ML20212F083
ML20212F083 | |
Person / Time | |
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Site: | Sequoyah ![]() |
Issue date: | 09/23/1999 |
From: | NRC (Affiliation Not Assigned) |
To: | |
Shared Package | |
ML20212F082 | List: |
References | |
NUDOCS 9909280022 | |
Download: ML20212F083 (5) | |
Text
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION j
RELATED TO THE RELIEF FROM CERTAIN WELD INSPECTIONS SEQUOYAH NUCLEAR PLANT. UNITS 1 AND 2 4
TENNESSEE VALLEY AUTHORITY DOCKET NOS. 50-327 AND 50-328
1.0 INTRODUCTION
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The inservice inspection (ISI) of the Amer:can Society of Mechanical Engineers (ASME) Code Class 1,2, and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code (the Code) and applicable addenda as required by Title 10, Code of Federal Reoulations, Part 50,10 CFR 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). Relief is provided for j
by 10 CFR 50.55a(a)(3), which states that alternatives to the requirements of paragraph (g) may be used, when authorized by the U.S. Nuclear Regulatory Commission (NRC), if (i) the i
proposed alternatives would provide an acceptable level of quality and safety or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
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 ASME Code,Section XI, for Sequoyah Plant's second 10-year ISI interval is the 1989 Edition. The components (including supports) may meet the requirements set forth in subsequent editions and addenda of the ASME Code incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein and subject to Commission approval.
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Pursuant to 10 CFR 50.55a(g)(5), if the licensee determines that conformance with an examination requireme,t of Section XI of the ASME Code is not practical for its facility, ENCLOSURE 9909290022 990923 PDR ADOCK 05000327 P
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2-information shall be submitted to the Commission in support of that determination and a request made for relief from the ASME Code requirement. After evaluation of the determination, pursuant to 10 CFR 50.55a(g)(6)(i), the Commission may grant relief and may impose alternative requirements that are determined to be aathorized by law, will not endanger life, property, or the common defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed.
By letter dated October 21,1998, the Tennessee Valley Authority (TVA), licensee for Sequoyah Nuclear Plant, submitted a request for relief from the VT-3 visual examination ofintegrally welded attachments required by the applicable ASME Section XI Code for 1-inch and smaller auxiliary feedwater (AFW) piping during the second ISI interval for Sequoyah, Units 1 and 2.
The licensee has proposed to adopt to the 1991 Addenda to the 1989 Edition of the ASME Code,Section XI, which exempts the visual examination of the subject integrally welded attachments to 1-inch and smaller piping in the auxiliary feedwater system. At the NRC staff's request, TVA has provided a comparison of the components exempted from examination under the applicable 1989 ASME Section XI Code with that of the 1991 Addenda. The NRC staff has reviewed and evaluated TVA's request for relief and the supporting information pursuant to 10 CFR 50.55a(a)(3)(ii) for Sequoyah, Units 1 and 2.
2.0 DISCUSSION Component identification Integrally welded attachments of supports and restraints for the AFW system piping 1-inch nominal pipe size (NPS) and smaller.
Code Reouirement The ASME Code,Section XI,1989 Edition, Subparagraph IWD-1220.1 applicable to Class 3 components states: ' Integral attachments of supports and restraints to components that are NPS 4 and smaller within the system boundaries of Examination Categories D-A, D-B, and D-C of Table IWD-2500-1 shall be exempt from the visual examination VT-3, except for the PWR Auxiliary Feedwater System."
Ucensee's Reauest for Relief TVA requests relief from the exemption criteria provided by Subparagraph IWD-1220.1 of the 1989 Edition of Section XI for visual examination requirements associated with the integrally welded attachments of supports and restraints to piping that are NPS 1-inch and smaller in SON's AFW system.
Licensee's Proposed Alternative l
TVA proposes to apply the exemption criteria of Paragraph IWD-1220.1 of the 1991 Addenda to the ASME Section XI Code,1989 Edition, which exempts the VT-3 visual examination of integrally welded attachments to NPS 1-inch and smaller piping in the AFW system.
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3-3.0 EVALUATION With regard to TVA's reconciliation of changes to the requirements of Subsection IWD-1220
" Components Exempt from Examination" of the 1989 Edition of the Code and with that of the 1991 Addenda to the Section XI Code, the NRC staff finds that the 1991 Addenda to the Section XI Code has clarified the intent of the exemption criteria for VT-3 visual examination with the following provisions for Class 3 components:
- a. provision for exemption of components in systems other than the AFW system
- b. provision for exemption of components in the AFW system
- c. provision for exemption of components in systems based on operating pressure / temperature
- d. provision for exemption due to inaccessibility
- e. provision for exemption of components with multiple openings item (b) above exempts piping of nominal pipe size of 1-inch and smaller and components (vessels, pumps, and valves) with cumulative opening of 1-inch and smaller from visual examination of integrally welded attachments. The NRC staff has assessed the impact on safety due to the exemption of piping 1-inch NPS and smaller for the AFW system which TVA has proposed as an alternative to the applicable Code requirement during the second 10-year ISI interval.
Since the inception of the ASME Section XI Code for ISI, piping 1-inch NPS and smaller, excluding steam generator tubes, has been exempted from the inservice examination except being subjected to the system pressure test based on the rationale that upon postulated rupture, the resulting flow loss is within the system capacity. Up until the 1980 ASME Code,Section XI, Winter 1980 Addendum, this exemption was provided and, beginning with the 1991 Addenda, the later Editions of the ASME Section XI Code have also provided for exemption of the above components in the AFW system. However, TVA's requested relief on the exemption is based on the rationale that the consequence of pipe failure of this size has no effect on the plant's ability to achieve and maintain safe shutdown conditions.
While not in disagreement with TVA's rationale, the NRC staff has further taken into consideration the hardship to TVA resulting from radiation exposure to its examination personnel during VT-3 visual examination of integrally welded attachment of piping 1-inch NPS and smaller in the course of removal and replacement of insulation and in the conduct of inspection. Based on the industry record, the instances of service-related failures of integrally welded attachments to piping 1-inch and smaller, are so few that the elimination of visual examination will provide an acceptable risk. Nevertheless, the NRC staff believes that the system pressure test performed during each inspection period would continue to provide assurance of operational readiness by ensuring leaktight integrity of the subject components. The NRC staff, therefore, has determined that the compliance to the 1989 Edition, ASME Code,Section XI, with regard to the VT-3 visual examination of integrally welded attachment to piping 1-inch NPS and smaller in the AFW system, would result in hardship or unusual difficulty to TVA without a compensating increase in the level of quality and safety.
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4.0 CONCLUSION
Based on the licensee's assessment of consequence of failure of piping 1-inch NPS and smaller in the'AFW system that there is no effect on the plant's ability to achieve and maintain safe shutdown conditions, the NRC staff accepts exemption of VF-3 visual examination of integrally welded attachments to the above components in accordance with the 1991 Addenda to the ASME Code,Section XI. The NRC staff believes that the VT-2 visual examination during system pressure test conducted during each inspection period would provide assurance of operational readiness by ensuring leaktight integrity of the pressure boundary. The NRC staff
' has determined that compliance to the 1989 ASME Code,Section XI, in regard to the VT-3 visual examination of the subject components in the AFW system would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety and, therefore, relief is granted pursuant to 10 CFR 50.55a(a)(3)(ii) for Sequoyah, Units 1 and 2, for the second 10-year ISI interval.
Principal Contributor: Prakash Patnaik, NRR Date: September 23, 1999 J
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Mr. J. A. Scalice SEQUOYAH NUCLEAR PLANT Tennessee Valley Authority cc:
Mr. Karl W. Singer, Senior Vice President Mr. Pedro Salas, Manager j
Nuclear Operations Licensing and Industry Affairs Tennessee Valley Authority Sequoyah Nuclear Plant
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6A Lookout Place Tennessee Valley Authority 1101 Market Street P.O. Box 2000 4
' Chattanooga, TN 37402-2801 Soddy Daisy, TN 37379 Mr. Jack A. Bailey Mr. D. L. Koehl, Plant Manager Vice President.
Sequoyah Nuclear Plant Engineering & Technical Services Tennessee Valley Authority Tennessee Valley Authority' P.O. Box 2000 6A Lookout Place Soddy Daisy, TN 37379 1101 Market Street Chattanooga, TN 37402-2801 Mr. Melvin C. Shannon Senior Resident inspector Mr. Masoud Bajestani Sequoyah Nuclear Plant Site Vice President U.S. Nuclear Regulatory Commission 1
Sequoyah Nuclear Plant 2600 Igou Ferry Road Tennessee Valley Authority Soddy Daisy, TN 37379 P.O. Box 2000 Soddy Daisy, TN 37379 Mr. Michael H. Mobley, Director l
TN Dept. of Environment & Conservation l
General Counsel Division of Radiological Health Tennessee Valley Authority 3rd Floor, L and C Annex ET 10H 401 Church Street 400 West Summit Hill Drive Nashville, TN 37243-1532 Knoxville, TN 37902 County Executive Mr. N. C. Kazanas, General Manager Hamilton County Courthouse i
Nuclear Assurance Chattanooga, TN 37402-2801 Tennessee Valley Authority SM Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Mr. Mark J. Burzynski, Manager Nuclear Licensing Tennessee Valley Authority 4X Blue Ridge 1101 Market Street Chattanooga, TN 37402-2801
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