ML20154P680
| ML20154P680 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 05/31/1988 |
| From: | Beth Brown, Mudlin J EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | |
| Shared Package | |
| ML19318G897 | List: |
| References | |
| CON-FIN-D-6022 EGG-MS-8042, NUDOCS 8810030149 | |
| Download: ML20154P680 (33) | |
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EGG-NS-8042 4
TCCHNICAL EVALUATION REPORT ON THE SECOND 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAN PLAN:
BOSTON EDISON CONPANY, PILGRIN NUCLEAR POWEA STATION, UNIT 1 D0CKET NUNBER 50-293 I
- 8. W. Brown J. D. Nudlin l
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Published Nay 1988 Idaho National Engineering Laboratory J
8 EG1G Idaho, Inc.
l Prepared for:
U.S. Nuclear Regulatory consission J
Washington, D.C. 20555 under 00E Contract No. OE AC07-761001570 4
FIN No. 06022 (Project 5) l l
A9 goge O
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ABSTRACT This report presents the results of the evaluation of the Pilgria Nuclear Power Station, Unit 1. Second 10-Year Interval Inservice Inspection (!$1)
Program Plan, Revision 3, submitted December 12, 1986, and Amendment ISI 87-02 submitted March 2, 1988. The December 12, 1986 submittal included new and revised requests for relief from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI requirements which the Licensee has determined to be impractical. Revision 3 of the Program Plan reflects the current plant configuration including the recirculation pipe replacement made during the 1983-1984 outage. The Pilgrim Nuclear Power Station, Unit 1 Second 10-Year Interval ISI Program Plan, Revision 3, is evaluated in Section 2 of this report. The ISI Program Plan is evaluated for (a) compliance with the appropriate edition / addenda of Section XI, (b) acceptability of the examination sample, ( ) exclusion criteria, and (d) compliance with ISI-related commitments identified during the Nuclear Regulatory Commission (NRC) review of previous submittals by the Licensee. The new and revised requests for relief from the ASME Code requirements which the Licensee has determined to be impractical for the second 10-year inspection interval are evaluated in Section 3 of this report.
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l This work was funded under:
U.S. Nuclear Regulatory Commission FIN No. 06022, Project 5 1
Operating Reactor Licensing Issues Program, Review of ISI for ASME Code Class 1, 2, and 3 Components 4
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$UMMARY The Licensee, Boston Edison Company, has revised the_ Pilgrim Nuclear Power Station, Unit 1, second 10-Year Interval inservice inspection (151) Progra?.
Plan, Revision 3 to meet the requirements of the 1980 Edition, Winter 1980 Addenda (80W80) of the ASME Code Section XI. The second 10-year interval began December 8, 1982 and ends December 8, 1992.
The information in the Pilgrim Nuclear Power Station Unit 1, Second 10-Year Interval !$1 Program Plan, Revision 3, submitted December 12, 1936, and Amendment !$1 87-02 submitted March 2, 1988, was reviewed. The December 12, 1986 submittal contained ravised and new requests for relief from the ASME Code Section XI requirements which the Licensee has determined to be impractical. Revision 3 of the Program Plan reflects the current plant configuration including all modifications made during the 1983-1984 refueling outage (i.e. recirculation pipe replacement project) and utilizes i
the newer Code Edition and Addenda (80W80) for weld selection. As a result of this review, a Request for Additional Information (RAI) was prepared 3
describing the information and/or clarification required from the Licensee in order to complete the review.
Based on the review of the Pilgrim Nuclear Power Station, Unit 1, Second 10-Year Interval 151 Program Plan, Revision 3, the Licensee's response to the Nuclear Regulatory Cosmission's RAI, and the recoassendations for the granting of relief from the !$! examination requirements that have been determined to be impractical, it has been concluded that the Pilgrim Nuclear Power Station Unit 1, Second 10-Year Interval ISI Program Plan, Revision 3, isacceptableandincompliancewith10CFR50.55a(g)(4).
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CONTENTS i
A85 TRACT.................................................................
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SUMMARY
111 1.
INTR 000CTION..........................................................
1 2.
EVALUATION OF INSERVICE INSPECTION PROGRAN PLAN.......................
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2.1 Doc ume n t s Ev a l u a t ed................................................ 4 2.2 Compliance with Code Requirements..................................
5 2.2.1 Compliance with Applicable Code Editions.......................
5 2.2.2 Acceptabil i ty of the Examination Sampl e........................ 5 2.2.3 Exclusion Criteria.............................................
5 2.2.4 Augmented Exami nati on Commi tments.............................. 6 i
2.3 Conclusions........................................................
7 3.
EVALUATION OF RELIEF REQUEST5.........................................
8 3.1 Class 1 Components................................................. 8 l
3.1.1 Reactor Pressure Yesse1........................................
8 3.1.1.1 Request for Relief PRR 4, Revision 1, Examination i
Category 8 A. Items 81.11 and 81.12, Reactor Pressure Vessel Beltline Region We1ds...............................
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l 3.1.1.2 Request for Rollef PRR-5, Revision 1. Examination i
Category 8-A, Items 81.21 and 81.22 Reactor Pressure l
Vessel Botton Head We1ds...................................
8 3.1.1.3 Request for Relief PRA-9, Revision 1, Examination Category 8 0, Items 83.90 and 83.100, Reactor Pressure t
Vessel Nozzle-to-Vessel Weld and Inside Radius 5ection....................................................
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i 3.1.2 Pressurizer (DoesnotapplytoSWRs) 3.1.3 Heat E.xchangers (No relief requests) 3.1.4 Piping Pressure Boundary.......................................
9 3.1.4.1 Request for Relief PRR-1, Revision 3, Examination Category 8 J Items 89.11 and 89.21, and Examination Category 8 K 1, Item 810.10 Class 1 Circumferential i
Pressure Retaining Piping Welds and Integrally Welded i
Attachments Within Flued Head Penetrations................. 9 i
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3.1.4.2 Request for Relief PRA-6, haaination Category B-J.
Items 89.10 and 89.40, Pressure Retaining Welds in Class 1 Piping Systems....................................
12 3.1.5 Pump Pressure Boundary........................................
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3.1.5.1 Request for Relief PRA-2, Revision 0, hamination Category B L-2, Iten B12.20 Recirculation Pum Casings.......................................p 12 3.1.6 Val ve Pres sure Boundary....................................... 12 1
3.1.6.1 Request for Relief PRR-3, Revision 1, hamination i
Category B M 2, Iten B12.40, Class 1 Valve Bodies......... 12 3.1.7 General (No relief requests) 3.2 Class 2 Components................................................
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i 3.2.1 Pressure Vessels..............................................
13 3.2.1.1 Request for Relief PRA-8, Revision 1 hamination Category C-8, Ites C2.21 RHR Heat hchanger l
Nozzle-to-Vessel Welds and Inside Radius Sections......... 13 3.2.2 Piping........................................................
14 3.2.2.1 Request for Relief PRR 6, haaination Category C-F, i
Items C5.10 and C5.32. Pressure Retaining Welds in Class 2 Piping 5ystems....................................
14 3.2.2.2 Request for Relief PRR 7, Revision 1, hamination Category C F, Items C5.11 and C5.12. Pressure Reta'ning Welds in the Containment Atmospheric l
Control System...........................................
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t 3.2.3 Pumps (No relief requests) 3.2.4 Valves (No relief requests)
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3.2.5 General (No relief requests) l 3.3 Class 3 Components (Noreliefrequests) f 3.4 Pressure Tests....................................................
15 3.4.1 Class 1 System Pressure Tests (Noreliefrequests)
L 3.4.2 class 2 Systes Pressure Tests.................................
15 3.4.2.1 Request for Relief PRA 12. Revision 0, hamination i
Category C H Hydrostatic Test of the Control Rod Drive Hydraulic 5ystem....................................
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3.4.2.2 Request for Relief PRR-13. Revision 0. Examination Category C H Hydrostatic Test of the Class 2 Portions of the Contai nment Atmospheric Control System............. 16 3.4.2.3 Request for Relief PRR-15. Revision 0, Examination Category C-H, Hydrostatic Test of the High Pressure Coolant Injection Turbine Exhaust Drain Line..............
17 3.4.3 Class 3 System Pressure Tests.................................
19 3.4.3.1 Request for Relief PRR-10. Revision 0. Examination Cateqory D 8, Hydrostatic Test of Two 10-Liter Shie'ded Sample Chambers..................................
19 3.4.3.2 Request for Relief PRR-11, Revision 0. Examination Category 0-8, i.ydrostatic Test of the Salt Service Water Systis..............................................
20 3.4.4 Genera1.......................................................
22 3.4.4.1 Request for Relief PRR 14 Revision 0, F.xamination Cateqories C H, 0 A, 0 I, and D.C. Hydrostatic Test of C' ass 2 and 3 Systems Containing Relief Valves and Instrum6ntation.......................................
22 3.5 General (No relief requests) 4.
CONCLUS10N...........................................................
23 5.
REFERENCES........................................................... 25 vi
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r TECHNICAL EVALUATION REPORT ON THE SECONO 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN:
BOSTON EDISON COMPANY, PILGRIM NUCLEAR POWER STATION, UNIT 1 DOCKET NUMBER 50 293 1.
INTRODUCTION Throughout the service life of a water cooled nuclear power facility, 10 C7R 50.55a(g)(4) (Reference 1) requires that components (including supports) which are classified as American Society of Mechanical Engineers
(.45ME)BoilerandPressureVesselCodeClass1, Class 2,andClass3 meet the requirements, except the design and access provisions and the preservice examination (PSI) 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 shall comply with the requirements in the latest edition and addenda of the Code incorporated by reference in 10 CFR 50.55a(b) on the i
date 12 months prior to the start of the 120-month inspection interval, subject to the limitations and modificaticns listed therein. The components (including supports) may meet requirements set forth in subsequent editions and addenda of this Code which are incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein. The Licensee, Boston Edison Company, has prepared the Pilgrim Nuclear Poner Station, Unit 1, Second 10-Year Interval Inservice Inspection L
(ISI) Program Plan, Revision 3, to meet the requireesnts of the 1980 Edition Winter 1980 Addenda (80W80) of the ASME Code Section XI. The second 10-year interval began December 8, 1982 and ends December 8, 1992.
As required by 10 CFR 50.55a(g)(5), if the licensee determines that certain l
Code examination requirements are impractical and requests relief from them, I
the licensee shall submit information and justifications to the Nuclear Regulatory Ccan.ssion (NRC) to support that determination.
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Pursuant to 10 CFR 50.55a(g)(6), the NRC will evaluate the licensee's j
determinationsunder10CFR50.55a(g)(5)thatcoderequirementsare l
impractical. The Commission may grant relief and may impose alternative I
requirements as it determines are authorized by law and will not endanger life or property or the coauson defense and security and are otherwise in the i
public interest, giving due consideration to the burden upon the licensee t
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that could result if the requirements were imposed on the facility.
I The information in the Ptigria Nuclear Power Station, Unit 1, Second 10-Year t
Interval !$1 Program Plan, through Revision 3 (Reference 3), submitted
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December 12, 1986, and Amendment IM 97 02 (Reference 4), submitted l
Narch 2, 1988 was reviewed. This reiew included the requests for relief l
l from the ASME Code Section XI requirements which the Licensee has determined I
to be impractical. Revision 3 of the Program Plan contained six new requests for relief which document limitations in the implementation of the j
hydrostatic test program, as well as revisions to relief requests which had been evaluated by the NRC staff in previous safety Evaluation Reports i
(References 5 and 6). The review of the 15! Program Plan was performed I
using the standard Review Plans of NUREG 0400 (Reference 7), Section 5.2.4 l
' Reactor Coolant Soundary Inservice Inspections and Testing,' and i
l Section 6.6 ' Inservice Inspection of Class 2 and 3 Components'.
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i In a letter dated July 14, 1987 (Reference 8), the NRC requested additional information that was required in order to complete the review of the !$!
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Program Plan. The requested information was provided by the Licensee in
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l submittals dated October 2, 1987 (Reference 9), October 30, 1987 i
l (Reference 10), and December 28,1987(Reference 11).
In these responses, j
the Licensee provided an itemized listing of components being examined
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during the second 10 year interval, isometric drawings, a listing of the ultrasonic calibration blocks being used during the second 10-year interval, and clarifications on examinations. Two requests for relief from ASME Code i
l Section XI requirements which the Licensee had previously determined to be 7
l impractical were withdrawn.
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4 As a result of a telephone conversation on February 10, 1988, the Licensee submitted an April 15, 1988 letter (Reference 12)discussingaugmented examinations being performed at Pilgrim Nuclear Power Station, Unit 1 during the second 10-year interval.
The Pilgrim Nuclear Power Station, Unit 1. Second 10-Year Inter /a1 I5I Program Plan, Revision 3, is evaluated in Section 2 of this report. The 15!
Program Plan is evaluated for (a) compliance with the appropriate edition / addenda of Section II. (b) acceptability of examination sample.
l (c) exclusion criteria, and (d) compliance with ISI related comeitsents identified during the NRC's review of previous P11 grin. Unit 1. PSI and ISI Program Plans.
The new and revised requests for relief are evaluated in Section 3 of this report. The remaining relief requests, applicable to the second 10-year interval but not revised, are also listed in Section 3 and the document containing the staff evaluation is referenced. Unless otherwise stated, references to the Code refer to the ASME Code,Section XI, 1980 Edition including Addenda through Winter 1980. Specific inservice test (157) programs for pumps and valves are being evaluatt.d in other reports.
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2.
EVALUATION OF INSERVICE INSPECTION PROGRAN PLAN This evaluation consisted 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 (a) Pilgrim Nuclear Power station, Unit 1. Second 10-Year Interval 15!
Program Plan, Revision 3, submitted December 12, 1986 (b) Safety Evaluation Report related to requests for relief from ISI requirements for Pilgrim Nuclear Power station, Unit 1, Second 10-Year Interval !$1 Program Plan, dated August 13, 1985 (Reference 5):
(c) Safety Evaluation Report related to requests for relief from 15!
requirements for Pilgrim Nuclear Power Station Unit 1, second 10-Year Interval 151 Program Plan, dated March 26, 1987 (Reference 6):
(d) Letter dated October 2,1987, containing the Licensee's response to the NRC's July 14, 1987 request for additional information with regard to Revision 3 of the 15! Program Plant (e) letter dated October 30, 1987, containing additional response from the Licensee with regard to the NRC's July 14, 1987 request for additional informations (f) Letter from the Licensee dated December 28, 1987, containing the final information requested in the NRC's July 14, 1987 request for additional informations i
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(g) Letter dated March 2, 1988, containing Amendment !$1 87 02 to Revision 3 of the Pilgria Nuclear Power Station, Unit 1. Second j
10 Year Interval 15! Program Plant and (h) Letter dated April 15, 1988, containing information about the augmente<; examinations being performed at Pilgrim Nuclear Power
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Station, Unit 1, during the second 10-year interval.
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2.2 Conn 11ance with Code Reautrements l
j 2.2.1 C=aliance with Aonlicable Code Editions I
l The Inservice Inspection Program Plan shall be based on the Code editions t
o defined in 10 CFR 50.55a(g)(4) and 10 CFR 50.55a(b). Based on the j
l starting date of December 4,1982, the Code applicable to the second 10-year interval is the 1980 Edition with Addenda through Winter 1980. As stated in Section 1 of this report, the Licensee has prepared the Pilgria l
Nuclear Power Station, Unit 1. Second 10-Year Interval !$! Program Plan, Revision 3, to meet the requirements of the 1980 Edition, Winter 1980 Addenda (80W80) of the ASME Code section XI.
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2.2.2 &ccentability of the Examination Samole i
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Inservice volumetric, surface, and visual examinations shall be performed
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on ASME Code Class 1, 2 and 3 components and their supports using
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sampling schedules described in section XI of the ASME Code and t
10CFR50.55a(b). Sample size and weld selection have been implemented in 1
j accordance with the Code and appear to be correct.
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2.2.3 Exclusion Criteria The criteria used to exclude components from examination shall be consistent with Paragraphs !WS 1220, !WC-1220 !WD 1220, and l
l 10CFR50.55a(b). The exclusion criteria have been applied by the t
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Licensee in accordance with the Code as discussed in Section 2.1, ' Program Description,' of the Pilgria Nuclear Power Station Unit 1, Second 10-Year Interval ISI Program Plan, Revision 3, and appear to be correct.
2.2.4 Auemented Examinatien Comitments 1
The Licensee has stated in the October 2,1g87 and April 15,1g88 submittals that augmented examinations will be implemented during the second 10-year inspection interval in accordance with the following documents:
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Regulatory Guide 1.150, ' Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations,' Revision 1 (Reference 13):
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NUREG 061g, 'BWR Feedwater Nozzle and Control Rod Drive Return Line l
NozzleCracking,'(Reference 14):
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IE Bulletin 80-13, ' Cracking in Core Spray Spargers.'
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NUREG-0803, "Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping,' (Reference 14):
(e)
Generic Letter 88 01, 'NRC Position on ISSCC in BWR Austenttic Stainless Steel Piping' (Reference 17)
(NUREG-0313. Revision 2,
' Technical Report on Material Selc: tion and Processing Guidelines for SWR Coolant Pressure Boundary Piping * (Reference 18), describes the technical bases for the NRC positions on ISSCC problems.):
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Generic Letter 87 05, ' Assessment of Licensee Measures to Nitigate and/or Identify Potential Degradation of Mark 1 Drywells' (Reforence 1g): and (g)
NUREG 0800, Section 3.6.1, ' Plant Design for Protection Against Postulated Piping Failures in Fluid Systems outside containment.'
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The Licensee reports that these augmented examinations are being scheduled and implemented independently from the ASME Code Section XI required examinations, and that they are not conducted as part of the Second 10-Year Interval 15! Program Plan at Pilgrim Station.
2.3 Conclusions Based on the review of the documents listed above, it is concluded that the Pilgrim Nuclear Power Station Unit 1, Second 10-Year Intr'* val Inservice Inspection Program Plan, Revision 3 is acceptable and in compliance with 10CFR50.55a(g)(4).
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EVALUATION OF RELIEF REQUESTS The requests for relief from the ASME Code requirements which the Licensee has determined to be impractical for the second 10 year inspection interval are evaluated in the following sections. Relief requests which have been i
resubmitted in Revision 3 of the Program Plan without revision are i
documented in previous Safety Evaluation Reports. Therefore, for those f
relief requests, this report will only identify the report in which the evaluation is contained and the NRC staff conclusion.
3.1 Class 1 Consonants
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3.1.1 Reactor Pressure Vessel 1
i 3.1.1.1 Reauest for Relief PRR-4. Revision 1. hamination Cateeory B-A.
j 11331 Bl.11 and Bl.12. Reactor Pressure Vessel Beltline Reaion MilA1 l
E Relief Request PRR-4 has been previously granted by the
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NRC in the Safety Evaluation Report (SER) dated August 13, 1985 l
provided that the Licensees (a) examines the accessible weld i
i areas, and (b) performs a visual examination of the vessel and i
i shield annulus area durinr, systes pressure tests.
In Revision 1
j 1 of PRA 4, the Licensee cosnitted to the above conditions.
j Therefore, the relief request evaluation as reported in the SER I
j should remain unchanged.
3.1.1.2 Reeuest for Relief PRR-5. Revision 1. hamination Cateoory B.A.
Items Bl.21 and Bl.22. Reactor Pressure Vessel Botton Head j
UilA1 E: Relief Request PRR-5 has not been revised. Therefore.
I the relief request evaluation, as reported in the SER dated l
August 13, 1985, should remain unchanged and relief be granted as requested, c
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3.1.1.3 Reauest for Relief PRR-9. Revision 1. hamination Cateeory B-D.
Items B3.90 and B3.100. Reactor Pressure Vessel Nozzle-to-Vessel Weld and Inside Radius Section E
The revised Request for Relief PRR-9 (Revision 1), as received in the Licensee's December 12, 1946 submittal of Revision 3 of the Program Plan, has been evaluated by the NRC in a Supplemental SER issued Narch 26, 1947. The subject report granted relief provided thatt (a)theexaminationsare performed to the maximum extent possibles and (b) the Code required system pressure tests are performed in accordance l
with IWB 5000.
In Revision 1 of PRA 9, the Licensee cosmitted to the above conditions. Therefore, the relief request evaluation as reported in the SER should remain unchanged.
e 3.1.2 Pressurizar (Does not apply to IWRs) l 3.1.3 Heat behanaars (No relief requests) i 3.1.4 Pinino Pressure Boundary s
3.1.4.1 Reauest for Relief PRR-1. Revision 3. haaination Cateeory B-J.
Items 89.11 and B9.21. and hamination Cateoory B-K-1. Item B10.10. Class 1 Circumferential Pressure Retaintne Pinine Welds
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and Inteerally Welded Attachments Within Flued Head Penetrations
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l M: This evaluation supercedes the evaluation in the SER j
dated August 13, 1945.
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l Code Recuirement: Section XI. Table IWS-2500-1. Examination j
Category B J. Iten 89.11 requires a 100% surface and volumetric I
examination of circumferential pressure retaining piping welds
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f 4 inch and greater nominal pipe size.
Itse 89.21 requires a I
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100% surface examination on circumferential pressure retaining piping welds less than 4 inch nominal pipe size. These examinations are to be performed as defined by Figure IWS-2500 8.
Section XI, Table IW8-2500-1 Examination Category B K-1, Item 810.10 requires a 100% volumetric or surface examination, as applicable, on Class 1 piping integrally welded attachments as defined by Figures IW8-2500-13, 14, and 15.
Licensee's Code Relief Recuest: Relief is requested from performing the Code-required examination (s) on the inaccessible welds within the following flued head penetrations:
System Line Size Penetration RHR 20' X-12 18' X-51A, X-51B 4'
X-17 Core Spray 10' X-16A X-168 RCIC 3'
X-53 RWCU 6'
X-14 58LC 1.5' X-42 Feedwater 18' X-9A, X-98 Main Steam 20' X-7A, X-7B X-7C, X-70 3'
X-8 HPCI 10' X-52 Licensee's Pronosed Alternative Examination: The first accessible pipe weld outside the subject penetrations will be velumetrically examined each interval, except for the 1.5 inch SSLC line which will receive a surface examination. The examinations required by IWB 5000 will be conducted on the alternative weld in accordance with the code, i
A YT 3 examination of the subject penetrations will be conducted each interval, to the extent practical.
Licensee's Basis for Recuestino Relief As stated in 10 CFR 50.55a(g)(1) for plants whose construction permits were 10
issued prior to January 1, 1971, components shall meet Section XI requirements to the extent practical. Since examination requirements for these wlds did not exist at the time Pilgrim Unit 1, was designed, accessibility for their inspection was not censidered. The design constraints make it extremely impractical to examine the subject wlds by volumetric or surface techniques. Boston Edison Company feels that this constitutes a basis for relief from the examination requirements of Section XI.
Evaluation:
In Revision 3 of Relief Request PRR-1, one additional penetration has been added (penetration X-8. 3 inch line in Main Steam system) and, in addition to the inaccessible circumferential pressure retaining wlds, the relief request has been expanded to include inaccessible integrally wlded attachments within the flued head penetrations. The subject welds are completely inaccessible for volumetric and/or surface examination because the welds are located within the containment nenetration. These welds can only be examined by inspecting for evidence of leakage during the systes hydrostatic tests.
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Conclusions:==
Based on the evaluation of Revision 3 to PRR-1 l
and the previous staff evaluation in the SER, dated l
August 13, 1985, which granted relief for PRR-1, it is l
concluded that the proposed alternative examination, along with l
the Code-required pressure test, ensures an acceptable level of inservice structural integrity and that compliance with the specific requirements of Section XI would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety. Therefore, it is recosseended that relief continue to be granted as requested.
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O 3.1.4.2 Recuest for Relief PRR-6. Examination Cateaory B-J. Items 89.10 and 89.40. Pressure Retainina Welds in Class 1 Pinina Systems H211: The request for relief from performing surface examinations on all pressure retaining welds in Class 1 piping systems was denied by the NRC in the SER dated August 13, 1985. Therefore, this request for relief was withdrawn by the Licensee in the submittal dated December 12, 1986.
3.1.5 Pumo Pressure Boundary 3.1.5.1 Reauest for Relief PRR 2. Revision 0. Examination Cateaory B-L-2. Item B12.20. Recirculation Pume Casinos H211: Relief Request PRR-2 has not been revised. Therefore, the relief request evaluation, as reported in the SER dated August 13, 1985, should remain unchanged with relief granted as requested.
3.1.6 Valve Pressure Boundary 3.1.6.1 Recuest for Relief PRR-3. Revision 1. Examination Cateaory B-M-2. Iten B12.40. Class 1 Valve Rodiel HQII: This relief request was submitted to obtain relief from the requirement to examine Class 1 valve body internals except when the valves are disassembled for maintenance. For the first ten year interval, 56 Class 1 valves were divided into 20 groups according to manufacturer, manufacturing method, constructional design and function. This grouping was provided to the NRC and subsequently used in the earlier versions of the Second 10 Year Interval 151 Program Plan.
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Revision 3 of the Program Plan incorporated the later Code (80W80) which eliminates the requirement to group the valves by manufacturer. The valves included in the original 20 categories have been regrouped and reduced to seven categories according to the later code requirements. The original list of l
56 valves has been reduced to 48 with the deletion of eight f
valves removed during Refueling Outage #6. The Licensee also reports that 36 of the subject valves were inspected during the refueling outage, including valves from all seven groups.
c As Revision 1 of PRA 3 only changes the grouping of valves as j
outlinedinthelaterCodeEditionandAddenda(80W80),the relief request evaluation, as reported in the SER dated August 13 1985, should remain unchanged and relief should be granted as requested.
3.1.7 General (No relief requests) 3.2 Class 2 Comoonents 3.2.1 Pressure Vessels _
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3.2.1.1 Reauest for Relief PRA 8. Revision 1. Examination Cateeory C-8.
l Ites C2.21. RHR Heat Exchaneer Nozzle to-Vessel Welds and Inside Radius Sections 1
lEII: The revised Request for Relief PRA-8 (Revision 1), as received in the Licensee's December 12, 1984 submittal of Revision 3 of the Program Plan, has been evaluated by the NRC in a Supplemental SER issued March 26, 1987. The subject report granted relief provided that (a)theproposed alternative surface examination is performed on the reinforcing ring (saddle)weldsthatmakethenozzle-tovesselwelds f
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0 inaccessible and (b) visual examination of the saddle welds for leakage is performed during periodic system pressure and hydrostatic tests in accordance with Subsection IWC-5000 requirements.
In Revision 1 of PRR-8, the Licensee cosuitted to the above conditions. Therefore, the relief request evaluation as reported in the SER should remain unchanged.
3.2.2 Pinine 3.2.2.1 Ragest for Relief PRR 6. Examination Cateoory C-F. Items C5.10 and C5.32. Pressure Retainino Welds in Class 2 Pinino Systems ELII: Relief from performing surface examinations on all pressure retaining welds in Class 2 piping systems was denied by the NRC in the SER dated August 13, 1985. Therefore, this request for relief has been withdrawn by the Licensee in the j
submittal dated December 12, 1986.
3.2.2.2 Recuest for Relief PRR-7. Revision 1. Examination Cateoory C-F.
Items C5.11 and C5.12. Pressure Retainino Welds in the -
Containment Atmosnheric Control 5vsten MLII: Relief Request PRA 7 has not been revised. Therefore, the relief request evaluation as reported in the SER dated August 13, 1945, should remain unchanged with relief granted as requested.
3.2.3 EWER 1 (No relief requests) 3.2.4 Valves (No relief requests) 3.2.5 General (No relief requests) 14
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i 3.3 Class 3 conoonents (No relief requests) i j
3.4 Pressure Tests 3.4.1 Class 1 System Pressure Tests (Noreliefrequests) 3.4.2 Class 2 System Pressure Tests 3.4.2.1 Reauent for Relief PRR 12. Revision 0. Examination Catenary C.H. Hydrostatic Test of the control Rod Drive i
Hydraulic Systes ELII: Relief Request PRR 12 requested relief from the l
l Code-required hydrostatic test for Class 2 Control Rod Drive (CAD) piping from the hydraulic control units (HCUs) to the l
l Reactor Pressure Vessel due to design configuration and l
impracticality.
In the basis for relief, the Licensee stated
[
that: ' Portions of the lines cannot be isolated for Class 2 i
hydrostatic testing due to design configuration.
Isolation of the remaining piping at the HCus is impractical due to the j
large number of valves to be realigned.'
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In the NRC request for additional information, dated
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July 14, 1987, it was pointed out that the staff does not
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consider the determination of impracticality justified based j
solely on valve realignment. Therefore. Relief Request PRR 12 was withdrawn in the Licensee's response dated October 2. 1947.
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I 3.4.2.2 Reauest for Relief PRR 13. Revision 0, hamination Cateeory C-H. Hydrostatic Test of the Class 2 Portions of the_
Ggatainment Atmosaheric Control Tyltgg j
Code Recuirement: Section XI. Table IWC 2500-1. Examination
{
Category C H requires a hydrostatic test of all Class 2 i
pressure retaining components each 10-year interval as outlined by !WC-5222.
i Licensee's Code Relief Reauest: Relief is requested fres performing the Code-required hydrostatic test on the Class 2 portions of the Containment Atmospheric Control System.
Licensee's Pronosed Alternative hamination: The non-isolable portions of the Containment Atmospheric Control System will be tested for integrity during the Appendix J. Type A, integrated leak rate test once each period. Ths isolable portions of the sample lines will be tested during the performance of Appendix J. Type C, local leak rate tests once each period.
Lig.gnsee's Basis for Reeuestine Relief: The Licensee reports that the purge and vent lines are open to the primary containment atmosphere and are unable to be isolated for hydrostatic or pneumatic testing. The sample lines can be isolated outside of containment but would require that extensive supports be added for hydrostatic testing.
Evaluation: The Licensee's submittal has been reviewed and it has been determined that footnote (1) of Table IWC-25001 Examination Category C H. excludes the open ended portion of the purge and vent lines from the Code required hydrostatic test.
It is also determined that, as the Containment Atmospheric Control System is designed for operation with air, the Code-required hydrostatic testing requirement for the sample lines is impractical.
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Conclusions:==
Based on the above evaluation, it is concluded that: (a) Relief is not required for the open ended portions of the purge and vent lines: and (b) For the sample lines, the Code hydrostatic testing requirements are impractical and compliance with the specific requirements of Section XI would result in hardship without a compensating increase in the level of quality and safety. Therefore, it is recommended that relief be granted as requested.
3.4.2.3 Recuest for Relief PRR 15. Revision 0. Examination, Cateoory C-H. Hydrostatic Test of the Hioh Pressure Coolant injection Turbine Exhaust Orain Line code Recuirementi Section XI. Table IWC-2500-1. Examination Category C.H. requires a hydrostatic test of all Class 2 pressure retaining components each 10-year interval as outlined by IWC-5222.
Licensee's Code Relief Recuest: lased on impracticality, L
relief is requested from performing the Code-required hydrostatic test on the Class 2 High Pressure Coolant Injection Turbine Exhaust Drain System bounded by M0-3, CV-52, 2301 112, 2301-33, 2301-33A, and the second flange on the turbine exhaust line.
l Licensee's Pronosed Alternative Examination: The Licensee proposes that, within each inspection period, a VT-2 examination be performed on the components bounded by M0-3, CV-52, 2301 112, 2301-33, 2301-33A, and the second flange on the turbine exhaust line. This YT-2 examination will be conducted during a system functional test as required by the Code.
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Licensee's Basis for Raouestine Relief: The Class 2 High pressure Coolant Injection Turbine Exhaust Drain Systen l
collects condensate from the turbine exhaust line, turbine casing, turbine steam rings, body drains on the stop valve, and l
l the steam chest drain. The1 inch (reducedto3/4-inch)
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turbine casing drain line is non-isolable between the turbine j
and the turbine exhaust drain pot. This line prevents isolation of the following piping off the exhaust drain pots (1) The piping free the turbine exhaust nozzle flange to the i
first down stress flanges (2) The steam ring drain liness (3)
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The drain lines from H0 2 and the steam chest drain line f
downstream of H0-1:
(4) The 3/4 inch drain line from the i
l exhaust drain pot to 2301 112:
(5) The 1 inch line from the
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exhaust drain pot to 2301 1315: and (6)The2inchand1 inch f
I lines from the exhaust drain pot bounded by 2301-131A, 2301 33, i
2301-33A and CV-32.
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Evaluati2D: The Licensee's submittal has been reviewed l
including the referenced diagram. The portion of Class 2
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piping as outlined above cannot be isolated, therefore, for l
this portion of piping, the system hydrostatic testing f
l requirements for Class 2 piping are impractical. The proposed l
i VT-2 visual examination during a system functional test is j
j acceptable.
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Conclusions:==
Based on the above evaluation, it is concluded that, for the subject portions of piping, the Code required hydrostatic test is impractical and that the proposed VT 2 I
examination will ensure an acceptable level of inservice structural integrity. Compliance with the specific requirements of section XI would result in hardship without a compensating increase in the level of quality and safety.
4 Therefore, it is recossended that relief be granted as l
requested.
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1 3.4.3 Class 3 System Pressure Tests i
I 3.4.3.1 Raeuant for Relief PRR-10. Revision 0. Examination Cateeory D-R. Hydrostatic Test of Two 10 Liter Shielded C-la f
I Chambers I
Code Reeuirement: Section XI. Table IWO-2500 1. Examination
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j Category 0-8. requires a hydrostatic test of Class 3 pressure f
retaining components in support of Resit'ual Heat Removal Systes each 10-year interval as outlined by IWD 5223.
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I Licensee's Code Relief Reauest: Relief is requested from
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l performing the Code required hydrostatic test on two 10 Itter f
I shielded sample chambers which are a part of the Reactor
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BuildingClosedCoolingWater(RBCCW) System.
I Licensee's Prenesed Alternative tramination: The Licensee j
proposes that, within each inspection period, a VT-2 exam will 4
j be performed on the two 10 liter shielded sample chambers.
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This YT 2 examination will be conducted during the l
Code required inservice leakage test.
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Licensee's Basis for Reeuestine Relief: The Code required hydrostatic test pressure for the RSCCW system is 165 psig j
which exceeds the design pressure of 125 psig for the two I
10-11ter shielded sample chambers.
I Evaluatieg: As a result of the NRC request for addittenal I
information dated July 14, 1947, the Licensee revisse Relief f
Request PRA 10 in the October 30. 1947 submittal. This
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l revision requests relief for the two 10 liter shielded sample I
chambers only. Each of the two sample chambers carries a sidestream of Reactor Building cooling water past an immersed f
l radiation detector. The chamber holds ten liters of cooling
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water to increase the sensitivity of the detector, and a shield I
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t surrounds the chamber / detector to reduce background radiation.
i The design pressure of the chamber is 125 psig, which is lower l
than the system design pressure of 150 psig but 25% higher than the maximum operating pressure.
In addition, the ASCCW system is protected from loss of chamber integrity by a 1/4 inch l
restricting orifice on the high pressure side. The Licensee reports that there is no allowable hydrostatic test pressure in the design documentation, since the chambers are treated as 4
instruments.
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Conclusions:==
Based on the above evaluation it is concluded that, for the two sample chambers listed, failure during l
operation would not affect the RSCCW system function, the Code required hydrostatic test is impractical, and that the proposed VT 2 examination will ensure an acceptable level of
)i inservice structural integrity. Compliance with the specific requirements of Section XI would result in hardship without a f
compensating increase in the level of quality and safety, f
1 Therefore, it is reconnended that relief be granted as j
j requested.
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3.4.3.2 Reeuest for Relief ptR-11. Revision 0, hamination l
Catenary B-3. Hydrostatic Test of the talt Service Water 1rstem f
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Code Requirement: Section XI Table IWO 2500-1. Examination Category 0 8 requires a hydrostatic test of Class 3 pressure l
retaining components in support of Emergency Core Cooling.
Containment Heat Removal Atmosphere Cleanup, and Reactor l
1 Residual Heat Removal systems each 10 year interval as outlined j
by IWO 5223.
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Licensee's code Relief Recuest: Relief is requested from the j
j requirements to hydrostatically test the Salt Service Water l
System pumps up to the expansion jo*
M the pump discharge I
lines on the basis of impracticall' t
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Licensee's pronosed Alternative Examination: The Licensee proposes to perform a VT 2 examination of the pump discharge i
piping up to the expansion joints during the Code-required systes inservice leakage test.
i Licensee's Basis for Recuestine Relief 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 systes includes five pumps whose pump casings are l
located under water. The hydrostatic test of the pumps and the
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discharge line would require disassembly and removal of the l
pumps. The requirements to remove the pumps for the sole j
purpose of performing a test of the pressure boundary has only I
a very small potential of increasing plant safety margins and a r
disproportionate impact on expenditures of plant manpower.
J Evaluation: The disassembly and removal of the pumps for the
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l sole purpose of performing the Code required hydrostatic test j
is a major effert and, in addition to the possibility of f
additional wear or damage to the pumps, could result in
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personnel receiving large amounts of radiation exposure.
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Conclusions:==
Based on the above evaluation, it is concluded j
I that, for the submerged Salt Service Water System pumps, the Code required hydrostatic test is impractical and that compliance with the specific requirements of Section XI would
[
result in hardship without a compensating increase in the level of quality and safety. Therefore, it is recossended that a
i relief be granted as requested.
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t 3.4.4 General, 3.4.4.1 Rtugst for Relief PRR 14. Revision 0. Eyamination Cateeories C H. 0-A. 0 B. and 0 C. Hydrostatic Test of Class 2 and 3 Systems Containine Relief Valves and Instrimantation jglH Relief Request PRR 14. S submitted, was considered generic in nature as it did not provide info m tion regarding the specific systees or components involved. Therefore, as a result of the NRC request for additional information dated July 14, 1987, the Licensee withdrew Relief Request PRA 14 in the submittal dated October 2, 1987.
3.5 General (No relief requests) 22
4.
CONCLUSION Pursuant to 10 CFR 50.55a(g)(6), it has been v.etermined that certain 5cctinn XI required inservice examinations 4re impractical.
In these cases, the Licensee has demonstrated that either the proposed alternatives would provide an acceptable level of qua', tty and safety or that compliance with the requirements would result in rardships or unusual difficulties without a compensating increase in the Inol of quality and safety.
This technical eval.dation has not identified any practical method by which the existing Pilgria Nuclear Power Station Unit 1, can meet all the specific triservice inspection requirements of Section XI of the ASNE Code.
Requiring c w pliance with all the exact Section XI required inspections would requira 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. The reactor pressure vessel and a number ot' the piping and component support systems are examples of components that would require redesign to meet the specific inservice examination provisions. Even after the redesign efforts, complete compliance with the Section II 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. Pursuantto10CFR50.55a(g)(6),
relief is allowed from these requirements which are impractical to implement.
The development of new or improved examination techniques will continue to be monitored. As improvements in these areas are achieved, the NRC may l
require that these techniques be incorporated in the next inspection i
interval !$1 program plan examination requirements.
Based on the review of the Pilgrim Nuclear Power : Station Unit 1. Second 10 Year Interval !$1 Program Plan, Revision 3, the Licensee's response to the NRC's Request for Additional Information, and the recossendations for 23 l
o o
the granting of relief from the 1.i! s imt.t on requirements that have been determined to be impractical, is om.'a.ui that the Pilgrim Nuclear Power Station, Unit 1, Second 10-Year Interval Inservice Inspection Program Plan, Revision 3, is acceptable and in compliance with 10 CFR 50.55a(g)(4).
24
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5.
REFERENCES
- 1. Code of Federal Regulations, Volume 10. Part 50.
.]
- 2. American Society of Mechanical Engineers Soller and Pressure Vessel i
Code,Section XI. Division 1, 1980 Edition through Winter 1940 Addenda.
- 3. Pilgria Nuclear Power station, Unit 1, second 10 Year Interval Inservice Inspection Program Plan, Revision 3, submitted December 12, 1944.
- 4. Letter, dated March 2, 1988, R.G. Bird (Boston Edison company (BEce)) to MC, ' Amendment {!5! 87-02) to Second Ten Year Interval Inservice Inspection Program,'
I
- 5. Safety Evaluation Report Related to Requests for Relief free Inservice l
Inspection Requirements for the second 10-Year Inspection Program at Pilgrim Nuclear Power Station Unit 1, dated August 13, 1945.
i l
- 6. Supplemental Safety Evaluation Report Related to Requests for Relief j
from Inservice Inspection Requirements for the second 10 Year Inspection Program at Pilgria Nuclear Power Station, Unit 1. dated March 28, 1947.
l l
- 7. NURE8 0400, Standard Review Plans, Section 3.6.1, "Plant Design for l
Protection Against Postulated Piping Failures in Fluid Systems Outside Containment,' Section 5.2.4, ' Reactor Coolant Soundary Inservice I
Inspection and Testing ' and Section 6.6, ' Inservice Inspection of Class I and 3 Components,' July 1981.
- 8. Letter, dated July 14, 1876 R.H.Wessmen(NRC) tor.S. Bird (BEco),
l
' Request for Additional Information on Second 10 Year Interval Inservice Inspection Program Plan for Pilgria Nucidar Power Station, Unit 1.*
i l
- 9. Letter, dated October 2, 1987, R.8. Bird (BEco) to NRC, ' Response to NRC l
Request for Additional Information Second 10 Year Interval Inservice l
Inspection Program Plan.'
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9
- 10. Letter, dated October 30, 1947 R.G. Bird (SECo)toNRC,' Design Allowances for seal Coolers and Sample Chambers."
- 11. Letter, dated December 24, 1987. R.8. Bird (BECo) to NRC, ' Itemized component Listing - Second 10 Year Interval Inservice Inspection Program Plan.'
- 12. Letter, dated April 15, 1948 R.S. Bird (Btco) to NRC, ' Augmented Inspections.'
- 13. Regulatory tutde 1.150, ' Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations,' Revision 1 dated February 1983.
- 14. NURES 0819, 'BM Feedwater Nozzle and Control kod Drive Return Line Nozzle Cracking,' dated November 1980.
- 15. It Bulletin 80 13, ' Cracking in Core Spray Spargers.' issued December 15, 1980.
- 16. NURte 0803, '8eneric Safety tynluation Report Regarding Integrity of BWR Scram System Piping,' dated August 1981.
- 17. Generic Letter 88 01, 'NRC Position on ISSCC in BWR Austenttic stainless steel Piping,' dated January 25, 1988.
- 18. NURte 0313, ' Technical Report on Material selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping,' Revision 2 dated January 1988.
- 10. Generic Letter 87 05, ' Assessment of Licensee Nessures to Nitigate and/or Identify Potential Degradation of Mark 1 Drpells' dated March 12, 1947.
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Technical Evaluation Report on the Second 10 Year Interval Inservice Inspection Program Plan:
Boston Edison Company, Pilgrim Nuclear Power
. s.'...:..::.. a o Station. Unit 1. Docket Num.ber 50 293 e '-
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May 1988
........:.....n B.W. Browr., J.D. Mudlin l
May 1988
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EG1G Idaho, Inc.
P. O. Box 1625 Idaho Falls, 10 83415 2209 FIN 06022 (Project 5)
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Materials Engineering Branch Technical Office of Nuclear Reactor Regulation
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U.S. Nuclear Regulatory Comission Washington 0.C.
20555 5...........
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This report presents the results of the evaluation of the Pilgrim Nuclear Power Station, Unit 1. Second 10 Year Interval Inservice Inspection (ISI) Program Plan.
Revision 3, submitted December 12, 1986, including Amendment ISI 87 02 to the 15!
Program Plan, submitted March 2, 1988.
The December 12, 1986 submittal included new and revised requests for relief from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Yessel Code Section XI requirements which the Licensee has determined to be impractical. Revision 3 of the Program Plan reflects the current plant configuration including the recirculation pipe replacement made during the 1983 1984 outage. The Pilgrim Nuclear Power Station Unit 1, Second 10-Year Interval ISI Program Plan, Revision 3. is evaluated in Section 2 of this report. The IS!
Program Plan is evalua6ed for (a) compliance with the appropriate edition / addenda of Section XI, (b) acceptability of the examination sample. (c) exclusion criteria, and (d) compliance with ISI related comitments identified during the Nuclear Regulatory Comission (NRC) review of previous submittals by the Licensee.
The new and revised requests for relief from the ASME Code requirements which the Licensee has determined to be impractical for the second 10 year inspection interval are evaluated in Section 3 of this report.
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