ML091970477
| ML091970477 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 07/24/2009 |
| From: | Boyce T Plant Licensing Branch II |
| To: | Franke J Florida Power & Light Co |
| Saba F, NRR/DORL/LPL2-2, 301-415-1447 | |
| References | |
| TAC ME0227 | |
| Download: ML091970477 (6) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 24, 2009 Mr. Jon A. Franke, Vice President Crystal River Nuclear Plant (NA1B)
ATrN: Supervisor, Licensing & Regulatory Programs 15760 W. Power Line Street Crystal River, Florida 34428-6708
SUBJECT:
CRYSTAL RIVER NUCLEAR PLANT, UNIT NO.3 - RELIEF REQUEST 08-001-MX, REVISION 0, REGARDING INITIAL EXAMINATION OF TENDONS AFFECTED BY POST-TENSIONING SYSTEM REPAIR/REPLACEMENT ACTIVITIES (TAC NO. ME0227)
Dear Mr. Franke,
By letter dated December 10, 2008, as supplemented by letter dated April 16, 2009, Florida Power Corporation (the licensee) submitted Relief Request 08-001-MX, Revision 0 (RR 08-001-MX) for the fourth 10-year inservice inspection (lSI) interval at Crystal River Unit 3 Nuclear Generating Plant (CR-3). This submittal requested an alternative to Table IWL-2521-2, "Augmented Examination Requirements Following Post-Tensioning System Repair/Replacement Activities," of American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, 2001 Edition through the 2003 Addenda, which requires an initial examination of tendons affected by post-tensioning system repair/replacement activities within 1 year (plus or minus 3 months) follOWing the completion of repair/replacement activities. The licensee proposed to perform the initial inspection of vertical tendons affected by the steam generator replacement activities at 2 years (plus or minus 3 months) following repair/replacement activities in lieu of 1 year (plus or minus 3 months) as specified in Table IWL-2521-2.
The Nuclear Regulatory Commission (NRC) staff has evaluated the licensee's submittal for RR 08-001 MX for the fourth 1O-year lSI interval at CR-3. Based on the information provided in the licensee's submittals dated December 10, 2008, and April 16, 2009, the NRC staff concludes that the licensee's proposed alternative to the requirements of Table IWL 2521-2 of the ASME Code is acceptable because it will provide an acceptable level of quality and safety.
Therefore, pursuant to paragraph 50.55a(a)(3)(i) of Title 10 of Code of Federal Regulations the implementation of the alternative is authorized for the CR-3 fourth 10-year lSI interval which began on August 14, 2008.
J. Franke
- 2 The NRC staff's safety evaluation is enclosed. If you have any questions regarding this matter, please contact Farideh Saba at (301) 415-1447.
Sincerely, IRA by EBrown forI Thomas H. Boyce, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-302
Enclosure:
Safety Evaluation cc w/enclosure: Distribution via ListServ
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- i' SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST 08-001-MX REGARDING INITIAL EXAMINATION OF TENDONS AFFECTED BY POST-TENSIONING SYSTEM REPAIR/REPLACEMENT ACTIVITIES FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT DOCKET NO. 50-302
1.0 INTRODUCTION
By letter dated December 10, 2008, as supplemented by letter dated April 15, 2009, Florida Power Corporation (the licensee) submitted Relief Request 08-001-MX, Revision 0 (RR 08-001-MX), for the fourth 10-year inservice inspection (lSI) interval at Crystal River Unit 3 Nuclear Generating Plant (CR-3). RR 08-001-MX requested an alternative to Table IWL-2521-2, "Augmented Examination Requirements Following Post-Tensioning System Repair/Replacement Activities," of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, 2001 Edition through the 2003 Addenda, which requires an initial examination of tendons affected by post-tensioning system repair/replacement activities within 1 year (plus or minus 3 months) following the completion of repair/replacement activities.
The licensee is planning to replace the steam generators during the 16th refueling outage, which is currently scheduled to start in September 2009. The removal and replacement of steam generators from the containment building requires a temporary construction opening.
Creation and restoration of this temporary construction opening requires removal and reinstallation of post-tensioning tendons within the boundary of the opening and detensioning and retensioning of selected tendons adjacent to the opening.
The licensee, in the letter dated December 10, 2008, stated that (1) the total number of vertical tendons in the containment building is 144; (2) total of 30 vertical tendons (23V1, 23V2, 23V3, 34V1 through 34V24, 45V22, 45V23, and 45V24) are affected by the steam generator replacement activities; and (3) according to the percentage of tendons affected by the repair/replacement activities and as specified in Table IWL-2521-2, a sample of two vertical tendons are required to have initial inspection 1 year (plus or minus 3 months) following the completion of repair/replacement activities. The licensee requested to defer the initial inspection of the sample of two vertical tendons by 1 year to align this initial inspection with the 9th tendon surveillance (35th year) scheduled to be performed in 2011 during 17'h refueling outage (R17).
Enclosure
- 2 The Nuclear Regulatory Commission (NRC) staff's review addressed the ability of the licensee to maintain an acceptable level of quality and safety and to ensure integrity of the containment post tensioning system when the initial inspection of vertical tendons affected by the post tensioning repair/replacement activities is performed 2 years (plus or minus 3 months) in lieu of 1 year (plus or minus 3 months) following the repair/replacement activities.
2.0 REGULATORY REQUIREMENTS Paragraph 50.55a(g) of Title 10 of the Code of Federal Regulations (10 CFR) specifies that lSI of nuclear power plant components shall be performed in accordance with the requirements of the ASME Code,Section XI. Pursuant t01 0 CFR 50.55a(a)(3), alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (i) the 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.
Subsection IWL-2521.2, "Tendon Affected by Repair/Replacement Activities," and Table IWL-2521-2 of ASME Code Section XI, 2001 Edition including Addenda through 2003, require augmented examination of tendons affected by post tensioning system repair/replacement activities. Table IWL-2521-2 requires an initial inspection of tendons affected by post-tensioning system repair/replacement activities within 1 year (plus or minus 3 months) following the completion
3.0 TECHNICAL EVALUATION
3.1 ASME Code Components Affected
ASME Class CC Components, Vertical Post-Tensioning Tendons (23V1, 23V2, 23V3, 34V1 through 34V24, 45V22, 45V23, and 45V24) 3.2
Applicable Code Edition and Addenda
Current Code of Record at CR-3 for the fourth 10-year containment lSI interval is the ASME Code,Section XI, 200-1 Edition through the 2003 Addenda.
3.3 Applicable ASME Code Requirement Table IWL-2521-2 of the current Code of Record requires an initial examination of tendons affected by post-tensioning system repair/replacement activities within 1 year (plus or minus 3 months) following the completion of repair/replacement activities.
3.4 Licensee Proposed Alternative and Basis for Use In RR 08-001-MX, as an alternative to Table IWL-2521-2 of ASME Code,Section XI, the licensee requested to perform the initial inspection of vertical tendons affected by the steam generators replacement activities at 2 years (plus or minus 3 months) following repair/replacement activities in lieu of 1 year (plus or minus 3 months). This alternative is proposed for the fourth 10-year lSI interval which began on August 14, 2008.
- 3 repair/replacement activities in lieu of 1 year (plus or minus 3 months). This alternative is proposed for the fourth 10-year lSI interval which began on August 14, 2008.
The licensee stated, in its letter dated December 2008, that (1) 15 lSI examinations have been completed to date and the existing vertical tendons have met all the post-tensioning test and examination acceptance criteria, (2) the planned repair/replacement activities affecting the vertical tendons will not result in any unique condition that may subject the vertical tendons to a different potential for structural or tendon deterioration, and (3) the affected vertical tendons will be retensioned to the original tendon seating force (0.7 times the ultimate tendon strength) and will be sUbject to prestress losses smaller than those for the original tendons. As such, this alternative provides an acceptable level of quality and safety, and does not adversely impact the health and safety of the public.
3.5 NRC Staff's Evaluation The NRC staff identified areas in which additional information was necessary to complete the evaluation of the licensee's proposed alternative. Therefore, the staff issued a request for additional information (RAI) to obtain further information from the licensee relative to the magnitude of vertical tendon prestressing force and the inspection results of previous tendon surveillances.
In its response, dated April 16, 2009, to the NRC staff's RAI provided the required vertical tendon force, according to the CR-3 current licensing basis, and the minimum predicted force for the population of the affected vertical tendons at the time of the R17 in 2011 as 1149 kilopounds (kips) and 1512 kips, respectively.
Considering the above, the NRC staff determined that the minimum predicted prestressing force in the affected vertical tendons is approximately 30 percent higher than the design basis required tendon force. Therefore, the NRC staff determined that there will be sufficient vertical prestressing force to satisfy the containment structure design basis requirements.
In its response to the NRC staff's RAI, the licensee also provided the inspection results for the past eight tendon surveillances performed from 1977 through 2007. Vertical tendons 23V2, 34V1, 34V4, 34V6, and 34V19 that will be affected by the steam generator repair/replacement activities were selected for inspection during previous tendon surveillances. The results of lift-off tests, tendon wire inspections, and tendon end anchorage inspections were generally satisfactory.
The NRC staff requested information relative to any indication of water intrusion or corrosion in the vertical tendons. In its response to the staff's RAI, the licensee stated that during the second tendon surveillance performed in 1980, tendon 56V1 was found to have approximately 15 gallons of water. The water was drained and the grease was replaced. The licensee added that during previous tendon surveillances performed from 1981 through 2007, with the exception of finding a few drops of water in the horizontal tendon 51 H25 field-end, there was no evidence of water found in any end caps of inspected horizontal and vertical tendons. The NRC staff concluded that the licensee adequately responded the NRC staff's RAI regarding indication of water intrusion or corrosion in the vertical tendons.
-4 tendons that will be affected by the steam generator repair/replacement activities were generally satisfactory; (2) the minimum predicted prestressing force in the affected vertical tendons, at the time of the proposed deferred initial inspection, is sufficient to satisfy the containment structure design basis vertical prestress force; and (3) there was no indication of water intrusion in the vertical tendons from 1981 through 2007. Therefore, the staff finds the licensee's proposed alternative to the requirements of Table IWL 2521-2 of ASME Code,Section XI, 2001 Edition through the 2003 addenda to postpone the initial inspection of the vertical tendons affected by the steam generator replacement activities by one year to be acceptable.
4.0 CONCLUSION
Based on the information provided in the licensee's submittals dated December 10, 2008, and April 16, 2009, the NRC staff concludes that the licensee's proposed alternative to the requirements of Table IWL 2521-2 of ASME Code,Section XI, 2001 Edition through the 2003 addenda is acceptable because it will provide an acceptable level of quality and safety.
Therefore, pursuant to 10 CFR 50.55a(a)(3)(i) the implementation of the alternative is authorized for the CR-3 fourth 10-year in service inspection interval which began on August 14, 2008. This authorization is limited to those components described in Section 3.1.
All other ASME Code,Section XI requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: Farhad Farzam Date: July 24, 2009
J. Franke
- 2 The NRC staff's safety evaluation is enclosed. If you have any questions regarding this matter, please contact Farideh Saba at (301) 415-1447.
Sincerely, IRA by EBrown forI Thomas H. Boyce, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-302
Enclosure:
Safety Evaluation cc w/enclosure: Distribution via ListServ DISTRIBUTION:
PUBLIC LPL2-2 R/F RidsNrrPMCrystalRiver RidsNrrLACSola RidsNrrDorlLpl2-2 RidsNrrDeEmcb RidsOgcRp RidsRgn2MailCenter RidsNrrDorlDpr Farhad Farzam RidsAcrsAcnw MailCTR ADAMS Accession Number: ML091970477 NRR-028 OFFICE LPL2-2/PM LPL2-2/LA EMCB/BC*
OGC*Waiver LPL2-2/BC NAME FSaba CSoia MKhanna JGitter TBoyce (EBrown for)
DATE 07/22/09 07/22/09 07/24/09 03/04/09 07/24/09
- by memorandum OFFICIAL RECORD COPY