ML061070115
| ML061070115 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 04/18/2006 |
| From: | Moroney B Plant Licensing Branch III-2 |
| To: | Stall J Florida Power & Light Co |
| Moroney Brenden 301-415-3974 | |
| Shared Package | |
| ML061100347 | List: |
| References | |
| TAC MC5633 | |
| Download: ML061070115 (15) | |
Text
April 18, 2006 Mr. J. A. Stall Senior Vice President, Nuclear and Chief Nuclear Officer Florida Power and Light Company P.O. Box 14000 Juno Beach, Florida 33408-0420
SUBJECT:
ST. LUCIE PLANT, UNIT NO. 2 - ISSUANCE OF AMENDMENT REGARDING USE OF WESTINGHOUSE ALLOY 800 SLEEVES IN STEAM GENERATORS (TAC NO. MC5633)
Dear Mr. Stall:
The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 144 to Renewed Facility Operating License No. NPF-16 for the St. Lucie Plant, Unit No. 2. This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated January 6, 2005, as supplemented October 14, 2005, and February 13, 2006.
The amendment revises TS Section 3/4.4.5, Steam Generators, to allow repair of steam generator tubes by installing Westinghouse Electric LLC Alloy 800 leak limiting sleeves.
A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely,
/RA/
Brendan T. Moroney, Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-389
Enclosures:
- 1. Amendment No. 144 to NPF-16
- 2. Safety Evaluation cc w/enclosures: See next page
ML061070115 *No Legal Objection NRR-058 OFFICE LPL2-2/PM LPL2- CSGB/BC OGC LPL2-2/BC 2/LA NAME BMoroney BClayton AHiser Shamrick
- MMarshall by memo dated DATE 04/5/06 04/5/06 04/04/06 04/13/06 04/18/06 FLORIDA POWER & LIGHT COMPANY ORLANDO UTILITIES COMMISSION OF THE CITY OF ORLANDO, FLORIDA AND FLORIDA MUNICIPAL POWER AGENCY DOCKET NO. 50-389 ST. LUCIE PLANT UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 144 Renewed License No. NPF-16
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Florida Power & Light Company, et al. (the licensee), dated January 6, 2005, as supplemented October 14, 2005, and February 13, 2006, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2. Accordingly, Renewed Facility Operating License No. NPF-16 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and by amending paragraph 3.B to read as follows:
B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 144, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Michael L. Marshall, Jr., Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: April 18, 2006
ATTACHMENT TO LICENSE AMENDMENT NO. 144 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-16 DOCKET NO. 50-389 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
Remove Pages Insert Pages 3/4 4-12 3/4 4-12 3/4 4-14 3/4 4-14
3/4 4-14a 3/4 4-15 3/4 4-15 3/4 4-17 3/4 4-17
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 144 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-16 FLORIDA POWER AND LIGHT COMPANY, ET AL.
ST. LUCIE PLANT, UNIT NO. 2 DOCKET NO. 50-389
1.0 INTRODUCTION
By letter dated January 6, 2005, as supplemented October 14, 2005, and February 13, 2006, Florida Power and Light Company, et al., (FPL, the licensee) requested to amend Renewed Operating License NPF-16 for St. Lucie Unit 2, by revising the Technical Specifications (TSs).
The proposed amendment would revise TS Section 3/4.4.5, Steam Generators, to allow repair of steam generator (SG) tubes by installing Westinghouse Alloy 800 leak limiting sleeves as an alternative to tube plugging.
The licensees supplementary submittals dated October 14, 2005, and February 13, 2006, provided clarifying information that did not change the scope of the proposed amendment as described in the original notice of proposed action published in the Federal Register and did not change the initial proposed no significant hazards determination.
St. Lucie Unit 2 has two Combustion Engineering Model 3410 SGs. The mill-annealed Alloy 600 SG tubes have an outside diameter of 0.75 inches and a nominal wall thickness of 0.048 inches. The tubes are explosively expanded for the full-depth of the tubesheet at each end and are supported by a number of carbon steel lattice grid (i.e., eggcrate) tube supports, diagonal bars, and vertical straps.
Sleeving is a repair method for defective tubing, providing an alternative to removing such tubes from service by plugging. The sleeve is a piece of tube with a diameter smaller than the parent tube (i.e., the tube being repaired). The repair method involves inserting the sleeve into the parent tube to span the defective portion of the tube and then expanding each end of the sleeve against the tube to form the sleeve-to-tube joint. Thus, the sleeve replaces the adjacent portion of the parent tube as the pressure boundary. FPL proposes a tube repair method that uses Westinghouse Alloy 800 leak-limiting sleeves, which are not required to be leak tight. The Westinghouse Topical Report, WCAP-15918-P, Revision 2, Steam Generator Tube Repair for Combustion Engineering and Westinghouse Designed Plants with 3/4 Inch Inconel 600 Tubes Using Leak Limiting Alloy 800 Sleeves, dated July 2004 (hereinafter referred to as the WCAP),
documents information regarding the analysis, design, installation, and qualification tests of the Alloy 800 leak-limiting sleeves. Alloy 800 leak-limiting sleeves have been approved for use in several foreign and U.S. (i.e., Calvert Cliffs, Watts Bar, and Comanche Peak) nuclear plants.
2.0 REGULATORY EVALUATION
The applicable U.S. Nuclear Regulatory Commission (NRC) regulations and guidance used in review of the proposed sleeve repair method utilizing Westhinghouse Alloy 800 leak-limiting sleeves are:
General Design Criterion (GDC) 14 of Appendix A to Title 10 of the Code of Federal Regulations (10 CFR), Part 50 requires that the reactor coolant pressure boundary (RCPB) be designed, fabricated, erected and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture. SG tubes represent a significant part of the RCPB. When a tube is defective, the current TSs require that the tube be removed from service by plugging. The licensee is requesting a TS amendment that would permit repair of defective tubes by sleeving as an alternative to plugging the tube. Because the tubes are part of the RCPB, 10 CFR 50.55a requires that the sleeve repair method be qualified in accordance with the specifications in Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), which refers back to Section III of the ASME Code, which is part of the design basis for the SG tubing. The sleeve must satisfy all applicable ASME Code Section III limits for design, operating conditions, and accident loading conditions. In addition, the sleeve wall thickness needs to satisfy the minimum wall thickness requirement of the ASME Code.
Appendix B to 10 CFR Part 50, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, requires a quality assurance program for the design, fabrication, construction, and operation of structures, systems, and components (SSC) in nuclear plants.
The pertinent requirements of Appendix B apply to all activities (i.e., design, fabrication, installation, inspection, repair, etc.) affecting the safety-related functions of those SSCs.
Regulatory Guide (RG) 1.121, Bases for Plugging Degraded PWR [Pressurized-Water Reactor] Steam Generator Tubes, provides guidance for determining the plugging limit, which is the minimum wall thickness beyond which the degraded tube should be plugged. The RG is also applicable to the determination of plugging limits for sleeves. In accordance with RG 1.121, the margin of safety against tube rupture under normal operating conditions should not be less than 3 at any tube location where defects have been detected. The margin of safety against tube failure under postulated accidents, such as loss-of-coolant accident (LOCA), steam line break, or feedwater line break concurrent with the safe shutdown earthquake, should be consistent with the margin of safety determined by the stress limits specified in Section III of the ASME Code.
3.0 TECHNICAL EVALUATION
The NRC staff reviewed the following aspects of the tube repair method utilizing Westinghouse Alloy 800 leak-limiting sleeves: design, installation, material selection, qualification testing, structural analyses, and inspection (i.e., nondestructive examination). These topics are discussed below.
3.1 Sleeve Design The transition (TZ) and tube support Alloy 800 leak-limiting sleeve designs were proposed for tube repair. Tube degradation located near the top of the tubesheet is repaired by TZ sleeves, while degradation at tube support plate intersections or in the freespan region are repaired by tube support sleeves. The length of the TZ and tube support sleeves is determined by the size of the degraded region of the tube (i.e., sleeve encompasses degraded region). A TZ sleeve is attached to the parent tube by several hydraulic expansion joints in the upper region of the sleeve and a hard roll joint in the lower region of the sleeve (i.e., inside the tubesheet). Several hydraulic expansion joints in the upper and lower regions of the sleeve are used to attach a tube support sleeve to the parent tube. The TZ sleeve includes a nickel (Ni) band at the hydraulic expansions and a thermally sprayed Ni-alloy band at the rolled expansion. The Ni-band improves the sealing of the hydraulic joints while the thermally sprayed Ni-alloy band increases the strength of the rolled joint.
3.2 Sleeve Installation All sleeves will be installed in accordance with the WCAP and St. Lucie Unit 2 TSs. Prior to sleeve installation, the inside diameter (ID) of the candidate tube is mechanically conditioned with a high speed buffing tool. The buffing process removes raised material and some of the oxides on the ID of the candidate tube. In addition, buffing prepares the sealing surface of the tube in the sleeve hydraulic expansion areas. The WCAP indicated that the buffing process may be eliminated when a sufficient confidence level in candidate tube ID conditioning is developed. The required number of hydraulic joints within the parent tube is made with an expansion device that is inserted into the sleeve to expand the sleeve. Consistent diametral expansions are ensured by controlling and monitoring the expansion device. The hydraulic expansion joints are designed to ensure that structural and leakage integrity limits are satisfied.
In addition, the hydraulic expansion joints limit the residual stresses in the parent tube. A roll expander produces the hard roll in the lower end of a TZ sleeve, which is controlled and monitored for consistency. After sleeve installation, an eddy current technique will be used to perform an initial acceptance and baseline inspection for all sleeve-tube joints. FPL will plug any sleeve if (1) an unacceptable expansion spacing occurred, (2) the torque value for the rolled joint did not fall within the proper torque range, (3) the two sets of expanded joints were not positioned at the proper elevation, and (4) an unacceptable indication is found in the pressure boundary of the sleeve-tube assembly.
3.3 Sleeve Material Selection Alloy 800, a nickel-iron-chromium (NiFeCr) alloy has been chosen as the sleeve material.
Alloy 800 was chosen by Westinghouse for its favorable mechanical properties and corrosion resistance in both the primary and secondary side water chemistry. It is procured in accordance with the requirements of the ASME Code,Section II, Part B, SB-163, NiFeCr Alloy, Unified Numbering System No. 8800, and Section III, Subsection NB-2000. Alloy 800 is incorporated in ASME Code Case N-20 and is considered acceptable for use by RG 1.85, Materials Code Case Acceptability ASME Section III, Division I, Revision 24, dated July 1986.
Westinghouse also requires additional restrictions on the content of various chemical elements and specifies a specific annealing temperature and yield strength for the Alloy 800 sleeve material.
3.4 Sleeve Qualification Testing Sleeve qualification tests were performed by Westinghouse in accordance with Appendix B to 10 CFR Part 50. The testing program included mechanical load tests (i.e., axial load tests, collapse tests, load cycling tests, pressure tests), leakage tests, and corrosion tests.
Sleeve-tube mockups constructed to the same dimensions as the installed sleeves were used to perform the sleeve qualification tests.
3.4.1 Mechanical Testing Westinghouse performed mechanical tests on mockup SG tubes containing sleeves to ensure the sleeves met design requirements. A summary of these tests is provided below.
Axial load tests were performed to ensure the sleeve-to-tube joint had adequate axial load carrying capability during normal operating, transient, and postulated accident conditions. In this series of tests, the parent tubes were severed to ensure the full axial load was applied to both the upper and lower joint. Axial loads can result from pressure differentials and the differential thermal expansion of the sleeve and the parent tube. Axial load tests were performed not only on specimens that had not been cyclically loaded but also on specimens that had been cyclically loaded (see below for a description of the cyclic load tests). In this series of tests, the displacement of the upper and lower sections were monitored and determined to be acceptable. The actual displacements were minor and would not result in tube-to-tube contact.
Collapse tests were performed to show that the sleeve would not collapse following a LOCA.
These tests showed that, even under secondary-side pressures much larger than that experienced during a LOCA, the sleeve would not collapse.
Cyclic load tests were performed on the sleeve-to-tube joints to determine the effect of cyclic loading on leakage and to ensure the tube-to-sleeve joint retained adequate structural integrity.
Cyclic loads result from the differential thermal expansion of the steam generator tube and the sleeve as well as from the internal pressure differential. The cyclic load tests included fatigue tests at room temperature for both intact and severed sleeved tubes, fatigue tests at operating temperature for severed sleeved tubes, and thermal and mechanical load cycling. The number of cycles that the specimens were subjected to was chosen to be conservative. The results of these tests showed that thermal and mechanical cyclic loads do not cause wear or degradation of the strength or leak resistance of the sleeve-to-tube joint.
Leak tests were performed to determine the leak rate under various operating and postulated accident conditions. The leak tests were performed at room and operating temperature and from both the primary and secondary side of the SG tube/sleeve assembly. Leak tests were performed both prior to the cyclic load tests and at various stages during these tests (as discussed above). The leak rate decreased following load cycling. These tests indicated that the leak rates are sufficiently small which would allow for a large number of sleeves to be installed without exceeding plant leak rate allowances for either accident or normal operating conditions.
3.4.2 Corrosion Testing Various corrosion tests, using highly corrosive solutions, were performed with full-length sleeved-tube mockups. Residual stresses in the parent tube at the sleeve joint locations are increased due to radial expansion of the tube at the sleeve-tube joints, which may increase the parent tubes susceptibility to stress corrosion cracking. The relative time to cracking of the sleeve-tube joint was assessed during the corrosion tests. The corrosion tests showed that cracking did not develop in the Alloy 800 sleeves.
Westinghouse stated that the Alloy 800 sleeves have not experienced service-induced degradation or significant leakage in nuclear plants. Westinghouse also stated that besides Alloy 800 sleeves, Alloy 800 tubing has been used in PWR conditions in foreign nuclear plants without experiencing significant primary or secondary side stress corrosion cracking. This is based on experience with more than 200,000 tubes in service. However, the staff judges that the time for the initiation of corrosion in sleeve-tube assemblies is difficult to quantify accurately.
Although vendors traditionally conduct accelerated corrosion tests of sleeve-tube assemblies to predict service life, the staff finds this method is unreliable for deterministic predictions. The staff does consider that the corrosion tests give a viable indicator of potential performance.
However, given the above listed limitations to sleeve-tube assembly life predictions, the staff can only assume a limited life expectancy of the Alloy 800 sleeves (although the staff recognizes that the life expectancy for the Alloy 600 parent tube is probably more limiting).
Considering the uncertainties in sleeve life expectancy, the staff requires licensees to inspect a sample of sleeves at each refueling outage to ensure that any degradation in the sleeve-tube assembly is detected and addressed early.
3.5 Sleeve Inspection The licensee will inspect the pressure boundary portion of the original tube wall in the area where the sleeve-tube joint will be established prior to all sleeve installations. The inspection will be performed with eddy current equipment and techniques capable of detecting all potential flaw types that may be present at this location.
The licensee stated that the inservice inspection of the Alloy 800 sleeves will be performed with the +Point' probe. This eddy current technique was qualified in accordance with Appendix H of Electric Power Research Institute report, PWR Steam Generator Examination Guidelines, Revision 5, September 1997. Flaws at each transition and expansion zone were represented by axially and/or circumferentially oriented notches in the sleeve-tube qualification samples. In addition, the qualification samples included tube flaws away from the expansion regions but in the pressure boundary. Additional flaws included electro-discharge machined notches and parent tube cracking. However, none of the qualification samples contained flaws behind the Ni-band of the TZ sleeves.
Since none of the qualification samples contained flaws behind the TZ sleeves Ni-band, the licensee elected to include in its TSs that the TZ sleeves with a Ni-band will not be in operation for more than one cycle of operation. Given that the TZ sleeves with a Ni-band will only be in operation for one cycle of operation and the parent tube is inspected prior to sleeve installation, the staff finds that the limitations in the inspection capability are not a concern, since it is highly
unlikely that degradation that would affect the integrity of this region of the sleeve would develop during one cycle of operation.
Following sleeve installation, FPL will perform a full-length preservice baseline examination on all Alloy 800 leak-limiting sleeves, with the exception of the TZ sleeves Ni-band region (as discussed above). This inspection will be performed with eddy current equipment and techniques capable of detecting all potential flaw types that may be present.
3.6 Sleeve Structural Analysis In accordance with Section XI and Section III of the ASME Code (invoked by 10 CFR 50.55a),
Westinghouse performed structural analysis for the Alloy 800 leak-limiting sleeves to demonstrate that the sleeves conform to the original design basis. The structural analysis was performed considering the loads under normal and accident conditions. The structural analysis included evaluating the axial loads acting on the sleeve-tube joint. The effects of vibration on the sleeved tubes were also evaluated. Two bounding tube configurations were assumed in Westinghouses analyses: (1) tube is intact, and (2) tube is severed at flaw location. The following two bounding tube support plate configurations were also used in Westinghouses analyses: (1) tube free to move past tube support plates, and (2) tube locked in the first tube support plate with axial motion prevented. The stresses and fatigue usage factor in the worst sleeve-tube configuration satisfy the ASME Code,Section III requirements.
Minimum required sleeve-thickness calculations, in accordance with ASME Code Section III, were also performed as part of Westinghouses structural analyses. Based on calculations, the sleeve wall is structurally acceptable because the actual sleeve wall thickness is greater than the minimum required wall thickness. Westinghouse accounted for sleeve wall-thickness degradation (i.e., cracking and wall thinning) in its calculations. Sleeve defects/degradation will not remain in service since FPL will plug any sleeved tube with defects/degradation in the sleeve material or in the sleeve-tube joints. The licensees plug-on-detection approach conservatively meets the intent of RG 1.121.
Alloy 800 material properties are not significantly different from Alloy 600 material properties under severe accident conditions, therefore, the structural integrity of the Alloy 800 sleeve is commensurate with the structural integrity of the Alloy 600 parent tube under severe accident conditions.
3.7 Sleeve Leakage Integrity The Alloy 800 sleeve-tube joint is not designed to be leak tight, but laboratory testing has shown the amount of leakage through a sleeve joint is small. FPL assumes that all installed sleeves leak at the rate provided in the WCAP. The leak rate for each sleeve is based on the upper 95-percent confidence limit on the mean value of leakage for appropriate temperature and pressure conditions. The leakage rate from all sleeves will be included in the condition monitoring and operational assessment process.
3.8 Technical Specifications Changes The staff reviewed the proposed changes to Section 3/4.4.5, Steam Generators, of the St. Lucie Unit 2 TSs to determine their acceptability regarding Alloy 800 leak-limiting sleeve installation. The following changes were proposed:
3.8.1 TS Surveillance Requirement (SR) 4.4.5.2.b will be revised to add a new item 4 containing inservice inspection requirements for the Alloy 800 leak-limiting sleeves, as follows:
- 4. All inservice Leak Limiting Alloy 800 sleeves shall be inspected over their full length during each refueling outage. These inspections will include both the tube and the sleeve.
3.8.2 SR 4.4.5.4.a.6 will be revised to include Alloy 800 leak-limiting sleeves in the plugging limit definition, as follows
- 6. Plugginq or Repair Limit means the condition at or beyond which the tube shall be removed from service by plugging or repaired by sleeving using the method in Specification 4.4.5.4.a.10 in the affected area. The plugging or repair limits are as follows:
- i. In the non-sleeved portion of a tube, the plugging or repair limit imperfection depth is 40% of the nominal wall thickness. This Limit is not applicable in the portion of the tube that is greater than 10.3 inches below the bottom of the hot leg expansion transition or top of the tubesheet (whichever is lower) to the tube end. Degradation detected between 10.3 inches below the bottom of the hot leg expansion transition or top of the tubesheet (whichever is lower) and the bottom of the hot leg expansion transition or top of the tubesheet (whichever is higher) shall be plugged or repaired on detection.
ii. In the region of a tube sleeved using a Westinghouse Leak Limiting Alloy 800 sleeve, the tube shall be plugged upon detection of any service induced imperfection, degradation or defect in the (a) sleeve or (b) pressure boundary portion of the original tube wall in the sleeve/tube assembly (i.e., the sleeve-to-tube joint).
iii. All Leak Limiting Alloy 800 Sleeves that have a nickel band shall be plugged or removed from service after one cycle in operation.
3.8.3 SR 4.4.5.4.a.8 will be revised such that the requirements for the extent of tube inspection are consistent with Amendment No. 143, dated April 11, 2006, as follows:
- 8. Tube Inspection for a tube with no portion of a sleeve extending below 10.3 inches from the bottom of the hot leg expansion transition or the top of the tubesheet (whichever is lower) means an inspection of the steam generator tube from 10.3 inches below the bottom of the hot leg expansion transition or top of the tubesheet (whichever is lower) completely around the U-bend to the top support of the cold leg. Tube Inspection for a tube with a portion of a sleeve extending below 10.3 inches from the bottom of the hot leg expansion transition
or the top of the tubesheet (whichever is lower) means an inspection from the bottom of the sleeve completely around the U-bend to the top support of the cold leg.
3.8 4 SR 4.4.5.4.a will be revised to add a new item 10, a definition of tube repair that includes Westinghouse Alloy 800 leak-limiting sleeves, as follows:
- 10. Tube Repair refers to sleeving with Westinghouse Leak Limiting Alloy 800 sleeves described in WCAP-15918-P, Revision 2, which are used to maintain a tube in service.
Leak Limiting Alloy 800 Sleeves are applicable only to the original steam generators.
The pressure boundary portion of the original tube wall in the sleeve/tube assembly (i.e.,
the sleeve-to-tube joint) shall be inspected prior to installation of each sleeve.
3.8.5 SR 4.4.5.4.b will be revised to add the option of repair, as follows:
- b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-2.
The referenced Table 4.4-2 will be revised to conform with the preceding changes by adding, where appropriate, the option of repair in addition to tube plugging.
3.8.6 TS Section 4.4.5.5, Reports, will be revised to conform with the preceding changes by adding the requirement to report the number of sleeves inspected and the number of tubes repaired by sleeving in postinspection reports.
3.9 Conclusions Structural analyses and tests for a variety of loadings that enveloped plant-specific design, operating, transient, and accident loads were performed by FPL. The analyses, testing, and operating experience demonstrate that the integrity of defective tubes may be restored by use of Westinghouse Alloy 800 leak-limiting sleeves. Therefore, the NRC staff finds that FPL has demonstrated the acceptability of the sleeve repair in accordance with NRC regulations and guidance (e.g., 10 CFR 50.55a, GDC 14 of Appendix A to 10 CFR, Part 50, RG 1.121) and the proposed changes are acceptable. The staff also concludes that the proposed changes to the TSs are consistent with the analyses and the proposed requirements for sleeving and are acceptable.
4.0 STATE CONSULTATION
Based upon a letter dated May 2, 2003, from Michael N. Stephens of the Florida Department of Health, Bureau of Radiation Control, to Brenda L. Mozafari, Senior Project Manager, U.S. Nuclear Regulatory Commission, the State of Florida does not desire notification of issuance of license amendments.
5.0 ENVIRONMENTAL CONSIDERATION
This amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (70 FR 9993, dated March 1, 2005). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: Leslie Miller Date: April 18, 2006
Mr. J. A. Stall ST. LUCIE PLANT Florida Power and Light Company cc:
Senior Resident Inspector Mark Warner, Vice President St. Lucie Plant Nuclear Operations Support U.S. Nuclear Regulatory Commission Florida Power and Light Company P.O. Box 6090 P.O. Box 14000 Jensen Beach, Florida 34957 Juno Beach, FL 33408-0420 Craig Fugate, Director Mr. Rajiv S. Kundalkar Division of Emergency Preparedness Vice President - Nuclear Engineering Department of Community Affairs Florida Power & Light Company 2740 Centerview Drive P.O. Box 14000 Tallahassee, Florida 32399-2100 Juno Beach, FL 33408-0420 M. S. Ross, Managing Attorney Mr. J. Kammel Florida Power & Light Company Radiological Emergency P.O. Box 14000 Planning Administrator Juno Beach, FL 33408-0420 Department of Public Safety 6000 Southeast Tower Drive Marjan Mashhadi, Senior Attorney Stuart, Florida 34997 Florida Power & Light Company 801 Pennsylvania Avenue, NW. Mr. G. L. Johnston Suite 220 Acting Vice President Washington, DC 20004 St. Lucie Nuclear Plant 6351 South Ocean Drive Mr. Douglas Anderson Jensen Beach, Florida 34957-2000 County Administrator St. Lucie County Mr. Bill Parks 2300 Virginia Avenue Acting Operations Manager Fort Pierce, Florida 34982 St. Lucie Nuclear Plant 6351 South Ocean Drive Mr. William A. Passetti, Chief Jensen Beach, Florida 34957-2000 Department of Health Bureau of Radiation Control 2020 Capital Circle, SE, Bin #C21 Tallahassee, Florida 32399-1741 Mr. C. Costanzo Acting Plant General Manager St. Lucie Nuclear Plant 6351 South Ocean Drive Jensen Beach, Florida 34957 Mr. Terry Patterson Licensing Manager St. Lucie Nuclear Plant 6351 South Ocean Drive Jensen Beach, Florida 34957