ML20115D962

Technical Evaluation Rept on Second 10 Yr Interval ISI Program Plan:Gpu Nuclear Corp,Tmi Nuclear Station,Unit 1, Docket 50-289

ML20115D962
Person / Time
Site:	Three Mile Island
Issue date:	05/31/1992
From:	Beth Brown, Galbraith S, Porter A EG&G IDAHO, INC.
To:	NRC
Shared Package
ML20115D944	List: ML20115D962
References
CON-FIN-D-6022 EGG-MS-10169, NUDOCS 9210210262
Download: ML20115D962 (31)

Text

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May 1992

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i r 1 s ' t TECHNICAL EVALUATION REPORT j.'

TECHNICAL EVALUATION REPORT ON THE SECOND

/daho 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN:

GPU NUCLEAR CORPORATION, Nat/onal

. . THREE MILE ISLAND NUCLEAR STATION, UNIT 1, Engineenny DOCKET NUMBER 50-289 Laboratory

, 1 Managed .

. by the U.S.- B. W. Brown Department . S. G. Galbraith ofEnergy . A. M. Porter T

i )q EGcG_ Prepared for the  ;

j U.S. NUCLEAR REGULATORY COMMISSION Wart perfonned under DOE Conoact l No. DE-AC07 MID01520 i

I 9210210262 921008 PDR ADDCK 05000289 P PDR l

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EGG-MS-10169

-TECHNICAL EVALUATION REPORT ON THE SECOND 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAN PLAN:

GPU NUCLEAR CORPORATION, THREE MILE ISLAND NUCLEAR STATION, UNIT 1, DOCKET NUMBER 50-289 B. W. Brown S. G. Galbraith A. M. Porter Published May 1992 Idaho National Engineering Laboratory EG&G Idaho Inc.

Idaho Falls, Idaho 83415 Prepared for:

- U.S. Nuclear Regulatory Commission Washington, D.C. 20555 under DOE Contract No. DE-AC07-761001570 FIN No. D6022 (Project 5)

ABSTRACT This report presents the results of the evaluation of the Three #ile Island .

Nuclear Station, Unit 1 (THI-1), Second 10-Year interval Inservice inspection (ISI) Program Plan, Revision 0, submitted April 19, 1991, 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 Inservice Inspection Program Plan is evaluated in Section 2 of this report for (a) compliance with the appropriate edition / addenda of Section XI, (b) acceptability of examination sample, (c) correctness of the application of system or component examination exclusion criteria, and (d) compliance with ISI-related commitments identified during the Nuclear Regulatory Commission's (NRC) previous reviews. The requests for relief are evaluated in Section 3 of this report.

This work was funded under:

U.S. Nuclcer Regulatory Commission FIN No. D6022, Pruject S Operating Reactor Licensing Issues Program, Review of ISI for ASME Code Class 1, 2, and 3 Components ii

SUMMARY

. The Licensee, GPU Nuclear Corporation, has prepared the Three #ile Island Nuclear Station, Unit 1, Second 10-Year Interval Inservice Inspection (ISI)

Program Plan, Revision 0, to meet the requirements of the 1986 Edition of ASME Code Section XI. The second 10-year interval began on April 20, 1991 and ends April 19, 2001.

The information in the Three Mile Island Nuclear Station, Unit 1, Second 10-Year Interval Inservice Inspection Program Plan, Revision 0, submitted April 19, 1991, was reviewed. Included in the review were the requests for relief from the ASME Code Section XI requirements that the Licensee has determined to be impractical. As a result of this review, a request for additional information (RAI) was prepared describing the information and/or clarification required from the Licensee in order to complete the review. The Licensee provided the requested information in the submittal dated December 12, 1991.

Based on the review of the Three Mile Island Nuclear Station; Unit 1, Second 10-Year Interval Intervice Inspection Program Plan, Revision 0, the Licensee's response to the Nuclear Regulatory Commission's RAI, and the recommendations ,

for granting relief from the ISI examinations that cannot be performed to the extent required by Section XI of the ASME Code, it is concluded that the Three Mile Island Nuclear Station, Unit 1, Second 10-Year Interval Inservice Inspection Program Plan, Revision 0, is acceptable and in compliance with 10 CFR 50.55a(g)(4), with the exception of Requests for Relief Nos. 10, 13, 14 (in part), and 15 (in part), as discussed in Section 3.0 of this report.

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CONTENTS ABSTRACT ................................ ii- ,

SUMMARY

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1. INTRODUCTION ............................ I

2. EVALUATION OF INSERVICE INSPECTION PROGRAM PLAN . . . . . . . . . . . 4 2.1 Documents Evaluated ........................ 4' 2.2 Compliance with Code Requirements ................ 4 2.2.1 Compliance with Applicable Code Editions . . . . . . . . . . . 4 2.2.2 Acceptability of the Examination Sample ........... 4 2.2.3 Exemption Criteria . . . . . . . . . . . . . . . . . . . . . . 5 2.2.4 Augmented Examination Commitments .............. 5 2.3 Conclusions ........................... 5

3. EVALUATION OF RELIEF REQUESTS . . . . . . . . . . . . . . . . . . . . 7 3.1 Class 1 Components . . . . . . . . . . . . . . . . . . . . . . . . 7 3.1.1 Reactor Pressure Vessel .......... ........ 7 3.1.1.1 Request for Relief No. 1, Examination Category B-F, -

Item B5.130 and Examination Category B-J, Item B9 ll, Dissimilar Metal and Terminal End Piping Welds . . . . . . 7 3.1.1.2 Request for Relief No. 7. Examination Category F-A, Item F1.40, Reactor Vessel Support Skirt . . . . . . . . . 9 3.1.2 Pressurizer ......................... 10 3.1.3 Heat Exchangers and Steam Generators . . . . . . . . . . . . . 10 3.1.3.1 Request for Relief No. 2, Examination Category B-H, Item B8.30, Steam Generator Integrally Welded Attachments ........................ 10 3.1.3.2 Request. for Relief No. 3, ' Examination Category B-B, Item B2.70, Letdown Cooler (primary side) -

Longitudinal Shell Welds . . . . . . . . . . . . . . . . . 11 ,

3.1.4 Piping Pressure Boundary . . . . . . . . . . . . . . . . . . . 13 3.1.5 Pump Pressure Boundary . . . . . . . ............. 13 iv

3.1.5.1 Request for Relief No. 8, Examination Category B-L-2, Item B12.20, Reactor Coolant Pump Casings ........ 13 3.1.6 Valve Pressure Boundary ................... 14 3.1.6.1 Request for Relief No. ll, Examination Category B-M-2, Item. B12.50, Valve Body Internal Surfaces ........ 14 3.1.7 General ........................... 17 3.1.7.1 Request for Relief No.13 (Part 1 of 3), Paragraph IWB-2430, Additional Examinations ............ 17 3.2 Class 2 Compor. cats . . . . . . . . . . . . . . . . . . . . . . . . 19 3.2.1 Pr essure Vessel s . . . . . . . . . . . . . . . . . . . . . . . 19 3.2.1.1 Request for Relief No. 5, Examination Category C-8, Item C2.32, Decay Heat Removal (DHR) Coolers Nozzle-to-Vessel Welds ..................... 19 3.2.1.2 Request for Relief No. 6, Examination Category C-A, Item C1.10, DHR Cooler Shell Welds . . . . . . . . . . . . 21 3.2.2 Piping . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 3.2.2.1 Request for Relief No. 4, Examination Category C-F-1, Class 2 Piping Welds . . . . . . . . . . . . . . . . . . . 23 3.2.2.2 Request for Relief No. 9, Open Ended Class 2 Piping ... 24 7 2.3 Pumps ............................ 26 .

3.2.4 Valves . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 3.2.5 General ........................... 26 3.2.5.1 Request for Relief No.13 (Part 2 of 3), Paragraph IWC-2430, Additional Examinations ............ 26 3.3 Cl as s 3 Components . . . . . . . . . . . . . . . . . . . . . . . . 28 3.3.1 Piping . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 3.3.2 Pumps ............................ 28 3.3.3 Valves............................ 28 3.3.4 General ........................... 29 3.3.4.1 Request for Relief No.13 (Part 3 of 3), Additional Examination Requirements for Class 3 Components ..... 29 3.4 Pressure Tests . . . . . . . . . . . . . ............ 31 v

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3,4.1 Class 1 System. Pressure Tests' . . . . . . . .:.;. . . . . . .-

31 3.4.1.1 -Request for Relief No. 16.4, Examination Category B-P,-

Item B15.51, Hydrostatic Test of the Class 1 Reactor '

Coolant System-(Auxiliary Pressurizer. Spray) . . . . . , . .

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'3.4.2 Class 2. System Pressure Tests ................ 32 3.4.2.1 Request for Relief No.16.1, Examination Category C-H, Item C7.40, Hydrostatic isst of Class.2 Hakeup and i Purification System Piping . . . . . . .: . . . . . . . . . 32-3.4.2.2 Request 1for Relief No. 16.2, Examination Category C-H, Item C7,40, Hydrostatic Test o' Class 2 Core Flooding-System Piping ...................... 134~

3.4.2.3 Request for Relief No. 16.3,_ Examination Category C-H, Item C7.40, Hydrostatic Test of Class 2 DHR System Piping . . . . . . . . . . . . . . . . . . . . . . . . . .- 36 3.4.2.4 Reques' for Relief No.16.5, Examination Category C-H, Item C7.40, Hydrostatic Test of Class 2 Hakeup and Purification System Piping . . . . . . . . . . . . . . . . -

37 3.4.3 Class 3 System Pressure Tests ................ 38 3.4.4 General ........................... 39 3.4.4.1 Request for Relief No. 15, Paragraph IWA-5242(a), VT-2 Examination of Bolted Connections ............ 39 3.4.4.2 Request for Relief No. 14, Paragraph IWA-5250(a)(2), ,

Corrective Measures for Bolted Connections . . . . . . . . 41 3.5 General . . . . . . . . . . . . . . . . . . . . . . . ... . . . . . -

45 3.5.1 Ultrasonic Examination Techniques (No relief requests) 3.5.2: Exempted Components ................-..... 45 3.5.3 Other . . . . . . . . . . . . . . . . . . . . . . . . .. . . .. 45-3.5.3.1 Request for Relief No.10, Use of Form NIS-2 . . . . . . . 45 3.5.3.2 Request for Relief No.12, Categorization and 4

Selection of Component Supports . . . . . . . . . . . . . 48 -

4. CONCLUSION ............................. 50

5. . REFERENCES ....-........................ 52 vi

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-TECHNICAL EVALUATION' REPORT:0N THE:

SECOND 10-YEAR INTERVAL INSERVICE INSPFCTION PROGRAM PLANr GPU NUCLEAR CORPORATION, THREE MILE ISLAND NUCLEAR STATION, UNIT 1, 1 DOCKET-NUMBER 50-289--

1. INTRODUCTION Throughout the service life of a water-cooled nuclear power facility, 10 LFR 50.55a(g)(4) (Reference 1) requires that. components (including supports) that are classified a:;- American Society of Mechanical 1 Engineers (ASME) Boiler and Pressure Vessel Code Class 1, Class 2, and Class 3 meet the.

requirement's,.except the design and access provisions and the pres'ervice examination requirements, set forth in the ASME Code Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components" (Reference 2), to the extent practical within_ the limitations of design, geometry, and materials of

-construction of the components. This section of the regulations also requires that inservice examinations _of_ components and system pressure tests' conducted during successive 120-month inspection intervals comply with the requirements -

in the latest edition and addenda of the Code incorporated by reference in 10 CFR 50.55a(b) on the.date 12 months prior to'the start of the 120-month inspection interval . subject to the limitations and modifications listed --

therein. The components (including supports) may meet. requirements set _ forth l 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. The Licensee, GPU Nuclear-(GPUN) Corporation, has prepared i the Three Mile Island Nuclear Station, Unit 1, Second 10-Year Interval- l Inservice-Inspection (ISI)' Program Plan, Revision 0-(Reference 3), to meet the requirements of the 1986 Edition of the ASME Code-Section XI. The second 10-year interval began ~ April 20, 1991 and ends April' 19, 2001. d

- Astreq'uired-by-10 CFR h55a(g)(5), if the licensee determines that certain-Code examination requirements-are-impractical and requests relief from them, the ~ licensee'shall submit information and justifications to the Nuclear Regulatory Commission (NRC) to support that determination.

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Pursuant to 10 CFR 50.55a(g)(6), the NRC will evaluate the licensea's determination that Code requirements are impractical to implement.

Alternatively, pursuant to 10 CFR 50.55a(a)(3), the NRC will evaluste the .

Licensee's determination that either (i) the proposed alternatives provide ar.

acceptable level of quality and safety or that (ii) Code compliance would result in hardship or unusual difficulty without a compensating increase in safety. The NRC may grant relief and may impose alternative requirements that are determined to be authorized by law, will not endanger life, property, or the common defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.

Ths information in the Three Mile Island Nuclear Station, Unit 1, Second 10-Year Interval ISI Program Plan, Revision 0, submitted April 19, 1991, was reviewed, including the requests for relief from the ASME Code Section XI requirements that the Licensee has determined to be impractical. The review of the ISI Program Plan was performed using Standard Review Plan NUREG-0800 (Reference 4), Section 5.2.4, " Reactor Coolant Boundary Inservice Inspections and Testing," and Section 6.6, " Inservice Inspection of Class 2 and 3 Components."

In a letter dated September 19, 1991 (Reference 5), the NRC requested additional information that was required in order to complete the review of the ISI Program Plan. The requested information was provided by the Licensee in the response to request for additional information regarding the second ten year inservice inspection interval dated December 12, 1991 (Reference 6). In this response, the Licensee, GPU Nuclear Corporation, committed to the volumetric examination of 7.5Y. of the non-exempt Class 2 piping welds where the wall thickness does not meet the minimum requirements for examination in Table IWC-2500-1, Examination Category C-F-1 (i.e., wall thickness <3/8").

The Three Mile Island Nuclear Station, Unit 1, Second 10-Year Interval ISI Program Plan is evaluated in Section 2 of this report. The ISI Program Plan is evaluated for (a) compliance with the appropriate edition / addenda of Section XI, (b) acceptability of examination sample, (c) correctness of the rpplication of system or component examination exclusion criteria, and 2

(d) compliance with ISI-related comitments 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, 1986 Edition. Specific inservice test (IST) programs for pumps and valves 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 license conditions pertinent to ISI activities. This section describes the submittals reviewed and the results of the review.

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

(a) Three Nile Island Nuclear Station, Unit 1, Second 10-Year Interval Inservice inspection Program Plan (Revision 0), dated April 1991; and, (b) Letter, dated December 12, 1991, response to tne NRC request for additional information.

2.2 Compliance with Code Reouirements 2.2.1 Comeliance with Apolicable Code Eoitions The Inservice Inspection Program Plan shall be based on the Code .

editions defined in 10 CFR 50.55a(g)(4).and 10 CFR 50.55a(b). The Code applicable to the second 10-year ISI interval, based on the starting date of April 20, 1991, is the 1986 Edition. The Licensee has prepared the second interval 10-year ISI program, Revision 0, to meet-the 1986 Edition.

2.2.2 Acceptability of the Examination Samole Inservice volumetric, surface, and visual examinations shall be performed on- ASME Code Class 1, 2, and 3 components and their supports using sampling schedules described in Section XI of the ASME Code and 10 CFR 50.55a(b). Sample size and weld selection have been imnlemented in accordance with the Code and 10 CFR 50.55a(b) and appear to be correct.

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i 2.2.3 Exemotion Criteria i

I 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 ,

l t,+ micensee in accordance with the Code as discussed in the ISI program plan, and appear to be correct.

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

(a) The reactor pressure vessel examinations will comply with the requirements of Regulatory Guide 1.150, Revision 1 (Reference 7);

(b) Examinations of the reactor coolant pump motor flywheel assembly will be performed in accordance with THI-1 Technical Specification 4.2.4; (c) Examinations of certain main steam system welds outside of containment will be performed in accoroance with THI-l Technical Specification 4.15. This requires examination of welds MS-0001, ,

MS-0002, MS-0003 and MS-0004L at 3-1/2 year intervals or the nearest refueling outage.

(d) All load bearing welds will be examined over the normal inservice inspection interval using standard ISI techniques for the head and internals handling- fixture (Tripod) using NUREG-0612, Control of Heavy Loads at Nuclear Power Hants (Reference 8); and (e) Examinations of main feedwater Welds FW-0034 through FW-0039 will be performed every interval because of postulated break analysis to assure compliance with the h ak exclusion requirements as specified in ParagrapF. B.I.b (7) of Branch Technical Position MEB 3-1 (Reference 9).

2.3 Conclusions Based on the review of the documents listed above, it is concluded that the Three Mile Island Nuclear Station, Unit 1, Second 10-Year Interval 5

ISI Program Plan, Revision 0, is acceptable and in compliance with 10CFR50.55a(g)(4).

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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 CcmooneD11 3.1.1 Etactor Pressure Vessel 3.1.1.1 Reauest for Relief No.1. Examination Cateoory B-F. Item B5.QQ and Examination Cateaory B-J. Item B9.11. Dissimilar Metal and Terminal End Pioina Welds Code Reauirement: -Section XI, Table IWB-2500-1, Examination Category B-F, Item B5.130, and Examination Category B-J, Item B9.ll, requires both surface and volunietric examination of all dissimilar metal piping welds and terminal end piping welds at vessels as defined by Figure IWB-2500-8. ,

licensee's Code Relief Peouest: Relief is requested from performing 100% of the Code-required surface examination for Welds CF-0001, CF-0020, RC-0001, RC-0052, RC-0106, RC-0054, RC-0087, RV-0009BM, and RV-0010BM.

Licensee's Basis for Reauestina Relief: Relief is requested from performing the surface examinations because the above listed welds are-located inside the reactor vessel primary shield wall and performing these examinations would require removal. of sand plugs and insulation to gain access into this high radiation environment. GPUN estimates that a cumulative exposure of 87 Person-Rem would be necessary to complete the Code-required examinations.

Licensee's Proposed Alternative Examination: GPUh vill perform ultrasonic examination of the OD surface of these w lds from the 10 using techniques qualified by Babcock and Wilcox.

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6 Evaluation: The Licensee has stated'that the required surface examinations of the subject welds cannot be performed without i excessive personnel radiation exposure. The proposed alternative ,

of a voluntric inspection from the ID _is acceptable provided the Licensee meets the following conditions:-

(1) The remote volumetric examination includes the entire weld volume and heat affected zone instead of only the inner one-third of the weld.

(2) The ultrasonic testing instrumentation and procedures are demonstrated to be capable of detecting 00 surface connected defects, in the circumferential orientation, in a laboratory test block. The defects should be cracks .d not machined notches.

The proposed alternative examination, with the above supplemental conditions, will provide reasonable assurance that unallowable inservice flaws have not developed in the subject welds or that they will be detected and repaired prior to return of the plant to service.

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Conclusions:==

It is concluded that the Code-required surface examination of the subject welds would result in personnel radiation exposures contrary to ALARA principles. The examination is therefore impractical to perform at THI-1. In the December 12, 1991 response to the NRC request for additional information, the Licensee committed to meet the above stated conditions. When these conditions are met, the proposed alternative volumetric examination from the ID of the pipe will provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(g)(6)(1), it is recommended that relief be granted as requested. -

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e c, 3.'l.1.2 - Reauest - for Relief No. 7. ' Examination Cateaorv- F-A. Item F1.40. j R(actor- Vessel Suonort Skirt

- Code Reauirement: Section XI, Table-IWF 2500-1,-Examination.

Category F-A, Item F1.40 requires a 100% visual examination- 3 (VT-3) of pl_ ate- and shell-type supports to the extent indicated in paragraphs IWF-1300 and IWF-2510 and as defined by Figure-IWF-1300-1.

Licensee's Code Relief Reauest: Relief is. requested from -

performing 100% of the Code-required VT-3 examinati n on the-outside surface and approximately 50% on the inside of the reactor-vessel support skirt.

Licensee's Basis for Reauestino Relief: The Licensee states that access to the inside surface is obtained-by entering a-tunnel leading to the reactor vessel lowr head and incore instrument >

- guide tubes. The entire lower head and approximately half of the support shirt are obstructea by mirror insulation when accessed from beneath the reactor vessel. The'inside anchor area and a portion of the support _ skirt are visible using.a ' remote hand held.

camera for about 50% of the circumference. The remai_ning-circumferance, where the incore guide tubes exit the-reactor shield : area,- is inaccessible because. of the incore guide tubes. -

Access to-the outside surface of the skirt could only be_obtained g from the top of the reactor vessel by lowering remote camera equipment between the reactor vessel and insulation and between-

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l the primary shield wall and insulation. This approach requires removal of sand plugs and insulation. It would be necessary to

,- - navigate between insulation supports using remote crawlers of some type in order to view the anchor area.

L Licensee's ProDosed Alternative Examinatioq: 3erform a remote visual examination of the accessible areas of the support skirt inside surface.

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ti Evaluation: The Licensee's submittal has been reviewed,

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including the drawing which shows the examination limitations.-

. Based.on the design of the RPV support skirt, it is concluded ,- ,

-that visual examination to the extent required by?the Code-is ,

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impractical and would result in hardship or unusual difficulty.

without a compensating increase in the level of quality and safety. A recate visual examination-of the accessible portion of ,

the support skirt will give a rossonable assurance of continued 1 inservice structural integrity.

Conclusions:

Based on the -above evaluation, the impracticality of performing the Code-required examinstion, and pursuant to 10 CFR 50.55a(g)(6)(i), it is recommended that relief be granted as requested.

3.1.2 Pressurizer (No relief requests) 3.1.3- Heat Exchancers and Steam Generators 3.1.3.1 Reauest for Relief No. 2. Examination Cateoory B-H. Item B8.30.

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Steam Generator Inteorally Welded Attachments Code Reouirement: Section XI, Table IWB-2500-1, Examination "

Category B-H, Item 88.30 requires a 100% surface-examination as defined by Figure IWB-2500-13.

Licensee's Code Relief Reauest: Relief is requested from performing the Code-required surface examination of the area labeled C to D (Figure IWB-2500-13)-for Welds-SG-0006:and SG-0013 on Steam Generators RC-H-1A 'and!kC-H-1B, respectively. -

Licensee's Proposed Alternative Examination:

None. The Licensee proposes to perform a Code-required surface examination on area A to B and a visual examination of area C to D.

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6- . 4, Licensee's Basis for Reauestina Relief: The:1.icensee states that the steam generator is designed in such a manner-that access to

. this_. area for:a meaningful examination it impractical.

Substantia 1Lsurface^ preparation would be required -to: perform a surface examination and the weld face reduces to a-1-3/4 inchl  !

radius,. restricting access for weld preparation equipment-and-personnel. t

. Evaluation: Detail "X" of Drawing 131117E shows the restrictiorc associated with' the .inside surface of integral attachment Welds CG-0006 and SG-0013 on Steam Generators RC-H-1A and RC-H-18. The' 1-3/4 inch radius at the inside weld face limits access for personnel and. surface-preparation equipment. - The steam generator design makes the Code-required surface-examination of the inside-weld surface impractical to perform. Imposition of the Code I

requirement on GPUN would cause a burden that would not be compensated significantly by an increase in safety above that provided by the propos'ed examination.  !

Conclusions:

It is concluded that the combination of the Code-required surface examination.to the-outside of these welds and a 2 visual examination of the inside surface will provide reasonab1_e assurance of continued inservice integrity of the subject welds.

'Therefore, pursuant to 10 CFR 50.55a(g)(6)(1), it-is recommended that relief be granted as requested.

3.1.3.2 Reauest for Relief No. 3. Examination Cateoorv B B. Item B2.70.

Letdown Cooler f or' nary ~ side) - lonaitudinal Shell Welds Code Reauirem d : Section XI, Table IWB-2500-1, Examinatior. ~

Category B-B, Item B2.70 requires a 100% volumetric examination Lof 1. foot of one longitudinal weld at each end of the shell, at-the intersection of the circumferential weld, as defined by

-Figure IWB-2500-2. The examination-may be limited to one vessel among a group of vessels performing a-similar function.

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Licenseo's Code Relief Reauest:. Relief is requested from '

L ' performing the volumetric examination'of-one-foot at each end of-longitudinal Welds MU-1019L, MU-1020L, MU-1021L,-MU-1022L, MU-1034L, MU-1035L,- MU-1036L and MU-1037L on Letdown Coolers-

-MU-C-1A and MU-C 18.-

Licensee's Basis-for-Reauestina Relief: The Licensee states that the letdown coolers are of a helical coil design _using a split manifold to allow for tube expansion and seal aldit.g prior to ,

completing. manufacture of the manifold. The tube 8 undle and ~ all but approximately 1-1/4 inch of the inlot and 2 ~ 4 inch of. the outlet manifolds are then covered by-the Class 1(heat sink) side -

of the heat exchanger. Accest is limited to the short section of manifold outboard of the heat sink side of the vessel.

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Licensee's Procosed Alternative Examination: None. GPUN proposes to volumetrically examine the accessible length of all manifold welds on one cooier during the 10-year interval.

Evaluation: The drawings submitted by the Licensee have been reviewed and it has been determined that the volumetric th examination of one foot of longitudinal weld at each end of the \

shell welds is impractical to perfirm due to inaccessibility.

The accessible portions of the subject welds are the most highly ,

stressed areas-and will receive the Code-required volumetric examination. In order to complete the volumetric examination-to the extent required by the Code, the letdown coolers would require extensive modifications. The inservice' structural integrity will be assured by the performance of a volumetric examination to accessible portions of the subject manifolds.

Conclusions:

It is concluded that the volumetric examination of one foot of longitudinal ' weld on each end of the letdown cooler .

manifolds is impractical to perform at Three-Mile Island, Unit 1,

-to the extent required by the Code. The granting of relief will not endanger life, property, or the common defense and security and is:otherwise in the public interest, considering the burden 12

upon t'he. facility. Therefore, pursuant to-10 CFR 50.55a(g)(6)(1), it is recomended that relief be granted as requested.

3.1.4 Pioina Pressure Boundary (No relief requests) 3.1.5 Pumo Presture Boundary 3.1.5.1 Reauest for Relief No. 8. Examination Cateoorv B-t-2. Item B12.20. Reactor Coolant Pumo Casinas Code Reauirend: Section XI, Table IWB-2500-1, Examination.

Category B-L-2, Item B12.20 requires a 100% VT-3 visual I

examination of the internal surfaces of Class 1 pump casings by the end of the interval. The examination is limited to at least one pump in each group of pumps performing similar functions in the system.

Licensee's Code Relief Reauest: Relief is requested from performing the Code-required VT-3 visual examination of the

~

reactor coolant pump internal surface.

Licensee's Basis for Reauestina Relief: The Licensee believes that the Code requirement to disassemble a rea tor coolant pump strictly for the purpose of visual examination does not provide an increase in safety commensurate with the resultant personnel radiation exposure associated with the inspection.

Licensee's Proooted Alternative Examination: Visual examination of.the casing interior surface of one reactor coolant pump will be performed only if a pump is disassembled for repair or maintenance.

Evaluation: The visual examination is performed to determine if unanticipated severe degradation of the casing is occurring due to phenomena such as erosion, corrosion, or cracking. However, 13

. . .. \

experience with similar pumps at other plants has not shown any significant degradation of pump casings. Because of this, later editions and addenda of the ASME Code (1988 Addenda) state that only pumps that are disassembled for reasons such as maintenance, 4 repair, or volumetric examination must have their internal )

surfaces examined.

The disassembly of the reactor coolant pumps for the sole purpose of visual examination af the casing internal surfaces is a major effort and requires many manhours fn.' skilled maintenance and inspection personnel. The possibility of damage to the pump and the excessive radiation exposure would not have a compensating increase in safety.

Since no major problems have been reported in the industry with regard to pump casings, the Licensee's proposal will provide adequate assurance of the continued inservice structural integrity of the Reactor Coolant Pumps.

Conclusioni: The disassembly of a pump for the sole purpose of inspection is impractical to perform at THI-1. It is concluded '

that public health and safety will not be endangered by allowing the proposed examination to be performed in lieu of the Codo

! requirement. Therefore, pursuant to 10 CFR 50.55a(g)(6)(i), it is recommended that relief be granted provided that, if a pump j has not beer, disassembled, the Licensee should report this fact

in the ISI Summary Report at the end of the interval.

I 3.1.6 Valve Pressure Boundary 3.1.6.1 Reouest for Relief No. 11. Examination Cateoory B-M-2. Item B12.50. Valve Body Internal Surfaces Code Reouirement: Section XI, Table IWB-2500-1, Examination Category B-M-2, Item B12.50 requires a VT-3 visual examination of the internal surfaces of valve bodies exceeding 4-inch nominal 14

pipe size. This examination is limited to one valve within each group of valves that are of the same design and manufacturing method. and that perform similar functions in the system.

l Licensee's Code Relief Reauest: Relief is req"*Sted from performing the required VT-3 visual examination for valves DH-V1, DH V2, DH-V22A/B, CF-V4A/B, and CF-V5A/B unless they are disassunbled for maintenance.  :

Licensee's Basis for Reauestino Relief: The Licensee reports that the design of the valves is such that disassembly is Nguired to perform the examination. Disassembly of valves DH-V1 and DH-V2 would require full core offload. In addition, the refueling canal water would have to be drained or a freeze seal would have to be used in order to disassemble DH-V1. Lowering the refueling canal water level cob 1d extend an outage by approximately two days and would add approximately four person-rem to an outage. The Licenses states that use of a freeze seal on such a large pipe so close to the reactor vessel would not be advisable.

The Licensee also states that the CF-V4A/B, CF-VSA/B, and DH-V22A/B inspections could be performed without core offload but there would be an increased potential for a loss of decay heat if performed while the RCS is in a drained condition. IfCF-V5A/B were chosen to be disassembled while in a drained condition, a freeze seal would still be required because of their proximity to the RCS.

The Licensee points out that the 1989 Edition of Section XI requires these examinations only if a valve is disassembled for maintenance, repair, or volumetric examination. None of these valve designs utilize a valve body weld.

Licensee's Proposed Alternative ExaminatiQn: None. The Section XI required visual examinations of Valves DH-V1, DH-V2, 15

DH-V22A/B, CF V4A/8, and CF-VSA/B will be performed only if the valve is disassenbled for maintenance or repair.

Evaluation: The visual examination is performed to determine if unanticipated degradation of the valve body is occurring due to phenomena such as erosion, ccrrosion, or cracking. However, previous experience during examination of similar valves at other plants has not shown any significant degradation of valve bodies.

In order to examine the internal surfaces of a valve body in a' ordance with the requirements, complete disassembly of the valve would be required which, in addition to the possibility of damage to the valve, could result in personnel receiving excessive radiation exposure. Imposition of this requirement on GPUN would cause a burden that would not be compensated significantly by an increase in safety above that provided by the proposed examination.

Later editions and addenda of the ASME Code (1988 Addenda) state that the internal surface visual examination requirement is only applicable to valves that are disassembled for reasons such as

~

maintenance, repair, or volumetric examination.

Conclusions:

It is concluded that the disassembly of a valve for the sole purpose of inspection is impractical at TM1-1, and that compliance would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Therefore, pursuant to 10 CFR 50.55a(g)(6)(1), it is recommended that relief be granted provided that if a valve has not been disassembled, this fact should be reported by the Licensee in the ISI Sumary Report at the end of the inter tal.

I l-16 l

c I

3.1.7 General l 1

3.1.7.1 Reouest for Relief No. 13 (Part 1 of 3). Paraoraoh IWB 2430.

Additional Examinations ,

Code Reouirement: Section XI, Paragraph IWB-2430 states, (a) Examinations performed in accordance with Tsble IWB-2500-1 that reveal indications exceeding the acceptance standards of Table IWB 34101 shall be extended to include additional ,

examinations at thir outage. The additional examinations shall include the remaining welds, areas, or parts included in the inspection item listing and scheduled for this and the subsequent period. If the examinations for that inspection item are not scheduled in the subsequent period, the most immediate period containing scheduled examinations shall be taken as the s& sequent period.

(b) If the additional examinations required by (a) above reveal indications exceeding the acceptance standards of Table IWB-3410-1, the examinations shall be further extended to include additional examinations at this outage. The additional examinations shall include all the welds, areas, or parts of similar design, size, and function.

I (c) For the inspection period following the period in which the examinations of (a) or (b) above were completed, the examinations shall be performed as normally scheduled in accordance with IWB-2400.

Licensee's Code Relief Reouest: Relief is requested from performing Additional Examinations as required in IWB-2430 of the 1986 Code.

Licensee's Basis for Reouestino Rel.igf: The Licensee states that the requirements in the 1986 Code are very prescriptive and could actually not allow credit for examination of components that should be examined as part of an expanded sample. ASME has 17

recognized that these situations can exist and is in the process of changing the Code additional examination requirements. These changes allow selection of additional examinations baseri on similar materials and service thus the root cause of the condition requiring additional examinations is considered when expanding examination samples.

Licensee's Proposed Alternative Examination: The Licensee proposes to perform additional examinations based on proposed changes to Paragraph IWB-2430 of Section XI, as follows:

(a) Examinations performed in accordance with Table IWB-2500-1, except for Examination Categories B-E and B-P, that reveal flaws or relevant conditions exceeding the acceptance standards of Table IWB-3410-1 shall be extended to include additional examinations at this outage. The additional examinations shall include an additional number of welds, areas, or parts included

'n th6 inspection item equal in' number to the welds, aress, or p rts included in the inspection item that were scheduled to be performed during the present inspection period. The additional examinations shall be selected from welds, areas, or parts of similar material and service. This additional selection may require the inclusion of piping systems oth e than the one containing the flaws or relevant conditions.

(b) If the additional examinations required by (a) above reveal flaws or relevant conditions exceeding the acceptance standards of Table IWB-3410-1, the examinations shall be further extended to include additional examinations at thic outage. These additional examinations shall include the remaining number of welds, areas, or parts, included in the inspection item of similar material and service subject to the same type of flaws or relevant conditions.

(c) For the inspection perioo rollowing the period in which the examinations of (a) or (b) above were completed, the examinations 18

= ,

shall be performed as normally scheduled in accordance with IWB-2400. l (d) For steam generator tubing, additional examinations shall be governed by the Plant Technical Specification. I Evaluation: The proposed alternative to paragraph IWB 2430 appears in the 1991 Addenda. Although the proposed alternative may have technical merit, the NRC staff has not yet reviewed and endorsed th s change. Approval of this request for relief before NRC staff review would undermine the review process.

.)

Conclusions:

Based on the unapproved nature of the 1991 Addenda, it is recommended that this specific request for relief from the additional examination requirements of the ASME Code be denied.

3.2 Class 2 Componerts 3.2.1 Eressure Vessels 3.2.1.1 Reauest for Relief No. 5. Examination Cateoory C-B. Item C2.32.

Decay Heat hemoval (DHR) Coolers Nozzle-to-Vessel Welds Code Reauirement: Section XI, Table IWC-25001, Examination Category C-8, Item C2.32 requires a volumetric examination af the inner 1/3 wall thickness of nozzle welds and the nozzle inner radii as defined by Figure IWC-2500-4(c) when the inside of the vessel is accessible.

Licensee's Code Relief Reauest: Relief is requested from p forming the Code-required volumetric examination of welds DH 035. B, DH-0397B, DH 0401B, and DH-04038 and the associated inner radii on Decay Heat Removal Coolers DH C-1A and DH-C-1B.

Licensee's Basis for Reauestino Relief: The Licensee states'that nozzles of this design are difficult to examine using ultrasonic 19

L testing (UT) due to their inherently complex three dimensional nature and scan surface limitations, in addition to the complex geometry, the Licensee refers to the material of construction, 3n4 series stainless steel, which causes additional attenuation anu beam re-directional problems.

Licensee's Proposed Alternative Examinati.QD: A liquid penetrant examination (PT) of these nozzle welds and inner radii will be performed if the coolers are dissembled for maintenance.

EvaluatiGD: TM Code requires that the nozzle-to-shell (or head) welds and assvin k w e 7dit receive a volumetric examination when the insit i of th sh .; accessible. The Licensee proposes to per w o i v face er, amination (PT) of the inside of the nozzles if the co;1ers are disassembled for maintenance.

In the response to the NRC's request for additional information, dated December 12, 1991, the Licensee explained the impracticality associated with this Code requirement at THI-1.

Although the volumetric examination is nossible to perform, the radiation levels inside the DHR Cooiers are estimated to be 16 R/hr. The manual ultrasonic uamination is expected to require about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of scanning time inside the cooler. A liquid penetrant examination of the area of interest would require about ten minutes of time in the cooler. Based on the ALARA considerations associated with the examination of the interior of the DHR Coolers, the liquid penetrant examination would minimize personnel exposure. Since the objective of the required examination is to detect service induced flaws initiating at the ID, a PT examination would provide a reasonable assurance of the continued inservice structural integrity.

Conclusions:

In order to perform the Code requirement the .

Licensee would have to develop an automated data acquisition and analysis system with a remote operated manipulator. This technology is not considered to be reasonably available for such a specific nozzle design, therefore the Code requirement is 20

impractical and imposition would create a burden that would not be compensated by an increase in safety above that provided by the proposed alternative. It is recommended that, pursuant to 10 CFR 50.55a(g)(6)(i), relief be granted as requested. >

3.2.1.2 Egauest for Relief No. 6. Examination Cateaory C-A. Item C1.10.

DHR Cooler Shell Welds Code Reauirement: Section XI, Table IWC-25001, Examination Category C-A, Item C1.10 requires a 100% volumetric examination, as defined by Figure IWC-2500-1, of the heat exchanger shell welds at gross structural discontinuities.

Licensee's Code Relief Recuest: Relief is requested from performing 100% of the Code-required volumetric examination of welds DH 0399 and DH 0404 on DHR Coolers DH-C-1A and DH-C-1B.

Licensee's Basis for Reauestina Relief: The subject retaining ring to shell weld is covered by a sliding flange. The Licensee states that this design makes examination of these. welds impractical unless the cooler is disassembled. Even if -

disassembled, a meaningful volumetric examination could only be performed from the inside of the head. Ultrasonic examination from inside of the head would require personnel to remain in this high contamination and high dcse rate environment much longer than if a liquid penetrant examination were performed on this weld.

The Licensee reports that disassembly and inspection of this cooler to accommodate required examinations would require approximately four man weeks of labor and exposure of 4.4 Person-Rem. Disassembly of the cooler would require a decay heat removal-system train to out of service for an extended length of time.

21

a licensee's Procosed Alternative Examination: The Licensee  !

proposes to perform a liquid penetrant (PT) examination of the ID and OD surface of the weld on one cooler should one or both require disassembly for maintenance or repair during the interval. In addition, a VT 2 examination of the weld area will be performed, once each period, on each cooler while the cooler is in service.

Evaluation: The Code requires that one of the retaining ring to shell welds on the DHR coclers receive a volumetric examination during the 10-year interval. A volumetric examinatio'. of this '

weld is impractical because of the design and the high radiation _;

levels estiinated to exist inside the cooler. With the sliding flange design, the weld is inaccessible for ultrasonic examination in the assembled position, and essentially inaccessible when disassembled. A sketch of this area was previded with the Licensee's response to the NRC's request for additional information, dated December 12, 1991.

If disassembled, the only access to this weld for ultrasonic examination would be from the inside surface. The estimated dose

~

rate on the inside of the coolers is 16 R/hr. A volumetric examination of this weld would require approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of personnel-time at the weld. The Licensee's proposed alternative is to perform a liquid penetrant (PT) examination of both the 10 and OD. surfaces of the weld if one of the coolers requires disassembly for maintenance or repair during the interval.

Service induced defects would initiate from the surfaces of the weld, therefore the proposed alternative would provide reasonable assurance of the continued inservice structural integrity of-the weld.

Conclusions:

It is concluded that the Code-required volumetric examination is impractical for THI-1 and imposition would result in the Licensee having to redesign and/or replace the DHR l Coolers, thus creating a burden without a compensating increase in safety above that provided by the proposed alternative.

l 22 t

\

Therefore, pursuant to 10 CFR 50.55a(g)(6)(i), it is recomended that relief be granted as requested.

3.2.2 Pioino 3.2.2.1 Recuest for Relief No. 4. Examination Cateoory C-F-1. Class 2 Pipino Welds Code Reouirement: Section XI Table IWC-2500-1, Examination Category C-F-1, requires that piping welds greater than 4 inches in diameter but less than 0.375 inch wall thickness in non-exempt (IWC-1220) portions of systems be included in the total weld population, but not be examined.

For piping welds greater than 4 inches in diameter and greater than 0.375 inch wall thickness, both surface and volumetric examination is required for those welds requiring examination, as defined by figure IWC-2500-7.

Licensee's Code Relief Recuest: Relief is requested to substitute a sampling of seven welds with wall thicknesses less than 0.375 inch for welds 0.375 inch thick and greater _ in the decay heat removal system.

Licensee's Basis for Reguestino Relief: The Licensee states that complete ebmpliance with these Code requirements would leave a large portion of the non-exempt Class 2 DHR System piping from the DHR pumps downstream unexam.:dd. GPUN believes this portion of the system is important enough to plant safety that examination of these welds is warranted, licensee's Proposed Alternative Examination: GPUN proposes to categorize and examine non-exempt thin wall piping (greater than 4 inch NPS and-less than 0.375 inch wall thickness) from DH V4A and B to DH-PIA and B as if they were greater than 0.375 inch 23

wall thickness. These welds will be included in the 7.5% total number of welds to be examined without increasing the sample size for the Decay Heat Removal System.

Eyaluation: The Licensee's submittal has been reviewed, including the tables showing the total number of welds, by item

• number, of the Class 2 systems. The effect of this request for relief can be seen on the distribution of the 7.5% weld sample through the DHR System. Seven welds not required to be examined, based on wall thickness, will be surface and volumetrically examined and subtracted from the required weld sample size. The NRC requires that entire engineered safety syst^ms (i.e., RHR, CHR, ECCS) will not be excluded from volumetric examination based upon wall thickness. When entire systems are excluded, augmented volumetric examinations are required. In the response to the NRC's request for additional information, dated December 12, 1991, the Licensee expanded the potential sample population to include previously excluded sections of pipiag.

Since the DHR System is receiving the required total number of weld examinations, a reasonable assurance of the continued inservice structural integrity of the system will be retained.

Conclusions:

Based on the above evaluation, it is concluded that the Licensee's proposed alternative examination provides an equivalent level of quality and safety for the DHR system.

Therefore, pursuant to 10 CFR 50.55a(a)(3)(1), it is recommended that the proposed alternative examination be authorized.

3.2.2.2 Reauest for Relief No. 9. Open Ended Class 2 Pioina Code Reouirement: Section XI, Table IWC-2500-1, Examination Category C-F-1, Item Numbe C5.11 requires both 100% volumetric .

and surface examinations cf Class 2 circumferential piping welds greater than 4 inch NPS and wall thickness greater than 0.375 inch as defined by IWC-2500-7, 24

Section XI, Paragraph IWC-1221(f) exempts from examination piping and other components of any size beyond the last shutoff valve in open ended portions of systems that do not contain water during normal plant operating conditions.

Licensee's Code Relief Reouest: Relief is requested to consider piping and components from the reactor building sump to Valves DH-V6A and DH-V6B as exempt even though the system contains water.

Licensee's Basis for Reouestina Relief: The Licensee contends that the piping and components upstream of Valves DH-V6A/D are either buried in concrete or encased in welded steel guards.

Plant operating conditions are such that a minimum level of water is maintained in the sump such that this portion of the system will contain some water. GPUN states that this portion of the system is tested each refueling outage under the containment leak rate testing program requirements in accordance with 10 CFR 50, Appendix J.

Licensee's ProDosed Alternative Examination: None. The Licensee proposes to treat components upstream of Valves DH V6A and DH V6B as exempt components.

Evaluation: The Class 2 piping and components between the reactor building sump and Valves DH-V6A/B are buried in concrete or encased in welded steel guards. IWC-1221(f) exempts piping and components beyond the last shutoff valve in open ended portions of systems if they do not contain water under normal plant operating conditions. The subject piping and components are open ended and only contain the water resulting from the static head pressure of the water level maintained in the sump.

Because it is open ended, it is not considered pressure retaining. This portion of the system contains some water but is inaccessible and, therefore, it is impractical to perform Code-required examinations without redesigning the system. The 25

l l

Appendix J Containment Leakrate Testing Program will verify the

-integrity of this portion of the system.

Conclusions:

Based on the above evaluation, the burden on the 1 Licensee if the Code requirement were imposed, and pursuant to 10 CFR 50.55a(g)(6)(i), it is recomended that relief be granted as requested.

8 3.2.3 Egepa (No relief requests) 3.2.4 Valves (No relief requests) 3.2.5 General 3.2.5.1 Reouest for Relief No. 13 (Part 2 of 3). Paraaraoh IWC-2430.

Additional Examinations

[.0de Reouirement: Section XI, Paragraph IWC-2430 states:

(a) Examinations that detect indications exceeding the allowable standards of IWC-3400 shall be extended to include an additional number of components (or areas) within the same examination category, approximately equal te the number of components (or areas) examined initially during the inspection.

(b) If the additional examinations detect further indications exceeding the allowable standards of IWC-3000, the remaining number of similar components (or areas) within'the same examination category shall be examit,sd to the extent specified in .

Table IWC-2500-1, except as modified by (c) and (d) below.

(c) Where the required piping examinations are limited to one loop or branch run of an essentially similar piping configuration, examinations that reveal indications exceeding the allowable standards of IWC-3000 shall require the additional examinations of (a) above and shall be extended to include  ;

26

examination of a second loop or branch run to the extent specified in Table IWC-2500 1.

(d) If the examination of the second loop or branch run reveals further indications exceeding the allowabia standards of IWC-3000, the remaining number of loops or branch runs that perform similar functions shall be examined, licensee's Code Relief Reouest: Relief is requested from using the 1986 Code Additional Examinations requirements.

Licensee's Basis for Reauestino Relief: The Licensee states that the requirements in the 1986 Code are very prescriptive and could actually not allow credit for examination of components that should be examined as part of an expanded sample. ASME has recognized that these situations can exist and is in the process ,

of changing the Code additional Lamination requirements. These changes allow selection of additional examinations based on >

similar materials and service, thus the root cause of the condition requiring additional examinations is considered when expanding examination samples.

Listnsee's Proposed Alternative Examination: The Licensee proposes to perform additional examinations based on proposed changes to Paragraph IWC-2430 of Section XI.

(a) Examinations performed in accordance with Table IWC 25001, except for examination category C-H, that reveal flaws or relevant conditions exceeding the acceptance standards of Table IWC-34101 shall be extended to include additional examinations at this outage. The additional examinations shall include an additional number of welds, areas, or parts included in the inspection item eoual in number to 20% of the welds, areas, or parts included in the inspection _ item that are scheduled to be performed during the interval. The additional examinations shall be selected from welds, areas, or parts of similar material and service. This additional selection may require the inclusion of 27

)

piping systems other than the one containing the flaws or relevant conditions.

(b) If the additional examinations required by (a) above reveal flaws or relevant conditions exceeding the acceptance standards of Table IWC-3410-1, the examinations shall be further extended to include additional examinations at this outage. These additional examinations shall include the remaining number of -

welds, areas, or parts included in the inspection item of similar material and service subject to the same type of flaws or relevant conditions.

(c) For the inspection period following the period in which the examinations of (a) or (b) above were completed, the examinatic.4s shall be performed as normally scheduled in accordance with IWC-2400.

Evaluation: The proposed alternative to Paragraph IWC-2430 appears in the 1991 Addenda. Although the proposed alternative may have technical merit, the NRC staff has not yet reviewed and endorsed this change. Approval of this request for relief before

~

NRC staff review would uidermine the review process.

Conclusions:

Based on the unapproved nature of the 1991 Addenda, it is recommended that this specific request for relief from the additional examination requirements of the ASME Code be denied.

3.3 Class 3 Comoonents 3.3.1 Eipjng (No relief requests) 3.3.2 Pumps (No relief requests) 3.3.3 Valves (No relief requests) i 28

3.3.4 General 3.3.4.1 Reauest for Relief No.13 (Part 3 of 3). Additional Examination Reauirements for Class 3 Components Code Reauirement: The 1986 Edition of Section XI has no provisions for additional examinations of Class 3 components.

1 Licensee's Code Relief Recueti: Relief is requested to use the proposed changes to Section XI, Additional Examinations requirement for Class 3 components.

Licensee's Basis for Reqqtitina Re_llif: The Licensee states that the requirements in the 1986 Code for Class 1 and Class 2 components are very prescriptive and could actually not allow credit for examination of components that should be examined as part of an expanded sample. ASME has recognized that these situations can exist and is in the process of changing the Code additional examination requirements. These changes allow selection of additional examinations based on similar materials and service, thus the root cause of the condition requiring additional examinations is considered when expanding examination samples. The proposed Code change includes Class 3 components.

Licensee's Proposed Alternative Examination: The Licensee proposes to perform additional examinations based on the proposed addition of Paragraph IWD 2430 to Section XI.

(a) Examinations performed in accordance with Table IWD 25001 (except for examination Item Numbers D1.10, 02.10, and D3.10) that reveal flaws or relevant conditions exceeding the acceptance standards of IWD 3000 shall be extended to include additional examinations at this outage. The additional examinations shall inclose an additional number of welds, areas, or parts included in the inspection item equal in number to 20% of the welds,

. areas, or parts included in the inspection item that are scheduled to be performed during the interval. The additional 29

examinations shall be selected from welds, areas, or parts of similar material and service. This additional selection may require the inclusion of piping systems other than the one containing the flaws or relevant conditions.

(b) If the additional examinations required by (a) above reveal flaws or relevant conditions exceeding the acceptance standards of IWD-3000, the examinations shall be further extended to include additional examinations at-this outage. The extent of the additional examinations shall be determined by the Owner based upon an engineering evaluation of the root cause of the flaws or relevant conditions. The Owner's corrective measures shall be documented per IWA 6000.

(c) For the inspection period following the period in which the examinations of (a) or (b) above were completed, the examinations shall be performed as normally scheduled in accordance with IWD-2400.

Evaluation: Additional examinations for Class 3 components are r.at covered by the 1986 Code. The Licensee's proposal appears in the 1989 Edition, 1991 Addenda. The inclusion of these requirements into the THI-l Program Plan provides a level of quality and safety above that provided by the Code ecord for toe second ISI interval.

Conclusi_ons: Performing Class 3 additional examinations to the 1901 Addenda's basic requirements, when no requirements exist in the 1986 Code, is an acceptable option for the Licensee. It is determined that because there is no existing 1986 Code requirement, relief is not required.

30 a

3.4 Pressure Tests 1

3.4.1 Class 1 System Pressure Tests 3.4.1.1 Reauest for Relief No.16.4. Examination Cateoory B P. Item B15.51. Hydrostatic Test of the Class 1 Reactor Coolant System (Auxiliary Pressurizer Sorav)  !

l Code Reauirement: Section XI, Table IWB-2500-1. Examination l

Category B P, item B15.51 requires that Class I systems be subjected to a system hydrostatic pressure test, per Paragraph IWB-5222, at or near the end of each inspection interval.

Licensee's Code Relief Reauest: Relief is requested from performing the Code required hydrostatic pressure test on approximately 4 feet of 1-1/2 inch diameter piping between Valves RC-V23 and RC-V4.

Licensee's Basis for Reouestina Relief: The Licensee states that there is no pressure tap and, therefore, no way to connect a hydrostatic test pump to this short section of pipe.

Licensee's Proposed Alternative Examination: None. A VT-2 examination will be performed during the system inservice leak test.

Evaluation: Valve RC-V4 is the Class 1/ Class 2 system boundary.

RC-V23 is a check valve that allows flow from the Class 2 Decay Heat Spray Line to the Pressurizer. The subject section of pipe cannot be pressurized from the Class I side because the check valve prevents flow. It is not desirable to pressurize from the Class 2 side because of the possibility of pressurized thermal shock on the Class I pressurizer system. The system design, therefore, makes the Code required hydrostatic test impractical tu perform. In order to perform the hydrostatic test in accordance with the requirements, the subject 4 foot line would require design modifications or installation of a pressure tap.

31

Conclusion:

It-is concluded that the Code-required hydrostatic I pressure test'of the subject section of Class 1 piping is ,

impractical to perform at THI-1. The increase in plant safety ,

would not compensate for the burden placed on the Licensee that l would result from imposition of the requirement. Therefore, pursuantto10CFR50.55a(g)(6)(1),itisrecommendedthatrelief be granted as requested. l l

3.4.2 Class 2 System Pressure Tests , 3.4.2.1 Reauest for Relief No. 16.1. Examination Cateaort C-H. Item .

C7.40. Hydrostatic Test of Class 2 Makeuo and purification System

• Pioina Code Reauirement: Section XI, Table IWC-2500-1, Examination Category C-H requires that the system hydrostatic test pressure be at least 1.10 times the system pressure for systems with  ;

design temperature of 200< or less, and at least 1.25 times the- ,

system pressure for systems with design temperature above 200*F.

Licensee's Code Relief Reauest: Rel'ief is requested from '(

performing the_ Code-required hydrostatic test of the makeup and purification piping-section from MU-V64C to MU-V193C. ,

Licensee's Basis for Reauestina' Relief: The Licensee states that between Valves MU-V64C and MU-Vl93C, thre is no vent or instrument tap that will allow connection of. the hydrostatic . test pump. Hydrostatic testing of this particular'section of piping,_

approximately 8 feet of?l-1/2 inch and 2 inch diameter pipe, is therefore not considered practical. This piping experiences nominaloperatingpressure(approximately3000psig)duringthe performance of the quarterly inservice test. . i licensee's Proposed Alternative Examination: None. The Licensee i

states that a' system leakage test with makeup and purification-Pump MU-PIC_ operating (approximately 3000 psig) has assured that ,

32

. ___ __.____.u_.. ._-_ _ , . . _ . _ . . _ . _ _ . . - . ,

the pressure retaining capability, structural integrity, and leaktigntness f this piping is maintained.

Evaluation: As shown in ISI Drawing No. ID ISI-FD Orl, there are three streams associated with this portion of the Makeup and Purification System. Systems "A" and "B" receive the Code-required examination because the piping between MU-Y64A/B and MU Vl93A/B contains a vent that allows connection of the hydrostatic test pump. System "C" does not have a vent or instrument tap. TheDsystemdesign,therefore,makestheCode-required hydrostatic test impractical to perform. In order to perform the hydrostatic test in accordance with the requirements, the subject line would require design modifications and installation of a connection for the hydrostatic test pump. The increase in plant safety would not compensate for the burden placed on the Licensee that would result from imposition of the requirement.

Because the "C" system sees the same operating and environmental conditions as the "A" and "B" systems, which receive the Code-required hydrostatic test, a reasonable assurance of the continued structural integrity is attained.

Code Case N-498, " Alternative Rules for 10-Year Hydrostatic Pressure Testing for Class 1 and 2 Systems," is included in.

Revision 9 of Regulatory Guide 1.147 (Reference 10). This Code Case essentially eliminates the hydrostatic test for Class 1 and 2 systems. As an alternative, it allows other system pressure tests to be performed provided the systems are held at nominal operating pressure (NOP) for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for insulated systems and 10 minutes for noninsulated systems prior to performing the VT-2 visual examination.. The system is held at NOP during the performance of the VT-2 visual examination.

Conclusioni: The concept of performing pressure tests at system operating pressure has been accepted by the NRC. The above evaluation describes the impracticality of performing the Code-33

--. = - -- .- - _ .

required examination. Therefore, pursuant to 10 CFR 50.55a(g)(6)(i), it is recomended that relief be granted as requested provided the system is pressurized tu nominal operating pressure for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for insulated systems and 10 minutes for noninsulated systems prior to the VT-2 visual examination. The system should be held at N0P during the performance of the VT-2 visual examination.

3.4.2.2 Reauest for Relief No,16.2. Examination Cateoory C-H. Item C7.40. Hydrostatic Test of Class 2 Core Floodino System Pioino Code Reauirement: Section XI, Table L 2500 1, Examination Category C-H requires that the system hydrostatic test pressure be at least 1.10 times the system pressure for systems with design temperature of 200*F or less, and at least 1.25 times the system pr. essure for systems with design temperature above 200*F.

Licensee's Code Relief Reauest: Relief is requested from performing the Code-required hydrostatic pressure test of the Class 2 piping between Valves CF-V1A/B and r* V4A/B.

Licensee's Basis for Reouestino Relief: The Licensee states that the THI-l FSAR Section 6.1.2.1.c and Plant Operating Procedure (0P) 1102-11. " Plant Cooldown" and OP 1102-1, " Plant Heatup to 525'F" require that CF-VIA/B be open when the RCS is >650 to 700 psig. There it no isolation valve between CF-VIA/B and CF-V4A/B.

In addition, the Licensee reports that THI-l Technical Specification Section 3.1.2.1 oses not allow pressurization of the RCS to nominal operating pressure (2155 psig) without heatup which requires CF-VIA/B open. The core flood tanks are designed for 700 psig and there is no way to achieve a hydrostatic test pressure of 3125 psig between CF-VIA/B and CF-V4A/B.

Licensee's Proposed Alternative Examination: None. A system inservice leak test and VT-2 visual examination from C~-VIA/B to CF-V4A/B will be performed in accordance with IWA-5211(a) at 34

600125 psig. This pressure range is the same as the range of core flood tank operating pressure permitted by Technical Specifi< *. ion Sect?on 3.3.1.2.a. This pressure range represents the maH t ; range of pressure to which this section of piping would be cxposed during operation. The Licensee contends that leakage fiam this portion of the system would be quickly evident by a drop in core flood tank pressure and level.

Eyaluation: CF-VIA/B are motor operated valves that must remain open when the RCS is >650 to 700 psig; CF-V4A/B are check valves that will not allow RCS pressure into the core flood system piping. The core flood tanks are designed for 700 psig and there is no way to achieve a hydrostatic test pressure of 3125 psig between the subject valves. The system design, therefore, makes the Code-required hydrostatic test impractical to perform. In order to perform the hydrostatic test in accordance with the requirements, the subject system would require design modifications. Imposition of this Code requirement on GPUN would result in a burden without a compensating increase in safety.

Code Case N 498, " Alternative Rules for 10 Year Hydrostatic Pressure Testing for Class 1 and 2 Systems," is included in Revision 9 of Regulatory Guide 1.147. This Code Case essentially eliminates the hydrostatic test for Class 1 and 2 systems. As an alternative, it allows other system pressure tests to be performed provided the systems a e held at nominal operating pressure (NOP) for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for insulated systems and 10 minutes for noninsulated systems prior to performing the VT-2 visual examination. The-system M held at N0P during the performance of the VT-2 visual exanination.

J

[gngiusions: The concept of performing pressure tests at system operating pressure has been accepted by he NRC. The above evaluation describes the impracticality of performing the Code-required examination. Therefore, pursua:t'to 10 CFR 50.55a(g)(6)(1), it is recommended that relief be granted as requested provided the cystem is pressurized to nominal 35

_ _ _ _ - _ _ _ _ _ - _ - - _ _ - - _ _ - _ _ _ - - - - - _ . ---_----_--A

i

. . . l operating pressure for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for insulated systems and 10 minutes for noninsulated systems prior to the VT-2 visual examination. The system should be held at NOP during the performance of the VT-2 visual examination.

3.4.2.3 Reauest for Relief No. 16.3. Examination Cateaory C-H. Item C7.40. Hydrostatic Test of Cless 2 DHR System Pioina Code Reauirement: Section XI, Table IWC-2500-1, Examination j Category C H requires that the system hydrostatic test pressure l ba at least 1.10 times the system pressure for #fstems with l design temperature of 200*F or less, and at least 1.25 times the  !

system pressure for systems with design temperature above 200'F.

Licensee's Code Relief Reauest: Relief is requested from

]

performing the Code-required hydrostatic pressure test at the j Class 2 test pressure of 3125 psig for the DHR piping between Valves DH-V22A/B and DH-V4A/B.  ;

l Licensee's Basis for Reauestina Relief: The Code required hydrostatic test pressure is 3125 psig. The Licensee reports that.there is no isolation valve downstream of DH-V22A/B in the reactor coolant system (RCS) and, therefore, no way to hydrostatically test the pipe between-DH-V22A/B and DH-V4A/B.

Licensee's ProDosed Alternative Examination: The Licensee proposes to perform a hydrostatic pressure test of the subject piping to the Class 1 test pressure of 1.02 times the nominal operating pressure of 2155 psig, or 2200 psig.

Evaluation: As shown in Drawings ID-ISI-FD-005 and ID-ISI-fD 019 for THI-1, the system design does not permit pressurizing the subject portions of Class 2 piping to the Code-required test-pressure withcut overpressurizing the adjacent Class 1 piping.

Therefore, the Code-required hydrostatic test pressure for the Class 2 portion is impractical to attain. The system would 36

_. _ _ _1 _ _

require design modifications in order to perform the Code-required hydrosthtic test. Imposition of the requirement on GPUN would cause a burden that would not be compensated significantly by an increase in safety above that provided by the proposed alternative. The Licensee's alternative test, performed at the Class 1 test pressurr., will provide reasonable assurance of the continued inservice structural integrity.

Conclusions:

It is concluded that the Code-required hydrostatic test of the subject portions of Class 2 piping in the DHR system is impractical to perform at THI-1, and that the public health and safety will not be endanaered by allowing the alternative test to be performed in lieu of the Code requirement. Therefore, pursuant to 10 CFR 50.55a(g)(6)(i), it is recommended that relief be granted as requested.

3.4.2.4 Reauest for Relief No.16.5. Examination Cateaory C-H. Item C7.40. Hydrostati tTest of Class 2 Hakeuo and Purification System Pinina Code Recuirement: Section XI, Table IWC-2500-1, Examination Category C-H requires that the system hydrostatic test pressure be at least 1.10 times the system pressure for systems with design temperature of 200*F or less, and at least 1.25 times the system pressure for systems with design temperature above 200*F.

Licensee's Code Relief Reauest: Relief is requested from performing the Code-required hydrostatic pressure test at the Class 2 test pressure of 3125 psig for piping between Valves MU-V16A/B/C/D and MU-V107A/B/C/D.

Licensee's Basis for Reauestina Reliff: The Code-required hydrostatic test pressure is 3125 psig. The Licensee raports-that there is no isolation valve downstream of MU-V107A/B/C/D and, therefore, no way to attain the hydrostatic test pressure of 37

3125 psig for the piping between Valves MU-V16A/B/C/D and MU-V107A/B/C/D.

Licensee's Proposed Alternative Examination: The Licensee proposes to perform a hydrostatic pressure test of the subject piping to the Class 1 test pressure of 1.02 times the nominal operating pressure of 2155 psig, or 2200 psig.

Evaluation: As shown in Drawing ID ISI-FD-017 for THI-1, the system design does not permit pressurizing the subject portions of Class 2 piping to the Code-required test pressure without overpressurizing the adjacent Class 1 RCS piping. Therefore, the Code-required hydrostatic test pressure for the Class 2 portion is impractical to attain. The system would require design modifications in order to perform the Code required hydrostatic test. Imposition of the requirement on CPUN would cause a burden that would not be compensated significantly by an increase in safety above that provided by the proposed alternative. The Licensee's alternative test, performed at the Class 1 test pressure, will provide reasonable assurance of the continued inservice structural integrity. .

C.oncl usioS * : It is concluded that the Code-required hydrostatic test of the subject portions of C1tss 2 piping in the makeup and purification system is impractical to perform at THI-1, and that the public health and safety will not be endangered by allowing the alternative test to be performed in lieu of the Code requirement. Therefore, pursuant to 10 CFR 50.55a(g)(6)(i), it is recommended that relief be granted as requested.

3.4.3 Class 3 System Pressure Tests (No relief requests) 38-

i i

3.4.4 Deneral 3.4.4.1 Reauest for Relief No. 15. Paraaraoh IWA-5242fal. VT-2 Examination of Bolted Connections Code Reauirement: Section XI, Paragraph IWA-5242(a) states that for systems borated for the purpose of controlling reactivity,  !

insulation shall be removed from the pressure retaining bolted  !

connections for VT-2 visual examination.

Table IWA-5210-1 requires a VT-2 visual examination at nominal operating pressure after opening and reclosing cf a component in a Class 1 system.

Licensee's Code Relief Reauest:- Relief is requested from performing the Code-required VT-2 visual examination at operating pressure with insulation removed from any closure where a personnel hazard exists due to elevated temperature.

Licensee's Basis for Reauestina Relief: The Licensee states that THI-1 Technical Specification 3.1.2.1 does not allow pressurization of the Reactor Coolant System (RCS) to nominal operating pressure without heatup. The Licensee believes that it is impractical to remove and re-install the insulation on components where a personnel hazard exists due to elevated temperature.

Licensee's Proposed Alternative Etamination: The Licensee proposes to perform a VT-2 visual examination, per IWA-5241(b),

"Noninsulated Components," during system pressure tests, with th.e insulation installed. IWA-5241(b) stater that for ccmponents whose external surfa es are inaccestit% for direct visual examination (VT-2), on?] the examination of the surrounding area (including floor areas or cmipment surfaces located underneath the components) for evidence of leakage shall be required. The affected closures will also be 4xamined for evidence of leakage 39

1 i

during cold shutdown for refueling when personnel safety is not a concern.,

Evaluation: Table IWB-2500-1 Examination Category B-P, requires i that a VT-2 visual examination be performed during the Class 1 system leakage test each refueling outage. Table IWC-2500-1, Examination Category C-H, requires a VT-2 during the Class 2 system pressure test each period. Each of these sys+.em pressure tests shall be conducted at the test conditions (pressure and temperature) designated for that system. For systems borated for the purpose of controlling reactivity, insulation shall be removed from pressure retaining bolted connections. The Licensee is attempting to blanket Class 1 and 2 bolted connections under one request for relief with justification only for Class 1.

THI-1 Technical Specification 3.1.2.1 does not allow pressurization of the RCC to nominal operating pressure without heatup. Consequently, a personnel hazard exists when insulation is removed for VT-2 examination. There is no temperature restriction identified for pressurization of the Class 2 systeais.

The Licensee prcposes to perform a VT-2 visual examination, per IWA-5241(b), "Noninsulated Components" of the bolted connections, with the insulation installed. Based on the safety implications of reinstalling the insulation at elevated temperature, the Code-required examination is impractical to perform for the Class 1 systems. The Licensee's proposed alternative does not allow sufficient time, prior to VT-2 visual examination, for saturation of the insulation at a leaking bolted connection and detection before reactor startup. For this reason, the system should be pressurized to NOP for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> before performing the VT-2 visual examination of tht: insulated components.

GPUN's proposed alternative implies that during each refueling outage the affected clor,ures will have the insulation removed for VT-2 visual examination at zero or static pressure. Becaum borated water leaves a crystalline residue, this proposed 40

alternative examination provides a reasonable assurance that any previous leakage at bolted connections would be detected.

Conclusions:

It is concluded that installing insulation at bolted connections during reactor operating conditions creates a personnel hazard due to the elevated temperature and pressure.

Therefore, pursuant to 10 CFR 50.55a(g)(6)(1), it is recommended that relief be granted from the VT-2 visual examination for the Class 1 bolted connections provided the Licensee's alternative examinations require: (1) a minimum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at nominal operating pressure before the VT-2 visual examination of the insulated connections, and (2) that the insulation be removed from the affected Class 1 bolted connections during each refueling outage for a VT-3 visual examination.

Because the justification for determining impracticality was insufficient for Class 2 components, it is recommended that relief from the code required VT-2 visual examinations for Class 2 components be denied.

3.4.4.2 Reauest for Relief No. 14. Paraaraoh IWA-5250(a)(2). Corrective Measures for Bolted Connections a

Code Recuirement: Section XI, Paragraph IWA-5250(a)(2) requires removal and VT-3 visual examination of all the bolting at bolted connections identified as leaking during a system pressure test.

Licensee's Code Relief Reouest: Relief is requested from performing the Code-required VT-3 visual examination of bolting at bolted connections identified as leaking based on an engineering evaluation.

Licensee's Basis for Raouestino Rel' 7: The Licensee states that removal of all bolting at a bolted connection is not always required to assure the connection has not degraded. Some connections are more difficult to seal after disassembly'when 41

l 1 . . e compared to retorquing an already assembled connection. In addition, complete disassembly of a connection '.n a radiation area would increase personnel exposure. GPUN notes that the ASME has recognized such situations and changed this requirement to allow removal and evaluation of one bolt in a bolted connection identified as leaking.

If leakage is minor and can be corrected b retorquing, removal of the bolt or stud for VT-3 is not appropriate for low pressure and low temperature systems (s200 psig and s200'F). The river water system and the closed cooling water systems are typical low temperature, low pressure systems.

The Licensee states that bolting removed from a connection is visually examined by maintenance or quality assurance personnel.

These personnel are not VT-3 certified but ara qualified to perform a first line visual examination due to their experience and training. This first line examination will determine that no corrosion or corrosion degradation exists after cleanup prior to reinstalling the connection. Should corrosion or corrosion degradation be identified, the bolting will be replaced or be VT-3 examined and evaluated in accordance with IWA-3100.

Licensee's Procosed Alternative Examination: If leakage is found in any Class 1, 2, and 3 system components and plant availability would be affected by removal of the bolting to perform a VT-3 examiration, as an alternative an evaluation of the failure potential for the bolting will be performed. GPUN believes that it ir ' practical to shutdown the plant if an evaluation shows:

1) ,ct the leakage is acceptable, 2) that the bolting and the closure will not fail before the next scheduled system outage, and 3) the leakage will not cause unacceptable degradation to adjacent components.

The Licensee's proposed alternative is to replace the requirements of the 1986 Edition with the requirements of the 1989 Edition, 1990 Addenda for Paragraph IWA-5250(a)(2). In 42 l

~'

addition, for low temperature and low pressure systems, no bolting will be removed for VT-3 examination if retorquing corrects the leakage. Also, in some cases a VT-3 visual examination will-not be performed until the next system outage if evaluation shows that the leakage is acceptable, the bolting and closure will not fail, and adjacent components will not be degraded. GPUN will only perform VT-3 examinations of bolting removed from connections found to be leaking during a system pressure test if degradation has been identified as discussed above and the bolting is to be reused.

Evaluation: There are four separate parts to be considered in this request for relief. (1) The use of 1989 Edition 1990 Addenda in lieu of the 1986 Edition for Paragraph IWA-5250(a)(2).

(2) The performar,ce of an evaluation of the failure potential fnr leaking bolted connections that could postpone the VT-3 examination until the next system outage. (3) Visual examinations c. e e d bolting performed by non-certified VT-3 personnel. (4) For low pressure and low temperature systems (5200 psig 5200*F), no bolting will be removed fe C3 examination if retorquing corrects the leakage.

(1) The 1986 Edition requires that all bolting be removed and VT-3 examined at all leaking bolted connections. The 1990 Addenda was revised to require removal of one of the bolts closest to the leakage. When the removed bolt has evidence of degradation, all remaining bolting in the connection shall be remnved, VT-3 examined, and evaluated. The 1990 Addenda relies on the training and skill of the certified VT-3 inspector to detect degradation. This concept provides an acceptable level of quality and safety in that the area exposed to leakage is evaluated.

l (2) GPUN is requesting to use an engineering evaluation of the failure potential for a leaking bolted connection as an alternative to a VT-3 visual examination. A VT-3 of the bolting is required if the VT-2 visual examiration reveals 43

leakage during a system pressure test, These system pressure tests are usually performed during the mode changes associated with reactor startup. Any leakage detected during thisSection XI VT-2 examination is subject to the Corrective Measures described in IWA-5250 . ' would not be considered prudent to contit.de in the sti t- ode with a leaking bolted cennection in a borated sys:p..

Detecting leakage during the routine surveillance activities that occur during normal plant operation is not considered inservice inspection. In this case, an engineering ,

evaluation is warranted if leakage is detected. The Licensae indicates that when plant availability would be affected by ran. oval of the bolting, the engineering evaluttion would be performed in lieu of the VT-3 examination at that time, During reactor startup, the plant should be considered unavailable until the requirements of' ASME Section XI are completed.

Request for Relief No.15, evaluated previously in this Technical Evaluation Report, describes the impracticality of

~

removing the insulation from bolted connections for VT-2 visual examination. This could mean that the proposed engineering evaluation would be conducted without removing the insulation.

(3) As required by IWA-5250(a)(2) of 1990 Addenda, the Licensee proposes to have the single bolt removed from a leaking bolted connection, examined by non-certified VT-3 personnel to ' determine if a VT-3 examination is necessary on the l remainder of the bolted conner. tion. Certified VT-3 visual l examinations will be performed as the result of a system pressure test if degradation has been identified by non-certified personnel. As stated earlier, the concept of removing and examining only one bolt in a leaking connection

! is considered acceptable when relying on the training and skill of the VT-3 inspector. Since THI-I is proposint to do l

44 l

l

the examination without certified VT-3 personnel, the proposal is unacceptable.

(4) The Licensee proposes attempting to correct minor leakage in low temperature and low pressure systems (5200 psig and

$200*F) by retorquing. The Code does not differentiate between system pressures and temperatures for appropriate corrective measures. The Licensee has not demonstrated the imprar+icality of this requirement.

Conclusions:

Based upon the above evaluation, it is concluded that adequate technical justification has not been presented to demonstrate the-impracticality of complying with the Code requirement for items (2), (3), and (4) above. Therefore, it is recommended that relief for these items be denied. However, pursuant to 10 CFR 50.55a(a)(3)(i), it is recommendeo chat the proposed alternative described in (1) above be authorized.

3.5 General 3.5.1 Ultrasonic Examination Technioues (No relief requests) 3.5.2 Exemoted Components (No relief requests) 3.5.3 Other 3.5.3.1 Reouest for Relief No. 10. Use of Form NIS-2 Code Reouirenent: Section XI, IWA-4000, IWA-5000, IWA-6000, and IWA-7000 require specific examinations, notifications, and records for repairs and replacements to Class 1, 2, and 3 components and their supports. The Code requires the use of Form NIS-2 for each, licensee's Code Relief Reouest: Relief is requested from completing the Code-requir3d Form NIS-2 for: testing, repairs, 45

I l

and preventive maintenance work on snubbers; " Replacement in Kind" activities; and repairs of components 1 inch NPS or less.

GPUN is also requesting relief from the code-required notification of'the Authorized Inspection Agency prior to the above mentioned NIS-2 actions, Licensee's Basis for Reouestino Relief: The Licensee contends that snubber testing, repairs, and preventive maintenance work is performed under a work authorizing document. This document identifies, in detail, hat work was done and why it was done.

Should snubber failures be identified, THI-l Technical Specification Section 4.17 requires an engineering evaluation of the condition that has been identified. The Licensee states that the documentation resulting from the Snubber Test Program can be reviewed for determination of snubber replacements and history and that the Authorized Nuclear Inservice Inspector (ANII) v"1 be notified of the snubber test schedule and can verify replacements as deemed necessary.

The Licensee states that their " Replacements in Kind" activities should not be included in Section XI requirements. GPUN has established a practice that allows the maintenance department to replace components with similar components. Normal examples of such activities would be 1) replacement of valve disks or bonnets with spares, 2) replacement of bolting with new parts or spares that have been cleaned and examined, and 3) replacement of relief valves with spares that have been tested and refurbished.

" Replacements in Kind" activities are considered routine maintenance similar to disassembly of components for adjustment or cleaning. The Licensee's position is that the ANII may verify replacement activities by reviewing the VT-2 visual examination documentation if desired.

The Lk e see states that completion of the Form NIS-2 for routine mainter 2nce practices would result in a paperwork system that would wt enhance their ability to trend replacement activities.

The Licensee's opinion is that it may, in fact, detract from 46

their ability to identify any significant replacement activities by producing a redundant paperwork system.

Section XI, IWA-4700 requires completion of Form NIS-2 for_all repairs regardless of component nominal pipe size. The Licensee states that the Code has exempted components 1 inch NPS or less from hydrostatic tests followin3 repairs, from examinations of Class 1, 2, or 3 components, and from the replacement requirements of IWA-7000. Completion of Form l115-2 for these components is not in line with the importance the Code has_placed on these components for other requirements.

Licensee's Proposed Alternative Examinat<pm GPUN will notify the ANII prior to each refueling outage that snubber testing will be performed during the refueling outage. All records required by the surveillance procedure will be completed and available for review as an alternative to inclusion on Form NIS-2.

GPUN will notify the NRC of non-exempt " Replacements in Kind," of components not exempt from Section XI, via the Form NIS-1 submittal following each refueling outage.

GPUN will not complete a Form NIS-2 Owners Report for Repairs or Replacements for repairs on components 1 inch NPS or less.

Evaluation: Uniformity and consistency are paramount for the effective regulation of the nuclear industry. The Code-required Form NIS-2 is a mechanism for maintaining and evaluating the extent of repairs and replacements industry wide. Eliminating the Form NIS-2 essentially eliminates the ANII. Elimination of the ANII from certain repair or replacement activities effectively isolates the utility from a third party review by the Authorized Inspection Agency.

Conclusions:

Disagreement with a Code requirement is not a basis for granting relief. Imposition of this Code requirement on GPUN 47

does not result in unu:ual hardship, burden or impracticality.

Therefor 9, it is recommended that relief be denied.

3.5.3.2 Reauest for Relief No.12. Cateaorization and Selection of Component Sucoorts Code Reauirement: Section XI, Paragraph IWF-2510(a) requires that the component supports selected for examination be the supports of those components that are required to be examined under IWB, IWC, and IWD during the first inspection interval.

Licensee's Code Relief Reauest: Relief is reques6ed from using the categorization and selection requirements for component supports in the 1986 Edition of the ASME Code.

Licensee's Basis for Reauestina Relief: The Licensee reports that Table IWF-2500-1 of Section XI requires examination of 100%

of the non-exempt piping supports and at least one component support, where similar components exist, during each inservice inspection interval. In the 1990 Addenda to Section XI, the ASME recognized that component support integrity could be verified adequately by examining a percentage of the support population.

Significant manpower would be necessary to support scaffold erection, insulation removal, and the actual performance of the examinations. The sampling of supports in accordance with the 1990 Addenda would greatly reduce the manpower required to perform these examinations.

Licensee's Prooosed Alternative Examination: GPUN proposes to use IWF-2000 of the 1990 Addenda in lieu of IWF-2000 of the 1986 Edition of Section XI.

Evaluation: Portions of Subsection IWF in ASME Code Editions prior to the 1990 Addenda are unclear with regard to rules for the inservice inspection of component supports; the 1990 Addenda clarified these rules and added others. The Licensee's proposal 48

. , ,a is to use Subsection IWF-2000 of the 1989 Edition,.1990 Addenda of Section XI. We have reviewed the Licensee's proposed.

alternative and found it to be an acceptable approach for selection and categorization of component supports for examination.

Conclusions:

It is concluded that the Licensee's proposed alternative provides an equivalent level of quality and safety with adequate assurance of the continued inservice structural integrity of the component supports. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), it is recommended that the proposed alternative examination be authorized, i-l l

49 i

a

4. CONCLUSION Pursuant to 10 CFR 50.55a(g)(6) or, alternatively,10 CFR 50.55a(a)(3), it has been determined that certain inservice examinations cannot be performed to the-extent required by Section XI of the ASME Code. In these cases for which relief is requested, ext.ept Requests for Relief Nos. 10, 13, 14 (in part), and 15 (in part), the Licensee has demonstrated that specific Section XI requirements are impractical or that alternative examinations should be authorized. For the exceptions cited above, it is concluded that the Licensee has not provided technical justification to support the determination that the Code requirement is impractical and that requiring the Licensee to comply with the Code requirement would result in hardship.

This technical evaluation has not identified any practical method by which the Licensee can meet all the specific inservice inspection requirements of Section XI of the ASME Code for the existing Three Mile Island Nuclear Station, Unit 1, facility. Compliance with all the exact Section'XI required inspections would necessitate redesign of a significant number of plant systems, sufficient replacement components to be obtained, installation of the new components, and a baseline examination of 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 certain provisions of Section XI of the ASME Code that have been determined to be impractical. Pursuant to 10 CFR 50.55a(g)(6), relief is granted from the requirements which are impractical to implement, or alternatively, pursuant to 10 CFR 50.55a(a)(3),

alternatives to the Code-required examinations are authorized provided that either (i) the proposed alternatives provide an acceptable leval of quality and safety or (ii)- Code compliance would result in hardship or unusual difficulty without a compensating increase in safety. Relief may be granted only if granting the 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.

The Licensee should continue to monitor the development of new or imptoved examination techniques. As improvements in these areas are achieved, the 50

. . ~ -_ .,

I Licensee should incorporate these techniques in the ISI program plan examination requirements.

Based on the review of the Three Mile Island Nuclear Station, Unit 1, Second

.30-Year Interval Inservice Inspection Program Plan, Revision 0, the Licensee's Response to the NRC's Request for Additional Information, and the recommendations for granting relief from the ISI examination requirements that have been determined to be impractical, it is concluded that the Three #ile Island Nuclear Station, Unit 1, Second 10-Year Interval inservice inspection Program Plan, Revision 0, with the exception of Requests for Relief Nos. 10, 13,14 (in part), and 15 (in part), is acceptable and in compliance with 10 CFR 50.55a(g)(4).

51

5. REFERENCES

1. Code of Federal Regulations, Title 10, Part 50.-

2. American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Division.1:

1986 Edition 1989 Edition through 1991 Addenda

3. Three Mile Island Nuclear Station, Unit 1, Second 10-Year Interval Inservice Inspection Program Plan, Revision 0, dated April 17, 1991.

4. NUREG-0800, Standard Review Plan, 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 September 19, 1991, R. W. Hernon (NRC) to T. G. Broughton (GPU Nuclear), request for additional information on the Second 10-Year Interval ISI Program Plan.

6. Letter, dated December 12, 1991, T. G. Broughton (GPU Nuclear) to NRC, response to the NRC request for additional information.

7. Regulatory Guide 1.150, Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinationc, Revision 1, dated February 1983.

8. NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, July 1980.

9. NRC Mechanical Engineering Branch Technical MEB 3-1, Postulated Rupture Locations in Fluid System Piping Inside and Outside Containment, Revision 2, dated June 1987. .

10. Regulatory Guide 1.147, Revision 9, Inservice Inspection Code Case Acceptability, ASME Section XI Division 1, April 1992.

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<s . .,,, .,.e., e L TITLE ANo SveTiTLE EGG-MS-10169 Technical = Evaluation Report on the Second 10-Year '

Interval Inservice Inspection Program Plan: . I cATE asecaT.veus to

• o~' naa GPU Nuclear Corporation,.

l

Three Nile--Island Nuclear ~ Station,- Unit 1 May -1992

4. ** oa Ga Aur Nuveia Docket-Number 50-289 FTN nan?? (pent-sn

6. AufMOmt$1 6. TYPE OF REPORT Technical 1

7. rf aico cov aa eo u o..e.,

B.W. Brown, S.G. Galbraith, A.M. Porter ca ., - u ~~ - c - -a -,.,  :

.. .En. gay gnizArios - ~ AuE A~o accatss u, = c- o, EG&G Idaho, Inc.

P.O. Box 1625

-Idaho: Falls, ID. 82415-2209-

9. $PCN$ORWG oRGANIZATlON - NAME ANo AooRtss ne mac, eyee snae . s remeras,. sa se =ac o= on= e, a.,ma, u.A as.,ss or m.in aseer, can med massmut sedDWELJ Materials and Chemical Engineering Branch Office of Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555 10.$UPPMM&NTARY NQTES I1. A&ST R ACT (200 ,wi e, mers This-report presents the results of the evaluation of-the Three Mile Island Nuclear Station, Unit 1 (TMI-1), Second 10-Year Interval -Inservice Inspection (ISI) Program-Plan, Revision 0, submitted April 19, 1991,- including- the reqwsts' 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 Inservice InspectioniProgram Plan is evaluated in Section 2 of this report for=

(a) compliance with the appropriate edition / addenda of Section XI, (b)- acceptability.

-of examination sample, (c) correctness of the application of system or component examination exclusion criteria, and (d) compliance with~ISI-related-commitments identified during the Nuclear Regulatory Commission's (NRC) previous reviews. -The requests for relief are evaluated in Section 3 of this report.

12. K EY WOR os/DESCR:PTQR $ (Las .em, e, sa, e wwe ss a.s,se seuse,per,,e .usees,., rne eye.,,.s it avavoiuTy $T4T&ptNT Unlimited

... ..wa. r . cu u., o r.a~

s Fa.e # sped Unclassified a r . ,, . .

Unclassified l$. NUMSEk C, PAGES 16,

• RICE .

NIC ,Cas 13s $34g5

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