ML20196K719

Summary of 990624 Meeting with Nei,Epri & Industry to Discuss Technical Issues Involving Implementation of NEI 97-06, SG Program Guidelines. Meeting Attendees & Handouts Encl

ML20196K719
Person / Time
Issue date:	07/07/1999
From:	Tim Reed NRC (Affiliation Not Assigned)
To:	Bateman W NRC (Affiliation Not Assigned)
References
NUDOCS 9907120031
Download: ML20196K719 (26)

Text

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j NUCLEAR REGULATORY COMMISSION o

2 WASHINGTON, D.C. 2056fM001 July 7, 1999 MEMORANDUM TO: William H. Bateman, Ch:ef Materials and Chemical Engineering Branch Division of Engineering THRU:

Edmund J. Sc!!ivan, Chief b

NDE & Metallurgy Section 6

Materials and Chemical EngineeringBranc FROM:

Timothy A. Reed, Senior Project Manager NDE & Metallurgy Section Mater!als and Chemical Engineering Branch

SUMMARY

OF THE JUNE 24,1999, TECHNICAL MEETING WITH NEl/EPRl/ INDUSTRY TO DISCUSS ISSUES INVOLVING IMPLEMENTATION OF NEl 97-06 On June 24,1999, the NRC met with representatives of Nuclear Energy Institute (NEI), Electric Power Research Institute (EPRI) ed industry to discuss technical issues involving the implementation of NEl 97-06 " Steam Generator Program Guidelines." Meeting attendees are identified in Attachment 1. The meeting handouts are provided as Attachment 2.

The following key points were made during the meeting:

1. StructuralIntegrity Assessments of Steam Generator Tubes industry provided a revised white paper regarding its position on structural integrity assessments of steam generator (SG) tubes. Industry maintains the position that structural integrity assessments should be performed using a factor of safety of three, and that the loading condition should be the pressure across the SG tube (AP) st normal full power operation. The staff is continuing to evaluate industry's position that nominal full power AP is the appropriate value to use in the 3 times AP calculation. To support its review, the staff

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requested information (consistent with its previous request in the May 6,1999, technical meeting) concerning transients or startup conditions that may result in APs greater than what exists at nominal full power conditions. Industry agreed to provide the requested information in generic terms. However, industry indicated that the information may not be complete (i.e., may O) not represent a comprehensive survey of industry) and it may not be bounding. Industry indicated that it plans to provide the requested information as part of the safety analysis]q supporting its gennic CG technical specification (TS) revision request.

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2 Industry also agreed to revise the SG tube integrity guidelines to include language addressing the issue of secondary loads (consistent with industry's ir situ pressure test guidelines). In addition, industry will present the information during a telecon with the NRC on July 9,1999, at 1:00 p.m.

2. Definition of SG Tube Burst The staff and industry agreed on the definition of SG tube burst as proposed by industry.

j industry indicated that it will revise NEl 97-06 to include in situ test reporting criteria consistent with Chapter 9 of the in situ pressure test guidellic.ss.

1

3. Revisions to NEl 97-06 j

i NEl indicated that it plans to address industry comments in revision 1 of NEl 97-06 and submit revision 1 to the NRC on a schedule consistent with its submittal of the generic SG TS change requast. However, NE ' indicated that given the amount of interaction needed with industry on NEl 97-06 revision 1, it may be submitted somewhat later than the generic SG TS change requsst.

4. Twelve-Month Reporting of SG Inspection Activities in response to industry's request that the staff reconsider the need for the 12-month SG report from all pressurized water reactor (PWR) licensees, the staff suggested, as an alternative, a report that would be required only when a certain threshold is exceeded (for example, a report j

would be required if greater than 1% of the SG tubes inspected were found to be defective). In addition, the staff suggested that the report be submitted to the staff in a more timely manner (for example, within 90 days from plant startup). Industry agreed to consider the staff"s proposal and provide its position during the July 9,1999, telecon.

5. Revision to SG TS/ Technical Requirements Manual (TRM)

Industry presented a proposed revision to the SG TS/TRM. The staff identified concerns with the proposal, particularly with regard to the determination of SG operability. The staff will complete its review of the proposal and provide its specific comments during July 9,1999, telecon.

6. Changes to the Performance Criteria and 10 CFR 50.59 The staff asked industry whether or not changes to the performance criteria, which per the industry proposal are located in the TRM and subject to 10 CFR 50.59, would require prior NRC review and approval prior to implementation. It is the strif's view that changes to these criteria could potentially be made under 10 CFR 50.59 without prior staff approval, while at the same time result in an increase in plant risk. The staff is concerned about the significant contribution to overall PWR plant risk from SG related events (i.e., containment bypass due to SG tube failure) and the importance of tube structural and accident leakage integrity in maintaining the current level of risk. Industry will evaluate the staff's question and consider alternatives for ensuring that changes to performance criteria receive prior NRC review and approval prior to implementation.

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- 7. Grandfathering and First Time Approval of Repair Criteria and Repair Methods The staff indicated that grandfathering repair criteria and repair methods (i.e., enabling a licensee to continue to be able to utilize repair criteria and repair methods already approved for its plant under the new proposed SG framework) is conceptually a straight-forward process of simply relocating these methods and criteria to the TRM. The staff indicated that the TRM will need to contain language which is equivalent to the previous TS language, although this may result in significantly less description if the SG program already addresses portions of the approved approaches.

Regarding the issue of first-of-a-kind approvals (i.e., generically approving new or previously approved repair criteria or repair methods by establishing constraints that must be satisfied by licensees wishing to implement such approaches), the staff indicated that the plant-specific safety evaluation reports (SERs) developed in the past are not adequate for this purpose. This is based on the fact that in the past, SERs were not written with this objective in mind (i.e.,

establishing the list of constraints thr.t would enable generic application of the approved approach), and instead were written to grant approval for a specific approach for a specific

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plant application. The staff suggested that it could issue a formal position for previously q

approved repair criteria and methods as requested by industry.

The staff indicated that revisions to approved topicals may need to be reviewed by the staff unless process type revisions (for example, changes to a method of cieaning) can be separated from portions of the approach which directly relate to the acceptability of the approach. The staff suggested that industry consider this issue in its development of future topical reports and anticipate, wherever possible, such potential future changes. This could allow the staff to address these issues in its generic approval, thereby precluding the need for reviewing future revis. eons to topical reports.

-The staff commented that for situations involving free span oegradation or the use of new materials (which potentially increase risk), this type of generic approval might be difficult due to the risk implications of such approaches and its plant-specific nature.

8. Schedule NEl indicated that due to the concems raised by the staff with regard to its proposed revision to the SG TS/TRM, its schedule for submitting its generic SG TS change request (August 17, 1999), will be delayed. NEl will provide the staff with its revised schedule after the July 9,1999, telecon.

As indicated above, a telecon will be conducted on July 9,1999, at 1:00 p.m., between NRC

and industry to discuss further the open technical and regulatory issues.

Attachments: As stated i

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NEl/EPRl/lNDUSTRY TECHNICAL. MEETING NEl 97-06 IMPLEMENTATION ISSUES JUNE 24,1999 LIST OF ATTENDEES NAME ORG/ POSITION

1. Tim Reed NRC/NRR/DE/EMCB

2. Jim Riley NEl

3. Gaiy Henry EPRI

4. Don Streinz ABB

5. Emmett Murphy.

NRC/NRR/DE/EMCB

6. Steve Long NRC/NRR/DSSA/SPSB

7. BillBateman NRC/NRR/DE/EMCB

8. Phil Rush NRC/NRR/DE/EMCB

9. Gary Elder Westinghouse

10. Rich Barrett NRC/NRR/DSSA/SPSB

11. David Goetcheus TVA

12. Rick Mullins Southern Co.

13. Kris Thomas NRC/NRR/DLPM

14. Gary Boyers FPL

15. Bob Keating Westinghouse

16. Bob Tjader NRR/TSB

17. Edmund Sullivan NRC/NRR/DE/EMCB

18. Richard Pearson Northern States Power

10. Mohammad Behravesh EPRI

20. Kevin Gweeney Arizona Public Service

21. Mati Merito EPRI

22. Eileen McKenna NRC/NRR/ DRIP

23. Helen Cothron TVA

24. Edward Throm NRC/DSSA/SPSB

25. Michael Short Southern California Edison 1

4 ATTACHMENT 1 4

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Steam Generator Integrity Technical Meeting NRC Staff i

NEI Steam Generator Task Force June 24,1999 Rockville, MD

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Agenda Technical positions e

3 NO delta P

. Tube burst NEI 97-06, rev 1 m Regulatory Framework Technical Specifications Technical Requirements Manual First time approval SER constraints Grandfathering approved ARCS a Risk Issues

. Severe Accidents Schedule

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f Technical issues a Deterministic Loading Definition - 3 NO dP e Definition -

. For structural integrity assessments using a factor of safety of three, the loading condition sha!! be the pressure across the steam generator tube at normal full power operation as defined in the design or equipment specification.

e Discussion.

Definition conceptually accepted at 5/6/99 meeting White Paper revised to provide additional information f

'V' Technical Issues a Tube burst / tube rupture Burst is defined as the gross structural failure of the tube wall. The condition a

typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (plastic) tearing of the tube material at ti.e ends of the degradation, e Basis Definition accepted at 5/6/99 meeting Inconclusive test results

. In situ Pressure test guidelines provide guidance for testing and evaluating inconclusive results

. NEl 97-06 requires utilities to evaluate and report csrcumstances should the test results indicate a failure to satisfy structural and leakage performance criteria.

g White paper with detailed discussion 2

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t Definition cf 3AP Deterministic Structural Performance Criteria Pressure Loading Definition Recommended Definition -

For structural integrity assessments using afactor of safety of three, the loading condition shall be the pressure across the steam generator tube at normalfull power operation as defined in the design or equipment specification.

Discussion NEI has endeavored to define the appropriate, historically supported loading definition for the deterministic structural integrity performance criteria. It is NEI's positic n that the description contained in DG-1074 is not consistent with past NRC positions and the intent of the ASME Code for original design and evaluation of inservice components.

This paper provides a review of ASME Code requirements and a historical overview. A discussion of definition application is also provided.

ASME Code Review A review of multipM industry documents, including NSSS design reports, sleeving topical reports, and regulatory submittals, indicates that all NSSS designers and multiple utilities have utilized the same ASME Section III Code equation, from paragraph NB 3324.1 to define minimum allowable tube wall thickness. A typical form from Reference 1is as follows:

APR.

1. "* ~ P,,, - 0.5 AP '

where imi,, = minimum acceptable tube wall AP = primary to secondary pressure differential Rt = tube radius Pm = primary stress limit, i.e.,

P. normal s Su/3 P upset s Sy Pm faulted s 0.7Su

Definition of SAP In Reference 2, the same requirement was written as 3APR.

Su - 0.5(Pl+ P,)'

The ASME Section III design requirement is therefore, that tube wall thickness should be maintained at greater than t.<,. All the submittals surveyed used normal full power AP as the basis for primary-to-secondary pressure differential. It is important to note that these documents received Staff review and approval.

Historical Regulatory Perspective NEI has reviewed historical information regarding Staff positions on this issue. From a review of the testimony of James Knight 3, Raymond Maccary4, Atomic Safety and Licensing Board 5 and Regulatory Guide 1.1216, it is clear that the primary requirements for margin of safety for steam generator tubes were:

1 (1) tubes with detected acceptable defects would not be stressed during the full range of normal reactor operation beyond the elastic range of the tube

material, (2) the factor of safety against failure by bursting under normal operating conditions should not be less than three.

Meeting requirement (1) assures that at normal operation, the maximum stress in the degraded tube will never exceed the yield stress for the tube material.

Normal i

operation, as used here, includes start-up and operational transients that are included in the design specification for the steam generator. For requirement (2) an explicit description of normal operating conditions that includes the full range of operational transients is not specified. However, Mr. Knight does point out that "... ncw steam generator tubes are typically manufactured with a wall thickness much greater that the wall thickness required by the design rules of ASME Code,Section III" (see previous discussion) and Mr. Knight further indicates that approximately twice the thickness (nominal / minimum) is typical. If in fact minimum wall calculations were based on the full range of operational transients or RCS design pressure safety limits, the typical maigin above minimum wall would be on the order of 10-20%, meaning plugging limits would have to be lowered well below most licensees current 40% criteria.

Statements in Regulatory Guide 1.121 provide additional insight for industry's consistent application of this criterion. Section 3.a.1 of the regulatory guide states that loadings associated with normal plant conditions, including start-up, operation in the power range, hot standby, cooldown and all anticipated transients should not produce a primary member stress in excess of the yield stress. Once again, this statement is

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DefinitioncfSAP explicit, whereas the stmpants'in Section 2.a.2 and 2.a.4 of the RG simply state a

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margin of safety under normal operating conditions should not be less than three (3).

Definition Application - OriginalDesign In establishing the' safe limiting conditions for steam generator tubing, plugs and sleeves, the effects 'of loadings both at normal operation and postulated accident conditions are always evaluated. As such, all design requirements, specified in ASME Section III, Subsection NB, are applied to the initial design of RCS pressure boundary components. As shown in Equation 1, the steam generator components are addres ed in s

terms of remaining wall _ thickness for normal, upset and faulted conditions. For the documentation surveyed for this paper, the requirement for normal operation of S,/3 was limiting.

Consideration of primary membrane stress, bending loads, seismic, combined primary plus secondary stresses, and fatigue loads are addressed in the original design as required by the applicable Code paragraphs. Industry will continue to implement these requirements as appropriate for new steam generator designs and new repair methods proposed and submitted for NRC approval. However, as stated in Subsection NB-1110,

"....[the ASME Section III rules] do not cover deteriorations that may occur in service as a result of corrosion,' radiation effects, or instability of the material." Instead, the performance of inservice steam generator components is assessed via condition monitoring and operational assessment as required by NEl 97-% and nct the ASME

. Code.

Definition Application - Condition Monitoring Regulatory Guide 1.160 and 10CFR50.65 state, in part, that licensees shall monitor the performance and condition of safety related components against established goals in a manner to provide reasonable assurance that such components are capable of fulfilling their intended functions. These goals are to be established commensurate with safety and, where practical, take into account industry-wide operating experience. In the case of steam generators, the performance criteria and definitions in NEI 97-% provide the necessary protection against spontaneous burst, and burst as a condition of Chapter 15 accidents.

The information, provided in this paper, has further demonstrated that the use of full power normal operating pressure differential is historically supported both by operating experience and accepted regulatory positions. The use of this definition provides consistent application, and is measurable and comparable within the condition monitoring process. While the factor of safety applied during the condidon monitoring

. process is consistent with the ASME Code, it should be noted that there is no Code j

requirement (either in Section III or XI) to conduct proof testing of components to be removed from service. Industry has, instead, adopted the accepted historical definitions described above as a basis for monitoring steam generator performance.

Definition of SAP For example, industry, through the development of the Steam Generator In Situ Pressure Test Guidelines 8, has established the appropriate guidance for the verification of integrity performance through proof testing. In addition to a safety factor of three, the guidelines include adjustments to ensure representative testing. Typical adjustments include temperature, test equipment instrument error and bladder corrections. The L

guidelines further state: "That all appropriate loads should be considered for the d=:ge form. For example, axial loads due to tube-to-shell temperature differences in OTSGs, or axial loading associated with locked tube supports in RSGs should be addressed to ensure test conditions are at least as severe as those expected during operating and accident events." These examples indicate that the use of industry's 3AP definition is conservative and appropriate to demonstrate structural integrity j

performance of degraded steam generator tubing.

In summary, the criteria and definitions in NEI 97-06 and the supporting industry documents are intended to provide reasonable assurance of steam generator tube integrity. The structural and leakage criteria (including operational leakage) and assessment processes are measurable, achievable and provide appropriate defense-in-depth.

I Conclusion i

i The use of normal full power operating pressure differential has time and time again been utilized as the basis for compliance with the three AP criteria. The Technical Specification plugging limits for the plants surveyed were based on full power differential pressure. The surveyed design reports for minimum wall calculations for q

tubing and sleeves and Regulatory Guide 1.121 compliance calculations were based on j

normal full power differential pressure. Additionally, use of the values listed in the design or equipment specification ensures consistency of application.

As such, NEl believes the interpretation in DG 1074 constitutes a regulatory position 1

that is either new or different from a previously applicable staff position, which represents a backfit per 10CFR 50.109. Additionally, based on the consistent application by industry in regulatory submittals and subsequent approvals by the NRC since 1975, compliance exception is not considered applicable, since this exception was intended to address situations "in which the licensee has failed to meet known and established standarda of the Commission because of omission or mistake of fact." In conclusion, the current definition as provided by NEI in Reference 7 is supported for inclusion in NEI 97-06.

References:

1. WCAP 13698 Revision 1, Laser Welded Sleeves for % inch Diameter Tube Feed-Ring Type and Westinghouse Preheater Steam Generators, Westinghouse Electric Corporation, May 1993.

Definition ef 3AP

2. Letter LD-83-058, AB-CE to Mr. Darrell Eisenhut, Director, Division of Licensing USNRC, Rapid Depressurization and Decay Heat Removalfor System 80, dated June 29,

1983,

3. Testimony of James Knight before the Atomic Safety and Licensing Board, January 1975.

4. Testimony of Raymond Maccary before the Atomic Safety and Licensing Appeal Board, January 1976.

5. Atomic Safety and Licensing Board, LBP-75 27, May 1,1975

. 6. Dratt Regulatory Guide 1.121, ~ Basesfor Plugging Degraded PWRSteam Generator Tubes, August 1976

7. NEI letter to NRC (S.J. Collins), Industry Steam Generator Program Initiative, dated December 17,1998

8. EPRI TR 107620 R1 Final Report, Steam GeneratorIn Situ Pressure Testing Guidelines, Revision 1, June 1999.

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- Definition ofBurst for SG 'lkbes Definition of Burst

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i RecommendedDefinWon ofBurst-Bunt is defined as the gross structuralfailure of the tube wall. The condition typically corresponds to an unstable opening displaxment (e.g., opening area increased in rest

• eto constant pressure) accompanied by ductile (plastic) tearing of the tube material at the utds of the degradation.

Discussion Since a burst definition is a required component for condition monitoring and operational assessment, a definition that can be analytically defined and is capable of being assessed via in situ and laboratory testing is required. Furthermore, the definition must be consistent with ASME Code definitions, and one that applies to most forms of tube degradation. Additionally, the definition is intended to demonstrate accord with the testimony of James Knight 1, and compliance with the historical guidance of the Regulatory Guide 1.1212. The definition of burst per these documents is in relation to grossfailure of the pressure boundary, e.g., "the degree of loading required to burst or collapse a tube wall is consistent with the safety factor in Section III of the ASME B&PV Code". Burst, or gross failure, according to the Code would be interpreted as a catastrophic failure of the pressure boundary.

The proposed definition must also support field application of the condition monitoring process. For example, verification of structural integrity during condition monitoring may be accomplished via in situ testing. Since these tests do not have the' capability to provide an unlimited water supply, nor the capability to maintain pressure under certain leakage scenarios, opening area may be more a function of fluid reservoir rather than tube strength. Additionally, in situ designs with bladders may not be reinforced. In certain cases, the bladder may rupture-when tearing or extension of the defect has not occurred. This condition may simply mean the opening of the flanks of the defect was sufficient to permit extrusion of the bladder, and that the actual, or true, burst pressure was not achieved during the test.3 The definition is also intended not to characterize local instability, or for example,

" ligament pop-through", as a burst. The onset of ligament tearing need not coincide with the onset of a full burst. As an example of not having a burst, consider an axial l

crack about 0.5" long with a uniform depth at 98% of the tube wall. Deformation during pressurization would be expected to lead to failure of the remaining j

ligame~nt, (i.e., extension of the crack tip in the radial direction) at a pressure below that required to cause extension at the tips in the axial direction. Thus, this would

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Definition ofBurat far SG Thbes three against crack extension in the axial dire j

e y of Similar conditions have been observed for deep wear indications e demonstrated.

Additional information found in Reference 4, further supports definition of burst. It is noted that if tube failure, i.e., burst is e proposed deformation of the crack to the extent that t' s plastic parallel elliptical configuration, the tubingaides of the crack open to a non-beyond those values before reaching a condition of gross failurei s Applicatien - Condition Monitoring condition monitoring process. Verificatio r

e monitoring may be accomplished via in situ testing. Sinc on not have the capability to provide an unlimited water suppl o

maintain pressure under certain leakage scenarios, openin y to function of fluid reservoir rather than tube strength. Additiona designs include bladders that are not reinforced. As a result, s u tool bladder may rupture when tearing or extension of the defect es, the As such, the EPRI Steam Generator In Situ Pressure Test Guide industry guidance regarding the evaluation of field test results S provide Requirements recommends actions such as post-test ECT and n., ost-Test provides guidance for the evaluation of pr on in i

Industry Database provides the protocol for data entry to t generator database including information such as flaw size, targ steam maximum pressure achieved, tool type (including bladder) and w Requirements and Tool Qualifcation, ident the user in obtaining reliable and consistent test results o assist d

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R erence 7, ui 1 nes for Ste Test and Examination.

n ra or u co NEI 97-06 requires utilities to submit condition monitoring results structural and leakage performance criteria. evaluation ng an

Definition of Burst for SG 1hbes Summary In conclusion, it is industry's position that the proposed definition is consistent with the Staff's definition in DG-1074 and with historical regulatory references, while providing necessary clarification to support condition monitoring and operational assessment for most defect forms.

References:

1.

Testimony of James Knight Before the Atomic Safety and Licensing Board, Docket Nos. 50-282 and 50-306, January 1975.

2.

Regulatory Guide 1.121 (Draft), " Bases for Plugging Degraded PWR Steam Generator Tubes," United States Nuclear Regulatory Commission, August 1976.

3.

EPRI TR-105505, Final Report, Burst Pressure Correlationfor Steam Generator Tubes with Through-Wall Axial Cracks, October 1997.

4.

NUREG-0491," Safety Evaluation Report Related to the Preliminary Design of the Standard Reference System RESAR-414," Docket Number STN 50-572, United States Nuclear Regulatory Commission, November 1978.

5.

Draft Regulatory Guide DG 1074, Steam Generator Tube Integrity, December 1998.

6.

EPRI TR-107620 R1 Final Report, Steam Generator In Situ Pressure Testing Guidelines, Revision 1, June 1999.

7.

EPRI TR 016743-V4R1 Final Report, Guidelines for Steam Gen.erator Tube Section Removal, Test and Examination, December 1997.

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Technical Issues a NEI 97-06, Rev 1

. Pending changes

. Performance criteria

+ Burst definition

+ Special report contents

. Deviation guidance gti

= Endorsement ofNEI 97-06 Technical Issues a Steam GeneratorIntegrity Assessment Guideline scheduled to be completed by j

7/31/99 m Qualified sizing techniques

. Will revise NEI 97-06 to include industry resolution proposed in the 5/6 handout gl 3

19. NRCStafYComments (10f7M8): Plur on DMeMan NEIguidelines shouldstate that " plug on detection " applies when NDEsystems are not validatedforgiven defect type.

Industry Resnonse(12/17/98):

The intent of this requirement already exists in NEI 97-06, Section 3.3.1, Repair Limits. The section states in part:

"Forplants experiencing a damageform or mechanismfor which no depth sizing capability exists, tubes identified with such damage are

" repaired / plugged-on-detection " and integrity should be assessed. "

NRCResolution (2/10M9):

The NEIwords "no depth sizing capt.bility exists "should be clarified to say no qualyled sizing capability (per EPRIAppendices G &H, Section 6) exists. "

NEI nronosed resolution (5/6/99):

)

The word " qualified"in the NRC words implies an acceptance criterion for sizing. The sentence should read as follows:

"Forplants experiencing a damageform or mechanismfor which no depth sizing capability exists (per EPRI Appendices G &H, Section 6), tubes identified with such damage are " repaired / plugged-on-detection " and integrity should be assessed."

Industry considers this item closed.

i i

e m

t Regulatory Framework

= Tech Specs admin section:

. Reporting requirements i

.12 month report

. Referenced licensee controlled documents or TRM

. See draft

'9' Regulatory Framework e Licensee Controlled Documents:

. Content not format

. Deterministic performance criteria

. Includes NRC notification requirements

. Operability statement removed

. Reporting requirements

. Guidance in NEl 97-06

. Fall within 10CFR50.59

. See draft 4

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Pr psc:d SG Tech Sp:c Chengss j

Admin Section and Bases 5.5.9 Steam Generator Prorram 5.5.9.1 General Requirements A program shall be implemented through the licensee's procedures to ensure that the SG tube integrity performance criteria are maintained. The SG performance criteria are defined in licensee controlled documents (e.g., UFSAR, the Technical i

Requirements Manual).

'5.5.9.2 Condition Monitoring Assessment Condition Monitoring Assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means,' prior to the plugg.ng or repair of tubes. Condition Monitoring Assessments shall be conducted during each outage (scheduled or unscheduled) during which the SG tubes are inservice inspected, plugged, or repaired to confirm that the performance criteria are being successfully maintained.

-5.5.9.3 Tube Repair Criteria

- Tube repair criteria shall be described in and implemented by the steam generator program. Prior to use, the repair criteria must be reviewed and approved by the NRC. The licensee may use repair criteria reviewed and approved by the NRC for

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another facility (or generically) if the licensee demonstrates that the repair criteria i

are. applicable to the licensee's facility for specific defect types and any limitations and I

conditions set forth in the NRC's safety evahiation of the criteria are met by the licensee. The licensee's demonstration must be documented in a written report available on site for NRC inspection.

5.5.9.4 Tube Repair Methods

' Steam generator tubes must be repaired using repair methods describcd in and implemented by the steam generator program. Prior to use, the repair methods (excluding tube plugs) must be reviewed and approved by the NRC. The licensee may use repair methods reviewed and approved by the NRC for another facility (or generically) if the licensee demonstrates that the repair metheds are applicable to the l

licensee's facility for specific defect types and any limitations and conditions set forth in the NRC's safety evaluation of the methods are met by the licensee. The licensee's demonstration must be documented in a written report available on-site for NRC inspection.

As of 06/24/99,6:38 PM Page 1 of 2

Prep:s:d SG Tech Sp;c Chcngas Admin Section and Bases i

i 5.6.10 Steam Generator Tube Insnection Renort Steam generator tube integnty shall be documented in a post inspection steam generator tube integrity report, submitted within 12 months after completion of

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the inservice inspection. The report shall summarize the scope ofinspections performed, active degradation mechanisms found, NDE techniques utilized for each degradation mechanism; number of tubes plugged or repaired during the inspection outage for each active degradation mechanism; repair method utilized and the number of tubes repaired by each repair ynethod; total number and percentage of tubes plugged and/or repaired to date and the effective j

plugging percentage in each steam generator; and condition monitoring including the results of tube pulls and in-situ testing.

B LCO 3.4.13 RCS Operational LEAKAGE APPLICABLE SAFETY ANALYSIS:

The Technical Specification requirement to limit primary to secondary leakage through any one steam generator to less than 150 gallons per day is significantly less than the initial condition of the safety analysis.

A limit of 150 gallons per day is based on operating experience as an indication of one or more propagating tube leak mechanisms. This leakage rate provides additional assurance against tube rupture at normal and faulted conditions and provides additional assurance that j

cracks will not propagate to burst prior to detection by leakage monitoring methods and commencement of plant shutdown.

1 As of 06/24/99,6:38 PM Page 2 of 2

9.

Prcpos:d Licsnsso Centrcil::d Menuni Chengaa for Steam Generator Integrity Information listed here are the key elements to be included in the licensee controlled documents referenced in Technical Specification 5.5.9.1 TRM x.y Steam Generators Steam generator tube integrity performance criteria are:

(i) Structural criterion:

Steam generator tubing shall retain a margin of 3.0 a' gainst burst under normal full power operation and a margin of 1.4 against burst under the limiting design basis accident concurrent with a safe sbutdown earthquake.

(ii)~ Accident induced leakare criterion:

The primary to secondary accident induced leakage rate for the limiting design basis accident shall not exceed the leakage rate assumed in the accident analysis in terms of totalleakage rate for all steam generators and for leakage rate for an individual steam generator, not to exceed [1 gpm].

- (iii) Operational leakare criterion The operational primary-to secondary leakage for any one steam generator shall not exceed 150 GPD.

t Defects which exceed the repair criteria shall be repaired using an approved j

nethod or removed from service by plugging the tube. The repair criteria approved for use are:

[ existing NRC approved repair criteria, e.g.,40% nominal tube wall thickness]

[..........]

The repair methods approved for use are:

[ laser welded sleeves]

[TIG welded sleeves]

[.........]

REPORTS:

1. Ifit is determined during a steam generator inspection that any of the performance criteria are not met, a special report should be submitted to the NRC within 90 days after entering Mode 4 providing the following information for the degradation of concern:

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Propes:d Licancso Centrcilsd 'Menual Changas for Steam Generator Integrity A. Performance criteria exceeded B. Cause for exceeding the performance criteria C. Corrective actions taken to prevent recurrence D. Basis for planned operating period-2.' Ifit is determined during a steam generator inspection that a steam generator tube failed to meet a 1.0 margin against burst at the limiting design basis accident differential pressure, the NRC should be notified in accordance with 10CFR50.72(b).

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3. Ifit is determined in Modes 1 through 4 that any of the performance criteria or the requirements of this [TRM section] are not met, the NRC shall be notified in accordance with 10CFR50.72(b) and a special report should be submitted to the NRC within 30 days providing the information listed in 1. above.

TRM r.z Definitions The following definitions are applicable to Technical Requirement x.y only.

Accident induced leakage rate means the primary-to secondary leakage rate occurring during postulated accidents other than a steam generator tube rupture.

This includes the primary to-secondary leakage rate existing immediately prior to the accident plus additional primary to secondary leakage induced during the accident.

Burst is defined as the gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (plastic) tearing of the tube material at the ends of the degradation.

j Normal full power operation is defined as the conditions existing during Mode 1 operation at the maximum steady state reactor power as defined in the design or equipment specification.

Repair Criteria are those measured parameters at or beyond which the tube must be repaired or removed from service by plugging.

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. Repair Methods are those means used to reestablish the RCS pressure boundary integrity of SG tubes without removing the tube from service. Plugging of a steam

. generator tube is not a repair.

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Regulatory Framework

= SER constraints on first time generic approval of new repair methods and repair criteria

. NRC draft needed

= SER grandfathering approved repair criteria and repair methods

. Utilities will use latest existing SERs and demonstrate applicability based on 10CFR50.59

. Utility will request SER revision if required

'h'l Risk issues a Severe accident guidance

. When to consider

. Criteria for success 5

O Schedule License Package schedule

. License change package to NRC

~8/17/99

= Next Meetings

+Next technical meeting in early September

+Next senior management meeting?

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Grandfathering ARCS and sleeving methodologies may be moved to the TRM for plants where such ARCS and sleeving methodologies have been approved (i.e., as identified in current TS).

Conceptually straight forward. However, current TS requirements pertaining to ARCS and sleeving methods are frequently integrated into the TS in such a way 1

that existing TS wording cannot be transferred verbatim to the TRM. Licensees will need to propose words for TRM (with respect to ARCS and sleeves) which are equivalent to existing TS requirements when submitting the industry proposed TS change packages.

First-of-Kind This concept would be expressed in new TS. As currently proposed, it would permit licensees to use repair criteria or repair methods approved by NRC for another facility (or generically) if the licensee demonstrates that the repair criteria

]

or repair method is applicable to the licensee's facility for specific defect types and any limitations or conditions set forth in the NRC's safety evaluation of the methods are met by the licensee.

Problem:

SEs to date which approve plant specific repair criteria or repair methods l

generally do not provide a clear basis for establishing the gerieric applicability of the staff's plant-specific finding. In general, the generic applicability of staff conclusions in plant specific SEs has not been reviewed by CRGR.

Potential Solution:

Staff would issue formal position conceming the generic applicability of its plant-specific finding before the "first-of-kind" concept can be applied.

This could be in the form of a letter to the appropriate industry group (e.g.,

NEI) and/or could be incorporated into a regulatory guide which identifies all generic staff positions relative to repair criteria and repair methods approved for specific plants.

Staff would prepare such formal positions only when requested by the vender or industry or utility group.

The "first-of-kind" concept in TS would read "the licensee may use repair 4

criteria or repair methods approved generically by NRC subject to the limitations and conditions set forth in the staff's approving document.

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Other issues Application of the first of kind concept is most straight forward when there is a generic topical report which the staff can reference in its generic finding.

Should utilities desire to implement a later revision to a topical report approved by the NRC, the staff must first be requested to approve the later revision.

ABB/CE Leak Tight (TIG-welded) Sleeves Defining a regulatory position concerning the generically applicability of this sleeving method, already approved for some plants, seems relatively straight forward from purely technical standpoint. E.G.:

Sleeving methodology documented in CEN-629-P, Rev 2, and CEN-629-P, Rev 1, Addendum 1, is approved for use in Westinghouse Model 44 and 51 steam generators subject to the following:

The sleeving methodology shall be identified in the TRM and shall be equivalent to the TRM requirements approved for the first-of-kind application. [ Note, existing TS requirements for these sleeves require inspections of sleeves at each refueling outage with sampling in accordance with EPRI Guidelines. Applicable plugging limit is 31% throughwall.)

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Visualinspections shall be performed prior to sleeve installation to f

verify adequate tube cleaning at the intended weld location. This shall be a 100% visual examination until sufficient field experience

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at the subject plant is gained to support (in the licensee's judgement) and statistically based sampling plan.

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Post weld acceptance examinations shall include visual inspections, in addition to UT and eddy current, until sufficient field experience is gained that there is a high rate of weld acceptability.

Post-weld heat treatment is optional.

Eddy current methods qualified in accordance with ABB/CE report 97-FSW-001-P, Rev 0, are acceptable for welded sleeve inspections for probe speeds not exceeding 0.45 inches /sec for detection or 0.1 inches /sec for detection and characterization.

Higher probe speeds may be used subject to requalification for these speeds.

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7. Grandfathering and First Time Approval of Repair Criteria and Repair Methods The staff indicated that grandfathering repair criteria and repair methods (i.e., enabling a licensee to continue to be able to utilize repair criteria and repair methods already approved for its plant under the new proposed SG framework) is conceptually a straight-forward process of simply relocating these methods and criteria to the TRM. The staff indicated that the TRM will need to contain language which is equivalent to the previous TS language, although this may result in significantly less description if the SG program already addresses portions of the approved approaches.

Regarding the issue of first-of-a-kind approvals (i.e., generically approving new or previously approved repair criteria or repair methods by establishing constraints that must be satisfied by licensees wishing to implement such approaches), the staff indicated that the plant-specific safety evaluation reports (SERs) developed in the past are not adequate for this purpose. This is based on the fact that in the past, SERs were not written with this objective in mind (i.e.,

establishing the list of constraints that would enable generic application of the approved approach), and instead were written to grant approval for a specific approach for a specific plant application. The staff suggested that it could issue a formal position for previously approved repair criteria and methods as requested by industry.

The staff indicated that revisions to approved topicals may need to be reviewed by the staff unless process type revisions (for example, changes to a method of cleaning) can be separated from portions of the approach which directly relate to the acceptability of the approach. The staff suggested that industry consider this issue in its development of future topical reports and anticipate, wherever possible, such potential future changes. This could allow the staff to address these issues in its generic approval, thereby precluding the need for reviewing future revisions to topical reports.

The staff commented that for situations involving free span degradation or the use of new materials (which potentially increase risk), this type of generic approval might be difficult due to the risk implications of such approaches and its plant-specific nature.

8. Schedule NEl indicated that due to the concerns raised by the staff with regard to its proposed revision to the SG TS/TRM, its schedule for submitting its generic SG TS change request (August 17, 1999), will be delayed. NEl will provide the staff with its revised schedule after the July 9,1999, telecon.

As indicated above, a telecon will be conducted on July 9,1999, at 1:00 p.m., between NRC and industry to discuss further the open technical and regulatory issues.

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