ML100210321

License Amendments, Removed Structural Integrity Requirements and Technical Specification Improvement to Extend Inspection Interval for Reactor Coolant Pump Flywheels

ML100210321
Person / Time
Site:	Turkey Point
Issue date:	02/23/2010
From:	Jason Paige Plant Licensing Branch II
To:	Nazar M Florida Power & Light Co
paige, Jason, NRR/DORL,301-415-5888
References
TAC ME0701, TAC ME0702
Download: ML100210321 (19)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 23, 2010 Mr. Mano Nazar Executive Vice President and Chief Nuclear Officer Florida Power and Light Company P.O. Box 14000 Juno Beach, Florida 33408-0420 SUB~IECT: TURKEY POINT UNITS 3 AND 4 - ISSUANCE OF AMENDMENTS REGARDING TECHNICAL SPECIFICATION CHANGE ASSOCIATED WITH REMOVAL OF STRUCTURAL INTEGRITY REQUIREMENTS AND TECHNICAL SPECIFICATION IMPROVEMENT TO EXTEND THE INSPECTION INTERVAL FOR REACTOR COOLANT PUMP FLYWHEELS (TAC NOS. ME0701 AND ME0702)

Dear Mr. Nazar:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 242 to Renewed Facility Operating License No. DPR-31 and Amendment No. 238 to Renewed Facility Operating License No. DPR-41 for the Turkey Point Plant, Units Nos. 3 and 4, respectively. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated February 16, 2009.

The amendments remove the structural integrity requirements contained in TS 3/4.4.10, and its associated Bases from the Turkey Point TSs. The proposed amendment also relocates the reactor coolant pump (RCP) motor flywheel inspection requirements in Surveillance Requirement (SR) 4.4.10 to SR 4.0.5 and revises the RCP motor flywheel inspection frequency from the currently approved 10-year inspection interval, to an interval not to exceed 20 years.

A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Si~rIY' t 1 ; : j e c t Manager

~:~tnL~enSing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-250 and 50-251

Enclosures:

1. Amendment No. 242 to DPR-31

2. Amendment No. 238 to DPR-41

3. Safety Evaluation cc w/enclosures: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 FLORIDA POWER AND LIGHT COMPANY DOCKET NO. 50-250 TURKEY POINT PLANT, UNIT NO.3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 242 Renewed License No. DPR-31

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Florida Power and Light Company (the licensee) dated February 16, 2009, 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.

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2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Renewed Facility Operating License No. DPR-31 is hereby amended to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment 1\10. 242 are hereby incorporated into this renewed license. The Environmental Protection Plan contained in Appendix B is hereby incorporated into this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days within issuance.

F. 0. R~E NUCLEARr.. E EGG~U.~ATORY. COMMIS.SION

,

~~V~~

~renda L Mozafari, Chief (~iZ4 Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Operating License and Technical Specifications Date of Issuance: February 23, 2010

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 FLORIDA POWER AND LIGHT COMPANY DOCKET NO. 50-251 TURKEY POINT PLANT UNIT NO.4 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 238 Renewed License No. DPR-41

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Florida Power and Light Company (the licensee) dated February 16, 2009, 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.

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2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Renewed Facility Operating License No. DPR-41 is hereby amended to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 238 are hereby incorporated into this renewed license. The Environmental Protection Plan contained in Appendix B is hereby incorporated into this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

. YJ/fM'

~;enda L Mozafari, Chief Plant Licensing Branch 11-2 (A:Jli~

Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Operating License and Technical Specifications Date of Issuance: February 23,2010

ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. 242 RENEWED FACILITY OPERATING LICENSE NO. DPR-31 AMENDMENT NO. 238 RENEWED FACILITY OPERATING LICENSE NO. DPR-41 DOCKET NOS. 50-250 AND 50-251 Replace Page 3 of Renewed Operating License DPR-31 with the attached Page 3.

Replace Page 3 of Renewed Operating License DPR-41 with the attached Page 3.

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 Index-viii Index-viii 3/4 0-4 3/4 0-4 3/44-38

3 E. Pursuant to the Act and 10 CFR Parts 40 and 70 to receive, possess, and use at any time 100 milligrams each of any source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactively contaminated apparatus; F. Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of Turkey Point Units Nos. 3 and 4.

3. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect, and is subject to the additional conditions specified below:

A. Maximum Power Level The applicant is authorized to operate the facility at reactor core power levels not in excess of 2300 megawatts (thermal).

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 242 are hereby incorporated into this renewed license. The Environmental Protection Plan contained in Appendix B is hereby incorporated into this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

C. Final Safety Analysis Report The licensee's Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21 (d), as revised on November 1,2001, describes certain future inspection activities to be completed before the period of extended operation.

The licensee shall complete these activities no later than July 19, 2012.

The Final Safety Analysis Report supplement as revised on November 1,2001, described above, shall be included in the next scheduled update to the Final Safety Analysis Report required by 10 CFR 50.71(e)(4), following the issuance of this renewed license. Until that update is complete, the licensee may make changes to the programs described in such supplement without prior Commission approval, provided that the licensee evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

Renewed License No. DPR-31 Amendment No. 242 I

3 E. Pursuant to the Act and 10 CFR Parts 40 and 70 to receive, possess, and use at any time 100 milligrams each of any source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactively contaminated apparatus; F. Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of Turkey Point Units Nos. 3 and 4.

3. This renewed operating license shall be deemed to contain and is sUbject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect, and is subject to the additional conditions specified below:

A. Maximum Power Level The applicant is authorized to operate the facility at reactor core power levels not in excess of 2300 megawatts (thermal).

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 238 are hereby incorporated into this renewed license. The Environmental Protection Plan contained in Appendix B is hereby incorporated into this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

C. Final Safety Analysis Report The licensee's Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21 (d), as revised on November 1,2001, describes certain future inspection activities to be completed before the period of extended operation.

The licensee shall complete these activities no later than April 10, 2013.

The Final Safety Analysis Report supplement as revised on November 1,2001, described above, shall be included in the next scheduled update to the Final Safety Analysis Report required by 10 CFR 50.71 (e)(4), following the issuance of this renewed license. Until that update is complete, the licensee may make changes to the programs described in such supplement without prior Commission approval, provided that the licensee evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

Renewed License No. DPR-41 AMENDMENT NO. 238 I

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION 3/4.4.9 PRESSURE/TEMPERATURE LIMITS Reactor Coolant System................................................................................. 3/44-30 FIGURE 3.4-2 TURKEY POINT UNITS 3&4 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 32 EFPY 3/44-31 FIGURE 3.4.3 TURKEY POINT UNITS 3&4 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS (100°F/hr)

APPLICABLE UP TO 32 EFPY 3/44-32 TABLE 3.4.5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITHDRAWAL SCHEDULE 3/44-34 Pressurizer 3/44-35 Overpressure Mitigating Systems................................................................... 3/44-36 3/4.4.10 DELETED 3/4 4-38 3/4.4.11 REACTOR COOLANT SYSTEM VENTS....................................................... 3/4 4-39 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS 3/45-1 3/4/5.2 ECCS SUBSYSTEMS - Tavg GREATER THAN OR EQUAL TO 350°F.......... 3/45-3 FIGURE 3.5-1 RHR PUMP CURVE....................................................................................... 3/45-6 3/4.5.3 ECCS SUBSYSTEMS - Tavg LESS THAN 350°F 3/4 5-9 3/4.5.4 REFUELING WATER STORAGE TANK........................................................ 3/45-10 TURKEY POINT - UNITS 3 & 4 viii AMENDMENT NOS. 242 AND 238

APPLICABI L1TY SURVEILLANCE REQUIREMENTS (CONTINUED)

b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda and the ASME OM Code and applicable Addenda shall be applicable as follows in these Technical Specifications:

ASME Boiler and Pressure Vessel Code and the ASME OM Code and applicable Addenda terminology for Required frequencies for inservice inspection and testing performing inservice inspection activities and testing activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

c. The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing activities.

d. Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements.

e. Nothing in the ASME Boiler and Pressure Vessel Code or the ASME OM Code shall be construed to supersede the requirements of any Technical Specification.

f. Each reactor coolant pump flywheel shall be inspected at least once every 20 years, by either conducting an in-place ultrasonic examination over the volume from the inner bore of the flywheel to the circle of one-half the outer radius, or conduct a surface examination (magnetic particle and/or liquid penetrant) of exposed surfaces of the disassembled flywheel.

4.0.6 Surveillance Requirements shall apply to each unit individually unless otherwise indicated as stated in Specification 3.0.5 for individual specifications or whenever certain portions of a specification contain surveillance parameters different for each unit, which will be identified in parentheses, footnotes or body of the requirement.

TURKEY POINT - UNITS 3 & 4 3/40-4 AMENDMENT NOS. 242 AND 238

REACTOR COOLANT SYSTEM 3/4.4.10 DELETED TURKEY POINT - UNITS 3 & 4 3/44-38 AMENDMENT NOS. 242 AND 238

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 242 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-31 AND AMENDMENT NO. 238 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-41 FLORIDA POWER AND LIGHT COMPANY TURKEY POINT PLANT, UNIT NOS. 3 AND 4 DOCKET NOS. 50-250 AND 50-251

1.0 INTRODUCTION

By application dated February 16, 2009, Florida Power and Light Company (FPL, the licensee) proposed an amendment to the Technical Specifications (TSs) for Turkey Point Plant, Units 3 and 4. The requested changes would remove the structural integrity requirements contained in TS 3/4.4.10, and its associated Bases from the Turkey Point TSs. The proposed amendment also relocates the reactor coolant pump (RCP) motor flywheel inspection requirements in Surveillance Requirement (SR) 4.4.10 to SR 4.0.5 and revises the RCP motor flywheel inspection frequency from the currently approved 1O-year inspection interval, to an interval not to exceed 20 years. The Bases for TS 3/4.4.10 will be deleted and the Bases revised to address the change in the RCP flywheel inspection interval. These Bases changes will be performed under the Turkey Point TS Bases Control Program, and are not included with the submittal.

2.0 REGULATORY EVALUATION

2.1 Removal of TS 3/4.4.10 and Associated Bases Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TSs as part of the license. The Commission's regulatory requirements related to the content of the TSs are contained in Title 10, Code of Federal Regulations (10 CFR) Section 50.36. The TS requirements in 10 CFR 50.36 include the following categories: (1) safety limits, limiting safety systems settings and limiting control settings, (2) limiting conditions for operation, (3) surveillance requirements, (4) design features, and (5) administrative controls.

The licensee stated in its February 16, 2009 letter that on July 22, 1993, the Commission issued its Final Policy Statement on Technical Specifications for Nuclear Power Reactors, stating that satisfying the guidance in the policy statement also satisfies Section 182a of the Atomic Energy Act and 10 CFR 50.36. The Final Policy Statement gave guidance for evaluating the required scope of the TSs and defined the guidance criteria to be used in determining which of the limiting conditions for operation (LCOs) and associated SRs should remain in the TSs. The

- 2 Commission stated that, in allowing certain items to be relocated to licensee-controlled documents while requiring that other items be retained in the TSs, it was adopting the qualitative standard enunciated by the Atomic Safety and Licensing Appeal Board. 1 There, the Appeal Board observed:

[T]here is neither a statutory nor a regulatory requirement that every operational detail set forth in an applicant's safety analysis report (or equivalent) be subject to a technical specification, to be included in the license as an absolute condition of operation which is legally binding upon the licensee unless and until changed with specific Commission approval. Rather, as best we can discern it, the contemplation of both the Act and the regulations is that technical specifications are to be reserved for those matters as to which the imposition of rigid conditions or limitations upon reactor operation is deemed necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety.

Pursuant to 10 CFR 50.36(c)(2)(ii), a TS is required for an LCO meeting one or more of the following criteria:

(1) Installed instrumentation that is used to detect, and indicate in a control room, a significant abnormal degradation of the reactor coolant pressure boundary; (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; (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; and (4) A structure, system or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

As a result, existing LCO requirements which fall within or satisfy any of the criteria in 10 CFR 50.36(c)(2)(ii) must be retained in the TS while those LCO requirements which do not fall within or satisfy these criteria may be relocated to other licensee-controlled documents.

2.2 Extension of RCP Motor Flywheel Inspection Interval The function of the RCP in the reactor coolant system (RCS) of a pressurized water reactor plant is to maintain an adequate cooling flow rate by circulating a large volume of primary coolant water at high temperature and pressure through the RCS. Following an assumed loss of power to the RCP motor, the flywheel, in conjunction with the impeller and motor assembly, provides sufficient rotational inertia to assure adequate primary coolant flow during RCP coastdown, thus 1 Portland General Electric Co. (Trojan Nuclear Plant), ALAB-531 ,9 NRC 263, 273 (1979).

-3 resulting in adequate core cooling. A concern regarding the overspeed of the RCP and its potential for failure led to the issuance of Regulatory Guide (RG) 1.14, "Reactor Coolant Pump Flywheel Integrity," Revision 1, dated AUgust 1975. RG 1.14 describes a method acceptable to the Nuclear Regulatory Commission (NRC, the Commission) staff of addressing concerns related to RCP vibration and the possible effects of missiles that might result from the failure of the RCP flywheel. The need to protect components important to safety from such missiles are addressed in General Design Criterion 4, "Environmental and Dynamic Effects Design Basis," of Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, "Licensing of Production and Utilization Facilities," which is applicable to plants that obtained their construction permits after May 21, 1971.

Specific requirements to have an RCP Flywheel Inspection Program consistent with RG 1.14 or previously-issued relaxations from the RG are included in the Administrative Controls Section of the TSs. The purpose of the testing and inspection programs defined in the TSs is to ensure that the probability of a flywheel failure is sufficiently small such that additional safety features are not needed to protect against a flywheel failure. The RG provides criteria in terms of critical speeds that could result in the failure of an RCP flywheel during normal or accident conditions.

In addition to the guidance in RG 1.14, the NRC has more recently issued RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," which provides guidance and criteria for evaluating proposed changes that use risk-informed justifications.

A proposed justification for extending the RCP flywheel inspections from a 1O-year inspection interval to an interval not to exceed 20 years was provided by the Westinghouse Owners Group (WOG) in topical report WCAP-15666, "Extension of Reactor Coolant Pump Motor Flywheel Examination," transmitted by letter dated August 24, 2001. The topical report addressed the proposed extension for all domestic WOG plants. Turkey Point Plant, Units 3 and 4 are 3-loop Westinghouse plants. The NRC accepted the topical report for referencing in license applications by letter and Safety Evaluation (SE) dated May 5,2003 (Agencywide Documents Access and Management System Accession No. ML031250595).

The proposed changes would extend the RCP motor flywheel examination frequency from the currently approved 1O-year inspection interval to an interval not to exceed 20 years. These changes are based on Technical Specification Task Force (TSTF) change traveler TSTF-421 (Revision 0) that has been approved generically for the Westinghouse Standard Technical Specifications (STSs), NUREG-1431. A notice announcing the availability of this proposed TS change using the Consolidated Line Item Improvement Process was published in the Federal Register on October 22, 2003 (68 FR 60422).

3.0 TECHNICAL EVALUATION

3.1 Removal of TS 3/4.4.10 and Associated Bases The licensee stated that the purpose of TS 3/4.4.10, "Structural Integrity", is to specify the requirements of maintaining the structural integrity of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Class 1,2 and 3 components. This specification was originally intended to support assurance that structural integrity and operational readiness of these components are maintained at an acceptable level throughout the

- 4 life of the facility. The specification is applicable in all operational modes. However, the specification does not provide actions for plant shutdown if the LCO is not met. This is because the specification addresses the passive pressure boundary function of ASME Code Class 1, 2 and 3 components as established under the Inservice Inspection (lSI) program. The lSI program is required pursuant to 10 CFR 50.55a, Codes and standards and SR 4.0.5.

The licensee states that maintaining a program-type requirement within an LCO creates significant interpretation issues for Operations personnel. The RCS structural integrity TS was part of the original TS and, therefore no basis history is available regarding its intent. However, the TS 3/4.4.10 appears to have been included to help ensure that plant heatup and startup would not occur until all required portions of applicable systems were verified to meet lSI acceptance criteria following inspections performed during a plant outage (normally performed during refueling outages).

Meeting these acceptance criteria helps ensure the integrity of applicable systems during all modes of operation, including accident events. For instance, the RCS pressure boundary is purposefully breached during Modes 5 and 6 operations to support plant outage activities and such breaches are not historically considered a violation of TS 3/4.4.10. Furthermore, TS 3/4.4.10 contains no action suggesting it was designed to accommodate integrity concerns once plant heatup has commenced. Structural integrity lSI activities are performed only during plant outages when conditions exist that permit access to the RCS pressure boundary and are not monitored or controlled through application of the lSI program during the operational cycle.

The licensee stated other TSs are designed to monitor the structural integrity of the RCS during operation and provide actions to shut down the unit if compliance is not maintained. For example, reactor coolant heatup and cooldown rates (TSs 3.4.9.1 and 3.4.9.2) and the overpressure mitigation system (TS 3.4.9.3) protect against applying undue stresses as a result of pressure/temperature transients on RCS components and piping. The RCS leakage TSs (3.4.6.1 and 3.4.6.2) provides a means of protecting the RCS integrity by detecting and monitoring leakage. Therefore, the licensee stated it is not necessary to apply TS 3/4.4.10 when integrity issues become evident during plant operation above cold shutdown. Because TS 3/4.4.10 is redundant to other regulations, it is acceptable to remove TS 3/4.4.10 from the TSs.

Finally, removal of this specification does not reduce the controls that are necessary to ensure compliance with the ASME Code. Structural Integrity is maintained by compliance with 10 CFR 50.55a, as implemented through the Turkey Point lSI Program required by TS 4.0.5, as well as by compliance with TSs 3.4.6.1, 3.4.6.2, 3.4.9.1, 3.4.9.2 and 3.4.9.3 for the RCS.

The TS changes proposed by FPL in this license amendment request are required to be evaluated to confirm compliance with the regulatory requirements in Section 2.0 of this SE. The licensee's basis for each finding is discussed in the following paragraphs.

For Criterion 1, the licensee stated the RCS ASME Code Class 1, 2, and 3 components do not include any instrumentation. Therefore, the NRC staff finds that TS 3/4.4.10 does not meet Criterion 1.

-5 For Criterion 2, the licensee stated structural integrity is neither a process variable, design feature, or operating restriction that is an initial condition of a design-basis analysis (DBA) or transient analysis. Structural integrity is not monitored or controlled during plant operation; it is verified during periodic inspections. Therefore, the NRC staff finds that TS 3/4.4.10 does not meet Criterion 2.

For Criterion 3, the licensee stated ASME Code Class 1, 2, and 3 components that are part of the primary success path and function to mitigate DBAs or transients that either assume the failure of, or present a challenge to, the integrity/operability of these components, are included in the individual specification that cover these components. The portion of this TS that is proposed to be removed addresses only the passive pressure boundary function of these components.

Therefore, the NRC staff finds that TS 3/4.4.10 does not meet Criterion 3.

For Criterion 4, the licensee stated the requirements covered by this TS that are being removed have not been shown to be risk significant to pUblic health and safety by either operating experience or probabilistic safety assessment. In addition, failure modes of applicable structures, systems or components (SSCs) would not be identified from the requirements of this TS. Furthermore, the requirements of this TS do not affect the risk review/unavailability monitoring of applicable SSCs. Therefore, the NRC staff finds that TS 3/4.4.10 does not meet Criterion 4.

The review for the structural integrity LCO relocation was actually performed and presented in a split report from the Director of NRR, Thomas Murley, on May 8 1988. Originally, NUREG-0212, "Standard Technical Specifications for Combustion Engineering" contained provisions for the LCOs and SRs in reference to the structural integrity of ASME Code Class 1, 2, and 3 components. This split report identified Section 3/4.4.10 "Structural Integrity" as not meeting the criterion for 10 CFR 50.36 and was, therefore, removed from subsequent revisions of the STSs.

This conclusion is consistent with NUREG-1431, "Standard Technical Specifications Westinghouse Plants," Revision 3.0, dated March 2004.

Therefore, since this TS does not fulfill any of the 10 CFR 50.36(c)(2)(ii) criteria for items that TSs must be established, the staff finds that removing TS 3/4.4.10 and the associated bases is acceptable. Finally, the removal of TS 3/4.4.10 and its associated references to structural integrity eliminates the redundancy of structural integrity requirements from the TSs that are already covered under 10 CFR 50.55a. Therefore, the NRC staff finds this proposed change to be acceptable.

Normally in applying the Commission Final Policy Statement on Technical Specifications for Nuclear Power Reactors, the NRC staff would require that a licen'see identify both the licensee-controlled document receiving a relocated TS and the change control mechanism that governs that document. However, in this instance, the licensee proposes deletion without relocation of the TS. The NRC staff finds this proposed deletion without relocation to be acceptable because the ASME Code Class 1, 2, and 3 structural integrity requirements continue to be covered under 10 CFR 50.55a, with which the licensee must comply with.

-6 3.2 Extension of RCP Motor Flywheel Inspection Interval The justification for the proposed change was provided in WCAP-15666, which the NRC staff accepted for referencing in license applications by letter and SE dated May 5, 2003. The topical report addresses the three critical speeds defined in RG 1.14: (a) the critical speed for ductile failure, (b) the critical speed for non-ductile failure, and (c) the critical speed for excessive deformation of the flywheel. The NRC staff found that the topical report adequately addressed these issues and demonstrated that acceptance criteria, for normal and accident conditions defined in RG 1.14, would continue to be met for all domestic WOG plants following an extension of the inspection interval. The topical report also provided a risk assessment for extending the RCP flywheel inspection interval. The NRC staffs review, documented in the SE for the topical report, determined that the analysis methods and risk estimates are acceptable when compared to the guidance in RG 1.174.

The technical basis of the change, as described in the NRC-approved topical report WCAP-15666, remains valid for the licensee's proposal to adopt the 20-year inspection interval.

The licensee has maintained the inspection methods consistent with RG 1.14 currently in its TSs and is proposing to adopt only the increased inspection interval.

In conclusion, the NRC staff finds that the regulatory positions in RG 1.14 concerning the three critical speeds are satisfied. The potential for failure of the RCP flywheel is, and will continue to be, negligible during normal and accident conditions. The change is, therefore, 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

These amendments involve a change in the 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 (74 FR 18255). 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 the 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

-7 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.

Principal Contributor: Keith M. Hoffman Date: February 23, 2010

IVIL100210321 *via memo NRR-058 OFFICE LPL2-2/PM LPL2-2/LA ITSB/BC CPNB/BC OGC NLO LPL2-2/BC(A)

NAME JPaige BClayton RElliott TLupold* BMizuno BMozafari DATE 01/25/10 01/22/10 02/02/10 01/19/10 02/19/10 02/23/10