ML20149L826

Reactor Pressure Vessel Status Report

ML20149L826
Person / Time
Issue date:	10/31/1996
From:	Elliot B, Hackett E, Albert Lee, James Medoff, Strosnider J, Wichman K Office of Nuclear Reactor Regulation
To:
References
FACA, NUREG-1511, NUREG-1511-S01, NUREG-1511-S1, NUDOCS 9611180284
Download: ML20149L826 (42)

Text

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NUREG-1511 Supplement 1 Reactor Pressure Vessel Status Report l

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U.S. Nuclear Regulatory Commission i

Office of Nuclear Reactor Regulation B. J. Elliot, E. M. Hackett, A. D. Lee J. Medoff, J. R. Strosnider, K. R. Wichman pf" "%,

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AVAILABILITY NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:

1. The NRC Public Document Room, 2120 L Street, NW., Lower Level, Washington, DC 20555-0001

2. The Superintendent of Documents, U.S. Government Printing Office, P. O. Box 37082, Washington, DC 20402-9328

3. The National Technical Information Service, Springfield, VA 22161-0002 Although the listing that follows represents the majority of documents cited in NRC publica-tions, it is not intended to be exhaustive.

Referenced documents available for inspection and copying for a fee from the NRC Public Document Room include NRC correspondence and internal NRC memoranda: NRC bulletins, circulars, information notices, inspection and investigation notices; licensee event reports; vendor reports and correspondence; Commission papers; and applicant and licensee docu-ments and correspondence.

The following documents in the NUREG series are available for purchase from the Government Printing Office: formal NRC staff and contractor reports, NRC-sponsored conference pro-ceedings, international agreement reports, grantee reports, and NRC booklets and bro-chures. Also available are regulatory guides, NRC regulations in the Code o/ Federal Regula-tions, and Nuclear Regulatory Commission issuances.

Documents available from the National Technical information Service include NUREG-series reports and technical reports preparsd by other Federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission.

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Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library, Two White Flint North,11545 Rockville Pike, Rock-ville, MD 20852-2738, for use by the public. Codes and standards are usually copyrighted and may be purchased from the originating organization or, if they are American National Standards, from the American National Standards Institute,1430 Broadway, New York, NY 10018-3308.

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NUREG-1511 Supplement 1 2

Reactor Pressure Vessel Status Report I

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1 Manuscript Completed: October 1996 Date Published: October 1996 i

B. J. Elliot, E. M. Hackett, A. D. Lee J. Medoff, J. R. Strosnider, K. R. Wichman i

Division of Engmeermg Omcc of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Wahington, DC 20555-0001

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ABSTRACT his report describes the issues raised as a result of information needed to assess compliance with the staff's review of Generic letter (GL) 92-01, regulatory requirements and licensee commitments Revision 1, responses and plant-specific reactor regarding RPV integrity. GL 92-01, Revision 1, pressure vessel (RPV) assessments and the actions Supplement 1, was issued as a result of generic issues taken or work in progress to address these issues. In that were raised in the NRC staff's reviews of addition, the report describes actions taken by the licensee responses to GL 92-01, Revision 1, and staff and the nuclear industry to develop a thermal plant-specific RPV evaluations. In particular, an annealing process for use at U.S. commercial nuclear integrated review of all data submitted in response to power plants. This process is intended to be used as GL 92-01, Revision 1, indicated that licensees may a means of mitigating the effects of neutron radiation not have considered all relevant data in their RPV on the fracture toughness of RPV materials, assessments. This report is representative of submittals to and evaluations by the staff as of

%e Nuclear Regulatory Commission (NRC) issued September 30,1996. An update of this report will be GL 92-01, Revision 1, Supplement 1, to obtain issued at a later date.

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d CONTENTS page

, Abstract...................................................................................................................... iii ListofTables.............................................................................................................. vii

. ExecutiveSummary....................................................................................................... ix y;

Abbreviations.............................................................................................................. xiii 4 .

1 Introduction.............................................................................................................. 1-1 2 Generic 1.etter 92-01, Revision 1, Supplement 1: Reactor Vessel Structural Integrity ................... '2-1 4

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.l 2.I Background....................................................................................................... 2-1 '

i 2.2CurrentStatus...................................................................................................

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2.2.1 Reactor Pressure Vessels Fabricated by Babcock and Wilcox ................. .................. 2-2 i f

i 2.2.2 - Boiling-Water Reactor Pressure Vessels . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-2 2.2.3 Reactor Pressure Vessels Fabricated by Combustion Engineering ............................... 2-3 2.2.4 Generic Industry Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-3

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n j 3 Pressurized Thermal Shock (PTS) Evaluations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 4

! 3.1 Revi sion of 10 CFR 50.61. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 ...

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+. 3.2CurrentIssues.................................................................................................... 3-1 3.2.1 Best-Esti mate Chemi stry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.2.2 Use of S urveillance Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1

3. 3 Summary of Generic Assessment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-2 3.4 Plant-Speci fic PTS Assessments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . 3-3 3.4.1 Summary of the Palisades PTS Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-3 3.4.2 Summary of the Calvert Cli ffs PTS Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-4 3.4.3 Summary of the Ginna Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . 3-4 1

4 Reactor Pressure Vessel Thermal Annealing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 l 4.1 G eneral Back g round . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 v NUREG-1511, Supp. I

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t Page 4.2 Thermal Annaaling Process and Technical Background .................................................. 4-1 l

l- 4. 3 Previous E x perience . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3 i

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4.4 Technical Codes and Standards for Thermal Annealing .................................................. 4-3 4.4.1 ASTM Standard E 509 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3 i 8

4.4.2 ASME Code Case N-557 on Thermal Annealing .................................................... 4-3 3

4.5 NRC Annaaling Rule and Regulatory Guide . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3 i

4.6 Overview of Metallurgical and Engineering Issues ...................................................... 4-4 I

,. 4.7 Department of Energy Annealing Demonstration Project ................. .............................. 4-4  !

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4. 8 Palisades Thermal Annealing Report . . . . . . . . . . . . . . . . .. . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-5  ;

1 i 4.9 Summary.......................................................................................................... 4-5 5 Reactor Vessel Integrity Database . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1  !

5.1 Summa ry of Database Features . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 ,

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5.2 Revisions included in the RVID Version 1.1, Revision 1................................................ 5-1 i

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i 5.3 Future Revisions to the RVID . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-2 i

t 6 Summary and Conc lusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 61 i 7 References......................................................................................................... 7-1 i 1

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Table 5.1 - Sample RVID Summary File for Chemistry Data .................................................. 5-3 Table 5.1 (Continued) - Sample RVID Summary File for Chemistry Data ................................... 5-4 8

Table 5.2 - Sample RVID Summary File for Upper Shelf Energy (USE) .................................... 5-5 Table 5.2 (Continued)- Sample RVID Summary File for Upper Shelf Energy (USE) ........... ......... 5-6 Table 5.3 - Sample RVID Summary File for Pressurized normal Shock (PTS) ............................ 5-7 Table 5.3 (Continued)- Sample RVID Summary File for Pressurized Thermal Shock (PTS) ........... 5-8 1

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, EXECUTIVE

SUMMARY

' This NUREG presents the actions taken by the U.S. b results of applying generic values of chemistry  !

Nuclear Regulatory Commission (NRC), as well as and increased margin terms indicated that plants  ;

! nuclear industry owners groups and individual would be predicted to reach the FTS screening i licensees, regarding the ongoing assessment of criteria at an earlier date than that given by the' PTS i

reactor pressure vessel (RPV) integrity. Since the assessment methodology in 10 CFR 50.61.

issuance of Generic Letter (GL) 92-01, Revision 1

, (Ref.1), in March 1992, and NUREG-1511 (Ref. 2), However, with the exception of six RPVs, the staff's in December 1994, the staff has directed its actions generic ===a==enant indicated that the RTm values for

. toward determining the generic implications of the the limiting beltline materials in all pressurized-water larger-than-expected variability observed in the reactor (PWR) RPVs would still be below the FTS

! chemical composition of RPV welds at the Palisades screening criteria at end-of-license (EOL) for the  ;

i Nuclear Power Plant, and assuring that licensees take plants. b limiting RPV in this ===a==aient was the action to assure all relevant data are considered in RPV at the R.E. Ginna Nuclear Power Plant. The i their RPV assessments. W staff has also reviewed generic assessment did not consider plant-specific

the Palisades thermal annealing plan and associated data, which could demonstrate that these six' plants thermal annealing demonstration project activities. In could have longer periods of time to reach the FTS addition, the staff has completed several plant-specific screening criteria.

. pressurized thermal shock (FTS) evaluations, and has

developed and iricorporated enhancements to the Subsequently, the Rochester Gas and Electric ' '

Reactor Vessel Integrity Database (RVID). Company (RG&E, the licensee for Ginna) provided a plant-specific PTS =asa===nent for the Ginna RPV.

During the fall of 1994, the Consumers Power This ===a== ment included RG&E's surveillance

Company (CPCo, the licensee for the Palisades plant) capsule data and all chemistry data representing the i performed material property tests and chemistry Ginna RPV beltline welds. This data indicated that i analyses of newly acquired samples of weld materials the RTns value for the limiting material in the Ginna y i that were removed from the Palisades retired steam - RPV would be well below the FTS screening criteria l generators. When compared to the previous weld at EOL. b staff reviewed RG&E's assessment and j- data, the copper and nickel measurements from the concurred with its conclusions. The plant-specific j retired steam generator welds indicated that the PTS assessment for the Ginna RPV demonstrates that l variability in the weld chemistry was greater than the use of plant-specific data could extend the time  !

l previously anticipated during the development of the for RPVs to reach the FTS screening criteria.

i FTS rule, Section 50.61 to Title 10 of the Code of Federal Regulations (10 CFR), Part 50 (Ref. 3).8 'the staff compiled data from the responses to '

GL 92-01, Revision 1, in the RVID computerized t To address generic issues related to this larger-than- database. Based on review of the data in the database

,. expected variability in weld chemical composition, and plant-specific reviews, the staff concluded that 3

the staff performed a generic PTS assessment and licensees may not have considered all the relevant

issued Supplement I to GL 92-01, Revision i data in their RPV assessments. Therefore, the staff, (Ref. 4). h purpose of the staff's generic PTS issued Supplement I to GL 92-01, Revision 1, in assessment was to demonstrate that there is time to May 1995. & supplement required that all  ;

! address the variability in weld chemistry, h addressees identify, collect, and analyze the impact of l generic FTS assessment used generic chemistry any new data pertinent to the structural integrity of ,

values and increased margin terms to account for the their RPVs relative to the requirements of larger-than-expected variability in weld chemistry. 10 CFR 50.60 (Ref. 5), 10 CFR 50.61, and Appendices G and H to 10 - CFR Part 50 1 Henceforth, all sections to Title 10 of the Code of federal (Refs. 6 and 7), as well as any potential impact on

Regulasiou Part 50 will be abbreviated 10 CFR 50.XX or j low temperature overpressure protection (LTOP) ,

20 CFR 50.XXX, as appropriate.

limits or pressure-temperature (P-T) limits.  ;

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All licensees have responded to GL 92-01, result in recovery of 80% to 90% of the fracture i i Revision 1, Supplement 1. Some licensees have toughness lost as a result of radiation embrittlement.

provided additional data that were not provided in

, their initial response to the GL. However, in regard CPCo has projected May 1998 for the anneal of the

, to GL 92-01, Revision 1 Supplement 1, no licensee Palisades RPV. However, in a letter dated has yet to identify any significant RPV integrity April 4,1996 (Ref. 8), CPCo provided a revised PTS issue. Most licensees have indicated that they are assessment indicating that the RTm value for the I participating in the owners group activities that will limiting material in the Palisades RPV would not .

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e whether new information is available. He exceed the PTS screemag criteria until after EOL. I

industry is coordinating the owners group activities his revised ====-at utilized the best-estimate

through the Nuclear Energy Institute (NEI). The chemistry that was previously reviewed by the staff, Boiling Water Reactor Vessel and Internals Project but utilized a lower projected neutron fluence at (BWRVIP) is coordinating activities for boiling-water EOL. As a result of the reduced neutron fluence, the i reactors (BWRs). The Combustion Engineenng revised IrrS assesament indicated that the Palisades l I Owners Group (CEOG) and Babcock and Wilcox RPV could satisfy the requirements of the IrrS rule  !

Owners Genup (B&WOG) have instituted programs well after the plant's fourteenth refueling outage. l to resolve the issue concerning weld chemistry The staff is currently reviewing CPCo's revised j l variability. assessment. j j ne staff has also reviewed PTS assessments To provide a regulatory framework for thermal l submitted by the licensees for Palisades and Calvert annealing, the staff has issued a new regulation, j Cliffs Nuclear Power Plants (CCNPP). These 10 CFR 50.66 (Ref. 9), as well as Regulatory Guide j licensees provided chemistry data that had not been (RG) 1.162 (Ref.10). The Department of Energy j included in previous ====ments. The additional awarded two contracts to demonstrate the engineering

chemistry data for the CCNPP vessels indicated that feasibility of the thermal annealing technology. The

the RTrrs values for the limiting materials in the first demonstration project was performed at the J l CCNPP RPVs would remain below the PTS Marble Hill facility and employed an indirect, gas-

screemag criteria for up to 20 years after EOL. For fired heating method. The second demonstration

, the Palisades vessel, the additional chemistry data project has been tentatively scheduled to take place at

indicated that the embrittlement of the RPV could be the Midland facility and will employ an electric 4 greater than previously projected, but the RPV would resistance heating approach. The staff has been

still satisfy the requirements of the IrTS rule until the closely following these projects in order to be

! end of the plant's fourteenth refueling outage, in late prepared for the Palisades and other potential 1999. annealing applications. The staff is currently j reviewing the Palisades thermal annealing plan and e Since the Palisades license expires in 2007, CPCo has the two demonstration projects sponsored by the submitted its preliminary thermal annealing plan for Department of Energy.

the Palisades RPV (See Section 4.8 for details and j References). Dermal annealing is a process in The staff has also developed and incorporated 1 which the RPV beltline is heated significantly above enhancements to the Reactor Vessel Integrity its operating temperature for an extended period. Database (RVID). He RVID was developed

[ His process mitigates the effect of neutron radiation following the staff's review of licensee responses to l

. by recovering both . the increase in transition Generic Letter (GL) 92-01, Revision 1. The database temperature (TT) and the decrease in upper-shelf summarizes the properties of the reactor vessel 1

energy _(USE). CPCo's annealing plan for the beltline materials for each operating commercial ,

Palisades RPV addresses the critical engineering and nuclear power plant. The database has been issued to 1

. metallurgical aspects of thermal annealing. The plan all U.S. licensees and some foreign regulatory calls for the annealing to be performed using an authorities. The staff periodically enhances and j

indirect, gas-fired heating method that would heat the updates the database based on feedback from the l 1 reactor vessel beltline region to the 850 F - 900aF industry and revised data from the licensees. The l 4

temperature range for approximately 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />. De RVID enables users to compare data from different l

! licensee projects that this annealing treatment should licensees. In comparing the data, the staff observed 2

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that some licensees reported different data for welds i that were fabricated from the same heats of weld wire. This led the staff to conclude that some liemanaan had not considered all relevant data when )

performmg their RPV integrity assessroents. The next updates to the RVID will incorporate any new information provided by licensees in response to the ,

close<mt letters to GL 92-01, Revision 1, and to l GL 92-01, Revision 1, Supplement 1.

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ABBREVIATIONS l

10 CFR Title 10 to the Code ofFederal MOU memorandum of understanding Regulationr ADP Annenling Demonstration Project NEl Nuclear Energy lastitute AMES Aging Materials and Evaluation NRC Nuclear Regulatory Commission Study ASME American Society of Mechanical NRR Office of Nuclear Reactor Regulation Engineers ASTM American Society for Testing and OG Owners Group Materials B&W Babcock and Wilcox Nuclear P-T pressure-temperature Technologies B&WOG Babcock and Wilcox Owners Group PTS pressurized thermal shock BG&E Baltimore Gas and Electric Company PWR pressurized water reactor BNCS Board of Nuclear Codes and RAI request for additional information Standards (ASME)

BWR . Boiling Water Reactor RES Office of Research (NRC)

BWRVIP Boiling Water Reactor Vessel RG Regulatory Guide and Internals Project CCNPP Calvert Cliffs Nuclear Power Plant RG&E Rochester Gas and Electric Company CEOG Combustion Engineering Owners RPV reactor pressure vessel Group CPCo Consumers Power Company RVWG Reactor Vessel Working Group DOE Department of Energy RVID Reactor Vessel Integrity Database EFPY effective full power years SER Safety Evaluation Report EOL- end of license SNSC Southeast Nuclear Service Center EPRI Electric Power Research Institute TAR thermal annealing report GL Generic Letter TT transition temperature LTOP low temperature overpressure USE upper shelf energy protection xiii NUREG-1511, Supp. I

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l 1 INTRODUCTION N original version of the " Reactor Press are Vessel

• RTume is the initial reference temperature of l Status Report," .NUREG-1511 (December 1994), the unirradiated material '

described the reactor pressure vessel (RPV), and

discussed the effect of radiation embrittlement on

• ARTm is the mean adjustment in reference  !

l RPV materials. NUREG-1511 also identified two temperature caused by irradiation j i - indicators for measuring embrittlement: (1) an

, increase in the ail-ductility transition : c. .,

• M is the margin to be added to cover 11 and (2) a decrease in upper-shelf energy (USE). uncertainties in the initial reference Limits on radiation embrittlement to the RPV are temperature, copper and nickel contents, ,

j defined in the pressurized. thermal shock (PTS) rule, fluence, and calculational procedures. l l Section 50.61 to Title 10, Code of Federal l Regulations, Part 50 (10 CFR 50.61), as well as h screemng criteria are 270*F for plates, forgings,  ;

Appendix G to 10 CFR Part 50. The MS rule and axial welds and 300'F for circumferential welds.  !

contains screemag criteria that limit the increase in When these screening criteria are exceeded, the IrrS i i transition temperature (TT), and Appendix G contains rule requires that licensees perform further plant- i l screemag criteria that limit the decrease in USE. specific evaluations of their reactor pressure vessels NUREG-1511 also summarized the results of the (RPVs) to justify continued operation of their

. staff's review oflicensee responses to Generic letter reactors.

(GL) 92-01, Revision 1, as well as plant-specific

RPV evaluations for all 37 boiling-water reactor Based on the docketed information available at the j (BWR) plants and 74 pressurized-water reactor time NUREG-1511. was issued, Beaver Valley

(PWR) plants in the United States. W data Unit 1 and the Palisades Nuclear Power Plant were resulting from the staff's review are stored in the the only plants projected to exceed the PTS screemng i NRC's computerized Reactor Vessel Integrity criteria prior to EOL. At that time, Beaver Valley i

Database (RVID). Unit 1 (EOL 2016) and Palisades (EOL 2007) were projected to exceed the PTS screening criteria in i This ngvi=*~1 " Reactor Pressure Vessel Status Report" 2012 and 2004, respectively. W Duquesne Light i discusses the staff's basis for issuing Supplement I to Company (the licensee for Beaver Valley Unit 1),

j GL 92-01, Revision 1 (Section 2.1), the status of and the Consumers Power Company (CPCo, the j licensee responses to the supplement (Section 2.2), licensee for Palisades) indicated that the irrS results j the current status of licensee compliance with the for their plants were based on the most current j PTS rule (Section 3), thermal annealing (Section 4), information and were subject to change. In a j and the staff's development of the RVID (Section 5). subsequent PTS assessment for the Palisades RPV, ,

t CPCo provided the staff with additional data l 2

'Ibe PTS rule adopted on July 23,1985, and revised indicating that Palisades would reach the PTS

. on May 15,1991, and December 19, 1995, defines screening criteria as early as 1999. However, the J

screening criteria for embrittlement of RPV materials licensee has recently revised the assessment based on l and actions to be taken if these screening criteria are a new neutron fluence projection, & staff is exmat~I. These screening criteria are given in terms currently reviewing the neutron fluence projection of reference temperature, or RTm, at the end-of- and related PTS assessment. These data and related license (EOL) for the plants. b RTm is defined as analyses are discussed in Section 3.3.1. j

follows

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Appendix G to 10 CFR Part 50 specifies that the  !

RTm = RTnare + ARTm + M USE (as measured from the results of Charpy impact tests) must be greater than 102 joules (75 P.-lb)in the where- unirradiated condition. Furthermore , Appendix G specifies that the USE should remain above 68 joules (50 ft-lb) during the operating lifetime, unless 1-1 NUREG-1511, Supp. I l

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analyses are performed to demonstrate that margins specific analyses. In NUREG/CR-6023, " Generic of safety exist for lower energies. Moreover, these Analyses for Evaluation oflew Charpy USE Effects safety margins must be equivalent to those specified on Safety Margins Against Fracture of RPV in Appendix G to Section XI of the American Society Materials" (Ref.12), the NRC staff concluded that of Mechamcal Engineenng (ASME) Boiler and PWR and BWR RPV materials can have USE values Pressure Vessel Code (Ref.11). less than 68 joules (50 ft-lb) and still provide the required margins of safety against fracture. On the Arough owners groups, the industry performed basis of the industry's equivalent margins analyses analyses to det setrate the USE values necessary to and the NRC's generic study, the staff concluded in satisfy the ASlut! Code for generic groupings of NUREG-1511 that all RPVs will have adequate plants. In addition, some lican=aan performed plant- upper-shelf toughness throughout their current licensed operating life.

NUREG-1511, Supp. I 1-2

2 GENERIC LE' ITER (GL) 92-01, REVISION 1, SUPPLEMENT 1:

REACTOR VESSEL STRUCTURAL INTEGRITY 2.1 Background integrity assessments for their plants:

After evaluating licensee responses to Generic Letter (1) describe the actions taken or planned to locate (GL) 92-01, Revision 1, the staff entered the data all data relevant to the evaluation of RPV from the responses into the Reactor Vessel Integrity integrity Database (RVID). The staff then used the RVID to compare the data received from different licensees. (2) assess any change in best-estimate chemistry As a result, the staff observed that some licensees based on consideration of all relevant data reported different data for welds that were fabricated from the same heat of weld wire. In addition, the (3) determine the need for use of the ratio staff noted that the variability in the amount of copper procedure in RG 1.99, Revision 2, when in welds fabricated from copper-coated electrodes applying surveillance data to RPV integrity was greater than previously estimated. The staff assessments therefore concluded that some licensees had not considered all relevant data when they performed (4) assess the need for revision of the existing RPV their RPV integrity assessments and that welds integrity evaluations, including PTS, USE, fabricated from copper-coated electrodes had larger- P-T limit and LTOP limit evaluations.

than-expected variability in chemical composition.

The staff's review of data from several plant-specific The ratio procedure is defined in RG 1.99, pressurized thermal shock (PTS) assessments Revision 2. When there is clear evidence that the  !

confirmed these conclusions. copper and nickel content of the surveillance weld is  !

different from that of the beltline weld, the The variability in chemical composition highlighted surveillance weld data should be adjusted when the sensitivity of RPV embrittlement to small changes determining the effect of neutron radiation on the in the chemical composition of beltline materials. It beltline weld. The adjustment in the surveillance also emphasized the need for licensees to use all weld is dependent upon the amount of copper and relevant data in their RPV assessments and to adjust nickel in the surveillance and beltline welds.

the surveillance data to the best-estimate chemistry, According to the ratio procedure (as defined in in accordance with the procedures in Regulatory RG 1.99, Revision 2), the measured increases in Guide (RG) 1.99, Revision 2 (Ref.13). transition temperature (IT) from the surveillance data are to be adjusted by multiplying them by the ratio of To obtain information needed to assess the the chemistry factor for the vessel weld to that of the significance of these issues, the NRC issued surveillance weld. These chemistry factors are Supplement I to GL 92-01, Revision 1 (Ref. 4). dependent upon the amount of copper and nickel, and This supplement, dated May 19,1995, requested that are determined from a table in the RG.

all addressees identify, collect and analyze the impact of any new data pertinent to the structural integrity of their RPVs relative to the requirements of 2.2 Current Status 10 CFR 50.60 (Ref. 5),10 CFR 50.61 (Ref. 3),

and Appendices G and H to 10 CFR Part 50 All licensees have responded to GL 92-01, (Refs. 6 and 7), as well as any potential impact on Revision 1, Supplement 1, and some licensees have low temperature overpressure protection (LTOP) provided new data that were not considered in their limits or pressure-temperature (P-T) limits, responses to GL 92-01, Revision 1, in 1992. The staff is currently reviewing the data; however, the More specifically, in GL 92-01, Revision 1, licensees have concluded that the data have no affect Supplement 1, the staff requested that licensees take on previously submitted RPV integrity evaluations.

the following actions with respect to the RPV Owners groups (OGs) have undertaken activities to 2-1 NUREG-1511, Supp. I

I t

9 1

t identify, collect, and report any previously unreported contended that its members have appropriately data that may be relevant to the integrity of RPV considered the relevant data in regard to their reactor l materials. His data search requires the review of vessel integrity evaluations, and have previously  ;

many years of welding records and is not scheduled reported the best-estimate weld metal chemistry i to be completed until the summar of 1997, ne values and valid RPV integrity evaluations. In l

=chantules and associated activities of the OGs are addition, the B&WOG RVWG asserted that its  !

described in the following sections. members need not use the ratio procedure f (as defined in RG 1.99, Revision 2), because the  ;

On March 20, 1996, the staff met with variability in chemistry for the surveillance welds was representatives of Nuclear Energy Institute (NEI), the representative of the variability in chemistry for the l Electric Power Research lastitute (EPRI), industry beltline welds. However, the B&WOG RVWG t OGs, and licensees. A letter (Ref.14) from provided a comparison of RTn, . values when l B. Sheron (NRC) to A. Marion (NEI), dated applying and not applying the ratio procedure. ,

April 18,1996, summarizes the outcome of that Regardless of the computational method, the RTns  !

meeting. The BWR Vessel and Internals Project values for all participating plants were shown to be >

(BWRVIP), Combustion Engineering Owners Group below the PTS screening criteria at the EOL. The l (CEOG) Reactor Vessel Working Group (RVWG), B&WOG RVWG listed 36 reports, previously and EPRI informed the staff of their programs submitted to the NRC, that form the basis for their  ;

i implemented in response to GL 92-01, Revision 1, conclusions. The staff is currently reviewing the Supplement 1. In turn, the staff informed the information reported in BAW-2257, Revision 1.  :

participants of its objectives related to GL 92-01, f Boiling-Water Reactor Pressure Vessels  ;

Revision 1, Supplement 1. As a result, the 2.2.2 I

participants identified two topics where further discussions would prove to be beneficial, including: On August 10,1995, the BWRVIP submitted a report (Ref.17) to address the RPV integrity of all ,

(1) discussions regarding methods being used by BWR/24 plants. According to the report, the long [

industry to arrive at best-estimate chemistry term plan of the BWRVIP is to participate in database j values for families of welds; and activities and cooperate with industry efforts to  ;

develop best-estimate chemistry values and methods l (2) discussions regarding the industry's to account for variability in weld chemistry. His i development and maintenance of a database for report was reclassified to non-proprietary status on l information related to RPV materials. June 27,1996 (Ref.18). [

t L

Reactor Pressure Vessels Fabricated The near-term response is discussed in EPRI Report  :

. 2.2.1 by Babcock and Wilcox No. TR-105908NP, " Bounding Assessment of BWR/24 Reactor Pressure Vessel Integrity Issues  ;

Licensees with pressurized-water RPVs fabricated by (BWRVIP-08NP)* (November 1995, Ref.19). This l Babcock & Wilcox (B&W) are members of the B&W report addresses the effects of chemistry variability Owners Group's (B&WOG) Reactor Vessel Working on USE and pressure-temperature (P-T) limits.2 he Group (RVWG). In a letter from D.L. Howell to BWRVIP evaluations, which used bounding values of .

the NRC, dated August 1,1995 (Ref.15), and in the available weld chemistry data, were performed to accompanying topical report (Ref. 16), the determine the worst case impact of weld chemistry  ;

B&WOG RVWG provided its response to GL 92-01, variability on USE and P-T limits. nese evaluations Revision 1 Supplement 1, on behalf of the indicate that all BWR licensees have satisfied the [

participating licensees in the owners group, existing RPV regulatory requirements. De staff has reviewed and approved the P-T limits and USE  ;

la that report (BAW-2257, Revision 1, dated October evaluations for all plants discussed in the report. l 1995), the B&WOG RVWG indicated that some f 2 EPRI Report No. TR 105908NP. November 1995. does not  ;

additional data were available from domestic BWRs >

address the effects of chemiary variability on FTS and LTOP and foreign PWRs in regard to weld chemistries, and 'b.c.u.e m oper. ting characteristics preclude Pts and initial Charpy V-notch and Drop Weight impact p , '

toughness values. Nonetheless, the B&WOG RVWG NUREG-1511, Supp. I 2-2 i

EPRI Report No. TR 105908NP indicated that weld The task group will also establish a model for chemistry variability had no impact on previously determinmg the best-estimate copper and nickel reviewed BWR USE evaluations. The report also values for each best, based on the available data and indicated that, if bounding chemistry values are used weld process. The final report to the participating (rather than best-estimate values as required in RPV licensees is scheduled for December 31,1996.

integrity mamanamanta), some P-T limits may not meet the safety margins of 10 CFR Part 50, Appendix G. 2.2.4 Generic Industry Activities However, for the limiting BWR operating condition (the leek rate test), the Appendix G safety factor of EPRI has developed an industry dataham entitled 1.5 could only potentially be reduced in the worst "RPVDATA". The database has the following case to 1.3. This demonstrates that the safety factors objectives:

required by Appendix G provide adequate margins even for a plant postulated to have an upper bound

• Combine all available RPV materials data into chemistry. The P-T limits may require revision if the an integrated, common material datah==a.

best-estimate chemistries are revised as a result of new data being collected in response to GL 92-01,

• Develop special data sear:h and retrieval Revision 1. Supplement 1. b industry's effort is capabilities for ease of use.

scheduled for completion in 1997.

• Develop tools to assist utilities in performing 2.2.3 Reactor Pressure Vessels Fabricated vessel evaluations.

by Combustion Engineering

• Establish a convenient mechanism to b CEOG RVWG has initiated a task related to incorporate new information into the datahama, better definition of weld chemistry. This task will compile heat specific information, including copper b initial version of RPVDATA was available and nickel values, and will document the source (s) of through EPRI as of March 1996. RPVDATA the information. The compilation of the data requires included the RVID database that was assembled by the assembly and evaluation of fabrication records in the staff. However, it also included data that the the ABB/CE Southeast Nuclear Service Center staff has not reviewed. b staff plans to review all (SNSC), located in Chattanooga, Tennessee. The existing data, as well as any data resulting records include over 450 boxes of original vessel from OG activities related to GL 92-01, Revision I, fabrication records and drawings. The documents Supplement 1. & staff's goal is to include all RPV include material certifications, procedures, data in a RPV database that can be maintained and specifications, fabrication records, laboratory log updated by the industry with oversight by the NRC book entries, and mspection and test records. (Refer to Section 5.3 for further discussion).

2-3 NUREG-1511, Supp.1 4

3 PRESSURIZED THERMAL SHOCK (PTS) EVALUATIONS 3.1 Revision of 10 CFR 50.61 chemistry compositions.

On December 19, 1995, the NRC revised the FTS Several factors need to be considered for rule (10 CFR 50.61). He revisions to the rule determination of the most accurate estimates of incorporated the following changes: chemical composition, including:

e permission to use generic values of unirradiated e sources of variability (copper coating processes, reference temperatures different from those separate nickel wire feeds, etc.);

specified in the rule, ifjustification is provided; e sample types (i.e. surveillance weld, nozzle o a requirement that the results from plant-specific dropout, etc.);

surveillance programs be integrated into the RTn: estimate if plant specific surveillance data e quantity and pedigree of the data; are deemed credible; e weld wire sources; e incorporation into the rule of the credibility criteria specified in Regulatory Guide e appropriate weighting techniques for the data.

(RG) 1.99, Revision 2; These factors are the subjects of owners group (OG) e a requirement that chemistry factors and margin research programs. He NRC staff and the OGs values be calculated using the methodology and meet regularly to discuss progress on these issues.

values specified in RG 1.99, Revision 2, if credible surveillance data are used to estimate A issue of particular regulatory concern has been the the RTns. fact that a simple average of the data does not represent a best-estimate of the amount of copper in As a result of the revised rule and GL 92-01, welds fabricated from copper-coated filler wire. The Revision 1, Supplement 1, licensees may need to staff's reviews of RPVs with copper-coated filler revise their best-estimate chemistries. In addition, wire indicate that there could be significant coil-to-licensees will need to review their surveillance data coil variability in the amount of copper because of to determine whether the data satisfy the credibility variability in the copper coating of the filler wire.

criteria in the revised rule or RG 1.99, Revision 2 The licensees for Calvert Cliffs and Palisades and whether application of the ratio procedure, as accounted for this variability by determining the best-outlined in the revised rule or RG 1.99, Revision 2, estimate for copper content from a weighted average is warranted. of the test results. In a weighted average, the average copper value from samples that represent more than one coil are weighted in accordance with 3.2 Current Issues the number of coils used to fabricate the weld.

3.2.1 Best-Estimate Chemistry 3.2.2 Use of Surveillance Data He FTS rule requires the use of the best-estimate The revised PTS rule requires that licensees chemical composition (percent copper and percent determine the RTns values from surveillance data nickel) for evaluating embrittlement. The best- when the data meet the credibility criteria defined in estimate for a weld is normally interpreted to be the RG 1.99, Revision 2, or in the revised rule. The use mean of the measured values for weld deposits made of plant-specific surveillance data may result in a from the same heat of weld wire as was used to RTn, value that is higher or lower than the RTns fabricate the critical weld. However, this approach value which would be determined by using the Tables may not always yield the most accurate best-estimate in the revised rule or the RG. The revised rule also 3-1 NUREG-1511, Supp. I

ruguires that the surveillance data must also be ===aammanes are in progress, the staff evaluated all a4usted in accordance with the ratio procedure PWR RPVs using generic chemistry values and specified in the revised rule or RG 1.99, Revision 2, increased margin terms in order to account for the when there is clear evidence that the copper or nickel potential variability in chemistries.' The results of content of the surveillance weld is different from that applying generic values of chemistry and increased of the beltline weld. The staff is continuing to margin terms predict that plants would reach the PTS evaluate the implications of variability in material screening criteria at an earlier date than would be properties and chemistry on the determmation of predicted by applying the PTS assessment credible surveillance data and integrated surveillance methodology in 10 CFR 50.61 to the plant-specific programs. This evaluation is being performed as part data. The staffs generic assessment is documented of the staffs overall RPV integrity program and will in Commission Paper SECY-95-119 (Ref. 20)*.

be addressed as warranted through revisions to RG 1.99, Revision 2, and Appendix H to According to the staffs conservative generic 10 CFR Part 50. ==amanvet, no plant would be predicted to reach the PTS screening criteria in less than 7 effective full-power years (EFPY) from 1995, and most plants l 3.3 Summary of Genanc Assessment would reach the MS screening criteria after the expiration of their current licenses.

Subsequent to the issuance of NUREG-1511, the Iwan=aa for Palisades submitted a revised PTS It is important to emphasize that the staffs generic evaluation for staff review. As part of the revised assessment was an extremely conservative analysis evaluation, the licensee submitted chemistry data for performed solely to demonstrate that there was welds in its retired steam generators. These welds sufficient time available to address the issues were fabricated from the same procedure and weld identified in GL 92-01, Revision 1, Supplement 1.

wire heat lot as were used for fabrication of the The evaluation did not consider plant-specific limiting welds in the Palisades RPV. These data information or surveillance data which the staff 1 indicated that significant variability existed in the deems necessary to accurately assess the life of RPVs reported chemistries (i.e., copper and nickel contents) in the industry.

for welds fabricated from the same heat of weld wire.

The staffs generic assessment was not intended to The staff confirmed that the significant variability in establish the operating life relative to the PTS weld chemistry was a generic issue by searching the screening criteria of 10 CFR 50.61, and the results of RVID and compenng the chemistry data reported by the assessment should not be interpreted in that way. l different licensees The results of the staffs efforts This caution has been substantiated by a subsequent I revealed that different licansaae had reported different plant-specific PTS assessment for the Ginna RPV, chemistry data for welds fabricated from the same which was predicted by the generic assessment to be heats of weld wire. This led the staff to conclude the first plant to reach the screening criteria. The that the variability in the amount of copper in welds PTS assessment, which was based on plant-specific fabncated using copper-coated electrodes was greater data, demonstrated that the Ginna RPV would not than previously estimated. reach the PTS screening criteria prior to EOL. The plant-specific PTS assessment for the Ginna RPV is Section 3.4.1 of this report discusses the staffs plant- documented in a letter from the A. Johnson (NRC) to specific nama=== ant of the implications of the Dr. R. Meeredy, dated March 22,19% (Ref. 21).

significant variability observed in the Palisades Aamierry data. The staff recognized that this 3 nees margin terms were increased using generic data for vi "'

• l""' f **3d **""* -

significant vari-bility could impact other plant RPV evaluations.1he staff is addressing this issue as part 4 3, ,,,,,; ,,,,,;, ,,,,,,,, 3, ,,,,,;,,, g, , y,,,,,,,,,

of its review of plant-specific PTS evaluations, and from Jac k R. stro.nider, nr.nch chief, Maieriet. .nd chemical its ongoing -mant of the PTS rule. Engineering nr nch, Divi. ion of Engmeerins, NRR, to Ashok C. Thadani, Anociate Director for Technology NRR.

dated May 5,1995. This memorandum is included as prt of To ensure that all plants will maintain adequate Co W o e ptSEC W W.

protection against MS events while the plant-specific NUREG-1511, Supp.1 3-2

This evaluation is summarized in Section 3.4.3. The The staff concluded that the coil-weighted aversge industry will continue to perform plant-specific PTS method is the preferable method of determmmg the.

assessments as required by 10 CFR 50.61. The NRC best-estimate percent copper for welds fabricated staff will review plant-specific assessments, and will from weld wire containing highly variable copper perform a systematic reassessment of all the coatings. These tests and analyses indicated that the industry's RPV evaluations as part of its review of degree of embrittlement of the Palisades RPV could the industry's responses to GL 92-01, Revision 1, be higher than calculated in the July 1994 interim Supplement 1. safety evaluation. With the new data included in the evaluation, analyses performed in accordance with the PTS rule indicate that the Palisades RPV will satisfy 3.4 Plant-Soecific IrrS Asseumants the requirements of the FTS rule until the end of the plant's fourteenth refueling outage, scheduled for late 3.4.1 Summary of the Palisades PTS Review 1999.

The staffissued an interim PTS safety evaluation for As part of its review the staff noted that significant the Palisades plant in a letter dated July 12, 1994 variability existed in the reported chemistry data (Ref. 22). In that evaluation, the staff concluded that (i.e., copper and nickel) for the limiting RPV weld, the RTn, value for the limiting weld in the Palisades To assess this concern, the NRC staff employed the RPV would reach the PTS screening criterion in the Palisades plant-specific chemistry and fluence data, year 2004, - prior to the expiration of the Palisades and performed RPV failure frequency calculations operating license in 2007. The staff based this similar to those in SECY-82-465 (Ref. 23), which conclusion on evaluation of the data available at that established the basis for the PTS screening criteria.

time. The staff indicated that this conclusion could These analyses demonstrated that the safety margins change on the basis of test results from the retired intended by the PTS rule will be satisfied through the steam generators, which contained weld metal Palisades fourteenth refueling outage, even when fabricated from the same heats of weld wire (heats considering the variability observed in the Palisades W5214 and 34B009) as were the limiting welds in the chemistry data.

Palisades RPV beltline.

As a result of its evaluations, the NRC determined During the fall of 1994, the Consumer Powers that the Palisades RPV can be operated in compliance Company (CPCo, licensee for the Palisades Nuclear with the requirements of 10 CFR 50.61 through the Power Plant) performed material properties tests and plant's fourteenth refueling outage. The staff's chemistry analyses of newly acquired samples of safety evaluation is contained in a letter from weld material that had been removed from its retired E.G. Adensam to K.M. Haas, dated April 12,1995 steam generators. The copper and nickel (Ref. 24). In October 1995 (Ref. 25), CPCo measurements from the retired stor.m generators were proposed a thermal anneal of the Palisades RPV for added to the previous weld data to determine the best- the plant's thirteenth refueling outage. Thermal estimate values of copper and nickel for the Palisades annealing mitigates the effects of neutron radiation on limiting welds. the RPV materials, and would allow CPCo to operate the Palisades plant beyond the fourteenth refueling To provide a common basis for comparing the copper outage.

measurements from different samples, and to determine a best-estimate weight percent copper, the However, in a letter dated April 4,1996 (Ref. 26),

licensee determined whether the measurements from the licensee provided a revised PTS assessment a sample represented weld metal from one or more indicating that the RTns value for the limiting weld coils of weld wires. The number of coils of weld in the Palisades RPV would not exceed the PTS wire was determined by examining the weld record screening criteria until after the expiration of for the sample and the locations of the measurements Palisades operating license. This revised assessment from the sample. The licensee determined the best- utilies the best-estimate chemistry that was discussed estimate value for copper from a coil-weighted in the safety assessment of April 12, 1995, and average of the samples. utilizes a lower projected neutron fluence at EOL.

The staff is currently reviewing this revised 3-3 NUREG-1511 Supp. I

(

l J

assessment.- If the revised neutron fluence calculation difference in chemistry between the surveillance weld is found acceptable, the Palisades RPV could satisfy and the best-estimate chemistry of the beltline weld, the requirements of the FTS rule well after the BG&E adjusted the surveillance data by applying the plant's fourteenth refueling outage. As a result, CPCo ratio procedure of RG 1.99, Revision 2.

could defer the date for annealing the Palisades RPV beyond 1999. In its safety evaluation of January 2,1996, the staff concluded that the RTns values for the beltline 3.4.2 Su==ary of the Calvert Cliffs PTS Review materials in the CCNPP RPVs would remain below the PTS screening criteria 20 years after the In a letter dated January 2,1996 (Ref. 27), the staff expiration of the operating licenses for the plants.

provided a PTS evaluation of the Calvert Cliffs However, since this conclusion is Ap-6t upon the Nuclear Power Plant (CCNPP), Units 1 and 2. available chemistry and surveillance data, it could be Subsequently, the Baltimore Gas and Electric subject to change as new data become available.

Company (BG&E, the licansaa for CCNPP) updated the RTns values for many of the CCNPP beltline 3.4.3 Summary of the Ginna Review materials as a result ofinformation received from the following sources. In a letter dated March 22,1996 (Ref. 21), the staff provided its safety evaluation of the PTS assessment e Combustion Engineering fabrication records for the Ginna RPV. The MS assessment by the Rochester Gas and Electric Company (RG&E, the e chemical analyses from samples of Shoreham licensee for Ginna) is based on plant-specific RPV RPV weldments and an archived surveillance data. These data included the chemical composition block from the Pilgrim Nuclear Power Station, data frcm two weld dropouts, its surveillance weld and a weld qualification test sample, which were all e surveillance capsule data from McGuire Unit I fabricated from the same heat of weld wire as were and CCNPP Unit 2 used to fabricate the limiting beltline weld in the Ginna RPV. In addition, the assessment included e the most recent flux reduction measurements for irradiated Charpy-V test data from the Ginna CCNPP Units 1 and 2. surveillance capsule welds.

The chemical analyses from the Shoreham weldments Since the Ginna surveillance capsule weld was i and the Pilgrim surveillance block were used to fabricated from the same heat of weld wire as was deterame the hest-estimate chemistries for the used for fabrication of the limiting beltline weld, and CCNPP matenals that were fabricated from the same since the surveillance data met the credibility criteria heats of weld wire. To account for the variability in in RG 1.99, Revision 2, RG&E utilized the the amount of copper, the licensee used a weighted surveillance data to determine the RTns value of the average based on the number of coils used to limiting weld in the Ginna RPV. However, since the fabricate the test welds. A simple average was used best-estimate chemical composition of the limiting 1 to determine the best-estimate for nickel since the beltline weld was different from the best-estimate variability of this element was low. chemical composition of the Ginna surveillance capsule weld, RG&E also applied the ratio procedure BG&E also used surveillance weld data from the (as recommended in RG _ l.99, Revision 2) in the surveillace capsules in the McGuire, Unit i RPV. Ginna PTS evaluation. On the basis ofits evaluation

%e McGuire surveillance weld was fabricated from of Ginna's plant-specific RPV data, RG&E concluded the same heats of weld wire as were used to fabricate that the RTns value of the limiting beltline weld in a weld in the CCNPP, Unit No. I beltline. De the Ginna RPV would be below the PTS screening licensee compared the neutron and thennel criteria at EOL environ = ants in the CCNPP, Unit i vessel to those in the McGuire, Unit i vessel to demonstrate the The staff evaluated the chemical composition data in equivalency of the environments, and to demonstrate a different manner than RG&E. The licensee used a that the McGuire surveillance data were applicable to simple average of the measured copper content values CCNPP, Unit 1. In addition, to account for the to determine the best-estimate for copper. As NUREG-1511, Supp. I 3-4

.1 I

4 discussed in Section 3.2.1, the staff is concerned that to the average copper values from the weld dropouts a simple average of the data may not always than to the average copper values from the represent a best-estimate for copper. He staff's surveillance weld and the weld qualification sample.

review of other R.PVs fabricated from copper-coated Since the thickest weld dropout had the greater filler wire similar to that used in the Ginna RPV number of measurements and should have been indicates that there could be significant coil-to-coil fabricated with more coils, the staff concluded that variability in the amount of copper because of the thicker weld dropout should provide a best-variability in the copper coatings of the filler wires. estimate copper value for the limiting Ginna beltline The licensees for Calvert Cliffs and Palisades weld by conservatively accounting for the coil-to-coil accounted for this variability by determining the best- variability in copper.

estimate for copper from a weighted average of the test results. In a weighted average, the average The weighted-average of the copper measurements copper value from samples that represent more than from the thickest weld dropout resulted in a best-one coil are weighted according to the number of estimate for copper that was slightly greater than that coils used to fabricate the weld, ne staff discussed calculated from a simple average of the this issue with RG&E, RG&E indicated that it had measurements. As a result, the staff calculated a insufficient information to accurately determine the RTns value at expiration of the Ginna operating number of coils used to fabricate the four welds that license that was greater than that calculated by the represent the Ginna data base. licensee (268"F by the staff vs. 265*F by the licensee). However, the staff noted that both Therefore, for the assessment of the Ginna RPV, the calculated RTn, values are significantly less than the staff determined a best-estimate copper on the basis 300"F screening criterion (as stated in 10 CFR 50.61) of the data source, and the number and location of used for evaluation of the limiting beltline weld in the the measurements. Based on the number and location Ginna RPV. The staff therefore concluded that the of the measurements, the two weld dropouts contain Ginna RPV would satisfy the requirements of data from many more weld coils than the data from 10 CFR 50.61 until the EOL of the plant. This the surveillance weld and weld qualification sample. conclusion is predicated upon available chemistry and Hence, to account for the coil-to-coil variability in surveillance data, and could be subject to change as .

the amount of copper, the staff gave a greater weight new data become available.

3-5 NUREG-1511, Supp. I

4 REACTOR PRESSURE VESSEL THERMAL ANNEALING 4.1 General Background screenmg criteria that conservatively limit the allowable decrease in USE.

Reactor pressure vessels (RPVs) are fabricated from thick steel plates and/or forgings that are subject to A licensee can use a staff approved analysis tojustify embrittlement from neutron irradiation in the RPV operation beyond the embrittlement screening criteria beltline region. De embrittlement is manifested as of 10 CFR 50.61 or Appendix G to 10 CFR Part 50, a decrease in the fracture toughness of these or else choose to thermally anneal the RPV.

matenals. His decrease in fracture toughness is Publication of an NRC rule in 10 CFR 50.66 (Ref. 9) pdmarily a function of the following factors: and RG 1.162 (Ref.10) on thermal annealing was completed in February 1996, along with overall

• total amount of neutron irradiation (fluence) revisions to the RPV integrity regulations. However, '

the previous version of Appendix G to e chemical composition of the steels 10 CFR Part 50 recognized that " reactor vessels for which the predicted value of USE at end of life is

• W.eure of the irradiation below 50 ft-lbs or for which the predicted value of adjusted reference temperature at end oflife exceeds In order to limit the amount of neutron irradiation 200"F must be designed to permit a thermal annealing -

damage to the RPV beltline materials, many utilities treatment at a sufficiently high temperature to recover have redesigned their fuel loading patterns to reduce material toughness properties of ferritic materials of the amount of neutron leakage from the core, or have the reactor vessel beltline."

used neutron poisons or shielding to protect the RPV in regions of high neutron flux. However, these Annealing is also an option for extending the service techniques have only a limited effect if incorporated . lives of RPVs beyond the current end-of-license late in the life of the RPV. (EOL) or for establishing less restrictive plant operational pressure-te.mperature (P-T) limits for As discussed in previous sections, the level of startup and shutdown.

embrittlement is particularly sensitive to the chemical composition (specifically, the amounts of copper and nickel) of these steels. He NRC regulations 4.2 hrmal Aaa- line Pracam (10 CFR 50.61, and Appendix 0 to 10 CFR Part 50) and Techmcal Background and Regulatory Guide (RG) 1.99, Revision 2, provide methodologies to conservatively estimate the increase Thermal annealing is a process whereby the RPV in the transition temperature (TT) and decrease in the beltline is heated to a temperature significantly above upper abelf energy (USE) of the beltline materials as the operating temperature and held for an extended a result of neutron irradiation. period. Thermal annealing can be performed either

" dry" or " wet." Dry annealing is performed with the An increased TT makes the RPV beltline materials vessel drained and the fuel and internals removed.

more susceptible to rapid crack growth during startup - Wet annealing is typically performed with the full or shutdown and under accident conditions such as complement of primary coolant using the reactor pressurized thermal shock (ITS). De PTS rule coolant system pumps to provide the heat. The (10 CFR 50.61) contains screening criteria to recovery in a wet anneal is usually limited, since it is conservatively limit the amount of the shift in the TT. practically difficult to achieve a large differential between the operating and annealing temperatures.

De decrease in USE resulting from neutron The present discussion focuses oo dry annealing.

irradiation can create the potential for ductile crack growth under normal operating and accident The success of an annealing heat treatment in conditions. Appendix G to 10 CFR Part 50 contains mitigating the effects of irradiation embrittlement is 4-1 NUREG-1511, Supp.1

typically measured by the percent recovery in both manaaling of commercial RPVs. Server reviewed

'IT and USE. De recovery of as-fabricated data on annealing recovery and reitradiation effects j properties depends on the RPV steel chemistry, as for high-copper welds, and concluded that thermal well as the annealing time and temperature. For a annenling at 850*F can cause significant recovery in specified steel composition, the lower the annealing both the TT shift and reduction in USE. Server also  ;

temperature, the more time is required to achieve a reviewed engineering studies on thermal annealing, given level of recovery. However, the Carbon- Server concluded that annenling of U.S. reactors at Manganese steels and weldments used in western 850"F is feasible using existing conunercial heat RPVs are also potentially susceptible to metallurgical treating methods, but also that plant-specific engi-degradation (e.g., creep, temper embrittlement) when neering problems would need to be resolved. Server l subjected to higher temperatures and longer best also presented a thermal and structural analysis for a treatment durations. Thus a " window" of potcatial typical pressurized water RPV annealed at 850"F.

temperatures and Omes &ta for thermal annenling. This analysis predicted that vessel dimensional his window is typically further constrained by stability would be maintained and that post-anneal

. plant-specificandheating-method-specificoperational residual stresses would not be significant. However, and economic considerations. Server's results also indicated that differential thermal expansion of the RPV during annealing can For a given annenling time and temperature, the potentially lead to excessive bending of attached amount of recovery primarily depends on the level of piping. Server has stated that careful temperature the irradiation ' embrittlement and the chemical control is required during the annealing treatment in composition of the steel. Server (Ref. 28) has shown order to prevent this problem.

that an annealing temperature 100"F above the RPV irradiation temperature is not high enough to obtam Mager and others (Refs. 30 and 31) reported on substantial mechanical property recovery. Therefore, research to determine the extent of fracture toughness to achieve a menazole amount of recovery in a remery as a function of annealing time and relatively short r me, a practical minimum for the temperature for materials that are sensitive to neutron annealmg temperature would be on the order of at embrittlement, ney concluded that excellent least 150"F above the RPV irradiation temperature, recovery of all properties could be achieved by annealing at 850"F for 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />, and that the For typical westem RPVs with a nominal irradiation reembrittlement after annealing would follow the temperature of 550*F, this would imply a minimum same trend as the pre-annealing embrittlement rate.

annealing - temperature of. 700*F. . A maximum These reports also describe a thermal annealing i temperature has not been defined; however, 940*F procedure developed for field application.  :

was agreed upon as the upper limit for ASME Code .

. Case N-557 (Ref. 29; refer to Section 4.4), The The Department of Energy, Sandia National Labs, i 940*F limit was set to limit the potential for creep and EPRI conducted a

• Reactor Pressure Vesses and other founs of metallurgical degradation that can Hermal Annealing Workshop" in Fehniary 1994 result at elevated temperatures. (Ref. 32). The purpose of the workshao was to i provide a forum for U.S. utilities and interested  !

Durations of 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> have been typical for parties to discuss relevant experience er.d aues, and experimental annualing treatments that have been to identify potential solutions and appubes related ,

conducted in the 700*F - 900*F range. For western to thermal annealing of RPVs. An extensive amount RPV steels and weldments, these treatments can of information was prescated ranging from restore the TTs and USEs to more than 90 perce.nt of mechanistic studies to an economic analysis of their initial values. Due to the relative scarcity of potential annealing benefits, data from annealing treatments in the 700"F - 800*F range, proposed annealing treatments in the U.S. In 1995, Eason et al. performed analyses of existing have been focused to occur in the 800"F - 900*F - data on annealing of irradiated pressure vessel steels range. using both mechanistic and statistical considerations. .

Dese analyses led to the development of improved Server (Ref. 28) also summarized the overall state of correlation models for estimating Charpy USE and knowledge (as of 1985) for in-place thennal TT after radiation and annealing. This work is l

NUREG-1511, Supp.1 4-2

rep 6a in NUREG/CR-6327 (Ref. 33) and provides procedures for conducting an in-service thermal the basis for the equations (in RG 1.162) used to anneal of an RPV and for demonstrating the  ;

estimate the degree of recovery in fracture toughness effectiveness and degree of recovery. ASTM properties following annealing. Standard E 509-86 also provides direction for a post-annealing vessel radiation surveillance program.

Also in 1995, Pelli (Ref. 34) performed a " State of ASTM Standard E 509 is currently (1996) undergoing the Art Review on Thermal Annealing." This review a major revision to provide updated gt.idance, led to the conclusion that, although annealing particularly in the areas of technical references and technology has been used successfully on' Russian verification of recovery and re-irradiation VVER-440 RPVs (See Section 4.3), application to embrittlement.

Western-style reactors is more difficult hac- of the need to heat the entire beltline of plate-fabricated 4.4.2 ASME Code Case N 557 vessels. - Furthermore, for the Russian materials, on Thermal Annealing phosphorus was the limiting steel constituent for embrittlement as opposed to copper in U.S. steels. At the ASME Section XI meetings in Chicago in August 1995, the Task Group on Thermal Annealing undertook development of a Code Case on Thermal 4.3 Previous Exnerience Annealing of Reactar Vessels on a high priority basis. The development of the Code Case was Although thermal annealing has not yet been applied requested by the Consumers Power Company (CPCo, to a U.S. commercial power reactor, it has been the licensee for the Palisades plant) and supported by successfully applied to other reactors. Two Western- the NRC. The Task Group appointed a special team style RPVs that have been successfully annealed are to write the Code Case and technical basis document the Army's SM-1 A in 1%7 (Ref. 35), and the BR-3 on an expedited basis. The Code Case (designated in Moi, Belgium, in 1984 (Ref. 36). Both of these N 557) was passed by the ASME main committee on reactors operated at temperatures low enough to December 1,1995 (Ref. 29).

permit " wet annealing" at a temperature of 650"F using the reactor coolant pumps as the heat source. Code Case N-557 addresses annealing conditions (temperature and duration), temperature monitoring, in addition,14 Russian-designed VVER-440 PWRs evaluation of loadings, and non-destructive in Russia, Finland, and Eastern Europe, which examination techniques. Code Case N-557 received operate at temperature conditions similar to those at final approval by the ASME Board of Nuclear Codes U.S. PWRs, have been annealed at temperatures of and Standards (BNCS) on March 19, 19%.

approximately 850"F, using dry air and radiant The supporting technical basis document for heaters as the heat source (Ref. 34). Details of the Code Case N-557 will be published in an appropriate thermal annealing of the Novovoronezh Unit 3 RPV technical journal in 1996.

were reported by a U.S. delegation that witnessed the operation (Ref. 37). An NRC team also witnessed the annealing of the Lovissa Unit 1 RPV in 4.5 NRC Annealine Rule and Regulatory Guide Finland in August 1996.

The thermal annealing rule (10 CFR 50.66) was approved by the Commission and published in the 4.4 Technical Codes and Standards Federal Register on December 19, 1995. The r:ule for Thermal Aa== lina addresses the critical engineering and metallurgical aspects of thermal annealing. The regulatory process 4.4.1 ASTM Standard E 509 outlined in the proposed rule consists of several elements:

General guidance for in-service annealing is provided ,

in Am;rican Society for Testing and Materials e a thermal annealing report (TAR, describing the (ASTM) Standard E 509-86 (Ref. 38). Specifically, licensee's plan for conducting the anneal) to be ASTM Standard E 509-86 prescribes general submitted to the NRC prior to annealing 4-3 NUREG-1511, Supp. I

i 1

e' requirements for detersuning the percent Engmeerms issues that need to be addressed for l

recovery of RPV fracture toughes _ due to thermal annealing include, but are not limited to the annealing and requirements for Hermining, following reembrittlement trends occurring durinpeac*.sr I operations aAer annaaling e development of appropriate thermal and structural models for predicting limiting stress

• confirmation that thermal annealing was conditions and providing guidance for the performed in accordance with the TAR placement and quantity of instrumentation submitted in advance by the licensee e control of thermal gradients during bestup and

• public meetings to be held both before and aAer cooldown to minimize stresses and deformations the anneal to allow the NRC to respond to in the vessel and attachei piping questions from interested parties or individuals e adequate instrumentation (for temperature, 1 The regulatory guide on thermal annealing strains and displacements) for monitoring the  !

(RG 1.162) was proceaned in parallel with the rule response of the RPV and piping to the anneal J package and was published on February 15, 1996. ,

NUREG/CR 6327 (Ref. 33), which provides the e adequate onsite fire protection and proper l supporting technical basis . for irradiation adherence to National fire codes and standards  !

embrittlement recovery from thermal annealing, was (particularly with regard to gas-fired heating  !

issued in March 1995. The work in this report methods)  ;

provides the basis for - the computational embrittlement recovery models in RG 1.162.

• protection of personnel from radiation hazards,  !

including those associated with air-lifting internals within the contamment and placement ,

4.6 Overview of Metallurgical of instrumentation inside the bio-shield cavity and Engineermy lasues l

t

• protection of other equipment, components, and l RG 1.162 contains a detailed listing of metallurgical stavetures a.Tected by the annealing (e.g.,

and engineering issues that need to be addressed for minimizing biv-shield wall temperatures) thermal annealing of an RPV. (Background related t to these issues was presented in Section 4.2.) Details Valuable insight regarding these and other  ;

regarding fracture toughness recovery and engineering issues will be obtained from the j reembrittlement trends are covered in Section 3.0 of Annealing Demonstration Project (s)(see Section 4.7).  !

RG 1.162. Specifically, RG 1.162 presents three r acceptable methods for estimating recovery: I 4.7 Denartment of Enerny Annealme .

(1) use of the vessel surveillance materials Demonstration Proiect l (2) removal of specimens from the RPV beltline The Department of Energy (DOE) is currently supporting thermal annealing for U.S. light water '

(3) a generic computational method power reactors in three phases:

l The RG also provides methods for predicting post- Phase 1: Evaluate Engineering Feasibility l annealing reembrittlement trends. The potential for and Material Property Recovery elevated temperature degradation (e.g., creep, temper embrittlement) of western RPV steels was addressed Phase 2: Assist in Establishing Regulatory in a technical basis document prepared for ASME Requirements Using Experimental Code Case N-557. Elevated temperature degradation and Analytical Data of material properties for Westem-style reactors was not considered to be an overriding concern for Phase 3: (a) Assist in Anneal of an thermal annealing treatments in the range of Operating Plant and (b) Evaluate 700"F -- 900"F. The potential for creep, in Post-Annealing Operability particular, can be minimized by following the guidelines of ASME Code Case N-557. ,

i NUREG-1511, Supp.1 4-4 i

The evaluation of engineering feasibility is referred to regarding thermal annealing of the Palisades RPV as tiae annealing demonstration project (ADP). Two would obviate the need for the plant-specific analysis.

contract awards for the ADP were announced on May 25,1995. Dese contracts ass with two separate In October 1995, CPCo initiated submittal of a report consortia for demonstration of the feasibility of describing the planned thermal annealing of the thermal annaaling technology at two cancelled plant Palisades RPV (Ref. 25). CPCo's plan calls for the  !

sites, Marble Hill, Indiana (CE RPV) and Midland, annealing to be performed using an indirect, gas-fired Michigan (B&W RPV). De first annealing heating method, which would heat the reactor vessel demonstration was performed at the Marble Hill site beltline region to the 850*F - 900*F tempenture and employed an indirect, gas-fired heating method. range for approximately 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />. The licensee De Marble Hill annealing demonstration was projects that this annealing treatment should result in completed in July 1996. De second demonstration recovery of 80 percent to 90 percent of the fracture at the Midland site will employ an electric resistance toughness lost as a result of radiation embrittlement.

heating' approach and is tentatively scheduled for December 1996. De NRC staff has been During the summer outage (May-August,1995) at represented at mestags of both the Mart >le Hill and Palisades, the licensee obtainad baseline information Midland Steering Committees and Design Reviews, on the condition of the vessel insulation and the temperatures of the RPV supposts and the cavity he NRC Office of Research (RES) has the lead in between the vessel and the bio-shield wall. The final representing the NRC's interests in the ADPs. sections of the preliminary TAR (Appendices A and RES has prepared a memorandum of understanding B to Section 1.7 of the prelimmary TAR) were (MOU) regarding NRC participation in the ADPs. submitted to the NRC on April 29,1996 (Ref. 42).

This MOU was signed by NRC and DOE Dese appendices completed CPCo's submittal of the on August 4,1995 (Ref. 39). The NRC's Office of preliminary TAR for the Palisades RPV. The report Nuclear Reactor Regulation (NRR) staff has worked is currently being reviewed by the staff. CPCo will closely with RES as observers of the .;nealing be relying heavily on the results of the Marble Hill demonstration projects in order to be prepared for the demonstration anneal (described previously) for Palisades and other potential annealing applications, completion and verification of the Palisades submittal.

The submittal process is expected to be completed by end of 1996, when the results from the Marble Hill 4.8 Pali-t- Hermal Aaa=tian Reoort demonstration anneal are expected to be published.

The licensee is currently projecting that the anneal of la the fall of 1994, the Consumers Power Company the Palisades RPV will commence in May 1998.

(CPCo), the licensee for the Palisades plant, developed chemical composition and mechanical In addition, on April 4,1995, CPCo submitted to the property data for welds removed from their retired NRC a revised neutron fluence analysis for the steam generators (Refs 40 and 41). This new Palisades RPV (Ref. 26). The analysis projects a information changed the best estimate chemistry of significantly reduced neutron fluence at EOL for the the limiting RPV beltline weld, his information also RPV. If approved by the NRC, this analysis could indicated an increased variability in chemical enable operation of the plant well beyond 1999 composition of the weld when compued to that without annealing.

assumed for the development of the PTS rule. In combination, this information indicated that the plant would exceed the PTS screening criteria prior to 4.9 Smaa-v EOL (2007). The staff issued a safety evaluation report (SER) regarding the variability of the Palisades The future is difficult to predict regarding thermal RPV weld properties on April 12.1995 (Ref. 23). annealing of U.S. commercial nuclear power The staff agreed with the licensee's best-estimate reactors. The commitment to anneal an RPV analysis of the chemical composition of the RPV involves significant engineering and regulatory welds and concluded that continued operation through analyses and the assignment of substantial resources.

late 1999 was acceptable. 10 CFR 50.61 requires However, the. approach can reverse neutron submittal of a plant-specific analysis justifying irradiation embrittlement, thereby decreasing operation beyond the screening criteria at least constraints on plant operation. This approach can 3 years before exceeding the criteria. In the SER, the enable operation to EOL for plants potentially staff recognized that submission of information challenged by the PTS screening criteria, and extend 4-5 NUREG-1511, Supp. I

oper' tion beyond EOL for others. In Sandia National diff:rences in the v:lue of annealing for a plant.

Laboratories Report SAND 94-1515/1 (Ref. 32),

Griesbach examined the value of annealing, and Successful demonstration of the engineering concluded that each plant-specific case. must be feasibility of annealing technology in the DOE evaluated and compared to other alternatives. programs (see Section 4.7) will greatly facilitate important in this regard is the fact that uncertainties future considerations for thermal annealing in the in RPV material properties can resultin significant United States.

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NUREG-1511, Supp.1 4-6

5 REACTOR VESSEL INTEGRITY DATABASE 5.1 S-v of Database Feelgrsg The RVID will be updated periodically to reflect the latest available information. Responses to GL 92-01, ne Reactor Vessel lategrity Database (RVID) was Revision 1. Supplement 1, and to the close-out letters devek. ped fol!cwing the NRC staff review of licensee to GL 92-01, Revision 1, are not necessarily reflected responses to GL 92-01, Revision 1. The database in the current version, but will be included in a future summarizes the properties of the reactor vessel version of the RVID. Revisions that were made to beltline materials for each operating commercial make the original version of the RVID (Version 1.1) nuclear power plant, more user friendly are described in the following section.

In addition to the licensee responses to GL 92-01, Revision 1, various documents were included in the review process and development of the RVID. These 5.2 Revisions included in the documents include surveillance capsule reports; RVID Version 1.1. Revision I documents referenced in licensee responses to GL 92-01, Revision 1, submittals; PTS submittals, The RVID was revised and the user's manual was P-T limits reports; and responses to staff's requests expanded as a result of the staffs assessment of for additional information (RAls). The data from Version 1.1 and comments from database users. The these source documents were reviewed and database with the current changes was issued on the documented in the RVID tables. world-wide web as Revision 1 in June 1996. The staff's revisions of the RVID included the following he RVID was designed and developed to reflect the changes:

current status of RPV integrity, and the data and information is consolidated in a convenient and e ne table headings for the PTS summary files accessible manner. Some of the data categories are were revised to read " Summary File for PTS" inputs of docketed information; others are computed for PWRs and " Summary File for Pressure-values, which are not necessarily docketed. He Temperature" for BWRs in order to reflect that programming logic used for calculations in the RVID BWRs do not have PTS evaluations.

follows the methodology in Regulatory Guide 1.99, Information regarding P-T limits for BWR Revision 2. beltline materials (not PTS evaluations) are contained in the RVID.

The RVID includes four tables (summary files) for each plant:

• A means of identifying surveillance data as credible or non-credible was added to the (1) background information table, database.8 (2) chemistry data table (3) upper-shelf energy table, 5 At present, with the exceptionof the surveillance data for eight units, each surveillance data file is defauhed to Credible "Y",

implying mat, for me net pf me units wim suneinana data, (4) P-T limits or PTS table all data are credible. %c eight units were assessed by the staff because the RVID reDected that surveillance data were being References and notes following each table document utilized for calculation of the chemistry factor for the linsing the source (s) of data and provide supplemental material in the RPV. He licensee's for Kewaunee. Indian Point 2. Indian Point 3. Maine Yankee, Robinson 2 and North information. Additionally, the RVID includes sort Anna 1 had two or more sets of credible surveillance capsule and data search capabilities. Users can select a data for their limiting material (reflected as credible *Y" in the des, i red grouping of plants and then specify database). he licensee's for Haddam Neck and Diablo information categories to search and list. Tables 5.1, Canyon I were determined to have less than two credible sets 5.2, and 5.3 provide examples of the Chemistry of surveiHance capsule data for the limiting material in their Data, USE, and PTS /P-T Limit Summary Files that vessels (reDected as credible *N* in the database). He staff

• I" * "U""* ' ""I'* '""*i"'"" d *"d ** "PP"P"

may be accessed by the user in the RVID. changes for non-credible surveillance data wdl be made accordingly in future versions of the RVID, 5-1 NUREG-1511, Supp.1

O ne code for deternuni:3 the percent drop in 5.3 Future Revisions to the RVID i USE was modified to correctly calculate USE ~

values for beltline materials with very high There are short-term as well as long-term goals for copper values. improving the RVID. Revision 1 was issued on the world wide web in June 1996, incorporating the The staffs revisions to the user's manual included the changes described above. However, for many units followmg changes: the information provided in licensee responses to the staffs close-out letters to GL 92-01, Revision 1, was e .The appropriate configuration of the not included in Revision 1 to the RV10 ne "config.sys" file was outlined, and an information provided in licensee responses to explanation included to indicate why the file and GL 92-01, Revision 1, Supplement I was also not buffer configurations in the computer file are included in Revision 1 of the RVID. Licensees have important, nis "config.sys" file should be joined wig owner's groups in order to address the stored in memory for the user to be capable of issues raised in the GL. The new information is successfully runnmg the RVID. expected to be available at the end of 1997, and the RVID will be updated accordingly.

e ne user's manual was modified to include a description of the types of problems that could ne data environment will eventually be switched to arise while runmng the RVID in a networked a more user-friendly format. Microsoft Access 7 )

environment, and specifically during software will be used to bring the database into the I simultaneous, multi-user access of the system.' Windows environment (as opposed to the current  ;

DOS-based FoxPro software)'. The original 1 e lastructions were added regarding how to cmate version of the RVID is now contained in the industry j a batch file that is needed for successful running database known as "RPVDATA," which is

]

of the RVID.- maintamed by the EPRI. RPVDATA runs under >

Microsoft Access and is a flat file (no computations e The user's manual was revised to clarify that are performed within the database).

P-T limits reports, and not just PTS reports, are available in the RVID. RPVDATA also contains several " tiers" of l information that are not in the RVID. Examples j e Expanded information was provided regarding include details of weld property sampling and  ;

how to find a specific plant record while in the comparisons of licensee "best-estimate" chemical  ;

plant information screen. composition values with docketed information. The NRC staff and industry are currently working to e A paragraph was added stating that the database resolve inconsistencies between the two databases.

could contain information that has not been Subsequent to the completion of the GL 92-01, submitted to the NRC's Document Control Desk Revision 1 Supplement I effort, the NRC will make

(" docketed" information), revisions to the RVID. At that time, it will be possible to compare the NRC and industry databases, nese changes to the RVID and to the user's manual and establish an NRC approved database, are outlined in NRC Administrative letter 95-03, Subsequently, it may be possible to have maintenance Revision 1, dated July 10,1996 (Ref. 43). of the database performed by the industry with NRC oversight. The NRC will be able to verify the updates by comparisons with docketed plant 6 The RVID was not designed to work in a networked information.

environment. ' h can function under this scenario; however.

error menenges will appear when muhiple users are logged into 7 MicrosoA Access" and FoxPro are trademarks of the the symern.

MicrosoA Corporation.

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NUREG-1511, Supp.1 5-2 l

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Chemestry f acter for plate CS114-1 mes cattuteted ussng amot sur westeense eeste esported en taas 2015. Rev. 1

.L Table 5.2 - Sample RVID Summary File for Upper Shelf Energy (USE) 90

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Chemistry f atter for W 112 sold esse teleutetod frem Ocense 1. Point Desch 2, MtEIE, esO-1, and Sanche late surwettlere easte that mes reported in BAu 1833, Rev. 1. These surweettence notes essee febr6ceted meth the seen heet h as NF-112.

Stindt, for matte bett forging is e seen watue from 24 forings steiler to ATE 131. The date is reported in Bau 190ser asui has a stambard eleveietten of b SteF.

f tuence ente report en tant-N22, uueE for ptetes C5120-2, CS114-2, C5120-1 wul C $114-1 reported in enet-2222.

UWSE for forgtg Afa 131 was 95/95 tetorere toeit that ens reported en RAW-2222.

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tanE for forsteg afu 131 use 95/95 toleroice timet that una reported in gau-2222.

Table 5.3 (Continued) - Sample RVID Summary File for Pressurized Thermal Shock (PTS)

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6

SUMMARY

AND CONCLUSIONS De NRC staff has reviewed licensee m r..ac; to information reported by the licensees and are subject GL 92-01, Revision 1, Supplement 1. The staff has to change. The dates at which the RTns values for also performed a generic PTS assessment of all the limiting materials in the vessels are projected to pressurized-water RPVs. The generic assessment exceed the screening criteria may change as a result was evaluated by the staff using plant-specific of new surveillance data and additional analyses.

surveillance data. On the basis of these reviews, the NRC staff has confirmed that the RTns values for Also, by implementing different fuel management most of the domestic RPVs are not projected to techniques and inserting special neutron absorbing exceed the pressurized thermal shock (NS) screening materials in the reactor core, licensees may be able to criteria prior to the end of their current operating reduce the irradiation levels sufficiently to stay below licenses. the screening criteria. In addition, licensees may anneal the RPV to recover a large percentage of the  :

As discussed in the original " Reactor Pressure Vessel vessel's fracture toughness lost to neutron irradiation. )

Status Report" (NUREG-1511), the RTns values for  !

the limiting materials in the Beaver Valley Unit I and la their responses to GL 92-01, Revision 1, Palisades vessels were the only plants projected to Supplement 1, the industry's Owners Groups (OGs) exceed their IrTS screening criteria prior to expiration informed the NRC that additional data and of the plant operating licenses.' However, based on information will be submitted for review by the staff.

the results of subsequent chemical composition and The OGs' programs include extensive searches for.

mechanical properties tests of weld materials from relevant data and the development of methodologies Palisades' retired steam generators, the Consumers for determining the best-estimate chemistries for I

Power Company (CPCo, the licensee for the welds fabricated using copper-coated electrodes. The Palisades plant) projected that the degree of last of the OG's programs is not expected to be embrittlement of the Palisades RPV could be greater completed until the summer of 1997. The staff's than previously predicted. As a result, CPCo review of the OGs' data will include a reassessment concluded that the RTns value for the limiting of each licensee's RPV. Atler completing this material in the Palisades RPV would exceed the PTS review, the staff will incorporate any new data into screening criteria in 1999. the RVID and prepare another update of this report.

Recently, however, in a letter dated April 4,1996, it is difficult to predict what the status will be CPCo provided a revised PTS assessment that regarding thermal annealing of commercial RPVs in l projected a lower neutron fluence at the expiration of the United States. He commitment to anneal an i the Palisades operating license. As a result of the - RPV involves significant engineering and regulatory l reduction in neutron fluence, CPCo concluded that analys,es and the assignment of substantial resources. l the RTm value for the limiting material in the However, the approach can recover a large Palisade RPV would not reach the PTS screening percentage of the fracture toughness lost to neutron ,

criteria until many years after 1999. This revised irradiation, thereby decreasing constraints on plant PTS assessment is being reviewed by the staff, operation. This approach will enable plant operation to EOL for RPVs potentially challenged by the PTS It is important to note that the. staff and licensee screening criteria, and extend operation beyond EOL assessments are based on currently available for others.  ;

Successful demonstration of the engineering 8 Based on information available in 1994, the RTm values for feasibility of annealing technology in the DOE the limiting materials in the Beaver Valley 1 and Paliandes programs will greatly facilitate future considerations RW8 MM Projected to exceed the FTS screening limit in 2012 '

for thermal annealing in the United States.

and 2004 prior to EOL in 2016 and 2007, napectively.

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l 7 REFERENCES

1. Generic letter 92-01, Revision 1, ' Reactor Vessel Structural Integrity,10 CFR 50.54(f)," March 6,1992.

2. NUREG-1511, " Reactor Pressure Vessel Status Report," December 1994.

3. Title 10, Code ofFederal Regularlow, Section 50.61, " Fracture Toughness Requirements for Protection

, Against Pressurized 'liermal Sbock Events.'

4. Generic Letter 92 01, Revision 1, Supplement 1, " Reactor Vessel Structural Integrity," May 19, 1995.

5. Title 10, Code ofFederal Regularlow, Section 50.60, "A~t e Criteria for Fracture Prevention 4 Measures for Lightwater Nuclear Power Reactors for Normal Operation."

6. Title 10, Code ofFederal Regularlow, Past 50, Appendix G, " Fracture Toughness Requirements."

4

7. Title 10, Code offederal Regulationr, Part 50, Appendix H, " Reactor Vessel Material Surveillance Program Requirements.'

s

! 8. Letter from R.W. Smedley, Licensing Manager, Consumers Power Company, to the U.S. Nuclear

Regulatory Commission Document Control Desk, " Docket 50-255 - License DPR Palisades Plant

Updated Reactor Vessel Fluence Values," April 4,1996. j i

9. Title 10, Code ofFederal Regulationr, Section 50.66, ' Requirements for Thermal Annealing of the Reactor Pressure Vessel."

10. Regulatory Guide 1.162, " Format and Content of Report for Thermal Annealing of Reactor Pressure i Vessels," February 1996. I

11. Appendix 0 to Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code.

12. NUREG/CR-6023, " Generic Analyses for Evaluation of Low Charpy USE Effects on Safety Margins Against Fracture of RPV Materials,' Oak Ridge National Laboratory, July 1993,

13. Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," May 1988.

14. Letter (untitled) from B.W. Sheron, Director - Division of Engineering, Office of Nuclear Reactor Regulation, to A. Marion, Nuclear Energy Institute, enclosing internal NRC Memorandum of April 15, 1996, ' Summary of Meeting Held with the Nuclear Energy Institute (NEI) and Industry Representatives to Discue Reactor Vessel Integrity," April 18,1996.

g

15. Letter JHT/95-77 from D.L. Howell, Acting Manager - Licensing Services, the Babcock and Wilcox Owners Group, to the U.S. Nuclear Regulatory Commission, "B&W Owners Group Reactor Vessel Integrity Program," August 1,1995.

16. Babcock and Wilcox Topical Report BAW-2257, Revision 1 "B&W Owners Group Reactor Vessel Working Group Response to Generic Letter 92-01, Revision 1, Supplement 1,* October 1995.

17. Ietter 95-404 from the J.T. Ber4 ham, Chairman - BWR Vessel and Intemals Project, to the U.S.

Nuclear Regulatory Commission, forwarding 'BWRVIP Response to NRC Generic letter 92-01, Revision 1, Supplement 1, Reactor Vessel Structural Integrity," August 10,1995.

7-1 NUREG-1511, Supp.1

)

i i

V i

i

18. Letter %-335 from Vaughn Wagner, Technicci Chairman, BWRVIP Integration Committee, to the U.S.

I Nuclear Regulatory Commission, " Reclassification of a BWRVIP Report Submitted in Response to Generic letter 92-01, Revision 1, Supplement 1," June 26,1997. ,

l 19. EPRI TR-105908NP, "BWR Vessel and laternals Project, Bounding Assessment of BWR/2-6 Reactor i

Pressure Vessel Integrity Issues (BWRVIP-08NP)," November 1995. l 4

l

20. . SECY-95-117 " Status of Reactor Pressure Vessel Issues," May 8,1995. l

, 21. Letter f om A.R. Johnson, Project Manager, Project Directorate I-1, Division of Reactor Projects I/II, Office of Nuclear Reactor R.egulation, U.S. Nuclear Regulatory Commission, to Dr. R.C. Meeredy, Vice President - Nuclear Operations, Rochester Gas & Electric Company, "R.E. Ginna Nuclear Power Plant, Pressurized Thermal Shock Evaluation (TAC No. M93827)," March 22,1996, i

h ' 22. Letter from A.H. Hsia, Project Manager, Project Directorate Ill-1, Division of Reactor Projects, Office of i i Nuclear Reactor Regulation, to R.A. Fenech, Consumers Power Company, " Palisades Plant - Interim )

- Safety Evaluation of Pressurized Thermal Shock (PTS) and Upper-Shelf Energy (USE) Analyses," July 12, l l- 1994 i

23. SECY-82-465, " Pressurized nermal Shock," November 23,1982.

24. letter from E.G. Adensam, Acting Director - Division of Reactor Projects Ill/IV, Office of Nuclear Reactor Regulation, to K.M.' Haas, Plant Safety and Licensing Director, Palisades Plant, " Palisades Plant

- Pressurized Thermal Shock Safety Evaluation (TAC No. M83227)," April 12,1995.

t

25. letter from R.W. Smedley, Manager - Licensing, Consumers Power Company, to the  !

t U.S. Nuclear Regulatory Commission, " Docket .50-255 - License DPR Palisades Plant 10CFR50.61

- Prelinunary Thermal Annealing Report Section 3," October 12,1995. +

26. Imater from R.W. Smedley, Manager - Licensing, Palisades Plant, to the U.S. Nuclear Regulatory i^

Commission Docurnent Control Desk, " Docket 50-255, License DPR-20, Palisades Plant Updated Reactor Vessel Fluence Values," April 4,1996.

27. letter from D.G. Mcdonald, Jr., Senior Project Manager, Project Directorate I-1, Division of Reactor  !

Projects I/II, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, to R.E.

Denton, Vice President - Nuclear Energy, Baltimore Gas and Electric Company, " Updated Values for Pressurized normal Shock Reference Temperatures - Calvert Cliffs Nuclear Power Plant, Unit Nos.1 and 2 (TAC Nos. M93230 and M93231)," January 2,1996.

28. NUREG/CR-4212 "In-Place Thermal Annealing of Nuclear Reactor Pressure Vessels," W.L. Server, April 1985.

l

29. ASME Code Case N-557, "In Place Dry Annealing of a PWR Nuclear Reactor Vessel," March 1996.

30. T.R. Mager, " Feasibility of and Methodology for Thermal Annealing of an Embrittled Reactor Vessel,"

EPRI NP-2712, Vol. 2, Electric Power Research Institute, Palo Alto, CA, November 1982.

31. - T.R. Mager et al., " Thermal Annealing of an Embrittled Reactor Vessel, Feasibility and Methodology,"

EPRI NP-6113-SD, Electric Power Research Institute, Palo Alto, CA, January 1989.

I 32.' SAND 94-1515/1, " Proceedings of the DOE /SNL/EPRI Sponsored Reactor Pressure Vessel Thermal l Annealing Workshop," September 1994. 1 l

33. NUREG/CR-6327. "Models for Embrittlement Recovery Due to Annealing of Reactor Pressure Vessel I Steels," E. Eason, J. Wright, R. Odette, E. Mader, February 1995.

NUREG-1511, Supp.1 7-2

- - - - - . - . . - . - - - ~ . - . _ . _ - _ . _ - - - . _ . . - - - , - . _ -

f

34. R. Pelli and K. Torronnen, ' State of the Art Rrview on 'Ihermal Annealing," European Commission, DG XII - Joint Research Centre, Institute for Advanced Materials, DG XI - Safety of Nuclear Installations,

, Aging Materials and Evaluation Studies (AMES) Report No. 2, December,1994.

l 35. U. Potapovs, J.R. Hawthorne, and C.Z. Serpen, Jr., " Notch Ductility Properties of SM-1 A Reactor Pressure Vessel Following the In-Place Annealing Operation," Nuclear Applications, Vol. 5, No. 6, 3

pp. 389-409,1%8.

7

36. A. Fabry et al., " Annealing of the BR-3 Reactor Pressure Vessel," in Proceedings of the 7keffth Water

Reactor Sqfety Research I.; formation Meeting, NUREG/CP-0058, Vol. 4, pp.144-175, NRC, January 1985.

1-j

37. N. M. Cole and T. Friderichs, " Report on Annealing of the Novovoronezh Unit 3 Reactor Vessel in the i USSR," NUREG/CR-5760 (MPR Associates, Inc., MPR-1230), NRC, July 1991.

38. American Society for Testing and Materials, " Recommended Guide for in-Service Annealing of Water-l Cooled Nuclear Reactor Vessels," ASTM E 509-86, Philadelphia,1986.

39. letter from J.M. Taylor, Executive Director for Operations, U.S. Nuclear Regulatory Commission, to Dr. T.R. lash, Director - Office of Nuclear Energy, Science and Technology, 3

U.S. Department of Energy, enclosing 2 copies of the Memorandum of Understanding for the DOE l Annealing Demonstration Project, August 4,1995.

40. Consumers Power Company submittal to the U.S. Nuclear Regulatory Commission, Revision I to l

Calculation No. EA-RDS-94-02, " Evaluation of Palisades Current PTS Screening Criteria Margin," dated )

l November 8,1994.

i

41. letter from R.A. Fenech, Vice President - Nuclear Operations, Consumers Power Company, to the j U.S. Nuclear Regulatory Commission, " Docket 50-255 - License DPR Palisades Plant 10CFR50.61 l - Pressurized Thermal Shock - Additional Preliminary Information," November 10,1995

42. Letter from R.W. Smedley, Manager - Licensing, Consumers Power Company, to the U.S. Nuclear Regulatory Commission, " Docket 50-255 - License DPR Palisades Plant 10CFR50.61 l

- Preliminary Thermal Annealing Report, Thermal Annealing Operating Plan, Section 1.7, Thermal and Stress Analysis, Appendix A and Appendix B," April 29,1996.

! 43. NRC Administrative Letter 95-03, Revision 1, " Availability of Reactor Vessel Integrity Database,"

July 10,1996.

73 NUREG-1511 Supp. I

(ZC F01u 336 U.S. NUCLE AR REGULATORY COMMIS$10N 1. hEPOAT NUMBE M RC i nc2, N $fM 22ci.22n BIBLIOGRAPHIC DATA SHEET (See snstructoong On the rowerse) j

2. TITLE AND SUBTITLE )

Reactor Pressure Vessel Status Report 2. DATE REPORT PuBoiSsED ,

MONTH VEAR l I

October 1996

4. FIN OR GR ANT NUMBER

5. AUTHOR (S) 6. TYPE OF REPORT B.J. Elliot, E.M. Hackett, A.D. Lee Technical J. Medoff, J.R. Strosnider, K.R. Wichman

1. PE R l00 COV E R E D sincouso e carest B. P FORMI G ANIZATiON - N AME AND ADDR ESS ist Nnc. preem Dessoa, Offsce or nerica, u.E Nurmar neeuderory commssanoa, and marsene saareas. st contractor, prov or Division of Engineering Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001

9. PONSO G AN1ZATION - N AME AND ADDR ESS tot Nnc. tree "some es anose"; or contracror, proem Nnc Osutinen, Orrue or nopton, u 1 enemar noeuurary comm, won.

Same as above 1

10. SUPPLEMENTARY NOTES

11. ABSTR ACT L200 worar er mas This report describes the issues raised as a result of the staff's review of Generic Letter (GL) 92-01, Revision 1, responses and plant-specific reactor pressure vessel (RPV) assessments and the actions taken or work in progress to address these issues. In additiori, th'e report describes actions taken by the staff and the nuclear industry to develop a thermal annealing process for possible use at U.S. commercial nuclear plants to mitigate the effects of neutron radiation on the fracture toughness of RPV materials.

The Nuclear Regulatory Commission (NRC) issued GL 92-01, Revision 1, Supplement 1, to obtain information needed to assess compliance with regulatory requirements and licensee commitments regarding RPV integrity. GL 92-01, Revision 1, Supplement 1, was issued as a result of generic issues raised in the NRC staff's review of licensee responses to GL 92-01, Revision 1, and plant-specific RPV evaluations.

In particular, an integrated review of all data submitted in response to GL 92-01, Revision 1, indicated that licensees may not have considered all relevant data in their RPV assessments.

12, KE Y WORDS/DESCR:PTOR $ ttest woron er pa, es ener w,n essar asse crases m ancerme rne ,voorr.s 13. ava LAasuiv sTAituENT Generic Letter (GL) 92-01, Revision 1; GL 92-01, Revision 1, , ,,

]f", ,"f,te{ ,, ,,,o, Supplement 1; Pressure-Temperature Limits; Pressurize,d Thermal """

Shock; Radiation Embrittlement; Reactor Pressure Vessel; l Upper-Shelf Energy; 10 CFR 50.61; 10 CFR 50.66; Appendix G to y, U", classified 10 CFR Part 50 i Thermal Annealing Unclassified is. NUMBER OF PAGES

16. PRICE M.C FORM 335 (249)

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