un^-SA NOD ROTSTRIBUTION:Uenrialr miNRC PDRPDI-1 ReadingDMcDonaldRACapraRIngram, 12/1H/2EMullinix, 12/H/5BElliot, 7/0/4KWichman, 7/D/4UI. IUI, g -IOFC :LA:PDI-l :2PE-l T1:FlT:T:TA:DRPE .T:AP----: ~~ ----------- :, r-]------:- --------- _- --- -- _-__--_-_-_ -NAME :C~oqan ,inald/vsb :RCapra ." :MBoyle :ELeedDATE : /.'/92 :1 / 0/92 :1 /15/92 :1/1i/92 : 1/6/92CF -~p a ~ --~ .~--- -.-- -~.--- ~....*F : e.OF--------:------ --------:------ --------:------ ---------NAME :JPartlow \ :DATE :Z / /92UFFICIAL KRLURD COPYDocument Name: ELLIOT