ML20198S194
| ML20198S194 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 11/30/1998 |
| From: | Mary Anderson, Charles Brown, Galbraith S IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | NRC (Affiliation Not Assigned) |
| Shared Package | |
| ML20198S163 | List: |
| References | |
| INEEL-EXT-98-01, INEEL-EXT-98-01007, INEEL-EXT-98-1, INEEL-EXT-98-1007, NUDOCS 9901110203 | |
| Download: ML20198S194 (36) | |
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INEEL/ EXT-98-01007 8.
Technical Evaluation Report on the Second 10-Year Interval inservice inspection Program Plan:
Southern Nuclear Operating Company, Vogtle Electric Generating Plant, Units 1 and 2, Docket Numbers 50-424 and 50-425 1
M. T. Anderson, C. Brown, S. G. Galbraith, A. M. Porter '
l Published November 1998 Idaho National Engineering and Environmental Laboratory Materials Physics Department Lockheed Martin Idaho Technologies Company Idaho Falls, Idaho 83415 1
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Prepared for the Division of Engineering Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 JCN J2229 (Task Order A28) l l
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d ABSTRACT This report presents the results of the evaluation of the Vogtle Electric Generating Plant Seconci Ten-Year Intervalinservice Inspection Program submitted May 29,1997, including the requests for relief from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, requirements that the licensee has determined to be e impractical. The Vogtle Electric Generating Plant Second Ten-YearIntervalinservice Inspection Program is evaluated in Section 2 of this report. The inservice inspection (ISI) 4 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 commitrnents identified during previous Nuclear Regulatory Commission reviews. The requests for relief are evaluated in Section 3 of this report.
l This work was funded under:
U.S. Nuclear Regulatory Commission JCN J2229, Task Order TWA-A28 Technical Assistance in Support of the NRC Inservice inspection Program ii
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SUMMARY
l The licensee, Southern Nuclear Operating Company, prepared the Vogtle E/ectric Generating Plant Second Ten-Year IntervalInservice Inspection Program to meet the requirements of the 1989 Edition of the American Society of Mechanical Engineers (ASME)
Boiler and Pressuro Vessel Code,Section XI. The second 10-year interval began May 31, s 1997, and will end on May 30,2007.
The information in the Vogtle Electric Generating Plant Second Ten-Year Interval Inservice Inspection Program submitted May 29,1997, was reviewed. The review included requests for relief from the ASME Code Section XI requirements that the licensee ,
has determined to be impractical. As a result of this review, a request for additional information (RAl) was prepared describing the information and/or clarification required from the licensee in order to complete the review. The licensee provided the requested information in a submittal dated June 26,1998.
Based on the review of the program plan, the licensee's response to the Nuclear Regulatory Commission's RAl, and the recommendations for granting relief from the ISI l examinations that cannot be performed to the extent required by Section XI of the ASME !
Code, no deviations from regulatory requirements or commitments were identified in the Vogtle Electric Generating Plant Second Ten-Year IntervalInservice Inspection Program.
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' CONTENTS
SUMMARY
._..................................................... iii j 1. I NTR O D U CTI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 ,
De 4
. 2.L EVALUATION 'OF INSERVICE INSPECTION PROGRAM PLAN . . . . . . . . . . . . . . . . . 3
- 2.14 Documents Evaluated . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
. 2.2 Compliance with Code Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 -
4 . -2.2.1 Compliance with Applicable Code Editions .................... 3 2.2.21 Acceptability of the Examination Sample ..................... 5
. 2.2.3 ; Exemption Criteria . . . . . . . . . . . - . . . . . . . . . . . . . . . _ . . . . . . . . . . . 5 -
. 2.2.4 - Augmented Examination Commitments . . . . . . . . . . . . . . . . . . . . . . . 5 x 2. 3 C oncl u si on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 3.' EVALUATION OF RELIEF REQUESTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3.1 Cla ss 1 Com ponents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
- 3.1.1 Reactor Pressure Vessel : . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ._ . . 8 3.1.1.1 . Request for Relief No. RR-2, Examination Categories B-D'and B-F, items B3.90, B3.100, and B5.10, Reactor Pressure Vessel Nozzle-to-Shell, Nozzle inner Radius Sections, and -
Nozzle-to-Safe End Examinations . . . . . . . . . . . . . . . . . . . . . 8
'3.1'1.2 Request for Relief No. RR-3, Examination Category B-A, items-B1.11, B1.12, and 81.21, Reactor Pressure Vessel (RPV) -
Weld s . . . . . . . . . . . . . . . . . . . . . . . . . .' . . . . . . . . . . . . . . - 1 0 -
3.1.1.3: Request for Relief No. RR-4 (Part 1) Examination Category B-A, item B1.40, Reactor _ Pressure Vessel RPV Closure Head We l d s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 0 3.1.1.4 - Request for Relief No. RR-_4 (Part 2) Examination Category B- ,
K, item B10.10, Integrally Welded Attachments to Pressure i Ve s se l s ' . . . . . . . . . . . . . . . . . . . . . . _ . , . . . . . . . . . . . . . . 1 1 3.1.1.5 Request for Relief No RR-31, Examination Category B-A, item B1.30, Reactor Pressure Vessel (RPV) Shell-to-Flange Weld . 13 3.1.1.6 Request for Relief No. RR 5, Examination Category B-G-1, item B6.10, Reactor Pressure Vessel (RPV) Closure Head N ut s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 6 3.1. 2 Pre ssurizer . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 W 3.1.2.1 Request for Relief No. RR-7, Examination Categories B-B, and B-D, items 82.11, B3.110, and B3.120, Pressurizer Circumferential Shell-to-Head Welds, Nozzle-to-Shell Welds, and Nozzle inner Radius Sections . . . . . . . . . . . . . . . . . . . . 16 g 3.1.2.2 Request for Relief No. RR-8 Examination Category B-K, item
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B10.10, Integrally-Welded Attachments to Pressure Vessels . 18
- 3.1.3 Heat Exchangers and Steam Generators . . . . . . . . . . . . . . . . . . . . . 19 3.1.3.1 Request for Relief No. RR-6, Examination Categories B-B and
,- B-D, items 82.40 and B3.140, Steam Generator Tubesheet-to-Shell Welds and Nozzle inside Radius Sections . . . . . . . . . .- 19 13.1.4. Piping Pressure Boundary ...............................22 iv m -
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3.1.4.1 Request for Relief No. RR-9, Ultrasonic Techniques and Calibration Blocks for Examination of Cast Stainless Steel (Grade SA 351-CF8A) for Reactor Coolant System (RCS)
Piping Welds ..................................22 3.1.4.2 Request for Relief No. RR-10, Examination Category B-F, item B5.130, Dissimilar Metal Piping Welds . . . . . . . . . . . . . 25 3.1.4.3 Request for Relief No. RR 11, Examination Category B-J, ,
item B9.31, Class 1 Branch Connection Welds .......... 27 3.1.4.4 Request for Relief No. RR 12 (Part 1), Examination Category B-J, Item B9.11, Class 1 Circumferential Welds . . . . . . . . . . 29 3.1.4.5 Request for Relief No. RR-13, Examination Category B-J, Class 1 Pressure Retaining Wolds ................... 32 3.1.5 Pump Pressure Boundary . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 3.1.6 Valve Pressure Boundary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2 3 .1. 7 G e n e r a l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2
- 3. 2 Cla ss 2 Com pone nts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 3.2.1 Pressure Vessels .....................................32 3.2.1.1 Request for Relief No. RR-14, Examination Categories C-A, and C B, items C1.20, C2.21, C2.22, and C2.31, Class 2 Vessel Welds, Nozzle-to-Vessel Welds and Nozzle Inside Radius (IR) Sections ....................... ..... 32 3 . 2 . 2 Pi pi n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ....... 38 3.2.2.1 Request for Relief No. RR-12 (Part 2), Examination Category C-F-1, item C5.11, Class 2 Circumferential Welds . . . . . . . . 38 3.2.2.2 Request for Relief No. RR-15, Examination Category C-F-1, Class 2 Pressure Retaining Welds ................... 40 3.2.2.3 Request for Relief No. RR-16, Examination Category C-F-2, item C5.51, Class 2 Carbon Steel Piping Welds ......... 40 3.2.2.4 Request for Relief No. RR-17, Examination Category C-F-1, item C5.11, Class 2 Piping Welds ...................41 3.2.2.5 Request for Relief No.18, Examination Category C-F-1, Items C5.11, C5.30 and C5.41 Class 2 Austenitic Stainless Steel Piping Welds in the Nuclear Service Cooling Water (NSCW) System ...............................42 3.2.2.6 Request for Relief No.19, Use of Code Case N-524, Altemative Examination Requirements for I.ongitudinal Welds in Class 1 and 2 Piping, for Examination of Class 2 Piping Welds.......................................44
- 3. 2. 3 P u m p s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 4
- 3. 2. 4 Va lve s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 4 -.
- 3. 2. 5 G e ne ral . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 44 3.3 Cla ss 3 Com ponents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 44 3.4 Pr e s su re Te sts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 44 l 3.4.1 Class 1 System Pressure Tests . . . . . . . . . . . . . . . . . . . . . . . . . . . 44 3.4.2 Class 2 System Pressure Tests . . . . . . . . . . . . . . . . . . . . . . . . . . . 45 3.4.3 Class 3 System Pressure Tests . . . . . . . . . . . . . . . . . . . . . . . . . . . 45 i
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3.4. 4 G ene ral . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 5 3.4.4.1 Request for Relief No. RR 23, Use of Code Case N-498-1, Alternative Rules for 10-Year System Hydrostatic Testing for Class 1, 2, and 3 Systems . . . . . . . . . . . . . . . . . . . . . . 4 5 3.4.4.2 Request for Relief No. RR-24, Use of Code Case N-416-1, Altemative Pressure Test Requirements for Welded Repairs s or Initialization of Replacement items by Welding, Class 1 2, and 3 Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 5 3.4.4.3 Request for Relief RR-26, IWA-5242(a), Visual Examination of
_ insulated Components . . . . . . . . . . . . . . . . . . . . . . . . . . . . 45
- 3. 5 G e n e r a l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 4 3.5.1 Ultrasonic Examination Techniques ........................54 3.5.2 Exempted Components . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54 3.5.2.1 Request for Relief No. RR-22, Use of Code Case N-544,
. Repair and Replacement of Smallitems ...............54
- 3. 5. 3 O t he r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 4 3.5.3.1 Request for Relief No. RR-1, Subsection lWA-2413, Proposed Alternative to Successive Interval Scheduling .... 54 3.5.3.2 Request for Relief No. RR 20, Use of Code Case N 509, Alternative Rules for the Selection and Examination of Class 1, 2, and 3 Integrally Welded Attachments . . . . . . . . . . . . . 57 3.5.3.3 Request for Relief No. RR-25 (Revision 9/17/97), IWA-5250(a)(2), Corrective Actions for Bolted Connections . . . . . 57 3.5.3.4 Request for Relief No. RR-28, Use of Code Case N-508-1, Rotation of Serviced Snubbers and Pressure Relief Valves for the Purpose of Testing ........................58 3.5.3.6 Request for Relief No. RR-30, Use of Code Case N-532, Alternative Requirements to Repair and Replacement and Inservit e Summary Report Preparation and Submission as Required by IWA-4000 and IWA-6000 . . . . . . . . . . . . . . . . 58
- 4. C O N C LU S I O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 61
- 5. R E F E R E N C E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 3 L
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TECHNICAL EVALUATION REPORT ON THE i SECOND 10-YEAR INTERVAL 1
INSERVICE INSPECTION PROGRAM PLAN:
SOUTHERN NUCLEAR OPERATING COMPANY, VOGTLE ELECTRIC GENERATING PLANT, 1
.. UNITS 1 AND 2, 1
DOCKET NUMBERS 50-424 AND 50-425 '
l
- 1. INTRODUCTION 1
Throughout the service life of a water-cooled nuclear power facility,its components (including supports) that are classified as American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Class 1,2, and 3 are required by
- 10 CFR 50.55a(g)(4) (Reference 1) to meet the requirements, except the design and access provisions and the preservice examination requirements, of the ASME Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, ,
(Reference 2) tJ the extent practica! within the limitations of design, geometry, and i materials of construction of the components. This section of the regulations also requires that inservice examinations of components and system pressure tests conducted during successive 120-month inspection intervals comply with the requirements in the latest
. . edition and addenda of the Code incorporated by reference in 10 CFR 50.55a(b) on the date 12 months prior to the start of the 120-month inspection interval, subject to the limitations and modifications listed therein. The components (including supports) may meet requirements set forth in subsequent editions and addenda of this Code that are incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein, and subjec' to Nuclear Regulatory Commission (NRC) approval. The licensee, Southern Nuclear Operating Company (SNC) has prepared the Vogtle Electric Generating Plant Second Ten-Year Intervalinservice Inspection Program (Reference 3) to meet the requirements of the 1989 Edition of the ASME Code,Section XI. The second 10-year interval began May 31,1997, and will end on May 30,2007.
., Pursuant to 10 CFR 50.55a(a)(3), proposed alternatives to the Code requirements may be used when authorized by the NRC. The licensee must demonstrate either that the proposed alternatives provide an acceptable level of quality and safety, or that Code
- r. compliance would result in hardship or unusual difficulty without a compensating increase in safety. Pursuant to 10 CFR 50.55a(g)(5)(iii),if the licensee determines that conformance with certain Code examination requirements is impractical for its facility, the licensee shall submit information to the NRC to support that determination. Pursuant to 10 CFR 50.55a(g)(6)(i), the NRC will evaluate the licensee's determination that Code requirements are impractical. The NRC may grant relief and may impose alternative requirements that it determines to be authorized by law, will not endanger life, property, or the common defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
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The intormation in the Vogtle Electric Generating Plant Second Ten-YearInterval Inservice !nspection Program, submitted May 29,1997, was reviewed, including the requests for relief from the ASME Code Section XI requirements that the licensee has determined to be impractical. This review was performed using the standard review plans l
of NUREG 0800, Section 5.2.4, " Reactor Coolant Boundary Inservice Inspections and Testing," and Section 6.6, " Inservice inspection of Class 2 and 3 Components" (Reference 4), s i The NRC requested additionalinformation that was necessary to complete the review of the inservice inspection (ISI) program plan. The requested information was provided by the licensee in a letter dated June 26,1998, (Reference 5). Two additional requests for relief were submitted in a letter dated August 14,1998 (Reference 6).
The Vogtle Electric Generating Plant Second Ten-Year IntervalInservice Inspection Program is evaluated in Section 2 of this report. The ISI program plan is evaluated for (a) compliance with the appropriate editionladdenda of Section XI, (b) acceptability of examination sample, (c) correctness of the application of system or component examination exclusion criteria, and (d) compliance with ISI-related commitments identified during the NRC's previous reviews. The requests for relief are evaluated in Section 3 of this report. Unless otherwise stated, references to the Code refer to the ASME Code,Section XI,1989 Edition. Ir. service test programs for snubbers and for pumps and valves t are being evaluated in other reports.
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- 2. EVALUATION OF INSERVICE INSPECTION PROGRAM PLAN This evaluation consists of a review of the applicable program documents to determine whether or not they are in compliance with the Code requirements and any previous i license conditions pertinent to ISI activities. This section describes the submittals '
reviewed and the results of the review.
t 2.1 Documents Evaluated Review has been completed on the following information from the licensee:
- Vogtle Electric Generating Plant Second Ten-YearIntervalinservice Inspection Program, dated May 29,1997 (Reference 3).
- Licensee's response to NRC RAI dated June 26,1998 (Reference 6).
- Letter, dated August 14,1998, containing two additional relief requests (Reference 6).
2.2 Compliance with Code Requirements 2.2.1 Compilance with Applicable Code Editions inservice inspection program plans are to be based on Section XI of the ASME Code editions defined in 10 CFR 50.55a(g)(4) and 10 CFR 50.55a(b). The second interval for Vogtle Electric Generating Plant, Unit 1 began May 31,1997, therefore, the Code applicable to the Unit 1 second interval ISI program is the 1989 Edition. As addressed in Request for Relief No. RR-1 (Section 3.5.3.1), the licensee requested to update the Program for Unit 2 concurrently with Unit 1. Under the proposed change, the VEGP-2 second 10-Year interval will begin two years ahead of schedule such that the two VEGP units will be under the same edition of the Code. As stated in Section 1 of this report, the licensee has prepared the Vogtle Electric Generating Plant Second Ten-Year Interval Inservice Inspection Program to meet the requirements of the 1989 Edition of the Code.
In accordance with 10 CFR 50.55a(c)(3),10 CFR 50.55a(d)(2), and 10 CFR 50.55a(e)(2), ASME Code cases may be used as alternatives to Code
., requirements. Code Cases that the NRC has approved for use are listed in Regulatory Guide 1.147, inservice Inspection Code Case Acceptability, (Reference 7) with any additional conditions the NRC may have imposed. When used, these Code Cases must be !
". implemented in their entirety. The licensee may adopt an approved Code Case by providing written notification to the NRC. Published Code Cases awaiting approval and '
subsequent listing in Regulatory Guide 1.147 may be adopted only if the licensee requests, l
and the NRC authorizes, their use on a case-by-case basis, i
The~ licensee's second 10-year ISI program includes the Code Cases listed below.
These Code Cases either have been approved for use in Regulatory Guide 1.147 or have been included as requests for relief. !
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Code Case N-408 2 Altemative Rules for Examination of Class 2 Piping (RR-21 evaluated and authorized in NRC SER dated March 24,1998)
' Code Case N-4161 Alternative Pressure Test Requirement for Welded Repairs or Installation of Replacement items by Welding, Class 1, 2, and 3 (RR-24 evaluated and authorized in NRC SER dated March 24,1998) e Code Case N-491 Alternative Rules for the Examination of Class 1, 2, and 3 MC Components and Supports of Light Water Cooled Power Plants ..
Code Case N-498 Alternative Rules for 10-Year System Hydrostatic Testing for Class 1, and 2 Systems Code Case N-498-1 Alternative Rules for 10-Year System Hydrostatic Testing for Class 1, 2, and 3 Systems (RR-23 evaluated and authorized in NRC SER dated March 24,1998)
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l Code Case N-5061 Rotating of Serviced Snubbers and Pressure Relief Valves for t
the Purpose of testing (RR-28 evaluated and authorized in NRC SER dated December 9,1997)
Code Case N-509 Alternative Rules for the Selection and Examination of Class 1, 2, and 3 Integrally Welded Attachments (RR 20 evaluated and authorized in NRC SER dated March 24,1998)
Code Case N-521 Altemative Rules for Deferral ofInspections of Nozzle-to-Vessel Welds, inside Radius Sections, and Nozzle-to-Safe End Welds of a Pressurized Water Reactor (PWR) Vessel (Evaluated in Section 3.1.1.1 of this document)
Code Case N-524 Alternative Examination Requirements for Longitudinal Welds in Class 1 and 2 Piping (RR-19 evaluated and authorized in NRC SER dated March 24,1998)
Code Case N-532 Altemative Requirements to Repair and Replacement Documentation Requirements and Inservice Summary Report Preparation and Submission as Required by IWA-4000 and IWA-6000 (Evaluated in Section 3.5.3.6 of this document)
Altemative Requirements for VT-2 VisualExamination of Class :
Code Case N 533 1 Insulated Pressure-Retaining Bolted Connections (Evaluated in Saction 3.4.4.3 of this report) ,
Code' Case N-544 Repair / Replacement of Smallitems (RR-22, Rev.1, evaluated and authorized in NRC SER dated March 24,1998)
Code Case N-546 Altemative Requirements for Qualification of VT-2 Examination Personnel (Evaluated in Section 3.4.4.5 of this document)
~ Code Case N-566 Corrective Action for Leakage identified at Bolted Connections (RR-25 evaluated and authorized in NRC SER dated October 24,1997)
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2.2.2 Acceptability of the Examination Sample Inservice volumetric, surface, and visual examinations shall be performed on ASME Code Class 1,2, and 3 components and their supports using sampling schedules described i in Section XI of the ASME Codo and 10 CFR 50.55a(b). Sample size and weld selection
, procedures have been implemented in accordance with the Code and 10 CFR 50.55a(b) and appear to be correct with the following exceptions:
The volumetric examination of reactor pressure vessel (RPV) closure head studs, required by Examination Category B-G-1, item B6.20, " Closure Studs, in place", have not ;
been included in the Program. In addition, all RPV studs are scheduled to be examined during the third period. Under Examination Category B-G-1, Note 1, the examinations may be performed in-place or with the studs removed. Considering the latitude provided by the Code, these volumetric examinations should not be deferred until the end of the interval; and a sample of the studs should be examined each period.
It appears that the licensee is not meeting the percentage requirements for some items contained in the itemized listings (i.e., C5.21, C5.51). It also appears that many components are not scheduled in the tables. Components requiring examination in :
accordance with ASME Section XI should be identified and scheduled in the Plant Program documentation. These and other Class 1 and 2 components should be reviewed and Code sampling and scheduling requirements verified.
2.2.3 Exemption Criteria The criteria used to exempt components from examination shall be consistent with Paragraphs lWB-1220, IWC-1220, IWC-1230, IWD-1220, and 10 CFR 50.55a(b). The exemption criteria have been applied by the licensee in accordance with the Code, as discussed in the ISI program plan, and appear to be correct with the following exception:
It appears that the licensee may be exempting Class 2 components with cumulative inlet / outlets exceeding 4-inch NPS. As allowed by Code Case N-408-2, which was evaluated and authorized in Request for Relief RR-21 in an NRC SER dated, March 24, 1998, vessels, pumps and valves, and their connections in piping NPS 4 and smaller, are exempt from the examination requirements of the Code. /n piping is defined as having a
_ cumulative inlet and a cumulative outlet pipe cross-sectional area neither of which exceeds the nominal OD cross-sectional area of the designated size. However, it appears that several vessels (i.e, Regenerative Heat Exchanger and Letdown Reheat Heat Exchanger) may exceed the nominal cross-sectional area of the designated size (4-inch NPS). The criteria used at VEGP to exempt Class 2 vessels should be reviewed 2.2.4 Augmented Examination Commitments l
l In addition to the requirements specified in Section XI of the ASME Code, the licensee has committed to perform the following augmented examinations:
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h (a) Volumetric examination of piping welds per NRC Branch Technical Position MEd 3-1 and VEGP Improved Technical Specification 5.5.16; (b) Examination of reactor coolant pump flywheels in accordance with NRC Regulatory
~ Guide 1.14, Reactor Coo / ant Pump f/ywheelIntegrity (Reference 8), and VEGP TS5.5.7*; and o
(c) Ultrasonic examination of the RPV in accordance with NRC Regulatory Guide 1.150, Ultrasonic Testing of Reactor Vesse/ Welds During Preservice andInservice '
Examinations (Reference 9);
- By letter dated May 8,1998, the licensee requested to use WCAP-14535A in lieu of RG ,
1.14. As stated by the license, evaluation is pending.
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2.3 Conclusion Based on the review of the documents listed in Section 2.1, no deviations from regubtory requirements or commitments were identified in the Vogtle E/ectric Generating Plant Second Ten Yearintervalinservice Inspection Program except 1) for the scheduling of Examination G G 1, RPV bolting,2) percentages of certain items that do not appear to
[* meet the minimum sangle sizes of the Code, and 3) the potentially incorrect exemption of certain Class 2 vessels. Note ths: this report does not include a review of the implementation of the augmented exa.ninations,it merely records that the licensee has committed to perform them.-
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- 3. EVALUATION OF RELIEF REQUESTS The requests for relief from the ASME Code requirements that the licensee has determined to be impractical for the second 10-year inspection interval are evaluated in the following sections.
3.1 Class 1 Components 3.1.1 Rwactor Pressure Vessel ,,
3.1.1.1 Request for Relief No. RR 2, Examination Categories B-D and B-F, items B3.90, B3.100, and B5.10, Reactor Pressure Vessel Nozzle-to-Shell, Nozzle Inner Hadius Sections, and Nozzle-to-Safe End Examinations Code Requirement-Section XI, Table IWB-25001, Examination Category B-D, items B3.90 and B3.100, and Examination Category B-F, item B5.10, require a 100% volumetric examination of all reactor vessel nozzle-to-shell welds, nozzle inner radius sections, and nozzle-to-safe end butt welds each inspection interval as defined by Figures IWB-2500-7 and IWB 2500-8. At least 25% but not more than 50% (credited) of the nozzles shall be examined by the end of the first inspection period and the remainder by the end of the inspection interval. RPV nozzle safe end welds may be performed coincident with the vessel nozzle examinations.
Licensee's Proposed Alternative-4n accordance with 10 CFR 50.55a(a)(3)(i), the licensee proposed to implement Code Case N-521, Alternative Rules for DeferralofInspections of Nozzle-to-Vessel Welds, Inside Radius Sections, and Nozzle-to-Safe End Welds of a Pressurized Water Reactor Vessel. The licensee stated:
"ASME Section XI Code Case N 521 accepts the alternative scheduling of the subject examinations provided that certain conditions are met. These conditions include the following: (1) No inservice repairs or replacements have been performed to the RPV, (2) no existing flaws requiring successive inspections exist, and (3) the unit is not in the first inspection interval.
"The required types of examinations will continue to be performed except that they will be performed at the end of the ten-year interval."
Licensee's Basis for Proposed Alternative-
" Historically, pressurized water reactors such as VEGP-1 and 2 examine the RPV outlet nozzle to shell welds, their inside (inner) radius sections, and associated nozzle to safe .-
and welds during the first inspection period in order to comply with the requirements of the ASME Section XI Code. These examinations are performed from the inside surfaces of the RPV using submerged ultrasonic examination techniques with an automated reactor vesselinspection tool. These examinations are generally performed whila defueled with water in the refueling canal, thus typically being a critical path activity. The similar reactor vesselinlet nozzles are accessible only with the RPV core barrel removed, as during the " Ten-Year Inservice inspection" at which time they are examined. In order to consolidate the examination of the reactor vesselinlet and outlet nozzles such that they are examined at one time, SNC proposes the use of ASME 8
Section XI Code Case N-521 which allows rescheduling of all such RPV examinations l to the end of the inspection interval. The rescheduling allows SNC significant opportunities for savings in contractor cost, critical path, radiation exposure, and internal manpower requirements while still maintaining compliance with the examination requirements of the Code.
, "In order to facilitate possible future use of Code Case N-521, each of the subject RPV l areas for which relief is being requested were examined during VEGP-1 Outage 1R6 in Spring 1996 during which the " Ten-Year Inservice Inspection" was performed. This
!- included re-examining those areas, e.g., the outlet nozzles and their associated components, which were examined during the first inspection period in the First Ten-Year Interval. This was done voluntarily by the licensee so that no more than ten years would elapse before being examined again at the end of the Second Ten-Year Interval.
In the case of VEGP-2, a similar course of action is planned for Outage 2R6, currently scheduled for March 1998, at which time the remaining RPV examinations for its first Ten-Year interval will be completed. Each of the outlet nozzles and their associated components will be re-examined on VEGP-2 during its ' Ten-Year Inservice Inspection' so that all of the RPV examinations may be performed at the end of the Second Ten-Year Interval. No rejectable indications have been observed n >r have any repairs been made to either reactor pressure vessel at VEGP-1 and 2 which would preclude the use of Code Case N-521.
" Based on the foregoing, SNC requests that this request for relief be authorized pursuant to 10 CFR 50.55a(a)(3)(i) permitting the use of ASME Section XI Code Case N-521 at VEGP-1 and 2."
Evaluation-The Code requires the examination of at least 25%, but not i 3 e than 50% of RPV nozzles and associated inside radius (IR) sections and nozzle safe t,.T s oing the first inspection period. The licensee has requested to use Code Case N-521 and r fer the examination of these areas until the end of the third 10-year interval.
Code Case N-521 states that the examination of RPV nozzles, IR sections and nozzle-to safe end welds may be deferred provided (a) no inservice repairs or replacements by welding have ever been performed on any of the subject areas, (b) none of the subject areas contain identified flaws or relevant conditions that currently requirc successive inspections in accordance with IWB-2420(b), and (c) the unit is not in the first interval.
The licensee has confirmed that these conditions have been met. in addition, the licensee
, examined all the subject areas during the final refueling outage (the third period) of the first 10-year interval for Unit 1 and plans to do the same for Unit 2. By performing the nozzle examinations and associated IR sections and nozzle-to-safe end welds at the end of the l'- previous 10-year interval, the licensee has established a new sequence of examinations
- and will not exceed 10 years between examinations. By meeting the conditions in the l Code Case and by repeating the examinations at the end of the previous interval, the
! licensee's proposed alternative will provide an acceptable level of quality and safety since the maximum time of 10 years between inspections will not be exceeded.
Conclusion-Considering that the licensee has met all the conditions stated in the Code Case and has examined all of the effected areas at the end of the previous interval, a new sequence of examinations has been established. Furthermore, since the time between 1
. . _ . . _ _ . _ . _ . _ __ _-.-_______.-_.______._m ._
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examinations will not exceed 10-years, the licensee's proposed alternative will provide an :
[ . acceptable level'of quality and safety. Therefore, it is recommended that the licensee's l
proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i). The use of Code
!- Case N-521 should be authorized for the second 10-year interval at VEGP, or until the
- l. ~ Code Case is approved for general use by reference in Regulatory Guide 1.147. After that l time, the licensee may continue to use the Code Case with the limitations, if any, listed in l Regulatory Guide 1.147. ,
l 3.1.1.2 Request for Relief No. RR-3, Examination Category B-A, items B1'.11, B1.12, and B1.21, Reactor Pressure Vessel (RPV) Welds Note: Request for Relief RR-3 was withdrawn by the licensee in_the June 26,1998, ,
i submittal as the result of an NRC RAI question.
3.1.1.3 Request for Relief No. RR4 (Part 1) Examination Category B-A, item B1.40, Reactor Prsssure Vessel RPV Closure Head Welds l
Code Requirement--Examination Category B-A, item B1.40, requires 100% volumetric and surface examination of RPV head-to-flange welds, as defined by Figure IWB-2500-5. ,
I: Licensee's Code Re#ef Request--in accordance with 10 CFR 50.55a(g)(5)(iii), the licensee t I requested relief from performing volumetric examination to the extent required by the Code l l' for the RPV closure head-to-flange weld . listed in the table below. ;
- y 4 Q LX@'i'V' '^5,'
- 'A:}l:',Talke RR4dettN:'i;? l' 'T D % ;:A?
-,n ; ,. ,nno ~ ~ ~ : ~ ~ .. ~ r l 'e(Jnttr /;), p'f1 M eklIN X % m 9 MJ;lH Smitationi,1h'74l N M Coverage'v ;
L 1_ 11201-V6-OO1-WO2 RPV closure head lifting lugs and 65 %
~2 21201-V6-OO1-WO2 !
Licensee's Besis for Requesting ReWef-I~ " Physical obstructions, e.g., RPV closure head lifting lugs and the RPV closure head ;
l torus flange configuration, limit or otherwise prohibit movement of the transducer used for the ultrasonic examination of the RPV closure head to flange < weld along the ,
required scan region thereby limiting the amount of examination coverage that can be [
p attained. Full Code coverage of ninety percent (90%) or greater (as defined by ASME ~
Section XI Code Case N-460 and as accepted by the NRC) is not possible for the !
referenced welds that are in the vicinity of the obstructioris. '
" Relief was granted by the NRC for the RPV closure head torus to flange welds and ,
. closure head lifting lugs during the First Ten-Year Interval. These included First Ten-
!- Year interval Requests for Relief RR-7 and RR-52 for VEGP-I and 2. NRC approval was documented in correspondence dated November 26,1991 and December 17,1991 for ,
VEGP I and 2, respectively.
"AILhough there are physical obstructions which limit the amount of examination coverage for those welds identified in Attachment 1, reasonable assurance still exists '
r that an acceptable level of quality and safety will be maintained since there have been c 10 l ,
no catastrophic failures of reactor pressure vessels. As a result, SNC requests that l l relief be authorized pursuant to 10 CFR 50.55a(g)(6)(i) since it is inspractical to perform l these examinations to the extent required by the Code".
l t Licensee's Proposed Altemative Examination-
"No alternate examination is proposed. The RPV closure head to flange weld will be
. volumetrically examined to the extent practical".
Eve /uetion-The Code requires 100% volumetric and surface exa nination of the subject L- RPV head-to-flange welds. However, complete examination is restricted by adjacent lifting
! lugs and the flange configuration, which make the volumetric exoination impractical to i- p'erform to the extent required by the Code. To meet the Code requirements, the RPV head would require design modifications to allow access for examination imposition of this requirement would create a burden on the licensee.
The licensee can examine a considerable portion (65%) of each of the subject welds.
L in addition, the surface examination can be completed to the extent required by the Code.
L The partial ultrasonic examination, in combination with the Code-required surface 1 examination should detect any significant patterns of degradation and provide reasonable assurance of the continued structuralintegrity of the RPV closure head-to-flange weld. !
l Conclusion-Based on the impracticality of meeting the Code coverage requirements for the subject welds, and the reasonable assurance provided by the examinations that were ,
completed,it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i). ;
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3.1.1.4 Request for Relief No. RR-4 (Part 2) Examination Category 8-K, item B10.10,
- Integrally Welded Attachments to Pressure Vessels Code Requ/remont-Code Case N-509*, Examination Category B-K, Item B10.10 requires i 100% surface examination of integrally welded attachments to pressure versels as defined by Figures IWB-2500-13, -14 and -15, as applicable.
Licensee's Code Re#e/ Request-in accordance with 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from performing the surface examination, to the extent required by the i Code, for the RPV integral attachment welds listed in the table below. -
l 0 e " r q'^ )) '
' /1 h ; ; 9TableARMlartMble ' ;l.c D '[ if W -
' UrNt: : 1/'l' Weld'iQ'l' $ '[~kdss( U5'dMWbOI'b 'k $$MlI 7 1 11201-V6-001-W204 RPV closure head lifting device and 50 % !
CRD braces I I' 11201-V6-001-W205 1 11201-V6-001-W206 1
)
- Use_ of Code Case N-509 was previously evaluated and authorized in an NRC SER dated ;
l- March 24,1998.
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l a sq gggggg y Ma JOnit? w$W5Id]6) %
<M {ddijMiddji < M {USkerassh 2 21201 -V6-001-W204 RPV closure head lifting device and 50 %
CRD braces 21201-V6-001-W205 e 21201-V6-001-W206 Licensee's Basis for Requesting Relief-
"The geometric configuration and location of the Control Rod Drive (CRD) braces and RPV closure head lifting device prevents a complete surface examination of the RPV closure head lifting lugs.
" Relief was granted by the NRC for the RPV closure head torus to flange welds and closure head lifting lugs during the First Ten-Year Interval. These included First Ten-Year interval Requests for Relief RR-7 and RR-52 for VEGP-1 ard 2. NRC approval was documented in correspondence dated November 26,199i and December 17, 1991 for VEGP-1 and 2, respectively.
"Although there are physical obstructions which limit the amount of examination coverage for those welds identified in Attachment 1, reasonable assurance still exists that an acceptable level of quality and safety will be maintained since there have been no catastrophic failures of reactor pressure vessels. As a result, SNC requests that relief be authorized pursuant to 10 CFR 50.55a(g)(6)(i) wnce it is impractical to perform these examinations to the extent required by t7e Code".
In the June 26,1998, submittal, the licenses stated:
"Section V, Article 7, Paragraph 741 of the 1989 Edition of the ASME Code states:
'Two separate examinations shall be performed on each area. During the second examination, the lines of magnetic flux shall be approximately perpendicular to those used during the first examination.' Approximately fifty percent (50%) of the total Code coverage is obtainable due to the Vessel Head Clevis, Braces and Unremovable insulation. The lifting device, vessel head clevis, braces and unremovable insulation are part of an integrated head package designed for rapid head removal. Disassembly ~
of this head package is not practical."
Licensee's Proposed Altemative Examination- ~
"No alternate examination is proposed. The RPV closure head to flange weld will be volumetrically examined to the extent practical. Likewise, the RPV closure head lifting lugs will be examined by surface means to the fullest extent practical."
Evaluation-The Code requires 100% surface examination of the subject integrally welded attachments. However, complete examination is restricted by an integrated head package, consisting of the vessel head clevis, braces and nonremovable insulation, which makes the surface examination impractical to perform to the fxtent required by the Code. To meet the Code requirements, the integrated head package would require design modifications to 12
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allow access for examination. Imposition of this requirement would create a burden on the licensee.
l The licensee can examine a considerable portion (50%) of each of the subject welds.
The partial surface examination should detect any significant patterns of degradation that ,
l may occur and provide reasonable assurance of the continued structuralintegrity of the l 4_ RPV closure head lifting lugs. )
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l Conclus/an -Based on the impracticality of meeting the Code coverage requirements for
- - the subject welds, and the reasonable' assurance provided by the examinations that were completed, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i).
3.1.1.5- Request for Relief No. RR-31, Examination Category B-A, item B1.30, Reactor Pressure Vessel (RPV) Shell-to-Flange Wald i
Code Requirement- Examination Category B-A, items 81.30 requires 100% volumetric l examination of the RPV shell-to-flange weld as defined in Figure IWB-2500-4. In accordance with Examination Category B-A, Note (4), at least 50% of the length of the RPV flange welds must be examined by the end of the first inspection period. If the partial examination is conducted from the flange face, the remaining volumetric examinations required to be conducted from the vessel wall may be performed at or near the end of the inspection interval.
L/censee's Proposed Altemative-In accordance with 10 CFR 50.55a(a)(3)(i), the licensee p;oposed an alternative to the examination requirements of the Code for the RPV shell-to-flange weld. The licensee stated:
"The volumetric examinations of the RPV shell-to-flange welds will L a,tinue to be performed at VEGP-1 and 2. However, the entire length of these welds will be scheduled and examined from the flange face at or near the end of the Second Ten-Year Interval, instead of 50% of the weld length being examined from the flange face in the first inspection period and the remainder in the third inspection period. The entire length of the RPV Shell-to-Flange welds will be scheduled and examined from the vessel at or near the end of the Second Ten-Year Interval. The proposed altemative will provide an acceptable level of quality and safety. The welds on VEGP-1 and 2 will still be volumetrically examined during the Second Ten-Year Interval except that the subject welds will have their entire length examined from both the flange face and
- o. vessel wall at or near the end of the interval rather than examining just a portion of the
~ welds from the flange face during the first inspection period. Approval of the proposed altemative will not result in more than ten years elapsing between the RPV Shell-to-
- Flange weld examinations conducted during the First Ten-Year interval and those scheduled for examination during the Second Ten-Year Interval. Therefore, it is requested that this alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i). Denial of this request for relief would eliminate potential opportunities for savings in contractor costs, radiation exposure, and interval manpower requirements."
Licensee's Basis for Requesting ReRef-
"To comply with Table IWB-2500-1 during the first ten-year interval, 50% of the RPV
-shell-to-flange weld length was scheduled and examined from the flange face during 13
the first inspection period. The remaining 50% of the weld length was scheduled and examined from the flar.3e face during the third inspection period as allowed by Footnote 4 of Table IWB-2 BOO-1. With the partial examinations being conducted from the flange face, the remaining volumetric examinations from the vessel wall, were scheduled and examined at or near the end of the first ten-year interval as allowed by Footnote 3 of Table IWB-2500-1.
6 "At the end of the first ten-year intervals during the 1R6 and 2R6 maintenance / refueling outages at VEGP-1 and 2, respectively, one hundred percent (100%) of the h? / shell-to-flange weld length was examined from the flange face, and 100% of the same weld length was examined from the vessel wall. These examinations were part of those performed for the first ten-year interval.
" Southern Nuclear Operating Company has concluded that rescheduling of these examinations such
- hat they are performed at or near the end of the inspection interval will have little, if any, effect on the quality of examinations while providing substantial benefit to SNC. This conclusion is based on the following:
- 1. Examination Quality - This request for relief involves rescheduling the first period examinations (50% of the RPV Shell-to-Flange weld length, from the flange face) such that the entire weld length will be examined at or near the end of the inspection interval. Fifty percent of the weld length examined from the flange face during the first inspection Period of the First Ten-Year Interval was re-examined during maintenance / refueling outages IR6 and 2R6. This was done voluntarily by SNC in order to 're-zero' the examination so that no more than ten years would elapse before being examined again at or near the end of the Second Ten-Year Interval contingent upon NRC approval of this request for relief. By 're-zeroing' the RPV Shell-to-Flange weld during the First Ten-Year Interval such that no more than ten years would elapse before being examined again at or near the end of the Second Ten-Year Interval, the proposed rescheduling of this examination should have no effect on the quality of the overall RPV Shell-to-Flange examinations.
- 2. Examination Continuity - The potential evaluation of flange face examinations presents difficulties due to limited transducer manipulation, long metal paths, and extensive beam spread limitations (Refer to Figure 1 which depicts the RPV Shell-to-Flange weld). As a result, accurate sizing techriology is not as practical for examinations conducted from the flange face when compared to those conducted from the vessel wall. By using proven automated sizing technology from she vessel c wall such as that used during the ' Ten-Year Inservice Inspection', more accurate sizing results are achievable. With approval of this request for relief, SNC would only use a RPV flange face examination tool and a mechanized RPV inspection tool once at or near the end of the inspection interval. In addition, approval of this '
request for relief would allow SNC to re-synchronize the examination schedulo, thus providing a readily available means of characterizing any indications observed from the flange face examination should they occur.
"In addition to the foregoing, SNC is anticipating NRC approval to re-schedule the RPV
' Examination Category B-D and B-F welds or appurtenances,i.e., Nozzle-to-Vessel, Inside Radius Sections, and Nozzle-to-Safe End welds, to the end of the Inspection l
l 14
i I
i interval, as submitted to the NRC in Request for Relief RR-2. In that request for relief, )
SNC requested permission from the NRC to use ASME Section XI Code Case N-521, !
' Alternative Rules for Deferral of Inspections of Nozzle-to-Vessel Welds, inside Radius Sections, and Nozzle-to-Safe End Welds of a Pressurized Water Reactor (PWR) Vessel'.
l Use of that Code Case would allow for the rescheduling of Examination Category B-D l and B-F welds or appurtenances to be examined at or near the end of the inspection I
. interval provided that certain conditions were met, e.g., no inservice repairs or replacements by welding, no identified flaws or relevant conditions that currently require successive inspections, and the unit is not in the first inspection interval.
Although a different Examination Category is involved, i.e., B-A, than those addressed in Code Case N-521, the VEGP-1 and 2 RPV Shell-to-Flange welds meet those same conditions in that examinations from both the flange face (100%) the vessel wall (100%) were performed at the end of their respective first inspection intervals. In addition, no inservice repairs or replacements by welding have occurred and no flaws or relevant, conditions that require successive inspections were identified for that particular weld on either VEGP unit.
" Finally, the proposed rescheduling allows SNC significant opportunities for savings in i contractor cost, critical path time, radiation exposure, and internal manpower ;
requirements while still maintaining compliance with the examination requirements of j the Code. Approval of this request for relief for the RPV Shell-to-Flange (B-A) welds, in i conjunction with approval of Request for Relief RR-2, would allow the B-A, B-D, and B-F welds or appurtenances to be examined at the same time, By performing the examinations at the same time, additional radiation exposure reduction can be realized, as well as reductions in mobilization and coordination efforts, set-up time, and ;
examination time, in addition, performance of the examinations at one time at or near ;
the end of the Second Ten-Year Interval constitutes a Cost Beneficial Licensing Action (CBLA)in that savings in excess of $100,000 are expocted to be realized over the l remaining lives of the two VEGP units due in part to savings from not having to )
mobilize the RPV inspection vendor in the first inspection period of the inspection intervals."
i Evaluation-The Code requires 100% volumetric examination of the RPV shell-to-flange )
weld each inspection interval. The weld must receive at least a partial examination from I the flange face by the end of the first inspection period. In lieu of the Code scheduling i requirements, the licensee has proposed complete deferral of the RPV shell to-flange weld l to the end of the 10-year interval.
This proposal parallels Code Case N-521 which allows the deferral of other RPV "alds and examination areas provided (a) no inservice repairs or replacements by welding +
ever been performed on any of the subject areas, (b) none of the subject areas contai.'
identified flaws or relovant conditions that currently require successive inspections in accordance with IWB-2420(b), and (c) the unit is not in the first interval. The licensee confirmed that these conditions have also been met for the shell-to-flange weld. In ,
addition, the licensee has re-examined the subject welds during the third period of the first 10-year interval. By repeating this examinations at the end of the previous 10-year interval, the licensee has established a new sequence of examinations and will not exceed 10 years between examinations. By meeting the conditions contained in Code Case N-521 and by repeating the examinations at the end of the previous interval, the licensee's 15
. - . . ..- . - -_ -. - ~_ - --. . .__ . . - .- _ - . ~ . . - -
a l
l proposed alternative will provide an acceptable level of quality and safety since the
. maximum time of 10 years between inspections will not be exceeded.
Conclusion-Considering that the licensee has met all the conditions discussed above and i
~ has examined the effected areas at the end of the previous interval, a new sequence of examinations has been established. Furthermore, since the time between examinations will not exceed.10-years, the licensee's proposed alternative will provide an acceptable ,
level of quality and safety. Theiafore,it is recommended that the licensee's proposed alternative be authorizea pursuant to 10 CFR 50.55a(a)(3)(i). .-
3.1.1.6 Request for Relief No. RR 5, Examination Category B-G-1, item B6.10, Reactor Pressure Vessel (RPV) Closure Head Nuts Note: Evaluated and authorized in an SER dated March 24,1998.
3.1.2 Pressurizer 3.1.2.1 Request for Relief No. RR-7, Examination Categories B B, and B-D, items B2.11, B3.110, and B3.120, Pressurizer Circumferential Shell to Head Welds, Nozzle-to-Shell Welds, and Nozzle Inner Radius Sections j
Code Requirement-Examination Category B-B, item B2.11 requiies 100% volumetric examination of pressurizer circumferentia! shell-to-head welds as defined by Figure IWB- ,
2500-1. Examination Category B-D, items B3.110 and B3.120 require 100% volumetric 1 examination of pressurizer nozzle to vessel welds and nozzle inside radius sections as defined by Figure IWB-2500-7.
Licensee's Code Re#ef #eguestHn accordance with 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from the Code coverage requirements for the pressurizer welds listed in the table below.
, a>
J + Q , " 2 ' 3 , ' ' Table RR-7
.'Jm'-
":' , ^ si ' J c ',, -
Urut/ttom , w , ' , Wold ID '@ v Desenption N '- Cc Limitation , , , ~ ' Coverage
!' 1/B2.11 11201 V6-002-WO1 Upper head-to-upper Nozzles, supports & 68%
shell weld ID plate
~
l 2/B2.11 21201-V6-00? wn1 linnar head to-uoper Nozzles, supports & 68% -
shell weld ID plate {
(
1/B2.11 11201-V6-002 WO5 Lower shell to lower Support skirt 84 % .-
head weld i 2/B2.11 21201 V6-002-WO5 Lower shell to lower Support skirt 84 %
head weld l 1/B3.110 11201-V6-002-W10 Upper head to 6" Nozzle configuration 50%
11201-V6-002-W11 safety nozzle weld 2/83.110 21201-VG-002 W10 Upper head to 6' Nozzle configuration 50 %
21201-V6-OO2-W11 safety nozzle wald 16 i.
n - - . - - - . . . - . - . - . - . . . . - _ . - -- _._ _ _
l'
+ MMgh@@M%g5l%RhWlgBG%%:$1eW$MQQjfQQGR MGy%nMi;W;a%M w
h$: 'I I f '
1/B3.110 11201-V6-002-W16 14" surge nozzle Nozzle config./ heater 15%
l penetrations l 1/B3.110 11201 V6-002-W12_ 6" and 4" safety Nozzle configuration 50 %
,, 11201-V6-002-W13 nozzles i 11201 V6-002-W14 2/B3.110 21201 V6-002-W12 6" and 4" safety Nozzle configuration 50 %
t 21201-V6-002 W13 nozzles 21201-V6-002 W14-2/83.110 21201-V6-002-W16 14* surge nozzle Nozzle config./ heater 15%
penetrations j 1/B3.120 11201-V6-002-lR-06 16" surge nozzle Heater penetrations 0% i 2/83.120 21201-V6-002-lR-06 16" surge nozzle Heater penetrations 0% l ll IUcensee's Basis for Requesting ReWef-
"(a) Geometric limitation of the pressurizer skirt obstructs and/or prevents transducer movement along the required scan region of the Category B-B lower shell to lower head weld. Full Code coverage is not possible in the vicinity of the
, obstructions.
l
"(b) _ Geometric configuration of the 16" Category B-D pressurizer surge nozzle prevents scanning from the nozzle side. In addition, the pressurizer lower head heater penetrations obstructs and/or prevents transducer movement along the required scan region from the shell side. Limited coverage can be obtained for the 16" Category B-D nozzle to shell weld. However, no coverage can be
- obtained for the 16" nozzle inner radius due to the transducer " set back" '
dimension required.
'l
"(c)- . Geometric configuration of the 4" and 6" relief and spray nozzles prevent
- scanning from the nozzle side for the Category B-D upper head to nozzle welds. l
"(d) Geometric limitations of the pressurizer supports, ID plates and instrumentation !
nozzles obstructs and/or prevents transducer movement along the required scan !
region of the' Category B-B upper head to upper shell weld. Full Code coverage is not possible in the vicinity of the obstructions, e "The examinations are being conducted to the extent practical.
i " Relief was initially granted by the NRC during the First Ten-Year interval for those
[ welds / components in Attachment 1 [ paraphrased abovel having requests for relief
' submitted for them. These included First Ten-Year Interval Requests for Relief RR-12, RR-14, and RR-15. NRC approval was documented in correspoXence dated j November 26,1991 and December 17,1991 for VEGP-1 and 2, respectively.
~
i "Although there are physical obstructions which limit or prohibit the amount of examination coverage for those welds / components identified in Attachment 1, l
h
reasonable assurance still exists that an acceptable level of quality and safety will be achieved. As a result, SNC requests that this request for relief be authorized pursuant tc 10 CFR 50.55a(g)(6)(i) since it is impractical to perform the examinations to the extent required by the Code."
Licensee's Proposed A/temative Examination-
"No alternate examinations will be performed for Category B-B welds. A 100% surface ,
examination will be performed on the 4" and 6" nozzle to shell welds, and approximately 85% on the 16" nozzle to shell (B-D) weld (heater penetration obstruction)." ,
l Evaluation-The Code requires 100% volumetric examination of the subject pressurizer
- nozzle-to-vessel welds and inside radius (IR) sections. However, complete volumetric examination is limited by physical obstructions, such as nozzle configuration, heater penetrations and adjacent supports, that obstruct access to the examination area.
Therefore, the Code coverage requirements are impractical for these welds. To meet the Code requirements, the pressurizer and associated obstructions would require design modifications to allow access for examination. Imposition of the Code requirements would result in a considerable burden on the licensee.
To supplement the limited volumetric examinations, the licensee proposed to perform surface examinations of the nozzle-to-vessel welds (Examination Category B-D welds).
The limited volumetric examination, in conjunction with the supplemcntal surface examinations should detect any significant patterns of degradation that may occur and provide reasonable assurance of the structuralintegrity of pressurizer nozzle-to-vessel welds.
The pressurizer surge nozzle IR sections are completely obstructed and no alternative examinetions were proposed. However, there are other pressurizer nozzle IR sections that can be examined that provide an indication of patterns of degradation that could affect the surge riozzles. The examination of the other pressurizer nozzles provides reasonable assurance of the structuralintegrity of the surge nozzles.
Conclusion--Based on the impracticality of meeting the Code coverage requirements for the subject examination areas, and the reasonable assurance provided by the examinations that can be complettd,it is recommended that relief be granted pursuant to 10 CFR
50.55a(g)(6)(i).
3.1.2.2 Request for Relief No. RR-8 Examination Category B-K, item B10.10, Integrally Welded Attachments to Pressure Vessels .,
Note: Request for Relief RR-8 was withdrawn by the licensee in the June 26,1998, submittal as the result of an NRC RAI question.
l 18
. . m_ _ . . _ . -. _ _ ._ _ . . _ . . . _ . _ . _ . - . _ . . _ _ _ _ _ _ . _
1 l
l 3.1.3 Heat Exchangers and Steam Generators 3.1.3.1 Request for Relief No. RR 6, Examination Categories B-B and B-D, items 82.40
- and B3.140, Steam Generator Tubesheet to-Shell Welds and Nozzle inside Radius Sections .
l Code Requiremont-Examination Category B-8, item B2.40 requires 100% volumetric examination of steam generator tubesheet-to-shell welds as defined by Figure IWB-2500-6. l p Examination Category B-D, item B3.140 requires 100% volumetric examination of steam generator nozzle inner radius sections as defined by Figure IWB-2500-7.
Licensee's Code Railef Request-in accordance with 10 CFR 50.55a(g)(5)(iii), the licensee i p requested relief from the Code coverage requirements for the steam generator examination l
- areas listed in Table RR 6 below.
!- 1 n' h(<' # ,d,Te M / ' a, , o
$$h :ssd <,'3 s ,4 in: e::uder+a4. 1 Unit # tem 7 MP; WelifID' e '
O'cc';"' #J'Unillation ':' , ' :"c* a Costa 0*m:>
1/82.40. 11201-B6-001 -WO8 Tube sheet configuration 80 % l 11201-B6-002-WO8 11201-B6-003-WO8 11201-B6-004-WO8 2/B2.40 21201-B6-001 WO8 Tube sheet configuration 80%
21201 B6-002-WO8 <
21201-B6-003-WO8 l 21201 86-004-WO8 )
1/B3.140 11201 86-001-lR-01 Nozzle Configuration 0%
1120' %001-in-tu 112C t d6 002-IR-01 11201-B6-002-IR-02 11201-B6-003-IR-01 i
11201-B6-003-IR-02 i
- 11201-B6-004-lR-01 I
- 11201-B6-004-IR-02 2/B3.140 21201-B6-001-IR-01 Nozzle Configuration 0%
! 21201-B6-001-lR-02
!. 21201-B6-002-lR-01 l*'
21201-86-002-lR-02 l 21201-B6-003-IR-01 l c, 21201-B6-003'IR-02 l 21201-B6-004-IR-01 21201 B6-004-IR-02 i
19 i
l -- . - - - , , . - , , - - , , .,
Licensee's Basis for Requesting Relief-
"(a) Physical limitation of the steam generator tube sheet obstructs and/or prohibits transducer movement along the required scan region of the channel head to tube sheet weld. Full Code coverage is not possible in the vicinity of the obstructions.
"(b) The steam generator primary side nozzles are integrally cast as part of the channel head. The steam generator nozzle radius sectino cannot be i volumetrically examined from outside of the nozzle or channel head because the rough, as-cast contact surface is not suitable for ultrasonic coupling, and the '
geometric configuration requires an excessively long test metal distance resulting in high ultrasonic attenuation. The inside of the nozzle and channel head areas are covered with cladding in the 'as-welded' condition; therefore, meaningful volumetric examination cannot be performed from the 'as-welded' surface. Even with proper preparation of the inside surface for volumetric examination, an adequate examination of the area of interest (base metal just below the cladding) could not be achieved due to the resulting ultrasonic response at the clad-to-base metal interface. Refer to Attachment 2' for a depiction of the nozzles at VEGP.
"The examinations of the steam generator tubesheet welds are being conducted to the extent practical.
" Compliance with the requirements of ASME Section XI for the specific nozzle inside radius section is impractical. Due to the high radiation field present (generally greater than 10 REM / hour), visual examinations are not practical and dose acquired is contrary to the principles of "As Low As Reasonably Achievable" (ALARA). During the First Ten-Year Interval, visual examinations and the Code-required pressure tests were performed and no evidence of degradation was observed.
" Relief was initially granted by the NRC during the First Ten-Year Interval for those we'ds/ components in Attachment 1 [ paraphrased above] having requests for relief submitted for them. These included First Ten-Year Interval Requests for Relief # RR-19 and RR-42. NRC approval was documented in correspondence dated November 26,1991 and December 17,1991 for VEGP-1 and 2, respectively."
In the June 26,1998, response to the NRC RAl, the licensee stated:
" Visual examinations will not be performed as an alternative examination to volumetric -
l requirements for the steam generator nozzle inner radii. Requirements for inspection of nozzle inner radius regions in Class 1 systems have been in effect for many years, and ,
according to a recent study performed by the ASME Section XI ISI Optimization Group, have not resulted in any inspection findings for vessels in which relief has been requested herein. This study also surveyed a total of twenty-five (25) utilities in an
- effort to gain perspective on the state of primary nozzle inner radius examinations. For steam generator primary nozzle inner radius examinations, the survey population
( included two hundred thirty (230) nozzles. From that population, one hundred forty-four (144) volumetric examinations have been performed, or are planned. The
- Attachment not included in this report.
l 20 l
l - - - - _ - _ - _ _ _
l service-induced flaws have been detected in all examinations performed.
" Risk Assessment
" Evaluations have shown the effects of in-service examinations on the risk of failure due to cracking in the nozzle inner radius. Since the applied stress intensity factor
. does not exceed the fracture toughness,it could be argued that leakage would occur l
from a through-wall flaw before any integrity problems would occur, at any of these nozzles.
"The question to ask is whether in-service inspection can change risk of failure by identifying in-service flaws. There are no mechanisms of damage other than fatigue for the nozzle corners. Therefore, the only scenarios of concern are for a flaw which was not found in the pre-service examination to grow during service, or for a flaw to !
iritiate during service and propagate.
"In conclusion, the evaluations have demonstrated that it is highly unlikely that the nozzles in questior.3 would fail under any anticipated service conditions, in-service, SNC inspections can hardly benefit plant safety for something that is very unlikely tc happen. The examinations are very difficult to perform because of access and high radiation environment in Tiany cases. Examinations which have been done have not led to the discovery or any indications."
i licensee's Proposed Alternative Examination-
"(a) No alternate examination is proposed. The volumetric examination of the referenced steam generator tube sheet welds are being conducted to the fullest extent practical.
"{b) No alternate examination is proposed for steam generator primary side nozzles inner raru."
Evaluation-The Code requires 100% volumetric examination of the subject steam i generator tubesheet-to-shell welds and nozzle inside radius (IR) sections. However, I complete volumetric examination is limited by component configuration which obstructs access to the examination area. Therefore, the Code coverage requirements are impractical for these areas. To meet the Code requirements, the steam generator would require design modifications to allow access for examination imposition of the Code
. requirements would result in a considerable burden on the licensee.
For the tubesheet-to-shell welds, a significant portion (80%) of the examinations can be completed. Therefore,it is concluded that any significant patterns of degradation that may occur should be detected, and reasonable assurance of the continued structural integrity provided. For the nozzle IR sections, the Code-required examination is completely inaccessible. However, the licensee has not proposed any examination in lieu of the Code-required volumetric examination. In the previous interval, a VT-1 visual examination was performed, but, as stated by the licensee, is not proposed for the current interval. The l licensee's basis for not performing the visual examination is an evaluation that considered the history of cracking and the low risk associated with steam generator primary nozzle IR sections. Although this may be reasonable grounds for a change in the Code requirements, l
21
requests for relief are not the proper mechanism for proposing Code changes. Therefore, it is concluded that relief be granted only if a VT-1 visual examination is performed on each of the subject steam generator nozzle IR sections.
Conclusion-Based on the impracticality of meeting the Code coverage requirements for the subject examination areas, and the reasonable assurance provided by the examinations that can be completed, it is recommended that relief be granted pursuant to 10 CFR ,
50.55a(g)(6)(i) for the tubesheet-to-shell welds. For the primary nozzle IR sections, it is concluded that the Code coverage requirements are impractical, but that a VT-1 visual -
examination should be performed as proposed during the first 10-year interval. Therefore, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i) provided the subject nozzle IR sections receive a VT-1 visual examination in lieu of the Code-required volumetric examination.
3.1.4 Piping Presst:re Boundary 3.1.4.1 Requen.t for Relief No. RR-9, Ultrasonic Techniques and Calibration Blocks for Examination of Cast Stainless Ste::! (Grade SA 351-CF8A) for Reactor Coolant System (RCS) Piping Wolds Code Requ/rement-Article ill-2430 of ASME Section XI requires that manual ultrasonic examination scanning shall be done at twice (+6dB) the primary reference level as a minimum. ASME Section XI Article I, Supplement 1 (b) requires the calibration block thickness to be of a size sufficient to contain the entire examination path. For weld thicknesses over 2 inches but less than 4 inches,Section V, Article 5; Figure T-542.2.1 requires calibration block thickness to be either 3 inches, or actual thickness.Section XI, Article 111, Paragraph lll-3410 requires that basic calibration blocks be made from material of the same nominal diameter and nominal wall thickness as the pipe to be examined.
- Licensee's Proposed Altemative-In accordance with 10 CFR 50.55a(a)(3)(i), the licensee proposed an alternative tc the Code requirements for the examination of Grade SA 351-CF8A, cast stainless steel RCS piping welds. Calibrations will be performed utilizing two calibration blocks, one with a 2.45-inch nominal wall thickness and one with a 3-inch nominal wall thickness. The licensee stated:
" Ultrasonic examination scanning will be performed at the primary reference level rather than twice the reference level for the centrifugally-cast stainless steel piping and static-cast elbows in the Reactor Coolant System. In addition, calibrations for e examining the subject piping and fittings will be performed using the method described above." .-
Licensee's Bssis for Proposed Attemative-
"(a) The cast SA 351, CF8A material contains a banded microstructure that consists of a duplex grain size ranging from extremely coarse to very fine. This irregular grain structure causes significant attenuation snd some angular variations during a typical shear-wave ultrasonic examination. Therefore, a 1.0 inch dual-clement focused ultrasonic transducer, utilizing a 45 degree refracted longitudinal wave with a frequency of 1.0 megahertz is used. During calibration, the primary 22 j
l l
l reference level is set using side-drilled holes, with the notch brought to the Distance Amplitude Correction (DAC) curve, as allowed by ASME Section XI i Article 111-3230. Scanning is possible only at the primary reference level due to l excessive noise associated with the higher gain levels and the metallurgical I structure of the material. A demonstration using this technique was performed for NRC Region 11 and was determined to be a conservative method of detecting
- , reflectors from the inside diameter (10). (Reference NRC report numbers 50-l 425/85-24 and 50-425/85-25).
6 "b) Calibration / examination with the two calibration blocks of 29.0" Diameter /
2.45" Thickness (Block No. 331 A) and 32.0" Diameter /3.00" Thickness (Block No. 329A), along with the use of ASME Code Case N-461, satisfy Code requirements for wall thickness. Also, ASME Section V, Paragraph T-542.2.1 allows a tolerance of 11". The applicable SNC cast stainless ultrasonic examination procedure, as well as procedures for ultrasonically examining piping, will address the requirement for verification of actual wall thickness for appropriate screen range determination. The curvature of the calibration blocks also meet the requirement of ASME Section XI Article I, Appendix 1, Supplement 1.
" Relief was granted by the NRC for using the primary reference level for ultrasonic scanning for the First Sn-Year Interval. These included First Ten-Year interval Requests for Relief RR-22 for VEGP-1 and 2. NRC approval was documented in correspondence dated November 26,1991 and December 17,1991 for VEGP-1 and 2, respectively.
" Relief was granted by the NRC for use of these calibration blocks during the First Ten-Year interval. These included First Ten-Year Interval Requests for Relief RR-23 for VEGP-1 and 2. NRC approval was documented in correspondence dated November 26, 1991 and December 17,1991 for VEGP-1 and 2, respectively. A subsequent revision of the original requests for relief were approved by the NRC as documented in correspondence dated March 8,1996 and August 13,1996 for VEGP-1 and 2, respectively.
"The examinations of the centrifugally-cast stainless steel piping and static-cast elbows in the Reactor Coolant System are being conducted to the fullest extent practical. As a result, SNC requests that relief be authorized pursuant to 10 CFR 50.55a(a)(3)(i) since
. th > proposed alternatives provide an acceptable level of quality and safety."
In the June 26,1998, submittal, the licenseo stated:
o "For cast stainl3ss steel examinations, Appendix Vill requirements (Supplement 9) are in the course of preparation. Past demonstrations performed by VEGP perscnnel to the NRC will not satisfy the Code if performed similarly with the requirements of Appendix Vlli Supplement 2 (wrought austenitic). Current Performance Demonstration Initiative
! (PDI) sensitivity levels for wrought austenitic are based on material noise, if the same l techniques are used for cast stainless steel examinations, relief due to the scanning sensitivity will not be required per Requests for Relief RR-9 or RR-12.
i 23
.- ~__._____m _ _ _ _ _ ~ . . - - -
l "When the Appendix Vill PDI Program was initiated, specific, unique plant designs were I
not to be included in generic specimen sets. Plants with these unique designs would l
need _to implement their own testing parameters either within the requirements of l Appendix Vill or, with NRC approval, an alternative to the requirements. The 10-inch SA-376 Class 1 Safety injection System piping at VEGP, indicated in Request for Relief l
l RR-12, falls in the category of requiring on alternative to the requirements.
( "For weld or material configurations that are not covered by the PDI generic sample set, SNC intends to address on a case-by-case basis. SNC will require the
[
nondestructive examination (NDE) personnel to be qualified to Appendix Vill, and will -
i have either a mock-up or calibration block to demonstrate the procedure. Once the l procedure has been demonstrated for the plant specific configuration, the NDE l personnel will demonstrate the procedure to the satisfaction of the SNC Lead Level lit.
I. A limited demonstration will take place, as the individual will have previously met the l requirement for identifying the number of flaws, etc. as required by Appendix Vill.
! Computer modeling may be used by SNC for the demonstration of specific applications, L
such as inner radius, nozzle to shell, or other areas where it is applicable."
l Evaluation-ASME Section XI requires ultrasonic examinations to be performed at two l' times reference sensitivity (+ 6 db).Section V, Article 5 requires calibration block thickness to be either 3 inches or actual thickness for welds greater than 2 inches but less l
than or equal to 4 inches thick. In lieu of these requirements, the licensee has proposed to l perform examinations at reference sensitivity and calibrate on blocks of 2.45-inch and 3- ,
inch nominal wall thickness. A demonstration using this technique was performed for NRC Region 11 and was determined to be a conservative method of detecting reflectors from the
- inside diameter (ID). Frcm the demonstration, the NRC inspectors determined that the l
calibration provided by the 2.45-inch thick block was as sensitive, if not more sensitive, l than the calibration of the thicker block. This alternative is further supported by Code i . Case N-461, which has been approved for general use by reference in Regulatory Guide !
1.147, inservice Code Case Acceptability, ASME Section XI, Division 1, and allows calibration blocks to be used that are within 25% of the actual thickness. Specifically, that would allow the use of blocks ranging from 2.25-inches thick to 3.75-inches thick.
Based on the successful demonstration that was documented in the October 18,1994, NRC letter, and the guidance provided in Code Case N-461, the INEEL staff concludes that the examinations performed using the thinner calibration block will provide an acceptable level of quality and safety. Regarding demonstrating the proposed alternative to scan at reference sensitivity, current revisions of ASME Section XI, Appendix Vill do not provide f requirements for demonstrating techniques and precedures for cast stainless steel materials. Therefore, the licensee will require inspection personnel to be qualified to ~
Appendix Vill and will demonstrate the procedure on a mock-up or calibration block.
Considering the lack of guidance provided by the Code and that the licensee will enhance
[- the examination with demonstrations "in-the-spirit" of Appendix Vlli, the proposed
- examination at reference sensitivity should provide an acceptable level of quality and safety.
Conclusion-Based on the evaluation above, it is concluded that the proposed alternative will provide an acceptable level of quality and safety. Therefore, it is recommended that the licensee's proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i) for 24
l the first 10-year ISI interval.
3.1.4.2 Request for Relief No. RR 10, Examination Category B-F, item B5.130, !
Dissimilar Metal Piping Welds Code Requ/rement-Examination Category B-F, item B5.130 requires 100% surface and
, volumetric examination, as defined by Figure IWB-2500-8, for Class 1 dissimilar metal piping welds greater than or equal to 4-inch nominal pipe size.
t-I F Ucensee's Code Re#cf Request--in accordance with 10 CFR 50.55a(g)(5)(iii), the licensee <
l requested relief from the Code coverage requirements for the dissimilar metal piping welds listed in the table below. I i
c'e ,
a ,: Py '^ , , i g g, , 4 ', , * /' ; ' s, L ^' 'Talite RR M $ ' d?' :' . ' ' 5 2 <s <
2 , '? '
' Unit k ' ' 's , Wald ID .', 3 ' M > c'. Deic'ription G li i b linitation,:/,4 207Coveragedi 1 11201-001-5 31" Elbow-to-SG Nozzle Nozzle 50 %
I 11201-002-5 configuration 11201-003-5 11201-004-6 11201-005-1
- 11201-006-1 I 11201-007-1 11201-008-1 2 21201-001-5 31" Elbow-to-SG Nozzle Nozzle 50 %
i 21201-002-5 configuration 21201-003-5 21201-004-6 21201-005-1 21201-006-1 21201-007-1 21201 008 L Ucensee's Basis for Requesting ReWef-
"The elbows are manufactured from centrifugally cast stainless steel with a coarse l grain structure which prevents the use of normal ultrasonic shear wave techniques (Reference RR-9). The configuration of these elbow to nozzle weld joints is such that s complete examinations are impractical to perform. The nozzle side of these welds has a sharp taper immediately adjacent to the weld, plus the surface of the nozzle is in a
- rough, as cast condition. These conditions prevent any meaningful ultrasonic examinations from b6ing performed on the nozzle; therefore, scanning of these welds l- can only be performed on the elbow side of the weld and on a portion of the weld l itself. With this configuration, examinations will be performed as follows
i l
l Reflectors Parallel to the Weld - For reflectors parallel to the weld, scanning will be performed only from the elbow side using a one half-node refracted longitudinal wave; therefore, only one beam direction will be attainable.
i
- Reflectors Transverse to the Weld - For reflectors transverse to the weld, scanning will L
25 g
l -- - -
be performed on the elbow side and on a portion of the weld using a one half-node refracted longitudinal wave. Transducer interference by the nozzle taper will prevent total weld coverage.
" Relief was initially granted by the NRC during the First Ten-Year Interval for these welds in Attachment 1 (paraphrased abovel. This included First Ten Year Interval Requests for Relief RR-17. NRC approval was documented in correspondence dated .
November 26,1991 and December 17,1991 for VEGP-1 and 2, respectively.
"These welds are' located in a low oxygen, PWR primary system water environment outside of areas considered subject to thermal stratification; therefore, there should be a low potential for cracking in these welds. Complete examinations of other welds in the system experiencing similar environmental conditions have and will be performed, using similar examination techniques. The overall examinations performed on welds in the primary system water environment, in conjunction with the low potential for cracking, should provide reasonable assurance of the operational readiness of the' welds for which relief is requested. Compliance with Code covera'ge requirements would necessitate refabrication of the elbows and/or the Steam Generator primary nozzles, which is completely impractical; therefore, approval should be granted pursuant to 10 CFR 50.55a(g)(6)(i).
Licensee's Proposed Attemative Examination-
"No alternative examinations are proposed."
Eve /uetion-The Code requires 100% volumetric examination of the subject dissimilar metal welds. However, complete volumetric examination is limited by nozzle configuration that restricts access to the examination area. Therefore, the Code coverage requirements are impractical for these welds. To meet the Code requirements, the nozzle welds would require design modifications to allow access for examination. Imposition of the Code requirements would result in a considerable burden on the licensee.
The licensee can examine 50% of these welds, in addition, these welds are part of a larger sample of similar welds that can be examined. The limited volumetric examination, in conjunction with the examination of other similar welds should detect any significant patterns of degradation that may occur and provide reasonable assurance of the structural integrity of the subject dissimilar metal welds.
Conclusion-Based on the impracticality of meeting the Code coverage requirements for /
the subject examination areas, and the reasonable assurance provided by the examinations that can be completed, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i).
3.1.4.3 Request for Relief No. RR-11, Examination Category B-J, item B9.31, Class 1 Branch Connection Welds Code Requirement-Examination Category B-J, Item B9.31 requires 100% surface and volumetric examination, as defined by Figures IWB-2500-9, -10, and -11, for Class 1 branch connection welds 4-inch nominal pipe size and larger.
26
Licensee's Code Relief Request-in accordance with 10 CFR 50.55a(g)(5)(iii), the licensee l requested relief from the Code coverage requirements for the Class 1 branch connection i welds listed in the table below.
l l >& >m sa -
m . M% Table'RRel_1P Ag@M , JE aug; , - ~, a
~
l $Unid !(Weld lDli - sSizeib , am W Limitation , ^ agswg [C6verede .
!O 1 11201-009 4 4" Circ scan (0% cw/ccw) 20%
!' 11201-012-4 Ax scan (0% pipe /80% branch)
F
.1 11201-002 2 6" Cire scan (0% cw/ccw) 22 %
11201-003-2 Ax scan (40% pipe /50% branch)
I 1 11201-009-6 10" Cire scan (50% cw/50% ccw) 50 %
l 11201 010-4 Ax scan (100% pipe /0% branch) 11201-011-5 11201-012-6 1 11201-001-2 12" Circ scan (0% cw/ccw) 22 %
11201-004-3 Ax scan (40% pipe /50% branch) 1 11201-004-2 16" Circ scan (0% cw/ccw) 22 %
Ax scan (40% pipe /50% branch) l 2 21201-009-4 4" Circ scan (0% cw/ccw) 20 %
21201-012-4 Ax scan (0% pipe /80% branch) 2 21201-002 2 6" Circ scan (0% cw/ccw) 22 %
21201-003-2 Ax scan (40% pipe /50% branch) 2 21201-009-6 10" Cire scan (50% cw/50% ccw) 50 %
21201-010-4 Ax scan (100% pipe /0% branch) 21201-011-5 21201-012-6 2 21201-001-2 12" Cire scan iO% cw/ccw) 22 %
21201 004-3 Ax scan (40% pipe /50% branch) 2 21201 004-2 16" Circ scan (0% cw/cew) 22 %
Ax scan (40% pipe /50% branch)
Licensee's Besis for Requesting ReNef-
" Examination of branch connection welds is typically difficuit due to the configuration of the branch connection fitting and the weld design. For branch connection welds in
- the VEGP Reactor Coolant System (RCS) primary loop piping, these problems exist in addition to the problems of examining cast stainless steel pipe material. (See s
Attachment 2*).
"Two basic weld configurations are used for the branch connection designs on the RCS primary loop piping. The 4-inch branch connections were installed using a ' set-on' l weld design, while the 6,10,12, and 16-inch branch conncctions were installed using a " set-in" design.
" Typically, examination coverage of branch connection welds can be obtained by I
- Licensee attachment not included in this report.
27 4
1 I
\
scanning from the main run of pipe using a 45 degree shear wave technique. In this f case, due to the cast stainless steel material used in the RCS primary loop piping, the I examination is limited to a 1/2 node examination using a 45 degree refracted '
longitudinal (RL) wave technique developed for this piping material.
" Examination of the 4-inch branch connection welds from the main run of piping using the 1/2 node RL wave is not possible due to the geometry of the ' set-on' .
configuration. However, partial coverage of the 6,10,12, and 16-inch branch connection welds from the main run using the 1/2 node RL technique is possible because of the ' set-in' configuration. The VEGP branch connection configuration is depicted in Attachment 2*.
"For the 4,6,12 and 16 inch, the ' Calculated Percentages' are less than reported during the First Ten Year interval. Although the same technique is used, circumferential scan limitations were not initially included in the calculations which reduce the accumulative Code Coverage obtained by one-half. Performing circumferential scans for the applicable branch connections is not possible due to weld configurations. While the examination coverages that are listed in Attachment 1
[ paraphrased above] vary from those shown for the First Ten-Year Interval, the actual extent and quality of the examinations remain the same.
"The examinations are being conducted to the fullest extent practical.
" Relief was initially granted by the NRC during the First Ten-Year interval for those welds in Attachment 1 having requests for relief submitted for them. These included First Ten-Year interval Requests for Relief RR-21 for VEGP-1 and 2. NRC approval was documented in correspondence dated November 26,1991 and December 17,1991 for VEGP-1 and 2, respectively.
"These welds are located in a low oxygen, PWR primary system water environment outside of areas considered subject to thermal stratification; therefore, there should be a low potential for cracking in these welds. Complete examinations of other welds in the system experiencing similar environmental conditions have and will be performed, using similar examination techniques. The overall examinations performed on welds in l the primary system water environment, in conjunction with the low potential for cracking, should provide reasonable assurance of the operational readiness of the welds i for which relief is requested. Compliance with Code coverage requirements would l
necessitate refabrication of the elbows and/or the branch connections and main loop f l piping, which is completely impractical; therefore, approval should be granted pursuant to 10 CFR 50.55a(g)(6)li).
Licensee's Proposed A/temative Examination-i "It was deterrnined that the only feasible examination from the branch connection side l would be a refracted shear wave technique from the taper of the fitting. This technique relies on the ability of the shear wave to reflect off of the inner wall of the fitting bore.
Calibration blocks were built with a 10% notch for sensitivity and 3/16 inch side-drilled I . holes for establishing a DAC curve.
"The forged stainless steel material used in all but the 10-inch branch connections 28
allows the use of shear wave ultrasonic techniques; however, the geometry of the fittings still presents problems in obtaining Code-required examination coverage. The fitting side of the 10-inch branch connection is cast stainless steel which precludes the use of a shear wave technique. Also, due to the geometry of the part, examination from the branch connection side using the 1/2 node RL technique is not feasible."
, Eve /uetion-The Code requires 100% volumetric examination of the subject branch
- connection welds. However, complete volumetric examination is limited by component configuration and weld design that restrict access to the examination area. Therefore, the
- Code coverage requirements are impractical for these welds. To meet the Code -
requirements, the nozzle welds would require design modifications to allow access for examination. Impositica of the Code requirements would result in a considerable burden L on the licensee.
The licensee can perform volumetric examination on a portion of each of these welds along with the complete surface examination. In addition, these welds are part of a larger sample of welds that can be examined. The limited volumetric examination, in conjunction with the Code surface examination and the complete examination of other similar welds should detect any significant pattems of degradation that may occur and provide reasonable assurance of the structuralintegrity of subject branch connection welds.
Conclus/on--Based on the impracticality of meeting the Code coverage requirements for the subject examination areas, and the reasonable assurance provided by the examinations that can be completed, it is recommended that relief be granted pursuant to 10 CFR l l . 50.55a(g)(6)(i). '
3.1.4.4 Request for Relief No. RR-12 (Part 1), Examination Category B J, item B9.11, Class 1 Circumferential Wolds Code Requirement-Examination Category B-J, item B9.11 requires 100% surface and volumetric examination of Class 1 pressure retaining circumferential piping welds as defined by Figure IWB-2500-8.
, Licensee's Code Re#ef Request-in accordance with 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from the Code coverage requirements for the item B9.11 circumferential welds listed in the table below.
! *- +
,. J , s i . .:4 g ^' E ETableM12 (PartjE WFF WS " , "WS l A s Uniti G Wald 80' l s Descrip6 cal @j$ 0@9MMAIJmitation$##W < CoveradP l*+ 1 11204 124-1 10" Valve-to-pipe Geometry of valve 50 %
11204-124 7 weld (100% pipe /0% valve) 11204-124-8 1 11204-124-11 10" Pipe-to-tee weld Geometry of tee 88 %
y 11204-124-12 l
i 29
, @d ,
<M- * * '
+ @TaNo RR 12 Part(1)! ** MM - deme U@@ $ .
(DessriptW kg n IJmitadoW)' < + , MCoyoisss?
[Uniti m' il W eld .10 L a e, J, 11204 124 15 11204-124-16 11204-125-1 11204-125-7 ,
11204-125-8 <
11204 125-15 11204-125-16
1 11204 126-1 10" Valve-to-pipe Component Geometry 50%
11204 126-7 weld (100% pipe /0% valve) 11204-126-8 11204-126-15 11204-126-16 11204 127-1 11204 127-7 11204-127-8 11204-127 19
~ 11204-127-20 21204-124-1 10" Valve-to-pipe Geometry of valve 2 21204-124-7 weld (100% pipe /0% valve) 50 %
21204 124-8 2 21204-124 12 10" Pipe-to-tee weld Geometry of tee 88%
21204-125 12 21204 126-12 21204-124-15 21204-124-16 10" Valve-to-pipe Geometry of valve 2 21204-125-1 weld (100% pipe /0% valve) 50 %
21204-125-7 21204-125-8 21204-125-15 21204 125 10" Valve-to-pipe Geometry of valve 50 %
2 21204-126-1 weld (100% pipe /0% valve) 21204 126-7 21204-126-8 21204-126-15 21204 126-16
.2 21204 127-1 10" Valve-to-pipe Geometry of valve 50 % ,.
21204-127 7 weld (100% pipe /0% valve) 21204-127-8 21204 127-19 -
21204-127 20 Licensee's Basis for Proposed Altemative-
"The SA-376 and SA-312 materials exhibit severe angular variations and significant attenuation problems during a typical shear-wave ultrasonic examination. This was determined to be caused by severely banded microstructure. A meaningful ultrasonic examination could not be accomplished on the SA-376 and SA-312 materials using ;
. conventional shear-wave techniques. In addition, physical limitations exist due to the l geometric configuration of some joints, l
30
i l
l " Relief was granted by the NRC for the First Ten-Year Interval. These included First l Ten-Year Interval Requests for Relief RR-26 for VEGP-1 and 2. NRC approval was l documented in correspondence dated November 26,1991 and December 17,1991 for l VEGP-1 and 2, respectively. Subsequent revisions to the requests for relief were l approved by the NRC as documented in correspondence dated January 6,1993 and August 13,1996 for VEGP-1 and 2, respectively. The examinations of the Class 1 and
, 2 Safety injection System piping are being conducted to the fullest extent practicalin the same manner as for the Firs Ten-Year interval.
l- The examinations performed using the half-node technique, while rmt meeting full Code coverage requirements, are considered to be a viable means of detecting any significant circumferential cracking and should provide reasonable assurance of the operation readiness of the subject welds. Compliance with Code coverage requirements would necessitate replacement of the piping, which is completely impractical, therefore, approval should be granted pursuant to 10 CFR 50.55a(g)(6)(i).
Licensee's Proposed A/temative Examination-
"A refracted-longitudinal (L) wave, which is a 1/2 node examination technique, was found to be the best technique during preservice inspection activities. During calibration for the RL-wave examination, the primary reference level will be set using side-drilled holes and appropriate notches. Scanning was possible only at the primary reference level due to the excessive noise associated with the metallurgical structure of the material. On pipe to valve configurations, the weld was examined on the pipe side using the half-node technique. Scanning from the valve side was not possible due to valve geometry. On pipe to tee configurations, the weld was examined on the pipe side using the half node technique. Only partial scanning from the tee was possible due to geometry."
Evaluation-The Code requires 100% volumetric and surface examination of the subject welds. However, complete ultrasonic examination is limited by component configuration and weld geometry which restrict access to the Code-required examination volume.
Therefore, the Code coverage requirements are impractical for these welds. To meet the Code requirements, design modifications would be necessary to allow access for examination. Imposition of these requirements would create a burden on the licensee.
The licensee can examine at least 50% of each of the welds,in addition to the complete surface examination. In addition, these welds are part of a larger sample of
., welds that are being examined to the extent required by the Code. The limited volumetric examination in conjunction with the complete surface examination and the complete examination of other similar welds should detect any significant patterns of degradation
- *- that may occur and provide reasonable assurance of the structuralintegrity for the subject welds.
Conclusion-Based on the impracticality of meeting the Code coverage requirements for the subject welds and the reasonable assurance of the structuralintegrity provided by the examinations that can be performed,it is recommended that relief be granted pursuant to l 10 CFR 50.55a(g)(6)(i).
31
, 3.1.4.5 Request for Relief No. RR-13, Examination Category B-J, Class 1 Pressure Retaining Welds l:
Note: Request for Relief RR 13 was withdrawn by the licensee in the June 26,1998, submittal as the result of an NRC RAI question.
3.1.5 Pump Pressure Boundary.
t
(. No relief requests.
i r
3.1.6 Valve Pressure Boundary.
T l No relief requests.
l 3.1.7 General No relief requests.
3.2 Class 2 Components 3.2.1 Pressure Vessels 3.2.1.1: Request for Relief No. RR-14, Examination Categories C-A, and C-B, items C1.20, C2.21, C2.22, and C2.31, Class 2 Vessel Welds, Nozzle-to-Vessel ;
Welds and Nozzle Inalde Radius (IR) Sections Code Requirement-Examination Category C-A, item C1.20 requires 100% volumetric examination, as defined by IWC 2500-1, of circumferential head welds. Examination ;
Category C-8, item C2.21 requires surface and volumetric examination, as defined by Figure IWC-2500-4(a) or (b), as applicable, for nozzle-to-vessel welds without reinforcing i plates in vessels greater that 1/2-inch nominal wall thickness. Examination Category C B, j ltem C2.22 requires 100% volumetric examination, as defined by Figure IWC 2500-4(a).of l (b), as applicable, for nozzle IR sections in nozzles without reinforcing plates in vessels ;
greater that 1/2-inch nominal wall thickness. Examination Category C-B, item C2.31 ;
requires 100% surface examination, as defined by Figure IWC-2500-4(c) for nozzle ]'
reinforcing plate welds in vessels 1/2-inch nominal wall thickness.
licensee's Code Re#ef Request--In accordance with 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from performing the surface and/or volumetric examinations, to the extent *
, I required by the Code, for the welds listed in Table RR-14 below. l 3
r l
32
__ _ = _ . - _ _ _ _ _ _ - . __ _ - - _ _ -
Stable FIR-14.1 >
Category / - iWeld IDI [ Description ? l
- lLimitstion) (Covd g ,
- m s CA/C1.20 11204-V6-001-WO2 VEGP-1 BIT Head-to-shell Vessel supports and taper 57 %
weld (76% head /0% shell)
, 11204-V6-001-WO3 Vessel Taper 62 %
(100% head /0% shell)
C-B/C2.21 11201-86-001-W18 VEGP-1 SG MS outlet Nozzle configuration 50%
nozzle-to-head we.ld (100% shell/0% nozzle)
C-B/C2.21 11201-86-002-W19 VEGP-1 SG FW nozzle- Nozzle configuration 50 %
to-shell weld (100% shell/0% nozzle)
C-B/C2.21 11201-86-004-W26 VEGP-1 SG Aux. FW Nozzle configuration 50 %
nozzle-to-shell weld (100% shell/0% nozzle)
C-B/C2.21 11201 -V6-001 -WO1 VEGP-1 BIT nozzle-to- Nozzle configuration 50 %
11201-V6-001-WO4 head weld (100% shell/0% nozzle)
C-B/C2.21 21201-B6-001-W18 VEGP-2 SG MS outlet Nozzle configuration 50 %
nozzle-to-head weld (100% shell/0% nozzle)
C-B/C2.21 21201-B6-002-W19 VEGP-2 SG FW nozzle- Nozzle configuration 50 %
to-shell weld (100% shell/0% nozzle)
C-B/C2.21 21201-86-002-W26 VEGP-1 SG Aux. FW Nozzle configuration 50 %
nozzle-to-shell weld (100% shell/0% nozzle) l C-B/C2.22 11201-86-001-IR03 VEGP-1 SG MS outlet Nozzle configuration 0%
nozzle IR (No IR)
C-B/C2.22 11205-E6-001-IR01 VEGP-1 RHR HX 1R Nozzle configuration 0%
11205-E6-001-IR02 (0% shell/0% nozzle) i C-B/C2.22 11205-E6-002-IR01 VEGP 1 RHR HX IR Nozzle configuration 0%
11205-E6-002-IR02 (0% shell/0% nozzle)
C-8/C2.22 21201-86-001-lR03 VEGP-1 SG MS outlet Nozzle configuration 0%
nozzle IR (No IR)
C-B/C2.22 21205-E6-001-IR01 VEGP-2 RHR HX IR Nozzle configuration 0%
21205-E6-001-IR02 (0% shell/0% nozzle) 21205-E6-002-IR01 21205-E6-002-lR02
. C-B/2.31 11205-E6-001-WO9 VEGP-1 RHR HX Nozzle configuration 0%
thru W12 reinforcement plate (Inaccessible) 11205-E6-002-WO9 l '- thru W12 C-B/2.31 21205-E6-001-WO9 VEGP-2 RHR HX Nozzle configuration 0%
thru W12 reinforcernent plate (inaccessible) 21205-E6-002-WO9 l thrujV12 Licensee's Basis for Requesting Relief- l
"(a) The configuration of the steam generator (SG) steam outlet nozzle is such that no inner radius exists. The nozzle is manufactured from a forging-that is a solid l 33 i l
i
l l
l block of steel. Seven (7) holes, each 81/2 inch diameter, have been drilled through this forging to provide an outlet for the steam. Thus, this nozzle does not have a conventional inner radius.
(b) Geometric configuration of the SG main steam, auxiliary feedwater and main feedwater nozzles presents physical limitations that prevent complete coverage during ultrasonic examination. Scanning from the nozzle side is not feasible. ,
l (c) The configuration of the VEGP Residual Heat Removal (RHR) heat exchanger (Hx) nozzles differs from that shown in ASME Section XI Figure IWC-2500-4(c). - ,
I Although the reinforcing plate welded to the vessel has a rounded configuration in the flow it is not a true nozzle inner radius when compared with the configuration in Figure IWC-2500-4(c). Please refer to Attachment 2' for a figure depicting the VEGP configuration it is not possible to perform an inner radius ultrasonic examination since the interface between the reinforcing plate and the RHR heat exchanger vessel wall prohibits volumetric examination. Although the reinforcement plate is welded to the inside diameter of the heat exchanger wall, the reinforcing plate-to-vessel welds are inaccessible. Therefore, it is impractical to perform a surface examination on the reinforcing plate-to-vessel welds.
(d) For the VEGP-1 Boron Injection Tank (BIT), the nozzle and shell configuration presents physical limitations that prevent complete coverage during ultrasonic examinations.
"The examinations are being conducted to the fullest extent practical.
Relief was initially granted by the NRC during the first ten-year interval for those welds / components in Attachment 1 having requests for relief submitted for them.
These included First Ten-Year Interval Requests for Relief RR-28, RR-29, RR-30 and RR-32. NRC approval was documented in correspondence dated November 26,1991 and December 17,1991 for VEGP-1 and 2, respectively, it was determined during the first ten-year interval review process that geometric configurations and interferences make the volumetric examinations of these welds impractical to perform to the extent recuired by the Code. The subject components would require extensive modifications in order to obtain complete compliance with the specific requirements of ASME Section XI. The increase in plant safety would not compensate for the burden placed on the licensee that would result from imposition of the requirement.
"Although there are physical obstructions which limit the amount of examination '
coverage for those components identified in Attachment 1 [ paraphrased abovel, reasonabla assurance still exists that an acceptable level of quality and safety will be maintair,ed since there have beer no catastrophic failures of Class 2 pressure vessels. -
SNC requests that relief be authorized pursuant to 10 CFR 50.55a(g)(6)(i) since imposing the Code requirements is impractical."
In the June 26,1998, submittal, the licensee stated:
"The Boron injection Tank head to shell welds in question (11204-V6-WO2 and WO3)
- Included in licensee's submittal, but not in this report.
34
j have a configuration such that scanning from the shell side is impractical. The l' machined transition angle combined with the shell being of a centrifugal cast material does not allow the necessary Code examination coverage. However, the combined l coverage obtained during ISI examinations has improved to an accumulative total of fifty-seven percent (57%) for WO2 and sixty-two percent (62%) for WO3. This represents an increase in examination coverage from the previously reported thirty-eight percent (38%) and fifty percent (50%) for welds WO2 and WO3, respectively.
- Weld WO2 has 4 support leg interferences in addition to the aforementioned 4 restrictions, m
Licensee's Proposed Attemative Examination-
"No alternate examination is proposed. The affected Class 2 vessel welds are being examined to fullest extent practical."
Evaluation-The Code requires 100% volumetric examination of the subject examination areas. However, access to these areas is limited by component configuration or designed such that examinations cannot be performed. Therefore, the Code-required volumetric examinations are impractical for these areas. Design modifications would be required to allow access for examination. Imposition of the Code requirements would result in a burden on the licensee.
For the boron injection tank head-to-shell welds and the steam generator nozzle-to- I vessel welds, the licensee can examine at least 50% of each weld. These examinations should detect any significant patterns of degradation. Therefore, reasonable assurance of the structural integrity will ba provided by the partial examinations.
The steam generator outlet nozzles were designed with an internal multiple hole type flow restrictor rather than a single radiused nozzle as described by IWC-2500-4.
Therefore, there are no IR sections to examined. The RHRHX nozzles were fabricated with a reinforcement plate that is radiused, but is not a true inner radius and is separate from the vessel wall. Therefore, degradation that could occur should be limited to the reinforcement plate and should not effect the vesselitself.
' Conclusion-Based on the evaluation above, it is concluded that the Code volumetric examinations are impractice! for the subjected examination areas. Therefore, it is recommended that relief be granted pur aant to 10 CFR 50.55c(g)(6 (i).
3.2.2 Piping 3.2.2.1 Request for Relief No. RR-12 (Part 2), Examination Category C-F-1, item C5.11,
- 4. Class 2 Circumferential Welds Code Requirement--Examination Category C-F-1, item C5.11 requires 100% surface and volumetric examination of pressure retaining circumferential welds in austenitic stainless steel piping as defined by Figure IWC-2500 7.
Licensee's Code Re//e/ Request--In accordance with 10 CFR 50.55a(g)(5)(iii), the licensee L requested relief from the Code coverage requirements for the Class 2 circumferential welds listed in the table below.
35
- --- _ . - ~ . - - . . - - ~ . - . . - - . - - - , . _ . . - - - - - - . . . - . - - _ .
4 a
ya'n e' t w d A M, iWTaWe512fartt2)$@NM H#@H@@ f!W 4MiliEI
, Unit " 'O ' W old ID ' d < DeshiptionM $CMi 1.lmitation MO W#
- yCdMad@
1 11204 120-6 10" Pipe-to-valve Geometry of valve 50 % :
11204 121-6 welds (100% pipe /0% valve) 11204-122-6 11204 123-6 '
i 2 21204 123-6 10" Pipe-to-valve Geometry of valve 50% !
welds (100% pipe /0% valve)
Licensee's Basis for Proposed Attumative-
"The SA-376 and SA-312 materials exhibit severe angular variations and significant attenuation problems during a typical shear-wave ultrasonic examination. This was determined to be caused by severely banded microstructure. A meaningful ultrasonic examination could not be accomplished on the SA-376 ed SA-312 materials using conventional shear wave techniques. In addition, physical limitations exist due to the -
geometric configuration of some joints.
" Relief was granted by the NRC for the First Ten-Year Interval. These included First Ten-Year interval Requests for Relief RR-26 for VEGP-1 and 2 NRC approval was ,
. documented in correspondence dated November 26,1991 and December 17,1991 for ,
VEGP 1 and 2, respectively. Subsequent revisions to the requests for relief were approved by the NRC as documented in correspondsnce dated January 6,1993 and August 13,1996 for VEGP-1 and 2, respectively. The examinations of the Class 1 and ,
2 Safety injection System piping are being conducted to the fullest extent practical in the same manner as for the First Ten-Year Interval.
1 The examinations performed using the half-node technique, while not meeting full Code coverage requirementr, are considered to be a viable means of detecting any significant ;
circumferential crack:ng and should provide reasonable assurance of the operation ;
readiness of the subject welds Compliance with Code coverage requirements would necessitate replacement of the piping, which is completely impractical; therefore, approval should be granted pursuant to 10 CFR 50.55a(g)(6)(i).
-Licensee's Proposed Attemative Examination- .
"A refrdcted longitudinal (L) wave, which is a 1/2 node examination technique, was found to be the best technique during preservice inspection activities. During calibration for the RL-wave examination, the primary reference level will be set using side-drilled .-
holes and appropriate notches. Scanning was possible only at the primary reference level due to the excessive noise associated with the metallurgical structure of the material. On pipe to valve configurations, the weld was examined on the pipe side using the half-node technique. Scanning from the valve side was not possible due to
- valve geometry. On pipe to tee configurations, the weld was examined on the pipe side using the half node technique. Only partial scanning from the tee was possible due to ;
geometry."
- EveAustion-The ' Code requires 100% volumetric and surface examination of the subject welds. However, complete ultrasonic examination is limited by component configuration and weld geometry which restrict access to the Code-required examination volume, ,
J 36
I.
Therefore, the Code coverage requirements are impractical for these welds. To meet the Code requirements, design modifications would be necessary to allow access for examination. Imposition of these requirements would create a burden on the licensee.
The licensee can examine at least 50% of each of the welds, in addition to the complete surface examination. In addition, these welds are part of a larger sample of
, welds that are being examined to the extent required by the Code. The limited volumetric examination in conjunction with the complete surface examination and the complete
. examination of other similar welds should detect any significant patterns of degradation L- that may occur and provide reasonable assurance of the structural integrity for the subject l welds.
I Conclusion--Based on the impracticality of meeting the Code coverage requirements for i the subject welds and the reasonable assurance of the structuralint~1rity provided by the I examinations that can be performed,it is recommended that relief ao e ranted pursuant to 10 CFR 50.55a(g)(6)(i).
3.2.2.2 Request for Relief No. RR-15, Examination Category C-F-1, Class 2 Pressure Retaining Welds Note: Request for Relief RR-15' was withdrawn by the licensee in the June 26,1998, submittal as the result of an NRC RAI question. I 3.2.2.3 Request for Relief No. RR-16, Examination Category C-F-2, item C5.51, Class 2 ,
Carbon Steel Piping Welds I Code Requiremant-Examination Category C-F-2, item C5.51 requires 100% surface and volumetric examination, as defined by Figure IWC-2500-7, for circumferential welds in carbon steel piping greater than or equal to 3/8-inch nominal wall thickness and greater than 4-inch nominal pipe size.
' L/consee's Code Re#e/ Request-in accordance with 10 CFR 50.55a(g)(5)(iii), the licensee requestad relief from the Code coverage requirements for the welds listed in the table below.
s'> n 3~- qw : %gg :YgTatytpt18l gcf " ;' gy >pm '*
@Waldfl02 - 4 EDesedption $$ ' 'W; #4 Omitation ':^ n ~;' C : [C64ksioj On16 1 11301-001-5 29.5" Pipe-to-valve 0% valve side,64% pipe side 74%
1' 11301-001-6 29.5" Valve-to-pipe 0% valve side,100% pipe side 74 %
1 11301-004-5 29.5" Pipe-to-valve 0% valve side,88% pipe side 82 %
1 11301-004-6 29.5" Valve-to-pipe - 0% valve side,82% pipe side 82 %
11301-004-8 29.5" Pipe-to-valve
\. Ucensee's Basis for Requesting Re6ef-I "Physicallimitations due to geometric configuration of the welded areas restrict coverage of the ex'aminations volume required by Figure IWC-2500-7. The volumetric l
37
examinations of the Class 2 piping welds are being conducted to the fullest extent ,
practical' As noted herein, physical access is restricted thereby preventing full Code examination coverage."
In the' June 26,1998, submittal, the licensee stated that the welds above have been examined to the extent practical and that due to OD configuration and/or material
- attenuation (2.812-inch thicki 29.5-inch diameter welds) examinations using extended ,
beam paths is not practical. .
Licensee's Proposed Altemative Examination-
"No alternate examination is proposed."
Evaluation-The Code requires 100% surface and volumetric examination of th9 subject Class 2 piping welds. However, the weld configuration (pipe / valve) prevents access from one side of.the welds and makes the Code volumetric examination impractical to complete for these welds. Design modifications would be required to provide access for examination, and imposition of the Code requirements would result in a burden on the licensee.
The licensee can examine a significant portion of each weld from the pipe side and in the circumferential direction.- In addition, the surface examination can be examined to the extent required by the Code. Therefore, the limited volumetric examination, in conjunction with the Code-required surface examination, should detect any significant patterns of. ,
degradation that may occur and provide reasonable assurance of the continued structural integrity for the subject welds.
Conclusion-Based on the impracticality of meeting the Code coverage requirements and the reasonable assurance of the structuralintegrity provided by the examinations that can be performed,it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i). ,
3.2.2.4 Request for Relief No. RR-17, Examination Category C-F-1, item C5.11, Class 2 Piping Welds Note: Request for Relief No. RR-17 was evaluated and authorized pursuant to 10 CFR 50.55ata)(3)(i) in an NRC SER dated October 24,1997. ,
4 3.2.2.5 Remst for Relief No. "18, Examination Category C-F-1, items C5.11, C5.30 and C5.41 Class 2 Austenitic Stainless Steel Piping Walds in the Nuclear Service Cooling Water (NSCW) System Code Requiremont-Examination Category C-F-1, item C5.11 requires 100% surface and volumetric examination, as defined by Figure IWC-2500-7, for Class 2 circumferential piping welds greater than or equal to 3/8-inch nominal wall thickness and greater than 4- l inch nominal pipe size (NPS). Item C5.30 requires surface examination, as defined by Figure IWC-2500-7, for socket welds in piping greater than or equal to 3/8-inch nominal wall thickness and greater than 4-inch NPS. Item C5.41 requires surface examination, as defined by Figures IWC-2500-9 to -13 inclusive, for circumferential branch connections of ,
I 38
branch piping 2-inch NPS and greater.
! Licensee's Proposed Attemative-In accordance with 10 CFR 50.55a(a)(3)(i), the licensee l proposed an alternative to the Code requirements to exclude the NSCW system from the Class 2 weld population. The licensee stated:
l y " System pressure testing will be performed on the Class 2 portions of the NSCW
'- system piping as described in IWA-5000 and IWD-5000 or in accordance with ASME Section XI Code Case N-498-1. In addition, the associated component supports will be h examined per the requirements of ASME Section XI and Code Case N-491. The Class 2 pressure-retaining welds of NSCW will be eliminated from the total population of ASME Category C-F-1 welds from which the 7.5% sample of welds are to be nondestructively examined. To insure containment integrity, Primary Reactor Containment Leakage Testing as required by 10 CFR 50, Appendix J will continue to be performed.
Licensee's Basis for Proposed Attemative-
"The NSCW system is a cooling water system that performs a Class 3 function as defined in NRC Regulatory Guide 1.26, Revision 3, 'Ouality Group Classifications and Standards for Water , Steam , and Radiological-Waste-Containing Components of Nuc!aar Power Plants', and therefore per Table 2500-1, Examination Categories D-A, D-8, and D-C, no ncndestructive examinations are required to be performed on the pressure-retaining welds. There is a small portion of this piping that is classified ASME Class 2 as it provides a containment isolation function as well the Class 3 cooling water function. These Class 2 pressure-retaining welds are greater than 4 inches nominal pipe size (NPS) and less than 3/8 inches nominal wall thickness and do not require nondestructive examination per ASME Section XI Table IWC-2500-1, Examination Category C-F-1.
"The NSCW system is a cooling water system that performs a Class 3 function as defined in NRC Regulatory Guide 1.26, Revision 3, and therefore, per Table IWD-2500-1, Examination Categories D-A, D-8, and D-C, no nondestructive examinations are required to be performed on the pressure-retaining welds. To satisfy the Class 3 function Inservice inspection (ISI) requirements, this piping receives system pressure test as described in IWA-5000 and IWD-5000 or in accordance with ASME Section XI Code Case N-498-1 (See Request for Relief RR-16) and the component supports are examined per the requirements of ASME Section XI Code Case N-491. There is a small portion of this piping that is classified ASME Class 2 as it provides a containment L
isolation function as well the Class 3 cooling water function. (Note: To insure its containment integrity, the VEGP-1 and 2 containments receive 10 CFR 50, Appendix J Primary Reactor Containment Leakage Tests.) The piping wolds in this portion of the NSCW system receive the required Class 3 examinations described above and because they are classified Class 2, are required to be counted as part of the total population of ASME Category C-F-1 welds from which the 7.5% sample of welds are to be
- nondestructively examined. These NSCW welds, however, are not required to be l
nondestructively examined per Examination Category C-F-1 because of a NPS greater
- than 4 inches and a nominal wall thickness less than 3/8 inches. Therefore, the Class 2 l portion of the NSCW system is required to receive the same level of ISI as the Class 3 portion and performs primarily a cooling water function. The net impact of the
- Examination Category C-F-1 requirement is that it increases the quantity of pressure-39
, . - -- .- - - - ~ ~ - -.- - . - . _ .-.-- _ . - - -
P A
tretaining welds to be examined and resultsi n those exami nati onsbi e ng requ i red on :
Class 2 systr ms other than NSCW. Because the primary function of these welds is a Class 3 cooling' water function SNC believes that the ISI requirements for Class 3 are i adequate for providing an acceptable level of quality and safety. Also,in the First Ten-year interval, SNC performed nondestructive examinations on 7.5% of the Class 2 NSCW pressure-retaining welds and did not find any significant problems with the .
welds examined. Therefore, it is requested that the r:oposed attemative be authorized , ,
pursuant to 10 CFR 50.55a(a)(3)(i). ..
In the June 26,1998, submittal,'the licensee stated: -;
l "In the First Ten-Year Interval, SNC performed nondestructive examinations on seven-
'and-one half percent (7.5%) of the Class 2 NSCW pressure-retaining welds
-(approximately.162 welds for the two VEGP units) and did not find any significant problems with the welds examined. There are 1096 and 1068 welds in the VEGP-1 and VEGP-2, respectively, in this portion of the NSCW piping.
"Although < 'sssified as Class 2, the primary purpose of the subject NSCW welds is that they provic. ; Class 3 cooling water function. Request for Relief RR-18 (Basis for Relief) indicates that the subject piping is greater than 4-inches nominal pipe size and .
has a nominal wall thickness of less than 3/8-inch; therefore, nondestructive examination per ASME Section XI Table IWC4500-1, Examination Category C F-1 is not required. As a result, SNC believes that the ISI requirements for Class 3 are
- adequate for providing an acceptable level of quality and safety thereby justifying the exclusion of these welds designated as Class 2 from the total weld count for true Class 2 welds. The small portion of this piping that is classified as ASME Class 2 cannot be reclassified since it provides a containment isolation function. There is only one isolation valve, thus the Class 2 classification." ,
l EvaAustion-The Code requires surface and/or volumetric examination of Class 2 piping welds greater than or equal to 3/8-inch nominal wall thickness and greater than 4-inch NPS. Piping greater than 4-inch NPS but less than 3/8-inch nominal wall thickness are excluded from examination but must be included in the Class 2 weld population. The licensee has proposed to eliminate the NSCW system, which contains over 1000 welds in each unit, from the Class 2 weld population. As stated by the licensee, this system '
- pr'ovides a Class 3 cooling water function. It cannot be reclassified as Class 3 because it also provides a containment isolation function and has only one isolation valve.
The licensee contends that tne proposed alternative will provide an acceptable level of quality and safety, and the INEEL staff agrees that the alternative will have little if any .
- l' impact on the NCSW system. However, eliminating these welds from the Class 2 weld -
- l. ~ population will effect examinations of other Class 2 systems and it is not clear how l reducing the number of examinations performed on other systems provides an acceptable level of quality and safety. Therefore, the proposed alternative should not be authorized.
Conclusion-Based on the evaluation above, it is recommended that the licensee's
- proposed alternative not be authorized. The licensee has not demonstrated that the
! proposed alternative will provide an acceptable level of quality and safety.
40
3.2.2.6 Request for Relief No.19, Use of Code Case N-524, Altemative Examination l Requirements for Longitudinal Welds in Class 1 and 2 Piping, for Examination of Class 2 Piping Welds -
Note: Request for Relief No. RR-19 was evaluated and authorized in an NRC SER dated March 24,1998.
3.2.2.6 Request for Relief No. 21, Use of Code Case N408-2, A/temative Rules for Examination of Class 2 Piping Welds 1
Note: Request for Relist No. RR-21 was evaluated and authorized in an NRC SER dated I March 24,1998. l 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 i No Relief Requests 3.4 Pressure Tests 3.4.1 Class 1 System Pressure Tests No relief requests 3.4.2 Class 2 System Pressure Tests No relief requests 3.4.3 Class 3 System Pressure Tests No relief requests i
T.I 41 i-
3.4.4 General 3.4.4.1 Request tor Relief No. RR-23, Use of Code Case N-4981, A/temative Rules for 10-Year System Hydrostetic Testing for Class 1, 2, and 3 Systems Note: Request for Relief No. RR-23 was evaluated and authorized in an NRC SER dated March 24,1998. ,
3.4.4.2 Request for Relief No. RR 24, Use of Code Case N-416-1, A/temative Pressure
- Test Requirements for Welded Repairs or initialization of Replacement items by Welding, Class 1, 2, and 3 Systems Note: Request for Relief No. RR-24 was evaluated and authorized in an NRC SER dated March 24,1998.
3.4.4.3 Request for Pelief RR-26, IWA 5242(a), Visual Examination of Insulated Components Code Requirement-For systems borated for the purpose of controlling reactivity, Subparagraph IWA-5242(a) requires removal of insulation from pressure-retaining bolted connections for VT-2 visual examination during system pressure testing.
Licensee's Proposed A/temative-Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee has proposed an alternative to the requirements of IWA-5242(a). The licensee stated:
".CJass 1 comoonents:
" Insulated Class 1 pressure-retaining bolted connections will be uninsulated an) VT-2 examined once each refueling outage while the connections are at atmospheric or static pressures. The bolted connections mil be examined again with the insulation installed during the regularly scheduled system pressure test at nominal system operating temperature and pressure as required by Table IWB-2500-1, Examination Category B-P.
" Class 2 and 3 comoonents:
" Insulated Class 2 and 3 pressure-rete ning bolted connections will be uninsulated and VT-2 examined once each inspectior period, vehen the connection is not at pressure. ,.
In addition, where this request for re ief is applied. the frequency between the insulation removals (including the V1-2 examination) will not exceed forty (40) months except as follows. If a reactor refooling outage is in progress or is scheduled to start -
within six months when the 40 moaths expires, insulation removal and the VT-2 examination would be allowed to be deferred provided that they are performed prior to plant startup following the reactor refueling outage. These examinations may be performed when the connections are not at the pressures required by IWA-SOOO, IWC-5000 and IWD-5000. The bolted connections will be examined again with the insulation installed during the regularly scheduled (once per inspection period) system pressure test as required by Table IWC-2500-1, Examination Category C-H and Table IWD 2500-1, Examination Category D-A."
42
I l
l 1
l licensee's Basis for Proposed Attemative-
" Subparagraph IWA-5242(a) specifies that insulation must be removed from pressure-retaining bolted connections for VT-2 visual examination during the performance of system pressure testing. This is applicable to the following systems:
- Reactor Coolant System (System consists of Class 1 and 2 components),
- Chemical and Volume Control System (System consists of Class 1,2, and 3 i components), l Residual Heat Removal System (System consists of Class 1 and 2 components),
- Safety injection System (System consists of Class 1 and 2 components), and l
- Nuclear Sampling System - Liquid (System consists of Class 2 components). .
l
!. " Class 1 Components:
" Table IWB-2500-1, Examination Category B-P requires a system leakage test (lWB-
. 5221) and corresponding VT-2 visual examination of Class 1 components each refueling outage prior to plant startup. This system leakage test is performed in Mode 3 when the Reactor Coolant System is at Nominal Operating Pressure (=2235psig) and Nominal Operating Temperature (=550'F to 650'F). The majority of the Class 1 components are in the Reactor Coolant System however some portions of the Class 1 i boundary extend to include portions of Safety injection, Chemical Volume Control, !
and Residual Heat Removal systems. All Class 1 components are in containment.
The rmoval and installation of insulation during the performance of system pressure testing inside containment piesents the following hazards:
- Increased potential for debris to be in containment which could migrate to the Containment Emergency Sumps and restrict the suction of the Emergency Core Cooling System during accident (LOCA). All debris is required to be removed from containment prior to entering Mode 4,
- Increased potential for personnel heat stress since the containment ambient temperature may be as high as 100*F, e increased potential for personnel burn injuries due to installation of insulation in proximity of extromely hot components,
- Increase personnel safety hazard since ladders would have to be used to inspect i
many of the bolted connections and replace the insulation. Temporary work
- ,, platforms /scaffol ding inside containment are removed prior to entering Mode 4, e increased radiation exposure to personnel since temporary shielding is removed prior to entering Mode 4, and
- Increased potential for impacting outage duration due to amount of manpower required to support insulation removal and examinations during Mode 3 following refueling outage activities.
I 43
a
" Class 2 Components:
" Table IWC-2500-1, Examination Category C-H requires a system pressure test (IWC-5221) during system functional and system inservice tests and correspcoding VT-2 visual examination on Class 2 components once each inspection period. The following discusses tne applicable systems and the basis for relief for each Class 2 system: ,
- 1. " Basis for Relief of Reactor Coolant System (RCS):
The Class 2 portions of RCS are located adjacent to the Class 1 boundary and are classified as Class 2 based on line size and isolation valve criteria. The system inservice tests for these Class 2 portions of RCS are VT-2 examined in Modes 1,2 and 3 and therefore the same basis for relief as provided for Class 1 cpplies. These Class 2 pressure boundaries are located in containment.
- 2. " Basis for Relief for the Chemical and Volume Control System (CVCS):
For those portions of CVCS which are located inside containment (e.g., Charging, Letdown, Excess Letdown, Alternate Pressurizer Spray, Reactor Coctant Pump Seal Leakoff) the same basis for relief as piovided for Class 1 above applies except that the system operating temperatures are less resulting in less potential for bum-related injuries. The VT-2 examinations are performed in Modes 1,2, and 3.
For those portions of CVCS which are located outside containment, radiation levels, high component temperatures (e.g., approx. 290*F for Letdown), and availability of personnel during non-outage times may preclude removing insulation while the pressure-rctaining bolted connections are pressurized. CVCS is inservice during power operation and, as such, many,if not all, components will be uninsulated, VT-2 examined, and reinsulated when the pressure-retaining bolted connections are pressurized. However, as previously addressed, conditions may be present wnich may not allow insulation removal except during refueling outages.
- 3. " Basis for Relief for the Residual Heat Removal System (RHR):
The RHR is placed inservice during shutdown prior to refueling activities in Mode ,.
4 when the Reactor Coolant System is =350
- F and =350 psig. RHR remains inservice in Modes 5 and 6 during the refueling outage and remains inservice in Mode 4 during startup following the refueling outage. The VT-2 examinations .
are performed in either Mode 4 or Mode 5 when the RCS is =350 psig.
l For those portions of RHR which are located inside containment, the samn basis for relief as provided for Class 1 above applies except that the system operating temperatures are less resulting in less potential for bum-related injuries. It is impractical to attempt to limit VT-2 examinations to Mode 5 in order to avoid the complications of performing RHR pressure test in Mode 4. The VT-2 examinations are performed in Modes 4 or 5.
44
p i
For those portions of RHR which are located outside containment, radiation l levels, high component temperatures, availability of personnel, and increased l L thermal loads on chilled water room cooling systems may preclude removing l
insulation while the pressure-retaining bolted connections are pressurized. It is i significantly more prudent to uninsulate, VT-2 examine, and reinsulate the i
[ pressure-retaining bolted connections in RHR when the system is not pressurized during non-outage times or during refueling outages when the system is not at l the required pressure.
l
- 4. " Basis for Relief for the Safety injection System (SI): i l
The system pressure tests performed on Si are either system functional tests or system inservice tests as follows:
Some of the system functional tests are performed during Modes 1,2, and 3 when RCS pressure is greater than Si pump discharge pressure. VT-2 examinations are performed on portions of SI during various activities and tests t which require a SI pump to be in operation. The scope of these VT-2 L examinations includes components both inside cnd outside containment. The 1 performance of the VT-2 examinations during these activities is generally
-performed in less than one hour to minimize run time on the Si pumps.
The remainder of the system functional tests are performed during Mode 6 and
' defueled conditions with the reactor head removed. VT-2 examinations are performed on portions of SI which are pressurized during check valve flow testing activities which involve injection of water into the reactor pressure vessel. The scope of these VT-2 examinations includes components both inside and outside containment. The performance of the VT-2 examinations during >
these activities is generally performed in less than one hour to minimize run time on the applicable pumps and to hMhnize the impact on critical path testing i
during refueling outages.
Some'of the system inservice tests are performed during Modes 1,2, or 3 on portions of Sl which are pressurized by the Si accumulator tanks. The SI
. . accumulator tanks are generally depressurized during refueling outages. The scope of these VT-2 examinations includes components which are located only inside containment.
^
.' e,
- . - The remainder of the system inservice tests are performed on portions of SI 5 . which are pressurized by the static head of the refueling water storage tank. '
R . The VT-2 examinations on these portions of Si are generally performed during power operation (Mode 1) but may be performed in other Modes if tank levels are adequate. The scope of these VT-2 examinations includes components which are located only outside containment.
t For those portions of Sl which are located outside containment radiation levels
!. and availability of personnel during non-outage times or during system functional l testing may preclude removing insulation while the pressure-retaining bolted
- i. connections are pressurized.
45
,4
. . . . . - . ., . . - - , - - , a
For those portions of SI which are located inside containment and VT-2 examined during Modes 1,2, and 3 the same basis for relief as provided for Class 1 above applies except that the system operating temperatures are less, resulting in less potential for burn-related injuries.
For those portions of SI which are located inside containment and VT-2 examined during Mode 6 and defueled conditions with the reactor vessel head removed ,
containment radiation levels and availability of personnel during system '
functional testing may preclude removing insulation while the pressure-retaining bolted connections are pressurized.
- 5. Basis for Relief for the Nuclear Sampling System - Liquid:
The liquid portions of the Nuclear Sampling System are used for providing samples for analysis purposes of the RCS, CVCS and RHR. This system is located both inside and outside containment and is subject to the same system pressure tests as the systems for which it used to provide samples. Therefore, the same basis for relief as discussed above for RCS, CVCS and RHR is applicable to the liquid portions of the Nuclear Sampling System.
" Class 3 components:
" Table IWD-2500-1, Examinatiun Category D-A requir6s a system inservice test (LWD-5221) and corresponding VT-2 visual examination on Class 3 components once each inspection period. Subparagraph IWA-5242(a) is applicable to the boric acid storage tank and boric acid transfer portions of the CVCS. System inservice test are performed as follows:
"Some of the system inservice tests are performed on portions of CVCS which are pressurized by the static head of the boric acid storage tank. The scope of these VT-2 examinations includes components which am located only outside containment.
The VT-2 examinations on these portions of CVCS are generally performed during power operation (Mode 1) but may be performed in other Modes if tank levels are adequate. These system inservice tests are generally performed during power operation and, as such, many, if not all, components will be uninsulated, VT-2 examined, and reinsulated when the pressure-retaining bolted connections are pressurized. However, availability of personnel during non-outage times may preclude removing insulation while the pressure-retaining bolted connections nre pressurized.
.3 "The remainder of the system inservice tests are performed on portions of CVCS which are pressurized when a boric acid transfer pump is operating. The scope of '
these VT-2 examinations includes components which are located outside containment. The VT-2 examinations on these portions of CVCS are generally performed during power operation (Mode 1) with a boric acid transfer purnp running with system valves aligned in a recirculation flowpath which precludes injecting high concentrations of borated water into CVCS and ultimately into the RCS. The boric acid transfer pumps are operated as necessary to perform system functions and necessary testing and, as such, are not continuously in operation. Since these pumps are not continuously in operation availability of personnel during non-outage times may preclude removing insulction while the pressure-retaining connections are 1
l 46
~ .. . ., . - .-.- .-.- - - __ . - - . - . -.-.- - -.- - . -
l l
! pressurized.
"The 1983 Edition through Summer 1983 Addenda of ASME 3ection XI was applicable for the First Ten-year interval at VEGP lWA-5242 of the 1983 Edition throuch Summer 1983 Addenda of ASME Section XI did not require insulation removal; therefore, this request for relief was not needed at VEGP during the First-
. Year Interval, i
+ " Justification for Grantina Relief:
[ " Class 1 Comoonents t
l "The following items are justification for granting relief for Class 1 Components:
- 1. . Evidence of leakage through pressure-retaining bolted connections which are in systems which are borated for the purpose of controlling reactivity is readily detectable by visual observation when systems are not at operating temperature and pressure. The boric acid concentrations are sufficiently high such that boric acid residues will be present if leakage has occurred at the pressure-retaining
- bolted connection.
- 2. In addition, the ASME Section XI Code Committee has issued Code Case N-533 (copy provided as Attachment 1 to this request for relief) which allows as an alternative for Class 1 pressure-retaining bolted connections that insulation may l be removed and VT-2 examined when the connection is not pressurized. The Code Case also requires that any evidence of leakage be' evaluated in accordance H with IWA-5250.~ Refer to Request for Relief RR-25 for details concerning relief -
from IWA-5250(a)(2).
- 3. Compliance with the Code presents hardships previously discussed lin " Basis for-Proposed Alternative" abovel.
- 4. For the reasons discussed above, SNC has determined that implementation of the proposed alternatives to the Code requirements provides an acceptable level of L -ouality and' safety and therefore requests that the proposed alternative pursuant j to 10 CFR 50.55a(a)(3)(i).
P
'" Class 2 and 3 Components t.
l' "The' proposed alternative provides the following two-phars methodology for ensuring
'I the integrity of the pressure-retaining bolted connect.ons.
o
'1. " Removing insulation and performing a visual examination (at modified time period specified in the Alternative Examination), when the pressure-retaining L bolted connection is not at pressure, will allow for detection of previously occurring leakage through the presence of boric acid crystals. Boric acid concentrations are sufficiently high such that boric acid residues will be present
'. and can be visually obrerved if leakage has occurred at the pressure-retaining i- bolted connection. .
i-F 47 i
+ g e w-e- p - - - ~y wT--- -e-
- 2. " Performing a system pressure test on the bolted connection with the insulation in place, utilizing the Code specified holding time for any leakage to penetrate the insulation, will provide a means of detecting any significant leakage.
"The proposed alternative provides reasonable assurance that the structural integrity of the pressure-retaining bolted connections will be maintained, thereby, continuing to provide an acceptable level of quality and safety. Therefore, approval of this ,
proposed attemative should be authorized pursuant to 10 CFR 50.55a(a)(3)(i)."
Evaluation-The Code requires the removal of allinsulation from pressure-retaining bolted connections in systems borated for the purpose of controlling reactivity when performing VT-2 visual examinations during system pressure tests. As an alternative, the licensee has proposed to perform the VT-2 visual examination required by the Code with the insulation in place. In addition, a separate, direct visual examination with the insulation removed but without pressurizing the system will be performed during each refueling outage for Class 1 components and overy 40 months for Class 2 and 3 components, except when a reactor refueling outage is in progress or is scheduled to start within six months when the 40 -
months expires. In that case, insulation removal and subsequent VT-2 examination will be deferred until prior to plant startup following the reactor refueling outage. The licensee's proposed alternative is essentially equivalent to Code Case N 533, Alternative Requirements for VT-2 Visual Examination of Class 1 Insulated Pressure-Retaining Bolted Connections,Section XI, Division 1, except the proposed alternative was extended to address Code Class 2 and 3 bolted connections. Code Case N-533 is currently under review by the NRC staff and has not yet been approved for use by incorporation into Regulatory Guide 1.147, inservice Inspection Code Case Acceptability.
The fi:ensee's proposed alternative provides a thorough approach to ensuring the leak-tight integrity of systems borated for the purpose of controlling reactivity. First, by perforrning the leakage test at operating pressure with the insulation in place, any significant leakage will be detected when the leakage either penetrates the insulation, or is detected ct joints or low points. Second, by removing the insulation each refueling outage for Class 1 components and at 40 month intervals for Class 2 and 3 components, the licensee will be able to detect minor leakage indicated by the presence of boric acid crystals or recidue. This two-phase approach will provide an acceptable level of quality and safety for bolted connections in borated systems.
Conclusion-Based on the evaluation above, it is concluded that the proposed alternative will provide an acceptable level of quality and safety. Therefore, it is recommended that the licensee's proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i) for use on Class 1,2, and 3 systems. The use of this alternative should be authorized for the second interval at VEGP, Units 1 and 2, or until Code Case N-533 is approved for general use by reference in Regulatory Guide 1.147. After that time, the licensee should follow the conditions,if any, specified in the regulatory guide.
48
- i. i 3.4.4.5 Proposed Altemative to use Code Case N 546, A/temative Requirements for QueRticetion of VT 2 Exemination Personnel Code Requirement-Section'XI, IWA-2300, requires that personnel performing VT-2 and VT-3 visual. examinations be qualified in accordance with comparable levels of competency as defined in ANSI N45.2.6. Additionally, the examination personnel shall have natural or l~
corrected near distance acuity, in at least one eye, equivalent to a Snellen fraction of 20/20. For far vision, personnel shall have natural or corrected far distance visual acuity lo of 20/30 or equivalent.
Licensee's Proposed Attemative-Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed to use Code Case N-546, Altemative Requirements for Qualification of VT-2 Examination Personnel,Section XI, Division 1, as an alternative to the ASME Section XI l- qualification requirements for VT-2 visual examiners. The licensee stated: ;
" Southern Nuclear Operating Company will comply with the requirements of Code Case N-546 with the exception of using IWA-2321 of the 1995 Editica for visual acuity. SNC is requesting to use IWA-2321 of the 1989 Edition as an alternative."
in the June 26,1998, submittal, the licensee stated that: 1 l
1
- 1) " Examination personnel will utilize existing VT-2 examination procedures which
, assures obtaining consistent and quality VT-2 visual examinations.
1
- 2) " Examination personnel qualifications will be documented and records will be' l maintained.
- 3) " Independent reviews and evaluations of detected leakage will be performed by persons other than those that performed the VT-2 visual examinations."
Licensee's Besis for Requesting ReWef-
" Code Case N-546 provides an alternative to IWA-2300 for qualification of VT-2 examination personnel. SNC reviewed Code Case N-546 and cetermined its
- implementation will substantially reduce the burdens required by IWA-2300 for qualification of VT-2 examination personnel.
- l. The ASME Code Committee evaluateri and approved Code Case N-546 as an
- y acceptable alternative for qualification of VT-2 examination personnel.- Code Case N-546 contains the following requirements for VT-2 examination personnel
R
$ 1. At least 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> plant walkdown experience, such as that gained by licensed and non-licensed operators, local leak rate personnel, system engineers, and inspection and nondestructive examination personnel.
- 2. - At least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of training on ASME Section XI requirements and plant specific procedures for VT-2 visual examination.
l
- 3. Vision test requirements of IWA-2321,1995 Edition.
49
"An exception is requested to the requirements of IWA-2321 of the 1995 Edition for the vision test requirements, it is requested that the requirements of IWA-2321 of the 1989 Edition be used in lieu of the 1995 Edition. The Second Ten-Year Intervals for VEGP-1 and 2 are required by 10 CFR 50.55a to be in compliance with the requirements of the 1989 Edition. The use of the 1.089 Edition is sufficient to assure the visual acuity of examination personnel and, as such, there is not need to impose the 1995 Edition on those examination personnel qualified to the requirements of ,
Code Case N-546. 4 a
"The implementation of Code Case N 546, including the exception of IWA-2321 of the 1995 Edition, will not affect the level of quality and safety, nor decrease the margin of public health and safety. Therefore,it is requested that the proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i)."
Evaluation-The Code requires that VT-2 visual examination personnel be qualified to levels of competency comparable to those identified in ANSI N45.2.6. The Code also requires that the examination personnel be qualified for near and far distance vision acuity. In lieu of the Code requirements, the licensee proposed to implement Code Case N-546 for personnel performing VT-2 visual examinations. This Code Case includes the following requirements:
- 1. At least 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> plant walkdown experience, such as that gained by licensed and nonlicensed operators, local leak rate personnel, system engineers, and inspection and nondestructive examination personnel.
- 2. At least four hours of training on Section XI requirements and plant specific procedures for VT-2 visual examination.
- 3. Vision test requirements of IWA-2321,1995 Edition.
The qualification requirements in Code Case N-546 are not significantly different from those for VT-2 vi"Jal examiner Certification. Licensed and nonlicensed operators, local leak rate personnel, system engineers, and inspection and nondestructive examination personnel typically have a sound working knowledge of plant components and piping layouts. This knowledge makes them acceptable candidates for performing VT 2 visual examinations.
In addition to meeting the rcquirements contained in Code Case N-546, the licensee ,
has committed to use procedural guidelines for consistent, quality VT-2 visual '
examinations, verify and maintain records of the qualification of persons selected to perform VT-2 visual examinations, and perform independent reviews and evaluations of leakage found by a person (s) other than those that performed the VT-2 visual examination.
Regarding the use of the visual acuity requirements of the 1989 Coda, the INEEL staff concurs that the use of the currently approved Code is sufficient to assure the visual acuity of examination personnel and, there is no need to impose the 1995 Edition on those examination personnel qualified to the requirements of Code Case N-546. Based on a review of Code Case N-546 and the additional commitments made by the licensee, the INEEL staff believes that the proposed alternative to the Code requirements will provide an acceptable le%. of quality and safety.
50
- - . . - - - - . . . _ - . . - . . - . - - - - . . - . - - . . . - - . . - ~ . . . , . . -
V
- ConcAus/an--Based on the evaluation above, the INEEL staff concludes that the licensee's proposed alternative provides an acceptable level of quality and safety. Therefore, it is
- j. recommended that the licensee's request to implement Code Case N-546 with the ,
additional commitments be authorized pursuant to 10 CFR 50.55a(a)(3)(i). The use of this
- Code Case should be authorized for the second interval at VEGP, Ur.ds 1 and 2, or until the l
Code Case is approved for general use by reference in Regulatory Guide 1.147. After that )
L[ time, the licensee must follow the conditions, if any, specified in the regulatory guide. j r
! 3.5 General i l
-3.5.1 Ultrasonic Examination Techniques No relief requests L
3.5.2 ' Exempted Components I l
3.5.2.1 Request for Relief No. RR-22, Use of Code Case N-544, Repe/r amt Replacement .
of SmeNItems j l . . . .i Note: Request for Relief RR-22 was evaluated and authorized an NRC SER dated March :l 24,1998.
3.5.3' Other l
3.5.3.1 Request for Relief No. RR-1, Subsection IWA 2413, Proposed Alternative to )
Successive interval Scheduling j l
Code Requirement-Part J0-55a(g)(4)(ii) of Title 10 of the Code of Federal Regulations (CFR) states: " Inservice examination of components and system pressure tests conducted L .during successive 120-month . inspection intervals must comply with the requirements of the latest edition and addenda of the Code incorporated by reference in paragraph (b) of this section 12 months prior to the start of the 120-month inspection interval, subject to the limitations and modifications listed in paragraph (b) of this section".
f4 j Paragraph IWA-2413, " Successive Inspection intervals", of the 1989 Edition of ASME Section XI states: "The inspection plan for each successive inspection interval shall comply 9 '. with the Edition and Addenda of this Division that has been adopted by the regulatory authority.12 months prior to the start of.the inspection interval, or subsequent Editions and l ' Addenda that have been adopted by the regulatory authority. Specific portions or such l : subsequent Editions or Addenda may be used provided all related requirements are met".
i l Licensee's Proposed Attemetive--in accordance with 10 CFR 50.55a(a)(3)(i), the licensee
' proposed to start the VEGP-2 Second Ten-Year interval (and subsequent intervals, for the i remainder of the plant life) ahead of schedule such that the two VEGP units will be under
- the same edition and/or addenda of the Code, i.e., the 1989 Edition of ASME Section XI.
51
Further, an enemption is requested from the requirements of 10 CFR 50.55a(g)(4)(ii) for updating the VEGP-2 ISI program at its normal update time, i.e., May 20,1999, for its Second Ten-Year Interval. The licensee stated:
" Alternate examinations are not applicable to this request for relief involving the early update of the VEGP-2 ISI Program and exemption from updating at the normal time for VEGP-2. As noted in the Georgia Power Company letter of August 16,1996, the .
remaining examinations for the final two years of the current interval for the VEGP-2
- ISI Program will be rescheduled for the first two years of the new ten-year interval "'
which will be concurrent with the update for VEGP- 1. These examinations will continue to be scheduled as required by ASME Section XI Paragraph IWA 2413."
Licensee's Basis for Proposed Attemative-
"By letter LCV-0861 dated August 16,1996, Georgia Power Company (the former licensee for VEGP and sister company to Southern Nuclear Operating Company, the current licensee and operator of VEGP) requested NRC concurrence to update the VEGP-2 ISI program for the Second Ten-Year Interval approximately two years ahead of schedule.
"The early update of the VEGP-2 ISI program would be coincident with the update for VEGP-1 and would allow the two VEGP units to be performed to the same edition and/or addenda of the Code. Otherwise, a different edition and/or addenda of the Code would have to be used for the ISI programs for each unit at VEGP because of the differences in commercial operation dates on which the ISI program interval dates are based. Specifically, the ISI program for VEGP-1 would be performed to the 1989 Edition of ASME Section XI while that for VEGP-2 would continue to be performed to the 1983 Edition of ASME Section XI with Addenda through Summer 1983 untilits ISI program was updated in 1999. The commercial operation dates for VEGP-1 and 2 are May 31,1987 and May 20,1989, respectively. In its August 16,1996 letter to the NRC, Georgia Power Company indicated that a formal request for relief would be submitted requesting permission to update the VEGP-2 ISI program concurrent to that for VEGP-1 and requesting exemption from having to update VEGP-2 at its normal update time.
"By concurrently updating the VEGP-1 and 2 ISI programs, a more comprehensive examination will be achieved which will enhance the possibility of detecting a generic i
problem and will also reduce costs involved with maintaining two separate lui
! programs should the use of different Code editions /addenoa be required. In addition, e use of the same edition and/or addenda of the Code for the two units would help
- prevent possible errors associated with maintaining two separate programs with different requirements.
"The practice of early updates has been used by Georgia Power Company since 1985 with the updates at Hatch Nuclear Plant, Unit 2, for the Second Ten-Year interval and most recently for the Third Ten-Year Interval update.
"In its November 27,1996 response to the Georgia Power Company letter, the NRC indicated its concurrence with the concurrent update of the VEGP-2 ISI program with that for VEGP-1. Further, it was indicated that the applicable ASME Code, i.e., the 52
t-L 1989 Edition, for the ISI program during the adjusted interval for VEGP-2 was acceptable in accordance with 10 CFR 50.55a(g)(4). !
l "By concurrently updating the VEGP-1 and 2 ISI programs, a more comprehensive examination will be achieved which will enhance the possibility of detecting a generic problem and will also reduce costs involved with maintaining two separate ISI programs should the use of different Code editions / addenda be required in addition,
[ use of the same edition and/or addenda of the Code for the two units would help i P prevent possible errors associated with maintaining two separate programs with b different rer Mrements. Examinations which remain to be performed to complete the i First Ten-Year Interval on VEGP-2 will be rescheduled such that they are performed during the first two years of the adjusted interval on that plant unit. This will help ensure that no more than ten years will elapse between examinations.
1 i
Southern Nuclear Operating Company requests that relief and the exemption discussed herein be authorized pursuant to 10 CFR 50-55a(a)(3)(i) since an acceptable level of quality and safety will have been achieved and public health and
- safety will not be endangered."
In the -June 26,1998, letter, the licensee stated: )
! "As indicated in Request for Relief RR-1, the VEGP-2 examinations which complete the First Ten-Year Interval will remain as scheduled such that they are performed ,
during the first two years of the adjusted interval. Because of the early update, the 1 2R6 Maintenance / Refueling Outage becomes the first outage of the Second Ten-year Interval for VEGP-2. None of the examinations performed in 2R6 will be counted toward the Second Ten-Year examination requirements, except in the case of ASME Code Categories B-G-1, B-G-2, B-l.-2 and B-M-2. The examinations for those l
particular Code category components are required only once during the inspection interval when a component, e.g., a pump or valve, is disassembled for maintenance or
- repair. As allowed by the Code, such examinations are limited to at least one pump in each group of valves that are of the same size, construction design, and 1 manufacturing method that perform similar functions in the system. Examinations and tests scheduled for 2R2 were performed using the requirements of the 1989 Edition of ASME Section XI (to the extent practical).
"There are seven maintenance / refueling cutages scheduled for VEGP-2 between May' 31,1997 and May 30,2007. The Second Ten-Year examinations will be scheduled in the six remaining maintenance / refueling outages,2R7 through 2R12. The RPV b
examinations will be performed in the third period of the interval provided that the k NRC approves Request for Relief RR-2 for use. During 2R6, one hundred percent L (100%) of the RPV welds were nondestructively examined since we wished to use ASME Section XI Code Case N-521 (as presented in RR-2) for the second ten-year inspection interval. A 100% RPV examination was performed during VEGP-1 Outage 1R6 for the same reason. If Request for Relief RR-2 is approved by the NRC, all RPV
- . examinations for VEGP-2 will be performed in 2R12. This falls well within the ten-l year requirement for performing the RPV examinations.
l- " Credit for either the First or the Second Ten-year interval will be so noted in the I
53
L f I VEGP 2 Outage 2R6 " Owner's Report for Inservice inspection (ASME Form NIS-1',
which is to be submitted to the NRC by July 17,1998. Thus, examinations originally
! scheduled for the end of the First Ten-Year interval have been completed and are ,
being documented in the aforementioned report. Those examinations will be repeated in approximately ten years, as allowed by the Code. The sequence of examinations will not be sufficiently altered from the First ten-Year Interval to the second Ten-Year Interval. The First Ten-Year interval requirements and commitments have been met for VEGP-2 and we will meet the necessary requirements and commitments for the Second Ten-Year Interval for that unit."
Evaluation-The Regulations require that licensee's prepare their ISI Program to meet the requirements of the latest Code edition in effect 12 months prior to the start of the120 month inspection interval. By letter dated August 16,1996, the licensee requested concurrence from the NRC to update VEGP, Unit 2, two years ahead of schedule so that Unit 2 updates would coincide with Unit 1 updates. The NRC staff found this request acceptable in accordance with 10 CFR 50.55a(g)(4). Subsequently, the license has e requested an exemption from the requirements of 10 CFR 50.55a(g)(4)(ii) for updating the VEGP-2 ISI program at its normal update time in May of 1999 so that the Unit 1 and Unit 2 intervals will continue to coincide. In the June 26,1998, response to the NRC RAI, the licensee confirmed that all the requirements for both the first and second intervals would
- be met.
By aligning the intervals for Units 1 and 2, the licenses can perform ISI using the same Code Edition for both units. Since the two units are nearly identical, this will improve the efficiency of ISI performed at VEGP. Furthermore, since the overall number of examinations performed will not be reduced, the licensee will maintain an acceptable level
- of quality and safety for the plant.
I Conclusion-Based on the evaluation above, it is concluded that the licensee's proposed alternative provides an acceptable level of quality and safety. Therefore, it is recommended that the proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i). -
3.5.3.2 Request for Relief No. RR-20, Use of Code Case N 509, A/temative Rules for the Selection and Examination of Class 1, 2, and 3 Integrally Welded Attachments i
Note: Request for Relief RR-20 was evaluated and authorized in an NRC SER dated March 24,1998. # -
i
~3.5.3.3 Request for Relief No. RR-25 (Revision 9/17/97), lWA-5250(a)(2), Corrective ,
a Actions for Bolted Connections .
l Notei Request for Relief RR-25 was evaluated and authorized an NRC SER dated October 24,1997.
3.5.3.4 ' Request for Relief No. RR-28, Uce of Code Case N-508-1, #otation of Serv /ced
. Snubbers and Pressure ReWef Valves for the Ptsrpose of Testing ,
! Note: Request for helief RR-28 was evaluated and authorized an NRC SER dated l 54 i
December 9,1997.
l 3.5.3.5 Request for Relief No. RR-29, IWF-5000, inservice inspection and Testing of I Hydraulic and Mechanical Snubbers f
l Note: As previously stated, snubbers are evaluated elsewhere.
t 3.5.3.6 Request for Relief No. RR-30, Use of Code Case N-532, Attemative Requirements to Repair and Replacement andinservice Summary Report y
I Preparation and Submission as Required by IWA-4000 and IWA-6000 Code Requirement-Section XI, Paragraph IWA-6220 requires that the licensee prepare rt. ports using NIS-1, Owner's Report for Inservice inspections, and NIS-2, Owner's Report for Repair or Replacements; IWA-6230 requires that these reports be filed with the j
- enforcement and regulatory authorities having jurisdiction at the plant site within 90 days 1 of the completion of the inservice inspection conducted during each refueling outage.
i Licec te's Proposed Attemative--In accordance with 10 CFR 50.55a(a)(3)(i), the licensee proposed to use Code Case N-532, Altemative Requirements to Repair and Replacement and Inservice Summary Report Preparation and Submission as Required by IWA-4000 and l IWA-6000,Section XI, Division 1, in lieu of the Code reporting requirements. The licensee L stated:
i 1 l " Southern Nuclear Operating Company will comply with the requirements of Code I L
Case N-532, with the following clarification regarding reporting of corrective measures. Code Case N-532, paragraph 2.0(c), requires an abstract for repairs, replacements, and corrective measures required due to an item containing a flaw or relevant condition exceeding the acceptance criteria of ASME Code Section XI.
According to Section XI, the term, ' corrective measures' has two applications. One application involves repair and replacement activities on pressure-retaining components (e.g., metal removal and welding). The other application involves maintenance-type activities, such as tightening of bolting, replacing gaskets / packing, cleaning surface corrosion products, and adjusting component supports. For Code Case N-532 reporting, SNC considers ' corrective measures' to involve only repair and l replacement activities.
Licensee's Basis for Proposed Attemative-h4 "Previously, voluminous NIS-1 Owner's Data Reports for Inservice inspection were submitted which included the NIS-2 Owner's Reports for Repairs and Replacements.
9- The process of preparing, reviewing, and submitting the rer. orts within the 90-day P~ time requirement often resulting in difficulties for the licensee's staff. Implementation of Code Case N-532 allows the submittal of abstracts versus the reports required by l
the ASME Code Section XI. In addition, the implementation of tisat Code Case is consistent with the NRC's philosophy found in NRC letter SECY-94-093 dated May 10,1995. Per SECY-94-093, the NRC is to take a proactive role through its
' representetives in the ASME Code to modify reporting requirements and to eliminate I- the need to submit inservice inspection reports following each refueling outage.
i "The ASME Code Committee evaluated the proposed attornative reporting l-55
requirements and determined that the requirements of Code Case N 532 are acceptable for replacing the existing documentation and reporting requirements.
Since Code Case N-532 only affects documentation and reporting requirements,its implementation will not affect the level of quality and safety, nor decrease the margin of. public health and safety. While the cost savings associated with Code Case N-532 ;
have not been quantified as a Cost Beneficial Licensing Action item, its implementation is consistent with the intent to eliminate work activities which are not 'i beneficial and costs therefore. Therefore, it is requested that the proposed alternative
~
be authorized pursuant to 10 CFR 50.55a(a)(3)(i). ".
Evaluation-The use of Form NIS-1, Owner's Report For /nservice /nspections, and Form '
NIS-2, Owner's Report for Repairs or Rep /acemersts, and submittal of the 90-day Summary Report are Code requirements. Alternatives contained in Code Case N-532 allow the
. licensee to submit these records in an abstract format on Form NIS-2A, Repair / Rep /acement Certification Record, and Form OAR-1, Owner's Activity Report, following the completion of an inspection period.
The requirements associated with documentation of inservice examinations and ,
repairs / replacements and the subsequent submittal of Forms NIS-1 and NIS-2 within 90
. days following a re'ueling outage are administrative only. It is noted that repair and replacement docur mtation reviews and approvals by the Authorized Nuclear inspector continue to be re ed by this Code Case and that the licensee is required to establish a Repair /Replaceme . Plan in accordance with IWA-6340 of the 1992 Edition of Section Xf.
The licensee has implemented /nspection Program B of the Code. Under this program, ,
examination schedules are satisfied on a "per period" basis. Considering the milestones associated with /nspection Program 8, submittal of the results of examinations and an abstract of repairs / replacements on a periodic basis is a reasonable alternative. In addition, the INEEL staff believes that the forms contained in Code Case N-532, which '
provide a' summary of the status of repairs / replacements and a more detailed status of examinations by period and interval, are an improvement over report forms currently f
required by the Code. For example, OAR 1 includes the status of examinations credited for the period and percent credited to date for the interval, by Examination Category. This type of information provides the regulatory authorities a more comprehensive report on the status of the inservice inspection program.
Conclus/on-Considering the administrative nature of the Code recording and reporting ,
criteria, the INEEL staff believes that use of the altematives to Code requirements .
contained in Code Case N-532 will continue to provide an acceptable level of quality and I' safety. Therefore, it is recommended that the licensee's proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i). The use of alternatives contained in Code '
Case N-532 should be authorized for the current interval or until such time as the Code
- . Case is published in a future revision of Regulatory Guide 1.147. At that time, if the L
, licensee l~ntends to continue to implement the alternatives of this Code Case, the licensee is to follow all provisions in Code Case N-532 with limitations issued in Regulatory Guide l
! , 1.147, if any.
l:'
56
- 4. CONCLUSION Pursuant to 10 CFR 50.55a(g)(6)(i), it has been determined that certain inservice examinations cannot be performed to the extent required by Section XI of the ASME Code.
In the cases of Relief Requests RR-4 (Parts 1 and 2), RR-7, RR-10, RR-11, RR-12 (Parts 1 and 2), RR-14, and RR-16, the licensee has demonstrated that specific Section XI 4 requirements are impractical. It is, therefore, recommended that relief be granted as j requested. For Request for Relief RR-6, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i) provided that the licensee meets the conditions stated in the evaluation. Granting relief will not endanger life, property, or the common defense and security and is otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
Pursuant to 10 CFR 50.55a(a)(3), it is concluded that for Relief Requests RR-1, RR-2, RR-9, RR-26, RR-30, RR-31, and implementation of Code Case N-546, the licensee's proposed alternatives will (a) provide an acceptable level of quality and safety, or (b) Code compliance will result in hardship or unusual difficulty without a compensating increase in safety, it is recommended that the proposed alternative be authorized.
For Relief Request RR-18, the licensee did not provide adequate justification to support the determination that the proposed alternative provides an acceptable level of auality and safety. Therefore, it is recommended that the proposed alternative not be authorized.
By letter dated June 26,1998, the licensee withdrew Relief Requests RR-3, RR-8 and RR-15 and deleted them from the ISI Program Plan.
Requests for Relief RR-5, RR-17, RR-19, RR-20, RR-21, RR-22, RR-23, RR-24, RR-25, RR-28 and RR-29 have been or will be evaluated elsewhere.
This technical evaluation has not identified any practical method by which Southern Nuclear Operating Company can meet all the specific inservice inspection requirements of Section XI of the ASME Code for the existing Vogtle Electric Generating Plant, Units 1 and
- 2. Compliance with all of the Section XI examination requirements would necessitate redesign of a significant number of plant systems, procurement of replacement components, installation of the new components, and performance of baseline examinations for these components. Even after the redesign efforts, complete compliance 3 with the Section XI examination requirements probably could not be achieved. Therefore, g it is concluded that the public interest is not served by imposing provisions of Section XI of the ASME Code that have been determined to be impractical.
Southern Nuclear Operating Cornpany should continue to monitor the development of new or improved examination techniques. As improvements are achieved, Southern Nuclear Operating Company should incorporate these techniques in the ISI program plan j
examination requirements.
Based on the review of the Vogtle Electric Generating Plant Second Ten-YearInterval Inservice Inspection Program, the licensoe's response to the NRC's request for additional information, and the reccmmendations for granting relief from the ISI examinations that 57
I cannot be performed to the extent required by Section XI of the ASME Code, no deviations
- from regulatory requirements or commitments were identified except for the scheduling of Examination B-G-1, RPV bolting, percentages of certain items that do not appear to meet the minimum sample sizes of the Code, and the incorrect exemption of certain Class 2 vessets as described in Section 2.2 of this document, P-o j
.)
l
?
h
.4 4
58 P ____ _ _ __ _ _ _ ______ _ ______. __
- 5. REFERENCES i
- 1. Code of Federal Regulations, Title 10, Part 50.
=
- 2. American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Division 1,1989 Edition.
4
- f' 3. Vogtle Electric Generating Plant Second Ten-YearIntervalInservice Inspection Program submitted May 29,1997.
- 4. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Section 5.2.4, " Reactor Coolant Boundary inservice inspection and Testing," and Section 6.6, " Inservice Inspection of Class 2 and 3 Components,"
July 1981.
- 5. Letter dated June 26,1998, J. D. Woodward (SNC) to Document Control Desk (NRC), containing response to the NRC RAl.
- 6. Letter, dated August 14,1998, J. B. Beasley (SNC) to NRC Document Control Desk, containing two requests for additionalinformation for the second 10-year ISI Program.
- 7. NRC Regulatory Guide 1.147, inservice Inspection Code Case Acceptability, Revision 11, October 1994.
- 8. NRC Regulatory Guide 1.14, Reactor Coolant Pump flywheelintegrity, Revision 1, dated August 1975.
^ 9. NRC Regulatory Guide 1.150, Ultrasonic Testing of Reactor Vessel Welds During i Preservice andinservice Examinations, Revision 1, February 1983.
l h
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l l
l I
a
- 59 i
l
1:RC Form 335 U.S. Nuclear Regulatory Commusion 1, REPORTNUMBER NPCM i102 (Assigned by NRC, Add Vol., Supp., Riv., and 320ia202 "*""" '"#'
HIBLIOGRAPIIIC DATA SIIEET INEEllEXT-98-01007
- 2. 'ITILE AND SUBTTTLE 3. DATE REPORT PUBLISIIED Technical Evaluation Report on the Second 10-Year Interval Month Year Inservice Inspection Program Plan:
Southern Nuclear Operating Company, November 1998 Vogtle Electric Generating Plant, Units I and 2,
, Docket Numbers 50-424 and 50-425 JCN J2229 (Task Order A28)
[
A 5. AUTIIOR(S) 6. TYPE OF REPORT M. T. Anderson Technical C. T. Brown S. G. Galbraith 7. PERIOD COVERED (inclusive Dates)
A. M. Porter
- 8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division,00 ice or Region, U.S. Nuclear Regulatory Commission, and mailing addren; if contractor, provide name and mailing address)
Idaho National Engineering and Environmental laboratory Materials Phy:ics Lockheed Idaho Technologies Company Idaho Falls, Idaho 83415 l l
- 9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type Same as above"; if contractor, provide NRC Ditision, Odice or Region, U.S. I Nuclear Regulatory Commission, and mailing addren)
Civil and Geosciences Branch Division of Engmeering Oflice of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission - ;
Washington,D.C. 20555 1 1
- 10. SUPPLEMENTARYNOES
- 11. ABSTRACT (200 Wards or less)
This report presents the results of the evaluation of the Vogtle Electric Generating Plant Second Ten-Year Interval Inservice Inspection Program submitted May 29,1997, including the requests for relief from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, requirements that the licensee has determined to be impractical. He Vogtle Electric Generating Plant Second Ten-Year Interval Inservice Inspection Program is evaluated in Section 2 of this report. He inservice inspection (ISI) plan is evalua'.11 for (a) compliance with the appropriate edition / addenda of Section XI,(b) acceptability of exammation sample,(c) correctness of the application of system or component examination exclusion criteria, and (d) compliance with ISI-related commitments identified during previous Nuclear Regulatory Commission reviews. The requests for relief are evaluated in Section 3 of this report.
t 12. KEY WORDS/DESCRIPTORS (114 wants or phrases that will amist researchers in locating the report) 13. AVAILABILITY STATEMENT I Unlimited
- e
- 14. SECURfrY CLASSIFICATION (This page) Unclassified (This report) Unclassified
- 15. NUMBER OF PAGES l
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- b. Raouest for Relief No. RR-21. lWC-1220. Class 2 Comoonents Exemot from l Examination Code Raouirement: IWC-1220, Components Exempt from Examination, contains the exemption criteria for Class 2 components. IWC-1222 contains the requirements for components within systems other than the Residual Heat Removal (RHR), Emergency Core Cooling (ECC), and Containment Heat Removal (CHR) systems. In accordance l
with IWC-1222, components exempt from the surface and volumetric examination requirements of IWC-2500 are as follows:
(a) Vessels, piping, pumps, valves, and other cernponents NPS 4 and smaller.
(b) Component connections NPS 4 and smaller (including nozzles, socket fittings, and other connections) in vessels, piping, pumps, valves, and other components of any size.
(c)' Vessels, piping, pumps, valves, and other components of any size in systems or portions of systems that operate (when the system function is required) at a pressure equal to or less than 275 psig and at a temperature equal to or less than 200'F.
(d) Piping and other components of any size beyond the last shutoff valve in open ended portions of systems that do not contain water during normal plant operating conditions.
Licensee's Pronosed Alternative: Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee has - -
proposed to use the exemption criteria found in IWC-1222 of the 1989 Addenda of the 1989 Edition of ASME Section XI for components within systems other than the RHR, ECC, and CHR systems. This Addenda does not require surface and volumetric examinations of vessels and their connections in piping 4 inch nominal pipe size (NPS) and smaller for Examination Category C-A, items C1.10, C1.20, and C1.30 (Pressure-Retaining Welds in Vessels) and Examination Category C-C, item C3.10 (Integrslly Welded Attachments to Pressure Vessels). Relief is specifically requested for the _
following components:
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l Reaenerative Heat Exchanner (Tan Nos.1- & 2-1208-E6-001)
Excess Letdown Heat Exchanner(Tan Nos 1- &. 21208-E6-002)
Letdown Heat Exchanner(Tan Nos. 6 & 2-1208-E6-003)
Letdown Reheat Heat Nxchanaer (Tan Nos. 1- & 21208-E6007)
Suction Damner (Tao Nos. 1-& 2- 1208 V4-001)
Discharae Damnener(Tao Nos. 1- & 2-1208-V4-0021 l The licensee stated: .
"These exemptions exclude the applicable vessels from the surface and volumetric !
examinations required by IWC 2500. The remainder of the Code-required examinations (i.e., pressure tests) would be performed to assure that an acceptable level of safety i l and quality is maintained for the applicable components." i L
- Licensee's Basis for Reauestino Relief (as stated):
"Subarticle IWC-1220 of the 1989 Addenda of ASME Section XI allowed the exemption of selected components from the surface and volumetric examination requirements of IWC-1220 [ Table IWC-2500-1]. The 1996 Addenda of ASME Section XI a!ao includes these exemptions in IWC-1220. The NRC granted these exemptions to VEGP in the L first interval through correspondence dated March 8,1996 and August 13,1996 for j VEGP-1 and 2, respectively.
! "These exemptions will be allowed when the newer Addenda and Editions of the Code are authorized in 10 CFR 50.55a. SNC sees no benefit in performing examinations on components which the Code has determined can be exempted. The other requirements l in the Code are therefore acceptable to assure an acceptable level of safety or quality, i lt is impractical to perform examinations which do not provide a compensating increase l in the level of safety or quality. i "These added exemptions would apply to several components which are in high dose l
i rate areas. The most signifbant of these components is the regenerative heat l exchanger. A conservative whole body dose in the range of one to two Rem is a L reasonable estimate for examining the regenerative heat exchanger. The dose rate L surveys for the regenerative heat exchanger indicate a contact dose rate of two to three l Rem / hour and a dose rate at eighteen inches away from the heat exchanger of one to l one and-one-half (1 to 1-1/2) Rem / hour. The estimated stay time to perform the Code-required examinations on the regenerative heat exchanger is one hour. Guch exposure is contrar/ to the principles of ALARA to perform examinations on components.without !
L a compensating increase in safety or quality. For the reasons discussed above, SNC I
has determined that implementation of the Code requirements is impractical. Therefore, SNC requests that the proposed alternative be authorized pursuant to 10 CFR ~ ~
50.55a(a)(3)(i).
Evaluation: The licensee has requested to use the exemption criteria of IWC-1222 of 3
the 1989 Addenda in lieu of the exemption requirements of the Code of record. In t
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accordance with the 1989 Code, piping NPS 4 and smaller is exempt from examination, but connected components are not. In the 1989 Addenda of Section XI, lWC-1222 was revised to exempt vessels, pumps and valves, and their connections in piping NPS 4 and smaller, with the following note. "/n piping is defined as having a cumulative inlet and a cumulative outlet pipe cross sectional area neither of which exceeds the nominal OD cross sectional area of the designated size." This exemption is also contained in Code Case N-408-2, Altemative Rules for Examination of Class 2 Piping, 3 Section X/, Division 1, which has been approved for general use in Revision 11 of Regulatory Guide 1.147, inservice Inspection Code Case Acceptability - ASME Section XI, Division 1.
The change in the Code described above parallels the logic used for the exemption of Class 1 systems. Specifically, IWB-1220(b)(2) exempts " components and their connections in piping in 1-inch nominal pipe size and smaller", where "in piping" is defined as having one inlet and one outlet pipe, each of which is 1-inch NPS or smaller.
The discrepancy between Class 1 and 2 systems was recognized by the Code committee, which patterned the exemption criteria for Class 2 in the 1989 Addenda after existing exemption requirements for Class 1 systems.
The INEEL staff has reviewed this request and concludes that the licensee's alternative, to use the exemption criteria of the 1989 Addenda for the above specified Class 2 systems, will provide an acceptable level of quality and safety. The approach used for the Class 2 exemption criteria, found in the 1989 Addenda is similar to that used for exemption of Class 1 systems, in addition, the criteria has been approved by the NRC as part of Code Case N-408-2. Therefore, it is recommended that the licensee's proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i).
- c. Reauest for Relief RR-22. Use of Code Case N-544. Reosir and Reo/acement of Sma//
Items.Section XI. Division 1 Code Reauirement: The Code requires that repairs and replacements be performed in accordance with IWA-4000 and IWA-7000, respectively.
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