ML20217K462

Technical Evaluation Rept on Second 10-Yr Interval ISI Program Plan:Tva,Sequoyah Nuclear Plant,Units 1 & 2

ML20217K462
Person / Time
Site:	Sequoyah
Issue date:	01/30/1998
From:	Mary Anderson, Beth Brown, Feige E IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:	NRC (Affiliation Not Assigned)
Shared Package
ML20217K423	List: ML20217K420 ML20217K447 ML20217K462
References
CON-FIN-J-2229 INEEL-EXT-97-01, INEEL-EXT-97-01185, INEEL-EXT-97-1, INEEL-EXT-97-1185, NUDOCS 9805010223
Download: ML20217K462 (47)

Text

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lNEEL/ EXT-97-01185 January 1998

/deho Technical Evaluation Report on the Nations /

Engineering Second 10-Year interval inservice Laboratory Inspect. ion Program Plan:

Tennessee Valley Authority, Sequoyah Nuclear Plant, Units 1 and 2, Docket Numbers 50-327 and 50-328 M. T. Anderson B. W. Brown E. J. Folge A. M. Porter A

s. a c x x u s o m a n ri ng 7885 iM M oE E527 G PDR

INEEL/ EXT-97-01185 Technical Evaluation Report on the Second 10-Year Interval inservice inspection Program Plan:

Tennessee Valley Authority, Sequoyah Nuclear Plant, Units 1 and 2, Docket Numbers 50-327 and 50-328 M. T. Anderson B. W. Brown E. J. Feige A. M. Porter Published January 1998 Idaho National Engineering and Environmental Laboratory Materials Physics Department Lockheed Martin Idaho Technologies Company Idaho Falls, Idaho 83415 Prepared for the Division of Engineering Office of Nuclear Reactor Regulation U.S. Nr.. War Regulatory Commission W 4hington, D.C. 20555 JCN No. J2229 (Task Order A13)

i ABSTRACT This report presents the results of the evaluation of the Sequoyah Nuclear Mant,

\' Units 1 and 2, Second 10-Year Intervalinservice inspection Program Man, submitted l November 21,1995, including the requests for relief from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, requirements that the licenses has determined to be % practical. The Sequoyah Nuclear Mant, Units 1 and 2, Second 10-Year Intervalinservice Inspection Program Man 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, 1 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.

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i This work was funded under:  ;

U.S. Nuclear Regulatory Commission l

. JCN No. L2229, Task Order A13 Technical Assistance in Support  !

of the NRC Inservice Inspection Program il

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SUMMARY

, The licensee, Tennessee Valley Authority, prepared the Sequoyah Nuclear Plant, Units 1 and 2, Second 10-Year Intervalinservice Inspection Program Plan to meet the requirements of the 1989 Edition of the American Society of Mechanical Engineers (ASME)

Boiler and Pressure Vessel Code, Section Xil, except that the extent of examination for Class 1, Examination Category B J, piping welds 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 IO year interval began December 16,1995.

Tha information in the Coquoyah Nuclear Plant, Units 1 and 2, Second 10-Year Interval Inservice Inspection Progran' Plan, submitted November 21,1995, was reviewed. The mview included requests for relief from the ASME Code Section XI requirements that the licansee has determined to t:e impractical. As a result of this review, a request for auditional information (RAl) was prepared describing the information and/or clarification required from the licensee in order to complete the review. Conference calls were held March 4,1996 and March 18,1996, to clarify information required from the licensee. The licensee provided the requested information in a letter dated May 9,1996. In addition to the response to the NRC's request, the licensee submitted two additional requests for relief. After review of this information,it was determined that further clarification was required and a second RAI was sent to the licensee in a letter dated July 9,1996. After two additional conference calls including the licensee and the NRC, a response to the second RAI was provided in a letter dated September 6,1996. This response included three additional requests for relief and one revised request, in addition to the schedule of examinations for Unit 2. After review of this information,it was determined that a third RAl was necessary to complete the review of the second ten-year program plan. This RAI was sent to the licensee by letter dated December 17,1996. The licensee provided the requested information in letters dated March 4,1997, and August 28,1997. The latter submittal contained documentation on the completion of eight TVA commitments contained in previous submittals.

Based on the review of the program plan, the licensee's responses to the Nuclear Regulatory Commission's RAls, 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 Sequoyah Nuclear Plant, Units 1 and 2 Second 10-Year IntervalInservice Inspection Program Plan, except as noted in the evaluation of Requests for Relief 1-1S11,21S11, 1-1512 (Part 2), 21512 (Part 2),1-ISI 3,21S1-3, and ISPT-03.

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CONTENTS

, A B ST R A CT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii

SUMMARY

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1. I NT R O D U CTI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

2. EVALUATION OF 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 ...................... 5 2.2.3 Exem ption Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2.2.4 Augmented Examination Commitments . . . . . . . . . . . . . . . . . . . . . . . . 5 2.3 Conclusion ................................................6

3. EVALUATION OF RELIEF REQUESTS .................................7 3.1 Class 1 Components .........................................7 3.1.1 Reactor Pressure Vessel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 3.1.1.1 Requests for Relief 1-ISl-1 and 21S1-1 Examination Category B A, item B1.30, Reactor Vessel Shell-to Flange Weld . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 3.1.1.2 Requests for Relief 1-1S12 and 21S12, (Part 1),

Examination Category B A, item B1.40, Reactor Vessel Closure Head Weld.To-Flange Weld . . . . . . . . . . . . . . 8 3.1.1.3 Requests for Relief 1 ISI 2, and 2 ISI 2 (Part 2),

IWB-2420(a) Successive inspections of Examination  !

Category B A, item B1.40, Reactor Vessel Closure Head Weld-To Flange Welds . . . . . . . . . . . . . . . . . . . . . . . 10 3.1.1.4 Requests for Relief 1151-6 and 2 ISI 6, Examination Categories B D and B-F, items B3.90, B3.100, and B5.10, Reactor Vessel Nozzle-to Vessel Welds, Nozzle Inside Radius Sections, and Nozzle-to-Safe End Welds . . . . . . . . . . 11 3.1. 2 Pre s suriz e r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 3.1.3 Heat Exchangers and Steam Generators . . . . . . . . . . . . . . . . . . . . . . 13 3.1.4 Piping Pressure Boundary . . . . .......................... 13 3.1.5 Pump Pressure Boundary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 3.1.6 Valve Pressure Boundary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 3.1. 7 G e ne r al . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 4 3.2 Class 2 Components ........................................14

. 3.3 Cla ss 3 Components . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 i

3.3.1 Pre ssure Vessels . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14

3. 3 . 2 Pi pin g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 4 f 3.3.2.1 Revised Requests for Relief 1-ISI 3 and 21S13, j IWD-1220.1 Class 3 Components Exempt iv

From Examination . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 3.3.2 Pumps

.............................................16 3.3.3 Vr'ves ............................................. {

3. 3.4 Je ne ral . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16. . . . . .

3.4 Pre s sure Test s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 3.4.1 Class 1 System Pressure Tests . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 3.4.1.1 Request for Relief ISPT-05, Rev 2, Table IWB-2500-1, Examination Category B P, Pressure Retaining C om p one nt s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 6 3.4.2 Class 2 System Pressure Tests . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 4 3.4.2.1 Request for Relief ISPT-04, Use of Code Case N 522 for Pressure Testing Class 2 Components at Containment }

Pe ne trati ons . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 3.4.3 Class 3 System Pressure Tests . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 3.4.4 G e ne ral . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 0 3.4.4.1 Request for Relief ISPT-01, Request for Authorization to Use ASME Code Case N 4981, Attemative Rules for 10-Year System Hydrostatic Testing for Class 1, 2, and 3 Systems,Section XI, Division 1 . . . . . . . . . . . . . . . . . . . . . 20 3.4.4.2 Request for Relief No. ISPT-02, Use of Code Case N-4161 Altemative Pressure Test Requirements for Welded Repairs or Installation of Replacement items by Welding,Section XI, Division 1 . . . . . . . . . . . . . . . . . . . 22 3.4.4.3 Request for Relief ISPT-03,IWA 5250(a)(2), System Pressure Test Corrective Measures . . . . . . . . . . . . . . . . . . . 24 3.4.4.4 Request for Relief ISPT 08, IWA 5250(a)(2), System Pressure Test Corrective Measures . . . . . . . . . . . . . . . . . . . 25 3 . 5 G e ne ra1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 6 3.5.1 Ultrasonic Examination Techniques . . . . . . . . . . . . . . . . . . . . . . . . . 26 3.5.2 Exempted Components . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 6 3.5.2.1 Requests for Relief 1-1S1-4 and 2 ISI 4, IWC-1221, IWC-1222(b) and IWD-1220.1, Class 2 and 3 Items Exempt From Examination . . . . . . . . . . . . . . . . . . . . . 26

3. 5 . 3 O the r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 6 3.5.3.1 Request for Relief 1 ISI-5 and 2 ISI 5, Use of Code Case N 509, Altemative Rules for the Selection and .

Examination of Class 1, 2, and 3 Integrelly Welded 1 j

A ttachmen ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 6 '

3.5.3.2 Requests for Relief 1-1517 and 2-ISI-7, Use of Code Case N-524 Altemative Examination Requirements For Longitudinal Welds in Class 1 and 2 Piping '

Section XI, Division 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 3.5.3.3 Request for 9elief No. ISPT-06, Use of Code Case N-546, Altemative Requirements for Oustification of VT-2 Visual Examination Personnel . . . . . . . . . . . . . . . . . 30 v

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4. C O N C LU S I O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 5

5. R E F E R E N C E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 7 l

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. o TECHNICAL EVALUATION REPORT ON THE SECOND 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN:

TENNESSEE VALLEY AUTHORITY, SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2, DOCKET NUMBERS 50-327 AND 50-328

1. INTRODUCTION Throughout the service life of a water-cooled nuclear power facility,its components (including supports) that are classified as American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Class 1,2, and 3 are required by 10 CFR 50.55a(g)(4) (Reference 1) to meet the requirements, except the design and access provisions and the preservice examination requirements, of 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 a successive 120-month inspection intervals comply with th'e requirements in the latest i 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 Sequoyah Nuclear Plant, Units 1 and 2, Second 10-Year IntervalInservice Inspection Program Plan (Reference 3) to meet the requirements of the 1989 Edition of the ASME Code, Section Xil, 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 December 16,1995.

Pursuant to 10 CFR 50.55a(s)(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 impractic61 for its facility, the licensee shall submit inforrnation 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 1

requirements that it determines to be authorized by law, will not endanger life, ptreerty, 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 Sequoyah Nuclear Plant, Units 1 and 2, Second 10-Year Interval inservice Inspection Program Plan, submitted November 21,1995, was reviewed, including the requests for relief from the ASME Code Section XI requirements that the licensee has determined to be impractical. This review 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 a letter dated February 14,1996 (Reference 5) the NRC requested additional information that was necessary to complete the review of the inservice inspection (ISI) program plan. Conference calls were held March 4,1996 and March 18,1996, to clarify information required from the licensee. The requested information was provided by the licensee via a letter dated May 9,1996 (Reference 6). In addition to the response to the NRC's request, the licensee submitted two additional requests for relief. After review of this information,it was determined that further clarification was required and a second RAI was sent to the licensee in a letter dated July 9,1996 (Reference 7). The licensee's response to this RAl, provided in a letter dated September 6,1996 (Reference 8), included three additional requests for relief and one revised request,in addition to the schedule of examinations for Unit 2. After review of this information, it was determined that a third RAI was necessary to complete the review of the second ten-year program plan. This RAI was sent to the licensee in a letter dated December 17,1996 (Reference 9). The licensee provided the requested information in letters dated March 4,1997 (Reference 10), and August 28,1997 (Reference 11). The latter submittal contained documentation on the completion of eight TVA commitments contained in Reference 10. This documentation included the schedule for examinations for Unit 1, and eight revised requests for relief.

The Sequoyah Nuclear Plant, Units 1 and 2, Second 10-Year Intervalinservice inspection Program Plan is evaluated in Section 2 of this report. The ISI program plan is evaluated for (a) compliance with the appropriate editionladdenda 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 snubbers and for pumps and velves are being evaluated in other reports.

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2. EVALUATION OF 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 submittels reviewed and the results of the review.

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

• ' Sequoyah Nuclear Mant, Units 1 and 2, Second 10-Year Intervalinservice Inspection Program Man, dated November 21,1995 (Reference 3)

Licensee's responses to requests for additional information: Sequoyah Nuclear Plant (SQN) - AdditionalInformation for Second 10-Year Interval- Inservice Inspection (ISI) Program Man, dated May 9,1996 (Reference 6),

September 6,1996 (Reference 8), and March 4,1997 (Reference 10); and Sequoyah Nuclear Plant (SON) - AdditionalInformation for Second Ten-Year Interval- Inservice Inspection (ISI) and Inservice Pressure Test (ISPT) Program Nans, dated August 28,1997 (Reference 11).

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 Sequoyah Nuclear Plant, Units 1 and 2 began December 16,1995: 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 prepar'ed the Sequoyah Nuclear Plant, Units 1 and 2, Second 10-Year intervalinservice Inspection Program Man to meet the requirements of 1989 Edition of the 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 12) with any additional conditions the NRC may have imposed. When used, these Code cases must be implemented in their entirety. The licenses may adopt an approved Cods case by providing written notification to the NRC. Published Code cases awaiting approval and subsequent listing in Regulatory Guide 1.147 may be adopted only if the licensee requests, and the NRC authorizes, their use on a case by case basis.

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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.

Request Code for Case Relief Title N 307-1 Revised Ultrasonic Examination Volume for Class 1 Bolting, Table IWB-2500-1, Examination Category B-G-1, When the Examinations Are Conducted From the Center-Drilled Hole N 457 Qualification Specimen Notch Location for Ultrasonic Examination of Bolts and Studs N 460 Altemative Examination Coverage for Class 1 and CIsss 2 Welds N-461 Altemative Rules for Piping Calibration Block Thickness N-463-1 Evolustion Procedures and Acceptance Criteria for Flows in Class 1 Ferritic Piping That Exceed the Acceptance Standards ofIWB-3514.2 N-4 81 Alternative Rules for Examination Requirements for Cast Austenitic Pump Casings (Unit 1 only)

N 491 Attemative Rules for Examination of Class 1, 2, 3, and MC Component Supports of Light Water Cooled Power Plants N-494 1 Pipe Specific Evaluation Procedures and Acceptance Criteria for Fisws in CIsss 1 Ferritic Piping That Exceed the Acceptance Standards ofIWB 3514.2 N-503 Limited Certification of Nondestructive Examination Personnel N 416-1 ISPT-02 Altemative Pressure Test Requirement for Welded Repairs or Installation of Replacement items by Welding N 4981 lSPT-01 Altemative Rules for 10-Year System Hydrostatic Testing for Class 1, 2, and 3 Systems N-509 1-ISl 5 Alternative Rules for the Selection and Examination of Class 1, 2, 2-lSi 5 and 3 Integrally Welded Attachments N 521 1ISI6 Altemative Rules for DeferralofInspections of Nozzle to Vessel 2-lSI 6 Welds, inside Medius Sections, and Nozzle-to Safe End Welds of a Pressurized Water Reactor (PWR) Vessel N 522 lSPT-04 Pressure Testing of Containment Penetration Piping N-524 1-ISI-7 Altemative Examination Requirements for Longitudinal Welds in 2-lSI 7 Class 1 and 2 Piping 4

Request Code for i

, Case Relief Title j N-533 ISPT-07 Altemative Requirements for VT-2 VisualExamination of Class t

, Insulated Pressure-Metsining Bolted Connections N-546 ISPT-06 Alternative Requirements for Qualification of VT-2 Examination Personnel I

I 2.2.2 Acceptability of the Examination Sample '

i Inservice volumetric, surface, and visual examinations shall be performed on ASME j Code Class 1,2, and 3 components and their' supports using sampling schedules described I 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) i and appear to be correct.

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2.2.3 Exemption Criteria l 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 the reactor pressure vesselin accordance with Regulatory Guide 1.150 (Reference 13)

Volumetric examination of the reactor coolant pump flywheel high stress areas every 3 years, as well as volumetric and surface examinations with the flywheel removed at 10 year intervals, satisfying NRC Regulatory Guide 1.14, Aesctor Coolant Pump FlywheelIntegrity (Reference 14)

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• Ultrasonic examination of steam generator feedwater norries per IE Bulletin 7g-13, Cracking in Feedwater System Piping (Reference 15)

Eddy current examination (100%), each refueling outage, of all reactor vessel in-core detector thimble tubes that are in service per IE Bulletin 88 09, Thimble Tube Thinning in Westinghouse Reactors (Reference 16) 5 ,

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Volumetric examination of the reactor pressure vessel nozzles in accordance with Section 3.0 of Attachment 10 to the Sequoyah Nuclear Plant, Units 1 and 2, Second 10-Year Intervalinservice Inspection Program Plan.

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 Sequoyah Nuclear Plant, Units 1 and 2. Second 10-YearIntervalinservice inspection Program Plan. Note that this report does not include a review of the implementation of the augmented examinations,it merely records that the licensee has committed to perform them, 6

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3. EVALUATION OF 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 Requests for Relief 1-1811 and 2-1811, Examination Category B A, item 81.30, Reactor Vessel Shell-to-Flange Weld i

Code Aequfrement--Examination Category B A, item B1.30 requires 100% volumetric examination of the reactor vessel shell-to-flange weld as defined in Figure IWB 2500-4.

Licensee's Code Aelie/Neguest--4n accordance with 10 CFR 50.55a(g)(5)(iii), the licensee requested classification of the reactor vessel shell-to flange weld as a reactor vessel shell weld.

Licensee's Desis for Requesting ReNef-

"The reactor vessel flange-to-upper shell weld is located behind the core barrel and is therefore inaccessible until the core barrel is removed. The vessel flange to-upper shell weld is 3g inches below the flange face. Due to the location of vessel flange-to-upper shell weld, TVA intends to classify the weld as a reactor vessel shell weld.

(Examination Category B-A, item No. B1.11)

"The Sequoyah Reactor Vessel Stress Report entitled Analysis of the Main Closure including Core Support Ledge (Document No. 30616-1105) has been reviewed to determine a fatigue usage factor for the vessel flange to shell weld. This analysis does not provide a usage factor specifically for the weld because the analysis considers weld and base material to be homogeneous and equal in elasticity,

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strength, and fatigue properties. Instead, the analysis provides usage factors at i critical locations.

"The maximum fatigue usage factor in the vessel in the vicinity of the flange to shell weld as found in the above analysis is 0.00662. This value can be conservatively used for the weld and is considered extremely low compared to the  ;

code allowed fatigue usage factor of 1.0.

"Due to the distance (3g") from the flange face to the flange-to-upper shell weld, limitations of present ultrasonic techniques, and the very low fatigue usage factor, the flange-to-upper shell weld should be classified as a reactor vessel shell weld."

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• Licensee's Proposed Attemative Examinetlen-

"A remote ultrasonic examination of the weld will be conducted from the vessel inside diameter near the end of the inspection interval."

Eve /uet/on-The Code requires volumetric examination of reactor vessel shell to-flange welds. One third of the examination is to be performed each period. The licensee proposed to re classify the shell to-flange weld as a shell weld This would allow deferral of this examination until the end of the interval.

The Code classifies this weld as a shell to flange weld. The Code has no provision for reclassification of welds. The INEEL staff believes that the subject weld is indeed a shell-to-flange weld and should be examined as one. -

The licensee seeks to defer the examination until the third period. This is allowed by the Code, provided that a partial examination is conducted from the flange face.

Conclus/on-The licensee's alternative, to reclassify the subject weld, is not in accordance with standard Section XI practices. The Code provides a method to defer examination of the shell to flange weld. Therefore, the INEEL staff recommends that the request to use the licensee's proposed alternative be denied.

3.1.1.2 Requests for Relief 1-ISl 2 and 2-ISl 2, (Part 1), Examination Category 8 A, item 31.d'J, Reactor Vessel Closure Head Weld To-Flange Wold Code Requ/rement-Examination Category B-A, item B1.40 requires 100% volumetric examination of the reactor vessel closure head-to-flange weld as defined in Figure IWB-2500-5.

Licensee's Code Relief Aeguest-4n accordance with 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from 100% volumetric examination of the reactor vessel closure head-to-flange weld.

Licensee's Basis for Roguesting MeWef-

"The design configuration of the closure head precludes full ultrasonic (UT) examination of the reactor pressure vessel (RPV) closure head to flange weld (WO8-09). Examination of the closure head to flange weld is obstructed by lifting lugs (three) and the tapered portion of the flange. The design configuration allows for ultrasonic testing only approximately 50% of the required volume of coverage for the RPV closure head to-flange weld.

"The design configuration of the closure head precludes ultrasonic (UT) examination from the flange side of the closure head to flange weld. In order to examine the welds in accordance with the requirement, tne reactor vessel would require extensive design modifications. TVA will perform a best-effort ultrasonic examination and a surface examination (MT) of essentially 100% of the closure head to-flange weld area in the third inspection period."

8

In the May 9,1996 response to the NRC's RAI the licensee provided the following information.

"The ASME Section XI code requirements for reflectors oriented parallel to the weld stipulate that the angle beam search units shall be aimed at right angles to the weld axis, with the search unit manipulated so that the ultrasonic beams pass throughout the entire volume of weld metal. The subject weld configuration limits bi-directional  !

coverage from the flange side due to the adjacent flange junction. This junction restricts the search unit scan surface.

" SON does not plan to perform a UT examination from the flange face because the ,

examination would not provide meaningful results, as stated below: '

1 "The geometric configuration of the flange to head weld is not amenable for ultrasonic examination from the flange face. This is due to the geometric curvature of the head and the extensive metal path distance required to interrogate the required weld volume.

"The flange face contains two o ring grooves (0.6 inch) in width around the circumference and contains 12 recessed locations for o-ring clips, which limit complete scan coverage from the flange face.

"The flange face and head are highly radioactive. Examination from the flange face would result in considerable personnel radiation exposure.

Radiation levels at the flange area ate estimated to be .5 roentgen equivalent man (rem) per hour. Total dose estimate for performing a zero degree from the flange face is estimated to be 1.2 rom, which is in addition to the exposure for performing the examination from the outside diameter of the weld surface. Localized shielding would be of no value due to the close contact necessary for manually scanning the flange surface."

Ucensee's Proposed Attemative Examination-

"TVA will perform a best-effort ultrasonic examination to achieve as much code coverage as possible and achieve meaningful results."

EveAustion-The Code requires volumetric examination of reactor vessel closure head welds. However, the geometric configuration of the flange to-head weld precludes complete examination and makes the Code-required volumetric examination impractical.

To obtain complete volumetric coverage, modifications or replacement of the component with one of a design providing for complete coverage would be required. Imposition of this requirement would cause a considerable burden on the licensee.

The licensee proposed to perform a partial examination, which would include approximately 50% of the Code required volume. Based on the volume that can be examined, it is reasonable to conclude that a pattern of degradation, if present, will be 9

o -

l

)

detected. Thus, reasonable assurance of continued inservice structuralintegrity will be '

provided.

Conclus/on-The design configuration makes the Code-required volumetric examination impractical. Based on the significant amount of weld coverage obtained, reasonable assurance of structuralintegrity will be provided. Therefore,it is recommended that relief be granted as requested, pursuant to 10 CFR 50.55a(g)(6)(i).

3.1.1.3 Requests for Relief 1-1812, and 21S12 (Part 2). lWB 2420(a) Successive inspections of Examination Category B A, item 31.40, Reactor Vessel Closure Head Weld To-Flange Wolds Code #egu/rement-lWB-2420(a) requires the sequence of component examinations established during the first inspection interval to be repeated during each successive inspection interval.

L/consee's Code Relief Roguest-in accordance with 10 CFR 50.55afg)(5)(iii), the licensee requested relief from following the sequence of examinations set during the first interval for the reactor vessel closure head to-flange weld.

Licensee's Basis for Requesting ReNef-

"In the first interval TVA elected to examine the closure head-to-flange weld (WO8-09) in three equal segments during the first, second, and third inspection periods respectively. In the second interval TVA will perform the entire circumference of the weld during one examination at the end of the interval. This will require the closure head shroud and accompanying insulation to be removed only one time. The closure head is highly radioactive and eliminating two shroud removals would reduce the overall personnel radiation exposure considerably.

Therefore, TVA request deferral of the volumetric and surface examination of the closure head-to-flange weld to the third inspection period of the second interval.  ;

"The closure head is highly radioactive and shroud removal is required to access the l closure head to-flange weld. Shroud removalis manpower intensive and eliminating two shroud removals would reduce personnel radiation exposure considerably.

Therefore, TVA request deferral of the examination of the closure head-to-flange weld to the third inspection period of the second interval."

Licensee's Proposed Afternative Exerninetion-

"TVA will perform a best effort ultrasonic examination to achieve as much code coverage as possible and achieve meaningful results. In addition, TVA will perform a surface examination (MT) of essentially 100% of the closure head-to flange weld area in the third inspection period."

Ere/uetion-The Code requires 100% volumetric examination of the reactor vessel closure head weld. One third of this examination is to te performed each period. The licensee proposed to defer the entire examination until the third period.

10

Deferring the examination until the third period would allow two-thirds of the weld to exceed 10 years between examinations. This is in conflict with the sampling philosophy of Section XI. Furthermore,later editions of the Code, through the 1995 Edition, do not support this change. Therefore, the INEEL staff believes that insufficient technical justification exists and that this request should be denied. The INEEL staff recommends that the licensee take this issue to the ASME Section XI Code Committee for resolution through the industry process.

Conclus/on- Based on insufficient technical justification, the INEEL staff recommends that this request be denied.

3.1.1.4 Requests for Relief 1-1S16 and 2-1816, Examination Categories B-D and B-F, items 83.90, B3.100, and 85.10, Reactor Vessel Nozzle to Vessel Welds, Nozzle faside Radius Sections, and Nozzle-to Safe End Welds Code Requirement-Examination Category B D, items B3.90 and B3.100 require 100%

volumetric examination, as defined by IWB-2500-7, of.at least 25% but not more than 50% of reactor vessel nozzle to-vessel welds and nozzle inside radius sections by the end of the first period.

Examination Category B F, item B5.10 requires 100% surface and volumetric examination, as defined by Figure IWB 2500 8, of dissimilar metal nozzle to safe end welds: these may be examined coincident with the vessel nozzle examinations.

Licensee's proposed Altemat/ve--Pursuant to 10 CFR 50.55ala)(3)(i), the licensee requested authorization to use Code Case N 521, Alternate Rules for Deferra/ of Inspections of Nozzle to-Vessel Welds, inside Radius Sections, and Nozzle-to-Safe End Welds of Pressurized Water Reactor (PWR) Vessel,Section XI, Division 1. The licensee states:

" SON will perform examination in accordance with the requirements of Code Case N 521."

Licensee's Basis for proposed Attemative-

"The RV nozzle-to-vessel welds and inside radius sections are ultrasonically examined from the RV inside diameter using automated examination devices. With I

the RV core barrelin place, the outlet nozzles are accessible for examination from the nozzle bore only. The inlet nozzles are inaccessible for examination until the core barrelis removed.

"During the first inspection period, the accessible volumes of the nozzle-to-vessel welds and inside radius sections on four outlet nozzles would be examined from the nozzle bore. The remaining examinations for the four inlet and four outlet nozzles would be completed in the third inspection period, to satisfy the requirements of Table IWB-25001, Examination Category B D, item Numbers 83.90 and 3.100.

11

"The RV nozzle to-safe end welds on the four outlet nozzles are ultrasonically examined from the inside diameter when the outlet nozzles are examined from the nozzle bore during the first inspection period. The four inlet nozzles would be examined in the third inspection period to satisfy the requirements of Table IWB-2500-1, Examination Category B-F, Item Number B5.10.

" Performing all required examination for the nozzle to-vessel welds, inside radius sections and nozzle to-safe end welds during the third inspection period in lieu of ,

the first and third inspection periods provides several benefits for SON: l "1. A reduction in the number of times an automated device and associated materials and equipment must be installed on RV. - -

l 1

"2. Compliance with ALARA Program

a. Reduction in Radweste

b. Reduction in Personnel Exposure "3. Use of same vendor, personnel, procedures and equipment to examine welds would enhance quality of examination results and reduce possibility for errors.

" SON examined the nozzle to-vessel welds and inside radius sections, and nozzle-to safe end welds of the four outlet nozzles in both the first and third inspection periods of the first inspection interval. These additional examinations were performed to maintain a ten year frequency.

" SON Units 1 and 2 satisfied the following conditions as required by the ASME Code Committee for use of Code Case N 521:

"(a) No inservice repairs or replacements by welding have ever been performed on any of the Nozzle-to-Vessel Welds,inside Radius Sections, or Nozzle to-Safe End Welds.

"(b) None of the Nozzle to-Vessel Welds, inside Radius Sections, or Nozzle to-Safe End Welds contains identified flaws or relevant conditions that currently required successive inspections in accordance with IWB 2420(b).

"(c) The unit is not in the first inspection interval."

EveAustion-The Code requires volumetric and surface examination of the subject nozzle-to-vessel welds, inside radius sections, and nozzle-to pipe welds during each 10 year ISI interval. At least 25%, but not more than 50% (credited), of the nozzle to-vessel welds and inside radius sections must be examined by the end of the first inspection period, and the balance completed by the and of the 10 year interval. The sequence of examinations established for the subject welds during the first inspection interval shall be repeated during each successive interval.

12  !

The licensee proposes to use the provisions in Code Case N-521: and defer the examinations to the end of the inspection interval.

The licensee examined the outlet nozzle-to vessel welds,insida radius sections, and j

outlet , nozzle-to pipe welds during the first period of the first 10 year interval to meet the l Code requirements. In addition, the licensee repeated the examination of these welds )

during the third period of the first interval. The subject welds will be reexamined during the third period of the third interval.

Paragraph IWB-2420(a), " Successive Inspections," states that the sequence of component examinations established in the first inspection interval shall be repeated during successive inspection intervals, to ths extent practical. Thus,' examinations are performed at intervals of not more than 10 years. The licensee reexamined the subject welds during the third period of the first interval. This reexamination of the outlet nozzle welds during the first interval established a new sequence of examinations for the Reactor Pressure Vessel. Since the subject welds were examined in the third period of the first interval, 10 years will not be exceeded if the examinations are deferred to the third period of the second interval. Therefore, this . schedule will provide an acceptable level of quality.

Conclus/on-The licensee's proposed alternative will provide an acceptable level of quality and safety since all conditions in the Code Case have been met, and there will be no more than 10 years between inspections, except where the length of a 10 year intervalis adjusted in accordance with IWA-2430. Therefore,it is recommended that the proposed 5

alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i). The use of afternatives l contained in Code Case N 521 should be authorized for the current interval or ur.;il such time as the Code Case is published in Regulatory Guide 1.147. At that time, if the license intends to contiriue to implement this Code Case, the licensee is to follow all the provisions in Code Case N 521 with limitations issued in Regulatory Guide 1.147,if any.

1 3.1.2 Pressurizer No relief requests. l 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.

13

3.1.6 'Velve Pressure Boundary No relief requests.

3.1.7 General No relief requests.

3.2 Class 2 Components

[

l No relief requests.

l L 3.3 Class 3 Components 3.3.1 Pressure Vessels i No relief requests.

l 3.3.2 Piping 3.3.2.1 Revised Requests for Relief 1-1S13 and 2-1813, IWD-1220.1 Class 3 Components Exempt From Examination l

l Code Megu/rement-LWD-1220.1 allows integral attachments of supports and restraints to '

components that are NPS 4 and smaller within the system boundaries of Examination L Categories D A, D 8, and D-C of Table IWD 25001 to be exempt from the VT-3 visual examination, except for PWR Auxiliary Feedwater Systems.

l Licensee's proposed Altemative--4n accordance with 10 CFR 50.5Ea(a)(3)(i), the licensee h

proposes to use the exemption criteria contained in lWA 5250 of the 1991 Addenda for NPS 1 and smaller piping in the Auxiliary Feedwater System. The licensee states:

" Integral attachments of supports and restraints that are NPS 4 and smaller but j greater than NPS 1 in the Auxiliary Feedwater System within the system boundaries of Examination Categories D-A, D B and D-C of Table IWD 25001 shall be t

examined in accordance with the visual examination method."

Licensee's Basis for Proposed Attemative-

"The majority of NPS 1 and smaller auxiliary feedwater piping is field routed.

. ldentification of supports and the integrally welded attachments on this piping would be labor intensive. The identification process would require walkdowns of vent lines, drain lines, instrumentation lines and sampling lines. Subsequent to the walkdowns, substantial resources will be required to sketch the applicable lines.

Further resources will be required to check, verify and maintain the resulting sketches.

14

"NPS 1 and smaller piping is not analyzed for pipe rupture, due to the low safety significance associated with rupture of these lines. Sequoyah did not identify any significant problems with integrally welded attachments during the first inspection interval.

"ASME Section XI Codes 1991 Addenda and later incorporated the NPS 1 and smaller exemption requirements for Auxiliary Feedwater System. The exemption includes piping NPS 1 and smaller and exemptions for vessels, pumps, and vales and their connections in piping NPS 1 and smaller.

"Sequoyah believes that it is an unnecessary burden to inspect supports and integrally attachments on NPS 1 and smaller piping in the Class 3 Auxiliary Feedwater System. The inspection of these items willimpose costs without compromising the safety or quality of the unit.

"This request is to use only the NPS 1 and smaller exemption on auxiliary Feedwater System piping.

"It is unnecessary burden to examine supports and integrally welded attachments on NPS 1 and smaller piping in the Class 3 Auxiliary Feedwater System. Even for the Class 1 and 2 systems in the 1989 Edition of ASME Section XI, there are no requirements to examine supports and integrally welded attachments on NPS 1 and smaller. Prior to the 1980 Addenda, there was no requirement to examine Class 3 NPS 1 and smaller Auxiliary Feedwater System piping supports and integrally welded attachments. The later 1991 Addenda IWD 1220 incorporated the NPS 1 and smaller exemption to provide clarification on the exemptions.

" Problems with these integral attachment are not anticipated to cause any loss of function of the associated NPS 1 and smaller line."

(The licensee supplied a listing of their NPS 1 and smaller Auxiliary Feedwater Lines.)

Evaluet/on4n lieu of the Code-required VT 3 visual examinations of integral attachment welds in piping NPS 1 and smaller in the PWR Auxiliary Feedwater System, the licensee proposes to use the requirements of the 1991 Addenda, which exempts NPS 1 and smaller. Based on the licensee's proposed use of Code Case N 509 (evaluated elsewhere in this report), the INEEL believes that no significant burden exists. Code Case N 509 requires the licensee to inspect only a minimum of 10 percent of all Class 1, Class 2, and Class 3 integrally welded attachments to piping, pumps, and valves, uniformly distributed in all systems. Use of Code Case N 509 constitutes a significant reduction in the total number of components requiring examination, therefore, the licensee's basis for this request for relief is not justified.

Conchss/on--8ased on the licensee's proposed use of Code Case N-509 (see Requests for Relief 11S15 and 2 ISI 5), the INEEL staff believes that it has not been demonstrated that 15 ,

I 1

a significant burden exists. Therefore it is recommended that this request for relief be denied.

3.3.2 Pumps No relief requests.

3.3.3 Valves No relief requests.

3.3.4 General No relief requests.

3.4 Pressure Tests 3.4.1 Class 1 System Pressure Tests 3.4.1.1 Request for Relief ISPT-05, Rev 2, Table IWB 2500-1, Examination Category B P, Pressure Retaining Components Code Requ/rement-Examination Category B-P, Footnote 2, defines the system's pressure-retaining boundary for the hydrostatic test as including all Class 1 components within the system boundary.

Licensee's Proposed Attemer/ve-Pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee proposed an alternative to the Code required system hydrostatic test on open-ended tailpipes that serve as vent, drain, test, or fill lines in the Reactor Coolant System (RCS).

The licensee states:

"These piping segments will continue to be visually inspected following each refueling outage for leakage and evidence of past leakage during the RCS leakage test. This test is conducted with the RCS at full operating temperature and pressure."

Licensee's Basis for Proposed Attemative-

"Various piping segments are locsted in open-end tailpipes that serve as vent, drain, test, or filllines Manual valves and flanges bound these piping segments to provide the design-required double isolation at the reactor coolant pressure boundary. These piping segments are not normally pressurized. Pressure testing of these piping segments at nominal operating pressure in Mode 3 would require that the inboard isolation valve be opened when the reactor coolant system (RCS) is at full temperature and pressure (547 F and 2245 psig). The action would violate the design requirement for double isolation valve protection. The potential for spills when opening the system presents a significant risk of personnel contamination.

Pressure testing in Mode 6 would require that a hydrostatic pump be connected at 16

1 each segment location. However, for some segments there is no connection available and would require a modification for installation of a pump connection.

These piping segments are located in high-radiation areas and testing would result in high personnel radiation exposure. A breakdown of the dose estimates for each radiation area in the plant is provided below:

"1. RCS Loop Drains 6 items at 10 person hours per item at 300 mR/ hour.

"2. Reactor Vessel Head Vents 2 items at 10 person hours per item at 150 mR/ hour and 2 items at 8 person-hours per item at 20 mR/ hour.

"3. Pressurizer Spray Vents 2 items at 10 person-hours per item at 200 mR/ hour.

"4. Excess Letdown Drain 1 item at 8 person-hours per item at 50 mR/ hour.

"5. RCS Seal Drains and Vents 4 items at 8 person-hours per item at 20 mR/ hour and 4 items at 8 person-hours per item at 50 mR/ hour.

"Where mR' stands for millirem. This results in a total of 27.960 Rem of dose accumulated for performing these tests. This data is based on estimated durations and actual survey data from SON's cycle 5 outages. These radiation exposure estimates are based on a pressure test in Mode 6 when each of the blind flanges weuld have to be removed, a test flange installed, and a hydrostatic pump connected. Personnel would remain in the area to perform the test, disconnect the test equipment, and reinstall the blind flange.

"These piping segments are visually inspected each refueling outage as the unit returns to operation. These segments are not specifically pressurized past the first isolation valve for this inspection. It is possible that the piping is pressurized because of leakage at the first isolation valve. With these inspection being performed approximately six times in each inspection interval, the increase in safety achieved from the required nominal operating pressure test is not commensurate with the hardship of performing such testing.

EveAust/co-The Code requires a system hydrostatic test to be performed once per interval in accordance with IWA 5000 for Class 1 components. The pressure test is to be applied to the components within the Class boundary. Vents and drains are typically designed with redundant isolation valves. The first valvs is the primary isolation while the second '

provides redundant isolation. These valves are only opened and closed for system draining and venting. Under normal operating conditions, these valves would be closed. Requiring the licenses to open the primary valve for the purpose of pressurizing the second valve, L

17

and piping segments between valva, results in unnecessary radiation exposure and additional radioactive waste (the contaminated water that will be introduced between the isolation valves). Therefore, the INEEL staff believes that requiring the licensee to open the primary isolation valve, and short piping segment, for the sole purpose of pressurizing 4 the secondary isolation valve serves no practical purpose and results in a burden with no I compensating increase in quality and safety.

Conclus/an-Compliance with the Code's hydrostatic testing reg'uirements for the subject  !

piping segments results in hardship and/or unusual difficulty without a compensating increase in the level of quality and safety. Performing a visual examination on the outermost isolation valve of the subject lines for evidence of leakage will provide reasonable assurance of operational readiness. Therefore,it is recommended that the licensee's proposed alternative be authorized for'Sequoyah Nuclear Plant, Units 1 and 2, pursuant to 10 CFR 50.55a(a)(3)(ii).

3.4.2 Class 2 System Pressure Tests 3.4.2.1 Request for Relief ISPT-04, Use of Code Case N 522 for Pressure Testing Class 2 Components at Containment Penetrations Code #egu/rement-Examination Category C H, items C7.30, C7.40, C7.70, and C7.80, in conjunction with Code Case N-498-1, require system leakage testing of Class 2 piping and valves once each inspection period.

L/censee's proposed Attemer/ve-Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposes to use Code Case N-522 as an alternative to the Code-required system leakage test for all Code Class 2 containment penetration piping when the remaining inboard and outboard piping is outside the scope of Section XI. The licensee states:

"The requirements of ASME Code Case N 522 shall be used. When using Code Case N-522, TVA will perform such tests 1) at peak calculated containment pressure and 2) the test procedures will include methods for detection and location of through wall leakage in containment isolation valves (CIVs) and pipe segments between the CIVs."

Licensee's Bes/s for Proposed Attemet/ve-

"The portion of piping that penetrates containment and the associated inboard and outboard containment isolation valves are required to be constructed in accordance with Class 1 or 2 design requirements. In the instance where the penetration is associated with a non safety system, the sole function of the penetration and the associated valves are to provide containment isolation capability for the protection of containment integrity during the unlikely event of a loss of the attached non-safety piping. In all cases the isolation valves associated with these penetrations are maintained, dt. ring normal operation, in the locked closed position or close upon receipt of a containment isolation signal. The safety function of these penetrations is verified by 10CFR50 Appendix J leakage testing. The performance of additional pressure testing, as required by Table IWC-2500-1, Category C H, is considered 18 l

impractical in the sense that there is minimal enhancement to quality or safety resulting from addition testing. It is TVA SON's position that testing pursuant to 10CFR50 Appendix J requirements provides an acceptable level of quality and safety as the additional pressure testing requirements of the Code."

Eve /uet/on-The licensee proposed to implement the alternatives contained in Code Case N 522, Pressure Testing of Containment Penetration Piping, for portions of the lines that are Class 2 at the containment penetration. These segments of lines are safety related only because they function as part of the containment pressure boundary and are relied on for containment integrity. Therefore, it is logical to test the penetration piping portion of the associated systems to the containment test criteria found in 10 CFR 50.55a, Appendix J.

Appendix J pressure tests are local leak rate and integrated leak rate tests that verify the leak tight integrity of the primary reactor containment and of systems and components th6t penetrate containment. In addition, Appendix J test frequencies provide assurance that the containment pressure boundary is being maintained at an acceptable level while monitoring for deterioration of seals, valves, and piping. Use of Appendix J, Option B results in tests being performed at intervals not exceeding 60 months versus 40 months as required by the Code. The staff has determined that these containment testing frequencies are acceptable, therefore they should also be considered acceptable for the subject piping.

The Class 2 containment isolation valves (CIVs) and connecting pipe segments must withstand the peak calculated containment internal pressure related to the maximum design containment pressure. The INEEL finds that the pressure-retaining integrity of the CIVs and connecting piping and their associated safety functions may be verified with an Appendix J, Type C test if it is conducted at the peak calculated containment pressure.

The seal between the connecting pipe segment and containment may be verified using an Appendix J, Type B test. Therefore, when the connecting pipe segment is subjected to either a Type B or C test, its safety function is verified by the Appendix J test.

Section XI, IWC-5210(b) requires that, where air or gas is used as a testing medium, the test procedure include methods for detection and location of through-wallleakage. If the licensee's test procedure uses air as a testing medium, the procedure should meet the above requirement for the CIVs and pipe segments between the CIVs.

The INEEL staff believes that an acceptable level of quality and safety will be provided by Appendix J tests, when the leak test is performed at the peak calculated containment design pressure and the test procedure provides for detection and location of through wall leaks.

l Conclus/on-It is recommended that the licensee's proposed alternative to the Code-required pressure tests, the use of Code Case N-522 with the stipulations that the leak l test will be performed at the peak calculated containment pressure and that the test procedure will provide for detection and location of through wallleaks, be authorized

{

19

pursuant to 10 CFR 50.55a(a)(3)(i). The use of alternatives contained in Code Case N 522 should be authorized for the current interval or entil such time as the Code Case is published in Regulatory Guide 1.147. At that time, if the licensee intends to continue to implement this Code Case, the licensee is to follow all the provisions in Code Case N 522 with limitations issued in Regulatory Guide 1.147, if any.

3.4.3 Class 3 System Pressure Tests No relief requests.

3.4.4 General 3.4.4.1 Request for Relief ISPT 01, Request for Authorization to Use ASME Code Case N-4981, Altemative # des for 10-Year System Hydrostatic Testing for Class 1, 2, and 3 Systems,Section XI, Division 1 Code Regdrement-Table IWD 25001, Examination Category D A, items D1.10, D2.10, and D3.10 require that a system hydrostatic test be performed in accordance with IWA-5000 once each 10-year interval.

Licensee's Proposed Altemative-Pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee proposed an alternative to the Code-required hydrostatic pressure test. The licensee states:

"The regt.irements of ASME Code Case N 498-1 will be used for the requirements ,

of system pressure testing for Class 3 systems."

Licensee's Basis for Proposed Attemative-

"Since the acceptance of Code Case N-498 it has been common recognition that 10 year hydrostatic pressure testing can be somewhat burden some, and in some instances, may expose components to unnecessary stress levels without significant enhancement to quality or safety. TVA SON employs a very comprehensive corrosion monitoring and control program which periodically performs visual examinations and wall thickness examination by volumetric on selected Class 3 piping and components as a very accurate means of monitoring for pressure boundary degradation. It is SON's position that performing system pressure test on Class 3 systems consistent with the requirements of N 498-1 in conjunction with the Corrosion Monitoring Program provides equivalent or superior assurance of system integrity as that provided by the Code."

' Evaluation-The Code requires the performance of a system hydrostatic test once per intervalin accordance with IWA 5000 for Class 1,2, and 3 pressure retaining systems, in lieu of the Code hydrostatic testing requirements, the licensee has requested authorization to use Code Case N 498-1, Alternative Rules for 10-Year System Hydrostatic Testing for Class 1, 2, and 3 Systems, dated May 11,1994.

20

Code Case N-498 Alternative Rules for 10-Year System Hydrostatic Testing for Class 7 and 2 Systems, was previously approved for general use on Class 1 and 2 systems in Regulatory Guide 1.147, Rev. 9. For Class 3 systems, Revision N 498-1 specifies requirements identical to those for Class 2 components (for Class 1 and 2 systems, the alternative requirements in N-4981 are unchanged from N-498). In lieu of 10-year hydrostatic pressure testing at or near the end of the 10-year interval, Code Case N-498-1 requires a VT-2 visual examination at nominal operating pressure and temperature in conjunction with a system leakage test performed in accordance with paragraph IWA-5000 of the 1992 Edition of Section XI.

The system hydrostatic test, as stipulated in Section XI,is not a test of the structural integrity of the system but rather an enhanced leakage test (Reference 17). Hydrostatic testing only subjects the piping components to a small increase in pressure over the design pressure; therefore, piping dead weight, thermal expansion, and seismic loads present far greater challenges to the structuralintegrity of a system. Consequently, the Section XI hydrostatic pressure test is primarily regarded as a means to enhance leak detection during the examination of components under pressure, rather than as a method to determine the structuralintegrity of the components. In addition, industry experience indicates that leaks are not being discovered as a result of hydrostatic test pressures propagating a preexisting i flaw through the wall-in most cases leaks are being found when the system is at normal operating pressure.

Class 3 systems do not normally receive the amount and/or type of nondestructive examinations that Class 1 and 2 systems receive. While Class 1 and 2 system failures are relatively uncommon, Class 3 leaks occur more frequently and are caused by different failure mechanisms. Based on a review of Class 3 system failures requiring repair during the last 5 years,' the most common causes of failures are erosion-corrosion (EC),

microbiologically-induced corrosion (MIC), and general corrosion. In general, licensees l have implemented programs for the prevention, detection, and evaluation of EC and MIC:

therefore, Class 3 systems receive inspection commensurate with their functions and expected failure mechanisms.

System hydrostatic testing entails considerable time, radiation dose, and dollar resources. The safety assurance provided by the enhanced leakage gained from a slight increase in system pressure during a hydrostatic test may be offset or negated by the necessity to gag or remove Code safety and/or relief valves (placing the system, and thus the plant, in an off normal state), erect temporary supports in steam lines, and expend resources to set up testing with special equipmerit and gages. Therefore, performance of system hydrostatic testing represents a consider.eble burden without a compensating increase in quality and safety. Giving consideration to the minimal amount of increased  !

assurance provided by the increased pressure associatet, with a hydrostatic test versus the pressure for the system leakage test and the hardship associated with performing the

a. Documented in Licensee Event Reports and the Nuclear Plant Reliability Data System databases.

21

hydrostatic test, the INEEL staff finds that compliance with the Section XI hydrostatic testing requirements results in hardship and/or unusual difficulty without a compensating increase in the level of quality and safety.

Conclus/on-Considering the minimal amount of increased assurance provided by the increased pressure associated with a hydrostatic test versus the pressure for the system leakage test and the hardship associated with performing the hydrostatic test, the INEEL staff finds that compliance with the Section XI hydrostatic testing requirements results in hardship and/or unusual difficulty without a compensating increase in the level of quality and safety. Therefore, it is recommended that the use of Code Case N-4981 for Code Class 1,2, and 3 systems be authorized pursuant to 10 CFR 50.55a(s)(3)(ii) for the second interval or until such time as 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 this Code Case, the licensee must follow all provisions in Code Case N-4981, with limitations issued in Regulatory Guide 1.147, if any.

3.4.4.2 Request for Relief No. ISPT-02, Use of Code Case N.41641, Attemet/ve Pressure Test Reqdrements for Welded Repeles or insteMetion of Replacement items by Welding,Section XI, Division 1 Code Reqdrement-Section XI, Paragraph IWA-4400, Pressure Test, requires a system hydrostatic test in accordance with IWA-5000 after repairs by welding on a pressure-retaining boundary.

Following welding, the Code requires volumetric and/or surface examination (depending on wall thickness) of repairs or replacements in Code Class 1 and 2 systems, but only requires a surface examination of the final weld pass in Code Class 3 piping components.

There are no ongoing nondestructive examination (NDE) requirements for Code Class 3 components except for VT-2 visual examination for leaks in conjunction with the 10 year hydrostatic tests and the periodic pressure tests.

Licensee's Proposed A/temet/ve-Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed to use Code Case N 4161, Alternative Pressure Test Requirements for Welded Repairs or Instellation of Replacement items by Welding,Section XI, Division 1. The licensee states: '

"The requirements of ASME Code Case N-416-1 shall be used. TVA will also perform

~

an additional surface examination on the root (pass) layer of butt and socket welds on the pressure retaining boundary of Class 3 welds which require a surface examination as required by ASME Section lit, Subsection ND 5222."

22

Ucensee's sesis for Proposed Attemative-

" Class 1,2, and 3 pressure boundary replacements requiring installation by welding are normally constructed and supplied in accordance with the requirements of ,

ASME Section ill which provides for hydrostatic testing at the manufacturers.

l Subsequent to installation hydrostatic pressure testing is a means of proving weld integrity. Performing NDE and invoking acceptance criteria in accordance with current ASME lll requirements in addition to a system leakage test provides 3 reasonable assurance that weld integrity is maintained at an acceptable level of quality, it is TVA SON's position that the performance of hydrostatic testing subsequent to weld repairs and the installation of welded replacements is an impractical requirement which could not only expose components to unnecessary

. stress levels, but provides little or no enhancement to the level of quality or safety."

Eve /uet/on-6ection XI of the Code requires a system hydrostatic test to be performed in accordance with IWA 5000 after repairs made by welding on the piessure retaining boundary. The licensee has proposed the use of Code Case N-416-1 in lieu of the Code requirements. Code Case N-416-1 specifies that NDE of the welds be performed in accordance with the applicable subsection of the 1992 Edition of Section Ill. The Code Case also allows a VT-2 visual examination to be performed at nominal operating pressure and temperature in conjunction with a system leakage test, in accordance with Paragraph IWA 5000 of the 1992 Edition of Section XI.

The 1989 Editions of Sections til and XI are the latest Code editions referenced in 10 CFR 50.55s. The NRC staff previously compared the system presst.re test requirements of the 1992 Edition of Section XI to those of the 1989 Edition. In summary:

1) The test frequencies and the pressure conditions associated with these tests have not changed:

2) The hold times have either remained unchanged or increased;

3) The terminology associated with the system pressure test requirements for all three Code classes has been clarified and streamlined; and

4) The NDE requirements for welded repairs remain the same.

I Piping components are designed to withstand the loading mechanisms that are postulated to occur under the various modes of plant operation. Hydrostatic testing I subjects the poing components to a smallincrease in pressure over the design pressure and, thereforo. does not present a significant challenge to pressure boundary integrity.

Accordingly, hydrostatic pressure testing is primarily regarded as a means to enhance leak detection during the examination of components under pressure rather than a measure of i the structuralintegrity of the components.

i Considering the NDE performed on Code Class 1 and 2 systems and that the  ;

hydrostatic pressure tests rarely result in pressure boundary leaks that would not have occurred during system leakage tests, the INEEL staff believes that the added assurance of

~

integrity provided by the hydrostatic test is not commensurate with the associated burden, which typically includes the installation of blanks, cutting and removing supports for '

23

access, and removing insulation to prepare and restore the systems, all of which increase radiation exposure for plant personnel.

For Class 3 components, there are no ongoing NDE requirements except for the visual examination for leaks in conjunction with the 10-year hydrostatic test and periodic pressure tests. Therefore, eliminating the hydrostatic test and only performing the system pressure test for Class 3 components is considered acceptable because the licensee is performing an additional surface examination on the root pass layer of butt and socket welds on the pressure retaining boundary during repair and replacement activities for Class 3 systems.

ConcAusion-The licensee's proposed alternative, use of Code Case N 416-1 with an additional surface examination of the root pass layer of repair and replacement butt and socket welds in Class 3 systems, provides an acceptable level of quality and safety.

Therefore, it is recommended that the proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i). The use of this Code Case should be authorized during the second 10 year interval until such time as the Code Case is approved for general use by reference in Regulatory Guide 1.147. After that time, the licensee may continue to use Code Case N 4161 with the limitations, if any, listed in Regulatory Guide 1.147.

3.4.4.3 Request for Relief ISPT-03, lWA 5250(a)(2), System Pressure Test Corrective Measures Code Requirement -4WA-5250(a)(2) states that if leakage occurs at a bolted connection during a system pressure test, then all bolting must be removed and a VT-3 visual examination performed to detect corrosion.

Lkensee's Proposed Altemative-Fursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed an alternative to performing the Code-required removal and VT-3 visual examination of bolting if leakage occurs during a system pressure test of Class 1,2, and 3 systems. The licensee stated:

"The requirements of IWA 5250(a)(2) of the 1990 Addenda to the 1989 Edition will be used".

Licensee's Basis for Requesting ReMef-

" Current Codo requirements specify that all bolting must be removed in the event of a bolted connection leak for the purpose of VT-3 examination and evaluation. This would require placing the associated component or portion of piping out of service possibly resulting in plant shutdown, delaying plant startup, or placing the plant in an unsafe condition for continued operation. Additionally, removal of all botting is impractical if there is reasonable assurance that bolting materialis of a specification which is not susceptible to corrosion when in contact with leaking fluid. It is TVA SON's position that proposed alternative testing provides an acceptable level of quality and safety as that provide by the Code."

24 j

Eve /uet/on-The Code of record requires that, for leakage at bolted connections, all the bolting be removed for VT 3 visual examination. in lieu of this requirement, the licensee has proposed to remove only the bolt closest to the source of leakage for VT 3 visual examination. The 1990 Addenda requires the removal of one bolt closest to the leakage; if there is evidence of degradation, all the remaining bolting at that connection must be removed and evaluated. Based on the licensee's proposed Request for Relief ISPT 08 (as evaluated in following Section 3.4.4.4), the INEEL believes that this request for relief should not be necessary in that the licensee has proposed that an engineering evaluation be done prior to any corrective measures.

ConcAus/on-Based on the licensee's Request for Relief ISPT 08 (see following Section 3.4.4.4), the INEEL staff believes that this request for relief is redundant and that no significant burden exists. Therefore, it is recommended that this request for relief be denied.

3.4.4.4 Request for Relief ISPT-08, IWA 5250(a)(2), System Pressure Test Corrective '

Measures Code #eguirement-4WA 5250(a)(2) states that if leakage occurs at a bolted connection during a system pressure test, then all bolting must be removed and a VT-3 visual examination performed to detect corrosion.

L/consee's Proposed Altomer/ve-Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed an alternative the Code'-required removal and VT 3 visual examination of bolting if leakage occurs during a system pressure test of Class 1,2, and 3 systems. The licensee stated:

"When evidence of leakage is discovered at bolted connection during a Section XI inservice pressure test, the connection will be evaluated for structural integrity using the following factors:

"1. Size of the leak "2. Duration of leak "3. Bolting and flange materials "4. Visual evidence of corrosion with the connection assembled "5. Corrosivenuss of the process fluid "6. Experience with similar bolting material in similar environment "7. Degradation cf other components in the vicinity of the leakage "When the engineering evaluation determines that the bolted connection possesses sufficient structural integrity for continued operation, removal of the bolt may be deferred to the next unit outage of sufficient duration to allow the system to be removed from service and depressurized. Should the engineedng evaluation determine that sufficient structural integrity does not exist at the connection, provisions will be made to remove one bolt for visual examination (VT-3) in '

compliance with IWA 5250(a)(2)."

25

Licensee's Basis for Proposed Attemative-

"ASME Section XI inservice pressure tests are, as a rule, performed with the system inservice, in particular, the Code Class 1 leakage tests are performed as the unit is returning to service following each refueling outage. The requirement to immediately remove bolting from a mechanical connection when evidence of leakage is detected can create a significant hardship on the plant which is not commensurate with the increase in the level of quality and safety that is provided.

For systems that are aligned to normal plant operating configuration during testing, paragraph IWA 5250(a)(2) may require the system be taken out of service and depressurized to permit removal of one of the bolts prior to any type of engineering analysis of the connection. An engineering evaluation of the leak and the affected mechanical connection may be able to determine that sufficient structural integrity exists in the connection and that removal of bo! ting for visual examination in compliance with paragraph IWA 5250(a)(2) can be deferred to the next unit outage without a reduction in component safety margin."

Eve /uetion--4n eccordance with the 1989 Edition of the Code, when leakage occurs at bolted connections, all boltino is required to be removed for VT-3 visual examination. It i should also be noted that later Addenda of the Code (i.e.1990 Addenda) requires the  !

removal of one bolt closest to the leakage; and if there is evidence of degradation, all the remaining bolting at that connection must be removed and evaluated. This concept focuses on the area most prone to degradation and evaluates the bolt exposed to the leakage, in lieu of the Code-required removal of bolting, the licensee has proposed to evaluate the bolted connection to determine the susceptibility of the bolting to corrosion and the potential for failure.

1 The proposed alternative is a systematic approach that allows the licensee to apply sound engineering judgement provided that, as a minimum, all seven evaluation factors j listed in the proposed alternative are considered. Furthermore,if the initial evaluation indicates the need for a more in-depth evaluation, the bcit closest to the source of leakage will be removed, VT-1 examined, and evaluated in accordance with IWA-3100(a).

Therefore, the licensee's alternative to the Code-required removal of botting at a joint when leakage occurs will provide an acceptab's level of quality and safety, as the integrity of the joint will be maintained.

Conchs /on-The licensee's proposed attemative, to use a systematic approach and sound engineering judgement for corrective measures when leakage at bolted connections is observed, will provide an acceptable level of quality and safety. Therefore, it is

. recommended that the proposed altemative be authorized pursuant to 10 CFR 50.55sta)(3)(i).

26

3.5 General 3.5.1 Ultrasonic Examination Techniques No relief requests.

3.5.2 Exempted Components 3.5.2.1 Requests for Relief 1-1S1-4 and 2-1S14, IWC-1221, IWC-1222(b) and IWD 1220.1, Class 2 and 3 Items Exempt From Examination in the licensee's August 28,1997, response to an NRC request for additional information, the licensee withdrew Requests for Relief 1-ISI 4 and 2-1S1-4.

3.5.3 Other 3.5.3.1 Request for Relief 1-ISI 5 and 2 ISI-5, Use of Code Case N 509, Altemat/ve Rules for the Selection and Examination of Class 1,2, and 3 Integrelly Welded Attachments Code Requ/rement-ASME Section XI, Examination Categories B-H and B K 1, items B8.10, B8.20, B8.30, B8.40, B10.10, B10.20, and B10.30 require a volumetric or surface examination of integrally welded attachments as defined by Figures IWB-2500-13, -14, or

-15, as applicable. Examination Category C-C, items C3.10, C3.20, C3.30, and C3.40 require surface examination as defined by Figure IWC 2500-5. Examination Category D-A, D-B, and D-C, items D1.20 through D1.60, D2.20 through D2.60, and D3.20 through D3.60 require VT-3 visual examination as defined by Figure IWD-25001.

Licensee's proposed Altemat/ve-Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed an alternative to the Code requirements associated with the selection and examination of integrally-welded attachments. The licensee proposes to implement Code Case N-509, Alternate Rules for the Selection and Examination of Class 1, 2, and 3 Integrally Welded Attachments,Section XI, Division 1. The licensee stated:

" Relief is requested to use Code Case N-509, " Alternative Rules for the Selection and Examination of Class 1,2, and 3 Integrally Welded Attachments,Section XI, Division 1 in lieu of the 1989 Edition of ASME Section XI."

"The examination requirements of Code Case N-509 will be used as an altemative to the examination requirements of the 1989 Edition of ASME Section XI, Division 1.

" Note: Code Case N-509 requires selection of component supports for examination in accordance with IWF of the 1990 Addenda. Sequoyah willimplement Code Case N-509 utilizing Code Case N-491, which contains the same requirements for selection of component supports for examination as the 1990 Addenda. Code Case N-491 has been approved for use in Regulatory Guide 1.147.

27

" SON will perform surface examinations on a minimum of 10 percent of the total number of nonexempt Class 1,2, and 3 integral attachments on piping, pumps and valves.

"In the case of multiple vessels of similar design, function, and service, only one integral attachment of only one of the multiple vessels will be examined.

" Examinations will be performed on integral attachments when a component support member deformation (e.g. broken, bent, or pulled out parts) is identified during operation, refueling, maintenance, examination, inservice inspection, or I testing."

Licensee's Basis for the Proposed Attemative-

"In many cases, performing a surface examination on Class 1 and Class 2 integrally welded attachments requires the removal of pipe clamps. Pipe clamp removalis labor intensive and often destroys bolting in the process. Pipe clamp removal on larger components such as those found on the Main Steam and Feedwater Systems are particularly difficult.

"Sequoyah did not identify any significant problems with integrally welded attachments in the first ISI Inspection Interval. The costs for preparing and examining integrally welded attachments are significant, as is the ottendant additional radiation exposure. Code Case N-509 provides an overall cost savings and exposure reduction without compromising safety or quality.

"By utilizing Code Case N.509 the number of integral attachments requiring i examination is reduced for Class 1 integrally welded attachments where base I material design thickness is 5/8" or greater, Class 2 integrally welded attachments where base material design thickness is 3/4" or greater, and Class 3 integrally welded attachments.

"However Code Case N-509 increases the Class 1 and 2 integral attachment population for certain areas due to the elimination of material design base thickness exemption and for Class 3 integrally welded attachments a more stringent visual examination method (VT-1 in lieu of a VT-3) is required." {

)

l Evaluation-in lieu of Code requirements for selection and examination of integral I attachment welds, the licensee proposes to apply Code Case N 509, A/temative Rules for j the Selection and Examination of Class 1, 2, and 3 Integrally Welded Attachments, which I allows e significant reduction in the total number of examinations performed. The licensee stated that at least a 10% sample of all nonexempt integral attachment welds in Class 1 2, snd 3 systems will be examined.

i Considering that many of the Code examination requirements are ba' sed m sen,r> ling to assure that service-related degradation is not occurring, it is logical to extend the san pling process to welded ir$tegral attachments. Based on the licensee's proposed sampie of a 28

minimum of 10% of allintegral attachment welds, uniformly distributed among all Code j Class 1,2, and 3 systems, the INEEL staff believes that degradation, if occurring, will be I detected. Therefore, the use of the alternatives contained in Code Case N 509, with a minimum 10% selection of allintegrally welded attachments distributed in each Code Class / system, will provide an acceptable level of quality and safety.

ConcAusion-The licensee has proposed to examine integral attachments in accordance with Code Case N 509, with a minimum 10% selection of all nonexempt Code Class 1,2, and 3 integrally welded attachments on piping, pumps, and valves. The INEEL staff believes that the licensee's proposed alternative will provide an acceptable level of quality and safety. Therefore,it is recommended that the licensee's proposed altemative be authorized pursuant to 10 CFR 50.55a(a)(3)(i). Use of alternatives contained in Code Case N 509, with the selection provision noted above, should be authorized for the current interval or until such time as 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 this code case, the licensee should follow all provisions in Code Case N 509, with conditions issued in Regulatory Guide 1.147, if any.

3.5.3.2 Requests for Relief 1-1817 and 2-1S17, Use of Code Case N 524, A/temative Examination Requirements for Longitudinal Welds in Class 1 and 2 Mping,Section XI, Division 1 Code Requirement-4xamination Categories B J, C-F-1, and C F-2, items 89.12, C5.12, C5.22, C5.52, and C5.62, require surface and volumetric of examination Class 1 and 2 longitudinal piping welds. Items 89.22, C5.42, and C5.82 require surface examination of Class 1 and 2 longitudinal piping welds. The examination volume / surface area includes 2.5t at the intersection with circumferential welds required to be examined.

I Licensee's Proposed A/temative-Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed to use the alternatives contained in Code Case N 524, A/temative Examination Requirements for Longitudinal Welds in Class 1 and 2 Piping,Section XI. Division 1, in lieu of 100% of the Code-required surface and volumetric examinations on Class 1 and 2 longitudinal piping welds. The licensee stated:

"As an alternative to the requirements of the 1989 Edition of ASME Section XI, SON will perform examinations in accordance with the requirements of Code Case N-524 as follows:

"(a) When only a surface examination is required, examination of longitudinal piping welds is not required beyond those portions of the welds within the examination boundaries of intersecting circumferential welds.

"(b) When both surface and volumetric examinations are required, examination of longitudinal piping welds is not required beyond those portions of the welds within the examination boundaries of intersecting l circumferential welds provided the following requirements are met:  ;

29 A-

C *

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"(1) Where icngitudinal welds are specified and locations are known, examination requirements shall be met for both transverse and parallel flaws at the intersection of the welds and for that length of longitudinal weld within the circumferential weld examination volume

"!2) Where longitudinal welds are specified but locations are unknown, or the existence of longitudinal welds is uncertain, the examination requirements shall be met for both transverse and parallel flaws within the entire examination volume of intersecting circumferential welds."

Licensee's Basis for Proposeef Altsmetive-

" 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.

"Throughout the industry, 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 longitudinal piping welds can require acid etching, eddy current ,

examination, or a combination of NDE methods for location. This increases I radiological exposure, radweste generation, and overall costs for performance of )

ASME Section XI examinations." l EveAustion-The licensee proposed to implement the alternatives contained in Code  !

Case N-524, Attemative Examination Requirements for longitudinal Welds in Class 1 and 2 Piping, Section X/, Division f, in lieu of performing the applicable surface and volumetric examinations of longitudinal welds in Class 1 and 2 piping, as required by the Code. Code Case N 524 allows the licenses to examine only the potentially critical portions of the  !

longitudinal welds (the portions that intersect circumferential welds) in conjunction with examination of the circumferential welds.  ;

The licensee's altamative is based on the position that, due to fabrication controls and lack of susceptibility to the conditions that lead to failure, longitudinal welds are unlikely to ,

' fail. The potentially critical portions of the longitudinal welds are the portions that j 1 intersect circumferential welds; these regions will be examined in conjunction with the circumferential welds. Additionally, where the longitudinal weld cannot be identified, 30

O 4 100% of the circumferential weld will be examined for flaws parallel and transverse to the weld to ensure that the longitudinal /circumferential weld intersection is examined. With this alternative, the most critical area of the longitudinal weld is examined, thus providing an acceptable level of quality and safety. Based on the fabrication quality of longitudinal welds and the extent of examinations performed, this provides an acceptable level of

, quality and safety.

Conclusion-The licensee's proposed altemative, the use of Code Case N-524 for examination of Class 1 and 2 piping longitudinal welds provides an acceptable level of quality and safety. Therefore,it is recommended that the use of Code Case N 524 be approved pursuant to 10 CFR 50.55a(a)(3)(i). Use of Code Case N 524 should be authorized for the current interval or until such time as the Code Case is published in Regulatory Guide 1.147. At that time, if the lictnsee intends to continue to implement this Code Case, the licensee is to follow all provisions in Code Case N 524 with limitations issued in Regulatory Guide 1.147,if any.

3.5.3.3 Request for Relief No. ISPT-06, Use of Code Case N 546, A/temative Requirements for QueMcation of W-2 WsualExamination Personnel Code #eguirement-Section XI, IWA 2300, requires thrs personnel performing VT-2 and VT-3 visual examinations be qualified in accordance with comparable levels of competency as defined in ANSI N45.2.6. Additionally, the examination personnel shall have natural or corrected near distance vision acuity, in at least one eye, equivalent to a Snell fraction of 20/20. For far vision, personnel shall have natural or corrected far distance visual acuity of 20/30 or equivalent.

Licensee's prqposed Attemative +ursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposes to use Code Case N 546, Altemative Requirement for Qualification of W-2 Examination Aersonnel, as an alternative to the qualification requirements for VT 2 visual examiners. The licensee stated:

" SON will use ASME Code Case N-546 as altemative requirements for the qualification of VT-2 examination personnel. TVA will 1) develop procedural guidelines for obtaining consistent, quality VT-2 visual examinations,2) document, and maintain records to verify, the qualification of persons selected to perform VT-2 I visual examinations, and 3) implement independent review and evaluation of leakage by persons other than those that perform the VT 2 visual examination."

Licensee's Basis for Prpposed Attemative-

"ASME Section XI Code Case N 546 has been approved by ASME but has not been published nor has it been accepted for use by the NRC through Regulatory Guide 1.147. This Code Case provides alternative requirements for the qualification of personnel to perform VT 2 examinations. The opinion of the Coinmittee is that  !

VT-2 visual examination personnel need not be qualified nor certified to comparable j levels of competence in accordance with the referenced standard (i.e., ANSI '

N45.2.6, ASNT SNT-TC-1 A, or ASNT CP-199) provided the examination personnel 1 are qualified in accordance with the following requirements. I l

31  !

I "1. At least 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> plant walkdown experience, such as that gained by licensed and nonlicensed operators, local leak rate personnel, system engineers, and i mspection and nondestructive examination personnel.

"2. At least four hours of training on Section XI requirements and plant specific i procedures for VT-2 visual examination.

l

) "3. Vision test requirements of IWA 2321,1995 Edition.

I "The use of this Case will allow SON to use experienced plant personnel to perform l the VT 2 examination during the performance of system pressure tests. These personnel are knowledgeable of the plant systems and routinely perform walkdowns of the plant systems looking for abnormalities such as leaks in piping systems. By l performing routine walkdowns, they are more familiar with the location of piping systems and can therefore perform a VT-2 examination in a more timely manner.

Plant operators have a keen sense of duty while performing their daily rounds to l ensure that plant systems are operating correctly and reporting problems such as leakage from piping systems. System Engineering perform routine walkdowns as part of their job to ensure that plant systems are operating correctly. Their l experience, responsibility, and knowledge should be used and can help the plant l perform system pressure testing in a more timely and responsible manner which j

helps to ensure that plant piping systems operate correctly which is the intent of performing ASME Section XI system pressure tests. Using these personnel will also allow SON to eliminate the need for hiring additional personnel certified to the requirements of IWA 2300 especially during refueling outages when system pressure tests are typically performed. Since the VT-2 examination is an examination for the evidence of leakage, the use of plant personnel, certified to the attemative requirements, who typically perform this type of examination during their daily job duties will not compromise the quality or safety of the VT-2 l examinations."

Eve /uet/on-The Code requires that VT-2 visual examination personnel be qualified to comparable levels of competency as defined in ANSI N45.2.6. The Code also requires that the examination personnel be qualified for near and far distance vision acuity. The licensee proposes to implement the alternatives contained in Code Case N 546 which allows reduced qualifications for VT-2 personnel. The licensee stated that VT-2 visual examiners will meet the following requiremsats, as included in the Code Case:

1. At least 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> plant walkdown experience, such as that gained by licensed and nonlicensed operators, local leak rate personnel, system engineers, and inspection and nondestructive examination personnel.

2. At least four hours of training on Section XI requirements and plant specific procedures for VT-2 visual examination.'

3. Vision test requirements of IWA 2321,1995 Edition.

32

Based on a review of Code Case N 546, the INEEL staff believes that these alternatives to the Code qualification requirements for examination personnel will provide an acceptable level of quality and safety. The qualification requirements in Code Case N 546 are not significantly different from the qualifications required for VT 2 visual examiner certification.

Licensed and nonlicensed operators, local leak rate personnel, system engineers, and inspection and nondestructive examination personnel typically have a sound working knowledge of plant components and piping layouts. This knowledge makes them acceptable candidates for performing VT-2 visual examinations.

The licensee's proposal is to use' employees with a variety of occupational backgrounds to perform the VT-2 visual examinations. Because of the varied experience levels and potentially different interpretations of leakage, the licensee has committed to:

Developed procedural guidelines to obtain consistent VT-2 visual examination results.

Implement a program that documert3 the qualifications, training, and visual acuity of persons selected to perform the VT-2 visual examinations.

To perform an independent review and evaluation of findings by r, arsons other than those that performed the VT-2 visual examinations.

Conclusion-The licensee's proposed alternative requirements for cue.... cation of VT-2 examination personnel will provide an acceptable level of quality and safety. Therefore, it is recommended that the proposed altemative be authorized pursuant to 10 CFR 50.55s(a)(3)(i). The use of Codo Case N 546 should be authorized for the current interval, or until such time as the Code Case is published in a future revision of Regulatory Guide 1.147. After that time, if the licensee continues to implement this Code Casc. the licensee is to follow all provisions in Code Case N 546 with limitations issued in Regulatory Guide 1.147, if any.

3.5.3.4 Request for Relief No. iSPT-07, Use of Code Case N 533, Attemative Requirements for W-2 VisualExamination of Class 1 AnsulatedPressure Retaining Botted Connections,Section XI, Division 1 Code Regu/rement--lWA 5242(a) reauires that, for systems borated for the purpose of controlling reactivity, insulation shall be removed from pressure-retaining bolted connections for VT-2 visual examination.

f.icenroe's proposed Attemative-Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed to implement the altamatives contained in Code Case N 533, Altemative Requirements for W-2 VisualExamination of Class 1 insulated Pressure-Retaining Bolted Connections,Section XI, Division 1, in lieu of the AS'ME Section XI requirements for removing insulation from Class 1 and Class 2 pressure-retaining bolted connsetions located

~

inside containment during VT-2 visual examinations. The licensee stated:

"The alternative test of Code Case N 533 will be performed at each refueling outage. A four hour hold time at test conditions will be observed prior to the VT-2 visual examination."

1 33

c ,

Licensee's Basis for Proposed Attemative- *

"The leakage test of Code Class 1 components is performed at the completion of each refueling outage with the unit in hot standby and the reactor coolant system at full pressure and temperature. The Section XI Code Class 1 leakage test is generally the final activity before unit restart following the refueling outage.

Compliance with the Code requirements would involve holding the unit in hot standby mode after completion of the VT-2 examination of all Code Class 1 components to enable replacement of insulation at each Code Class 1 bolted connection and the removal of scaffolding necessary for reinstallation of insulation.

This situation places a hardship on the plant for the following reasons:

"1. Entering containment to replace thermalinsulation and to remove scaffolding when the unit is at full temperature jeopardizes the safety of personnel due to heat stress and the potential for burns resulting from contact with hot components.

> "2. Insulation replacement activities require holding the unit in hot standby mode until all work is completed and ali personnel have exited containment. These activities will delay the return of the unit to production for several hours.

"The purpose of removing insulation from pressure retaining bolting for visual examination is to inspect for borated water leakage that could cause corrosion of the bolting. Due to the deposits of boron crystals that remain where borated water leakage occurs, it is not necessary to actually see the fluid leakage in order to determine where leakage has occurred. Therefore, borated water leakage inspections can be effectively performed when the system is depressurized. For this reason the hardships resulting from Code compliance are mot commensurate with the increase in safety achieved.

"The leakage testing of the Code Class 2 components inside containment is performed l in conjunction with Code Class 1 components during unit startup following each refueling outage when the components are pressurized to full system operating l pressure. The basis for deferral of insulation removal for these components is the i same as that for Code Class 1."

Eva/ustion-Paragraph IWA 5242(a) requires the removal of all insulation from pressure- i retaining bolted connections in systems borated for the purpose of controlling reactivity s

when performing VT-2 visual examinations during system pressure tests. The licensee has proposed to implement the alternative to Code requirements contained in Code Case N 533, Altemative Requirements for VT-2 VisualExamination of Class 1 Insulated Pressure- Metaining Solted Connections, for Class 1 and Class 2 bolting in borated systems located inside containment. This Case allows the VT 2 visual examination to be performed in conjunction with startup following a 4-hour hold time at operating pressure with the insulation in place. In addition, a separate VT-2 visual examination is then performed each l tofueling outage during cold shutdown with the insulation removed. Use of this Case l

significantly reduces the personnel hazerds associated with the extreme heat and radiation I 34

o

• O exposure that would occur during a VT-2 examination with the insulation removed a.'.d subsequent replacement of the insulation prior to the run cycle.

When bolted connections are examined in accordance with Code Case N 533, the joints are VT 2 visually examined during the start-up pressure test (insulation in place) and again each refueling outage (insulation removed). Based on this frequency of examinations, it can be concluded that the bolted joint integrity will be verified at the same frequency currently required by the Code. Significant leakage during the pressure test would be detected by the VT-2 visual examination performed with the insulation in place.

If leakage occurs, corrective action would be necessary to meet minimum technical specification requirements. Therefore, an acceptable level of quality and safety is provided by the proposed alternative.

ConcAusion-The proposed alternative, the use of Code Case N 533, provides an acceptable level of quality and safety. Therefore, it is recommended that the use of Code Case N-533 be authorized pursuant to 10 CFR 50.55a(a)(3)(i). Use of Code Case N 533 should be authorized for the current interval or until such time as the Code Case is published in Regulatory Guide 1.147. At that time, if the licensee intends to continue to implement this Code Case, the Ucensee is to follow all provisions in Code Case N 533 with limitations issued in Regulatory Guide 1.147, if any.

35

4. CONCLUSION Pursuant to 10 CFR 50.55a(g)(6)(i), it has been determined that certain inservice examinations cannot be performed to the extent required by Section XI of the ASME Code.

In the cases of Relief Requests 1-1S11 (Part 1) and 21S11 (Part 1), the licensee has demonstrated that specific Section XI requirements are impractical. It is, therefore, recommended that relief be granted as requested. Granting relief will not endanger life, property, or the common defense and security and is 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.

Pursuant to 10 CFR 50.55a(s)(3),it is concluded that for Relief Requests 11S15, 2 ISI 5,1 1516, 2 ISI 6,1-ISI 7, 21S17, ISPT-01, ISPT-02, ISPT-04, ISPT-05, ISPT-06, ISPT-07, and ISPT-08 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 the proposed alternatives be authorized.

l For Requests for Relief 1 1S13, 2 ISl 3, and ISPT-03, it has been determined that, based on multiple requests for relief submitted by the licensee to accomplish the same j objectives, these relief requests no longer represent a significant burden and relief should not be required. Therefore,it is recommended that relief be denied.

For Relief Requests 1-1S11,2-lSI 1,1-ISI-2 (Part 2), and 2-1S12 (Part 2), the licensee did not provided sufficient justification to support the determination that the Code requirement is impractical or that compliance with the Code requirement would result in hardship. Therefore, it is recommended that relief be denied.

In the August 28,1997, response to an NRC request for additionalinformation, the licensee withdrew Relief Requests 1-ISI 4 and 21S1-4.

This technical evaluation has not identified any practical method by which TVA can meet all the specific inservice inspection requirements of Section XI of the ASME Code for the existing Sequoyah Nuclear Plant, Units 1 and 2. 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 XI 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.

TVA should continue to mor% the development of new or improved examirstion techniques. As improvements v , achieved, TVA should incorporate these tech..iques in the ISI program plan examination requirements.

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\

1 Sased on the review of the Sequoyah Nuclear Plant, Units 1 and 2. Second 10 Year Intervelinservice Inspection Program Plan TVA's responses to the NRC's requests 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 except those noted in the evaluation of Requests for Relief 1-ISI-1,21511,1 ISI 2 (Part 2),

21S1-2 (Part 2),1-ISI 3,2-1S13, and ISPT-03.

37

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• 5

5. REFERENCES i

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 and 1974 Edition through Summer 1975 Addenda.

3. Sequoyah Nuclear Mant, Units 1 and 2, Second 10-Year Intervalinservice Inspection Program Man, submitted November 21,1995.

4. NUREG-0800, Standard Review Man for the Review of Safety Analysis Reports for Nuclear Power Mants, 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 February 14,1996, D. LaBar9a (NRC) to O. D. Kingsley, Jr. (TVA) containing request for additional information.

6. Letter dated May 9,1996, R. H. Shell (TVA) to Document Control Desk (NRC),

containing response 'ISequoyah Nuclear Plant (SON) - AdditionalInformation for Second 10-Year Interval- Inservice inspection (ISI) Program Man) to the NRC RAI dated February 14,1996.

7. Letter dated July 9,1996, R. W. Hernan (NRC) to O. D. Kingsley, Jr. (TVA) containing request for additionalinformation.

8. Letter dated September 6,1996, R. H. Shell (TVA) to Document Control Desk (NRC),

containing response (Sequoyah Nuclear Plant (SQN)- AdditionalInformation for Second 10-YearInterval-Inservice Inspection (ISI) Program Man) to the NRC RAI dated July 9,1996.

9. Letter dated December 17,1996, R. W. Hernan (NRC) to O. D. Kingsley, Jr. (TVA) containing request for additional information.

10. Letter dated March 4,1997, R. H. Shell (TVA) to Document Control Desk (NRC),

containing response lSequoyah Nuclear Mont (SGN)- AdditionalInformation for Second 10-Year interval-Inservice Inspection (ISI) Program Man) to the NRC RAI dated December 17,1996.

11. Letter dated August 28,1997, Pedro Salas (TVA) to Document Control Desk (NRC),

containing response (Sequoyah Nuclear Mont (SON) - AdditionalInformation for Second 10-Year Interval- Inservice inspection (ISI) and inservice Pressure Test (ISPT)

Program Mans] to the NRC RAI dated December 17,1996.

12. NRC Regulatory Guide 1.147, inservice inspection Code Case Acceptability, Revision 11, October 1994 38

c. o

13. NRC Regulatory Guide 1.150, Ultrasonic Testing of Reactor Vessel Welds During Preservice andInservice Examinations, Revision 1, February 1983.

14. NRC Regulatory Guide 1.14, Reactor Coolant Pump flywheellnteprf'/, Revision 1 August 1975.

15. IE Bulletin 7913, Cracking in feedwater System Piping, August 30,1979.

16. IE Bulletin 88 09. Thimble Tube Thinning in Westinghouse Reactors, July 26,1988.

17. S. H, Bush and R. R. Maccary, Development ofIn Service Inspection Safety Philosophy for U.S.A. Nuclear Power Plants, ASME,1971.

4 1

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i 39

f

' NRC Form 335 U.S. Nuclear Rrgulator) Comrmasion 1. REPORT NL'MBER y,cy 3,,,

yy,Nac.ade v.t.s..aw,mid..d i 320s.3202 BIBLIOGRAPHIC DATA SHEET INEEtJEXT a7.e11s5

~ .

2. TITLE AND SUBTMLE

3. DATE REPORT PUBLISHED

, Technical Evaluation Report on the M *" Y" Second 10 Year interval inservice inspection Program Plan: January 1998 Tennessee Valley Authortty, Sequoyah Nuclear Plant, Units 1 and 2 4. m R RANT NUMBER Docket Numbers 50-327 and 50-328 JCN J2229 (TWA-A13)

5. ALTrHOR(S) 6. 'IYPE OF REPORT M.T. Anderson B. W. Brown Technical I

7. PERIOD COVERED m om.)

8. PERFORMING ORGANIZATION . NAME AND ADDRESS nrNaC,guewids Deuuman,00so m Bagmen.U S Nester magubesty r , and medug edeus. sammwear, p swide name and mehg adeus)

INEEL/LMITCO P. O. Box 1625 Idaho Falls, ID 83415-2209

9. SPONSORING ORGANIZATION.NAME AND ADDRES$ntwac. type-same sashwo .desmenner.p id.Nac ommen.oe .am u.s Musema nagiemory comm and samhrg adeens)

Civil and Geosciences Branch Ofrece of Nuclear Regulatory Commission U. S. Nuclear Regulatory Commission Washington D.C. 20555

10. SUPPLEMENTARY NOTE 3 II.ABSTRACTomow *=lms)

This report presents the results of the evaluation of the Sequoyah NuclearPlant, Units 1 and 2, Second 10-YearIntervalInservice Inspection Program Plan, submitted November 21,1995, (

including the requests for relief from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, requirements that the licensee has determined to be impractical. The Sequoyah NuclearPlant, Units 1 and 2, Second 10-YearIntervalInservice Inspection Program Plan, is evaluated in Section 2 of this report. The, Inservice Inspection (ISI)

Program Plan is evaluated for (a) compliance with the appropriate edition / addenda of Section XI, (b) acceptability of the 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 (NRC) reviews. The requests for relief are evaluated in Section 3 of this report.

, 12. KEY WORDS 'DESCR!rTORS rtai sse w penne sa se mum mesmeima m ineme en nysn) 13. AVAILABIUTYSTATEMENT Unlimited

14. SECURITY CLASSIFICATION rne , ) Unclassified m.m.mo Unclassifed

15. NUMBER OF PAGES 16 PRICE