ML20209A374
| ML20209A374 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 04/05/1999 |
| From: | Beasley J SOUTHERN NUCLEAR OPERATING CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LCV-1307, NUDOCS 9904140231 | |
| Download: ML20209A374 (26) | |
Text
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J.BariliBeasley.Jr PL Southern Nucle:r Vice President Operating Company,1:c.
i Wgtle Preject 40 Inamess Center Parkway i
L R0. Box 1295 Birmingharn. Alabarna 35201 Tel 205 992.7110 Fax 205.992.0403 SOUTHERN COMPANY Energy to Serve Your %rld*
LCV-1307 April 5, 1999' Docket Nos.: 50-424, 50-425 U. S. Nuclear Regulatory Commission
~ ATTN: Document Control Desk Washington, D. C.
20555 Ladies and Gentlenien:
VOGTLE ELECTRIC GENERATING PLANT REQUESTS FOR RELIEF - CONTAINMENT INSERVICE INSPECTION PROGRAM By Federal Register notice dated August 8,1996, the U. S. Nuclear Regulatory Commission (NRC) amended 10 CFR 50.55a, Codes and Standards, to incorporate by reference Subsections IWE and IWL of the 1992 Edition and Addenda of Section XI to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. Subsections IWE and IWL provide the inservice inspection requirements for Class.
MC (metallic containments, including the metallic liners of Class CC containments) and Class CC (concrete containments), respectively. The containment for Vogtle Electric Generating Plant (VEGP), Units I and 2, each consists of a prestressed reinforced concrete cylinder and hemispherical dome supported on a flat, conventionally reinforced concrete basemat with a central cavity and instrumentation tunnel to house the reactor vessel. The inside face of the containment is lined with steel plates which are welded together to form a leak-tight barrier and which are anchored to the concrete. The containment structure was designed in accordance with ASME Section III, Division 2, 1975 Edition through Winter 1975 Addenda, Article CC-3000 only. The rules of both Subsection IWE and IWL apply to the VEGP containment vessels.
l Certain of the requirements of Subsections IWE and IWL necessitate that vi alternative examination be performed or present either a hardship or an impracticality. As a result, Southern Nuclear Operating Company (SNC) submits herein several requests for relief h
O for VEGP from the Code requirements pursuant to 10 CFR 50.55a(a)(3) and (g)(5)(iii).
Specifically, the requests for reliefinvolve the following areas:
f l
Visual examination of containment penetration seals and gaskets (Request for Relief e
RR-E-1),
9904140231 990405 PDR ADOCK 05000424 PE G
l w-
e 4
%a 4
U. S. Nuclear Regulatory Commission l
LGV-1307 Page Two Successive examinations after containment repairs (Request for Relief RR-E-2),
e Bolt torque or tension testing on Class MC pressure-retaining bolting (Request for Relief RR-E-3),
Preservice inspection of new paint or coatings (Request for Relief RR-E-4),
Visual examination of paint or coatings prior to removal (Request for Relief RR-E-5),
e I
Visual examination (VT-3/VT-3C) illumination and resolution (Request for Relief a
RR-L-1 ),
Alternative schedule for tendon and concrete examinations (Request for Relief RR-L-2),and Alternative for tendan testing (VEGP-2 only)(Request for Relief RR-L-3).
Each of the requests for relief referenced above are found in Enclosure 1 and should be referred to for details. SNC is a participant in the Electric Power Research Institute fPRI) project for the preparation of a Generic IWE/IWL Program and Guide, including generic requests for relief. This program was described to the NRC on October 23,1997, during a meeting between the NRC, EPRI, Nuclear Energy Institute (NEI), and several utilities. Except for Requests for Relief RR-L-2 and RR-L-3, which are unique to VEGP, the remaining enclosed equests for relief from the Code requirements are similar to those that have been previously submitted to the NRC by other licensees for review and approval. The VEGP requests for relief, with the exception noted herein, a:e bcsed on requests for relief developed as part of the EPRI project. The page numbers noted at the bottom of each page of the enclosed requests for relief correspond to the page numbers in the VEGP Containment Inservice Inspection Program Plan document that will be maintained at the plant site for NRC review upon request.
In addition to the foregoing, SNC requests relief from the reporting requirements of 10 CFR 50.55a(b)(2)(ix)(A) (as interpreted by SNC) and (D) relative to tendon grease leakage. In lieu of reporting grease leakage per the aforementioned regulatory ret;uiremem(s), SNC requests that reporting of grease lealage continue to be permitted per our existing plant Technical 5pecifications. Details on this request for relief from the regulatory requirements are provided in Enclosure 2.
l The NRC is respectfully requested to approve these requests for relief from the Code and/or regulatory requirements prior to January 1,2000. This request is based on the fact
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s U. S. Nuclear Regulatory Commission LC,V-1307 Page Three that only one or two outages remain on each VEGP unit in order to complete the expedited examinations that are required by the rulemaking to be completeu oy September 9,2001. T he results of the NRC review are required to facilitate adequate planning for these examinations.
Should there be any questions in this regard, please contact this office.
Sincerely, fg 8
1
/. B. Bea y
r.
JBB/JAE/
Enclosures:
- 1. VEGP Units 1 and 2, Requests for Relief from the Code Requirements of the ASME Section XI Code,1992 Edition with 1992 Addenda (includes RR-E-1, -2, -3, -4, -5, RR-L-1, -2, and -3)
- 2. VEGP Units 1 and 2, Request for Relief from the Regulatory Requirements of 10 CFR 50.55a(b)(2)(ix)(A) and (D) xc: Hartford Steam Boiler Inspection and Insurance Company Mr. F. E. Bellais, Jr. (w/ enclosures)
Southern Nuclear Onerctng Company Mr. W. L. Burnteister (w/o enclosures)
Mr. J. T. Gasser (w/o enclosures)
Mr. M. Sheibani (w/euclosures)
SNC Document Management (w/ enclosures)
U. S. Nuclean Regulatory Commission l
Mr. R. R. Assa, Project Manager, NRR (w/ enclosures) l Mr. L. A. Reyes, Regional Administrator (w/ enclosures)
Mr. J. Zeiler, Senior Resident Inspector, Vogtle (w/enelosurey
E 1
4 3
I ENCLOSURE 1 TO SOUTHERN NUCLEAR OPERATING COMPANY LETTER LCV-1307 i
VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 REQUESTS FOR RELIEF' FROM Tile CODE REQUIREMENTS OF THE ASME SECTION XI CODE, 1
1992 EDITION WITH 1992 ADDENDA 1
l l
l 1
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Includes Requests for Relief RR-E-1, RR-E-2, RR-E-3, RR-E-4, RR-E-5, RR-L-1, RR-L-2, and RR-L-3 (VEGP Containment insersice Inspection Program Plan Jocument pages 4-3 through 4-20) that follow this cover page.
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SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 FIRST CONTAINMENT INSPECTION INTEP. VAL REQUEST FOR RELIEF NO. RR-E-1 1.
System (s)/ Component (s) for Which Relief is Requestad:
Seals and gaskets of Class MC (Metallic Containment) pressure-retaining components, Examination Category E-D, item Nos. E5.10 and E5.20 of IWE-2500, Table IWE-2500-1, ASME Section XI,1992 Edition,1992 Addenda, as amended by 10 CFR 50.55a.
This request for relief applies to the following components that incorporate seals and gaskets as the containment pressure boundary:
i Electrical penetrations fittea with metal compression fittings.
Airlock door seals, including door operating mechanism penetrations that are part of the containment pressure boundary.
Containment penetrations whose design incorporates resilient seals, gaskets, or sealant compounds.
Doors with resilient seals or gaskets except for seal-welded doors.
II.
Code Requirement :
IWE-2500, Table IWE-2500-1 requires that seals and gaskets on airlocks, hatches, and other devices receive a VT-3 visual examination once each interval to ensure containment leak-tight integrity.
Ill.
Code Requirement from Which Relief is Requested:
Relief is requested from performing the Code-required VT-3 visual examination on the components identified in section 1.
IV.
Basis for Relief:
Practical VT-3 visual examination consideratior,s of these seals and gaskets would require the joints to be oisassembled since many of the surfaces of seals and gaskets are normally inaccessible. The ASME Code Committee recognized that disassembly of the joints to perform visual examinations was not warranted, and th e 1998 Edition of ASME Secthn XI removed the examination requirement.
4-3 Rev.0
a SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 FIRST CONTAINMENT INSPECTION INTERVAL REQUEST FOR RELIEF NO. RR-E-1 (con +inued)
IV.
Basis for Relief: (continued)
The proposed alternate examination (Appendix J, Option B) provides a periodic, non-intrusive test method which will ensure that the integrity of the seats and gaskets is being maintained. As noted in 10 CFR 50 Appendix J, the purpose of the testing is to ensure that leakage of containment penetrations whose design incorporates resilient seals, gaskets, sealant compounds, and electrical penetrations fitted with seal assemblies remains below established limits. Damage to seals or gaskets, which could affect containment integrity, is best detected with this type of test and will be performed as follows:
Electrical Penetrations And Containment Penetrations Whose Design incorporates Resilient Seals, Gaskets, Or Sealant Compounds Those penetrations that are not disassembled during the 10-year interval will receive an Aopendix J Option B test at least once in the 10-year interval. For those penetrations that are disassembled or opened, an Appendix J test is required upon final assembly (prior to start-up). Additionally, if a seal (including O-rings) or gasket is replaced, it will be visually incpected by maintenance personnel before re-assembly or closure. These tests and inspections will assure the leak tightness of primary containment and provide an acceptable level of quality and safety.
Airlocks and the Containment Equipment Hatch The Personnel Airlocks are opened as needed during maintenance outages and refueling outages. Prior to final closure, the accessible portions of gaskets and the door sealing faces are inspected for damage that could affect the leak tightness of the seal. If gasket replacement is necessary, the new gasket will be visually inspected by maintenance personnel before re-assembly or closure. Door seals wi! be Appendix J tested in accordance wiih Technical Specification requirements within sevcn days of opening and once every 30 days during periods of frequent opening.
4-4 Rev.O
o J
SOUTHERN NUCLEI ~1 OPERATING COMPANY VOGTLE ELECTRIC GENEi TING PLANT, UNITS 1 AND 2 FIRST CONTAINMENT n SPECTION INTERVAL j
REQUEST FOR FttELIEF NO. RR-E-1 (continued)
IV.
Basis for Relief: (continued)
The Containment Equipment Hatch is normally removed during refueling outages.
If gasket replacement is necessary, the new gasket will be visually inspected by maintenance personnel before re-assembly or closure. Prior to establishing containment integrity following the refueling outage, the Containment Equipment Hatch is leak rate tested in accordance with Appendix J.
These tests and inspections will assure the leak tightness of primary containment and provide an acceptable level of quality and safety.
V.
. Alternative Examination:
The leak-tightness of the seals and gaskets identified in section I will be tested in accordance with 10 CFR 50, Appendix J. If a seal or gasket becomes visually accessible, it will be visually inspected by maintenance personnel before reassembly or closure and 10 CFR 50, Appendix J leakage testing.
VI.
Justification for Granting Relief:
The functionality of the containment penetration seals and gaskets (including those of electrical penetrations) will continue to be verified during the Type B testing as required by 10 CFR 50, Appendix J. Proving the integrity of the seals and gaskets through Type B testing provides an acceptable alternative to the Code requirements. The alternative examinations are adequate to ensure the integrity of the subject seals and gaskets and will provide an acceptable level of quality and safety. As a result, relief should be granted under 10 CFR 50.55a(a)(3)(i).
- Vil, implementation Schedule:
The alternative examination will be performed, as conditions warrant, beginning with the First inspection Intervul for ASME Section XI Code, Subsection IVE-classified components or their equivalent that commenced prior to the end of the first inspection period (September 8, 2001).
4-5 Rev.0
o SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 1
FIRST CONTAINMENT INSPECTION INTERVAL REQUEST FOR RELIEF NO. RR-E-2 1.
System (s}iComponent(s) for Which Relief is Requested:
Class MC metallic components.
ll.
Code Requirement:
ASME Section XI,1992 Edition,1992 Addenda, IWE-2420(b) states: 'When component examination results require evaluation of flaws, areas of degradation, or repairs in accordance with IWE-3000, and the component is found to be acceptable for continued service, the areas containing such flaws, degradation, or repairs shall be reexamined during the next inspection period listed in the
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schedule of the inspection program of IWE-2411 or IWE-2412, in accordance with Table IWE-2500-1, Examination Category E-C."
ll1.
Code Requirement for Which Relief is Requested:
Relief is requested from the reexamination requirement after repair during the next inspection period.
IV.
Basis for Relief:
1 10 CFR Part 50.55a was amended in the Federal Register (61 FR 41303) to require the use of the ASME Section XI,1992 Edition,1992 Addenda, when performing containment examinations. The purpose of a repair is to restore the component to an acceptable condition for continued service in accordance with j
f the acceptanci. tandards of IWE-3000. When making repairs, IWA-4150 requires the owner to conduct an evaluation of the suitability of the repair including consideration of the cause of failure.
i Repairs are performed in accordance with IWA-4000, the intent of which is to use the construction code to restore the component to its original conditico where
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i practical. If a repair has restored the component to an acceptable condition, successive examinations are not warranted. If the repair was not suitable, then the repcir does not meet code requirements and the component is not acceptable for continued service; further repair work would be necessary. No similar requirement is found for Class 1,2, or 3 ASME Section XI repairs. Conducting successive examinations on components that have been repaired would result in hardship without a compensating level of qustity and safety. In addition, if the repair area is subject to accelerated degradation, the repair would still require augmented examinatica in accordance with Table IWE-2500- 1, Examination Category E-C.
4-6 Rev.O m
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SOUTHERN NUCLEAR OPERATING COMPANY i
l VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 FIRST CONTAINMENT INSPECTION INTERVAL REQUEST FOR RELIEF NO. RR-E-2 l
(continued) l V.
Proposed Alternative:
I Repairs will be performed in accordance with IWA-4000, to restore the component to an acceptable condition. Successive examinations as required by IWE-2420(b) will not be performed; however, successive examinations will continue to be done on those flaws or areas of degradation which have been accepted for continued service by evaluation.
VI.
Justification for Granting Relief:
Repairing components to restore the component to an acceptable condition provides adequate assurance of the integrity of the repair. Ccmpliance with the specified requirements of this section would result in hardship or unusual difficulty l
without a compensating increese in the level of quality and safety; therefore, relief should be granted under 10 CFR 50.55a(a)(3)(ii).
Vll.
Implementation Schedule:
The alternative examination will be performed, as conditions warrant, beginning with the First Inspection Interval for ASME Section XI Code, Subsection IWE-l classified components or their equivalent that commenced prior to the end of the l
first inspection period (September 8,2001).
l 4-7 Rev.0 1
o a4 SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 FIRST CONTAINMENT INSPECTION INTERVAL REQUEST FOR RELIEF NO. RR-E-3 1.
System (s)/ Component (s) for Which Relief is Requested:
)
Class MC (Metallic Containment) piessure-retaining bolting associated with the following components:
Bolted flanges on containment airlocks, Bolted flanges on small piping penetrations, and Bolted flanges on electrical penetrations.
11.
Code Requirement:
A3ME Section Xl,1992 Edition,1992 Addenda, Table IWE-2500-1, Examin'ation Category E-G, Pressure-retaining Bolting, item E8.20, requires bolt torque or tension testing on bolted connections that have not been disassembled and reassemb!ed during the inspection interval.
Ill.
Code Requirement from Which Relief is Requested:
Relief is requested from performing the Code-required bolt torque or tension testing on Class MC (Metallic Containment) pressure-retaining bolting.
IV.
Basis for Relief:
10 CFR Part 50.55a was amended in the Federal Register (61 FR 41303) to require the use of the ASME Section XI,1992 Edition,1992 Addenda, when performing containment examinations. Bolt torque or tension testing is required on IWE bolted connections that have not been disassembled and reassembled during the inspection interval, but is not required on any other ASME Section XI Class 1,2, or 3 bolted connections. This type of testing was removed from the 1998 ASME Section XI Code.
Bolted flanges on containment airlocks and bolted flanges on small piping penetrations each receive a 10 CFR Part 50, Appendix J, leak-rate test. The performance of the 10 CFR Part 50, Appendix J test itself proves that the bolt torque or tension remains adequate to provide a leak-rate that is within acceptable limits. The torque or tension value of bolts only becomes an issue if the leak-rate is excessive. Once a bolt is torqued or tensioned, it is not subject to dynamic f
Icading that could cause it to experience significant change. Appendix J testing and visualinspection is adequate to demonstrate that the design function is met, j
4-8 Rev.O
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 FIRST CONTAINMENT INSPECTION INTERVAL REQUEST FOR RELIEF NO. RR-E-3 (continued)
V.
Alternate Examination:
The following examinations and tests required by Subsection l\\AE ensure the
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structuralintegrity and the leak-tightness of Clase MC (Metallic Containment) pressure-retaining bolting, and, therefore, no acaitional alternative examinations 4
are proposed:
l (1)
When accessible, exposed surfaces of bolted connections shall be visually examined in accordance with the requirements of Table IWE-2500-1, l
Examination Category E-G, Pressure-retaining Bolting, Item No. E8.10, and (2)
Bolted connections shall meet the pressure test requirements of Table IWE-2500-1, Examination Category E-P, All Pressure-retaining Components (10 CFR 50, Appendix J leak-rate test).
VI.
Justification for Granting Relief:
The alternative examinations provide adequate assurance of the integrity of Class CC pressure-retaining bolting. As a result, relief should be granted under 10 CFR 50.55a(a)(3)(i) because the proposed alternative provides an acceptable level of quality and safety.
)
- Vil, implementation Schedule:
1 The attemative examination will be performed, as conditions warrant, beginning with the First inspection Interva' for ASME Section XI Code, Subsection IWE-classified components or their equivalent that commenced prior to the end of the first inspection period (September 8, 2001).
i I
l 4-9 Rev.O U
1 a
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 FIRST CONTAINMENT INSPECTION INTERVAL REQUEST FOR RELIEF NO. RR-E 4 1.
System (s)/ Component (s) for Which Relief is Requested:
All Class MC, Subarticle IWE-2200(g), preservice examination requirements of reapplied painted or coated containments.
11.
Code Requirement :
Subsection IWE-2600(b) of ASME Section XI,1992 Edition,1992 Addenda, requires that reapplied paint and coating systems shall be examined in accordance with IWE-2200(g). IWE-2200(g) requires that when paint or coatings are reapplied, the condition of the new paint or coating shall be documented in the preservice examination records.
Ill.
Code Requirement from Which Relief is Requested:
Relief is rinuested from the requirement to perform a preservice inspection of new paint or coatings.
IV.
Basis for Relief:
The paint and coatings on the containment pressure boundary were not subject to Code rules when they were originally applied and are not subject to ASME Section XI rules for repair or replaceme7t in accordance with IWA-4111(b)(5). The VEGP coating program, which is applied and inspected in accordance with a quality assurance program which meets the requirements of 10 CFR 50, Appendix B, verifies the adequacy of applied coatings. Recording the condition of reapplied coating in the preservice record does not substantiate the containment str.:ctural integrity. However, SNC acknowledges that the quality and integrity of coatings applied is relevant to the containment's functional integrity. This assurance is best accomplished by visually inspecting the coating, which is accomplished through the VEGP " Qualified (N) Coatings" program. Should deterioration of the coating in the reapplied area occur, the area would require additional evaluation regardloss of the preservice record. Recording the condition of new paint or coating in the preservice records does not increase the level of quality and safety of the containment.
In NRC SECY 96-033 dated April 17,1996, the Commission responded to Comment 3.2, which involves IWE-2200(g), by stating, "In the NRC's opinion, this does nct mean that a visual examination must be performed with every application of paint or coating. A visual examination of the topcoat to determine the i
soundness and the condition of the topcoat should be sufficient." The visual examination is currently accomplished through the VEGP " Qualified (N) Coatings" program.
4-10 Rev.0 I
7 1
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SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 FIRST CONTAINMENT INSPECTION INTERVAL REQUEST FOR RELIEF NO. RR-E-4 (continued)
V.
Alternative Examination:
The reapplied paint and coatings on the containment vessel will be examined in accordance with the VEGP " Qualified (N) Coatings" program. If degradation of the coating is identified, additional measures will be applied to determine if the containment pressure boundary is affected. Although repairs to paint or coatings are not subject to the repair / replacement rules of ASME Section XI based on the Code's response to inquiry 97-22, repairs to the primary containment boundary, if required, would be conducted in accordance with ASME Section XI Code nJles.
VI.
Justification for Granting Relief:
Coating inspection programs currently restore the coating to its original condition thereby providing adequate assurance of the integrity of the coating. As a result, relief should be granted under 10 CFR 50.55a(a)(3)(i) because the proposed alternative provides an acceptable level of quality and safety.
Vll.
Implementation Schedule:
The alternative examination will be performed, as conditions warrant, beginning with the First inspection Interval for ASME Section XI Code, Subsection lWE-classified components or their equivalent that commenced prior to the end of the first inspection period (September 8,2001).
4-11 Rev. O
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4 SOUTHERN NUCLEAR OPERATING COMPANY VO3TLE ELECTRIC GENERATING PLANT, UNITS 1 AND_2 FIRST CONTAINMENT INSPECTION INTERVAL
)
REQUEST FOR RELIEF NO. RR-E-5 1.
System (s)/ Component (s) for Which Relief is Requested:
)
All visual examinations of painted or coated containment components prior to removal of paint or coatings.
11.
Code Requirement:
Subarticle IWE-2500(b) of ASME Section XI,1992 Edition,1992 Addenda, requires that when paint or coatings are to be removed, the paint or coatings shall be visually examined in accordance with Table IWE-2500-1 prior to removal.
Ill.
Code Requirement from Which Relief is Requested:
Relief is requested from Subarticle IWE-2500(b), which requires that when paint or coatings are to be removed, the paint or coatings shall be visually examined in accordance with Table IWE-2500-1 prior to removal.
IV.
Basis for Rellef:
10 CFR 50.55a was amended, as cited in the Federa/ Register (61 FR 41303) to require the use of the 1992 Edition,1992 Addenda, of ASME Section XI when performing containment examinations. Paint and coatings are not part of the containment pressure boundary under current Code rules because they are not associated with the pressure-retaining function of the component. The interiors of containments are painted to prevent corrosion and to aid in contamination removal efforts. Paint and ccatings on the containment pressure boundary were not subject to Code rules when they were originally applied and are no' sut, ject to ASME Section XI rules for repair or replacement in accordance with IWA-4111(b)(5). Deterioration of the paint or coating materials, e.g., flaking, scaling, etc., on containment would be an indicator.of potential degradation of the containment pressure boundary. Additional measures would be employed to determine the nature and extent of any degradation, if present. The application of ASME Section XI rules for removal of paint or coatings when unrelated to an ASME Section XI repair or replacement activity, does not provide a compensating increase in the level of quality and safety.
j 4-12 Rev.O
r SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 FIRST CONTAINMENT INSPECTION INTERVAL REQUEST FOR RELIEF NO. RR-E-5 l
l (continued)
V.
Alternative Examination:
i The paint and coatings on the containment vessel will be examined in accordance with tne VEGP " Qualified (N) Coatings" program. If degradation of the coating is identified, additional measures will be applied to determine if the containment pressure boundary is affected. Although repairs to paint or coatings are not subject to the repair / replacement rules of ASME Section XI based on the Code's response to Inquiry 97-22, repairs to the primary containment boundary, if required, would be conducted in accordance with ASME Section XI Code rules.
VI.
Justification for Granting Relief:
Coating inspection programs as used at VEGP currently restore the coating to its original condition thereby providing adequate assurance of the integrity of the l
coating. As a result, relief should be granted under 10 CFR 50.55a(a)(3)(i) because the proposed alternative provides an acceptable level of quality and safety.
I Vll.
Implementation Schedule:
The alternative examination will be performed, as conditions warrant, beginning with the First Ir.cpection Interval for ASME Section XI Code, Subsection IWE-classified components or their equivalent that commenced prior io the end of the first inspection period (September 8,2001).
I l
4-13 Rev.O
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 FIRST CONTAINMENT INSPECTION INTERVAL REQUEST FOR RELIEF NO. RR-L-1 1.
System (s)/ Component (s) for Which helief is Requested:
The exterior portion of VEGP-1 and 2 Containment Buildings fabricated from concrete.
II.
Code Requirement:
ASME Code,Section XI,1992 Edition,1992 Addenda, requirement IWL-2310,
' Visual Examination and Personnel Qualification," and IWA-2210, ' Visual Examinations", require specific minimum illumination and maximum direct examination distance for all concrete surfaces and maximum procedure demonstration lower case cnaracter height. The concrete portion of the l
containment buildings at VEGP-1 and 2 are subject to the rules and requirements for Inservice inspection of Class CC (Concrete Containment) components, Examination Category L-A, Concrete, item L1.11, as applicable to IWL-2310,
' Visual Examination and Personnel Qualification," and IWA-2210, " Visual Examinations."
lli.
Code Requirement from Which Relief is Requested:
l Relief is requested from the IWL-2310 requirement to use the minimum illumination, maximum direct examination distance, and maximum procedure demonstration lower case character height specified in !WA-2210 and Table IWA-2210-1 for VT-3 examinations, when performing visuai examinations (VT-3C) of the concrete containment.
IV.
Basis for Relief:
The VT-3 requirements specified in IWA-2210 and Table IWA-2210-1 were developed for the examination of components such as Class 1 pump and valve bodies, the Class 1 reactor pressure vessel interior, Class 3 welded attachments, and Class 1,2, and 3 supports. VT-3 examinations are conducted to detenr.ine i
the general mechanical and structural condition of components and their supports j
by verifying parameters such as clearances, settings, and physical displacements.
j Additionally, VT-3 examinations are conducted to detect discontinuities and iinperfections, such as loss of integrity at bolted or welded connections, loose or j
missing parts, debris, corrosion, wear, or erosion. For these Class 1,2, and 3 components, small amounts of corrosion / erosion or small crac!;-like surface flaws may be detnmental to the structural integrity of the component; therefore, the stringent requirements of IWA-2210 and Table IWA-2210-1 are generally appropriate.
4-14 Rev 0
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 FIRST CONTAINMENT INSPECTION INTERVAL REQUEST FOR RELIEF NO. RR-L-1 (continued)
IV.
Basis for Relief: (continued)
However, it was recogn, zed by the industry and the NRC during the development of the implementing 10 CFR 50.55a rules that IWA-2210 and Table IWA-2210-1 requirements were excessively stringent for the IWE-required examination of the j
metal portion of the containment. Therefore, the NRC changed the recuirements to allow that "When periorming remotely the visual examinations required by Subsection IWE, the maximum direct aistance specified in Table IWA-2210-1 may be extended and the minimurn illumination requirements specified in Table IWA-2210-1 may be decreased provided that the conditions or indications for which the visual examination is performed can be detected at the chosen distance and illumination" SNC has concluded that, similar to the consideration used for the IWE examinations, the use of the VT-3 requirements found in IWA-2210 and Table IWA-2210-1 when performing VT-3C examinations of the concrete surfaces is also excessively stringent and should not ce applied. This is based on the recognition that due to the nature of concrete, a concrete containment will have numerous, small" shrinkage-type" surface cracks or other imperfections that are not detrimental to the structural integrity of the containment. The application of IWA-2210 and Table IWA-2210-1 " minimum illumination requirements", " maximum direct visual examination distance requirements", and " maximum procedure demonstration lower case character height requirements" to attempt to identify these small " shrinkage-type cracks" or other imperfections is considered to be unnecessary and cou!d result in a large number of man-hours erecting scaffolding, using lifts, evaluating insignificant indications, etc.
Per the requirements of IWL-2320, the Registered Professional Engineer (RPE) is experienced in evalueting the inservice condition of structural concrete and is knowledgeable of the design and Construction Codes and other criteria used in design and construction of concrete containments. The RPE will use experience and training to determine the necessary requirements to detect indications that are detrimental to the containment integrity.
V.
Proposed Alternative:
)
VT-3C examinations will be performed as required by IWL-2310 except that instead of using the minimum illumination, maximum direct examination distance, and maximum procedure demonstration lowt case character height requirements specified in IWA-2210 and Table IWA-2210-1 for VT-3 examinations, the recommendations of the RPE 4
for illumination and riistance will be implemented.
4-iG Rev.0
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 FIRST CONTAINMENT INSPECTION INTERVAL REQUEST FOR RELIEF NO. RR-L-1 (continued)
VI.
Justification for Granting Relief:
Section XI relies on the knowledge and experience of the RPE as a key element for an IWL visual inspection program. Examining the concrete surfaces using distances and illumination requirements, established by a knowledgeable RPE, would provide for detection of flaws of sufficient size to assure that the structural integrity of the concrete containment is being maintained. As a result, relief should be granted under 10 CFR 50.55a(a)(3)(i) because the proposed alternative provides an acceptable level cf quality and safe.y.
Vll.
Implementation Schedule:
The alternative examination will be performed, as conditions warrant, beginning with the First inspection interval for ASME Section XI Code, Subsection IWL-classified components or their equivalent that commenced prior to the end of the first inspection period (September 8,2001).
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4-16 Rev.O
j SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 f
FIRST CONTAINMENT INSPECTION INTERVAL
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REQUEST FOR RELIEF NO. RR-L-2 i
1.
System (s)/ Component (s) for Which Relief is Requested:
The exterior portion of the VEGP-1 and 2 Containment Buildings fabricated from concrete and their post-tensioning systems.
II.
Code Requirement:
Subarticle IWL-2400 of ASME Code,Section XI,1992 Edition,1992 Addenda, requires that the concrete and post-tensioning system be examined at dates coinciding with the anniversary of the Structural Integrity Test date.
Ill.
Code Requirement from Which Relief is Requested:
Concrete shall be examined in accordance with IWL-2510 at 1,3, and 5 years following the completion of the containment Structural Integrity Test CC-6000 and every 5 years thereafter.
Unbonded post-tensioning systems shall be examined in accordance with IWL-2520 at 1,3, and 5 years following the completion of the containment Structural Integrity Test and every 5 years thereafter.
IV.
Basis for Relief:
Due to the expense of : cpairing, inspecting, and testing the tendon surveillance platforms, Georgia Power Company, the former operator and licensee of VEGP and sister company to SNC, the current operator and licensee, requested a change to the Technical Specifications for VEGP-1 and 2 to allow testing of both units during the same time period. The requested change was authorized by the NRC in their letter dated September 12,1989, to Georgia Power Company, and resulted in Amendments 23 and 4 to the Technical Specifications for VEGP-1 and 2, respectively. The use of the above approved Technical Specification change will continue to provide assurance of containment structuralintegrity.
V.
Proposed Alternative:
The next tendon and concrete examinations will be performed on VEGP-1 and 2 by August 1, 2000 i 1 year and every 5 years i 1 year thereafter.
4-17 Rev.O
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SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRid GENERATING PLANT, UNITS 1 AND 2 FIRST CONTAIN'.1ENT INSPECTION INTERVAL REQUEST FOR RELIEF NO. RR-L-2 (continued)
)
VI.
Jystification for Granting Relief:
Yhe frequency of testing is equivalent to the requirements of the referenced edition and addenda of the ASME Section XI Code. As a result, relief shoui<1 be granted under 10 CFR 50.55a(a)(3)(i) because the proposed attemative provides an acceptable level of quality and safety.
Vll.
Implementation Schedule:
VEGP Technical Specifications SR 3.6.1.2 and 5.5.6 currently implement the alternative examination schedule.
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i 4-18 Rev.O 4
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SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 FIRST CONTAINMENT INSPECTION INTERVAL REQUEST FO' LIEF NO. RR-L-3 1.
System (s)/ Component (s) for Which Relief is Requested:
I VEGP-2 tendon strands.
II.
Code Requirement:
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Iv02523 of ASME Code,Section XI,1992 Edition,1992 Addenda, requires that a strand sample be visually examined and tensile tested.
Ill.
Code Requirement from Which Re"ef is Requested:
IWL-2523.1 requires that one sample tendon, from each type, be detensioned and a single strand removed.
IWL-2523.2 requires that the strands selected in IWL-2523.1 are tensile tested and visually examined.
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IV.
Basis for Relief:
The VEGP post-tensioning systems are designed so that no tendons on Unit 2 and only a sample on Unit 1 can be detensioned without creating voids in the sheathing filler material. VEGP was originaii/ icensed with t' e NRC so that l
n tendon lift-off and strand testing would be performed on Unit 1 and that no lift-off and strand testing would be performed or required on the post-tensioning system of Unit 2. Presently, even though not originally required, VEGP is performing lift-off testing on the Unit 2 tendons but cannot perform strand testing in conjunction with lift-off testing due to the Unit 2 tendon design.
V.
Proposed Alternative:
Amendments 23 and 4 to the Plant Technical Specifications for VEGP-1 and 2, respectively, were authorized by the NRC in their letter dated September 12,1989, to Georgia Power Company, the former licensee and operator of VEGP and sister company to SNC, the current licensee and operator. Based on these particular amendments being issued for the VEGP Technical Specifications, the following testing is proposed as an alternative to the ASME Section XI Code requirements:
During lift-off testing of Unit 2, one tendon on Unit 1 will be strand tested in accordance with IWL-2523 as credit for Unit 2. The type tendon strand tested will be alternated between hoop and inverted-U tendons.
4-19 Rev.O
i SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 FiRST CONTAINMENT INSPECTION INTERVAL REQUEST FOR RELIEF NO. RR-L-3 (continued)
VI.
Justification for Granting Relief:
The VEGP post-tensioning system was reviewed and licensed with the NRC as reflected by the issuance of Amendments 23 and 4 to the Plant Technical Specifications for VEGP-1 and 2, respectively. Because of the design of the tendons on Unit 2, they can not be detensioned thereby preventing compliance with the requirements of the ASME Section XI Code for tendon strand testing. As a result, relief from the Code requirements should be granted under 10 CFR 50.55a(g)(6)(i) because they are impractical and cannot be complied with without a redesign of the plant due to the design of the post-tensioning system for Unit 2.
The proposed alternative, which is based on testing currently approved by the NRC in the VEGP Technical Specifications, ensures that the structural integrity of the containment is being maintained.
Vll.
Implementation Schedule:
VEGP Technical Specifications SR 3.6.1.2 and 5.5.6 currently implement the l
alternative testing and will be used during the first inspection interval for i
containment testing.
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4-20 Rev.O
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e, ENCLOSURE 2 TO SOUTliERN NUCLEAR OPERATING COMPANY LETTER LCV-1307 VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 REQUEST FOR RELIEF FROM THE REGULATORY REQUIREMENTS OF 10 CFR 50.55a(b)(2)(ix)(A) and (D) 1 I
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a, ENCLOSURE 2 TO SOUTHERN NUCLEAR OPERATING COMPANY LETTER LCV-1307 (continued)
Regulatory Requirement for Which Reliefis Requested:
Southern Nuclear Operating Company (SNC) hereby proposes an alternative to the requirements of 10 CFR 50.55alb)(2)(ix)(A) (as interpreted by SNC) and (D) for Vogtle Electric Generating Plant (VEGP), Units 1 and 2, as they pertain to the area of tendon grease leakage and the reporting thereof. SNC requests NRC authorization for the alternative discussed herein which is based on the requirements of the existing VEGP l
Technical Specifications.
Basis for Relief:
10 CFR 50.55a(b)(9(ix)(A) requires:
" Grease caps that are accessible must be visually examined to detect grease leakage or grease cap deformation. Grease caps must be removed for their examination when there is evidence of grease cap deformation that indicates deterioration of anchorage hardware."
10 CFR 50.55a(b)(2)(ix)(D) requires:
"The licensee shall report the following conditions, if they occur, in the ISI Summary Report required by IWA-6000:
(1) The sampled sheathing filler grease contains chemically combined water exceeding 10 percent by weight or the presence of free water; l
(2) The absolate difference between the amount removed and the amount replaced exceeds 10 percent of the tendon net duct volume; (3) Grease leakage is detected during general visual examination of the containment surface."
10 CFR 50.55a(b)(2)(ix)(A) does not explicitly state that grease caps that are accessible, visually examined, and are observed to have grease leakage must be addressed in the ISI Summary Report required by IWA-6000. However, SNC believes that it is the intent of the NRC that licensees address any such tendon grease leakage identified in the ISI E2-1
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ENCLOSURE 2 TO SOUTilERN NUCLEAR OPERATING COMPANY LETTER LCV-1307 (continued)
Basis for Relief (continued):
Summary Report. Conversely,10 CFR 50.55a(b)(2)(ix)(D) explicitly states that certain conditions, including grease leakage detected during general visusi examinations of the containment, must be addressed in the ISI Summary Report. The Rule does not limit ISI reporting of grease leakage to only that leakage that is determined to be excessive.
VEGP Technical Specification 5.6.9 requires that "any abnormal (emphasis added) degradation of the containment structure detected during the tests required by the Prestrer. sed Concrete Containment Tendon Surveillance Program shull be reported to the NRC within 30 days". Further, that Technical Specification requires that "the report shall include the condition of the concrete (especially at tendon anchorages). the inspection procedure, the tolerances in cracking, and the corrective action taken". Excessive grease leakage could be viewed as abnormal degradation. Ilowever, small amounts of grease leakage are not considered by SNC to be abnormal and are believed to be common to unbonded post-tensioned concrete structures such as the containments at VEGP.
The Concrete Containment Tendon Surveillance Program used at VEGP is performed to the requirements of NRC Regulatory Guide (RG) 1.35, Revision 2, with the VEGP position on the regulatory guide noted in VEGP Updated Final Safety Analysis Report sections 1.9.35.2 and 3.8.1.7.2. This program and the reporting requirenients therefor have been previously approved by the NRC for use at VEGP. As a result, their continued use should be acceptable. Visual examinations of containment concrete to fulfill the requirements of ASME Section XI, Subsection IWL, are to be performed concurrently with the containment tendon testing.
The tendon testing program used at VEGP is implemented in accordance with VEGP Maintenance Department Manual GEN-100," Tendon Surveillance Program Manual",
which provides the tendon surveillance program requirements as referenced in the Bases for VEGP Technical Specification Surveillance Requirement (SR) 3.6.1.2. SR 3.6.1.2 states:
"For ungrouted, post-tensioned tendons, this SR ensures that the structural integrity of the contaimnent will be maintained in accordance with the provisions of the Containment Tendon Surveillance Program. Testing and Frequency are consistent with the recommendations of Regulatory Guide 1.35 (Ref. 4) and approved exceptions "
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ENCLOSURE 2 TO SOUTIIERN NUCLEAR OPERATING COMPANY LETTER LCV-1307 (continued)
Basis for Relief (continued):
i i
Reference 4 as cited in SR 3.6.1.2 is NRC Regulatory Guide 1.35, Revision 2.
J Associated plant procedures for conducting the containment tendon surveillance program I
included 25044-1 (for VEGP-1)," Containment Tendon Grease Leakage Tracking",
25044-2 (for VEGP-2) which is similarly titled to 25044-1, and 28235-C, " Containment Tendon Structural Integrity Test". The 25044-designated procedures are unit-specific and log locations in the respective unit to have grease leakage in excess of 7 gallons.
Seven gallons equates to approximately three percent (3%) of the tendon net duct volume. Procedure 28235-Chrovides administrative controls to meet the intent of the Technical Specifications surveillance requirements, including repor'ing degraded conditions, and to ensure that the structure 1 integrity of the containment is maintained.
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l Excessive grease leakage, as addressed in GEN-100, is considered to be an indication of potential abnormal containment degradation and is required to be reported to the NRC in a 30-day report pursuant to VEGP Technical Specification 5.6.9.
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Alternative to Regulatory Requirement:
In lieu of submitting information to the NRC in the ISI Summary Report concerning tendon grease leakage, SNC will report only those occurrences where leakage exceeds 7 gallons. Such occurrences of grease leakage exceeding seven gallons will be reported to l
the NRC pursuant to existing VEGP Technical Specification 5.6.9, i.e., within 30 days of l
identifying any abnormal degradation of the containment structure. The Technical i
Specification-required repon will constitute the report required by 10 CFR 50.55a(b)(2)(ix)(A) (as interpreted by SNC) and (D). The 30-day report required by Technical Specifications will be referenced in the ISI Summary Report by SNC letter number and date; however, a separate report will not be submitted to the NRC for fulfilling the requirement (s) of 10 CFR 50.55a(b)(2)(ix)(A) (as interpreted by SNC) and
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(D) since to do so would be redundant.
Justification for Granting Relief:
Abnormal degradation of the containment structure, including that due to any excessive tendon grease leakage, is already reportable to the NRC as required by VEGP Technical Specification 5.6.9. Grease leakage in excess of 7 gallons is currently reported by SNC for VEGP. Such reports are required to be submitted to the NRC within 30 days after identifying any abnormal degradation of the containment structure. Submittal of E2-3
r cc.s 3.
ENCLOSURE 2 TO SOUTIIERN NUCLEAR OPERATING COMPANY LETTER LCV-1307 (continued)
Justification for Granting Relief (continued):
i reports within the timeframe specified by VEGP Techriical Specification 5.6.9 is more conservative than the reporting requirement (s) of 10 CFR 50.55a(b)(2)(ix)(A)(as interpreted by SNC) and (D) that require submittal ofinformation to the NRC by means of the ISI Summary Report. ISI Summary Reports are required to be submitted pursuant to ASME Section XI, IWA-6000, within 90 calendar days from the end of each refueling outage in which inservice inspection activities may be conducted. As noted herein, tendon testing activities at VEGP are currently scheduled such that they will likely be performed during a non-outage period unlike normal inservice inspection activities which are typically performed immediately prior to or during maintenance / refueling outages.
The alternative discussed herein for the submittal of reports purmant to VEGP Technical Specification 5.6.9 is more conservative than the requirements.a 10 CFR
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50.55a(b)(2)(ix)(A)(as interpreted by SNC) and (D). As a result, the alternative reporting discussed herein maintains, and perhaps, even increases, the level of quality and safety over that required by the aforementioned 10 CFR 50.55a parag aphs. Accordingly, NRC authorization of the alternative grease leakage reporting and submittal of reports in accordance with existing VEGP Technical Specification requirements, i. e., within 30 days, is requested pursuant to 10 CFR 50.55a(a)(3)(i) in lieu ofi svoking the regulatory requirements. SNC will reference the 30-day report in the ISI Summary Report by SNC letter number and date; however, no separate report will be provided in the ISI Summary Report.
Schedule for Implementation I
VEGP Technical Specification SR 3.6.1.2 and 5.6.9 currently implement the alternative reporting requirements, including reperting of excessive grease leakage, and will continue to be used during the first interval for containment testing.
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