ML040440111
| ML040440111 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 04/28/2004 |
| From: | Moroney B NRC/NRR/DLPM/LPD2 |
| To: | Stall J Florida Power & Light Co |
| Moroney B, NRR/DLPM, 415-3974 | |
| Shared Package | |
| ML040440119 | List: |
| References | |
| TAC MB5667, TAC MB5668 | |
| Download: ML040440111 (16) | |
Text
April 28, 2004 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 UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS REGARDING THE RELOCATION OF SPENT FUEL POOL CRANE TECHNICAL SPECIFICATION REQUIREMENTS (TAC NOS. MB5667 AND MB5668)
Dear Mr. Stall:
The Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment Nos. 190 and 134 to Facility Renewed Operating License Nos. DPR-67 and NPF-16 for the St. Lucie Plant, Units 1 and 2. These amendments consist of changes to the Technical Specifications in response to your application dated July 18, 2002, as supplemented on November 14, 2002, and December 11, 2003.
These amendments implement an administrative change to relocate certain technical specification requirements regarding the spent fuel and cask handling cranes to the respective units Updated Final Safety Analysis Reports (UFSARs).
Changes to the description of the spent fuel cask handing cranes and the spent fuel handling accident analysis should be included in the next required update of the St. Lucie Unit 1 and 2 UFSARs.
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, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-335 and 50-389
Enclosures:
- 1. Amendment No. 190 to DPR-67
- 2. Amendment No. 134 to NPF-16
- 3. Safety Evaluation cc w/enclosures: See next page
ML040440111 NRR-058 OFFICE PDII-2/PM PDII-2/LA PDII-1/PM DSSA/S RORP OGC PDII-PLB 2/SC NAME BMoroney BClayton CGratton DSolorio TBoyce LZacardi WBurton DATE 3/4 /04 3/4/04 03/9/04 3/19/04 3/24/04 4/1/04 4/28/04 FLORIDA POWER & LIGHT COMPANY DOCKET NO. 50-335 ST. LUCIE PLANT UNIT NO. 1 AMENDMENT TO FACILITY RENEWED OPERATING LICENSE Amendment No. 190 License No. DPR-67
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Florida Power & Light Company (the licensee),
dated July 18, 2002, as supplemented on November 14, 2002, and December 11, 2003, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions 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 Commissions 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 Commissions regulations and all applicable requirements have been satisfied.
- 2. Accordingly, Facility Renewed Operating License No. DPR-67 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. 190, are hereby incorporated in the license. FPL 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 issuance. In addition, the licensee shall include the relocated information in the Updated Final Safety Analysis Report submitted to the NRC, pursuant to 10 CFR 50.71(e), as described in the licensees application and evaluated in the staffs safety evaluation attached to this amendment.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
William F. Burton, Acting Chief, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: April 28, 2004
ATTACHMENT TO LICENSE AMENDMENT NO. 190 TO FACILITY RENEWED OPERATING LICENSE NO. DPR-67 DOCKET NO. 50-335 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the area of change.
Remove Pages Insert Pages VIII VIII 3/4 9-7 3-4 9-7 3/4 9-15 3/4 9/15
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 FACILITY RENEWED OPERATING LICENSE Amendment No. 134 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 July 18, 2002, as supplemented on November 14, 2002, and December 11, 2003, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions 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 Commissions 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 Commissions regulations and all applicable requirements have been satisfied.
- 2. Accordingly, Facility Renewed 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. 134, are hereby incorporated in the license. FPL 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 issuance. In addition, the licensee shall include the relocated information in the Updated Final Safety Analysis Report submitted to the NRC, pursuant to 10 CFR 50.71(e), as described in the licensees application and evaluated in the staffs safety evaluation attached to this amendment.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
William F. Burton, Acting Chief, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: April 28, 2004
ATTACHMENT TO LICENSE AMENDMENT NO. 134 TO FACILITY RENEWED 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 marginal lines indicating the area of change.
Remove Pages Insert Pages IX IX 3/4 9-7 3-4 9-7 3/4 9-13 3/4 9-13
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 190 AND 134 TO FACILITY OPERATING LICENSES NOS. DPR-67 AND NPF-16 FLORIDA POWER AND LIGHT COMPANY, ET AL.
ST. LUCIE PLANT, UNITS 1 AND 2 DOCKET NOS. 50-335 AND 50-389
1.0 INTRODUCTION
By letter dated July 18, 2002, as supplemented on November 14, 2002, and December 11, 2003, Florida Power and Light Company, et al. (FPL, the licensee), requested amendments to Renewed Operating Licenses DPR-67 and NPF-16 for St. Lucie Units 1 and 2, respectively, to relocate the following technical specifications (TSs) related to the operation of the spent fuel and cask handling cranes to the respective units Updated Final Safety Analysis Reports (UFSARs):
TS 3/4.9.7 Crane travel - Spent Fuel Storage Pool Building [Units 1 and 2]
TS 3/4.9.13 Spent Fuel Cask Crane for Unit 1 TS 3/4.9.12 Spent Fuel Cask Crane for Unit 2 The licensees letters dated November 14, 2002, and December 11, 2003, providing supplemental information for this amendment request did not affect the original proposed no significant hazards determination, or expand the scope of the request as noticed in the Federal Register on August 6, 2002 (67 FR 50954).
2.0 REGULATORY EVALUATION
Section 182a of the Atomic Energy Act of 1954, as amended (the Act) requires applicants for nuclear power plant operating licenses to include the TSs as part of the license. The Commissions regulatory requirements related to the content of TSs are set forth in Section 50.36 of Title 10 of the Code of Federal Regulations (10 CFR). The regulation requires that the TSs include items in specific categories, including: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs);
(3) surveillance requirements; (4) design features; and (5) administrative controls. The regulation does not specify the particular requirements to be included in the TSs.
The four criteria defined by 10 CFR 50.36(c)(2)(ii) for determining whether particular items are required to be included in the TS LCOs, are as follows:
(A) Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
(B) Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; (C) Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
(D) Criterion 4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
The design basis accidents and transient analyses discussed in Criteria 2 and 3 include any design basis event described in the UFSAR, not just those events described in Chapters 6, Engineered Safety Features, or Chapter 15, Accident Analysis. The initial conditions captured under Criterion 2 should not be limited to only process variables assumed in the safety analyses, they should also include certain active design features and operating restrictions needed to preclude unanalyzed accidents. In this context, active design features include only those design features under the control of operations personnel (i.e., licensed operators and personnel who perform control functions at the direction of licensed operators).
Furthermore, should the TSs involve physical, designed-in features that prevent operations staff from immediately exceeding the assumptions in the bounding analysis in the course of operations, then the TSs would not meet Criterion 2 and could be relocated to the UFSAR or other similarly controlled document. The NRC staff documented its decisions on the relocation of TSs in the NRC Staff Review of Nuclear Steam Supply System Vendor Groups Application of the Commissions Interim Policy Criteria to Standard Technical Specifications, transmitted to the various nuclear industry owners groups on May 9, 1988.
Existing TSs that fall within or satisfy any of the above criteria must be retained in the TSs; those that do not fall within or satisfy these criteria may be relocated to other licensee-controlled documents.
In NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, the NRC staff provided regulatory guidelines for control of heavy load lifts to assure safe handling of heavy loads in areas where a load drop could impact on stored spent fuel, fuel in the reactor core, or equipment that may be required to achieve safe shutdown or permit continued decay heat removal. Section 5.1.1 of NUREG-0612 provides guidelines for reducing the likelihood of dropping heavy loads and provides criteria for establishing safe load paths; procedures for load-handling operations; training of crane operators; design, testing, inspection, and maintenance of cranes and lifting devices; and analyses of the impact of heavy load drops.
The guidelines in Sections 5.1.2 through 5.1.6 address alternatives to either further reduce the probability of a load-handling accident or mitigate the consequences of heavy load drops.
These alternatives include using a single-failure-proof crane to improve reliability through
increased factors of safety and through redundancy or duality in certain active components.
Criteria for design of single-failure-proof cranes are included in NUREG-0554, Single-Failure-Proof Cranes for Nuclear Power Plants.
3.0 TECHNICAL EVALUATION
The NRC staff performed the following evaluations of the TS LCOs listed in Section 1.0 of this safety evaluation (SE) to determine whether it is appropriate to relocate the TSs to the appropriate units UFSAR: (a) the application of Criterion 1, 3, and 4, to the TSs proposed for relocation, (b) application of Criteria 2 to the TSs proposed for relocation, and (c) the proposed level of control of the relocated requirements. TSs that meet any of the four criteria listed in 10 CFR 50.36, as summarized in Section 2.0 above, cannot be relocated to the UFSAR.
3.1 Criteria 1, 3, and 4 In its submittal, the licensee pointed out that the systems, equipment, and limits contained in TS 3/4.9.7 for each unit, TS 3/4.9.13 for Unit 1, and TS 3/4.9.12 for Unit 2, were not related to any installed instrumentation that is used to detect reactor coolant pressure boundary leakage or to indicate such degradation in the control room. Therefore, the licensee concluded that these TSs do not meet Criterion 1.
Likewise, the licensee stated that the systems, equipment, and criteria in these TSs are not a part of the primary success path that functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. As a result, the licensee concluded that TS 3/4.9.7 for each unit, TS 3/4.9.13 for Unit 1, and TS 3/4.9.12 for Unit 2 do not meet Criterion 3.
Finally, the licensee found that the systems, equipment, and limits contained in the aforementioned TS LCOs were not components listed as risk significant under either the St. Lucie probabilistic risk assessment (PRA) program, or the St. Lucie Maintenance Rule Program. From this, the licensee has concluded that TS 3/4.9.7 for each unit, TS 3/4.9.13 for Unit 1, and TS 3/4.9.12 for Unit 2 do not meet Criterion 4; that is, they are not a structure, system, or component that operating experience or PRA has shown to be significant to public health and safety.
The staff reviewed the licensees bases for relocation against the criteria in 10 CFR 50.36. In addition, the NRC staff independently evaluated the TS proposed for relocation against the requirements in 10 CFR 50.36 and concluded that TS 3/4.9.7 for each unit, TS 3/4.9.13 for Unit 1, and TS 3/4.9.12 for Unit 2 do not meet Criterion 1, 3 or 4.
3.2 Criterion 2 Criterion 2 applies to a process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis described in the UFSAR that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Initial conditions captured under Criterion 2 are not limited to only process variables assumed in safety analyses, but also include certain active design features and operating restrictions needed to preclude unanalyzed accidents. Active design features are intended to be those design features under the control of operations personnel (i.e., licensed operators and
personnel who perform control functions at the direction of licensed operators). Should a specific LCO considered for relocation involve a physical, designed-in plant feature that prevents operations staff from immediately placing the plant in an unanalyzed condition in the course of operations (one that would require a design change before operators could exceed the limits of the LCO) that LCO would not satisfy Criterion 2.
Crane Travel - Spent Fuel Storage Pool Building The NRC staff reviewed the design and operation of the spent fuel handling equipment for Units 1 and 2 against the requirements in 10 CFR 50.36 to determined whether TS 3/4.9.7 could be relocated to the UFSAR. The NRC staff considered loads that could be transported over the spent fuel pool by the refueling machine and the cask-handling crane.
The licensee evaluated several fuel-handling accident (FHA) scenarios in the past, including a FHA in the refueling building. The licensees current design basis FHA is described in St. Lucie UFSAR Section 15.4.3 for Unit 1 and Section 15.7.4.1.2 for Unit 2. The current FHA assumes the release of volatile fission products in the gas gap of the fuel pins from a single assembly in the containment with the equipment door and the emergency and personnel airlocks open for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following the event. The licensee determined that the in-containment FHA bounded the FHA in the refueling building. The staff documented its review of the licensees revised FHA in amendments dated February 27, 2001, for Unit 1 and October 22, 2001, for Unit 2.
Even though the refueling building FHA is not the bounding FHA described in the UFSAR, operating restrictions on the maximum loads that can be carried over the stored fuel in the spent fuel pool are necessary to ensure the current in-containment FHA remains bounding.
Dropping a load in excess of the current TS load limits onto the stored fuel in the spent fuel pool is unanalyzed and could result in consequences that exceed the limits evaluated in the current FHA. The load limit operating restriction in TS 3/4.9.7 for both units ensures loads exceeding the analyzed limits are not transported over stored spent fuel.
The NRC staff reviewed the design and operation of the Unit 1 and 2 refueling machines and came to the following conclusions regarding the application of Criterion 2 to the crane travel load limit. The licensee evaluated several FHA scenarios in the refueling building and determined that the most limiting scenario occurs when a fuel assembly is dropped vertically to the fuel pool floor and then strikes a protruding structure within the pool as it rotates and falls to the pool floor. The licensees analysis indicated that under these conditions, 14 outer fuel rods of the fuel assembly would be crushed, and the next interior row of fuel rods would bend. The licensee also evaluated the consequences of a dropped fuel assembly end-to-end impact with another fuel bundle stored in a spent fuel rack, but found it to be nonlimiting compared with the fuel assembly that drops and strikes a protruding object. However, for the purposes of performing the FHA analyses for each unit, the licensee assumed that the accidents radiological release would be based on the non-mechanistic failure of all fuel rods in one assembly. The accident considers the design of the refueling machine in that it prevents the operator from handling more than one fuel assembly at a time. The licensees assumption that the entire fuel assembly fails bounds the worst-case damage postulated from all possible fuel assembly drop analyses. For the licensee to exceed the assumptions in the fuel handling accident using the refueling machine, changes to the machines design or the weight of the fuel assemblies used at St. Lucie, Units 1 and 2, would be required. Prior NRC approval would be
required to make changes to the fuel-handling machine or to the type of fuel used if those changes could result in more than a minimal increase in the consequences of the FHA.
It could be postulated that a nonfuel load drop could produce an equivalent amount of fuel damage (i.e., equal to the non-mechanistic failure of all fuel rods in a single fuel assembly) if a sufficiently large nonfuel load were carried over the stored spent fuel by the refueling machine hoist. However, the licensee was not required to perform this analysis in the UFSAR, therefore, 10 CFR 50.36 does not require that the licensee include a TS establishing a LCO for the nonfuel load drop initial conditions. Under 10 CFR 50.59(c)(2)(iii), the licensee is required to evaluate the handling of nonfuel loads in the spent fuel pool that have not been previously evaluated. Prior NRC approval would be required to move a nonfuel load over irradiated fuel where dropping the load could result in more than a minimal increase in the consequences of the FHA.
Loads exceeding the limits established in TS 3/4.9.7 could also be transported in the vicinity of the irradiated fuel stored in the spent fuel pool by the cask-handling crane. However, the physical design of the refueling building allows loads on the main hook to travel only over the cask storage area of the spent fuel pool. The cask crane bridge is also outfitted with restricted zone limit switches which limits the operation of the cask crane bridge to areas away from stored spent fuel. A change in the design of the refueling building, or design or operation or the cask crane equipment would be required to allow the cask crane to transport loads in excess of TS 3/4.9.7 for Units 1 and 2 over stored spent fuel. Prior NRC approval would be required to make such modifications if the changes could result in more than a minimal increase in the consequences of the FHA.
Because the plant contains design features that will prevent plant operators from immediately exceeding the load limits specified in TS 3/4.9.7 for St. Lucie Units 1 and 2, the NRC staff concludes that the load limits do not meet Criterion 2. Therefore, current TS 3/4.9.7 can be relocated to the UFSAR.
Cask Cranes The licensee stated that the spent fuel cask crane load limit restrictions in TS 3/4.9.13 for Unit 1, and TS 3/4.9.12 for Unit 2 were inherent elements of the cask drop event initiator, and that, in conjunction with other non-TS requirements that restrict cask crane operation, provide a defense-in-depth approach to handling heavy loads. The licensee concluded that TS 3/4.9.13 for Unit 1, and TS 3/4.9.12 for Unit 2, did not meet Criterion 2.
The NRC staff reviewed the design and operation of the Unit 1 and 2 cask handling cranes against the requirements in 10 CFR 50.36 and came to the following conclusions. The St. Lucie Unit 1 UFSAR, Chapter 9.1, provides descriptions and analyses of the design features of the fuel storage and handling systems, including components such as the cask cranes.
Within this chapter, the licensee evaluated a cask drop event, wherein the licensee concluded that sufficient energy could be obtained from the drop of a ten-element cask (105 tons) to void the leak tight integrity of the pool structure. Further licensee analysis indicated that the drop of a single fuel element cask (25 tons) did not affect the integrity of the fuel storage system. The licensee currently restricts the load carried by the spent fuel cask crane to 25 tons in St. Lucie Unit 1 TS 3.9.13. Similarly for Unit 2, UFSAR Chapter 9.1 discusses the results of a licensee analysis that concludes that the cask storage pool is designed to withstand the impactive load
from a cask drop of 100 tons on the pool floor or walls. Loads in excess of 100 tons have not been evaluated by the licensee for their impact on the structural integrity of the cask storage pool. The licensee currently restricts loads carried by the Unit 2 cask-handling crane to 100 tons in St. Lucie Unit 2 TS 3.9.12.
Dropping a cask that exceeds the assumptions of the cask drop analysis into the cask storage pool is an unanalyzed event that could cause damage to the pool structure and possibly result in uncontrolled loss of coolant from the spent fuel pool. Loss of coolant in the spent fuel pool could challenge the integrity of the fission product barrier of the stored fuel through the loss of cooling capability. The licensees cask drop analysis for Unit 1 determined that cask crane loads exceeding 25 tons, if dropped, could cause structural damage and uncontrolled leakage from the spent fuel pool. As a result, the spent fuel cask load limit currently in the Unit 1 TSs represents an operating restriction that is an initial condition of a design basis accident that presents a challenge to the integrity of a fission product barrier due to the consequences of load drops in excess of 25 tons. With no plant design feature available to prevent operators from immediately transporting loads in excess of those assumed in cask drop analysis over the cask storage pool, the staff concluded that the cask crane load limit meets 10 CFR 50.36, Criterion 2, and St. Lucie Unit 1 TS 3.9.13 cannot be relocated to the Unit 1 UFSAR.
Similarly, the licensee evaluated a 100-ton cask drop event for Unit 2, concluding that the spent fuel would remain covered with water even if the cask storage pool empties completely due to a cask drop event. However, the St. Lucie Unit 2 cask crane is capable of lifting loads up to 150 tons. It is not known what effect a cask drop in excess of the current load limit would have on the structural integrity of the cask storage pool or the spent fuel pool structure. Cask drops in excess of 100 tons could cause structural damage and uncontrolled leakage from the spent fuel pool, resulting in a challenge to the integrity of the fission product barrier of the stored fuel through the loss of cooling capability. With no plant design feature available to prevent operators from immediately transporting loads in excess of those assumed in the cask drop analysis over the cask storage pool, the current cask crane load limit would be an operating restriction needed to preclude operating in an unanalyzed condition and the 100-ton cask crane load limit would meet 10 CFR 50.36, Criterion 2.
However, subsequent to the initial submittal, the licensee initiated a modification to replace the original cask handling cranes for Unit 1 and Unit 2 with new single-failure-proof cranes. In its December 11, 2003, supplement, the licensee indicated that the replacement cranes were installed in the fourth quarter of 2003. The licensee states that main hoist for each replacement crane meets the applicable guidance for single-failure-proof handling systems found in NUREG-0612; the replacement cranes have been designed, fabricated, installed, and tested to the guidance of NUREG-0554; and the new cranes have been successfully load-tested to 150 tons.
With new single-failure-proof cranes installed, the licensee intends to revise the UFSAR for each unit to remove the cask-drop accident from the licensing basis. With the cask-drop accident removed from the licensing basis, Criterion 2 of 10 CFR 50.36 no longer applies, and the crane load limits may be relocated from the TSs to the UFSAR. Therefore, the proposed amendments are acceptable.
3.3 Proposed Control of Relocated Requirements The licensee proposed to relocate the current TS LCOs listed in Section 1.0 of this SE to the respective units UFSARs. Although the St. Lucie UFSAR is a licensee-controlled document, the design, operation, and control of the affected systems and equipment, as described in the UFSAR, cannot be changed without prior NRC approval unless the criteria of 10 CFR 50.59, Changes, tests and experiments, are met. For example, changes to the design of the fuel that results in the fuel assembly weight changing would be evaluated under 10 CFR 50.59 for impact on the FHA analysis.
For those requirements approved for relocation, the NRC staff considers the UFSAR to have an appropriate level of control under 10 CFR 50.59.
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
These amendments change 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 amendments involve 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 amendments involve no significant hazards consideration and there has been no public comment on such finding (67 FR 50954, dated August 6, 2002). Accordingly, these amendments meet 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 these amendments.
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 amendments will not be inimical to the common defense and security or to the health and safety of the public for changes to Units 1 and 2 TS 3.9.7.
Principal Contributors: C. Gratton, NRR R. Dennig, NRR B. Moroney, NRR Date: April 28, 2004
Mr. J. A. Stall ST. LUCIE PLANT Florida Power and Light Company cc:
Senior Resident Inspector Mr. G. L. Johnston St. Lucie Plant Plant General Manager U.S. Nuclear Regulatory Commission St. Lucie Nuclear Plant P.O. Box 6090 6351 South Ocean Drive Jensen Beach, Florida 34957 Jensen Beach, Florida 34957 Craig Fugate, Director Mr. Terry Patterson Division of Emergency Preparedness Licensing Manager Department of Community Affairs St. Lucie Nuclear Plant 2740 Centerview Drive 6351 South Ocean Drive Tallahassee, Florida 32399-2100 Jensen Beach, Florida 34957 M. S. Ross, Managing Attorney Mr. David Moore Florida Power & Light Company Vice President, Nuclear Operations Support P.O. Box 14000 Florida Power & Light Company Juno Beach, FL 33408-0420 P.O. Box 14000 Juno Beach, FL 33408-0420 Mr. Douglas Anderson County Administrator Mr. Rajiv S. Kundalkar St. Lucie County Vice President - Nuclear Engineering 2300 Virginia Avenue Florida Power & Light Company Fort Pierce, Florida 34982 P.O. Box 14000 Juno Beach, FL 33408-0420 Mr. William A. Passetti, Chief Department of Health Mr. J. Kammel Bureau of Radiation Control Radiological Emergency 2020 Capital Circle, SE, Bin #C21 Planning Administrator Tallahassee, Florida 32399-1741 Department of Public Safety 6000 SE. Tower Drive Mr. William Jefferson, Jr. Stuart, Florida 34997 Site Vice President St. Lucie Nuclear Plant Marjan Mashhadi, Senior Attorney 6351 South Ocean Drive Florida Power & Light Company Jensen Beach, Florida 34957-2000 801 Pennsylvania Ave., NW Suite 220 Washington, DC 20004