ML18039A787

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Technical Evaluation Rept on Second 10-Year Interval Inservice Insp Program Plan, for Browns Ferry Nuclear Plant,Unit 3
ML18039A787
Person / Time
Site: Browns Ferry Tennessee Valley Authority icon.png
Issue date: 01/31/1999
From: Mary Anderson, Charles Brown, Galbraith S
IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:
NRC (Affiliation Not Assigned)
Shared Package
ML18039A783 List:
References
CON-FIN-J-2229 INEEL-EXT-98-01, INEEL-EXT-98-01184, INEEL-EXT-98-1, INEEL-EXT-98-1184, NUDOCS 9905250309
Download: ML18039A787 (46)
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INEEL/EXT-98-01 1 84 Technical Evaluation Report on the Second 10-Year Interval Inservice Inspection Program Plan:

Tennessee Valley Authority, Browns Ferry Nuclear Plant, Unit 3, Docket Number 50-296 M. T. Anderson, C. T. Brown, S. G. Galbraith, A. M. Porter Published January 1999 Idaho National Engineering and Environmental Laboratory Materials Physics Department Lockheed Martin Idaho Technologies Company Idaho Falls, Idaho 83415 990S2S0309 99052i PDR ADQCK 05000296 P PDR j Prepared for the Division of Engineering Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 JCN No. J2229 (Task Order A23)

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ABSTRACT This. report presents the. results of the evaluation of the Second Ten-Year Inspection Interval Inservice Inspection Program for Browns Ferry Nuclear Plant, Unit 3, Revision 0, submitted by letter dated January 22, 1997, including the requests for relief from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, requirements that. the licensee has determined to be impractical. The. Second Ten-Year Inspection Interval Inservice Inspection Program for Browns Ferry Nuclear Plant, Revision 0, is evaluated in Section 2 of this report. The inservice inspection (ISI) plan is evaluated for (a) compliance with the appropriate edition/addenda of Section XI, (b) acceptability of examination sample, (c) correctness of the application of system or component examination exclusion criteria, and (d) compliance with ISI-related commitments identified during previous Nuclear Regulatory Commission reviews. The requests for relief are. evaluated in Section 3 of this report.

This work was funded under:

U.S. Nuclear Regulatory Commission JCN No. J2229, Task Order A23 Technical Assistance in Support of the NRC Inservice Inspection Program

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SUMMARY

The licensee, Tennessee Valley Authority, has prepared the Second Ten-Year Inspection Interval lnservice Inspection Program for Browns Ferry Nuclear Plant, Unit 3, Revision 0, to meet the requirements of the 1989 Edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI. The second 10-year interval began November 19, 1996.

The information in the Second Ten-Year Inspection Interval Inservice Inspection Program for Browns Ferry Nuclear Plant, Unit 3, Revision 0, submitted by letter dated January 22, 1997, was reviewed. Included in, the. review were the requests for relief from the ASME Code Section XI requirements that the licensee has determined to be impractical. As a

.result of this review, a request for additional information (RAI) was prepared describing the information and/or clarification required from the licensee in order to complete the review.

The licensee provided the requested information in a submittal dated October 29, 1998..

Based on the review of the Program Plan, the licensee's response to the Nuclear Regulatory Commission's RAI, and the recommendations for granting relief from the ISI examinations that cannot be performed to the extent required by Section XI of the ASME Code, no deviations from regulatory requirements or commitments were identified in the Second Ten-Year Inspection Interval Inservice Inspection Program for Browns Ferry Nuclear Plant, Unit 3, Revision 0.

ik CONTENTS ABSTRACT .. ~ ~ ~ ~ ~ ~ ~ ~ ~ II

SUMMARY

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ III 1 ~ INTRODUCTION ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1

2. EVALUATIONOF INSERVICE INSPECTION PROGRAM PLAN ............. ~.... 3 2.1 Documents Evaluated... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3 2.2 Compliance with Code Requirements ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3

,2.2.1 Compliance with Applicable Code Editions ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3 2.2.2 Acceptability of the Examination Sample . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 4 2.2.3 Exemption Criteria . ~ . 5 2.2.4 Augmented Examination Commitments .. ... 5 2.3 Conclusion ... 6

3. EVALUATIONOF RELIEF REQUESTS ~ ~

3.1 Class 1 Components .. ~ ~ ~ e 7 3.1.1 Reactor Pressure Vessel ~ ~ ~ ~ 7 3.1.1.1 Request for Relief 3-ISI-1,Section XI, IWB-2420(b),

Successive Inspections When Flaw or Relevant Conditions are Evaluated and Qualify for Continued Service........ ~ ~ ~ ~ 7 3.1.2 Pressurizer (Not applicable) ~ ~ ~ ~ 7 3.1.3 Heat Exchapgers and Steam Generators ~ ~ ~ ~ 7 3.1.4 Piping Pressure Boundary ....7 3.1.5 Pump Pressure Boundary................ ~ ~ ~ 7 3.1.6 Valve Pressure Boundary ~ ~ ~ 7

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3 .1.7 General ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 7 3.1.7.1 Request for Relief 3-ISI-4, Examination Categories B-K-1 and C-C, Items B10.10 and C3.20, Integrally Welded Attachments ~ ~ ~ 7 3.1.7.2 Request for Relief 3-ISI-5, Approval to Implement Alternatives Contained in Code Case N-547, Alternative Requirements for Pressure Retaining Bolting of Control Rod Drive (CRDJ Housings ..8 3.2 Class 2 Components .. ~... .. ~ ~...... ~....... ~ . ~... ~ .. ~..... ..8 3.3 Class 3 Components . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ . ~ 8 3.4 Pressure Tests..... ..8 3.4.1 Class 1 System Pressure Tests ..8 3.4.2 Class 2 System Pressure Tests ..8 3.4.3 Class 3 System Pressure Tests ..8 3.4.4 General ~........... ..8 3.4.4.1 Request for Relief 3-SPT-1, Alternative Pressure Test for Welded Repairs or Replacements in Class 1, 2, and 3 S ystems ............... ~ ~ . ~ .................... 8 IV

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3.4.4.2 Request fo'r Relief 3-SPT-2, 10-Year Hydrostatic Test Requirements for Code Class 1, 2, and 3 Systems . ~.... ~... 9 3 .5 General ~ ~ ~ 9 3.5.1 Ultrasonic Examination Techniques ... 9 3.5.2 Exempted Components .............. ... 9 3.5.3 Other . ~ . 9 3.5.3.1 Request for Relief 3-ISI-2, IWF-5300, Inservice Examination and Test Requirements for Snubbers............ ~.... ~ ~ ~ 9 3.5.3.2 Request for Relief 3-ISI-3, Use of Code Case N-524, Alternative Examination Requirements for Longitudinal Weldsin Class 1 and 2 Piping ... 9 3.5.3.3 Request for Relief 3-ISI-6, Use of Code Case N-532, Alternative Requirements to Repair and Replacement Documentation Requirements and Inservice Summary Report Preparation and Submission as Required by IWA-4000 and IWA-6000.... ~ . 11 3.5.3.4 Request for Relief 3-SPT-3,Use of Code Case N-546, Alternative Requirements for Qualification of VT-2 Visual Examination Personnel .. ~ 13 3.5.3.5 Request for Relief 3-SPT-4, IWA-5250(a)(2), Corrective Action Resulting from Leakage at Bolted Connections .. ~ ~ . ~ ~ 13 3.5.3.6 Request for Relief 3-SPT-5, Table IWC-2500-1, Examination Category C-H, Items C7.10, C7.30, C7.50, and C7.70, Pressure-Retaining Components . ~... . ~ 13 4s CONCLUSION ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 4

5. REFERENCES ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 5

0 TECHNICAL EVALUATIONREPORT ON THE SECOND 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN:

TENNESSEE VALLEYAUTHORITY, BROWNS FERRY NUCLEAR PLANT, UNIT 3 DOCKET'UMBER 50-296

1. INTRODUCTION Throughout the service life of a water-cooled nuclear power facility, 10 CFR 50.55a(g)(4) (Reference 1) requires that components (including supports) that are classified as American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Class 1, Class 2, and Class 3 meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, (Reference
2) to the extent practical within the limitations of design, geometry, and materials of construction of the components. This section of the regulations also requires that inservice examinations of components and system pressure tests conducted during successive 120-month inspection intervals comply with the requirements in the latest edition and addenda of the Code incorporated by reference in 10 CFR 50.55a(b) on the date 12 months prior to the start of the 120-month inspection interval, subject to the limitations and modifications listed therein. The components (including supports) may meet requirements set forth in subsequent editions and addenda of this Code that are incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein, and subject to Nuclear Regulatory Commission (NRC) approval. The licensee, Tennessee, Valley Authority (TVA), has-prepared the Second Ten-Year Inspection Interval Inservice Inspection Program for Browns Ferry Nuclear Plant, Unit 3, Revision 0, (Reference
3) to meet the requirements of the 1989 Edition of the ASME Code,Section XI, except that the extent of examination for Class 1, Examination Category B-J, has been determined by the requirements of the 1974 Edition through Summer 1975 Addenda (74S75) as permitted by 10 CFR 50.55a(b). The second 10-year interval began November 19, 1996.

Pursuant to 10 CFR 50.55a(a)(3), proposed alternatives to the Code requirements may be used when authorized by the NRC. The licensee must demonstrate either that the proposed alternatives provide an acceptable level of quality and safety, or that Code compliance would result in hardship or unusual difficulty without a compensating increase in safety. Pursuant to 10 CFR 50.55a(g)(5)(iii), if the licensee determines that conformance with certain Code examination requirements is impractical for its facility, the licensee shall submit information to the NRC to support that determination. Pursuant to 10 CFR 50.55a(g)(6)(i), the NRC will evaluate the licensee's determination that Code requirements are impractical. The NRC may grant relief and may impose alternative

0 requirements that it determines to be authorized by law, will not endanger life, property, or the common defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.

The information in the Second Ten-Year Inspection Interval Inservice Inspection Program for Browns Ferry Nuclear Plant, Unit 3, Revision 0, submitted by letter dated January 22, 1997, was reviewed, including the requests for relief from the ASME Code Section XI requirements that the licensee has determined to be impractical This review

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was performed using the standard review plans of NUREG-0800, Section 5.2.4, "Reactor Coolant Boundary Inservice Inspections and Testing," and Section 6.6, "Inservice Inspection of Class 2 and 3 Components" (Reference 4).

In letter dated October 6, 1997 (Reference'5), the NRC requested additional information that was necessary to complete the review of the inservice inspection (ISI) program plan, The requested information was provided by the licensee in "Response to Request for Information.- Inservice Inspection Program Plan", dated October 29, 1998 (Reference 6) ~

In this response, Tennessee Valley Authority provided the requested information, withdrew four requests for relief, and revised several requests for relief.

The Second Ten-Year Inspection Interval Inservice Inspection Plan for Browns Ferry Nuclear Plant, Unit 3, is evaluated in Section 2 of this report. The ISI program plan is evaluated for (a) compliance with the appropriate edition/addenda of Section XI, (b) acceptability of examination sample, (c) correctness of the application of system or component examination exclusion criteria, and (d) compliance with ISI-related commitments identified during the NRC's previous reviews. The requests for relief are evaluated in Section 3 of this report. Unless otherwise stated, references to the Code refer to the ASME Code,Section XI, 1989 Edition. Inservice test programs for pumps and valves and for snubbers are being evaluated in other reports.

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2. EVALUATIONOF INSERVICE INSPECTION PROGRAM PLAN This evaluation consists of a review of the applicable program documents to determine whether or not they are in compliance with the Code requirements and any previous license conditions pertinent to ISI activities. This section describes the submittals reviewed and the results of the review.

2.1 Documents Evaluated Review has been completed on the following information from the licensee:

Second Ten-Year Inspection Interval Inservice Inspection Program for,Browns Ferry Nuclear Plant, Unit 3, Revision 0, dated January 22, 1997 (Reference 3).

Licensee's "Response to Request for Additional Information - Inservice Inspection Program", dated October 29, 1998 (Reference 6).

2.2 Compliance with Code Requirements 2.2.1 Compliance with Applicable Code Editions Inservice inspection program plans are to be based on Section XI of the ASME Code editions defined in 10 CFR 50.55a(g)(4) and 10 CFR 50.55a(b). The second interval at Browns Ferry Nuclear Plant, Unit 3, began November 19, 1996; therefore, the Code applicable to the second interval ISI program is the 1989 Edition. As stated in Section 1 of this report, the licensee has prepared the Second Ten-Year Inspection Interval lnservice Inspection Program for Brown Ferry Nuclear Plant, Unit 3, to meet the requirements of 1989 Edition of the ASME Code, except that the extent of examination for Class 1, Examination Category B-J, has been determined by the requirements of the 1974 Edition through Summer 1975 Addenda as permitted by 10 CFR 50.55a(b).

In accordance with 10 CFR 50.55a(c)(3), 10 CFR 50.55a(d)(2), and 10 CFR 50.55a(e)(2), ASME Code cases may be used as alternatives to Code requirements. Code cases that the NRC has approved for use are listed in Regulatory Guide 1 147, Inservice Inspection Code Case Acceptability, (Reference 7) with any

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additional conditions the NRC may have imposed. When used, these Code cases must be implemented in their entirety. Published Code cases awaiting approval and subsequent listing in Regulatory Guide 1 147 may be adopted only if the licensee requests, and the

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NRC authorizes, their use on a case-by-case basis.

The licensee's second 10-year ISI program includes the Code cases listed below.

These Code cases either have been approved for use in Regulatory Guide 1.147 or are included as requests for relief.

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Code Case N-307-1 Revised Ultrasonic Examination Volume for Class 1 Bolting, Table IWB-2500- 1, Examination Category B-G- 1, When the Examinations Conducted From the Center-Drilled Hole

're Code Case N-435-1 Alternative Examination Requirements for Vessels with Wall Thickness 2in. or Less Code Case N-41'6-1 Alternative Pressure Test Requirement for Welded Repairs or Installation of Replacement Items by Welding, Class 1, 2, and 3 (Relief Request 3-SPT-1, was authorized for use by NRC Safety Evaluation Report dated March 10, 1997)

Code Case N-457 Qualification Specification Notch Location for Ultrasonic Examination of Bolts and Studs

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Code Case N-460 Alternative Examination Coverage for Class 1 and 2 Welds Code Case N-461 Alternative Rules for Piping Calibration Block Thickness Code Case N-491 Alternative Rules for the Examination of Class 1, 2, and 3 and MC Components and Supports of Light Water Cooled Power Plants Code Case N-503 Limited Certification of Nondestructive Examination Personnel Code Case N-524 Alternative Examination Requirements for Longitudinal Weldsin Class 1 and 2 Piping (Evaluated in Request for Relief 3-ISI-"',

Code Case N-532 Alternative Requirements to Repair and Replacement Documentation Requirements and Inservice Summary Report Preparation and Submission as Required by IWA-4000 and IWA-8000 (Evaluated in Request for Relief 3-ISI-6)

Code Case N-546 Alternative Requirements for Qualification of VT-2 Examination Personnel (Request for Relief 3-SPT-3 was authorized for use by NRC Safety Evaluation Report dated September 15, 1998) 2.2.2 Acceptability of the Examination Sample Inservice volumetric, surface, and visual examinations shall be performed on ASME Code Class 1, 2, and 3 components and their supports using sampling schedules described in Section XI of the ASME Code and 10 CFR 50.55a(b). Sample size and weld selection procedures have been implemented in accordance with the Code and 10 CFR 50.55a(b) and appear to be correct.

0 2.2.3 Exemption Criteria The criteria used to exempt components from examination shall be consistent with Paragraphs IWB-1220, IWC-1220, IWC-1230, IWD-1220, and 10-CFR 50.55a(b). The exemption criteria have been applied by the licensee in accordance with the Code, as discussed in the ISI program plan, and appear to be correct.

2.2.4 Augmented Examination Commitments In addition to the requirements specified in Section XI of the ASME Code, the licensee has committed to perform the following augmented examinations:

~ Examination of feedwater nozzles to the requirements of NUREG-0619, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking (Reference 8).

Examination of CRD return line reroute per NUREG-0619 (ultrasonic examination of welded connections joining the rerouted CRD return line to the reactor water cleanup system for three successive refueling outages).

Examination of austenitic stainless steel and dissimilar metal welds susceptible to IGSCC to the requirements of Generic Letter 88-01, NRC Position on Intergranular Stress Corrosion Cracking (/GSCCj in BWR Austenitic Stainless Steel Piping, and NUREG-0313, Rev. 2, Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping. All accessible welds will be examined in accordance with GL 88-01 and NUREG-0313, Rev. 2 commitments.

(Reference 9)

~ Visual inspection of the Core Spray spargers and piping inside the RPV each refueling outage to the recommendations of SIL No.289.

~ Visual inspection of the Core Spray sparger indications documented in NOI-CSB-057 and NOI-C5B-056.

~ Visual inspection of the Core Spray T-box front cover plate weld and T-box-to-thermal sleeve weld joint to the recommendations of SIL No. 289.

~ Visual or ultrasonic inspection of the core shroud per SIL No.572.

~ Ultrasonic examination of the shroud support access. hole covers per SIL No.462.

~ Ultrasonic examinations of the jet pump hold down beams per SIL No.330 and NUREG CR-3052, BWR Jet Pump Assembly Failure (Reference 10).

~ Visual examination of jet pump sensing lines per SIL No. 420.

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~ Visual inspection of jet pump throats per SIL No. 465.

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~ Visual inspection of jet pump riser brace per SIL No. 551 and jet pump riser brace clamp per GE letter No. BFSE 93-.143.

~ Visual inspection of jet pump adjusting screw per SIL No. 574.

~ Visual inspection of the top guide per SIL No. 554 and GE letter Nos. BFSE 94-001 and BFSE 94-002.

~, 'Visual examination of all accessible areas of the intermediate range monitor (IRM) and source range monitor (SRM) dry tubes per SIL,No. 409.

~ Visual inspection of steam dryer drain channel welds per GE'letter No.

BFSE 94-002.

~ Inspection of instrument nozzle safe ends and core differential pressure/standby liquid control nozzle safe end per SIL No. 571.

~ Visual inspection of the top guide and core plate per Sli No. 588.

~ Ultra'sonic examination of the core spray and recirculation inlet safe ends per GE letter No. BFSE 94-007.

~ Ultrasonic examinations on selected circumferential pipe welds to provide additional protection against pipe whip per Technical Specification Surveillance Requirement 4.6.G.

2.3 Conclusion Based on the review of the documents listed in Section 2.1, no deviations from regulatory requirements or commitments were identified in the Second'Ten-Year Inspection Inspection Program for Browns Ferry Nuclear Plant, Unit 3, Revision 0. 'nterva//nservice

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3. EVALUATIONOF RELIEF REQUESTS The requests for relief from the ASME Code requirements that the licensee has determined to be impractical for the second 10-year inspection interval are evaluated in the following sections.

3.1 Class 1 Components 3.1.1 Reactor Pressure Vessel 3.1.1.1 Request for Relief 3-ISI-1,Section XI, IWB-2420(b), Successive Inspections When Flaw or Relevant Conditions are Evaluated and Qualify for Continued Service Note: Request for Relief 3-ISI-1 was previously evaluated and denied in an NRC SER dated June 11, 1997.

3.1.2 Pressurizer (Not applicable) 3.1.3 Heat Exchangers and Steam Generators No relief requests.

3.1.4 Piping. Pressure Boundary No relief requests.

3.1.5 Pump Pressure Boundary No relief requests.

3.1.6 Valve Pressure Boundary No relief requests.

3.1.7 General 3.1.7.1'equest for Relief 3-ISI-4, Examination Categories B-K-1 and C-C, Items B10.10 and C3.20, Integrally Welded Attachments Note: In response to an NRC RAI question, Request for Relief 3-ISI-4 was withdrawn by the licensee in the October 29, 1998 submittal.

0 3.1.7.2 Request for Relief 3-ISI-5, Approval to Implement Alternatives Contained in Code Case N-547, Alternative Requirements for Pressure Retaining Bolting of Control Rod Drive (CRD) Housings Note: In response to an NRC RAI question, Request for Relief 3-ISI-5 was withdrawn by the licensee in the October 29, 1998 submittal.

3.2 Class 2 Components No relief requests.

3.3 Class 3 Components No relief requests.

3.4 Pressure Tests 3.4.'1 Class 1 System. Pressure Tests No relief requests.

3.4.2 Class 2 System Pressure Tests No relief requests.

3.4.3 Class 3 System Pressure Tests No relief requests.

3.4.4 General 3A.4.1 Request for Relief 3-SPT-1, Alternative Pressure Test for Welded Repairs or Replacements in Class 1, 2, and 3 Systems Note: Request for Relief 3-SPT-1 was evaluated and authorized in an NRC SER dated March 10, 1997.

IQ 3A.4.2 Request for Relief 3-SPT-2, 10-Year Hydrostatic Test Requirements for Code Class 1, 2, and 3 Systems Note: In response to an NRC RAI question, Request for Relief 3-SPT-2 was withdrawn by the licensee in the October 29, 1998 submittal.

3.5 General 3;5.1 Ultrasonic Examination Techniques No relief requests.

3.5.2 'Exempted Components No relief requests.

3.5.3 Other 3.5.3.1 Request for Relief 3-ISI-2, IWF-5300, Inservice Examination and Test Requirements for Snubbers Note: Request for Relief 3-ISI-2 is considered part of the Inservice Test (IST) Program and, therefore, is not included in this evaluation. It will be evaluated by the Mechanical Engineering Branch of the NRC.

3.5.3.2 Request for Relief 3-1SI-3, Use of Code Case N-524, Alternative Examination Requirements for Longitudinal Weldsin Class 1 and 2 Piping Code Requirement Table IWB-2500-1, Examination Category B-J, Item B9.12 requires 100% surface and volumetric examinations of longitudinal piping welds in Class 1 piping 4-inch nominal pipe size and larger to be performed in conjunction with examination of the circumferential welds selected for examination, as defined in Figure IWB-2500-8. The length of longitudinal weld required to be examined is at least one pipe diameter, but not more than 12 inches, from the circumferential weld intersection point.

Examination Categories C-F-1 and C-F-2, Items C5.12, C5.22, C5.52, and C5.62 require 100% surface and/or volumetric examinations of longitudinal piping welds in Class 2 piping to be performed in conjunction with examination of the circumferential welds selected for examination, as defined in Figure IWC-2500-7, for at least 2.5t of each longitudinal weld at the circumferential weld intersection. For Items C5.42 and C5.82, a surface examination is required for longitudinal piping welds intersecting circumferential welds selected for examination, as defined in Figure IWC-2500-7.

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Licensee's Proposed Alternative-In accordance with 10 CFR 50.55a(a)(3)(i), the licensee

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proposed an alternative to the volumetric and/or surface examination for the length of longitudinal piping welds required to be examined in accordance with Tables IWB-2500 and IWC-2500. The licensee stated:

"As an alternative to the requirements of the 1989 Edition (no addenda) of ASME Section XI, Browns Ferry Nuclear Plant will adopt the provisions of ASME Code Case N-524 for the examination of Class 1 and 2 longitudinal piping welds."

Licensee's Basis for the Proposed Alternative (as stated)

"The alternative examination requirements of Code Case N-524 include examination of the subject longitudinal piping welds at intersecting circumferential welds within the examination boundary of the circumferential weld. The following items support the basis for the revised longitudinal piping weld examination boundary:

~ "Longitudinal Piping welds are fabricated during the manufacturing process under controlled conditions, which produce higher quality welds and more uniform residual stress patterns.

~ "Longitudinal piping welds undergo heat treatment during the manufacturing process which enhances the material properties of the weld and reduces the residual stress created by welding.

~ "Results of previous weld inspections indicate that longitudinal welds have not been a safety concern, and there has been no evidence of longitudinal weld defects compromising safety at nuclear power plants.

~ "Longitudinal welds have not been shown to be susceptible to any particular degradation mechanism

~ "The areas of a longitudinal weld that would be the most susceptible to a potential failure mechanism would be at the intersection of a field fabricated circumferential weld.

~ "Locating longitudinal piping welds can require acid etching, eddy current examination, or a combination of methods. This increases radiological exposure, radwaste generation, and overall cost for performance of ASME Section XI examinations."

Evaluation ASME Section XI requires the examination of one pipe diameter, but not more than 12 inches, of Class 1 longitudinal piping welds. For Class 2 piping welds, the length of longitudinal weld required to be examined is 2.5 times the pipe thickness. These lengths are measured from the intersection with the circumferential weld. The licensee's proposed alternative is to examine only the portions of longitudinal weld within the examination area of'the intersecting circumferential weld in accordance with Code Case N-524, Alternative Examination. Requirements for Longitudinal Weldsin Class 1 and Class 2 Piping.

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ik Longitudinal welds are produced during the manufacture of the piping, not in the field as

~ is the case for circumferential welds. Consequently, longitudinal welds are fabricated under strict manufacturing standards, which provide assurance of structural integrity.

These welds have also been subjected to the preservice and initial inservice examinations, which provide additional assurance of structural integrity.

No significant loading conditions or material degradation mechanisms have been identified to date that specifically relate to longitudinal seam welds in Class 1 and 2 nuclear plant piping. The most critical region of the longitudinal weld is the portion that intersects the circumferential weld. If degradation associated with a longitudinal weld were to occur, it is expected. that it would be located at the intersection with a circumferential weld.

Since this region will be examined during the examination of the circumferential weld, the licensee's alternative provides an acceptable level of quality and safety.

Conclusion Based on the evaluation above, it is concluded that the use of Code Case N-524 provides an acceptable level of quality and safety. Therefore, it is recommended that the licensee's proposed alternative, to use Code Case'N-524, be authorized pursuant to 10 CFR 50.55a(a)(3)(i). The use of this Code Case should be authorized for the second 10-year interval at Browns Ferry, Unit 3, or until the Code Case is approved for general use by reference in Regulatory Guide 1.147. After that time, the licensee must follow the conditions, if any, specified in the regulatory guide.

3.5.3.3 Request for Relief 3-ISI-6, Use of Code Case N-532, Alternative Requirements to Repair and Replacement Documentation. Requirements and lnservice Summary Report Preparation and Submission as Required by IWA4000 and lWA-6000 Code Requirement Section XI, Paragraph IWA-6220 requires that the licensee prepare reports using NIS-1, Owner's Report for Inservice inspections, and NIS-2, Owner's Report for Repair or Replacements; IWA-6230 requires that these reports be filed with the enforcement and regulatory authorities having jurisdiction at the plant site within 90 days of the completion of the inservice inspection conducted during each refueling outage.

licensee's Proposed Alternative-Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposes to implement Code Case N-532. The licensee stated:

"For BFN Unit 3 TVA will invoke the requirements of ASME Code Case N-532,

'Alternative Requirements To Repair and Replacement Documentation Requirements and Inservice Summary Report Preparation and Submission as Required by IWA-4000 and IWA-6000;Section XI, Division 1,'s an alternative to the ASME'Section XI requirements listed above."

licensee's Basis for the Proposed Alternative (as stated)

"The 1989 Edition of ASME Section XI requires the preparation and submittal of NIS-1 and NIS-2 Summary Reports within 90 days following the completion of each refueling outage. This requires resources to be diverted from restart of the applicable unit to 11

preparation of the NIS-1 and NIS-2 Summary Reports within the. prescribed time limit.

It also presents an unnecessary time constraint on the licensee, for submittal of the summary reports to the NRC.

"Code Case.N-532 allows an alternate method for the certification of repairs and replacements in conjunction with the preparation and submittal of the Owner's Activity Report (OAR). Code Case N-532 requires submittal of the OAR to NRC at the end of the inspection period, which allows flexibility of resources at the end of a refueling outage. The information required for the OAR-1, 'Owner's Activity Report'nd Form NIS-2A, 'Repair/Replacement Certification Record,'rovide the same level of assurance and third party certification as the corresponding, NIS-1, 'Owner's Report for Inservice Inspections'nd NIS-2, 'Owner's Report for Repairs and Replacements'rom the 1989 Edition of ASME Section XI. Code Case N-532 requires that additional information be provided at the end of each inspection period regarding the percentage of examinations completed, which assures both the Owner and NRC that ASME Section XI Code compliance has been achieved.

"TVA has concluded that Code Case N-532 provides an equal or superior degree of information related to repairs/replacements and ISI examinations completed when compared to the NIS-1 and NIS-2 required by the 1989 Edition of ASME Section XI."

Evaluation The use of Form NIS-1, Owner's Report For Inservice Inspections, and Form NIS-2, Owner's Report for Repairs or Replacements, and submittal of an inservice inspection summary report within 90 days of the completion of the inservice inspection conducted during refueling outage are Code requirements. Alternatives contained in Code Case N-532 allow the licensee to submit these records in an abstract format on Font[

NIS-2A, Repair/Replacement Certification Record, and Form OAR-1, Owner's Activity Report, following the completion of an inspection period.

The requirements associated with documentation of inservice examinations and repairs/replacements and the subsequent submittal of Forms NIS-1 and NIS-2 within 90 days following a refueling outage are administrative only. It is noted that repair and replacement documentation reviews and approvals by the Authorized Nuclear Inspector continue to be required by this Code Case and that the licensee is required to establish a Repair/Replacement Plan in accordance with IWA-6340 of the 1992 Edition of Section XI.

The licensee has implemented Inspection Program B of the Code. Under this program, examination schedules are satisfied on a "per period" basis. Considering the milestones associated with Inspection Program B, submittal of the results of examinations and an abstract of repairs/replacements on a periodic basis is a reasonable alternative. In addition, the INEEL staff believes that the forms contained in Code Case N-532, which provide a summary of the status of repairs/replacements and a more detailed status of examinations by period and'interval, are an improvement over report forms currently required by the Code. For example, OAR-1 includes the status of examinations credited for the period and percent credited to date for the interval, by. Examination Category. This type of information 12.

provides the regulatory authorities a more comprehensive report on the status of the

~

inservice inspection program.

Conclusion Considering that, the Code recording and reporting criteria are only administrative requirements, the INEEL staff believes that use of the alternatives to Code requirements contained in Code Case N-532 will provide an acceptable level of quality and safety for Browns Ferry, Unit 3. Therefore, it is recommended that the licensee's proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i). The use of alternatives contained in Code Case N-532 should be authorized for the current interval or until the Code Case is published in a future revision of Regulatory Guide 1.147. At that time, if the licensee intends to continue to implement the alternatives of this Code Case, the licensee is to follow all provisions in Code Case N-532 with limitations issued in Regulatory Guide 1.147, if any.

3.5.3.4 Request for Relief 3-SPT-3;Use of Code Case N-546, Alternative Requirements for Qualification of VT-2 Visual Examination Personnel Note: Request for Relief 3-SPT-3 was evaluated and authorized in an NRC SER dated September 22, 1998.

3.5.3.5 Request for Relief 3-SPT-4, IWA-5250(a)(2), Corrective Action Resulting from Leakage at Bolted Connections Note: Request for Relief 3-SPT-4 was evaluated and authorized in an NRC SER dated September 22, 1998.

3.5.3.6 Request for Relief 3-SPT-5, Table IWC-2500-1, Examination Category C-H, Items C7.10, C7.30, C7.50, and C7.70, Pressure-Retaining Components Note: 'In response to an NRC RAI question, Request for Relief 3-SPT-5 was withdrawn by the licensee in the October 29, 1998 submittal.

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4. CONCLUSION Pursuant to 10 CFR 50.55a(a)(3), it is concluded that for Relief Request Nos. 3-ISI-3, 3-ISI-6, and 3-SPT-1, the licensee's proposed alternatives will (a) provide an acceptable level of quality and safety, or (b) Code compliance will result in hardship or unusual difficulty without a compensating increase in safety. It is recommended that these proposed alternatives be authorized.

Relief Request No. 3-ISI-1 was evaluated and denied in NRC SER dated June 11, 1997.

Relief Request No. 3-SPT-1 was evaluated and the proposed alternative authorized in NRC SER dated March 10, 1997. Relief Request Nos. 3-SPT-3 and 3-SPT-4 were evaluated and the proposed alternatives authorized in NRC SER dated September 15, 1998. Relief Request 3-ISI-2 will be evaluated in a separate report. by the Mechanical. Engineering Branch of the NRC.

By letter dated October 29, 1998, the licensee withdrew Relief Request Nos. 3-ISI-4, 3-ISI-5, 3-SPT-2, and 3-SPT-5 and deleted them from the ISI Program Plan.

This technical evaluation has not identified any practical method by which the licensee can meet all the specific inservice inspection requirements of Section XI of the ASME Code for the Browns Ferry Nuclear Plant, Unit 3. Compliance with all of the Section XI examination requirements would necessitate redesign of a significant number of plant systems, procurement of replacement components, installation of the new components, and performance of baseline examinations for these components. Even after the redesign efforts, complete compliance with the Section Xl.examination requirements probably could not be achieved. Therefore,'it is concluded that the public interest is not served by imposing provisions of Section XI of the ASME Code that have been determined to be impractical.

The licensee should continue to monitor the development of new or improved examination techniques. As improvements are achieved, the licensee should incorporate these techniques into the ISI program plan examination requirements.

Based on the review of the Second Ten-Year Inspection Interval Inservice Inspection Program. for, Browns Ferry Nuclear Plant, Unit 3, Revision 0, the licensee's response to the NRC's request for additional information, and the recommendations for granting relief from the ISI examinations that cannot be performed to the extent required by Section XI of the ASME Code, no deviations from regulatory requirements or commitments were identified.

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5. REFERENCES 1.. Code of Federal Regulations, Title 10, Part 50.
2. American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Division 1:

1989 Edition 1974 Edition, Summer 1975 Addenda

3. Second Ten-Year Inspection Interval Inservice Inspection Program for Browns Ferry Nuclear Plant, Unit 3, Revision 0, submitted January 22, 1997.
4. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Section 5.2.4, "Reactor Coolant Boundary Inservice Inspection and Testing," and Section 6.6, "Inservice Inspection of Class 2 and 3-Components,"

July 1981.

5. Letter dated October 6, 1997, J. F. Williams (NRC) to O. D. Kingsley (Tennessee Valley Authority) containing request for additional information.
6. Letter dated October 29, 1998, T. E. Abney tTennessee Valley Authority) to Document Control Desk (NRC), containing response to the NRC RAI dated October 6, 1997.
7. NRC Regulatory Guide 1 ~ 147, Inservice Inspection Code Case Acceptability,.

Revision 11, October 1994.

8. NUREG-0619, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking, November 1980.
9. NUREG-0313, Revision 2, Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping, January 1988.
10. NUREG CR-3052, BWR Jet Pump Assembly Failure, November 1984.

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NRC Form 335 U.S. Nuclear Regulatory Commission 1. REPORT NUMBER NPCM II02 (Assigned by NRC, Add VolSupp., Rcv., and 3201,3202 Addendum Numbers, ifany)

BIBLIOGRAPHICDATASHEET INE EL/EXT-98-01184

2. TITLE AND SUBTITLE 3. DATE REPORT PUBLISHED Technical Evaluation Report on the Second 10-year Interval Month Year Inservice Inspection Program Plan:

Tennessee Valley Authority, January 1999 Browns Ferry Nuclear Plant, Unit 3, Docket Number 50-296 4. FINORGRANTNUMBER JCN J2229 (Task Order A23)

5. AUTHOR(S) 6. TYPE OF REPORT M. T. Anderson Technical C. T. Brown S. G. Galbraith 7. PERIOD COVERED (Inclusive Dates)

A. M. Porter

8. PERFORMING ORGANIZATION- NAME AND ADDRESS (IfNRC, provide Division, Offic or Region, U.S. Nuclear Regulatory Commission, and mailing address; ifcontractor, provide name and mailing address)

Idaho National Engineering and Environmental Laboratory Materials Physics Lockheed Idaho Technologies Company Idaho Falls, Idaho 83415

9. SPONSORING ORGANIZATION- NAME AND ADDRESS (lfNRC, type "Same as above"; ifcontractor, provide NRC Division, OAicc or Region, U.S. Nuclear Regulatory Commission, and mailing address)

Civil and Geosciences Branch Division of Engineering Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

10. SUPPLEMENTARY NOTES
11. ABSTRACT (200 Words or less)

This report presents the results of the evaluation of the Second Ten-Year Inspection Interval Inservice Inspection Program for Browns Ferry Nuclear Plant, Unit 3; Revision 0, submitted by letter dated January 22, 1997, including the requests for relief from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, requirements that the licensee has determined to be impractical. The Second Ten-Year Inspection Interval Inservice Inspection Program for Browns Ferry Nuclear Plant, Revision 0, is evaluated in Section 2 of this report.

The inservice inspection (ISI) plan is evaluated for (a) compliance with the appropriate edition/addenda of Section XI, (b) acceptability of examination sample, (c) correctness of the application of system or component examination exclusion criteria, and (d) compliance. with ISI-related commitments identified during previous Nuclear Regulatory Commission reviews. The requests for relief are evaluated in Section 3 of this report.

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the 13. AVAILABILITY STATEMENT report)

Unlimited

14. SECURITY CLASSIFICATION (This page) Unclassified (This report) Unclassified
15. NUMBER OF PAGES
16. PRICE

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Browns Ferry Nuclear Plant, Unit 3 Page 1 of 2 Second 10-Year ISI Interval TABLE 1

SUMMARY

OF RELIEF REQUESTS

.,:. Relief;:,'";::.'

Re'q'uest:':::,: ':,:,",':;.':;.:;

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3-ISI-1 Reactor 8-A 81.10, Flaw or relevant conditions in RPV IW8-2420(b) Examine ar'eas containing Denied in SER dated Pressure Vessel 81.30 welds requires the subject indications June 11, 1997 successive once during the third examinations period of the second duiing the next interval.

three periods 3-ISI-2 Class 1, 2 and Snubbers Evaluated under a 3 separate report 3-ISI-3 Class 1 and 2 B-J, 89.12, Longitudinal Pipe Welds Surface and Code Case N-524 Authorized (a3i)

C-F-1, C5.1 2, Volumetric, as C-F-2 C5.22, applicable C5. 52, C5.62 3-ISI-4 Class 1 B-K-1, 810.10 Integrally Welded Attachments Surface or Perform the examination of Withdrawn C-C C3.20 Volumetric 8-K-1 support integrally welded attachments on accessible portion without removing support members.

3-ISI-5 Class 1 8-G-2 87.80 CRD Bolting VT-1 Visual Code Case N-547 Withdrawn 3-ISI-6 Class 1, 2 and N/A N/A NIS-1 and NIS-2 Summary Reports NIS-1 and NIS-2 Code Case N-532 Authorized (a3i) 3 forms 3-SPT-1 Class 1, 2 and N/A N/A Repair and Replacement Visual Code Case N-498-1 Authonzed in SER dated 3 March 10, 1997 3-SPT-2 Class 1, 2 and N/A N/A System Hydrostatic Testing Visual Code Case N-416-1 Withdrawn 3

3-SPT-3 Class 1, 2 and N/A N/A Visual Examiners VT-3 Visual Code Case N-546 Authorized in SER dated 3 September 22, 1998 3-SPT-4 Class 1, 2 and N/A N/A Leakage Corrective Action VT-2 Visual Code Case N-522 Authonzed in SER dated 3 September 22, 1998

~0 <5 Browns Ferry Nuclear Plant, Unit 3 Page 2 ol 2 Second 10-Year ISI Interval TABLE 1

SUMMARY

OF RELIEF REQUESTS

!,',;.",',"Relief!:

Iterlu'e'st! System.'or;':::,:"'::

,; Nunlbe'i;
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.;.: ,-',ReIIuiied Meth'od;'::i;:::;::i';:::::.:::;;i;::,:i';i;:AIteinatjve'::;::, .'-:.

3-SPT-5 Class 2 Piping C-H C7.10, Pressure Retaining Components VT-2 Visual during Code Case N-522 Withdrawn C7.30, System Pressure C7.50, Tests C7.70

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