ML16034A080

Issuance of Amendments Regarding Technical Specification Change to Remove Communications and Manipulator Crane Requirements and Relocate to Licensee-Controlled Documents

ML16034A080
Person / Time
Site:	Saint Lucie
Issue date:	03/07/2016
From:	Perry Buckberg Plant Licensing Branch II
To:	Nazar M Nextera Energy
Buckberg P, NRR/DORL/LPLII-2
References
CAC MF5835, CAC MF5836
Download: ML16034A080 (25)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 7, 2016 Mr. Mano Nazar President and Chief Nuclear Officer Nuclear Division NextEra Energy P.O. Box 14000 Juno Beach, FL 33408-0420

SUBJECT:

ST. LUCIE PLANT, UNIT NOS. 1 AND 2 - ISSUANCE OF AMENDMENTS REGARDING TECHNICAL SPECIFICATION CHANGE TO REMOVE COMMUNICATIONS AND MANIPULATOR CRANE REQUIREMENTS AND RELOCATE TO LICENSEE-CONTROLLED DOCUMENTS (CAC NOS. MF5835 AND MF5836)

Dear Mr. Nazar:

The U.S. Nuclear Regulatory Commission (NRC or the Commission) has issued the enclosed Amendment Nos. 230 and 180 to Renewed Facility Operating License Nos. DPR-67 and NPF-16 for the St. Lucie Plant, Unit Nos. 1 and 2, respectively. These amendments consist of changes to the Technical Specifications (TSs) in response to Florida Power & Light Company's application dated March 10, 2015, as supplemented by a letter dated December 15, 2015.

The amendments remove TS 3/4.9.5 related to communication during core alteration and TS 3/4.9.6 related to manipulator crane operability from the TSs and require inclusion of those specifications in the Updated Final Safety Analysis Report, which the licensee is required to control by the provisions set forth in Title 10 of the Code of Federal Regulations (10 CFR),

Section 50.59, "Changes, tests, and experiments." The removal of TS 3/4.9.5 and TS 3/4.9.6 is consistent with the requirements in 10 CFR 50.36, "Technical specifications."

M. Nazar The NRC staff's safety evaluation of the amendments is enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Since/,

.***tjt~*

Perry H. Buckberg, Senior Project Manager Plant Licensing Branch 11-2 Division of Operator Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-335 and 50-389

Enclosures:

1. Amendment No. 230 to DPR-67

2. Amendment No. 180 to NPF-16

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. 50-335 ST. LUCIE PLANT UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERA TING LICENSE Amendment No. 230 Renewed License No. DPR-67

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

A. The application for amendment by Florida Power & Light Company (FPL, the licensee), dated March 10, 2015, as supplemented by a letter dated December 15, 2015, 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;

8. 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. 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.230 , are hereby incorporated in the renewed license. FPL shall operate the facility in accordance with the Technical Specifications.

1. This license amendment is effective as of its date of issuance and shall be implemented within 60 days. 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 was described in the licensee's application dated March 10, 2015, and evaluated in the staff's safety evaluation attached to this amendment.

FOR THE NUCLEAR REGULATORY COMMISSION 73;-~__J;l~

Benjamin G. Beasley, Chief Plant Licensing Branch 11-2 Division of Operator Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: lVB:rdl 7 , 2016

ATIACHMENT TO LICENSE AMENDMENT NO. 230 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-67 DOCKET NO. 50-335 Replace Page 3 of Renewed Operating License DPR-67 with the attached Page 3.

Replace the following pages of Appendix A, Technical Specifications, with the attached pages.

The revised pages are identified by amendment number and contains a vertical line indicating the area of change.

Remove Page Insert Page VIII VIII 3/4 9-5 3/4 9-5 3/4 9-6 3/4 9-6

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 or incorporated below:

A. Maximum Power Level FPL is authorized to operate the facility at steady state reactor core power levels not in excess of 3020 megawatts (thermal).

B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 230 are hereby incorporated in the renewed license.

FPL shall operate the facility in accordance with the Technical Specifications.

Appendix B, the Environmental Protection Plan (Non-Radiological), contains environmental conditions of the renewed license. If significant detrimental effects or evidence of irreversible damage are detected by the monitoring programs required by Appendix B of this license, FPL will provide the Commission with an analysis of the problem and plan of action to be taken subject to Commission approval to eliminate or significantly reduce the detrimental effects or damage.

C. Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21 (d), as revised on March 28, 2003, describes certain future activities to be completed before the period of extended operation. FPL shall complete these activities no later than March 1, 2016, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement as revised on March 28, 2003, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50. 71 (e)(4), following issuance of this renewed license. Until that update is complete, FPL may make changes to the programs described in such supplement without prior Commission approval, provided that FPL evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

D. Sustained Core Uncoverv Actions Procedural guidance shall be in place to instruct operators to implement actions that are designed to mitigate a small-break loss-of-coolant accident prior to a calculated time of sustained core uncovery.

Renewed License No. DPR-67 Amendment No. 230

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION .............................................................................. 3/4 9-1 3/4.9.2 INSTRUMENTATION ..... :................................................................................... 314 9-2 3/4.9.3 DECAY TIME ...................................................................................................... 3/4 9-3 3/4.9.4 CONTAINMENT PENETRATIONS ..................................................................... 3/4 9-4 3/4.9.5 DELETED ........................................................................................................... 3/4 9-5 3/4.9.6 DELETED ........................................................................................................... 314 9-6 3/4.9.7 DELETED ........................................................................................................... 3/4 9-7 3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION ................................. 3/4 9-8 3/4.9.9 CONTAINMENT ISOLATION SYSTEM .............................................................. 3/4 9-9 3/4.9.10 WATER LEVEL- REACTOR VESSEL ............................................................ 3/4 9-10 3/4.9.11 SPENT FUEL STORAGE POOL ...................................................................... 3/4 9-11 3/4.9.12 FUEL POOL VENTILATION SYSTEM - FUEL STORAGE .............................. 3/4 9-12 3/4.9.13 DELETED ......................................................................................................... 3/4 9-15 3/4.9.14 DELETED ......................................................................................................... 3/4 9-16 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN .................................................................................... 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS ......................................................................................................... 3/4 10-2 3/4.10.3 DELETED ......................................................................................................... 314 10-3 3/4.10.4 DELETED ......................................................................................................... 3/4 10-4 3/4.10.5 CENTER CEA MISALIGNMENT ...................................................................... 314 10-5 ST. LUCIE - UNIT 1 VIII Amendment No. 4, 24, 23, 2:+, W, 4-W, ~, 230

INTENTIONALLY DELETED ST. LUCIE - UNIT 1 3/4 9-5 Amendment No. ~. ~ , 230

INTENTIONALLY DELETED ST. LUCIE - UNIT 1 3/4 9-6 Amendment No. ~ , 230

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 FLORIDA POWER AND LIGHT COMPANY DOCKET NO. 50-389 ST. LUCIE PLANT, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 180 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 (FPL, the licensee), dated March 10, 2015, as supplemented by a letter dated December 15, 2015, 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. 180, are hereby incorporated in the renewed 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. 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 was described in the licensee's application dated March 10, 2015, and evaluated in the staff's safety evaluation attached to this amendment.

FOR THE NUCLEAR REGULATORY COMMISSION 7llr~~-d ~/

Benjamin G. Beasley, Chief Plant Licensing Branch 11-2 Division of Operator Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: M:m:h 7, 201 6

ATTACHMENT TO LICENSE AMENDMENT NO. 180 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-16 DOCKET NO. 50-389 Replace Page 3 of Renewed Operating License NPF-16 with the attached Page 3.

Replace the following pages of Appendix A, Technical Specifications, with the attached pages.

The revised pages are identified by amendment number and contain a vertical line indicating the area of change.

Remove Page Insert Page IX IX 3/4 9-5 3/4 9-5 3/4 9-6 3/4 9-6

neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required.

D. Pursuant to the Act and 10 CFR Parts 30, 40, and 70, FPL to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and E. Pursuant to the Act and 10 CFR Parts 30, 40, and 70, FPL to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

3. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission's regulations: 10 CFR Part 20, Section 30.34 of 10 FR Part 30, Section 40.41 of 10 CFR Part 40, Section 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 FPL is authorized to operate the facility at steady state reactor core power levels not in excess of 3020 megawatts (thermal).

B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 180 are hereby incorporated in the renewed license.

FPL shall operate the facility in accordance with the Technical Specifications.

Renewed License No. NPF-16 Amendment No. 180

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.8.2 D.C. SOURCES OPERATING ......................................................................................... 3/4 8-10 SHUTDOWN ......................................................................................... 3/4 8-13 3/4.8.3 ONSITE POWER DISTRIBUTION SYSTEMS OPERATING ......................................................................................... 3/4 8-14 SHUTDOWN ......................................................................................... 3/4 8-16 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION BYPASS DEVICES ...................................................... 314 8-17 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION .............................................................................. 3/4 9-1 3.4.9.2 INSTRUMENTATION ......................................................................................... 3/4 9-2 3/4.9.3 DECAY TIME ...................................................................................................... 314 9-3 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS ................................................... 3/4 9-4 3/4.9.5 DELETED ........................................................................................................... 314 9-5 3/4.9.6 DELETED ........................................................................................................... 314 9-6 3/4.9.7 DELETED ........................................................................................................... 3/4 9-7 3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION HIGH WATER LEVEL ............................................................................. 3/4 9-8 LOW WATER LEVEL .............................................................................. 314 9-9 3/4.9.9 CONTAINMENT ISOLATION SYSTEM ............................................................ 3/4 9-10 3/4.9.10 WATER LEVEL- REACTOR VESSEL ............................................................ 314 9-11 3/4.9.11 SPENT FUEL STORAGE POOL ...................................................................... 3/4 9-12 3/4.9.12 DELETED ......................................................................................................... 3/4 9-13 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN .................................................................................... 3/4 10-f 3/4.10.2 MODERATOR TEMPERATURE COEFFICIENT, GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS ........................................ 314 10-2 3/4.10.3 REACTOR COOLANT LOOPS ......................................................................... 314 10-3 3/4.10.4 CENTER CEA MISALIGNMENT ...................................................................... 314 10-4 3/4.10.5 CEA INSERTION DURING ITC, MTC, AND POWER COEFFICIENT MEASUREMENTS ............................................................................................ 3/4 10-5 ST. LUCIE - UNIT 2 IX Amendment No.~ . .i.M, 180

INTENTIONALLY DELETED ST. LUCIE - UNIT 2 3/4 9-5 Amendment No.~.~, 180

INTENTIONALLY DELETED ST. LUCIE - UNIT 2 3/4 9-6 Amendment No. 89, 180

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 230 AND 180 TO RENEWED FACILITY OPERATING LICENSE NOS. DPR-67 AND NPF-16 FLORIDA POWER AND LIGHT COMPANY ST. LUCIE PLANT. UNIT NOS. 1 AND 2 DOCKET NOS. 50-335 AND 50-389

1.0 INTRODUCTION

By application dated March 10, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15084A141), as supplemented by a letter dated December 15, 2015 (ADAMS Accession No. ML16005A146), Florida Power & Light Company (the licensee) submitted license amendment requests (LARs) for the St. Lucie Plant, Unit Nos. 1 and 2 (St. Lucie 1 and 2), to remove Technical Specifications (TSs) Limiting Condition for Operation (LCO) 3/4.9.5, "Communications," from both the St. Lucie 1 and 2 TSs; LCO 3/4.9.6, "Manipulator Crane Operability," from the St. Lucie 1 TSs; and LCO 3/4.9.6, "Manipulator Crane," from the St. Lucie 2 TSs and relocate these TS requirements to the Updated Final Safety Analysis Reports (UFSARs) for St. Lucie 1 and 2.

2.0 REGULATORY EVALUATION

2.1. Regulatory Discussion 2.1.1 Technical Specifications Section 182a of the Atomic Energy Act of 1954 (the Act), as amended, requires applicants for nuclear power plant operating licenses to include the TSs as part of the license. The Commission's regulatory requirements related to the content of TSs are set forth in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, "Technical specifications." The regulations require that the TSs include items in specific categories, including: (1) safety limits, limiting safety system settings, and limiting control settings; (2) LCOs; (3) Surveillances 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 (RCPB).

(B) Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a Design-Basis Accident (OBA) 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 (SSC) that is part of the primary success path and which functions or actuates to mitigate a OBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(D) Criterion 4. An SSC which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

The regulations in 10 CFR 50.36(c)(2) specify that, when an LCO of a nuclear reactor plant is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition is met.

2.1.2 Human Factors NRC human factors reviews address programs, procedures, training, plant design features, and operator manual actions related to operator performance during normal and accident conditions.

The NRC Staff conducted a human factors evaluation to confirm that operator performance would not be adversely affected as a result of changes to the St. Lucie 1 and 2 TSs. The review was based on the following regulatory guidance:

(A) NUREG-0800, "Standard Review Plan" Chapter 13.5.1.1 "Administrative Procedures-General" (B) NUREG-0711, "Human Factors Engineering Program Review Model" 2.2 System Description As described in the March 10, 2015, LAR, the Manipulator Crane (also referred to as the Refueling Machine) is a traveling bridge and trolley running on rails set in the concrete on each side of the refueling cavity pool inside the reactor building. Motors on the bridge and trolley accurately position the machine over each fuel assembly location within the reactor core or fuel transfer carrier. The hoist assembly contains an air operated grappling device that, when rotated by the actuator mechanism, engages the fuel assembly to be removed. The hoist assembly and grappling device are raised and lowered by a cable attached to the hoist winch.

During withdrawal or insertion of either a fuel assembly, or a fuel assembly with a control element inserted, the load on the hoist cable is monitored at the console to ensure that movement is not being restricted.

TS 3/4.9.5:

TS 3/4.9.5, "Communications," provides communications capability to ensure that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity condition during core alterations. As indicated in the LAR, the components covered by this LCO include radios and associated power and transmission equipment necessary to establish and maintain communications between the control room and the refueling station.

This only applies during core alterations, which can only be conducted with the reactor head removed and the Reactor Coolant System depressurized. Additionally, St. Lucie 1 and 2 UFSAR Sections 9.1.4.3 indicate there is direct communication between the control room and the refueling machine console during fuel-handling operations, allowing the control room personnel to inform the refueling machine operator of any impending unsafe condition detected from the control room during fuel movement.

TS 3/4.9.6 - Unit 1:

TS 3/4.9.6, "Manipulator Crane Operability," provides requirements for the cranes used for movement of control element assemblies or fuel assemblies to ensure each crane has adequate capacity to lift a fuel assembly and has protection against excessive lifting force in the event that an assembly becomes inadvertently stuck during lifting operations.

TS 3/4.9.6 - Unit 2:

TS 3/4.9.6, "Manipulator Crane," provides requirements for the refueling machine to ensure manipulator cranes will be used for movement of fuel assemblies; each crane has sufficient load capacity to lift a fuel assembly with or without a control element assembly; and protection of the core internals and pressure vessel from excessive lifting force in the event that an assembly becomes inadvertently stuck during lifting operations.

3.0 TECHNICAL EVALUATION

3.1 10 CFR 50.36(c)(2)(ii) Criteria In the March 10, 2015, LAR the licensee stated that the proposed change is to relocate sections of the TSs relating to certain refueling requirements. The licensee proposed relocating these refueling sections of the TSs to the UFSARs since these requirements are neither within the scope of equipment included in NUREG-1432, "Standard TSs for Combustion Engineering Plants," nor satisfy the criteria of 10 CFR 50.36(d)(2)(ii) for establishment of an LCO.

NUREG-1432 provides equipment and systems that are typically required in the TSs for Combustion Engineering plants. Communications equipment for refueling and manipulator crane specifications are not listed in NUREG-1432. Accordingly, the NRC Staff determined that the requirements proposed for location to the USFARs are not within the scope of NUREG-1432.

The staff's evaluation of the proposed changes against the 10 CFR 50.36(c)(2)(ii) criteria is as follows:

Criterion 1 :

TS 3/4.9.5:

The requirement related to communications between the control room and the refueling station are included in LCO 3.9.5 and described in Section 2.2 of this safety evaluation (SE). The function of the communication equipment is not part of any installed instrumentation that is used to detect abnormal degradation of the RCPB boundary.

TS 3/4.9.6:

The requirements related to manipulator crane (refueling machine) operability are located in Unit 1 LCO 3.9.6 and Unit 2 LCO 3.9.6 and described in Section 2.2 of this SE. These requirements are applicable during movement of control element assemblies or fuel assemblies within the reactor pressure vessel, which can only take place with the reactor head removed and the RCS depressurized. None of these operability requirements apply to instrumentation used to detect abnormal degradation of the RCPB boundary.

Therefore, the NRC staff concludes that none of the subject LCOs meet Criterion 1.

Criterion 2:

TS 3/4.9.5:

Equipment used by personnel to establish and maintain communications between the control room and the refueling station is not a process variable, design feature, or operating restriction that is an initial condition of a OBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, the staff concludes current LCO 3.9.5 does not satisfy Criterion 2 for inclusion in the TSs.

TS 3/4.9.6:

The requirement related to manipulator crane operability LCO 3.9.6 applies when handling fuel assemblies inside containment since the manipulator crane is unable to access the fuel building.

As noted in the LAR, two DBAs that could involve the manipulator crane have been identified for the St. Lucie 1 UFSAR: (1) UFSAR Section 15.3.3, "Inadvertent Loading of a Fuel Assembly into the Improper Position," and (2) UFSAR Section 15.4.3, "Fuel Handling Accident." For St. Lucie 2, DBAs involving the manipulator crane are similarly identified as being in Sections 15.4.7 and 15.7.4.1.2 of the UFSAR.

The application indicates that the manipulator crane may be involved in an accident related to inadvertent loading of a fuel assembly into the improper position, which is analyzed in UFSAR Sections 15.3.3 and 15.4. 7 for Unit 1 and Unit 2, respectively. LCO 3.9.6 contains operability requirements for crane capacity or excessive lifting force. These operability requirements are not related to the process used to ensure fuel assemblies are moved into the proper position in the core. LCO 3.9.6 does not prevent the misleading of a fuel assembly or otherwise involve the initiating conditions for this accident. Therefore, the NRC staff concludes that LCO 3.9.6

does not meet Criterion 2 for the Inadvertent Loading of a Fuel Assembly into an Improper Position OBA.

As noted in UFSAR Section 15.4.3 for Unit 1 and Section 15.7.4.1.2 for Unit 2, the applicable accident involving the manipulator crane is the postulated Fuel Handling Accident (FHA), which assumes a fuel assembly is dropped as an initial condition. This event consists of the drop of a single fuel assembly either in the Fuel-Handling Building (FHB) or inside of Containment. The manipulator crane cannot physically access the FHB, so it plays no role in the FHA postulated to occur there. For an FHA to occur inside of containment, the fuel assembly would be dropped from the manipulator crane. The effects and consequences of the design basis FHA involving the manipulator crane occurring inside containment are described in St. Lucie UFSAR Sections 15.4.3 and 15.7.4.1.2 for Unit 1 and Unit 2, respectively. The radiological consequences of the FHA were determined in accordance with the guidance in Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating DBAs at Nuclear Power Reactors." The St. Lucie FHA analysis assumed all fuel rods in a single fuel assembly are damaged as a result of the drop. The failure of all fuel rods is a conservative assumption that provides a bounding condition for determining the potential radiological consequences resulting from any level of damage to a single fuel assembly.

The LCO 3.9.6 associated with the manipulator crane specifies that the crane be operable with a minimum capacity for movement of fuel assemblies and an overload cut-off to protect the reactor vessel internals from excessive uplift force. Since the FHA assumes the manipulator crane drops a fuel assembly, the capacity of the crane is not an initial condition associated with the design basis FHA. Similarly, operation of the overload cutoff is not an initial condition of the design basis FHA because the assumption that all fuel pins are damaged in a single assembly bounds the potential damage to a fuel assembly that could result from excessive uplift forces.

Therefore, the NRC staff concludes that LCO 3.9.6 does not meet Criterion 2 for an FHA.

For additional safe-handling operation, the manipulator crane has physical, designed-in features that prevent the operators from inadvertently placing the plant in an unanalyzed condition.

Unit 1 UFSAR Section 7.6.1.2 and Unit 2 UFSAR Section 9.1.4.2.1.1 describe interlocks and operational constraints provided with the refueling machine design that interrupts hoisting of a fuel assembly if the load increases above the overload set point. As an additional protective feature, hoisting load is visually displayed so that the operator can manually terminate the operation if an overload or other unsafe condition occurs.

Based on the above information and conclusions, the staff determined that LCO 3.9.6 is unrelated to inadvertent fuel loading accidents and does not present an initial condition for fuel drop accident scenarios, and the consequences of such scenarios are bounded by those already considered in the UFSARs. Therefore, the NRC staff concludes that none of the subject LCOs meet Criterion 2.

Criterion 3:

TS 3/4.9.5 and TS 3/4.9.6:

The communication and equipment proposed for removal from TSs and relocation to licensee-controlled documents are not SSCs that are part of the primary success path and

which function or actuate to mitigate a OBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Similarly, the manipulator crane is used solely during refueling operations with the reactor head removed and does not actuate to mitigate a OBA or transient analysis. Therefore, the NRC staff concludes that none of the subject LCOs meet Criterion 3.

Criterion 4:

TS 3/4.9.5 and TS 3/4.9.6:

The manipulator crane and communication equipment included in the proposed TS change have not been shown in any operating experience or probabilistic risk assessment to be significant to public health and safety. Therefore, the NRC staff concludes that none of the subject LCOs meet Criterion 4.

3.2 Human Factors NRC's human factors reviews address programs, procedures, training, plant design features, and operator manual actions related to operator performance during normal and accident conditions. The staff conducted a human factors evaluation to confirm that operator performance would not be adversely affected as a result of the licensee's proposed changes to the TSs. The review was based on the regulatory guidance in NUREG-0800, Chapter 13.5.1.1 and NUREG-0711. Chapter 13.5.1.1 of NUREG-0800 provides a regulatory basis for reviewing changes involving administrative controls of cranes. NUREG-0711 provides guidance regarding the review of procedures. The purpose of the human factors review is to ensure operators have the equipment and information necessary for safe operation and minimizing the likelihood of errors while operating the plant. Therefore, the human factors review focuses on integrated performance of operators with equipment, procedures, and administrative controls and, therefore, relies on the guidance in NUREG-0800 and NUREG-0711.

The March 10, 2015, LAR indicates that information currently provided in two TSs will be deleted and moved to the UFSAR. Technical Specification 3/4.9.5, "Communications" provides administrative assurance that communication systems are available so control room staff can warn staff at the refueling station of potential safety issues during core alterations. Surveillance requirements provide evidence that communication systems are operable prior to fuel movement and at specified intervals during fuel movement. Similarly, TS 3/4.9.6, "Manipulator Crane" documents weight limits for the manipulator crane and auxiliary hoist. It also describes interlock settings to prevent overloading both devices. Surveillance requirements in both TSs provide evidence that the necessary systems are operable during movement of drive rods and fuel assemblies.

The staff requested that the licensee provide additional information regarding the content of the information that would be placed in the UFSAR. The December 15, 2015, letter is a response to a staff request for additional information and indicates that the licensee intends to reproduce the information currently in TS 3/4.9.5, "Communications" and TS 3/4.9.6, "Manipulator Crane" in its entirety in the UFSAR. This response ensures that no relevant information will become unavailable to operators as a result of this amendment request. Implementation of this LAR will not cause changes or deletions to the information available to operators, only to the location of

the information. Similarly, there are no changes to communication system equipment or other physical equipment used during manipulator crane operation as a result of this LAR. Job tasks completed by operators are not expected to change (with the exception that the information currently found in TSs 3/4.9.5 and 3/4.9.6 will now be found in the UFSAR rather than in the TSs). Similarly, surveillance intervals will be unchanged, maintaining the current level of protection with the communications and crane systems.

The LAR explains that the process described in 10 CFR 50.59 will be used, if this LAR is approved, to provide administrative control of the information in the UFSAR. Section 50.59 of 10 CFR requires the licensee to determine whether any changes to the requirements relocated in the UFSAR will need prior NRC approval via a license amendment. An NRC approved license amendment is required if the changes to the UFSAR may result in an increase in occurrence or frequency of accidents, or a failure or malfunction of an SSC. This provides reasonable assurance that information necessary to the operators will not be inadvertently removed in the future without due consideration to potential consequences of doing so.

The proposed license amendment preserves the information needed for operators to safely communicate and conduct operations using the manipulator crane. The 10 CFR 50.59 process will prevent important information from being removed without appropriate analyses. The proposed LAR maintains the same level of safety margin and protects information using previously approved administrative controls. The conduct of operations provides a reasonable control to ensure operators will continue to utilize the information described here as needed to safely operate the plant. Therefore the staff finds this treatment acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, on February 3, 2016, the NRC staff notified the State of Florida official (Ms. Cynthia Becker, M.P.H., Chief of the Bureau of Radiation Control, Florida Department of Health) of the proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

These amendments change inspection or surveillance requirements or requirements with respect to installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff determined that the amendments involve no significant change in the types, or significant increase in, the amounts of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. By Federal Register notice dated October 13, 2015 (80 FR 61483), the Commission previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on these findings. Accordingly, the 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

Based on the evaluation above, the NRC staff finds that St. Lucie 1 and 2 TS 3/4.9.5 and TS 3/4.9.6 do not meet the criteria in 10 CFR 50.36(c)(2)(ii) requiring inclusion in the TSs.

These changes are in alignment with the latest version of NUREG-1432 and do not diminish the level of safety found in the current TSs.

Based on the inclusion in the UFSAR, future changes to relocated TS requirements will be subject to the provisions of 10 CFR 50.59. As such, the NRC staff finds that there is reasonable assurance that future changes to the relocated requirements will be made in a manner that continues to protect public health and safety. Based on the above findings, the NRC staff concludes that relocation of St. Lucie 1 and 2 TS 3/4.9.5 and TS 3/4.9.6 to the St. Lucie UFSARs is acceptable.

Principal Contributors: Gordon Curran Brian Green Perry Buckberg Date: March 7, 2016

M. Nazar The NRC staff's safety evaluation of the amendments is enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, IRA/

Perry H. Buckberg, Senior Project Manager Plant Licensing Branch 11-2 Division of Operator Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-335 and 50-389

Enclosures:

1. Amendment No. 230 to DPR-67

2. Amendment No. 180 to NPF-16

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

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