ML20128C732

Summary of 921223 Meeting W/Epri Re on-going Work in EPRI Steam Generator Degradation Specific Management Program. List of Attendees Encl

ML20128C732
Person / Time
Issue date:	01/28/1993
From:	Strosnider J NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:	Richardson J NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
References
PROJECT-669A NUDOCS 9302040103
Download: ML20128C732 (88)

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-JAN 2 01393 MEM00ANDUM FOR:1 James E. Richardson,--Director

-Division of Engineering-Of fice-of Nuclear Reactor Regulation

=FRON:

. Jack.R. Strosnider,. Chief.

- Materials and Chemical __ Engineering Branch Division.of Engineering-

SUMMARY

OF DECEMBER 23,-1992, MEETING WITH THE ELECTRIC POWER RESEARCH INSTITUTE (EPRI) REGARDING STEAM GENERATOR..

-DEGRADATION SPECIFIC MANAGEMENT A meeting was held on' December 23,'1992 wit'h representatives from the Nuclear Regulatory Commission (NRC), the Electric Power Research Institute-(EPRI), the-

.public, and several utilities to' discuss on-going. work in the 'EPRI_ Steam.

Generator Degradation Specific Management--(SGDSM) Program,-

A list _of meeting' attendees is provided in Enclosure _1._ ~ A summary of_ the -

meeting including NRC staff comments are contained in Enclosure-2.

Enclosure >

3 is a copy 'of the handout _ materials presented during the meeting.j (l!!ginalSigne:i By Jack R. Strosnider, Chief Materials and Chemical Engineering Branch Division of Engineering:

Enclosures:

1.

List of Attendees-

-2.

Meeting Summary 3.-

Meeting Handout cc:

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ENCLOSURE I; ATTENDANCE-LIST j

December 23. 19921 NRC MEETING WITH THE F1ECTRIC POWER RESEARCH INSTITUTE 1

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STEAM GENERATOR DEGR&DATION SPECIFIC MANAGEMENT.

HAliE ORGANIZATION Jim Richardson NRC/NRR/DE B.-D. Liaw NRC/NRR/DE

-Jack Strosnider NRC/NRR/DE/EMCB Emmett Hurphy

- NRC/NRR/DE/EMCB' Herb Conrad NRC/NRR/DE/EMCB-Ken Karwoski

- NRC/NRR/DE/EMCB-Ron Gamble Sartrex Corp (EPRI Consultant)-

David-A. Steininger

- EPRI

-Kenneth R. Craig-Florida Power and Light' Tom Gerber Structural-Integrity. Associates :(EPRI:-

Consultant).

J. C. Blomgren Coramonwealth Edison-Larry Steinert EPRI consultant.

Steve Long NRC/NRR/DSSA/SPSB

.Ed Hackett NRC/RES/MEB-Steve Hoffman NRC/NRR//DRPE/PDII Greg-Kammerdeiner Duquesne-Light Company-Yoshi Horikawa Kansai: Electric.

S. H. Hanauer-Oregon Energy-Facilities: Siting Council.

Jim Mark Consolidated-Edison-Co.

John F. Smith Rochester' Gas and Electric Robert _ D. Pollard Union'of; Concerned Scientists -

Ken Eccleston

- NRC/NRR/DRSS/PRPB.

John Lee Virginia; Power; LJoe Eastwood

.-Virginia Power Lynn Connor

- Southern -Technical - Services-John Jensen AEPSC r

Robin" Jones EPRI Dan Mayes Duke Power Co.

Rick Mullins Southern Nuclear' Operating Co..

J..W. Cliffcrd

- NRC/NRR/DRPW-George Johnson NRC/NRR/DE/EMCB'.

Bob Smith Rochester Gas and Electric

-David Goetcheus

-TVA' Donald Naujock NRC/NRR/DE/EMCB

_ Bill Long

' NRC/NRR/DSSA/SCSB Karl Toth AEPSC U

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e-4 ENCLOSURE 2

SUMMARY

0F DECEMBER 23, 1992 MEETING CONCERNING STEAM GENERATOR DEGRADATION SPECIFIC MANAGEMENT PROGRAM Representatives of the Electr_ic Power Research Institute (EPRI),- several utilities, the public, and the NRC met on December 23,-1992, at NRC:

Headquarters in Rockville, Maryland to discuss on-going work within the EPRI Steam Generator Reliability Project (SGRP) on steam generator tube alternate repair criteria (ARC) and the Inservice Inspection Guidelines, EPRI and its consultants described the EPRI Steam Generator Degradation Specific Management (SGDSM) Program as a process for evaluating and monitoring and/or repairing distinct, individual tube degradation mechanisms, which includes the following key elements:

Adequate margin against tube rupture Improved inspection procedures and data analysis Reduced allowable leakage during operation Site boundary doses within 10 CFR 100 limits This process is intended to define inspection and repair criteria for distinct, individual degradati>n mechanisms.

The inspection and repair criteria would be determined ', sing generic guidelines in the industry developed report, " Steam Generator Degradation Specific Management." Utility implementation of the indivi1ual inspection and repair criteria would follow-requirements contained in t',e SGDSM report and in reports prepared for each distinct degradation mechanism.

Furthermore, implementation of a SGDSM Program would require utility modification of plant Technical Specifications and development of a plant specific SGDSM Program plan.

The following general areas were discussed _during the meeting:

1) SGDSM -

Structural and Material Design Basis - Safety Margin Concept; 2) SGDSM -

Inservice Inspection (ISI)/ Nondestructive Examination (NDE) Requirements, Qualifications, etc.; 3) SGDSM - Design Basis Accident Leakage and Radiological Aspects; 4) Utility SGDSM Program _ Implementation and Technical Specification Change; 5) SGDSM Program Application for'0DSCC at Steam Generator Tube Support Plates; 6) SGDSM Program Application for PWSCC at the steam Generator Tube Expansion Zone; and 7) Severe Accident Issues.

l At the conclusion of the meeting, the NRC made several commW on the industry proposal. The NRC noted that the proposed SGDSM '

r.3 had several-L positive aspects including: 1) the use of deterministic ang 3, with L

supporting probabilistic analyses, to demonstrate the technical merits for the-proposed ARC; 2) the use.of improved NDE technology and' techniques; and 3) the development of a standardized protocol for SG NDE technique qualification for.

both equipment, procedures and personnel (i.e., performance demonstration).

The NRC noted that a large number of plant specific submittals regarding-alternate steam generator tube plugging criteria would be difficult to review l

with existing resources and said~that submission of the SGDSM Program as a generic industry approach would facilitate review of steam generator alternate l

plugging criteria proposals.

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I At the conclusio_n -of-the meeting,- the-SGRP commitied on_ a Ltentative: schedule:

to 1)-identify a: candidate; for? a :" lead plant"? in; February 1993, 2)-provide ;.

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another briefing-:on the status of the SGDSM Program to the WC staff-Lin March L

-8 1993J:3) submit-the SGDSM Program packagetin June 1993, and:4)Lsubmit a letter c

by June 1993' documenting the. technical differences.between:the: planned EPRI.

l SGRP SGDSM submittal--and. the submittals' currentlyLproposed by; individual?

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I ENCLOSURE-3 STEAM GENERATOR DEGRADATION SPECIFIC MANAGEMENT (an Industry Initiative)

MEETING PASSOUT 1

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Steam Generator Degradation Specific Management j

(an Industry Initiative) i David A. Steininger Technical Advisor Steam Generator Reliability Project Office Electric Power Research Institute (EPRI)

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b WHAT IS THE STEAM GENERATOR RELIABILITY PROJECT (SGRP)?L SGRP is an EPRI (Utility) funded program directed at solving Generic problems?related to steam; generator operation.

Managed by EPRI personnel.with oversight.provided byTutility management and technical personnel operating through.three--

standing committees.

Active participation by many foreign utilities" operating L'PWR nuclear' : plants.-

Ad-hoc committees within the SGRP focuses effortsion.high,

priorityLsteam generator issues"such 'as steam. generartor ' tube alternate repair; criteria.(ARC) and the Inservice Inspections

-(ISI) Guidelines.

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L-J WHAT IS THE PERSONNEL MAKE-UP OF THE.STEAMLGENERATOR-TUBE ALTERNATE REPAIR CRITERIALCOMMITTEE AND WHEN WASL 1

IT FORMED?

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Representatives from steam generator vendors, domestic and Foreign utilities and consultants to EPRl; formed in 1989.

a D. A.-.Steininger, Chairman J. Begley, D. Malinowski S. Brown,u C.' L Williams J.. Houtman, T. Pitterlo J. Hutin Electric Power ' Research Westinghouse Electric Corp.

Electricite de. France '

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' institute G. "Alrey J. Smith G.". Kammerdeiner

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F Nuclear Electric Rochester Gas & Electric Duquesne Light. Company '

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F. -- Anderson C. Smoker

. G. Kent Northeast Utilities

. Wisconsin - Public ' Service Portland: General Electric J. Blomgren '

J. Eastwood

.D. Mayes-Commonwealth Edison Virginia Power Company Duke Power l

K.T Craig P. Anderson.

T. Gerber-Florida-Power & Ught ABD-Combustion Engineering Structoral Integrity Assoc. Inc.:

'M. Danek

..J. Engstrom H.' Housorman,iR.i.Vollmer.

Public' Service Electric & Gas _

Swedish State Power Board Zetec, Inc.,

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~ J. : Miller.

R. Pearson

'J. Frick Southern - Nuclear Operating Co.

Northern States Power Co.

South Carolina Electric l& - Gas i J. Snider -

P. ! Hernalstaen B. Curtis :

3 Babcock & Wilcox

. Laborelec.

Allen Nuclear" Associates:

G. Bollini J. : Jensen R. Gamble,

Spanish ' Utilities -

'American Electric Power Sartrex Corp.-

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WHO ARE THE MEMBERS OF THE STEAM GENERATOR IN-SERVICE INSPECTION GUIDELINE COMMITTEE?

ii RepresentativesLfrom steam generator vendors, domestic utilities and-consultants to EPRl; formed in 1982.

Jim Benson Northeast Utilities Randy Lewis Wolf Creek'Nuc.

Oper.

Gustavo Bollini Tecnatom Dan Malinowski Westinghouse-

. Gary Boyers Florida. Power & Light Dick Marlow Conam Nuclear.-

Paul Brown Yankee Atomic Rick : Maurer ABB-CE -

Steve Brown EPRl' Consultant Dan Mayes Duke Power:

7 Blaine Curtis Allen Nuclear Scott Redner Northern States l1 Associates Power Dan Dobbeni Laborelec Todd Richards B&W Nuclear:

r Services Robert Dolansky New-York Power Auth.

Mike Sears Commonwealth '

Edison Doug Harris Anacapa Sciences Tom Smith Southern Co.

Servicesi

. Gary Henry EPRI NDE Center Mark Torborg _GPU - Nuclear-Guy - Holmes. Wisconsin Pub.,

Bob Vollmer Zetec'inc.

4 Serv. Corp.

Howard Zetec inc.

Kevin Wachter-~ Rochester Gas &-

L Houserman Elec.

Ron Ingraham Westinghouse

.Clayton Webber. TVA' Jeffrey Lendrum Siemens' Nuclear Pwr.

Mohamad Chairman, ;EPRI' Behravesh i

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WHAT RESULTS HAVE BEEN PRODUCED BY THESE AD-HOC COMMITTEES?

3 PWR steam generator tube : repair limits:

technical support

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document fort expansion zone' PWSCC in roll' transitions (Revision' 11, EPRI ' Report, EPRI NP-6864-L; Revision 1,.. December ~1991.

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-Referenced in recent Northeast Utilities. Connecticut Yankee's '

submittal for ARC.

j PWR steam generator-tube repair limits:

Technical support document for outside diameter stress corrosion cracking 'at tube -

support plates, EPRI final-report, EPRI TR-100407, ? March 1992.

Referenced in Southern Nuclear Company Farley's' plant-submittal for L ARC.

l PWR steam L nenerator examination-euidelines:

- Revision 2,. EPRI Finalf Report,!EPRI NP-6201, December-1988,. AppendicesL G L& HT and-Su'pplement sI 'added in 1992-L I

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i WHAT INITIATED THE DEVELOPMENT AND IMPLEMENTATION ~

OF STEAM GENERATOR DEGRADATION SPECIFIC MANAGEMENT (SGDSM)?-

Past SGRP/NRC meetings directed at specific ARC and LISI details.

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Individual utilities.' submitting site specific ARC to the NRC..with additional utilities considering similar proposals.

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SGRP/NRC ' Annual Review Meeting in March 1992 discussed-H possible generic-licensing implementation and applicable technical aspects of PWSCC and ODSCC ARC.

SGRP offere'd to initiate a generic effort on SGDSM.

NRC -requested a full day meeting to discuss the implementation of:-

i the1possible : generic' application: of ARC in the licensing; arena 4aridz certain technical Ldetails. associated' with specific ARC such as 1

PWSCC andEODSCC.

SGRP immediately initiated work on Lthe1 generic application!offARC 3

4 NRC*12/22/92*DAS '6

FROM MARCH OF 1992 TO NOW WHAT DEVELOPMENTS OCCURRED-WHICH SUPPORTED THE CONTINUED HIGH. PRIORITY DEVELOPMENT OF' A GENERIC APPLICATION OF ARC?

L NRC. Regulatory Information Conference - July '21-22,1992 in Jwhich the presentation titled " Staff Initiatives in Steam Generator Integrity" indicated "The staff is considering whether expansion of the flaw-type specific:

correctiveL measure route 'is a'more appropriate treatment for steamLgenerator Lintegrity. concernsi in the staff's view, a matrix of flaw-specific measures could-1 be developed that would consider flaw type, size, orientation an'd location.

L These flaw. parameters;would be analyzed to determine the appropriate v

corrective' measure strategy.

That: strategy wouldlintegrate.a specific repair (plugging orf sleeving).. criterion with a leakage rate limit and wouldispecifyf augmented inspection. requirements.

The! tube repair criterionLfrom the matrix would then be adjusted to account for flaw? growth rate ~ and NDE uncertainty!to-obtain the appropriate: field tube-repair limit."'

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Flaw Mechanism Matrix

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Tubesheet Led Rate:1gpen Inspection:3%

Circ:No flaws allowed Axial: Repair limit:9V Tube Support Leakrate simit: 200gpd.

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Intersections with MRPC U-Bends In Field Limit =

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NRC+12/22/92*DAS-1 8

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3,; a WHAT IS THE UTILITY INDUSTRY OFFERING SIMILAR TO

-THE PROPOSED NRC FLAW MECHANISM" MATRIX?

Steam Generator Degradation Specific Management (SGDSM).

-A report that.provides the generic requirements for tube inspection, repair criteria development and implementation which can be used to produce the most cost-effective means to maintain acceptable plant safety and steam generator-reliability.

1 SGDSM.is a process for evaluating, and monitoring and/or repairing distinct, individual tube degradation mechanisms, and includes the following

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key elements-j a

. Adequate margm agamst tube rupture l

Improved inspection ' procedures and data analysis:

Reduced allowable leaka'ge during operation LMaintain site lboundaryfdose within 10 CFR 100 limits

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,.. y SPECIFICALLY, WHAT DOES THE STEAM Gi!N'dRATOR DEGRADATION SPECIHC MANAGEMENT DOCUMENT CONTAIN7 2

Generic Methodology for :

-y Developing Degradation Specific Inspection and '

. Repair Criteria

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Degradation Specific Steam Generator Inspection and Re Tube Inspections

' Criteria Reports pair 3r.

Plant Degradation Specific Management Program. Plan

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Inspection,P_r.ocedures

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a HOW ARE WE GOING TO PROCEED IN THIS MEETING AND WHAT ARE THE ISSUES INVOLVED?

8:30-9:00 AM Introduction and background (David A.Steininger, EPRI) 9:00 - 9:45 Steam Generator Degradation Specific Management (SGDSM) Program, Concept and Objective (Ron Gamble, Sartrex Corp.)

9:45-10:30 SGDSM - Structural and Material Design Basis - Safety Margin Concept (Ron Gamble, Sartrex Corp.)

10:30-10:45 Break 10:45-11:30 SGDSM - Inservice Inspection (ISI)/ Nondestructive Examination (NDE) Requirements, Qualification, etc. (David A. Steininger, EPl!I) 11:30-12:00 SGDSM - Design Basis Accident Leakage and Radiological Aspects (Tom Gerber, Structural Integrity Associates) 12:00-1:00 Lundt 1:00-2:00 Utility SGDSM Program Implementation and Technical Specification Change (Larry Steinert, Consultant) 2:00-2:30 SGDSM Program Application for ODSCC at Steam GeneratorTube Support Plates (Tom Gerber, Structural Integrity Assoc.)

2:30-3:00 SGDSM Program Application for PWSCC at the Steam Generator Tube Expansion Zone (Ron Gamble, Sartrex Corp.)

3:00-3:15 Break 3:15-3:45 Severe Accident Issues (T.Gerber, StructuralIntegrity Assoc.)

3:45-4:00 NRC Plans for review and resolution of utility submittals of Alternate Tut e Repair Criteria (W. Russell, NRC) 4:00-4:15 NRC comments on industry initiative - Steam Generator Degradation Specific Management (SGDSM) and expected NRC/ Industry action (W. Russell, NRC)

NRC+12/22/92+DAs-1 11

WHAT DO WE WANT TO ACCOMPLISH BY THE END OF THIS M Initiate a dialog on the industry supported proposal for SGDSM.

the scope of oteam generator Develop a common understanding t degradation specific management.

Resolution on acceptable procedures for the implementation of SGDSM.

Identification of procedures to obtain NRC review and approval of this generic approach of steam generator degradation specific management.

-- Dates required, how should they be met and by who Schedule Documentation submitted to NRC Organization submitting documents 12 NRC+12/22/92-DAS4 Nl

L WHAT DOCUMENTS REQUIRE NRC REVIEW FOR SGDSM IMPLEMENTATION?

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Steam Generator Degradation Specific Management.

L PWR steam generator tube repair limits:

technical support document for expansion zone PWSCC in roll transitionsI(Revision l

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1),.EPRI Report, EPRI NP-6864-L, Revision 1, December 1991.

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Referenced in recent Northeast Utilities Connecticut Yankee's i

submittal' for' ARC.

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PWR - steam -generator tube repair limits:

Technical support document for outside -diameter stress corrosien cracking at tube-I support plates, EPRI final report, EPRI TR-100407, March 1992.-

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submittal for ARC.

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PWR Steam Generator Examination Guidelines:

Revision 2, EPRI-

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i Final -Report, EPRI NP-6201, December 1988, including Appendices L

(1991-1992).

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HAPPENS DURING THE INTERIM PERIOD OF NRC REVIEW. AND APPROVAL OF APPROPRIATE DOCUMENTATION?

Present plant submittals?:

Future plant submittals.w/o generic approach in-place?

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IF NOT WORKABLE:IN PRESENT FORM What is the possible alternative?.

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b Steam Generator Degradation Specific Management (SGDSM)

Program Concept and Objective l

NRC Meeting December 22, 1992 Ron Gamble RG-1 NRC-12/22/92

. a Steam Generator Degradation Specific Management i

Program Concept and Objective i

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Provide an overview of Steam

. Generator Degradation Specific Management (SGDSM) as contained in draft report " Steam Generator Degradation Specific Management," December 1992 e.

Purpose of report is to document the Ad-Hoc Alternate Repair Criteria-Committee's recommended: approach for:

Development of -inspection and repair criteria for distinct

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degradation mechanisms.

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Plant specific implementation of SGDSM-L Licensing implementation of SGDSM.

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What is SGDSM7 What-is the Goal of SGDSM7 I

What are the Objectives of SGDSM7 a

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Alternative Strategies l

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WHAT IS SGDSM?

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SGDSM is a process for evaluating, and monitoring and/or repairing distinct, individual tube degradation mechanisms, and includes the following key aspects.

adequate margin against tube rupture ISI guidelines implementation \\SGDSM ISI procedures reduced allowable leakage during normal operation mamtain site boundary dose within 10CFR100 limits 1

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l-This process: definds inspection and repair criteria for distinct, L

individual degradation mechanisms The. inspection and repair criteria are determined using generic-guidelines.

in' the -industry? developed

report,

" Steam Generator Degradation SpeciGc Management" e

Utility implementation of the individual; inspection and repair criteria follows requirements; contained in the SGDSM report and-in reports prepared;for each distinct: degradation mechanism.

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WHAT IS THE GOAL OF SGDSM?

Provide utilities with an industry developed and NRC approved steam e

generator tube degradation' specific management process that provides-an unambiguous, industry consistent method for inspection and repair.

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L Provide the necessary and sufficient process to justify including j-SGDSM as an : option in the technical specifications.

. Define the required content for industry prepared and NRC approved.

- a degradation specific inspection and repair criteria.

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.Specify the content of an acceptable plant specific defect management

. program.

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WHAT IS ACCOMPLISHED BY IMPLEMENTING SGDSM7 Reduce unnecessary tube repair.

Reduce personnel radiation exposure.

Minimize further reductions in thermal margins due to unnecessary tube repair.

Achieve increased steam generator reliability and availability.

Provide.a well documented industry-wide practice for determining steam generator operability.

I Reduce NRC review time.

l Reduce plant outage time.

Incorporate technical improvements and their understanding in a timely manner on an industry-wide basis.

Maintain flexibility for long term repair options.

RG-1 *NRC-12/22/92 7

ALTERNATIVE STRATEGIES I.

Current Depth Based Inspection and Repair Inspectionand Repair Requirements CurrentTecimical Specifications Current Depth Based Inspection and Repair widt Augmented Inspection Requirements II.

Inspection Requirements Industry Tube Inspection Guidelines Repair Requirements t

CurrentTechnical Specifications III. Degradation Specific Management Inspection Requirements Industry Tube Inspection Guidelines Degradation Specific Inspection Criteria RepairRequirements Degradation Specific Repair Criteria Current Technical Specifications (for non-degradation specific damage)

RG-1-NRC 12/22/92 8

O SGDSM ELEMENTS Tube Rupture Margin A data base to correlate rupture strength with the NDE parameter.

l Repair criteria to ensure adequate margins against tube rupture j

consistent with regulatory requirements.

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Inspection A NDE measured parameter (e.g., crack depth, crack length, eddy current voltage) to reliably asses structural and leakage integrity.

Implement induktry tube inspection guidelines.

Detc~rmination of the NDE measurement uncertainty.

Lsakage Limits A' data base to correlate the Icak rate with the NDE parameter.

-Repair l criteria to ensure off-site doses less than 10CFR100 limits during postulated faulted loads.

The allowable leak. rate to maintain doses within 10CFR100 limits during postulated faulted loads..

-Up-to-date Degradation Growth Rate Data Base.

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/ Degradation Specific. Inspection and Repair Criteria: Basis-Reports c

i PLANT SPECIFIC. IMPLEMENTATION

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l l Implementation' of a degradation. specific. management program on a plant specific basis requires the following utility actions:

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Modification of plant-Technical Specifications to incorporate t

degradation specific management.

Implementation of an allowable leak rate limit of 150 gpd per steam f

c gen'erator during normal plant operation.

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Development.sf a plant specific SGDSM program plan, which includes i

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Plant specific tubec. inspection in conformance with. industry i

approved : practice.

Confirmation of-the degradation specific mechanism.

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i Enhanced inspection 'in the affected region by the ' inspection techniques identified in the degradation specific: report.

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PLANT. SPECIFIC IMPLEMENTATION Plant specific degradation growth rate.

Generic growth rates used until sufficient plant specific data have been obtained.

1 A tube structure liinit in terius of an NDE measured parameter that corresponds to an acceptable snargin against tube rupture.

The leak rate predicted during postulated faulted loads froin tubes remaining in service.

l RG-I *NMC+12/22&2 m

I O-I E

e STRUCTURAL AND MATERIAL DESIGN BASIS NRC Meeting December 22,1992 Ron Gamble 1

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STRUCTURAL AND-MATERIAL 4

DESIGN BASIS t '

l Tube rupture correlation Structural limit

. Repair limit

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TUBE RUPTURE CORRELATION Establish a data base to determine the correlation between degraded tube..

rupture strength and the degradation specific NDE measurement parameter (e.g., flaw depth, flaw length, voltage).

Rupture' data obtained from laboratory and/or service degraded tubes.

Normalize correlation with respect to tube dimensions and strength properties for subsequent plant specific application of correlation.

e Define restrictions if any on application of correlation (e.g., flaw orientation).

i RG-2 NRC 12/22/92' 2

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STRUCTURAL LIMIT i

-L Determine the structural limit from the rupture correlation using The maximum pressure differential as the larger of 3 time ormal l

operation (RG 1.121 margin) or1.4 times postulated faulted loads e

.(ASME Code equivalent margin) t Tube material strength properties

. Tube dimensions g

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j REPAIR LIMIT

. Establish the repair limit by adjusting the structural limit for e

Degradation Specific NDE measurement uncertainty Degradation Specific' growth I

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l SGDSM INSERVICE INSPECTION L

I:ISI)/ NONDESTRUCTIVE EXAMINATION i:N D E)

REQUIREMENTS, QUALIFICATION i

I David A. Steininger NRC Meeting December 22,1992 i

l l

I NRC+12/22/92 DAS-2 1

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L SGDSM INSERVICE INSPECTION L

(ISI)/ NONDESTRUCTIVE EXAMINATION (NDE)

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NRC Regulatory information conference - July 21 and 22,1992 in which the presentation titled " Staff Initiatives in Steam Generator l

Integrity" indicated i

"The key to staff acceptance of this methodology (ARC) is achieving i

L high confidence in field inservice inspection (ISI) activities. To l

achieve this high confidence, ISI performance needs to be i

l demonstrated, in a quantifiable way, in the areas of flaw detection, sizing, characterization and flaw-growth monitoring. The demonstration should encompass the analysts teams, their

}

equipment, and the man-machines Interface."

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I NRC 12/22/92*DAS-2.

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DOES THE UTILITY INDUSTRY AGREE WITH THE STAFF POSITION?

Yes, in fact as a prerequisite to adoption of the SGDSM program option requires the submitting utility to implement plant specific tube inspection procedures in conformance with the utility industry tube inspection guidelines:

PWR Steam Gene'rator Examination Guidelines: Revision 2, EPRI Final Report, EPRI NP-6201, December 1988, including updates:

l Appendix G, " Qualification of Nondestructive Examination Personnel for Analysis of Eddy Current Data," September 21,1992; l

Appendix H, " Performance Demonstration for Eddy Current Examination," September 21,1992; and Supplement I, " Guidelines l

for Disposition of Bobbin CoilIndications Attributed to ODSCC at Non-Dented and Drilled Tube Support Plates," September 21, 1991.

I NRC*12/22/92-D AS-2 3

e 1.

t 4

Does Steam Generator Degradation Specific Management Specify Additional ISI Requirements?.

L i

l Yes v

In addi; ion to generic tube inspection procedures as represented by i

the IStguidelines, degradation specific Inspection procedures have been developed following degradation specific inspection n

j'

. requirements specific to the applicable ARC.

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i NilC+12/22/92*DAS-2 '

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What is The Historical Background To The Industry L

Developed: Tube Inspection Guidelines?

I i

In late 1970s and early 1980s the utility industry recognized that there L

existed deficiencies in steam generator ISI.

i In response to the above the SGRP developed and informally issued

}

ISI' guidelines to provide recommendations on steam generators ISI l

that supplemented Technical Specification requirements. The i

guidelines were initially prepared in 1981 and were updated in 1984.

j e

L In the 1984-1985 time frame, the NRC conducted round robin NDE l

examinations on the retired Surry #2 steam generator. The NDE l

results.were validated by destructive examination.

l l-The results indicated that there was wide variability in the way data l

analysts Interpreted the same data which led to large data scatter l

In the sizing and probability of detection modes.

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What Do The EPRI ISI Guidelines Contain?

Recommendations to assist utilities in establishing an " inspection l

i l

engineering" organization and in implementing a formal steam j

l generator examination program, l},

Recommended practices for steam generator examination, including l

Inspection sampilng rules, type of NDE equipment and data analysis i

procedures, i

L

- A summary of steam generator operating experience which documents L

damage' mechanisms, vulnerable locations within the steam generator j

[

and OE experience with various damage mechanisms, j

t Ind -

y-wide data on tube-leak outages in which specific causes are Idt

- iled.

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Performance Demonstration Requirements I

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NRC+12/22/92 DAS-2 7

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Appendix G & H of ISI Guidelines l

Personnel & Technology Qualiiication l

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. Establish minimum requirements for an industry-wide data ana yst i

training and qualification program.

Develop' a standardized protocol for S/G NDE examination technique qualification - e.g., equipment and procedures.

P 4.

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NRC 12/22/82 DAS-2 e..

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14

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Personnel Qualification - Appendix G Framework established by Appendix G to ISI Guidelines:

" Qualification of Non destructive Examination Personnel for Analysis of Eddy Current Data."

Establishes minimum requirements; provides a formal Industry-wide structure for data analyst training / qualification.

Document Contents 1.0 Scope 4.0 Qualification Requirements 2.0 Qualification Level 5.0 Qualification Records 3.0 Written Practice 6.0 Termination 10 N R C =12/22/92*D AS-2

~

m

Technique Qualification - Appendix H Framework established by Appendix G to ISI Guidelines;

" Performance Demonstration for Eddy Current Examination Techniques" Establishes minimum requirements; provides a formal protocol for eddy current data acquisition and analysis qualification.

Document contents 1.0 Scope 3.0 Performance Demonstration 2.0 Examination System 4.0 Essentiai Variable Tolerances Requirement 5.0 Record of Qualification Supplement 1 - Equiament Characterization Supplement 2 - QualLfication Requirements for S/G Tubing 11 NRC 12/22/92*DAS-2

e a

i i ndustryoData Base i

Centralized Data Base e

. Purpose is to provide uniform training and examination source-t material for employer. Implemented data analyst. qualification using protocols established in Appendix G.-

IDB. initialized using industry. developed source material with

~

continued updates - e.g., IDB maintenance, on a yearly basis User Responsibilities

^.

Implementation (training &' qualification)-

Certification Record keeping.(non-centralized)-

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- NRC+12/22/92 DAS-2 -

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What are the Degradation Specific Inspection Procedures?

Documented in Individual ARC reports for the specific damage mechanism - addressed For example, PWSCC ARC at the tubesheet/ tube expansion zone requires 100% inspection of tubes by RPC at every outage.

13 NRC+12/22/02 DAS-2

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i OVERVIEW g

3 With SGDSM, some small through-wall defects can remain in service following. Inservice inspection.

SGDSM altornate repair criteria (ARC) ensure that leakage-during design basis accidents results in off-site doses 'lesu than: 10CFR100 limits.

This irequirement.is metLby demonstrating that conservatively oredicted accident leak rate (Qp) from degraded l tubes that.

remain in serviceLis less than the plant ' allowable accident leak rate c(Qa).

Op.< Qa

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'l PREDICTED ACCIDENT LEAK RATE

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. Predicted accident leak rate (Qp) is the conservatively calculated primary to second'ary. leak ~ rate from all degraded-tubes that remain in. service under ' postulated accident

~

conditions.

Op is based on. correlations between NDE measured o

L degradation and degraded ~ tube leakage test results.

L L

3 TLG 3 NRC+12/22/92

4 i

PREDICTED ACCIDENT LEAK RATE 7

(cont.)

i j

Tube leakage tests are conducted with:

l.

Tubes with degradation representative of inservice degradation (including. tubes removed from service).

l

.i Loading conditions. representative of the: most limiting

[

de. sign ' basis accident condition (e.g., MSLB).-

~ TLG-3 NRC+12/22/92 !

i PREDICTED ACCIDENT LEAK RATE (cont.)

Allicorrelations must utilize an upper bound fit to the applicable. tube 'leakagefdata.

l In utilizing this fit, the NDE measured degradation is adjusted.

to' account for::

NDE measurement uncertainty Degradation growth j

i l

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4 i

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'TLG-3 NaC 12/22/92l.

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ALLOWABLE ACCIDENT LEAK RATE i

1 i

~

The allowable accident leak rate (Qa) is the plant specific leak.

rate that can be tolerated under postulated accident loading-conditions without exceeding 10CFR100 off-site dose limits.

i l

For low probability design basis accident conditions, standard l

l review plan methods-and assumptions for calculating radiological consequences are judged to.be ' overly conservative (Ref.. NUREG 0844).

3 ConsistentLwith NUREG. 0844,.'the SGDSM. report : recommends 3

that the allowable accident leak rate. calculation be ba' sed on 1

I best estimate ' assumptions for initial iodine concentration,.

i iodineispiking and meteorology.

TLG-3 NRC+12/22/92 -

b

il APPLICATION The plant specific allowable accident leak rate is established as part of the degradation specific management plan.

After each steam generator tube inservice inspe tion, tubes with NDE measured degradation which exceeds the alternate repair criteria will be removed from service.

The predicted accident leak rate (Op) at the end of the next operating interval is conservatively calculated for those tubes with acceptable NDE indications which remain in service.

7 TLG-3*NRC 12/22/92

r APPLICATION (cont.)

If. the predicted acdident leak rate' exceeds. the plant allow.able accident leak rate, Op < Qa,. additional tubes are repaired until' the calculation demonstrates'.that Op <.Qa in this manner ARC explicitly limits. tube leakage' under design basis accident loading conditions.

1 i

l

'TLG-3 NRC+12/22/92-'

'8'

- =

^

f l

UTILITY SGDSM PROGRAM IMPLEMENTATION 1

AND TECHNICAL SPECIFICATIONS CHANGE-h

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Larry.Steinert t

December 22,1992 l

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NRCol2/22/92*LDS-1 1

4 F

l UTILITY SGDSM PROGRAM IMPLEMENTATION AND TECHNICAL SPECIFICATIONS CHANGE EPRI TECH SPEC CHANGES SGDSM REPORT v

INDUSTRY PLANT SPECIFC DEGRADATION SPECIFIC INSPECTION TUBE PROGRAM INSPECTION PLAN AND REPAIR N

GUIDELINES CRITERIA REPORTS v

PLANT IMPLEMENTATION ACTIVITIES Calculate repair Criterio Develop inspection procedures Report results S1 A

"s_

l p

^

^

CHANGES TO l

TECHNICAL SPECIFICATIONS FOR SGDSM i

APPROACH:

i Define Generic Tech Spec Changes for SGDSM.

Utilize Exisiting Technical Specifications to Maximum Extent Practical Minimize: Additions, Revisions, Deletions Reference EPRI-SGDSM Report 1

i i

1.

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NRC 12/22'/92 ELDS-1 31 i

3

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CHANGES TO TECHNICAL SPECIFICATIONS FOR SGDSM 4

Retain current Tech Spec section 3/4.4.5.2, " Steam Generator Tube -

i Sample Selection and inspection" as is but add a new paragraph:-

i "When specific types of steam generator tube wall degradation have been diagnosed and verified, a degradation specific management program plan may be implemented.

j Degradation specific management includes limits of applicability, definition of affected '

tube segments, inspection methods and scope, inspection results classification, tube j

repair criteria, and augmented leak rate calculations.' For those tube segments where degradation specific management has been applied, the requirements of the EPRI:

i document " Steam Generator Degradation Specific Afanagement" shall be used

• Revise operational primary-to-secondary leakage through any one i

steam generator to.150 gallons per~ day j

)

r NRCe12/22/92*LDS-1 4

J y

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~

PLANT SPECIFIC PROGRAM PLAN ELEMENTS 1

1. Commitment to industry tube inspection guidelines

2. Degradation specific inspection and repair reports to be.

implemented

3. Commitment to develop plant specific inspection procedures in.

conformance with items 1 and 2

4. Plant specific structural limit in accordance with item 2

5. Allowable-accident leak rate limit

6. Post-inspection reports to be provided to NRC.

]

1 l

6

' NRCol2/22/92 ELDS-l 5

r PLANTIMPI FMENTATION ACTIVITIES Perform inspections per SGDSM

- -Calculate NDE growth rate and repair limit Repair. tubes if NDE measured level of degradation exceeds repair limit Calculate predicted accident leak rate If predicted accident leak rate for all degraded tubes left in'. service :

is not less than the allowable accident leak rate, additional tubes -

must be repaired Plant inspection reports to NRC:

-Within 15 days following completion of each inspection:-

1 Repair limit l

Number of tubes repaired l

Within 12 months following completion of each inspection:

l Repair limit i

Examination results l

--- Repaired tubes?

Tubes-with indicated degradation left in service.

J 1

Predicted leakage under accident c'onditions NRCe12/22/92*LDS 6

' :/

4 5

l PWR' STEAM GENERATOR TUBE REPAIR LIMITS:

TECHNICAL SUPPORT. DOCUMENT FORLOUTSIDE DIAMETER STRESS CORROSION CRACKING AT TUBE. SUPPORT PLATES EPRI TR-100407, March 1992 1

T. L. Gerber NRC Meeting

- December 22,1992-

. i 1

L TLC-2*NRC*12/22/92 ~

1; 1

OVERVIEW OFDEGRADATION SPECIFIC MANAGEMENT FOR-

.OUTSIDE DIAMETER' STRESS CORROSION CRACKING AT TUBE SUPPORTPLATES-(ODSCC @ TSPs)'

o i-1005 bobbin coil inspection of affected TSPs intersections to identify and-measure ODSCC.

4 R-Conservatively established bobbin coil voltage based tube repair criterion -

to assure tube integrity.

Operational leak rate limit of 150gpd to provide added assurance that1

[

margins against tube rupture and leakage are maintained.

i Limit number of' tubes with ODSCC indications to assure acceptable-leakage under accident loading conditions.

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Illustration of ODSCC in Steam Generator Tub ~es in the TSP Region.

1.;

6 TUBE REPAIR CRITERION IS BASED ON:

Tube burst experiments correlated with bobbin coil voltage amplitude.

Application of Reg. Guide l.121 safety margins Correction for NDE measurement error and degradation growth between inspections.-

i J

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j TUBE REPAIR CRITERION IN TERMS OF

. BOBBIN COIL VOLTAGE Vat = VSL - VNDE - VCG I

Where:

Voltage Repair Limit val

=

' Voltage Structural Limit-VSL

=

Voltage Correction for NDE Measurement Error VNDE

=

I Voltage Correction for Degradation Growth VCG

=

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TUBE RUPTURE TESTS L

Tubes ~ pulled from service and tubes with ODSCC produced in the?

laboratory.

Tubes l pressurized to failure.without presence'of a simulated TSP a

Tube burst behavior correlated with bobbin coil voltage.

LLower 95% bounding interval curve developed.

Bosn'dingLcurve corrected for temperature and material properties.

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MARGINS APPLIED TO TUBE LOADS TO DEFINE STRUCTURAL LIMIT 3 for normal service conditions (Reg. Guide 1.121) 1.4 for postulated accident conditions (consistent with ASME Section III)

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. CORRECTION FOR NDE MEASUREMENT ERROR Contributions to measurement error include:

Measurement repeatability based on experiments.

Calibration standard dimensional variations based on calculations.

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n-a CORRECTION FOR DEGRADATIONLGROWTH

--q Degradation growth is based on plant operating experience with ODSCC @ -

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i TSPs.

change in bobbin coil voltage between inspections Utilize industry or plant specific data (if available)

Correction!for degradation growth is.the expected increase in bobbin coil voltage at end o'f operating cycle.

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' LEAKAGE UNDER ACCIDENT CONDITIONSL Tube leakage based on-tube leakage tests.

Calculate leak rate under postulated accident loads (include NDE measurement error and degradation growth).

Show-that leak' age is less.than plant allowable accident leak rate *.

If calculated leak rate is high, limit the number of tubes in service with' ODSCC indications.-

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• Rate which assures off-site dose is less than 10CFR100 Limits s.w.

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i TUBE LEAKAGE TESTS 4

'l Tubes pulled from service and tubes with ODSCC produced in laboratory.

a

. Leakage measured at normal operating and accident differential pressures.

ithout the' presence of TSP.

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-w Results correlated with bobbin coil voltage.

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.t CORRELATION OF TUBE LFAKAGE DATA 1

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An approach being evaluated for predicting the accident leak l

rate is to base Qp on -correlations between:

i j;

Tube. leak rate.vs. NDE measured degradation for tube 1

l which leaked during the tube leakage tests (i.e., non-

[

leakers will not be included in the correlation)

[

r L

Probability of leakege vs. NDE measured. degradation for all j;

tubes tested in the tube leakage tests.

l With this approach, the predicted accident leak rate is the

[

sum of the product of; probability of leakage times the j

estimated leskage for each tube with identified degradation.

l i.'

TLG 3 NRC+12/22/92 -

j

PWR Steam Generator Tube Repair Limits:

Technical Support Document for Expansion Zone PWSCC in Roll Transitions (Rev.1)

EPRI Final Report, EPRI NP-6864-L, Rev.1, December 1991..

Electric Power Research Institute Steam Generator Reliability Project Ron Gamble NRC Meeting

December 22,1992 RG-3*NRCo 12/22/92

o-

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i V

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L PWSCC IN EZ ROLL TRANSITIONS i

i y

Degradation Mechanism L

Scope

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Inspection Criteria Margin Against Tube Rupture j

i i

l Leakage Limits t

Plant Specific Applications t

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SCOPE j-Repair limit is based on crack length not depth fi I

Criteria apply to axial cracks (part-throughwall/throughwall) i l

Tubes with circumferential. cracks or axial cracks having distinct I

circumferential components will be repaired.

L Tubes with closely spaced axial cracks will be repair.

j l

Applicable to PWSCC in tube roll transitions above and below the tubesheet.

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Case B. Partial depth roll tube N

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Case C. Ovei rolled tube a axiallength of E2 PWSCC crack R tube rnean radius t - tube thkkness Figure 1-1. Schematic of EZ Roll Transition Axial Cracks 1-2 2

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INSPECTION CRITERIA i

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Inspect 100% of tubes in affected region.

L Use RPC to increase detection reliability in EZ.

t Assessment of RPC m~easurement uncertainty.

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I Confirmation of NDE method is supported by significant service

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experience and laboratory experiments.

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6" RG-3eNRCo12/22/92 I

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i-MARGIN AGAINST TUBE RUPTURE i

Conservative correlation between tube rupture strength and crack i

length.

i Margins consistent with regulatory and ASME Code guidelines.

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3x Normal operating AP (R.G.1.121) 1.4x Accident load AP (ASME Code) j i

i Allowance for tube sheet constraint, NDE measurement uncertainty, and l

l crack growth to obtain repair limit.

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l Burst pressure for tubes with roll transitions within tubesheet equals-

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I undegraded tube burst pressure (small diametrical gap between tube and tubesheet).

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LEAKAGE LIMITS Normal operation - 150 gpd.

l Faulted load -limit number and size of cracks to remain with 10CRF100 dose limits.

Predicted accident leak rate determined from deterministic analysis based on leak rate vs. corrosion crack length correlation.

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RG-3*NRCol2/22/92

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SEVERE ACCIDENT ISSUES T. L. Gerber NRC Meeting 12/22/92 l

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h"stC+12/22/92*TIG3

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-7 At the NRC Reguiatory Information Conference in July 1992, it was reported that "the staff is evaluating the impact of

. steam -generator ' integrity issues as they affect the implications of severe accidents."

it was noted that "the staff is examining significant core damage 1 sequences that may result in tube failures at creep regime temperatures. and subsequent containment bypass."

The possibilities of alternative repair criteria (ARC) (1) increasing the likelihood of multiple tube rupture during.an MSLB design basis event.or (2) impacting the consequences of a TMLB' severe accident have been considered.

1 12/22/92 TG

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DESIGN BASIS ACCIDENT (MSLB) i'i ARC explicitly address tube integrity for the MSLB accident margins against rupture satisF; regulatory requirements and are. consistent with current depth based repair' criteria i

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-tube leakage -under accident design basis loading conditions t

is bounded by -limiting the total predicted. leakage for those tubes that remain in : service.

ARC do.not increase. the probability of multiple : tube rupture (or. significant-tube leakage) during a MSLB design basis accident.

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SEVERE ACCIDENTS (TMLB')

The maximum possible tube differential pressure during a TMLB' is the same as during a MSLB.

Unless tube temperature are raised significantly above operating temperatures

(>1000 F) tube mechanical properties will change little tubes will not rupture tube leakage will not appreciably increase over that aservatively predicted for a MSLB.

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SEVERE ACCIDENTS (TMLB') con't.

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4 MAAP-4..TMLB' severe accident analyses have been run with tube ' leakage at and above that conservatively predicted for.a i.

MSLB accident.

i Preliminary results suggest even -with tube leakage above that postulated lfor a' MSLB accident.

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tube temperatures lag hot leg temperatures

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. tube te.mperatures' at the time of hot. leg failure are i

<1000 F i

i ARC do:not impact the consequences of a TMLB' severe accident.

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