L-95-050, Technical Evaluation Rept on Second 10-Yr Interval ISI Program Plan:Sce,San Onofre Nuclear Generating Station, Units 2 & 3,Docket Numbers 50-361 & 50-362
| ML20097F968 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 11/01/1995 |
| From: | Mary Anderson, Feige E, Hall K IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | NRC (Affiliation Not Assigned) |
| Shared Package | |
| ML20097F955 | List: |
| References | |
| CON-FIN-L-2556 INEL-95-0503, INEL-95-503, NUDOCS 9602200296 | |
| Download: ML20097F968 (32) | |
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- I t Attachment 1 l INE-95/0503 l
l Technical Evaluation Report on the Second 10-year Interval inservice inspection Program Plan-Southem California Edison Company, I San Onofre Nuclear Generating Station, Units 2 and 3, Docket Numbers 50-361 and 50-362 l l
M. T. Anderson I E. J. Feige K. W. Hall I
l Published November 1995 1
I Idaho National Engineering Laboratory Materials Physics Lockheed Idaho Technologies Company Idaho Falls, Idaho 83415 Prepared for tf m
. Division of Engineering office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Under DOE Idaho Operations Office Contract DE-AC07-941D13223 FIN No. L2556 (Task Order 43) 9602200296 960213 PDR ADOCK 05000361 p PDR
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i ABSTRACT This report presents the results of the evaluation of the San Onofre Nuclear '
Generating Station, Units 2 and 3, Second 10-Year Interval Inservice Inspection Program Plan, Revision 3, submitted October 4,1993, including the requests for relief from the American Society of Mechanical Engineers Boiler 1 -
and Pressure Vessel Code Section XI requirements that the licensee has
' ' determined to be impractica). The San Onofre Nuclear Generating Station, Units 2 and 3, Second 10-Year Interval Inservice Inspection Program Plan is
! evaluated in Section 2 of this report. The ISI Program Plan is evaluated for l (a) compliance with the appropriate edition / addenda of Section XI, (b) acceptability of examination sample, (c) correctness of the application of system or component examination exclusion criteria, and (d) compliance with ISI-related commitments identified during previous Nuclear Regulatory Commission (NRC) reviews. The requests for relief are evaluated in Section 3 of this report.
This work was funded under:
U.S. Nuclear Regulatory Commission FIN No. L2556, Task Order 43 Technical Assistance in Support of the NRC Inservice Inspection Program 11
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SUMMARY
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, The licensee, Southern California Edison Company, has prepared the San Onofre Nuclear Generating Station, Units 2 and 3, Second 10-Year Interval Inservice
, Inspection Program Plan, Revision 3, to meet the requirements of the 1989 Edition of the American Society of Mechanical Engineers (ASME) Boiler and 3
Pressure Vessel Code Section XI. For both Units 2 and 3, the second 10-year
- interval began on April 1, 1994 and ends on August 17, 2003.
l The information in the San Onofre Nuclear Generating Station, Second 10-Year l Interval Inservice Inspection Program Plan, Revision 0, submitted October 4, 1993, was reviewed. Included in the review were the requests for
! relief from the ASME Code Section XI requirements that the licensee has
- determined to be impractical. As a result of this review, a request for additional information (RAI) was prepared describing the information and/or i clarification required from the licensee in order to complete the review. The licensee provided the requested information in the San Onoftw #uclear Generating Station, Second 10-Year Interval Inservice Inspection Program Plan, l Revision 3, submitted April 17, 1995.
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. Based on the review of the San Onofre Nuclear Generating Station, Units 2
- and 3, Second 10-Year Interval Inservice Inspection Progran Plan, Revision 3,
! the licensee's response to the Nuclear Regulatory Commission's RAI, and the l recommendations for granting relief from the ISI examinations that cannot be
- performed to the extent required by Section XI of the ASME Code, no deviations from regulatory requirements or commitments were identified in the San Onofre l Muclear Generating Station, Units 2 and 3, Second 10-Year Interval Inservice l Inspection Program Plan, Revision 3, except those noted in Sections 2.2.2 and
- 2.2.3 and in the evaluations of Requests for Relief 3.3.5 and 3.3.6.
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1 t' l CONTENTS ABSTRACT ................................ ii
SUMMARY
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- 1. INTRODUCTION ............................ 1
- 2. EVALUATION OF INSERVICE INSPECTION PROGRAM PLAN . . . . . . . . . . . 4 2.1 Documents Eval uated . . . . . . . . . . . . . . . . . . . . . . . . 4 2.2 Compliance with Code Requirements ................ 4 2.2.1 Compliance with Applicable Code Editions . . . . . . . . . . . 4 l 2.2.2 Acceptability of the Examination Sample ........... 4 2.2.3 Exemption Criteria . . . . . . . . . . . . . . . . . . . . . . 5 2.2.4 Augmented Examination Commitments .............. 5 2.3 Conclusion . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
- 3. EVALUATION OF RELIEF REQUESTS . . . . . . . . . . . . . . . . . . . . 7 3.1 Class 1 Components . . . . . . . . . . . . . . . . . . . . . . . . 7 3.1.1 Reactor Pressure Vessel ................... 7 3.1.1.1 Request for Relief 3.3.1, Examination Categories B-A and B-D, Items Bl.30, B3.90, and B3.100, Reactor Vessel Shell-to-Flange Weld, Nozzle-to-Vessel Welds, and Nozzle Inside Radius Sections ............ 7 I i
3.1.2 Pressurizer i
......................... 9 3.1.2.1 Request for Relief 3.3.4, (Part 1), Examination Category B-D, Item B3.110, Pressurizer Nozzle-to- l Ve s sel Wel d s . . . . . . . . . . . . . . . . . . . . . . . 9 3.1.3 Heat Exchangers and Steam Generators . . . . . . . . . . . . . 11 3.1.3.1 Request for Relief 3.3.4, (Part 2) Examination Category B-D, Item B3.130, Steam Generator Nozzle-to-l Ve s sel Wel d s . . . . . . . . . . . . . . . . . . . . . . . . 11 3.1.4 Piping Pressure Boundary . . . . . . . . . . . . . . . . . . . 13 3.1.4.1 Request for Relief 3.3.2, Examination Category B-J, Items B9.11 and 89.12, Circumferential and '
Longitudinal Welds in Class 1 Pressure Retainin Piping . . . . . . . . . . . . . . . . . . . . g ...... 13 3.1.5 Pump Pressure Boundary (No requests for relief) iv
i 3.1.6 Valve Pressure Boundary ................... 15 3.1.6.1 Request for Relief 3.3.3, Examination Category B-M-1, Item B12.40, Pressure Retaining Welds in Valve Bodies .. 15 3.1.7 General (No requests for relief) j -
3.2 Class 2 Components . . . . . . . . . . . . . . . . . . . . . . . . 17 3.2.1 Pressure Vessel s . . . . . . . . . . . . . . . . . . . . . . . 17
. 3.2.1.1 Request for Relief 3.3.6, Examination Categories C-A,
' C-C, and F-A, Items C1.20, C1.30, C3.10, and F1.40, Regenerative Heat Exchanger Examinations . . . . . . . . . 17 3.2.2 Piping (No requests for relief) 3.2.3 Pumps (No requests for relief) 3.2.4 Valves (No requests for relief) 3.2.5 General (No requests for relief) 3.3 Class 3 Components (No requests for relief) 3.4 Pressure Tests (No requests for relief) 3.5 General ............................. 19 3.5.1 Ultrasonic Examination Techniques (No requests for relief) 3.5.2 Exempted Components (No requests for relief) 3.5.3 Other ............................. 19 3.5.3.1 Request for Relief 3.3.5, Paragraph IWA-2312, Certification and Recertification ............ 19 3.5.3.2 Request for Relief 3.4, Paragraph IWA-4511, General Requirements for Repair Welding Using Half Bead Welding Technique .................... 21
- 4. CONCLUSION ............................. 23
- 5. REFERENCES ............................. 25 v
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l TECHNICAL EVALUATION REFORT ON THE SEC05 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAN PLAN:
SOUTHERN CALIFORNIA EDISON COMPANY, SAN ON0FRE NUCLEAR GENERATING STATION, UNITS 2 AND 3, DOCKET NUN 8ERS 50-361 and 362
- 1. INTRODUCTION l
Throughout the service life of a water-cooled nuclear power facility, 10 CFR 50.55a(g)(4) (Reference 1) requires that components (including supports) that are classified as American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Class 1, Class 2, and Class 3 meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code Section XI, Rules for Inservice inspection of Nuclear Power Plant Components (Reference 2), to the extent practical within the limitations of design, geometry, and materials of construction of the components. This section of the regulations also requires that inservice examinations of components and system pressure tests conducted during successivs 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 to Nuclear Regulatory Commission (NRC) approval. The licensee, Southern California Edison Company, has prepared the San Onofre Nuclear Generating Station, Units 2 and 3, Second 10-Year Interval inservice Inspection Program Plan, Revision 3, (Reference 3) to meet the requirements of the 1989 Edition of the ASME Code Section XI. The second 10-year interval began April 1, 1994 and ends August 17, 2003.
As required by 10 CFR 50.55a(g)(5), if the licensee determines that certain l ,
Code examination requirements are impractical and requests relief from them, l the licensee shall submit information and justification to the NRC to support that determination.
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j e l Pursuant to 10 CFR 50.55a(g)(6), the NRC will evaluate the licensee's determination that Code requirements are impractical to implement. The NRC may grant relief and may impose alternative requirements that are determined to be authorized by law, will not endanger life, property, or the common s
defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
Alternatively, pursuant to 10 CFR 50.55a(a)(3), the NRC will evaluate the licensee's determination that either (i) the proposed alternatives provide an acceptable level of quality and safety, or (ii) Code compliance would result in hardship or unusual difficulty without a compensating increase in safety.
Proposed alternatives may be used when authorized by the NRC.
The information in the San Onofre Nuclear Generating Station, Units 2 and 3, Second 10-Year Interval Inservice Inspection Program Plan, Revision 0, (Reference 4) submitted October 4, 1993 was reviewed, including the requests for relief from the ASME Code Section XI requirements that the licensee has determined to be impractical. The review of the Inservice Inspection (ISI)
Program Plan was performed using the Standard Review Plans of NUREG-0800 (Reference 5), Section 5.2.4, " Reactor Coolant Boundary Inservice Inspections and Testing," and Section 6.6, " Inservice Inspection of Class 2 and 3 Components."
In a letter dated November 14, 1994 (Reference 6), the NRC requested additional information that was required to complete the review of the ISI Program Plan. The requested information was provided by the licensee in Revision 3 of the San Onofre Nuclear Generating Station, Units 2 and 3, Second
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10-Year Interval Inservice Inspection Program Plan. In this response, the j licensee, Southern California Edison Company, revised the Code deviations (requests for relief) in Sections 3.3 and 3.4.
The San Onofre Nuclear Generating Station, Units 2 and 3, Second 10-Year
. Interval Inservice Inspection Progran Plan is evaluated in Section 2 of this report. The ISI Program Plan is evaluated for (a) compliance with the appropriate edition / addenda of Section XI, (b) acceptability of examination sample, (c) correctness of the application of system or component examination -
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I 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. Specific inservice test (IST) programs for pumps and valves are being evaluated in other reports.
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l 2. EVALUATION 0F INSERVICE INSPECTION PROGRAN PLAN This evaluation consists of a review of the applicable program documents to L
determine whether or not they are in compliance with the Code requirements and
! any previous license conditions pertinent to ISI activities. This section
- describes the submittals reviewed and the results of the review.
2.1 Documents Evaluated !
Review has been completed on the following information from the licensee: 1 1
' (a) San Onofre Nuclear Generating Station, Units 2 and 3, Second 10-Year Interval Inservice Inspection Progran Plan, Revision 0, (Reference 4) submitted October 4, 1993; and (b) San Onofre Nuclear Generating Station, Units 2 and 3, Second 10-Year Interval Inservice Inspection Progran Plan, Revision 3 (Reference 3) submitted April 17, 1995. -
2.2 Como11ance with Code Reauirements 2.2.1 Como11ance with Apolicable Code Editions The Inservice Inspection Program Plan shall be based on the Code editions defined in 10 CFR 50.55a(g)(4) and 10 CFR 50.55a(b). Based on the starting date of April 1,1994, the Code applicable to the second 10-year interval is the 1989 Edition of ASME Section XI. As stated in Section 1 of this report, the licensee has prepared the San Onofre Nuclear Generating Station, Units 2 and 3 Second 10-Year Inservice Inspection i Program Plan to meet the requirements of 1989 Edition. :
2.2.2 Accentability 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 in Section XI of the ASME Code and 10CFR50.55a(b). Sample size and wcld selection have been implemented in accordance with the Code and 10 CFR ??.55a(b) and appear to be correct with the exception of integrally welded atuchments.
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Code Case N-509 Alternative Rules for the Selection and Examination of Class 1, 2, and 3 Integrally Welded AttachmentsSection XI, Division 1, (Reference 7), has been adopted by the licensee without it being included in NRC Regulatory (Reg) Guide 1.147 (Reference 8). During a conference call on July 25, 1995, the licensee stated that the reference to code Case N-509 would stay in the program and committed that the Code Case would not be used until it is included in Reg Guide 1.147. The Staff believes this position to be acceptable provided that 1) if Code Case N-509 is accepted by the NRC and included in Reg Guide 1.147, the licensee must follow all conditions associated with the Code Case, or
- 11) if Code Case N-509 is not accepted by the NRC prior to the end of the first period of the second interval, a request for relief or a revised plan will be prepared.
2.2.3 Exemotion Criteria The criteria used to exempt components from examination shall be consistent with Paragraphs IWB-1220, IWC-1220, IWC-1230, IWD-1220, and 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 except for Sections 2.10.2, 2.10.3, and 2.10.4 of the Program Plan. These Sections contain tests used in lieu of ASME Section XI pressure testing. However, there is not sufficient technical description to ensure that ASME Section XI requirements are being satisfied.
2.2.4 Auomented Examination Commitments In addition to the requirements specified in Section XI of the ASME Code, the licensee has committed to perform the following augmented examinations:
(a) Reactor vessel examinations in accordance with the requirements of NRC Regulatory Guide 1.150, Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations, Revision 1, (Reference 9);
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, t (b) Volumetric examination of the reactor coolant pump flywheel high stress areas every 3 years, as well as volumetric and surface examinations with the flywheel removed at 10-year intervals, satisfying NRC Regulatory Guide 1.14, Reactor Coolant Pump flywheel Jntegrity, (Reference 10);
(c) Ultrasonic examination of all circumferential and longitudinal welds of the "No break zones" in the Main Steam lines and portions of other high energy piping that penetrate containment; and (d) Volumetric examination of all low-pressure turbine disc bores.
2.3 Conclusion Based on the review of the documents listed above, no deviations from regulatory requirements or commitments were identified in the San Onofre Nuclear Generating Station, Units 2 and 3, Second 10-Year Interval Inservice Inspection Program Plan, Revision 3, with the exception of those described in Sections 2.2.2 and 2.2.3 above and in the evaluation of Requests for Relief 3.3.5 and 3.3.6.
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- 3. EVALUATION OF RELIEF REQUESTS The requests for relief from the ASME Code requirements that the licensee has determined to be impractical for the second 10-year inspection interval are evaluated in the following sections.
3.1 Class 1 Components 3.1.1 Reactor Pressure Vessel 1 3.1.1.1 Reauest for Relief 3.3.1. Examination Cateaories B-A and B-0.
Items Bl.30. B3.90. and 83.100. Reactor Vessel Shell-to-Flanae Weld. Nozzle-to-Vessel Welds. and Nozzle Inside Radius Sections Code Reauirement: Examination Category B-A, Item Bl.30 requires a volumetric examination of at least 50% of the weld by the end of the first period.
Examination Category B-D, Items 83.90 and B3.100 require a volumetric examination of at least 25% but not more than 50% of I the weld and adjacent base metal by the end of the first period. ,
licensee's Code Relief Reauest: The licensee requested relief from performing the Code-required volumetric examinations of the following welds and inner radius sections during the first period of the second interval:
Weld ID Description Cateaory Item No.
02-001-020 Upper Shell-to-Flanga Weld B-A Bl.30 02-001-021 Hot Leg Nozzle-to-Shell Weld B-D B3.90 02-001-024 Hot Leg Nozzle-to-Shell Weld B-D B3.90 02-001-021 Hot Leg Nozzle Inner Radius B-D B3.100 02-001-024 Hot Leg Nozzle Inner Radius B-D B3.100 Licensee's Basis for Reauestina Relief (as stated):
. "This weld is examined from the inside diameter of the reactor vessel by a remote NDE technique that requires removal of the internals (including the core support barrel) to access the weld.
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The installed core barrel renders the shell-to-flange weld inaccessible as shown in Sketch 1-1".
- "These welds were examined during the last refueling outage (1992) which was the full volume first 10-year inservice inspection. No discrepancies were noted at that time and there are no successive inspections pending as a result of indications found in previous inservice examinations. Postponing this examination until the rest of them are done will result in a saving of approximately $1 million."
Licensee's Proposed Alternative Examination (as stated):
" Examination shall be performed by the end of the third inspection period with the rest of the reactor pressure vessel welds."
Evaluation: The licensee stated that the scheduling requirements of Examination Categories B-A and B-D, Items B1.30, B3.90 and B3.100 result in a hardship. Deferral of the first period examinations for these items is acceptable provided that the examinations are completed within the same period in which the preceding examinations were performed, or earlier, so that there is no more than 10 years between examinations.
The licensee has established an acceptable level of quality and safety for the subject welds by examinations performed during the third period of the first 10-year interval. Requiring examination during the first period of the second interval is regarded as a burden. The proposed alternative, performance of all Item Bl.30, B3.90, and B3.100 examinations in the third period of the second 10-year interval, should be authorized provided that there is no more than 10 years between examinations, except where the length of a 10-year interval is adjusted in accordance with IWA-2430.
Conclusion:
The INEL staff has reviewed the licensee's request for relief from the scheduling requirements for the subject examinations. For the subject welds, performing the required "Not included in this evaluation.
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examinations in the first period of the second interval would result in an unnecessary burden without a compensating increase in the level of quality and safety. Therefore, it is recommended that the alternative scheduling be authorized pursuant to
, 10 CFR 50.55a(a)(3)(ii), provided that there is not more than 10 years between examinations, except where the length of a 10-year interval is adjusted in accordance with IWA-2430.
3.1.2 Pressurizer 3.1.2.1 Reauest for Relief 3.3.4 (Part 1). Examination Cateaory B-D.
Item B3.110. Pressurizer Nozzle-to-Vessel Welds Code Reauirement: Examination Category B-D, Item B3.110 requires a 100% volumetric examination of pressurizer nozzle-to-vessel welds as defined by Figure IWB-2500-7. .
Licensee's Code Relief Reauest: The licensee requested relief from performing 100% Code-required volumetric examinations of the following nozzle-to-vessel welds:
Weld ID Descriotion 03-005-009 Surge Nozzle-to-Bottom Head Weld 03-005-010 Spray Nozzle-to-Top Head Weld 03-005-011 Safety Nozzle-to-Top Head Weld 03-005-012 Safety Nozzle-to-Top Head Weld 03-005-013 Safety Nozzle-to-Top Head Weld Licensee's Basis for Reauestino Relief (as stated):
"To achieve full UT coverage of the whole volume required by the code, examination has to be performed on the shell side to scan volume D-E-H-J (Sketch 4-1)" and on the nozzle side to scan volume A-B-C-J. The nozzle design of the San Onofre pressurizer has geometric configuration that precludes achieving the required volume A-B-C-J."
"Not included in this TER.
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Licensee's Proposed Alternative Examination (as stated):
" Perform volumetric examination (UT) of volume D-E-H-J, Sketch 4-2* from the shell or head side of the pressure vessel."
Evaluation: The Code requires volumetric. examinations of pressurizer nozzle-to-vessel welds. However, the San Onofre pressurizer nozzles have a geometric configuration that precludes achieving 100% of the required examination from the outside.
Examination is possible from the internal surface; however, radiation levels of approximately 6 Rem /hr make a manual internal examination impractical.
The design configuration restrictions and high internal dose levels make the Code-required volumetric examination impractical at San Onofre. To obtain complete volumetric coverage, modifications or replacement of the component with one of a design providing for complete coverage would be required.
Imposition of this requirement would cause a considerable burden on the licensee. l The licensee proposed no additional examinations. However, based j on the significant amount of volumetric coverage that appears to j have been obtained, it is reasonable to conclude that a pattern of degradation, if present, would have been detected. Thus, reasonable assurance of continued inservice structural integrity
( has been'provided.
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Conclusion:
The design configuration restrictions and high internal dose levels make the Code-required volumetric examination impractical. Based on the significant amount of weld l coverage obtained, reasonable assurance of operational readiness
- l. has been provided. Therefore, it is recommended that relief be l granted as requested, pursuant to 10 CFR 50.55a(g)(6)(i).
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( 'Not included in this TER.
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3.1.3 Heat Exchanaers and Steam Generators 3.1.3.1 Reauest for Relief 3.3.4 (Part 2) Examination Cateaory B-D.
Item B3.130. Steam Generator Nozzle-to-Vessel Welds
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Code Reauirement: Examination Category B-D, Item B3.130 requires a 100% volumetric examination of steam generator nozzle-to-vessel welds as defined by Figure IWB-2500-7.
Licensee's Code Relief Reauest: The licensee requested relief from performing 100% of the Code-required volumetric examinations I of the following nozzle-to-vessel welds:
Weld ID Descriotion 03-003-010 Inlet Nozzle-to-Head Weld 03-003-011 Outlet Nozzle-to-Head Weld 03-003-012 Outlet Nozzle-to-Head Weld 03-004-010 Inlet Nozzle-to-Head Weld 03-004-011 Outlet Nozzle-to-Head Weld 03-004-012 Outlet Nozzle-to-Head Weld Licensee's Basis for Reauestino Relief (as stated):
"To achieve full UT coverage of the whole volume required by the code, examination has to be performed on the shell side to scan !
volume D-E-H-J (Sketch 4-1)d and on the nozzle side to scan volume A-B-C-J. The nozzle design of the San Onofre steam generators has geometric configuration that precludes achieving '
the required volume A-B-C-J. l "UT scan of the nozzle from the inside surface is possible. I However, the dose rate in the primary side is about 6000 mr/Hr or 3000 mr per half hour just for the examiner alone not considering the support crew that would be required for this examination."
Licensee's Proposed Alternative Examination (as stated):
" Perform volumetric examination (UT) of volume D-E-H-J,
' Sketch 4-2* from the shell or head side of the pressure vessel."
"Not included in this TER.
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& 4 Evaluation: The Code requ!res volumetric examinations of the steam generator nozzle-to-vtssel welds. However, the San Onofre steam generator nozzles have a geometric configuration that precludes examining 100% of the required volume. Examination of the subject area from the outside is limited due to nozzle configuration. While examination from the internal surface is possible, radiation levels of approximately 6 Rem /hr make a manual internal examination impractical.
The design configuration restrictions and the high internal dose levels make the Code-required volumetric examination impractical at San Onofre. To obtain complete volumetric coverage, modifications or replacement of the component with one of a design providing for complete coverage would be required.
Imposition of this requirement would cause a considerable burden on the licensee.
The licensee proposed no additional examinations. However,' based on the significant amount of volumetric coverage that is being obtained, it is reasonable to conclude that a pattern of degradation, if present, would have been detected. Thus, reasonable assurance of continued inservice structural integrity has been provided.
Conclusion:
The design configuration restrictions of the steam I generator nozzles and the high internal dose levels make the Code-required volumetric examination impractical at San Onofre.
However, the significant volumetric coverage attained provides reasonable assurance of operational readiness. Therefore, it is recommended that relief be granted as requested, pursuant to 10 CFR 50.55a(g)(6)(1).
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l ~ a L g 3.1.4 Pinina Pressure Boundarv 3.1.4.1 Reauest for Relief 3.3.2. Examination Cateoory B-J. Items B9.11 and B9.12. Circumferential and Lonaitudinal Welds in Class 1 Pressure-Retainina Pioina
, ~ Code Reauirement: Examination Category B-J, Items B9.11 and 89.12 require volumetric and surface examination of Class 1 piping NPS 4 inches or larger per Figure IWB-2500-8.
Licensee's Code Relief Reauest: The licensee requested relief from performing the Code-required surface examination of the following welds:
Weld ID Descriotion 02-001-033 Hot' Leg Nozzle-to-Ext Piece Weld 02-001-034 Cold Leg Nozzle-to-Ext Piece Weld 02-001-035 Cold Leg Nozzle-to-Ext Piece Weld 02-001-036 Hot Leg Nozzle-to-Ext Piece Weld 02-001-037 Cold Leg Nozzle-to Ext Piece Weld 02-001-038 Cold Leg Nozzle-to-Ext Piece Weld 02-006-005 Pipe Long Weld 02-006-006 Pipe Long Weld 02-006-007 Pipe-to-Outlet Nozzle Ext Pc Weld 02-007-005 Pipe Long Weld 02-007-006 Pipe Long Weld 02-007-007 Pipe-to-Outlet Nozzle Ext Pc Weld 02-009-005 Pipe-to-Elbow 02-009-003 Pipe Longitudinal Weld 02-009-004 Pipe Longitudinal Weld 02-009-006 Elbow Long Weld 02-009-007 Elbow Long Weld 02-009-008 Elbow-to-Inlet Nozzle Ext Pc Weld 02-011-005 Pipe-to-Elbow 02-011-003 Pipe Longitudinal Weld 02-011-004 Pipe Longitudinal Weld 02-011-006 Elbow Long Weld
, 02-011-007 Elbow Long Weld 02-011-008 Elbow-to-Inlet Nozzle Ext Pc Weld 02-013-005 Pipe-to-Elbow
. 02-013-003 Pipe Longitudinal Weld 02-013-004 Pipe Longitudinal Weld 02-013-006 Elbow Long Weld 02-013-007 . Elbow Long Weld 02-013-008 Elbow-to-Inlet Nozzle Ext Pc Weld 13
Weld ID Description 02-015-005 Pipe-to-Elbow 02-015-003 Pipe Longitudinal Weld
, 02-015-004 Pipe Longitudinal Weld 02-015-006 Elbow Long Weld 02-015-007 Elbow Long Weld
. 02-015-008 Elbow-to-Inlet Nozzle Ext Pc Weld Licensee's Basis for Reauestino Relief (as stated):
"These welds are located inside.the primary reactor shield cavity. The surface examination is done manually while the volumetric examination (UT) is done using remote UT technique and equipment. Those performing the manual surface examinations (and supports such as builders of scaffolding, removal of insulation, preparing and grinding the welds, fire watch, HP among others) shall be exposed to a dose rate of from 250 to 8000 mRenVHr. The estimated number of hours required for these examinations are as follows: Build scaffolding: 64 hours7.407407e-4 days <br />0.0178 hours <br />1.058201e-4 weeks <br />2.4352e-5 months <br />, remove insulation:
32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />, weld prep: 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, NDE examination for 24 welds:
96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />, reinstall insulations: 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> and remove scaffolding:
32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />. Total man-hours is 304. For a minimum 1/4 or 0.25 R/Hr times 304 Hrs, this amounts to 76 Rem.
"The entire volume of these welds and heat affected zones were examined by Ultrasonic Test (UT) in the first 10-year interval.
No relevant . indications were found. This relief request will result in dose savings of 76 Rem for this interval."
Licensee's Proposed Alternative Examination (asstated):
" Examination surface area A-B as shown in sketch 2-3* shall be I examined by Ultrasonic Test (UT). This surface examination (A-B) !
shall be performed at the same time with remote UT examination of I volume C-D-E-F. This examination will cover 100% of the weld !
volume instead of the 33% volume now required by the code."
Evaluation: For the subject Category B-J welds, the Code
. requires a volumetric examination of the inner 1/3 of the weld and a surface examination. However, the licensee proposed to perform a remote UT examination of 100% of each of the subject held volumes. These welds are located inside the primary reactor shield cavity. The surface examination is typically done manually while the volumetric examination (UT) is done using remote UT techniques and equipment. However, using the proposed
- Not included in this TER.
14
o .
a' .
alternative will result in savings of approximately 76 Man-Rem for this interval.
The INEL staff believes that the alternative examination l technique is acceptable, provided that the ultrasonic examination l technique has been demonstrated to be capable of detecting 00 l surface-connected cracks.
Conclusion:
The alternative examination will provide an acceptable level of quality and safety. Therefore, it is ;
t recommended that this alternative be authorized, pursuant to 10 CFR 50.55a(a)(3)(i), provided that the ultrasonic examination technique has been demonstrated to be capable of detecting OD surface-connected cracks.
3.1.5 Pumo Pressure Boundarv (No requests for relief) 3.1.6 yalve Pressure Boundarv 3.1.6.1 Reauest for Relief 3.3.3. Examination Cateoory B-M-1.
Item B12.40. Pressure-Retainina Welds in Valve Bodies Code Reauirement: Examination Category B-M-1, Item B12.40 requires volumetric examination of Class I welds in valve bodies NPS 4 inches or larger per Figure IWB-2500-17.
Licensee's Code Relief Reauest: The licensee requested relief from performing the Code-required volumetric examination of the following welds:
Weld ID Descriotion 02-019-013 Valve Body Upper Section Weld
, 02-019-014 Valve Body Lower Section Weld 02-019-010 Valve Body Lower Section Weld l 02-019-010A Valve Body Upper Section Weld l
. 02-019-010 Valve Body Lower Section Weld 02-019-010A Valve Body Upper Section Weld 02-020-010 Valve Body Lower Section Weld 02-020-010A Valve Body Upper Section Weld 02-021-026 16" Valve Body Upper Section Weld 15 l
l
c .
Weld ID Description 02-021-027 10" Stop Valve Body Lower Section Weld 02-021-034 16" Stop Valve Body Lower Section Weld
. 02-021-035 16" Stop Valve Body Upper Section Weld 02-021-044A 10" Stop Valve Body Lower Section Weld l 02-021-044 10" Stop Valve Body Upper Section Weld 02-021-054E 10" Stop Valve Body Lower Section Weld 02-021-0548 10" Stop Valve Body Upper Section Weld Licensee's Basis for Reauestina Relief (as stated):
"These valves are made of cast austenitic stainless steels with body welds using the electroslag welding process. The large grain structure of the cast material results in sound dispersion and attenuation that will not provide meaningful examinations for the component being inspected.
" Ultrasonic Test (UT) of cast austenitic stainless steel materials have not provided meaningful results as of this date.
If and when a newly developed technique becomes available, it will be used and adopted for this examination."
Licensee's Proposed Alternative Examination (as stated):
"In lieu of the Volumetric Examination required, the following examinations will be performed:
"1. Surface Examination (PT) of the weld and heat affected zone (area L-M).
"2. Visual Examination (VT-3) of the valve internals when the valve is dissembled for maintenance or repair.
"3. Visual Examination (VT-2) of the component in conjunction with the reactor coolant system pressure test following each refueling or repairs to this component."
Evaluation: The Code requires volumetric examinations of Class I welds in valve bodies NPS 4 inches or larger. The licensee proposed to perform a surface examination (PT) of the weld and heat affected zone, a VT-3 visual examination of the valve
. internals when the valve is dissembled for maintenance or repair, and the Code-required VT-2 visual examination in conjunction with
, the reactor coolant system pressure test following each refueling outage or repairs to this component.
! 16
, . ~
i . .
The attenuative properties of the cast stainless steel valve' body and of electroslag welds make the Code-required volumetric
- examination impractical for the subject welds. To obtain f' complete volumetric coverage, modifications and/or replacement of l the valves with ones of a design providing for complete coverage
)'
j would be required. Imposition of this requirement would cause a considerable burden on the licensee.
j
Conclusion:
The valve body weld examination areas listed above
- have large grain structures, resulting in sound dispersion and
! attenuation, that make the Code-required volumetric examination i impractical. The licensee's alternative ixaminations will -
provide reasonable assurance of operational readiness.
Therefore, it is recommended that relief be granted, pursuant to j 10CFR50.55a(g)(6)(1).
i l
3.1.7 General (No requests for relief) 3.2 Class 2 Components
} 3.2.1 Pressure Vessels i
j 3.2.1.1 Reauest for Relief 3.3.6. Examination Cateaories C-A. C-C. and F-A. Items C1.20. C1.30. C3.10. and F1.40. Reaenerative Heat
- Exchanaer Examinations Code Reauirement
- Examination Categories C-A, C-C, and F-A,
! Items C1.20, C1.30, C3.10, and F1.40 require volumetric, surface, 4
or visual examination of Regenerative Heat Exchanger l pressure-retaining welds, integral welded attachments, and
- supports, respectively.
i
- Licensee's Code Relief Reauest: The licensee requested relief from performing the following Code-required Regenerative Heat j Exchanger examinations:
i 17
~. -. . _ . . , . - .-. _ _
9 e I
ISI ID Descriotion Cat./ Item Method 03-080-010 Tubesheet-to-Shell Weld C-A/C1.30 Vol l 03-080-020 Tubesheet-to-Shell Weld C-A/C1.30 Vol
, 03-080-030 Head Circumferential Weld C-A/C1.20 Vol 03-080-040 ~ Head Circumferential Weld C-A/C1.20 Vol 03-080-050 Head Circumferential Weld C-A/C1.20. Vol L ,
03-080-060 Head Circumferential Weld C-A/C1.20 Vol 1 03-080-070 Head Circumferential Weld C-A/C1.20 Vol 03-080-080 Tubesheet-to-Shell Weld C-A/C1.30 Vol 03-080-090 Tubesheet-to-Shell Weld C-A/C1.30 Vol 03-080-100 Integrally Welded Attachment C-C/C3.10 Surface i
! 03-080-110 Integrally Welded Attachment C-C/C3.10 Surface l 03-080-120 Integrally Welded Attachment C-C/C3.10 Surface l 03-080-130 Integrally Welded Attachment C-C/C3.10 Surface l 03-080-140 Integrally Welded Attachment C-C/C3.10 Surface
- 03-080-150 Integrally Welded Attachment C-C/C3.10 Surface l 03-080-160 Integrally Welded Attachment C-C/C3.10 Surface 03-080-170 Integrally Welded Attachment C-C/C3.10 Surface 03-080-180 Integrally Welded Attachment C-C/C3.10 Surface 03-080-190 Integrally Welded Attachment l C-C/C3.10 Surface '
03-080-200 Integrally Welded Attachment C-C/C3.10 Surface 03-080-210 Integrally Welded Attachment C-C/C3.10 Surface 03-080-220 Support, Cradle Bands - Upper F-A/F1.40 VT-3 03-080-230 Support, Cradle Bands - Lower F-A/F1.40 VT-3 Licensee's Basis for Reauestina Relief (asstated):
" Radiation exposure rates in the Regenerative Heat Exchanger room are approximately 300 mR/Hr. Estimates for the Regenerative Heat l Exchanger portion of the ISI program are approximately 160 l
manhours. The effective jobsite dose rate is 60 mR/Hr in the Regenerative Heat Exchanger room. The ALARA dose estimate for this work is 9.54 person Rem."
Licensee's Prooosed Alternative Examination (asstated):
" Perform Visual (VT-2) examination for Items C1.20 and C1.30 each Period in lieu of the surface examination required by the code."
l Evaluation: The Code requires volumetric examination for the t
l Item C1.20 and C1.30 welds, surface examination for the Item C3.10 attachments, and VT-3 visual exam for the supports. The licensee proposed to perform a VT-2 visual examination in lieu the Code-required surface examination. However, the proposed VT-2 visual examination is already required by the Code. The
! licensee'has not provided an adequate reason for impracticality I
18
I or provided technical bases why the alternative will provided a reasonable assurance of operational readiness.
i .
Conclusion:
The technical information currently provided by the licensee does not support a determination of impracticality or provide a reasonable assurance of operational readiness.
l Therefore, it is recommended that relief be denied.
3.2.2 Pinina (No requests for relief) 3.2.3 Pumps (No requests for relief) 3.2.4 Valves (No requests for relief) l j 3.2.5 General (No requests for relief) 3.3 Class 3 Comoonents (No requests for relief)
. 3.4 Pressure Tests (No requests for relief)
! 3.5 General 4
I 3.5.1 Ultrasonic Examination Techniaues (No requests for relief) i
{ 3.5.2 Exemoted Comoonents (No requests for relief) 3.5.3 Other l 3.5.3.1 Reauest for Relief 3.3.5. Paracraoh IWA-2312. Certification and Recertification t . Code Reauirement: Paragraph IWA-2312 requires personnel
- g performing visual examinations to be qualified and certified to
. qualification levels comparable to those defined in SNT-TC-1A.
4
, 19 -
g i
Licensee's Code Relief Reauest: The licensee requested to use
, non-certified personnel to perform the Code-required VT-3 visual
{, examinatiori on the following valve bodies:
- Weld ID Description 02-025-120A 6" Safety Valve Body 02-025-230A 6" Safety Valve Body
- I Licensee's Basis for Reauestina Relief (as stated)
- )
1
' "These relief valves are sent out to a qualified vendor testing l facility whose repair and replacement program was approved by the utility. The components are reworked, repaired, refurbished and tested using a qualified repair program. The vendor performing the visual examinations are not certified to SNT-TC-1A or other comparable level of qualification defined in SNT-TC-1A. However, they have field expertise equal or superior to the level of training required by the code, and the scope of the maintenance or investigation activities is superior to the scope of the Code required examination.
"In the past, components have been inspected during a maintenance overhaul or investigated by personnel highly qualified but not necessarily certified in the code requirements. In some cases a complete root cause analysis was performed in addition to those performed within the scope of the ISI program. In order to prevent redundant inspection and reduce exposure, credit should be taken for the inspection which these individuals do. The component inspection will remain in the ISI program, however, credit for the inspection will be taken utilizing the vendor's inspection report, or the root cause report as applicable, in lieu of an additional VT-3 report."
Licensee's Proposed Alternative Examination (as stated):
"In addition to the thorough visual inspections by the vendor, the valves will be dimensionally inspected, damaged parts will be replaced and the refurbished valves tested. Records of repair and replacement in accordance with IWA-7520 shall be maintained by the Owner."
Evaluation: The Code requires that personnel performing visual examinations be certified to levels comparable to those defined in SNT-TC-1A. The licensee requested that a vendor whose l personnel are not certified to SNT-TC-1A or other comparable standard be allowed to perform VT-3 visual examinations on the 20
l subject valves. The licensee did not provide sufficient l justification regarding burden or impracticality for relief.
Conclusion:
The technical information currently provided by the licensee does not support a determination of impracticality or provide a reasonable assurance of operational readiness.
Therefore, it is recommended that relief be denied.
3.5.3.2 Reauest for Relief 3.4. Paracraoh IWA-4511. General Reauirements for Repair Weldina Usina Half-Bead Weldina Technioue l
Code Reauirement: daragraph IWA-4511 requires that only the shielded metal arc welding (SMAW) process using low hydrogen j
electrodes be used.
! Licensee's Code Relief Reauest: The licensee requested relief to use IWA-4500, Alternative Repair Welding Methods, of the 1992 Edition of.ASME Section XI, whi s allows the use of gas tungsten arc welding (GTAW).
, Licensee's Basis for Reauestina Relief (as stated):
"SCE prefers the automatic / machine GTAW process over the manually operated SMAW process for temper bead repairs. The automatic / machine GTAW process is inherently superior for controlling important process variables that affect the overall quality of the weld. Additionally, the automatic / machine GTAW process is more suited for adverse field environment than manual welding is due to the remote operation feature of the automatic / machine GTAW equipment.
" Welding variables that affect the quality of the weld are automatically controlled by microprocessor or are controlled by the welding operator from a remote location with the automatic / machine GTAW process. On the other hand, with a manual welding process such as SMAW, the quality of the weld is contingent on the welder's ability to skillfully manipulate the welding electrode. The welder's freedom of movement is severely restricted by the multiple layers of protective clothing that must be worn in a radi1 active environment. Therefore, the automatic / machine GTAW process provides a higher level of confidence that the desired weld properties can be obtained, even under adverse environmental conditions.
21 l _ _ _ _ ,
e "SCE is aware of ASME Section XI Code Case N-432 that permits the
- use of the automatic / machine GTAW process, however, would point
! out that significant advances in using this technology and process since the Code's approval of N-432.
"The following are examples of the differences between Code Case N 432 and Section XI, IWA-4500, 1992 edition: The later Code requires the base materials for procedure qualification test
' coupons be the same P Number and Group Number whereas the N-432 Code Case requires they be the same specification and grade of the production material. The N-432 Code Case limits the repair
< area to a maximum of 54 square inches whereas the later code 5 rules permit repair areas up to 100 square inches. The N-432
- ' Code Case mandates a single set of acceptance criteria for impact testing of the procedure qualification test assembly that are more stringent than the original requirements of our design i
specification. The later Code rules, in recognition of this, specify that the impact acceptance criteria meet the requirements i of the design specification. These changes in the later Code save the users the expense and time of qualifying a multitude of
- temper bead weld procedures for the same P number and Group number materials and permit the temper bead weld procedures that are qualified to have a much broader scope than would otherwise i be possible under Code Case N-432" f Licensee's proposed Alternative Examination (as stated)
"ASME XI, IWA-4500, 1992 edition, No addenda (Temperbead Technique, Auto /MachineGTAW)"
j Evaluation: The Code of record only allows the shielded metal i
arc welding (SMAW) process to be used for alternative welding procedures contained therein, whereas the 1992 Edition allows the gas tungsten arc welding (GTAW) process to be used. The licensee l
! proposes to use the 1992 Edition.
l The NRC staff reviewed IWA-4500 of the 1992 Edition in an NRC Safety Evaluation Report dated June 10, 1993. It was concluded !
that those provisions, which allow use of the GTAW process, j I
provide an acceptable level of quality and safety.
y i
Conclusion:
The alternative welding process in the 1992 Edition
- . will- provide an acceptable level of quality and safety.
Therefore, it is recommended that this alternative be authorized as requested, pursuant to 10 CFR 50.55a(a)(3)(1).
22
-c -- --m -_v , , -
. -.- ---.-.------ ~ . - - . _ - . _
- , ?'
- 4. CONCLUSION 1
It has been determined that certain inservice examinations cannot be performed i to the extent required by Section XI of the ASME Code. In the cases of l
Requests for Relief 3.3.3, 3.3.4 (Part 1), and 3.3.4 (Part 2), the licensee j has demonstrated that specific Section XI requirements are impractical; it is i therefore recommended that relief be granted as requested, pursuant to ;
10CFR50.55a(g)(6)(1). The granting of relief will not endanger life, !
- property, or the comon defense and security and is otherwise in the public interest, giving due consideration to the burden upon the licensea that could ;
- result if the requirements were imposed on the facility.
Pursuant to 10 CFR 50.55a(a)(3)(1), it is concluded that the licensee's
- proposed alternative to the Code requirement stated in Request for Relief 3.4 i
- will provide an acceptable level of quality and safety; it is recommended that l l
the proposed alternative be authorized. In the case of Request for Relief l 3.3.2, it is recommended that the proposed alternative be authorized only if the licensee satisfies the conditions stated in the evaluation of the relief 4
request. .
In Request for Relief 3.3.1 the licensee has demonstrated that specific Section XI requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. In this case, it is recommended that the proposed alternative be authorized, pursuant to '
10 CFR 50.55a(a)(3)(ii), provided the licensee satisfies the conditions stated in the above request for relief evaluation.
For Requests for Relief 3.3.5 and 3.3.6, the licensee has not provided sufficient technical justification to support the determination that 1) the Code requirements are impractical, or that 11) requiring the licensee to comply with the Code requirements would result in hardship. Therefore, in
, these cases it is recommended that relief be denied.
. This technical evaluation has not identified any practical method by which the licensee can meet all the specific inservice inspection requirements of l Section XI of the ASME Code for the existing San Onofre Nuclear Generating Station, Units 2 and 3, facility. Compliance with all of the Section XI 23
.~ .
examination requirements would necessitate redesign of a significant number of plant systems, pror.urement of replacement components, installation of the new components, and performance of baseline examinations for these components.
Even after the redesign efforts, complete compliance with the Section XI examination requirements probably could not be achieved. Therefore, it is concluded that the public interest is not served by imposing certain i provisions of Section XI of the ASME Code that have been determined to be impractical.
The licensee should continue to monitor the development of new or improved examination techniques. As improvements in these areas are achieved, the licensee should incorporate these techniques into the ISI program plan.
i Based on the review of the San Onofre Nuclear Generating Station, Units 2 and 3, Second 10-Year Interval Inservice Inspection Program Plan, Revision 3, the licensee's response to the NRC's request for additional information, and the recommendations for granting relief from the ISI examinations that cannot be performed to the extent required by Section >*I of the ASME Code, no deviations from regulatory requirements or commitments were identified, except l
those noted in Sections 2.2.2 and 2.2.3 and in Oequests for Relief 3.3.5 and 3.3.6.
i e
a 24
- e. .
i
- 5. REFERENCES
- 1. Code of Federal Regulations, Title 10, Part 50.
- 2. American Society of Mechanica] Engineers Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Division 1, 1989 Edition.
- 3. San Onofre Nuclear Generating Station, Units 2 and 3, Second 10-Year Interval Inservice Inspection Program Plan, Revision 3, submitted Apri1 17, 1995.
- 4. San Onofre Nuclear Generating Station, Units 2 and 3, Second 10-Year Interval Inservice Inspection Program Plan, Revision 0, submitted October 4, 1993.
- 5. 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.
- 6. Letter, dated November 14, 1994, Mei Fields (NRC) to Mr. Haroid B. Ray (Southern California Edison Company), containing request for additional information on the second 10-year interval ISI program plan.
- 7. Code Case N-509 Alternative Rules for the Selection and Examination of Class 1, 2, and 3 Integrally Welded AttachmentsSection XI, Division 1, November 25, 1992.
- 8. NRC Reguintory Guide 1.147, Inservice Inspection Code Case Acceptability, AS#E Section XI, Division 1, Revision 11, October 1994. l
- 9. NRC Regulatory Guide 1.150, Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations, Revision 1, February 1983.
- 10. NRC Reguiatory Guide 1.14, Reactor Coolant Pump Flywheel Integrity, Revision 1, August 1975.
r 25 .
f /*' '
Nag.pvass u.s. NUCLE AR REGUL ATORY Commission n Ea
] EoEE' BIBLIOGRAPHIC DATA SHEET
($tt anst!\octens CA the rtverst1
- 2. TITLE AND SU8 TITLE .
INEL-95/0503 Technical Evaluation Report on the Second 10-Year Interval Inservice Inspection Program Plan 1 CATE REPORT PU8us-ED Southern California Edison Company a~'- na=
San Onofre Nuclear Generating Station, Units 2 and 3 Novembe;r 1995 Docket Number 50-361 and 50-362 vin oR GRANT NuveER
- 5. AUTHORLS)
FIN-L2556 (Task 43
- 6. TYPE OF REPCRT l,
Technical M. T. Anderson, E. J. Feige, K. W. Hall uERico CovEREo,, ,-
1 eP o- o" .a n v.s ~ r *<,.nc v e.-* - .. <- u o grgRyg,N,zATuoN - NAuE ANo AcoREss ur nac..
- ora. .ca. . .
- INEL/LITCO -
4 P.O. Box 1625 idaho Falls, ID 83415-2209 4
- 9. C -
G ANIZATION - N AME AND A00R ESS ist *dc. tree ~seme as so..v". ,/c.arreer.r. ar e **c o***a, 0"m. .c Aspen, y1 hveter Asyver.ry Commesw.n, i
Civil and Geosciences Branch Office of Nuclear Regulatory Commission y U.S. Nuclear Regulatory Commission i Washington, D.C. 20555
- 10. SUPPLEMENTARY NOTES
- 11. A85 TRACT (200weros .r mas
]
' This report presents the results of the evaluation of the San Onofre Nuclear Generating Station, Units 2 and 3, Second 10-Year Interval Inservice Inspection (ISI) Program Plan, submitted October 4,1993 including the requests for relief froi t the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI requirements that the Licensee has determined to be impractical. The Sai '
Onofre Nuclear Generating Station, Units 2 and 3, Second 10-Year Interval ISI Program Plan 1:; evaluated in Section 2 of this report. The ISI Program Plan is evaluated for (a) compliance with the appropriate edition / addenda of Section XI, (b;
- acceptability of examination sample, (c) correctness of the application of system oi component examination exclusion criteria,.and (d) compliance with ISI-related commitments identified during previous Nuclear Regulatory Commission (NRC) reviews, j The requests for relief are evaluated in Section 3 of this report.
i ,
)
12, KE Y WOR DSiDE SCR:PTOR S ites rse .r aarsses rast asn amst **'r***e m 'ac*'*8 r** *8'" 8 4 ***''''8'"
Unlimited 14 5tCw mit y CLA&&+ CATION
} ITkn Pa.es
~'
Unclassified ir~s a ,,
Unclassified
- 16. NUMBER OF PAGES
- 16. PRICE
.ac .o.u m m
4
'l
( n, 5All ORDFRE EENERATINE STATION, IBIITS 2 AIR 3 Page 1 of 1 Second 10-Year ISI Interval TABLE 1
SUMMARY
OF REUEF REQUESTS
' ^
Relief ..' # '
Relief flequest 15ystemer$' . Exami.. x Item f ,. .
ilicensee Proposed Request missher' re t' ~ Category '; lie. Tveheme or Area to be Examined - . Required Method ~ : Alternative States 3.3.1 Reactor B-A B1.30 Shell-to-Flange Welds Volumetric Deferral of examinations Authorized Vessel 8-D 83.90 Nozzle-to-Vessel Welds examination of 501 to third period Conditionally 83.100 Nozzle Inner Radius Sections during first period 3.3.2 Class 1 8-J 89.11 Circumferential and Volumetric and Volmetric exam of 1001 Authorized Piping 89.12 Longitudinal Welds surface of weld volume Conditionally Systems examination 3.3.3 Class 1 8-M-1 812.40 Pressure-retalning welds in Volumetric Surface, VT-3, and VT-2 Granted Valves valve bodies examination Visual examinations 3.3.4 Pressurtzer B-0 83.110 flozzle-to-Vessel Welds Volumetric Perform volumetric exam Granted (Part 1) examination to the extent practical 3.3.4 Steam 8-D B3.130 Nozzle-to-Vessel Welds Volmetric Perform wohanetric exam Granted (Part 2) Generator examination to the extent practical '
3.3.5 IWA-2312 Certification and Use of certified Use of non-certified Denied recertification personnel for Code personnel for Code VT-3 VT-3 visual visual examinations examinations 3.3.6 Regenerative C-A C1.20 Pressure-retalning welds. Volumetric, leone Denied N
m Heat C-C C1.30 integral welded attachnents, surface, or visual S Exchanger F-A C3.10 and supports examinations Q F1.40 m c+
to 3.4 IWA-4511 Half bead repair welding Shielded metal arc Perform Gas tungsten arc Authorized welding process to welding perform half bead repair welding 4
-wv _ _ _ - - _ _ . - . - - _ - . _ _ _ . - - . . - _ - - _ _ _ _ - . - . . . - _ _ . - _ - _ _ _ . - . _ - - _ - _ - - _ _ _ _ _ - _ - _ - - . - . _ - - . _ - . . _ _ _ . _ - _ _ . . _ _ _