ML20138F066

TER on Third 10-Yr Interval ISI Program Plan:Boston Edison Co,Pilgrim Nuclear Power Station

ML20138F066
Person / Time
Site:	Pilgrim
Issue date:	06/30/1996
From:	Mary Anderson, Feige E, Hall K IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:	NRC
Shared Package
ML20137D047	List: L-96-017, TER on Third 10-Yr Interval ISI Program Plan:Boston Edison Co,Pilgrim Nuclear Power Station ML20137D042 ML20137D051
References
CON-FIN-J-2229 INEL-96-0177, INEL-96-177, NUDOCS 9607090104
Download: ML20138F066 (43)

Text

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i INEL 96/0177

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I Technical Evaluation Report on the

. Third 10-year interval inservice inspection Program Plan:

Boston Edison Company,

Pilgrim Nuclear Power Station, e

Docket Number 50-293 M. T. Anderson E. J. Folge K. W. Hall A. M. Porter I

Published June 1996 Idaho National Engineering Laboratory Lockheed Idaho Technologies Company Idaho Falls, Idaho 83415 I

Prepared for the Division of Engineering Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Under DOE Idaho Operations Office Contract DE-AC07 941D13223

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J2229 (Task Order TWA-A08) ggran

l' l ABSTRACT 1

[ This report presents the results of the evaluation of the-Pilgrim Nuclear Power Station, Third 10-Year Interval Inservice Inspection Progran Plan, ,

Revision 0, submitted September 1,1995, including the requests-for relief l from'the American Society of Mechanical Engineers (ASME) Boiler and Pressure

] Vessel Code,Section XI, requirements that the licensee has determined to be i impractical. The Pfigria Nuclear Power Station, Third 10-Year Interval i

j Intervice Inspection Program Plan, Revision 0, is evaluated in Section 2 of

! this report. The inservice inspection (ISI) program plan is evaluated for (a) compliance with the appropriate edition / addenda of Section XI, (b) acceptability of the examination sample, (c) correctness of the application of system or component examination exclusion criteria, and

(d) compliance with ISI-related commitments identified during previous Nuclear Regulatory Commission (NRC) reviews. The requests for relief are evaluated in l Section 3 of this report.

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This work was fundLd under:

U.S. Nuclear Regulatory Commission JCN No. J2229 Task Order TWA-A08 Technical Assistance in Support of the NRC Inservice Inspection Program ii

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SUMMARY

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The licensee, Boston Edison Company (BECo), has prepared the Pf1 grim Nuclear l Power Station, Third 10-Year Interval Inservice Inspection Program Plan, l 1 Revision 0, to meet the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, 1989 Edition.

The third 10-year interval began July 1, 1995.

k l The information in the Pilgrin Nuclear Power Station, Third 10-Year Interval Inservice Inspection Program Plan, Revision 0, submitted September 1,1995, {

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. l l l l In the review of the Pilgria Nuclear Power Station, Third 10-Year Interval Inservice Inspection Program Plan, Revision 0, the licensee's responses to the l Nuclear Regulatory Commission's request for additional information (RAI), and y the recommendations for granting relief from the inservice examinations that 3

4 cannot be performed to the extent required by Section XI of the ASME Code, no j deviations from regulatory requirements or commitments were identified except l

[ as noted in the evaluation of Request for Relief PRR-24.

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CONTENTS ABSTRACT ................................ 11

SUMMARY

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

2. EVALUATION OF INSERVICE INSPECTION PROGRAM PLAN . . . . . . . . . . . 4 2.1 Documents Eval uated . . . . . . . . . . . . . . . . . . . . . . . . 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 ........... 5 2.2.3 Exemption Criteria . . . . . . . . . . . . . . . . . . . . . . 5 2.2.4 Augmented Examination Commitments . . . . . . . . . . . . . . . 6 2.3 Conclusions ........................... 7

3. EVALUATION OF RELIEF REQUESTS . . . . . . . . . . . . . . . . . . . . 8 3.1 Class 1 Components . . . . . . . . . . . . . . . . . . . . . . . . 8 3.1.1 Reactor Pressure Vessel ................... 8 3.1.1.1 Request for Relief PRR-9, Examination Category B-D, Item B3.90, Reactor Pressure Vessel Nozzle-to-Shell Weld Examinations .................... 8 3.1.2 Pressurizer (Not applicable) 3.1.3 Heat Exchangers and Steam Generators (No relief requests) 3.1.4 Piping Pressure Boundary . . . . . . . . . . . . . . . . . . . 10 3.1.4.1 Request for Relief PRR-1, Examination Category B-J, Item B9.ll, B9.12 and B9.21, Examination of Class 1 Pipe Welds . . . . . . . . . . . . . . . . . . . . . . . . 10 3.1.4.2 Request for Relief PRR-23, Examination Category B-J, Item B9.12, Examination of Class 1 Longitudinal Piping Welds .......................... 13 3.1.5 Pump Pressure Boundary (No relief requests) 3.1.6 Valve Pressure Boundary (No relief requests) 3.1.7 General (No relief requests) iv i

3.2 Class 2 Components ........................ 15 3.2.1 Pressure Vessels (No relief requests) 3.2.2 Piping . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 3.2:?.1 Request for Relief PRR-7, Revision 2, Examination Category C-F-2, Item C5.51, Class 2 Containment Ai.mospheric Control System Pipe Welds .......... 16 3.2.2.2 Request for Relief PRR-17, Examination Category C-F-2, Item C5.51, Examination of Class 2 Pipe Welds . . . . . . 18 3.2.3 Pumps (No relief requests) 3.2.4 Valves (No relief requests) 3.2.5 General (No relief requests) 3.3 Class 3 Components (No relief requests) 3.4 Pressure Tests ......................... 20 3.4.1 Class 1 System Pressure Tests ................ 20 3.4.1.1 Request for Relief PRR-21, Revision 1, Examination Category B-P, Items B15.50, B15.60, and B15.70,

. Pressure Tests of Class 1 Piping, Pumps and Valves . . . . 20 3.4.2 Class 2 System Pressure Tests ................. 23 3.4.2.1 Request for Relief PRR-13, Revision 1, Examination Category C-H, Item C7.30, C7.40, C7.70,-and C7.80, Pressure Tests of Class 2 Piping and Valves at Containment Penetrations Where the Balance of System is Nonclass ........................ 23 3.4.3 Class-3 System Pressure Tests ................. 27 3.4.3.1 Request for Relief PRR-11 Revision 1, Examination Category D-B, Item D2.10, System Functional and Hydrostatic Tests of Salt Service Water Pumps ...... 27 3.4.4 General (No relief requests) 3.5 General ............................. 29 3.5.1- Ultrasonic Examination Techniques (No relief requests) 3.5.2 Exempted. components (No relief requests) v ,

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s 3.5.3 Other ............................ 29 1

3.5.3.1 Request for Relief PRR-24, Paragraph IWA-2311(b),

Appendix VII, Ultrasonic Examination Personnel i Qualification Requirements . . . . . . . . . . . . . . . . 29

, 3.5.3.2 Request for Relief PRR-18, Code Case N-491, Table IWF-

2500-1, Category F-A, Items F1.10 and F1.20, Examination of Component Supports . . . . . . . . . . . . 31 3.5.3.3 Request for Relief PRR-22, Request for Relief From l

. Regulatory Requirement To Obtain Relief From Code  !

Requirements . . . . . . . . . . . . . . . . . . . . . . . 32

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4. CONCLUSION ............................. 33 i

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5. REFERENCES ............................. 35 l i

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l-TECHNICAL EVALUATION REPORT ON THE THIRD 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAN PLAN l BOSTON EDIS0N COMPANY l l PILGRIN NUCLEAR POWER STATION

DOCKET NUNBER 50-293 s

1. INTRODUCTION 1

Throughout the service life of a water-cooled nuclear power facility, 10 CFR 50.55a(g)(4) (Reference 1) requires that components (including

supports) that are classified as American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Class 1, Class 2, and Class 3 meet the 1 requirements, except the design and access provisions and the preservice l examination requirements, set forth in the ASME Code Section XI, Rules for I Inservice Inspection of Nuclear Power Plant Components (Reference 2), to tt e 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 1 in subsequent editions and addenda of this Code that are incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein and subject to Nuclear Regulatory Commission (NRC) approval.

The licensee, Boston Edison Company (BEco), has prepared the Pfigrim Nuclear Power Station, Third 10-Year Interval Inservice Inspection Program Plan, Revision 0 (Reference 3), to meet the requirements of the 1989 Edition of the Code. The third 10-year interval began July 1,1995.

As required by 10 CFR 50.55a(g)(5), if the licensee determines that certain Code examination requirements are impractical and requests relief from them, the licensee shall submit information and justification to the NRC to support that determination.

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Pursuant to 10 CFR 50.55a(g)(6), the NRC willl evaluate the licensee's

! determination that Code requirements are impractical to implement. The NRC 4 i

may grant relief and may impose al'ternative 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.

Alternatively, pursuant to 10 CFR 50.55a(a)(3), the NRC will evaluate the licensee's determination that either (i) the proposed alternatives provide an

acceptable level of quality and safety, or (ii) Code compliance would result in hardship or unusual difficulty without a compensating increase in safety. <

i Proposed alternatives may be used when authorized by the NRC.

The information in the Pilgrim Nuclear Power Station, Third 10-Year Interval

! Inservice Inspection Program Plan, Revision 0, submitted September 1, 1995,

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 the Standard Review Plans (SRP) of NUREG-0800 (Reference 4), Section 5.2.4, " Reactor Coolant Boundary Inservice

Inspections and Testing," and Section 6.6, " Inservice Inspection of Class 2 l and 3 Components."

In a letter dated December 6,1995 (Reference 5), the NRC requested additional

, information that was required to complete the review of the ISI Program Plan.

Boston Edison Company provided the requested information in a letter dated February 15, 1996 (Reference 6). In this response, the licensee withdrew

Request for Relief PRR-22.

a The Pilgrim Nuclear Power Station, Third 10-Year Interval Inservice Inspection Program Plan, Revision 0, is evaluated.in Section 2 of this report. The ISI Program Plan is evaluated for (a) compliance with the appropriate edition / addenda of Section XI, (b) acceptability of examination sample, (c) correctness of the application of system or component examination exclusion criteria, and (d) compliance with ISI-related commitments identified during the NRC's previous reviews.

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i 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, l 1989 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 PROGRAN PLAN This evaluation consists of a review of the applicable program documents to determine whether or not they are in compliance with the Code requirements and any previous license conditions pertinent to ISI activities. This section describes the submittals reviewed and the results of the review.

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

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(a) Pilgrim Nuclear Power Station, Third 10-Year Interval Inservice Inspection Program Plan, Revision 0, submitted September 1, 1995

(Reference 3). j

(b) Response to Request for Additional Information, Third 10-Year l 1

Interval Inservice Inspe: tion Program, submitted February 15, 1996  !

j (Reference 6).

2.2 Compliance with Code Reauirements

2.2.1 Como11ance with Apolicable Code Editions 4

, The ISI Program shall be based on the Code editions defined in 10 CFR 50.55a(g)(4) and 10 CFR 50.55a(b). Based on the starting date of 1

July 1,1995, the Code applicable to the third interval ISI program is the 1989 Edition. As stated in Section 1 of this report, the licensee l has prepared the Pilgrim Nuclear Power Station, Third 10-Year Interval l Inservice Inspection Program Plan, Revision 0, to meet the requirements of the 1989 Edition.

1 In accordance with 10 CFR 50.55a(c)(3), 10 CFR 50.55a(d)(2), and 10 CFR 50.55a(e)(2), ASME Code Cases may be applied to systems and components as alternatives to Code requirements. ASME Code Cases that have been found suitable for general use by the NRC are listed in the latest edition of Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability, with conditions for use, as applicable. These Code cases must be implemented in their entirety and the licensee may adopt them by providing written notification to the NRC. Alternatively, published 4

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. Code cases pending approval and subsequent listing in Regulatory Guide F

1.147'may be adopted only if the licensee requests, and the NRC authorizes, their use on a case-by-case basis.

l The licensee has adopted three Code Cases for use in the third interval inservice inspection program. Code Case N-491, Alternative Rules for l Examination of Class 1, 2, 3 and MC Component Supports of Light Water Cooled Power Plants, is approved for use in Regulatory Guide 1.147 and is acceptable for use as written. The licensee noted that two additional Cooe cases, not yet approved for generic use in Regulatory Guide 1.147 will be implemented during their third ten year interval. l These are Code Cases N-416-1, Alternative Pressure Test Requirement for Welded Repairs or Installation of Replacement Items by Welding, and Code Case N-498-1, Alternative Rules for 10-Year Systen Hydrostatic Testing for Class 1, 2 and 3 Systems. These Code Cases were approved for use by

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the NRC in a Safety Evaluation Report dated March 10, 1995. Based on this approval, it is acceptable for Boston Edison Company to implement the subject Code cases for the third interval. l 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). The sample size and weld selection have been implemented in accordance with the Code and 10 CFR 50.55a(b) and appear to be correct.

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

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2.2.4 Auamented Examination Commitments In addition to the requirements specified in Section XI of the ASME Code, a partial list of augmented nondestructive examinations that will 1 i

be performed during the interval is provided below. Augmented  !

examinations are being performed in response to NRC notices, General Electric Service Information Letters, and other licensee commitments.

(a) Ultrasonic examination of stainless steel pipe welds in accordance with Generic Letter 88-01, NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping (Reference 7).

I (b) Visual or liquid penetrant examination, as applicable, of the Reactor Recirculation (RR) pumps' shafts and hydrostatic bearings.

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Reference:

General Electric SIL No. 459 and RICSIL No. 038). .

l (c) Ultrasonic examination of all remaining old design, creviced, Inconel 600 Shroud Head Bolts (HSHBs) each refueling outage.

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Reference:

General Electric SIL No. 433 SI).

(d) Ultrasonic and/or visual examination of the Core Shroud (Reference General Electric SIL No. 572 Rev.1).

(e) Visual and surface examinations of the feedwater nozzles in accordance with NUREG-0619 (Reference 8).

(f) Visual examination of the Core Spray spargers and the Core Spray piping inside the RPV each refueling outage. (Reference IE Bulletin No. 80-13).

1 (g) Ultrasonic examination of components subject to thermal stresses in accordance with NRC Bulletin 88-08, 7herma? Stresses in Piping Connected to Reactor Coo 7 ant Systems (Reference 9).

(h) Visual examination of 50% of the Jet Pump Riser Braces each

"- refueling outage. (

Reference:

General Electric SIL No. 551).

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9 (1) Visual examination of Jet Pump Sensing Lines (Reference General

, Electric.SIL No. 420).

(j) Visual Examination of the Top Guide (

Reference:

General Electric SIL No. 554).

i 2.3 Conclusions s

Based on the review of the documents listed above, no deviations fr.om regulatory requirements or commitments have been identified for the A Pilgria Nuclear Power Station, Third 10-Year Interval Inservice Inspection Program Plan, Revision 0.

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! 3. EVALUATION OF RELIEF REQUESTS l l

, The requests for relief from the Code requirements that the licensee has determined to be impractical for the third 10-year inspection interval are

[ evaluated in the following sections.

3.1 Class 1 Comoonents 3.1.1 Reactor Pressure Vessel i

f 3.1.1.1 Reauest for Relief PRR-9. Examination Cateaory B-D. Item B3.90.

Reactor Pressure Vessel Nozzle-to-Shell Weld Examinations i

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Code Reauirement

Section XI, Table IWB-2500-1, Examination

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! Category B-D, Item B3.90 requires a 100% volumetric examination i of all reactor vessel nozzle-to-shell welds as defined by l Figure IWB-2500-7.

4 l Licensee's Code Relief Reauest: The licensee requested relief l from performing the Code-required volumetric examinations of

the reactor vessel nozzle-to-shell welds listed below. l l

! SYSTEM N0ZZLE % Ct~erage N0ZZLE-TO-  !

j VESSEL .i WELD  !

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i Recirculation Outlet NIA 81.39% RPV-NIA-NV i

NIB 71.39 RPV-NIB-NV i i Recirculation Inlet N2A-K 64.24% RPV-N2A-A i Through -K l Feedwater N4A 98.53% RPV-N4A-NV N4B RPV-N4B-NV

N4C RPV-N40-NV i i N4D RPV-N4D-NV l 4

Core Spray N6A 97.91% RPV-N6A-NV 4 N6B RPV-N68-NV-i

. Jet Pump N9A 63.52% RPV-N9A-NV j Instrumentation N98 RPV-N98-NV i CRD Hydraulic N10 82.79% RPV-N10-NV i

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Licensee's Basis for Reauestina Relief (as stated):  :

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! "As stated in 10CFR50.55a(g)(1) and (g)(4), for plants whose  !

i construction permits were issued prior to January 1,1971, -l components shall meet the requirements set forth in~ASME 1 Section XI to the extent practical within the limitations of u design, geometry and materials of construction of the  !

components. Since ASME Section XI examination requirements did not exist at the time. the Pilgrim Nuclear Power Station was dasigned, examination accessibility was not a primary consideration. This is evident from the fact that. 21 nozzle-to-vessel welds are not sufficiently accessible due to interference with the biological shield wall and vessel insulation to allow 100% examination coverage.

" Relief is requested for the 21 Code Item B3.90 nozzle-to-vessel welds referenced above on the basis that full access is not achievable due to the reactor vessel / biological shield configuration and associated non-removable vessel insulation.

Licensee's Pronosed Alternative Examination (as stated):

"The reactor nozzle-to-vessel welds will be volumetrically examined to the maximum extent possible."

Evaluation: The licensee requested relief from the Code-required 100% volumetric examination of the reactor pressure vessel nozzle-to-shell welds; the licensee can obtain greater than 63% volumetric coverage on the subject welds. Based on a review of the information provided, it has been determined that scanning limitations associated with permanent insulation, biological wall to vessel clearance, and the nozzle geometry  ;

preclude complete volumetric coverage. As a result, obtaining j the Code-required coverage is impractical. To perform the complete volumetric examination, design modifications would be necessary to eliminate the scanning limitations, causing a considerable burden on the licensee.

The licensee proposes to perform the volumetric examinations to j the extent practical. Based on the high percent of coverage that can be obtained, it can be concluded that significant degradation, .if present, will be detected. As a result,  !

reasonable assurance of structural integrity will be provided.

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i conclusions: Based on the above evaluation, it is concluded that obtaining complete Code-required volumetric coverage for

' the subject nozzle-to-shell welds is impractical for Pilgrim j Nuclear Power Station. Based on the percent of coverage that can be obtained on the subject examination areas (greater than

) 63%), it can'be concluded that significant degradation, if present, will be detected. Therefore, it is recommended that 1-relief be granted pursuant to 10 CFR 50.55a(g)(6)(1).

l 3.1.2 Pressurizer (Not applicable) j 3.1.3 Heat Exchanaers and Steam Generators (No relief requests) i i

3.1.4 Pinina Pressure Boundarv i

i 3.1.4.1 Reauest for Relief PRR-1. Examination Cateaory B-J.

I i Items 89.11. 89.12. and B9.21. Examination of Class 1 Pine l

} Welds l

l i Code Reauirement: Table IWB-2500-1, Examination Category B-J,

! Items B9.11, B9.12, and 'B9.21 require 100% volumetric and/or  ;

l surface examination of circumferential and longitudinal pipe

! welds as defined in Figure IWB-2500-8.

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Licensee's Code Relief Reauest

The licensee requested relief j j from examination of the Class I pipe welds that are within the 1 following containment penetration assemblies. l l

SYSTEM PENETRATION LINE NO. LINE ALTERNATE SIZE' WELD i- RHR (Shutdown X-12 20"-EL-10 20" 10-0-17 Cooling Return)

RHR (LPCI & X-51A 18"-DC-14 18" 10-IA-14  !

Shutdown X-51B 10"-DCA-10 18" 10-IB-14 i

! Cooling Return) s Core Spray. X-16A 10"-DC-14 10" 14-A-17 j X-16B 10"-DC-14 10" 14-B-17 l

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SYSTEM PENETRATION LINE NO. LINE ALTERNATE SIZE WELD RCIC X-53 3"EB-13 3" 13-0-18 RWCU X-14 6"-EA-12 6" 12-0-24 SBLC X-42 1 1/2"-DC-11 1 B-11-178 1/2" Feedwater X-9A 18"-DL-6 18" 6-A-10 X-9B 18"-DL-6 18" 6-B-8 Main Steam X-7A 20"-EB-1"A" 20" l-A-15 X-7B 20"-EB-1"B" 20" l-B-15 X-7C 20"-EB-1"C" 20" 1-C-15 X-7D 20"-EB-1"D" 20" l-D-15 X-8 3"-EL-1 3" l-50-9 l

l HPCI X-52 10"-EB-23 10" 23-0-17 Licensee's Basis for Reauestina Relief (as stated):

"Each of the lines identified in Table PRR-1.1 (above) penetrates the primary containment by means of a penetration l

assembly.similar in design to that shown in Figure PRR-1.1'.

Each of these lines have a pressure retaining circumferential weld that is inaccessible for surface and volumetric examinations due to the design of the penetration assembly.

I "As stated in 10 CFR 50.55a(g)(1) and (g)(4), for plants whose construction permits were issued prior to January 1,1971, components shall meet the requirements set forth in ASME ,

Section XI to the extent practical within the limitations of '

design, geometry and materials of construction of the components. Since ASME Section XI examination requirements did not exist at the time the Pilgrim Nuclear Power Station was designed, examination accessibility was not a primary consideration. As Figure PRR-1.1 clearly illustrates, the  !

penetration design prohibits the performance of surface or  !

, volumetric examination on the welds inside penetration. j l

" Based on the information provided above, Boston Edison Company requests relief from'the ASME Section XI requirements for i surface and volumetric examination of the subject welds." j i

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' Figures provided by the licensee are not included with this evaluation.

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Licensee's Proposed Alternative Examination (as stated):

" Surface and volumetric examination will be performed each interval on the first accessible pipe weld outside each penetration, as listed in Table PRR-1.1. Weld B-II-178 will only be examined by the surface examination method because it is a socket weld which cannot be volumetrically examined.

"The examinations required by Table IWB-2500-1, Examination Category B-P, and IWB-5000 will be conducted on the alternate weld in accordance with ASME Section XI Code.

"A VT-3 examination.of the penetrations listed in Table PRR-1.1 will be conducted each interval,- to the extent practical."

Evaluation: The Code requires that 25% of the subject Class 1 circumferential and lon;Mudinal pipe welds NPS 4 and larger receive a' volumetric and surface examination. For pipe welds less than 4 inches NPS, a surface examination is required. The licensee has requested relief from the Code-required i nondestructive examinations on welds that are inaccessible because they are within containment penetrations. Based on a review of the penetration design associated with the welds listed above, it has been catermined that the Code-required examinations are impractical. To perform the Code-required examinations, modifications to the containment penetration

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would be required. Imposition of this requirement would cause a considerable burden on the licensee.

The licensee proposes to substitute the first accessible pipe weld outside each penetration as listed in the above table and j perform a surface and volumetric examination. Based on the )

i licensee's alternative, to examine the first accessible weld l

outside of the penetration, and the system leakage tests, it  !

can be concluded that generic degradation, if present, will be  !

detected. As a result, reasonable assurance of operational readiness will be provided.

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Conclusion:

Based on the above evaluation, it is concluded that performing the Code-required examination of the welds within the containment penetration is impractical. Based on i

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9 the scheduled examinations on the first accessible weld outside each penetration, in combination with leak tests, reasonable assurance of operational readiness will be provided.

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

, 3.1.4.2 Egauest for Relief PRR-23. Examination Cateoory B-J.

, Item B9.12. Exarnination of Class 1 Lonaitudinal Pioina Welds

l Code Reauirement: Section XI, Table IWB-2500-1, Examination Category B-J, Item B9.12 requires 100% surface and volumetric I examination of 12 inches of Class 1 piping longitudinal welds in piping 4 inch NPS and larger as defined in j Figure IWB-2500-8.

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Licensee's Code Relief Reauest: The licensee requested relief from performing 100% of the Code-required volumetric and/or

, surface examination for Class 1 longitudinal piping welds.

Licensee's Basis for Reauestino Relief (as stated):

" Based on the reasons stated below, the performance of surface and volumetric examination on longitudinal piping welds has a a negligible compensating effect on the quality or safety of l Class 1 piping, In addition, there is little, if any,

technical benefit associated with the performance of these examinations, but they result in a substantial man-rem exposure ,

and cost. l "1) Throughout the nuclear industry, there has been no evidence of rejectable service induced flaws being attributed to longitudinal' piping welds.

l "2) During the first inservice inspection interval at the Pilgrim Station, no inservice flaws have been detected in longitudinal piping welds.

"3) There are distinct differences between the processes used in the manufacturing of longitudinal and circumferential welds which enhance the integrity of longitudinal welds. First, longitudinal welds are typically manufactured under controlled shop conditions whereas circumferential welds are produced in the field under less ' ideal conditions. Secondly, longitudinal welds usually undergo heat treatment in the shop which improves their material properties and relieves the residual stresses 13 j

$ created by welding. Finally, shop manufacturing inspections can be performed under more favorable conditions which further increase the confidence level of the longitudinal weld quality.

"4) During field installation of piping, the ends of the longitudinal welds may be-affected during welding of the intersecting circumferential field welds. This small area falls within the circumferential weld inspection boundaries.

Therefore, the ends of the longitudinal welds will'still be subject to examination.

"5) From an industry-wide standpoint, there has been no evidence of longitudinal weld defects compromising safety at nuclear generating facilities.

"6) No significant loading or known material degradation mechanisms have become evident to date which specifically relate to longitudinal seam welds in nuclear plant piping.

"7) There is a significant accumulation of man-rem exposure and cost associated with the inspection of Class 1 longitudinal piping welds.

"8) The alternative examinations proposed below provide an acceptable level of quality and safety without causing undue hardship or difficulties."

Licensee's Proposed Alternative Examine.tjm (as stated):

" Surface and volumetric examination shall be performed, as applicable, on the length of the longitudinal weld that is normally examined during inspection for the intersecting circumferential weld (s). The volumetric examination at the intersection of circumferential and longitudinal welds will include both transverse and parallel scans within the length of the longitudinal weld that falls within the circumferential weld examination boundary."

Evaluation: The licensee,has requested relief from 100%

volumetric and/or surface examination of the longitudinal welds to the extent required by the Code. The licensee contends that longitudinal welds are unlikely to fail; this is the result of fabrication control's and lack of susceptibility to conditions that lead to failure. Furthermore, increased radiation dose would be incurred in performing the Code-required volumetric examination. The licensee proposes to examine the potentially critical portions of the longitudinal welds (that portion that intersects the circumferential weld) in conjunction with the 14

4 circumferential welds. The volumetric examination will include both transverse and parallel scans of the length of longitudinal weld that falls within the circumferential weld examination volume.

Based on the extent of the surface and volumetric examinations that will be performed in conjunction with examination of the associated circumferential weld, .the INEL staff believes that an acceptable level of quality and safety will be provided.

! This position is supported by Code' Case N-524, Afternative

Examination Requirements for longitudinal Welds in Class 1 and Class 2 Piping; this Code case has been found by the NRC to be an acceptable alternative to the Code requirements.

Conclusion:

An acceptable level of quality and safety is provided by the licensee's proposed alternative because potentially critical portions of longitudinal welds will be examined in conjunction with circumferential welds. Therefore, it is recommended that the proposed alternative to the examination of longitudinal piping welds be authorized pursuant to 10 CFR 50.55a(a)(3)(i).

3.1.5 Pumo Pressure Boundary (No relief requests) 3.1.6 Valve Pressure Boundary (No relief requests) 3.1.7 General (No relief requests) 3.2 Class 2 Components 3.2.1 Pressure Vesegli (No relief requests) 15

l*

1 i

i

} 3.2.2 Pinina i

3.2.2.1 Reauest for Relief PRR-7. Revision 2. Examination j Catecory C-F-2. Ites C5.51. Class 2 Containment Atmospheric

[ patrol System Pi_pe Welds Code Reauirement: Section XI, Table IWC-2500-1, Examination j Category C-F-2, requires volumetric and surface examination of

7-1/2. percent of. Class 2 circumferential welds as defined by

Figure IWC-2500-7.

i

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

) performing the Code-required nondestructive examinations on j pipe welds in the following lines:

1

[ LINE(S) FUNCTION EXTENT OF-

EXAMINATIONS .

20"-HM-45/20"HE-45 Purge Line to From Torus j Torus Penetration X-205 to Valve A0-5036B t

20"-HM-45/20"-HBB-~ Vent Line from From Torus 45/8"-HSB-45 Torus Penetration X-227 j to: 1) Secondary i Containment Vacuum Breakers X-212A and .

4 X-212B 2) Valve

' A0-5042A 3) Valve A05025 1

20"-HM-45/20"-HE- Purge Line to From Drywell 45 Drywell Penetration X-26 to Valve A0-5035B i 18"-HM-45/ Vent Line from From Drywell 8"-HM-45/ Drywell Penetration X-25 to 8"-HE-45 Valve A0-5044A

' Licensee's Basis for Reouestina Relief (as stated):

" Boston Edison Company requests relief from the requirement to perform nondestructive examinations in accordance with ASME Section XI, Table IWC-2500-1, Examination ~ Category C-F-2 on Code Class 2 lines listed above. Based on the discussion below, these examinations are considered redundant and without a compensating increase in the level of quality and safety.

16

l

!. l "The lines listed abcve are portions of non-safety related

! piping systems that penetrate the primary reactor containment.

! At each containment penetration, the process' pipe i classification has been upgraded to Code Class 2 in accordance i .with commitments made in the Pilgrim Station FSAR. The piping and valves are considered part of the primary reactor containment and upgraded to Code Class 2 at the penetration only to support the primary reactor containment safety

function. Except for this, the lines listed above provide no j safety function.

4 "The function of Containment Atmosphere Control piping is to t

maintain an inert atmosphere inside containment and to provide i a flow path to release excess pressure build-up into secondary ,

! containment under extreme conditions. This system normally operates at a temperature of 50*F and a pressure of 1 psig.

Since the system is normally dry, and not subject to high

temperatures and or pressures, the probability of failure is negligible. Even in the highly unlikely event of a failure

occurring in a weld on one of these lines, the consequences of j the failure would be insignificant.

i i "The primary reactor containment integrity, including all

containment penetrations, is periodically verified by i performing leakage tests in accordance with 10 CFR 50,

( Appendix J. Each of the Code Class 2 segments of the lines ,

i listed above and their associated isolation valves are tested  !

during an Appendix J, Type A, B or C leakage test. The Type A leakage test is performed three times in a ten year interval

! and the Type B and C leakage tests are performed at intervals -

4 not greater than 24 months. Performance of the Appendix J 1eak  !

! tests will verify the integrity of the subject Code Class 2  !

lines at each respective penetration. -The performance of ASME

! Section XI, Examination Category C-F-2 nondestructive j examinations on these same lines will provide little, if any, '

! additional verification of primary reactor containment

integrity. Based on this, the performance of Examination

! Category C-F-2 examinations on these lines is considered by

! Boston Edison to be unnecessary and provides a negligible

! increase in the level of. quality or safety.

" Note: The portions of the Containment Atmosphere Control piping-that are NPS 8 have nominal wall thickness which is less than 3/8". In accordance with Table IWC-2500-1, Examination Category C-F-2, no nondestructive examinations are_ required on these small _ sections of piping."

l Licensee's Proposed Alternative Examination (as stated):  !

" Boston Edison shall perform 10 CFR 50, Appendix J 1eakage tests on the primary reactor containment penetration lines r.u _ listed above, and on their associated valves, in accordance with Pilgrim Station procedures. No further alternate examinations are necessary in this case." _

17

l.

} Eva'luation: The licensee noted in the basis for requesting

! relief that the subject system normally operates at a

}' temperature of 50*F and a pressure of 1 psig. When applying h the exemption criteria of IWC-1222(c), this piping can be exempted from volumetric and surface examinations. Therefore, it is concluded that relief is not required.

Conclusion:

Relief is not required.

3.2.2.2 Reauest for Relief PRR-17. Examination Cateaory C-F-2.

Item C5.51. Examination of Class 2 Pine Welds i

Code Reauirement: Table IWC-2500-1, Examination Category C-F-2, Item C5.51, requires 100% volumetric and surface  ;

examination of circumferential-pipe welds equal to or greater l than 3/8 inch wall thickness' as defined in Figure IWC-2500-7. I Licensee's Code Relief Reauest: The licensee requested relief from the examination of Class 2 pipe Welds HC 10-F65 and j GB-10-F244 because they are inaccessible.  ;

Licensee's Basis for Reauestina Relief (as stated):

"As stated in 10CFR50.55a(g)(1) and (g)(4), for plants whose construction permits were issued prior to January 1, 1971, 4 components shall meet the requirements set forth in ASME Section XI to the extent practical within the limitations of design, geometry and materials of construction of the components. Since ASME Section XI examination requirements did not exist at the time the Pilgrim Nuclear Power Station was  !

designed, examination accessibility was not a primary consideration.

" Accessibility for examination of the welds listed in I Table PRR-17.1 was not provided in the original plant design.

The limiting factor creating the inaccessibility of the welds listed in Table PRR-17.1 are as follows:

• Weld HB-10-F65 is located in a pipe chase enclosed by safety-related 30 inch thick blockwalls 63.10, 63.11 and 63.12.

. Boston Edison believes that no appreciable assurance of component or system integrity is gained by dissembling the blockwalls for inspection access every 10 years compared to the inspection of a similar accessible weld (HB-10-F63) on the same line.

18

Weld GB-10-F244 is inaccessible for surface examination due i to the close proximity of adjacent pipes in the ceiling of the

"A"' Auxiliary Bay. The selection of one of the nearest accessible welds (HB-10-F239) on the same line as an alternate, will provide an equivalent assurance of component and system integrity.

l i

"This constitutes a basis for relief from the examination of requirements of ASME Section XI.

" Note: Relief.was granted for these two welds during the Second Inservice Inspection Interval. Examination requirements in ASME Section XI have been revised since that time. Per the 1989 Edition of Section XI, there are-no examinations required for piping welds less than 3/8" nominal wall thickness. Since l

3 the nominal wall thickness for weld HB-10-F65 is less than l 4 3/8", examinations are not required for this weld and therefore  !

, a request for relief is not necessary. It has been retained in l

! Relief Request PRR-17 for reference only. Weld GB-10-F244 has '

a nominal wall thickness of 3/8" and therefore still requires i

relief."

I Licensee's Proposed Alternative Examination (as stated):

l

" Alternate welds have been scheduled for examination-during the i j Third Inservice Inspection Interval as shown on Table PRR-17.1. '

Continued reasonable assurance of component and system j integrity is provided because the selected alternate welds are i of the same system, pipe run, size and material as the i inaccessible welds." .

l Evaluation: The Code requires a volumetric and surface l_ examination of piping circumferential Weld GB-10-F244.

] However, accessibility for the examination of this weld was not l provided in the original plant design, which occurred prior to '

) the issuance of Section XI inservice inspection requirements. l This weld is inaccessible due to the c' lose proximity of other piping. Therefore, the Code examination ~ requirements are I impractical. To perform the volumetric and surface examination to the extent required by the Code, design modifications would be necessary. Imposition of this requirement would cause a considerable burden on the licensee.

As an alternative to examination of the subject weld, the licensee has selected for examination another weld located in the same pipe run of the same size, type, and material as the 19

, , . ---- -, . , , ,_-,-.-y-- ,

inaccessible weld. Based on the selection and examination of this alternative weld, reasonable assurance of structural integrity will be provided.

Conclusions:

The volumetric and surface examination of the subject Class 2 pressure-retaining weld is impractical. The selection of an alternate weld and the VT-2 visual examination associated with the Class 2 system pressure tests provide reasonable assurance of the continued structural integrity of the Class 2 welds. Therefore, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(1).

3.2.3 Pumos (No relief requests) 3.2.4 Valves (No relief requests)

~

3.2.5 General (No relief requests) 3.3 Class 3 Comoonenti (No relief requests) 3.4 Pressure Tests 3.4.1 Class 1 System Pressure Tests 3.4.1.1 Reauest for R'elief PRR-21. Revision 1. Examination Cateaory B-P. Items B15.50. B15.60. and B15.70. Pressure Tests of Class 1 Pioina. Pumos and Valves Code Reauirement: Section XI, Table IWB-2500-1, Examination Category B-P, Items B15.50, B15.60, and B15.70 require system pressure tests each period and system hydro' static tests once each interval.

Licensee's Code Relief Reauest: Relief is requested from the Code-required pressure tests following the opening and

• - ~~-

reclosing of mechanical connections.

20

l L- j j Licensee's' Basis for Reauestino Relief (as stated):

) " Relief is requested from the full pressure requirement of the

IWB-5221(a) leakage tests on the basis of impracticality as I

!- cit'ed in the cases below. l i

i

It is sometimes necessary to rework and examine mechanical  !

connections after the Class I system leakage test has been

completed. Examples include the following

Safety relief valve flanged connections which were blanked

j. off for the system pressure test l Leaking mechanical connections which were discovered during j the system leakage test and reworked following depressurization Control rod drive mechanism repair or change out

! "The leakage acceptability of reworked mechanical joints is i

presently determined at the nominal pressure associated with 5%

power just prior to drywell inerting. This translates to a reactor pressure of 930 psig rathe'r than the 1035 psig nominal pressure associated with 100% reactor power.

Inspection inside the drywell is not feasible above 5% power because of adverse radiation and temperature levels combined with an inert atmosphere. Requiring a VT-2 visual inspection during a system pressure test at operating pressure would require BEco to make a full power drywell entry. In this ,

operating mode, examiners would be subjected to adverse i conditions due to radiation 'and high temperature levels as well as the inert atmosphere. Imposing this requirement would require special precautions such as using ice packs and cool air supply lines to perform the VT-2 visual inspections. These conditions may compromise the quality of the inspection and '

impose potential safety concerns and hazardous conditions for examination personnel. Examiners performing VT-2 visual inspection in a less adverse environment are more likely to be able to perform a higher quality examination and see evidence of leakage, when occurring.

" Full pressure testing at 1035 psig would require an alternate method of pressurization which potentially would only have a marginal increase in leakage rates and a disproportionate impact on outage schedule. When considering the minimal increase in pressure applied to piping systems at operating pressure versus the pressure attained at 5% power, it is believed that 930 psig will result in a detectable leakage rate, if leakage is going to occur."

Licensee's Proposed Alternative Examination (as stated):

" Boston Edison will perform the system leakage test for the special situation mechanical joints at 930 psig. Disposition 21

1 1

L.

! of observed leakage will consider the marginal increase in leakage rates that would occur at the nominal operating pressure associated with 100% rated reactor power."

Evaluation

Paragraph IWB-5221(a) requires that system leakage j' test's be performed at a test pressure not less than the nominal l operating pressure associated with 100% rated reactor power.

i To obtain nominal operating pressure for Pilgrim Nuclear Power l Station, the reactor must achieve 100% power. Relief is being i requested from the Code-required. system pressure test at

j. nominal operating pressure following the reassembly of i nonisolable Class 1 mechanical connections.

t b

To perform a system leakage test at 100% power for nonisolable

{. portions of a system following the-disassembly and reassembly of Class 1 mechanical connections is a major effort requiring i i

many manhours from skilled maintenance and inspection personnel 1 while causing excessive radiation exposure. As an alternative l to the system pressure test at operating pressure, the licensee proposes to perform the system leakage test at a pressure of at least 930 psig, at approximately 5% power during startup. The l INEL staff believes that the proposed test pressure (approximately 90% of operating pressure) will be adequate to

)

cause leakage from the mechanical connection following opening j and reassembly of the component if. the leak-tight connection '

has not been established. l l

The INEL staff believes that requiring the licensee to perform a system pressure test at 100% reactor power will result in a hardship without a compensating increase in quality and safety.  !

Based on the proposed alternative test pressure of at least 930 psig, it is reasonable to conclude that leakage, if it should occur, will be detected, providing reasonable assurance of operational readiness.

Conclusion:

For Pilgrim Nuclear Power Station, a system pressure test at 5% reactor power (930 psig) should provide 22

i reasonable assurance of continued operational readiness of mechanical connections. Requiring the licensee to perform a system pressure test at 100% reactor power would result in a hardship without a compensating increase in quality and safety.

J Therefore, it is recommended that the proposed alternative, i

performing a system pressure test at 5% reactor power I (930 psig) following the reassembly of Class 1 mechanical 1 connections, be authorized pursuant to 10 CFR 50.55a(a)(3)(ii.).

i i 3.4.2 Class 2 System Pressure Tests 3.4.2.1 Reauest for Relief PRR-13. Revision 1. Examination

! Cateaory C-H. Item C7.30. C7.40. C7.70, and C7.80. Pressure Tests of Class 2 Pioina and Valves at Containment Penetrations j Where the Balance of System is Nonclass 1

Code Reauirement: Section XI, Table IWC-2500-1, Examination Category C-H, Items C7.30, C7.40, C7.70, and C7.80 require i system pressure tests each period and system hydrostatic tests once each interval.

Licensee's Code Relief Reauest: Relief is requested from l pressure tests of containment penetration piping where the balance of the piping system is nonclass.

, PENETRATION LINE DESCRIPTION 1 NUMBER NUMBER X-2 45-H0-143 Personnel Airlock Test Line X-6 45-H0-172 Drive Removal Test Line X-21 2"-JB-32 Service Air X-22 3"-HCB-31 Instrument Air X-23 6"-HE-30 RBCCW Supply X-24 8"-HE-30 RBCCW Return X-25 18"-HM-45 Vent from Drywell l X-26 18"-HM-9 Vent to Drywell i

23

PENETRATION LINE DESCRIPTION NUMBER NUMBER X-33A 1"-DC-10 Recirculation Pump Seal Leak Detector X-33B 1"-DC-10 Recirculation Pump Seal Leak

, Detector X-43 45-HO-166 Drywell Test Connection X-46A 1"-DCB-2 Recirculation Pump Seal Purge X-46B 1"-DCB-2 Recirculation Pump Seal Purge X-46D 1"-HCB-45 Drywell Pressure Line X-46E 1"-DC-45 Reference Vessel Pressure Line X-47 45-HO-102 Drywell Test Connection X-49A 1"-DC-2 Recirculation Pump Seal l X-49B 1"-DC-2 Recirculation Pump Seal X-205 20"-HM-45 Torus Purge, Inlet X-218 4"-HCB-45 Torus Spare X-219 4"-HCB-45 HPCI Turbine Exhaust Vacuum and Hydrogen Recombiner Vent X-227 20"-HM-45 Torus Purge Exhaust Vacuum Relief and Direct, Torus Vent  ;

X-228A 1"-DC-45 Torus Pressure  !

X-2288 1"DC-45 Reference Vessel Pressure Licensee's Basis for Reauestina Relief (as stated):

" Boston Edison Company requests relief from the requirement to perform a pressure' test in accordance with ASME Section XI, Table IWC-2500-1, Examination Category C-H on the Code Class 2 lines listed above. Based on the discussion below, these pressure tests are considered redundant and without a compensating increase in the level of quality and safety.

"The lines 'above are portions of non-safety related piping systems that penetrate the primary reactor containment. At each containment penetration, the process pipe classification has been upgraded to Code Class 2 in accordance with commitments made in the Pilgrim Station FSAR. The piping and valves are considered part of the primary reactor containment and upgraded to Code' Class 2 at the penetration only to support the primary reactor containment safety function. Except for this, the lines listed above provide no safety function.

24

]

"The lines shown on P&ID M227'Sht. I are containment atmospheric control lines which cannot be isolated for pressure testing because they are open to the primary containment atmosphere. The sample lines shown on M227 Sht. I can be isolated outside of containment but would require that l extensive supports, test taps, vent lines and drain lines be l added for pressure testing the Class 2 piping segments. This <

same problem applies to many of the lines listed above which l are isolable from both inside and outside containment. <

"The primary reactor containment integrity, including all containment penetrations, is periodically verified by performing leakage tests in accordance with a 10 CFR 50, Appendix J. Each of the Code Class 2 segments of the lines listed above and their associated isolation valves are tested during an Appendix J, Type A, B or C leakage test. The Type A leakage test is performed three times in a ten year interval and the Type B and C leakage tests are performed at intervals not greater than 24 months. Performance of these Appendix J 1eak tests will verify the integrity of the subject Code Class 2 lines at each respective penetration. The performance a of ASME Section XI, Examination Category C-H pressure tests on i;

these same lines will provide little, if any additional verification of primary reactor containment integrity. Based i on this, the performance of Examination Category C-H pressure '

! tests on these lines is considered by Boston Edison to be unnecessary and provides a negligible increase in the level of quality or safety."

4 Licensee's Prooosed Alternative Examination (as stated):

" Boston Edison shall perform 10 CFR 50, Appendix J 1eakage tests on the primary reactor containment penetration lines listed above, and on their associated valves, in accordance with Pilgrim Station procedures."

Evaluation: The portions of the subject lines for which the licensee has requested relief from the Code-required pressure  !

tests are Class 2 at the penetration only. These segments of j lines are safety-related because they fuilction as part of the l containment pressure boundary and are relied on for containment integrity. Therefore, it is logical to test the penetration l piping portion-of the associa.ted system to containment test criteria as specified in Appendix J.

Appendix J pressure tests verify the leak-tight integrity of the primary reactor containment and of systems and components 25

1 .

4 i-

, that penetrate containment by local leak rate and integrated

, leak rate tests. In addition, Appendix J test frequencies i provide assurance that the containment pressure boundary is being maintained at an acceptable level while monitoring for

deterioration of seals, valves, and piping.

I

The Class 2 containment isolation valves (CIVs) and connecting 4

pipe segment must withstand the peak calculated containment

[ internal pressure'related to the maximum design containment

] pressure. The INEL finds that the pressure-retaining integrity l

of the CIVs and connecting piping.and their associated safety functions may be verified with an Appendix J, Type C test if

! conducted at the peak calculated containment pressure. The-f' seal between the connecting pipe segment and containment may be

verified using an Appendix J, Type B test. Therefore, when the connecting pipe segment is subjected to either a Type B or C

! test, its . safety function is verified by the Appendix J test.

Section XI, IWC-5210(b) requires.that where air or gas is used as a testing medium, the test procedures shall include methods .

) for detection and location of through-wall leakage' in components of the system tested. If the licensee's test '

procedure uses air as a' testing medium, .the procedure should meet the above requirement for the CIVs and pipe segments between the CIVs.

i The INEL staff believes that an acceptable level of quality and safety will be provided by Appendix J tests, provided that the licensee performs the leak test at the peak calculated containment design pressure and that a test procedure is implemented that provides for detection' and location of  !

through-wall leakages in the pipe segments that are being tested.

I

Conclusion:

It is recommended that the licensee's proposed  !

alternative to the Code-required pressure tests be authorized pursuant to 10 CFR 50.55a(a)(3)(i), provided that the licensee l

performs the leak test at the peak calculated containment 26

t l pressure and uses a test procedure that provides for detection '

and location of through-rall leakages in the pipe segments

, being tested.

3.4.3 Class 3 System Pressure Tests 3.4.3.1 Reauest for Relief PRR-11 Revision 1. Examination Cateaory D- L Item D2.10. System Functional and System Hydrostatic Tests of

Salt Service Water Pumos 4

Code Reauirement: Section XI, Table IWD-2500-1, Examination Category D-B, Item D2.10 requires a system functional test each period and a system hydrostatic test each interval.

Ljcensee's Code Relief Reauest: Relief is requested from performing the hydrostatic system pressure test on the Salt j Service Water Pumps, P-208A, P-2088, P-208-C, P208-D, and i l

P208-E.

Licensee's Basis for Reauestina Relief (as stated):

"The Salt Service Water System has been designated Class 3 and provides cooling to the Reactor Building and Turbine Building Closed Cooling Water Systems. The system includes 5 suction

. pumps that are submerged under water.

. " System pressure testing of the Salt Service Water System pumps and associated discharge lines would require disassembly and removal, special testing, and reassembly of the pumps. The requirements to remove the pumps for the sole purpose of performing a test of the pressure boundary has a negligible impact on plant safety and a disproportionate impact on expenditures of plant manpower. The integrity and function of the pumps is already verified by the performance of quarterly full flow tests. In addition, in the unlikely event of a pump failure, there is multiple redundancy provided by the remaining 4 pumps.

" Relief is requested from the requirements to perform a system pressure test on the Salt Service Water pumps up to the expansion joints in the pump discharge lines on the basis of impracticality and a negligible affect on plant safety."

l 27

Licensee's Proposed Alternative Examination (as stated):

"VT-2 examination of pump discharge piping up the expansion joints will be conducted during system functional pressure tests performed on the remainder of the system once each period as required by ASME Section XI."

Evaluation: The Code requires that the subject Class 2 piping receive a system hydrostatic test once every interval. Based on the review of the licensee's submittal, which includes a 2

piping and instrumentation drawing , it appears that there are no test taps or other practical ways to isolate the section of pump discharge piping to perform a hydrostatic pressure test.

Therefore, the Code-requirad hydrostatic pressure test is impractical. Performing the pressure test on the subject pipe would require the . removal of the pump and isolation of a portion of the discharge pipe, resulting in.a burden on the licensee.

The licensee has proposed to perform a VT-2 visual examination on the discharge pipe from the pump to the expansion joint during system functional tests. Considering the function of the service water system--to provide cooling water under constant flow--performing a VT-2 visual examination while the system operates will provide reasonable assurance of operational readiness.

Conclusion:

It has been determined that performing the Code-required hydrostatic pressure test on the Salt Service Water pumps up to the expansion joints is impractical. Performing a VT-2 visual examination during functional tests will provide reasonable assurance of operational readiness. Therefore, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i).

3.4.4 General (No relief' requests) 2 Not included with this evaluation.

28 I

1:O i

3.5 General l 3.5.1 Ultrasonic Examination Techniaues (No relief requests) i i

3.5.2 Exemoted Components (No relief requests) 3.5.3 Other 3.5.3.1 Reauest for Relief PRR-24. Paracraoh IWA-2311(b). Anoendix VII.

Ultrasonic Examination Personnel Oualification Reauirements Code Reauirement: Section XI, Paragraph IWA-2311(b) requires that the training, qualification, and certification of ultrasonic examination personnel shall also comply with the requirements of Appendix VII.

Appendix VII specifies the employer /s written practice, qualification of u.ltrasonic examiners, qualification records, and the minimum content of initial training courses for the ultrasonic examination method.

Licensee's Code Relief Reauest: The licensee requested relief from the Appendix VII requirements for the qualification of l nondestructive examination personnel for ultrasonic examination. '

Licensee's Basis for Reauestino Relief (as stated): a

" Boston Edison requests relief from implementation of Appendix VII until the performance demonstration requirements of Appendix VIII are fully implemented. Implementation of Appendix VII prior to full implementation of Appendix VIII is considered impractical and without a compensating increase in quality of safety.

1

" Appendix VIE was first introduced in' the 1988 Addenda to Section TI. This Appendix represents a dramatic change from previous Code editi.ons and current industry practices in the requirements for qualification of ultrasonic examination personnel. New training programs must be developed and taught by trained instructors, employers written practices must be l

l c'9  !

j j

completely rewritten, examination question banks must be developed, flaw specimen containing actual or simulated flaws must be acquired, and performance demonstration (practical j examinations) must be completed.

a 2-

" Implementation of Appendix VII will require a substantial I

l' industry effort. Although work is progressing towards compliance with Appendix VII, full implementation has not yet

been achieved. Since Appendix VII provides for use of

specimens prepared for ultrasonic performance demonstrations per Appendix VIII, many NDE vendors are developing.these two programs concurrently in order to avoid duplicated effort.

! - Though currently not required, the nuclear industry anticipates i that the Appendix VIII performance demonstration requirements

will be mandated by a backfit ruling in the Federal Register.

i In anticipation of this ruling, the Performance Demonstration Initiative (PDI) Committee is currently leading an industry l wide effort to implement Appendix VIII. The tentative i completion dates for pipe weld performance demonstrations and i reactor vessel performance demonstration are January of 1996, i and January of 1997, respectively.

"The Boston Edison Company intends to fully implement i Appendix VII when the performance demonstrations of Appendix VIII are mandated by a backfit ruling in the Federal l Register."

f Licensee's Proposed Alternative Examination (as stated):

} "The Pilgrim Station shall utilize ultrasonic examination j personnel qualified in accordance with the requirements of

! IWA-2300, except for IWA-2311(b). The additional Appendix VII i training, qualification, and certification requirements

! referenced in IWA-2311(b) shall be fully implemented when the

{ performance demonstrations of Appendix VIII are mandated by the

! Federal Register."

j Evaluation: Appendix VII was incorporated in the 1988 Addenda j to the 1986 Edition of ASME Section XI to enhance ultrasonic l

examination flaw detection. This appendix specifies '!

{ administrative and examination qualification requirements.

l Although Appendices VII and VIII both have requirements related '

! to flaw detection in ultrasonic examinations, their concurrent implementation is not necessary. Certain requirements of

Appendix VIII may strengthen the efforts of Appendix VII, but

} they are not necessary for its implementation.

I l

't 2 30 4

..v,, . ,_ . -, ,, . , _ . _ _ ......,._.m, - , , , . y, - . , .

i The INEL staff believes that the licensee has had sufficient time to develop an Appendix VII program. Appendix VII has been generally adopted through-out the industry. Although Appendix VIII will further improve flaw detection confidence, its implementation is not required in conjunction with Appendix VII. An Appendix VII program will increase quality and safety and is not considered impractical.

Conclusion:

The INEL staff does not believe that the licensee j has sufficient basis for relief from the subject requirements.

i -

Therefore, it is recommended that relief be denied.

i 3.5.3.2 Reauest for Relief PRR-18. Code Case N-491. Table IWF-2500-1.

Cateaorv F-A. Items F1.10 and F1.20. Examination of Component j Sucoorts Code Reauirement: Code Case N-491, Table IWF-2500-1, Category j F-A, Items F1.10 and F1.20, Examination of Component Supports, requires VT-3 visual examination of component supports.

4 Licensee's Code Relief Reauest: The licensee requested relief from the Code-required VT-3 visual examination of component Supports H-10-1-13SA and H-4-1-6.

Licensee's Basis for Reauestina Relief (as stated):

"As stated in 10 CFR 50.55a(g)(1) and (g)(4), for plants whose construction permits were issued prior to January 1, 1971, components shall meet the requirements set forth in ASME Section.XI to the extent practical within the limitations of design, geometry and materials of construction of the components. Since ASME Section XI examination requirements did not exist at the time the Pilgrim Nuclear Power Station was designed, examination accessibility was not a primary consideration.

" Accessibility for the examination of the supports listed in Table PRR-18.1 was not provided in the original plant design.

The limiting factors creating the inaccessibility of the supports listed in Table PRR-18.1 are as follows:

31

• Support H-10-1-13SA is located.in a pipe chase enclosed by safety-related 30 inch thick blockwalls 63.10, 63.11 and 63.12.

Boston. Edison believes that no appreciable assurance of compon'ent or system integrity is gained by disassembling the blockwalls for inspection access every 10 years compared to the inspection of similar support on the same line.

• Support H-4-1-6 is located underneath the reactor pressure vessel. Due to the proximity of the control rod drive mechanisms, this support is not accessible.for inspection.-

"The above information constitutes a basis for relief from the examination requirements of ASME Section XI and Code Case N-491."  :

Licensee's Proposed Alternative Examination (as stated):

" Boston Edison Co. is already performing visual examinations on a sufficient nunaber of supports on the same system lines to meet the criteria of Code Case N-491. It is therefore reasonable to conclude that generic support problems, if existing, will be detected."

Evaluation: The licensee stated that they are implementing Code Case N-491, which requires a percentage of components to be selected for examination, and that they are examining a sufficient number cf component supports to meet the criteria of this Code case. Because Code Case N-491 is approved for use in Regulatory Guide 1.147, the INEL staff has determined that relief is not required.

Conclusion:

Because the licensee is satisfying the requirements of Code Case N-491, which has beer, approved for use as listed in Regulatory Guide 1.147, relief is not required.

3.5.3.3 Reauest for Relief PRR-22. Reauest for Relief From Reaulatory Reauirement To Obtain Relief' From Code Reauirements In the February 15, 1996, response to the NRC request for additional information, the licensee withdrew this relief request.

32 l

__ _ ._ ~ _ . .. . _ _ . _ . . _ _ _ _ _ _ _ _ .._.______.m

.- l i 4. CONCLUSION J

l Pursuant to 10 CFR 50.55a(g)(6)(i), it has been determined that certain inservice examinations cannot be performed to the extent required by q Section XI of the ASME Code. For Requests for Relief PRR-1, PRR-9, PRR-11 l Rev.1, and PRR-17, the licensee has demonstrated that specific.Section XI q requirements are impractical; it is therefore recommended ,that relief be l

j granted. The granting of 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 I l requirements were imposed on the facility.

l Pursuant to 10 CFR 50.55a(a)(3), it is concluded that for Requests for Relief l PRR-21 and PRR-23, the licensee's proposed alternatives provide an acceptable

level of quality and safety in lieu of the Code required examination and

_should be authorized.

In addition, pursuant to 10 CFR 50.55a(a)(3), it is recommended that the j licensee's proposed alternative be authorized for Request for Relief PRR-13,

provided that the licensee satisfies the conditions stated in the request for

! ' relief evaluation.

For Request for Relief PRR-24, it is concluded that the licensee has not provided sufficient justification to support the determination that the Code requirement is impractical, or that requiring the licensee to comply with the Code requirement would result in hardship. Therefore, it is recommended that relief be denied.

I Based on the review of Request for Relief PRR-7 and PRR-18, it has been determined that relief is not required. As a result of the request for additional information, the licensee withdrew Request for Relief PRR-22. ,

l 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 Pilgrim Nuclear Power Station.

Compliance with all these requirements would necessitate redesign of a '

significant number of plant systems, procurement of replacement components, 33

4?

1 3

[ installation of the new components, and performance of baseline examinations for these components. Even after the redesign efforts, complete compliance

~ with the Section XI examination requirements probably could not be achieved.

l- Therefore, it 1s concluded that the public interest is not served by imposing

certain provisions of Section XI of the ASME Code that have been determined to i i be impractical.

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

i i

Based on the review of the Pilgrin Nuclear Power Station, Thirti 10-Year

Interval Inservice Inspection Program Plan, Revision 0, the licensee's i

4 responses to the Nuclear Regulatory Commission's request for additional information (RAI), and the recommendations for granting relief from the ISI

{- examinations that cannot be p'erformed to the extent required by Section XI of L

the ASME Code, no deviation from regulatory requirements or commitments was i identified, with the exception of Request for Relief PRR-24.

i l

i i

j- I r

i a .-

34 i

l

• 1 l

I

5. REFERENCES

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

2. l American Society of Mechanical Engineers Boiler and Pressure Vessel Code,  ;

Section XI, Division 1: '

1989 Edition

3. Pilgria Nuclear Power Station, Third 10-Year Interval Inservice i Inspection Program Plan, Revision 0, submitted September 1, 1995.

i

4. NUREG-0800, Standard Review Plan for' the Review of Safety Analysis Reports for Nuclear Power Plants, Section 5.2.4, " Reactor Coo 3 ant i Boundary Inservice Inspection and Testing," and Section 6.6, " Ins.ervice Inspection of Class 2 and 3 Components," July 1981. '

5. Letter from NRC to BEro , dated December 6, 1995, containing request for additional information on the Third 10-Year Interval ISI Program Plan. '

r

6. Letter from L. J. Olivier (BECo), dated February 15, 1996, to the Document Control Desk, containing the response to the NRC's request for j additienal information.

i

7. NRC 1etter dated February 4,1992, NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping, (Generic Letter 88-01, Supplement 1).

8. NUREG-0619/Rev. 1, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking, July 1980.

9. NRC Bu13etin No. 88-08, Thernal Stresses in Piping Connected to Reactor Coolant Systems, June 22, 1988.

35

e N,2

• sv 333 U 5. NUCLE AR REGUL ATOAY COMMI5510N 1.

• E

. Es

• ~ '*"'"' ~

2oEN BIBLIOGRAPHIC DATA SHEET t$et ,nstr uct: ens ca rat reverseJ 2, TiTcE ANo susTirtE INEL-96/0177 Technical Evaluation Report on the Third 10-Year Interval Inservice Inspection Program Plan:

,e Boston Edison Company Pilgrim Nuclear Power Station Tune ^2"""I'*"t'[ 1996 Docket Number 50-293 W "fMf9 7 dA-A08)

5. AUTMOR(S) 6 TYPE OF REPORT Technical M.T. Anderson, E.J. Feige, K.W. Hall, A.M. Porter 3 , , , ,0 0 ca y , , , , ,,,,,,,,, ,,,,.,.

c .e,

s. Py,0,Rg ORG,4,NIZ , ATION - N AM E AND ADD R ESS ter hac. aroveer onessoa, ofrece er septen, u.1 Ausmer aeruserery commessnea, eae menser eeeess er eaerec Lockheed Idaho Technologies Company P.O. Box 1625 Idaho Falls, ID 83415-2209 S. SPONSOR ING OR G ANIZATION - N AM E AND ADDR ESS tir Nac, tree seme es ese.e , ,r eoarrecror. are .ee Nec 0 ea, orrece er aerea, u.& 4. cme, a ,verery comm,swea.

f6Yer"Tils and Chemical Engineering Branch Office of Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555

10. SUPPLEMENTARY NOTES 11 ASSTRACT (200woros er sesst This report presents the results of the evaluation of the Pilgrim Nuclear Power Station, Third 10-Year Interval Inservice Inspection (ISI) Program Plan, Revision 0, submitted September 1,1995, including the requests for relief from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI requirements that the licensee has determined to be impractical. The Pilgrim Nuclear Station, Third 10-Year Interval Inservice Inspection 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 application of system or component examinatiori exclusion criteria, and (d) compliance with ISI-related commitments identified during previous Nuclear Regulatory Commission reviews. The requests for relief are evaluated in Section 3 of this report.

J. Ava#6Assu 7 y 57mTiw&NT

12. K.E Y WOR DS/DESCR:PT ORS tc,se weres orpareses taer *,st ess,st reseseraers an socerent rae meerr.s Unlimited

14. $t Gwm T v G@b5'* tGa7 WN ITa,s Paees Unclassified

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Unclassified Ib. NUMBER OF P AGE S

16. PR ICE knc eonU 335 (249'