ML20217H997

Summary of ACRS Subcommittee on Matl & Metallurgy Meeting on 970415-16 in Rockville,Md Re GLs Associated W/Sg Tube Insp Techniques,Effective Use of Ultrasonic Testing Techniques in ISI Programs & Degradation of SG Internals

ML20217H997
Person / Time
Issue date:	04/28/1997
From:	Advisory Committee on Reactor Safeguards
To:	Advisory Committee on Reactor Safeguards
References
ACRS-3057, NUDOCS 9708130377
Download: ML20217H997 (8)

Text

/ IggIFIg ann CERTIFIED: May 1. 1997 issued: April 28, 1997 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS MINUTES OF ACRS SUBCOMMITTEE MEETING ON MATERIALS AND METALLURGY APRIL 15-16, 1997 ROCKVILLE. MARYLAND The ACRS Materials and Metalluray Subcommittee held a meeting on April 15-16 1997, at 11545 Rockville Pike, flockville. Maryland, in Room T-2 B3. The purpose of the meeting was to hold discussions with representatives of the NRC staff and Consumers Energy cincerning generic letters associated with steam gercrator tube inspection techniques, effective use of ultrasonic testing techniques in inservice inspection programs, degradation of steam generator internals, and degradation of reactor vessel head penetrations: and the status of issues related to reactor pressure vessel integrity. The entire meeting '

was open to public attendance. Mr Noel Dudley was the cognizant ACRS staff engineer for this meeting. The meeting was convened at 1:00 p.m. on April 15 and adjourned at 2:40 p.m. on April 16. 1997.

ARENDIES:

625 W. Shack. Chairman R. Seale. Member T. Kress. Member NRC STAFF J. Strosnider. NRR M. Mayfield. RES E. Sullivan, Jr., NRR J. Muscara, RES R. Hermann. NRR C. Fairbanks. RES S. Coffin. NRR D. Jackson. RES P. Rush, NRR E. Hackett. RES L. Lois. NRR G. Carpenter NRR A. Lee. NRR j2.5 0/

INDUSTRY REPRESENTATIVES i

J. Hanson. Consumers Energy K. Cozens Nuclear Energy Institute R. Snuggerud. Consumers Energy S. Anderson. Westinghouse Electric Corporation There were no written comments or requests for time to make oral statements received from members of the public. A list of meeting attendees is available in the ACRS office files. Dr. Shack had a conflict of interest regarding steam generator tube inspections steam generator internals degradation, and inspection of reactor vessel head penetrations. He did not participate in the deliberations on these issues. p g 1 ? ?10 yc;g,--G,,wt- g  ;

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. Minutes: .-April 15 16.-1997 2 Materials. & Metallurgy Subcomittee INTRODUCTION:

l Dr. William Shack. Materials and Metallurgy-Subcomittee Chairman, convened the meeting at 1:00 p.m. on April-15,-1997, explained the purpose of the meeting. and called on Mr. Jack Strosnider, Office of Nuclear Reactor Regulation (NRR), to provide opening comments for the staff. Mr. Strosnider explained the reorganization of the Materials and Chemical Engineering Branch, noted that several important topic would be presented by the staff, and comented on the good cooperation and coordination between the Offices of Nuclear Regulatory Research (RES) and NRR in the area of materials and metallurgy.

PROPOSED GENERIC LETTER REGARDING DEGRADATION OF STEAM GENERATOR INTERNALS -

Ms. Stephan1e Coffin. NRR

. Ms. Coffin stated that the purpose of the proposed generic letter concerning degradation of steam generator. internals was to-e communicate the types of damage found in foreign steam generators.

. . - emphasize the importance of examining steam generator internals, and e request information concerning the conditions of steam generator internals.

Ms. Coffin presented information on the damage found in foreign steam

. generators, the significance of possible events that could result from damaged internels, and the foreign response to the damage. She explained the results of domestic steam generator inspections, the regulatory requirements for conducting inspections, and the information requested by the proposed generic letter. Ms. Coffin summarized the public comments on the proposed generic letter and the staff responses. She concluded that after the generic letter is issued, the industry is expected to provide coordinated industry responses though the vendor owners groups, the Nuclear Energy Institute, or the Electric Power Research Institute. The Subcommittee members and the staff discussed the types of damage identified, the root cause of the degradation, and what the staff will do with the industry responses to the proposed generic letter.

ULTRASONIC INSPECTION RELIABILITY AND PERFORMANCE DEMONSTRATION -

Dr.. Joseph Muscara. RES Dr..Muscara presented the research program associated with the evaluation and improvement of nondestructive examination (NDE) reliability for inservice inspections at nuclear power plants. The objective of the program, which

- began in 1977, was to determine the reliability of ultrasonic inspections, recomend code changes, and formulate improved inservice inspection criteria.

- The results of the research program indicated that the ASME code prescriptive ultrasonic testing procedures were not providing adequate flow detection-reliability. Consequently, during the 1980s the staff, its contractors, and industry developed ASME code.Section XI. Appendices VII and-VIII, which require teams to demonstrate acceptabl.e flaw detection capabilities before-

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, Minutesi April 15-16, 1997 3 Materials'& Metallurgy Subcommittee

' performing field inspections. Dr. Muscara explained the Appendix VIII performance demonstration requirements for detection tests and crack. sizing .

tests, and the acceptance criteria.for wrought austenitic stainless steel piping. Dr. Muscara's summary and conclusions included the following points.

• Even as improvements have been made to techniaues and procedures, there is a large variability in the 3erformance of NDE systems for similar procedures and equipment used )y different inspectors.

• There is a need to qualify NDE systems, through 3erformance demonstrations, to screen poor performers from tie qualified pool of NDE systems that are used for inservice inspections. ,

• The acceptability-of missing deep flaws in a passing detection test should be addressed.

• The inclusion of a criterion to the sizing qualification test for the maximum acceptable undersizing of deep flaws, in addition to the root-mean square error requirement, should be considered.

The Subcommittee members and the staff discussed the following items:

e code requirements for ultrasonic testing examination of piping, e development of software to improve inspections.

• differences between foreign and domestic-inspection techniques, e techniques used during round robin tests, and

• European performance demonstrations.

PROPOSED GENERIC LETTER REGARDING ULTRASONIC TESTING IN INSERVICE INSPECTION PROGRAMS - Mr. Robert Hermann, NRR Mr. Rouert Hermann, NRR, presented background information related to the technical and regulatory issues associated with the ultrasonic testing of pipes feedwater nozzles, and reactor vessels. He explained the performance demonstration methods required by ASME code,Section XI. Appendix VIII. Mr.

Hermann provided justification for issuing a generic letter requesting information on how licensees were using Ap)endix VIII in their inservice ins)ection programs. The Subcommittee mem)ers and the staff discussed the lacc of-a statistical basis for the performance demonstration acceptance criteria and the different contributors to the performance demonstration paso-fail rates.

PROPOSED GENERIC LETTER REGARDING STEAM GENERATOR TUBE INSPECTION TECHNIOUES -

Mr. Phillip Resh, NRR Mr. Rush explained that eddy current methods are the 3rimary NDE technique for identifying steam generator tube defects. He noted tlat the technique has large uncertainties and in most cases is not qualified to size defects. Mr.

Rush presented the regulatory requirements for controlling special inspection processes and for repairing identified defects. He provided examples of how

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. Minutes:.-A 15 16, 1997 .4 Materials &pril Metallurgy Subcommittee

- - some licensees have used ' questionable inspection practices and noted the lack-

. of information available to the staff concerning the industry steam generator

- tube inspection programs.

Mr. Rush stated that the purpose of the proposed generic letter concerning

- steam generator. tube inspection techniques was to:

e notify licensees of the importance of using qualified inspection

-techniques, e request information on sizing techniques and the basis for those.

techniques -and e verify compliance with 10 CFR 50 Appendices A and B, and the technical specifications.

Mr. Rush summarized the public comments on the proposed generic-letter and the staff responses. He stated that the generic letter would be issued in May 1997. -

The Subcommittee members and the staff discussed the following items:

e how the staff will use the industry responses -to the proposed generic letter, e characterizing and sizing of defects, e the importance of staff and industry interaction in developing risk-informed and performance based regulatory criteria, e the differences between degradation mechanisms found in the different i . steam generator designs, and e the use of pulled tube data to develop statistical bases for alternate repair criteria.

MOM) SED GENERIC LETTER REGARDING DEGRADATION OF REACTOR VESSELCLOSURE HEAD -

)ENETRATIONS - Mr. C. E. Carpenter. Jr., NRR' 2 .

Mr. Carpenter presented background information on cracks identified in foreign reactor vessel control rod drive mechanism and other vessel closure head penetrations, and on domestic industry inspection-experience. He noted that the nuclear steam supply system vendors have developed susceptibility models

- for reactor vessel penetrations, but have not submitted the models or the results of analyses to the staff. ' Mr.- Carpenter stated that the purpose of the proposed generic letter was to:

e- verify licensee compliance with regulatory requirements, e determine if an augmented inspection program should be imposed, and

.-- ' request information concerning the potential for resin intrusion events.

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-Mr. Carpenter summarized the public comments on the proposed ' generic-letter and-the staff responses. He concluded that' vessel head penetration cracking

, is not an immediate safety concern and that an integrated, long-term licensee program.- including periodic inspectione and monitoring, is necessary.

The Subcommittee members and the staff discussed the following items:

i e other possible vessel inspection techniques besides eddy-current inspections,-

• . metallurgical structure of the identified cracks.

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• the availability of the results of the susceptibility studies performed by owners groups, and

• whether all reactor vessel head penetrations should be inspected.

PALISADES: PRESSURE VESSEL FLUENCE REEVALUATION - Mr. Lambros Lois, NRR Mr. Lois presented background information on and the staff position regarding a revised reactor vessel fluence analysis submitted-by Consumers Energy for the Palisades reactor vessel. The Palisades reactor vessel is calculated to exceed the pressurized thermal shock screening limits prior to the expiration of its license. The licensee revised its fluence analysis based on a reevaluation of four existing surveillance capsules, recalculation of early cycle neutron sources, new geometric and temperature data, cavity dos'.,etry data, and a new statistical analysis method. The submittal requestec 25

)ercent reduction of the fluence accepted in the 1993 staff review of the

)alisades pressure vessel fluence analysis.

. The staff prepared a safety evaluation report (SER) on the licensee submittal.

The staff accepted an 8 percent reduction due to the revised physical measurements of the plant. The staff, however, concluded that the licensee's i requested 17

> unacceptable. percentage reduction baser.' on a reevaluated bias wasThe bias, wh fluence. is derived from calculations, dosimetry data, and spectral-adjustments. The staff evaluation was based on the following concerns:

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• no physical explanation for the requested deviations has been provided.

a plant-specific data are inconsistent. and

• - spectral least square-fitting does not represent a best-estimate value.

The current fluence value allows plant operation till 1999. The 8 percent reduction will allow plant operation through 2003. The Subcommittee members and the staff discussed the location-of vessel capsules and the uncertainties associated with the measurement-to-calculation ratio of fluence bias.

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Materials &pril-15 16, Subcommittee Metallurgy 1997

)RFT RIGJLATORY GUIDE DG 1053."CA.CULATIONAL AND DOSIMETRY METHODS FOR i

)ETERMI114G PRESSURE VESSEL NEUTR04 FLUENCf" .- Ms. Carolyn Fairbanks, NRR ,

Ms. Fairbanks explained that the purpose of the draft regulatory guide was to

_ provide an acceptable state-of the-art method for fluence determination and to standardize met 1odologies for determining vessel fluence. She presented the regulatory requirements for which the regulatory guide was developed, and summarized the schedule for issuance of the regulatory guide. Ms, Fairbanks noted that the present draft 'of the regulatory guide includes the application of Monte Carlo transport to pressure vessel fluence benchmark aroblems, The Subcomittee members and the staff discussed the relationship 3etween surveillance capsule data and reactor vessel fluence.

EALISADES REACTOR VESSEL INTEGRITY STATUS - Mr. Jack Hanson, Consumers Energy Mr. Hanson provided background on the licensee submittal regarding reactor vessel fluence. He presented the options available to Consumers Energy including vessel annealing, materials testing, regulatory guide 1.154 analysis, and early plant shutdown. He identified the location and number of in-vessel and ex vessel capsules used to generate the fluence data. He noted that Monte Carlo calculations were performed to provide an independent evaluation of vessel fluence.

Mr. Hanson explained the use of the least squares adjustment procedure for calculating a best estimate value for the measurement-to calculation ratio of fluence bias. He compared the individual ca)sule biases for Palisades and the best-estimate bias for Palisades to average aiases for other plants. Mr.

Hanson concluded that a default conservative value, rather than a best-estimate value, places an undue burden on Consumers Energy.

The Subcommittee members, representatives of- Consumers Energy, and the staff discussed the following items:

• types of material used in the dosimetry capsules.

• use of other basis besides fluence to extend plant operations. ,

• reasons for the 17 percent change in the calculated value for fluence.

• how plant-specific data can be taken into account without averaging.

• how damage to the reactor vessel is derived based on fluence, and

• whether the best-estimate value represents the true fluence.

NUREG 1411. SUPPLEMENT 1. " REACTOR PRESSURE VESSEL STATUS REPORT" -

Ms. Andrea Lee, NRR Ms. Lee explained that NUREG-1511. Supplement 1. " Reactor Pressure Vessel

-Status Report." incorporated information gathered from licensee responses to Generic letter (GL).92-01. Revision 1. Supplement 1, " Reactor Vessel

. Structural Integrity." and updated the NRC reactor vessel integrity database i (RVID). She noted that Babcock & Wilcox and Combustion Engineering have not yet responded to Supplement 1 of the generic letter-. ,

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- Ms. Lee presented the reasons for issuing the generic letter sup)lement.- '

sumarized the staff generic assessment of pressurized thermal slock based on the new data, and explained the features of the-RVID. She stated that some

- planned staff activities included reviewing new best-estimate chemistry data, t resolving discrepancies between the RVID and the industry database. and

! eventual approval of- a database maintained by the industry with oversight from

_the staff.

- Ms. Lee demonstrated the features of the RVID. She stated that the RVID would be made available to the industry and the public. Mr. Strosnider highlight d i the magnitude of the industry and staff effort expended in developing the RVID.

DOE REACTOR VESSEL ANNEALING PROJECT UPDATE - Ms. Deborah Jackson. RES

- Ms. Jackson presented the objectives and present status of the Department of Energy annealing demonstration project involving the Marble Hill and Midland reactor vessels. She showed and provided commentary on a video tape, which

included selected activities associated with the Marble Hill reactor vessel

' - annealing demonstration. Ms.-Jackson stated that the demonstration of annealing the Marble Hill vessel with an indirect gas fired heating method was successful. She explained that the annealing of the Midland vessel had been delayed. The Subcommittee members, the staff, and Mr. Hanson, who had observed "

n 4 the vessel annealing demonstration, discussed the associated stress analyses.

code cases, and procedures.

SUBCOMMITTEE COMMENTS Dr, Kress stated that the details contained in the presentations were at the appropriate' level. Dr. Seale expressed a desire to work in parallel with the staff in developing technical- and policy issues instead of waiting until the issues became intractable. Dr. Kress questioned whether the Comittee should review and.coment on the staff position related to the reevaluation of reactor vessel fluence. Dr. Seale mentioned that vessel fluence is an 7

interesting case since there are elements of risk in the regulatory decision.  !

FOLLOWUP ACTIONS 1

These presentations were provided as information briefings. No followup actions were identified, i

SUBCOMMITTEE RECOMMENDATIONS 2 The'Subcomittee' decided that imediate Committee followup activities were unwarranted.

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. Minutes: A 15 16. 1997 8 Materials &pril-Metallurgy Subcommittee BACKGROUND MATERIAL PROVIDED TO THE SUBCOMMITTEE:

1. Memorandum dated December 23, 1996, from Thomas Martin NRR. to David-Meyer. 0A.-

Subject:

Notice of Opportunity for Public Comment-for a Proposed Generic Communication Regarding Effectiveness of Ultrasonic testing Systems-in Inservice Inspection Programs.

2. CRGR presentation slides. "UT Reliability / Performance Demonstration."

dated November 19. 1996, t 3. Memorandum dated December 23. 1996, from Thomas Martin. NRR. to David Meyer,0A.

Subject:

Notice of Opportunity for Public Comment for a Proposed Generic Communication ~Regarding Degradation of Steam Generator Internal.

4. CRGR presentation slides, " Degradation of Steam Generator Internals."

dated November 19, 1996.

5. Memorandum dated December 23. 1996, from Thomas Martin, NRR. to David Meyer,OA.

Subject:

Notice of Opportunity for Public Comment for a Proposed Generic Communication Regarding Steam Generator Tube Ins)ection Techniques.

6. CRGR presentation slides. " Steam Generator Tube Inspection Techniques."

dated November 19, 1996.

7 SECY-97 063, " Proposed NRC Generic letter 97 ##, ' Degradation of Control Rod Drive Mechanism and Other Vessel Closure Head Penetrations'." dated February 20, 1997.

8. Memorandum undated. Lambros Lois, NRR. to Robert Schaff, NRR

Subject:

Safety Evaluation Report on Palisades: Pressure Vessel Fluence Reevaluation

9. U. 5. Nuclear Regulatory Commission NUREG-1511, Supplement 1,

" Reactor Pressure Vessel Status Report. October 1996 NOTE: Additional details of this meeting can be obtained from a transcript of this meeting available in the NRC Public Document Room. 2120 L Street.

N.W., Washington, D.C. 20006. (202) 634-3274. or can be purchased from Neal R. Gross and Company Incorporated, Court Reporters and Transcribers. 1323 Rhode Island Avenue, N.W., Washington, D.C. 20005.

(202) 234-4433.

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