ML20141A143
| ML20141A143 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 06/16/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20141A134 | List: |
| References | |
| NUDOCS 9706200118 | |
| Download: ML20141A143 (9) | |
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c, UNITED STATES NUCLEAR REEULATORY COMMISSION If WASHINGTON, D.C. 2066fr0001 kg. ... . j/
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION THE THIRD TEN YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN.
_AND ASSOCIATED RE0 VESTS FOR RELIEF SOUTHERN NUCLEAR OPERATING COMPANY. INC.
EDWIN I. HATCH NUCLEAR PLANT. UNITS 1 AND 2 DOCKET NOS. 50-321 AND 50-366 l
1.0 INTRODUC110N The Technical Specifications for Edwin I. Hatch Nuclear Plant (Hatch),
Units I and 2, state that the inservice inspection of the American Society of Mechanical Engineers (ASME) Code Class 1, 2, and 3 components shall be I performed in accordance with Section XI of the ASME Boiler and Pressure Vessel
- Code (ASME Code) and applicable Addenda as required by Title 10 of the Code of federal Regulatfons (10 CFR) Section 50.55a, except where specific written
, relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(1),
or alternatives approved pursuant to 10 CFR 50.55a(a)(3). Section 50.55a(a)(3) states that alternatives to the requirements of paragraph (g) may .
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' be used, when authorized by the NRC, if (1) the proposed alternatives would )
provide an acceptable level of quality and safety or (ii) compliance with the i sper.ified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality .and safety.
l Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) shall 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," to the extent practical within the limitations of design,
, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The applicable edition of Section XI of the ASME Code for Hatch Units 1 and 2, third 10-year inservice inspection (ISI) interval is the 1989 Edition. The third 10-year ISI interval began January 1,1996, for Hatch Units 1 and 2.
. Enclosure 1 9706200118 970616 PDR ADOCK 05000321 G PDR
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i Pursuant to 10 CFR 50.55a(g)(5), if the licensee determines that conformance with an examination requirement of Section XI of the ASME Code is not practical for its facility, information shall be submitted to the Comission in support of that determination and a request made for relief from the ASME I Code requirement. After evaluation of the determination, pursuant to 10 CFR 50.55a(g)(6)(1), the Commission may grant relief and may impose alternative requirements that are determined to be authorized by law, will not i endanger life, property, or the common defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed.
1 j By letter dated October 17, 1995, as supplemented January 26, April 5, l June 4, August 13, and November 18, 1996, Georgia Power Company submitted to I the NRC its third 10-year ISI program plan and associated requests for relief, and responded to the NRC staff's requests for additional information for Hatch
. Units 1 and 2. A sumary of the requested reliefs is attached.
2.0 EVALUATION AND CONCLUSIONS l The staff, with technical assistance from its contractor, the Idaho National i Engineering and Environmental Laboratory (INEEL), has evaluated the information provided by the licensee in support its third 10-year interval ISI plans and associated requests for relief for Hatch Units 1 and 2. Based on the information submitted, the staff adopts the contractor's conclusions and recomendations presented in the Technical Evaluation Report (TER), INEL-96/0188, Revision 1, dated December 1996 (Enclosure 2). The staff's evaluation of Request for Relief (RR) No. 11 is contained in Enclosure 3.
Based on the information provided, the staff concludes that no deviations from regulatory requirements or comitments were identified in the third 10-year interval ISI program plan for Hatch Units 1 and 2.
For RR Nos. 03, 05, and 06 the staff concludes that the Code requirements contained in these requests are impractical and that the licensee's proposed ;
testing provides reasonable assurance of operational readiness of the subject systems. Based on the impracticality of complying with the Code requirements L and the burden on the licensee if the Code requirements were imposed, the staff has concluded that pursuant to 10 CFR 50.55a(g)(6)(1) relief is granted for Requests for Relief Nos. RR-03, RR-05, and RR-06 as requested. The relief granted is authorized by law and will not endanger life or property or the comon defense and security and is otherwise in the public interest giving due l
consideration to the burden that could result if the requirements were imposed on the facility.
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i Also, the staff has concluded that the alternatives contained in RR-01, RR-04, RR-07, RR-11, RR-12, RR-13, RR-14, provide an acceptable level of quality and 1 l safety. Therefore, the alternatives contained in the above requests for l
relief are authorized pursuant to 10 CFR 50.55a(a)(3)(1) as requested.
In addition, the staff has concluded that requiring the licensee to comply with the Code requirements contained in RR-02, RR-09, RR-10, RR-15, and RR-16 would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The licensee's proposed alternatives provide reasonable assurance of operational readiness of the subject systems.
l Therefore, the licensee's proposed alternatives are authorized pursuant to
- 10 CFR 50.55a(a)(3)(ii) as requested.
l Code Cases N-498-1, N-509, N-522, N-524, N-416-1, and N-523 contained in RR-02, RR-04, RR-07, RR-09, RR-13, and RR-14, respectively, are authorized for the current interval or until such time as the Code Cases are published in a future revision of Regulatory Guide 1.147. At that time, if the licensee intends to continue to implement these Code Cases, the licensee is to follow all provisions in the above Code Cases, with limitations issued in Regulatory Guide 1.147, if any. Relief is not required for RR-08.
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Attachment:
Summary of Relief Requests Print.ipal Contributor: T. McLellan Date: June 16,1997 1
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EDWIN I. HATCH NUCLEAR PLANT, UNITS 1 & 2 Page 1 of 3 Third 10-Year ISI Interval TABLE 1
SUMMARY
OF RELIEF REQUESTS Relief Relief Request System or Exas Itese iPoteme or Aree to be Recysired Respsest thmber Component Category No. Emmerined seethod Licerisee Preposed Alterviettw . Status RR-01 Reactor B-G-1 B6.10 Closure Mead Nuts Surface VT-1 Visual Emmuinstion Authorized Pressure Vessel RR-02 Ccde Ctess 1, B-E B4.11 Nydrostatic fest Bomderies Hydrostatic Apply Alternatives Contained Authorized -
2, and 3 84.12 Test in Code case N-498-1 Systens B-P 84.13 815.11 ,
815.51 815.61 815.71 ,
C-H C7.20 C7.40 C7.60 C7.80 0-A D1.10 D-B D2.10 0-C D3.10 RR-03 Reactor B-D B3.90 2Ns Bottom need Drain Voltmetric VT-2 visual examinetton Granted Pressure vessel 83.100 Vecset-to-Nozzle Welds N15 eruf 2N15 !
2 W S Bottom Head Drain Vessel Inside Radius Section .
N15 and 2N15 c+
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EDWIN I. HATCH NUCLEAR PLANT, UNITS 1 1 2 Page 2 of 3 - :
Third 10-Year ISI Interval TABLE I
SUMMARY
OF RELIEF REQUESTS s
Retlef Retlef Regsest system er Eman Iteus Volume or Area to be Respired Respeat .
Ihmber rW Category llo. Esamined Metfood Licensee Prupaaed Attemative Status RR-04 Class 1, 2, and B-K-1 810.10 Integretty Welded Votunetric or Apply Alternatives Contained Authorized 3 810.20 ' Attachments Surface in Code Case N-509 C-C C3.10 Surface C3.20 D-A D1.20 VT-3 Visuat thru Examination ,
D1.60 0-8 D2.20 thru i D2.60 D-C D3.20 thru D3.60 RR-05 Class 2 C-A C1.20 Residust nest Removat Heat Volumetric Votunetric and St@ptemental Granted Pressure Exchanger need Welds: Surface to the Extent Vessels Practicas Shett Need-to-Upper Shell Ring Welds 1E11-2MX-A-1, 1E11-2HX-B-1, 2MX-A-1, 2HX-B-1 RR-06 Reactor B-F B5.20 2.5" Core D.P. & Liquid Surface VT-2 Visual Examination Granted Pressure Vessel Control Nozzle-to-Safe-End Welds N10 and 2N10 3" Bottom Need Drain Nozzle-to-Safe-End Welds N15 ,
and 2N15 2.5" RPV Instrumentation Nozzle-to-Safe-end Welds N164, 2N16A, N168, and 2N168 RR-07 Class 1 & 2 B-J E9.12 Longitudinal Welds Volumetric Apply Code case N-524 Authorized Piping C-F-2 C5.52 and Surface C5.82 as applicable
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EDWIN I. MATCH NUCLEAR PLANT, UNITS 1 & 2 Page 3 of 3 - - ;
Third 10-Year ISI Interval TABLE 1 i SUMARY OF RELIEF REQUESTS r
-antief I setief . .
! Regsmet Systese er Itein Regsfred .
Bogsest Ihh Compenant Eeme Category to. Voteme or Aree to be Esemined Itethod Lleanese Preposed Attemative Statens !
l RR-10 Class 1 and 2 tum-2600 NA Weld Reference System NA Develop Reference for Each Authorized ,
Piping and Enemination Aree m en Enseined Components j RR-11 Snubbers NA IIA NA NA IIA Authorized
, (see evolustion I by r IIRC/9ES)
RR-12 class 1 and 2 IWA- IIA Botted Comections Remove Att Evolustions of Botted Connection Authorized Piping and 5250(e)(2) sotting At i Components Leeking i Connection For i Evoluetion t
RR-13 Ctess 2 Piping C-N C7.10 Penetration Piping Pressure Test Appendix J Authorized C7.30
. C7.50 C7.70 ;
RR 14 Recording and IWA-6220 NA NA NA Code Case N-523 Authorized Reporting IWA-6230 I
RR-15 Ctess 2 Piping C-H C7.40 Nigh Pressure cootent Pressure Test Perform pressure test in Authorized C7.60 Injection System Piping conjunction with the System '
Functional Test !
RR-16 Class 3 Piping D-A D1.10 Safety Relief Velve Piping Pressure Test 100 Alternettve Authorized a %
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20066 4 001
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j SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION 1
RELATED TO THE INSERVICE INSPECTION PROGRAM RE0 VESTS FOR RELIEF :
SOUTHERN NUCLEAR OPERATING COMPANY. INC. ,
l EDWIN I. HATCH NUCLEAR PLANT. UNITS 1 AND 2 i
DOCKET NOS. 50-321 AND 50-366 i
1.0 INTRODUCTION
Title 10 of the Code of federal Regulatfons (10 CFR) Section 50.55a, requires that inservice inspection (ISI) of certain Code Class 1, 2, and 3 components be aerformed in accordence with Section XI of the American Society of Mec1anical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code and applicable Edition and Addenda, except where specific written relief has been l requested by the licensee and granted by the Commission pursuant to 10 CFR i l 50.55a(g)(6)(1), or alternatives approved pursuant to 10 CFR 50.55a(a)(3). In proposed alternatives, the licensee must demonstrate that: (i) the proposed alternatives provide an acceptable level of quality and safety; or (ii) compliance would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. In requesting relief, the licensee must demonstrate that the requirement is impractical for ;
, their facility. NRC guidance contained in Generic Letter (GL) 90-09, I l Alternative Requirements for Snubber Visual Inspection Intervals and l l Corrective Actions, provides alternatives to the Code requirements determined J i to be acceptable to the NRC staff.
Section 50.55a authorizes the Commission to grant relief from ASME Code l requirements upon mak'ing the necessary findings. The NRC staff's findings l
with respect to granting or not granting the relief request as part of the ;
licensee's ISI program are contained in this Safety Evaluation (SE). 1 l This SE covers a request for relief from the ASME B&PV Code,Section XI, i Subarticle IWF-5300(a) and (c), inservice visual examination requirements.
! The licensee proposes an alternate visual examination method, as described in relief request RR-11, submitted by Georgia Power Company's (GPC) letter dated October 17, 1995, with additional information provided by letter dated August 13, 1996. The licensee's ISI program is based on the requirements of Section XI of the ASME B&PV Code, 1989 Edition. The 1989 Edition of the ASME B&PV Code,Section XI, Subarticles IWF-5300(a) and (c), require that snubber inservice examinations be performed in accordance with the first Addenda to ASME/ ANSI OH-1987, Part 4, published in 1988 (OM-1988, Part 4), using the VT-3 visual examination methods described in paragraph IWA-2213.
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The ASME Code,Section XI, requires personnel performing nondestructive
- examinations be qualified and certified as described in Subarticle IWA-2300.
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! 2.0 RELIEF REOUEST RR-11 t
l The licensee requests relief from the ASME B&PV Code,Section XI, requirement i i
to use certified personnel in the performance of the visual examination of 1 snubbers and their attachments, excluding welded attachments, using the VT-3 i methods described in paragraph IWA-2213. ,
- 2.1 Licensee's Basis For Reauested Relief l
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i A selected group of GPC Maintenance Department personnel have received special j training in the examination and testing of snubbers. The training consisted 3 of specialized classes, presented by various snubber vendors and contractors, 1
! that included: (1) training for identification of potential problems and )
deficient conditions relative to snubber operability; (2) instructions in '
4 snubber repair and complete overhaul; and (3) training in performance of i snubber functional testing. Individuals from the selected group have been far i more involved and have accumulated many more hours of experience in activities
- associated with the removal, installation, examination, repair, ove/ haul, and
! testing of snubbers than a typical ASME,Section XI, VT-3 certified inspector.
Utilization of selected maintenance personnel, specifically trained to
} inspect, test, repair, remove, install, and overhaul snubber supports to i perform the visual examinations in conjunction with the Site Snubber Program
! will: (1) eliminate a redundant inspection by VT-3 certified personnel, reduce resource expenditures, and radiation exposure; (2) provide personnel with i extensive experience in snubber applications and maintenance; (3) meet all aspects of Section XI, Article IWF requirements other than certification of i inspection personnel; (4) provide reasonable assurance that unallowable
- inservice flaws have'not developed or that they will be detected and repaired
- prior to returning the reactor to service; and (5) provide an acceptable level of quality and safety and not endanger public health and safety. -
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J 2.2 Alternate Method Maintenance personnel who have been trained to recognize potential problems and deficient conditions, specifically applicable to snubbers, and who have been hvolved with the snubber functional testing program will be used to '
i perta.a visual examinations required by Subarticles IWF-5300(a) and (c) and i Table IWF-2500-1 (excluding welded attachments).
1 The qualification of personnel will be in accordance with the GPC Quality Assurance Program for training and qualification of plant personnel. The selected personnel will have all training documented to the training program requirements. The inspector'sSection XI, Article IWF-5000, activities will i be documented in the Site Snubber Program. Further, the GPC snubber i examination procedures require that all data sheets be reviewed by the site
! snubber engineer for concurrence and resolution of any reported snubber I
condition.
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2.3 Evaluation 1
Section XI, Subarticles IWF-5300(a) and (b), titled, Inservice Examinations and Tests, in part, require that inservice inspection of snubbers and their integral and nonintegral attachments including lugs, bolting, pins, and clamps must be examined using the VT-3 visual examination method described in paragraph IWA-2213. Paragraph IWA-2213, titled, Visual Examinations VT-3, in part, requires that the VT-3 visual examination be conducted to determine l the general mechanical and structural condition of components and their i
supports and include examinations for conditions that could affect operability or functional adequacy of snubbers and constant load or spring-type supports.
Further, paragraph IAW-2321, titled, Visual Acuity, in part, provides the personnel vision requirements and qualifications for VT-3 examiners.
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- .The licensee proposes to use selected maintenance personnel, specifically l trained and experienced i'n activities of removal, installation, examination, I repair, overhaul, and testing of snubbers to perform the VT-3 visual examinations, in lieu of certified inspectors. This alternative will reduce personnel radiation exposure and resource expenditures. The inspections will be limited to the visual examination of snubbers and their associated attachments, excluding the welded attachments. The selected personnel's qualifications and training will be in accordance the GPC Quality Assurance Program and documented. All snubber examination results will be documented in the Site Snubber Program procedures. GPC procedures require the site snubber engineer to review and resolve any reported snubber deficiency condition.
Further, by GPC letter dated August 13, 1996, in response to the NRC's request for additional information, the licensee states that the maintenance personnel who will perform the snubber inspections are required to have annual eye examinations, which meet the ASME Code,Section XI, paragraph IWA-2321 requirements.
Based on a consideration of: (1) the licensee's proposed use of experienced personnel explicitly
- trained for snubber visual examination; (2) the GPC Quality Assurance Program's qualification, documentation, and training requirements; (3) the resulting reduction of unnecessary radiation exposure of i plant personnel; and (4) the level of quality and safety provided, the staff l has determined that the proposed alternative is acceptable in that it will l i provide a sufficient means to detect the condition of snubbers. Compliance !
l with the required ASME Code use of certified inspectors would result in l
hardship without a commensurate increase in the level of quality and safety.
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3.0 CONCLUSION
j The staff concludes that the licensee's proposed alternative to use selected qualified personnel for visual examination of snubbers and their attachments in lieu of Code-required certified inspectors is authorized pursuant to 10 CFR 50.55a(a)(3)(ii), based on the determination that compliance with the Code requirements would result in a hardship without a compensating increase in the l 1evel of quality and safety.
t l Principal Contributor: F. Grubelich Date: June 16, 1997 '
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