Text

Issuance of Amendments to Revise Limiting Condition for Operation 3.0.4 and Surveillance Requirement 4.0.4
	ML14343A918
Person / Time
Site:	Saint Lucie
Issue date:	02/27/2015
From:	Farideh Saba; Plant Licensing Branch 1
To:	Nazar M; NextEra Energy
	Tam P
References
	TAC MF3618, TAC MF3619
	Download: ML14343A918 (78)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 27, 2015 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 TO REVISE LIMITING CONDITION FOR OPERATION 3.0.4 AND SURVEILLANCE REQUIREMENT 4.0.4 (TAC NOS. MF3618 AND MF3619)

Dear Mr. Nazar:

The U.S. Nuclear Regulatory Commission (NRC or the Commission) has issued the enclosed Amendment Nos. 220 and 170 to Renewed Facility Operating License Nos. DPR-67 and NPF-16 for the St. Lucie Plant, Unit Nos. 1 and 2 (SL-1 and 2), respectively. These amendments consist of changes to the Technical Specifications (TSs) in response to the Florida Power & Light Company (the licensee, FPL) application dated February 26, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14077A265), as supplemented by letters dated May 29, 2014 (ADAMS Accession No. ML14161A145), and July 25, 2014 (ADAMS Accession No. ML14210A006).

These amendments revise the SL-1 and 2, TSs, modifying requirements for mode change limitations in Limiting Condition for Operation 3.0.4 and Surveillance Requirement (SR) 4.0.4 to adopt the provisions of Technical Specification Task Force (TSTF)-359, "Increase Flexibility in MODE Restraints." On April 4, 2003, the Commission announced in the Federal Register (68 FR 16579) the availability of a model application concerning TSTF-359, Revision 8, using the "Consolidated Line Item Improvement Process (CLllP)." By letter dated April 28, 2003, the Nuclear Energy Institute (NEI) submitted to the NRC TSTF-359, Revision 9, "Increase Flexibility in Mode Restraints." According to NEI, TSTF-359, Revision 9, incorporates the NRC's responses to comments in the April 4, 2003, Federal Register Notice of Availability, and clarifies some minor differences between the model safety evaluation and TSTF-359, Revision 8. SL-1 and 2 is proposing to adopt Revision 9 of TSTF-359. Revision 9 of TSTF-359 has been incorporated in the improved vendor-specific Standard Technical Specifications in the following NRC reports: NUREG-1430, "Standard Technical Specifications, Babcock and Wilcox Plants";

NUREG-1431, "Standard Technical Specifications, Westinghouse Plants"; NUREG-1432, "Standard Technical Specifications, Combustion Engineering Plants"; NUREG-1433, "Standard Technical Specifications, General Electric Plants, BWR/4"; and NUREG-1434, "Standard Technical Specifications, General Electric Plants, BWR/6." Therefore, the NRC staff finds the licensee's adoption of Revision 9 of TSTF-359 in SL-1 and 2, TSs is consistent with the staff's current position and acceptable.

M. Nazar The amendments also revise SL-1 and 2, SR 4.0.1 to be consistent with the model SR in NUREG-1432, "Standard Technical Specifications, Combustion Engineering Plants."

The NRG staff's related safety evaluation of the amendments is enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, Farideh E. Saba, 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. 220 to Renewed Facility Operating License No. DPR-67

2. Amendment No. 170 to Renewed Facility Operating License No. 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 OPERATING LICENSE Amendment No. 220 Renewed License No. DPR-67

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

A. The application for amendment by Florida Power & Light Company (FPL, the licensee), dated February 26, 2014, as supplemented by letters dated May 29 and July 25, 2014, 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.

Enclosure 1

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. 220, 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.

FOR THE NUCLEAR REGULATORY COMMISSION Shana R. Helton, 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: February 27, 2015

ATTACHMENT TO LICENSE AMENDMENT NO. 220 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.

Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by captioned amendment number and contain marginal lines indicating the area of change.

REMOVE INSERT 3/4 0-1 3/4 0-1 3/4 0-2 3/4 0-2 3/4 3-2 3/4 3-2 3/4 3-3 3/4 3-3 3/4 3-4 3/4 3-4 3/4 3-10 3/4 3-10 3/4 3-11 3/4 3-11 3/4 3-12 3/4 3-12 3/4 3-21 3/4 3-21 3/4 3-41 3/4 3-41 3/4 4-12 3/4 4-12 3/4 4-17 3/4 4-17 3/4 4-59 3/4 4-59 3/4 5-7 3/4 5-7 3/4 6-10 3/4 6-10 3/4 7-1 3/4 7-1 3/4 7-4 3/4 7-4 3/4 7-9 3/4 7-9 3/4 7-18 3/4 7-18 3/4 7-27 3/4 7-27 3/4 8-3 3/4 8-3 3/4 9-12 3/4 9-12 3/4 11-14 3/4 11-14 3/411-15 3/411-15

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).

8. Technical Specifications The Technical Specifications contained in Appendices A and 8, as revised through Amendment No. 220 are hereby incorporated in the renewed license.

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

Appendix 8, 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 8 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 Uncovery 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. 220

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation (LCO) contained in the succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.

3.0.2 Noncompliance with a specification shall exist when the requirements of the Limiting Condition for Operation (LCO) and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.

3.0.3 When a Limiting Condition for Operation (LCO) is not met, except as provided in the associated ACTION requirements, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action shall be initiated to place the unit in a MODE in which specification does not apply by placing it, as applicable in:

1. At least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

2. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

3. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the LCO. Exceptions to these requirements are stated in the individual specifications.

This specification is not applicable in MODES 5 or 6.

3.0.4 When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made:

a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time;

b. After performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate (exceptions to this Specification are stated in the individual Specifications); or

c. When an allowance is stated in the individual value, parameter, or other Specification.

This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

ST. LUCIE - UNIT 1 3/4 0-1 Amendment No. 4G, W, 4-00 220

APPLICABILITY SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be applicable during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement. Failure to meet a Surveillance Requirement, whether such failure is experienced during the performance of the Surveillance Requirement or between performances of the Surveillance Requirement, shall be failure to meet the LCO. Failure to perform a Surveillance Requirement within the allowed surveillance interval shall be failure to meet the LCO except as provided in SR 4.0.3. Surveillance Requirements do not have to be performed on inoperable equipment or variables outside specified limits.

4.0.2 Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25% of the specified surveillance interval.

4.0.3 If it is discovered that a Surveillance was not performed within its specified ftequency, then compliance with the requirement to declare the Limiting Condition for Operation not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and the risk impact shall be managed.

If the Surveillance is not performed within the delay period, the Limiting Condition for Operation must immediately be declared not met, and the applicable ACTION(s) must be taken.

When the Surveillance is performed within the delay period and the Surveillance is not met, the Limiting Condition for Operation must immediately be declared not met, and the applicable ACTION(s) must be taken.

4.0.4 Entry into a MODE or other specified condition in the Applicability of a Limiting Condition for Operation (LCO) shall only be made when the LCO's Surveillances have been met within their specified frequency, except as provided by Surveillance Requirement 4.0.3. When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4.

This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

4.0.5 Surveillance Requirements for inservice inspection of ASME Code Class 1, 2 and 3 components shall be applicable as follows:

a. lnservice inspection of ASME Code Class 1, 2 and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(g) (6) (i).

b. deleted ST. LUCIE - UNIT 1 3/4 0-2 Amendment No. ~. 4G, W, 98, 4-08, ~. 4-86 220

TABLE 3.3-1 REACTOR PROTECTIVE INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

1. Manual Reactor Trip 2 1 2 1, 2 and* 1

2. Power Level - High 4 2(a) 3(f} 1, 2 2

3. Reactor Coolant Flow - Low 4/SG 2(a)/SG 3/SG 1, 2 (e) 2

4. Pressurizer Pressure - High 4 2 3 1, 2 2

5. Containment Pressure - High 4 2 3 1, 2 2

6. Steam Generator Pressure - Low 4/SG 2(b)/SG 3/SG 1, 2 2

7. Steam Generator Water 4/SG 2/SG 3/SG 1, 2 2 Level-Low

8. Local Power Density - High 4 2(c) 3 1 2

9. Thermal Margin/Low Pressure 4 2(a) 3 1, 2 (e) 2 9a. Steam Generator Pressure 4 2(a) 3 1, 2 (e) 2 Difference - High 1O. Loss of Turbine - Hydraulic 4 2(c) 3 1 2 Fluid Pressure - Low ST. LUCIE - UNIT 1 3/4 3-2 Amendment No. -la,~ 220

TABLE 3.3-1 {Continued}

REACTOR PROTECTIVE INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

11. Wide Range Logarithmic Neutron Flux Monitor

a. Startup and Operating -- 4 2(d) 3 1, 2 and* 2 Rate of Change of Power -

High

b. Shutdown 4 0 2 3,4,5 3

12. Reactor Protection System 4 2 4 1, 2* 4 Logic

13. Reactor Trip Breakers 4 2 4 1, 2* 4 ST. LUCIE - UNIT 1 3/4 3-3 Amendment No. 4-6, U 220

TABLE 3.3-1 (Continued)

TABLE NOTATION

With the protective system trip breakers in the closed position and the CEA drive system capable of CEA withdrawal.

(a) Trip may be bypassed below 1% of RATED THERMAL POWER; bypass shall be automatically removed when Wide Range Logarithmic Neutron Flux power is~ 1% of RATED THERMAL POWER.

(b) Trip may be manually bypassed below 685 psig; bypass shall be automatically removed at or above 685 psig.

(c) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed when Power Range Neutron Flux power is ~ 15% of RA TED THERMAL POWER.

4 (d) Trip may be bypassed below 10- % and above 15% of RATED THERMAL POWER; bypass shall be automatically removed when Wide Range Logarithmic Neutron Flux power is~ 1o-4 %

and Power Range Neutron Flux power,::: 15% of RA TED THERMAL POWER.

(e) Deleted.

(f) There shall be at least two decades of overlap between the Wide Range Logarithmic Neutron Flux Monitoring Channels and the Power Range Neutron Flux Monitoring Channels.

ACTION STATEMENTS ACTION 1 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the protective system trip breakers.

ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in either the bypassed or tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . For the purposes of testing and maintenance, the inoperable channel may be bypassed for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from time of initial loss of OPERABILITY; however, the inoperable channel shall then be either restored to OPERABLE status or placed in the tripped condition.

ST. LUCIE - UNIT 1 3/4 3-4 Amendment No. 40, ~. 4-a, 400, 409 220

TABLE 3.3-3 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TOTAL NO. MINIMUM OF CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT CHANNELS TO TRIP OPERABLE MODES ACTION

1. SAFETY INJECTION (SIAS)

a. Manual (Trip Buttons) 2 1 2 1,2,3,4 8

b. Containment Pressure -

High 4 2 3 1,2,3 9 C. Pressurizer Pressure -

Low 4 2 3 1, 2, 3(a) 9

2. CONTAINMENT SPRAY (CSAS)

a. Manual (Trip Buttons) 2 1 2 1,2,3,4 8

b. Containment Pressure -

High-High 4 2(b) 3 1,2,3 10a, 10b, 10c

3. CONTAINMENT ISOLATION (CIS)

a. Manual (Trip Buttons) 2 1 2 1,2,3,4 8

b. Containment Pressure -

High 4 2 3 1, 2, 3 9 C. Containment Radiation -

High 4 2 3 1,2,3,4 9

d. SIAS ----------------------------(See Functional Unit 1 above)-------------------------------

4. MAIN STEAM LINE ISOLATION (MSIS)

a. Manual (Trip Buttons) 2/steam 1/steam 2/operating 1, 2, 3, 4 8 generator generator steam generator

b. Steam Generator 4/steam 2/steam 3/steam 1, 2, 3(c) 9 Pressure- Low generator generator generator ST. LUCIE - UNIT 1 3/4 3-10 Amendment No. 4-§, J+, ~ 220

TABLE 3.3-3 {Continued}

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

5. CONTAINMENT SUMP RECIRCULATION (RAS)

a. Manual RAS (Trip Buttons) 2 1 2 1,2,3,4 8

b. Refueling Water Tank- Low 4 2 3 1,2,3 13

6. LOSS OF POWER

a. 4.16 kv Emergency Bus Under-voltage (Loss of Voltage) 2/Bus 2/Bus 1/Bus 1, 2, 3 12

b. 4.16 kv Emergency Bus Under-voltage (Degraded Voltage) 2/Bus 2/Bus 1/Bus 1, 2, 3 12

c. 480 V Emergency Bus Under-voltage (Degraded Voltage) 2/Bus 2/Bus 1/Bus 1, 2, 3 12

7. AUXILIARY FEEDWATER (AFAS)

a. Manual (Trip Buttons) 4/SG 2/SG 4/SG 1, 2, 3 11

b. Automatic Actuation Logic 4/SG 2/SG 3/SG 1,2,3 11

c. SG Level (1A/1B) - Low 4/SG 2/SG 3/SG 1,2,3 14a, 14b, 14c

8. AUXILIARY FEEDWATER ISOLATION

a. SG 1A - SG 1B Differential Pressure 4/SG 2/SG 3/SG 1,2,3 14a, 14b, 14c

b. Feedwater Header 1A - 1B Differential Pressure 4/SG 2/SG 3/SG 1,2,3 14a, 14c ST LUCIE - UNIT 1 3/4 3-11 Amendment No. 4-a, a+, 58, +-2, -W2, ~. 4-88220

TABLE 3.3-3 (Continued)

TABLE NOTATION (a) Trip function may be bypassed in this MODE when pressurizer pressure is < 1725 psia; bypass shall be automatically removed when pressurizer pressure is~ 1725 psia.

(b) An SIAS signal is first necessary to enable CSAS logic.

(c) Trip function may be bypassed in this MODE below 685 psig; bypass shall be automatically removed at or above 685 psig.

ACTION STATEMENTS ACTION 8 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 9 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in either the bypassed or tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . For the purposes of testing and maintenance, the inoperable channel may be bypassed for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from time of initial loss of OPERABILITY; however, the inoperable channel shall then be either restored to OPERABLE status or placed in the tripped condition.

b. Within one hour, all functional units receiving an input from the inoperable channel are also bypassed or tripped.

c. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while performing tests and maintenance on that channel provided the other inoperable channel is placed in the tripped condition.

ST. LUCIE - UNIT 1 3/4 3-12 Amendment No . .:ta, 4-e, 200 220

INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING LIMITING CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels shown in Table 3.3-6 shall be OPERABLE with their alarm/trip setpoints within the specified limits.

APPLICABILITY: As shown in Table 3.3-6.

ACTION:

a. With a radiation monitoring channel alarm/trip setpoint exceeding the value shown in Table 3.3-6, adjust the setpoint to within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable.

b. With one or more radiation monitoring channels inoperable, take the ACTION shown in Table 3.3-6.

c. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.1 Each radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-3.

4.3.3.2 At lease once per 18 months, each Control Room Isolation radiation monitoring instrumentation channel shall be demonstrated OPERABLE by verifying that the response time of the channel is within limits.

ST. LUCIE - UNIT 1 3/4 3-21 Amendment No. 4-0, ~. ~ 220

INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.8 The accident monitoring instrumentation channels shown in Table 3.3-11 shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a. Actions per Table 3.3-11.

SURVEILLANCE REQUIREMENTS 4.3.3.8 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-7.

ST. LUCIE - UNIT 1 3/4 3-41 Amendment No. J.7, ~ 220

REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following RCS leakage detection systems will be OPERABLE:

a. The reactor cavity sump inlet flow monitoring system; and

b. One containment atmosphere radioactivity monitor (gaseous or particulate).

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With the reactor cavity sump inlet flow monitoring system inoperable with an operable containment particulate radioactivity monitor, perform a RCS water inventory balance at least once per 24* hours and restore the sump inlet flow monitoring system to OPERABLE status within 30 days; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With the reactor cavity sump inlet flow monitoring system inoperable with only the containment gaseous radioactivity monitor operable, perform an RCS water inventory balance at least once per 24* hours and analyze grab samples of the containment atmosphere at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and either restore the sump inlet flow monitoring system to OPERABLE status within 7 days or restore the containment particulate radioactivity monitor to OPERABLE status within 7 days and enter action a. above with the time in this action applied against the allowed outage time of action a.; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With the required radioactivity monitor inoperable, analyze grab samples of the containment atmosphere or perform a RCS water inventory balance at least once per 24* hours, and restore the required radioactivity monitor to OPERABLE status within 30 days; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d. With all required monitors inoperable, enter LCO 3.0.3 immediately.

SURVEILLANCE REQUIREMENTS 4.4.6.1 The RCS leakage detection instruments shall be demonstrated OPERABLE by:

a. Performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION of the required containment atmosphere radioactivity monitor at the frequencies specified in Table 4.3-3.

b. Performance of the CHANNEL CALIBRATION of the reactor cavity sump inlet flow monitoring system at least once per 18 months.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

ST. LUCIE - UNIT 1 3/4 4-12 Amendment No. 444, ~. 24-2 220

REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the primary coolant shall be limited to:

a. 5 1.0 µCi/gram DOSE EQUIVALENT 1-131, and

b. 5 518.9 µCi/gram DOSE EQUIVALENT XE-133.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With the specific activity of the primary coolant >1.0 µCi/gram DOSE EQUIVALENT 1-131, verify DOSE EQUIVALENT 1-131 is 5 60.0 µCi/gram once per four hours.

b. With the specific activity of the primary coolant > 1.0 µCi/gram DOSE EQUIVALENT 1-131, but 5 60.0 µCi/gram DOSE EQUIVALENT 1-131, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while efforts are made to restore DOSE EQUIVALENT 1-131 to within the 1.0 µCi/gram limit. LCO 3.0.4.c is applicable.

c. With the specific activity of the primary coolant> 1.0 µCi/gram DOSE EQUIVALENT 1-131 for greater than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval, or> 60.0 µCi/gram DOSE EQUIVALENT 1-131, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d. With the specific activity of the primary coolant> 518.9 µCi/gram DOSE EQUIVALENT XE-133, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while efforts are made to restore DOSE EQUIVALENT XE-133 to within the 518.9 µCi/gram DOSE EQUIVALENT XE-133 limit. LCO 3.0.4.c is applicable.

e. With the specific activity of the primary coolant> 518.9 µCi/gram DOSE EQUIVALENT XE-133 for greater than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.

ST. LUCIE - UNIT 1 3/4 4-17 Amendment No. 09, ~. ~ 220

REACTOR COOLANT SYSTEM POWER OPERATED RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.13 Two power operated relief valves (PORVs) shall be OPERABLE, with their setpoints selected to the low temperature mode of operation as follows:

a. A setpoint of less than or equal to 350 psia shall be selected during heatup, cooldown and isothermal conditions when the temperature of any RCS cold leg is less than or equal to 200°F.

b. A setpoint of less than or equal to 530 psia shall be selected during heatup, cooldown and isothermal conditions when the temperature of any RCS cold leg is greater than 200°F and less than or equal to 300°F.

APPLICABILITY: MODE 4 when the temperature of any RCS cold leg is less than or equal to 300°F, MODE 5, and MODE 6 when the head is on the reactor vessel; and the RCS is not vented through greater than a 1.75 square inch vent.

ACTION:

a. With one PORV inoperable in MODE 4, restore the inoperable PORV to OPERABLE status within 7 days; or depressurize and vent the RCS through greater than a 1.75 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b. With one PORV inoperable in MODES 5 or 6, either (1) restore the inoperable PORV to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or (2) complete depressurization and venting of the RCS through greater than a 1.75 square inch vent within a total of 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months .

c. With both PORVs inoperable, restore at least one PORV to operable status or complete depressurization and venting of the RCS through greater than a 1. 75 square inch vent within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

d. With the RCS vented per ACTIONS a, b, or c, verify the vent pathway at least once per 31 days when the pathway is provided by a valve(s) that is locked, sealed, or otherwise secured in the open position; otherwise, verify the vent pathway every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

e. In the event either the PORVs or the RCS vent(s) are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs or RCS vent(s) on the transient, and any corrective action necessary to prevent recurrence.

f. LCO 3.0.4.b is not applicable to PORVs when entering MODE 4.

SURVEILLANCE REQUIREMENTS 4.4.13 Each PORV shall be demonstrated OPERABLE by:

a. Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and

b. Performance of a CHANNEL FUNCTION TEST, but excluding valve operation, at least once per 31 days; and

c. Performance of a CHANNEL CALIBRATION at least once per 18 months.

ST. LUCIE - UNIT 1 3/4 4-59 Amendment No. 00, 3-i, .w4, ~.

~220

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a. In MODES 3* and 4#, one ECCS subsystem composed of one OPERABLE high pressure safety injection pump and one OPERABLE flow path capable of taking suction from the refueling water storage tank on a safety injection actuation signal and automatically transferring suction to the containment sump on a sump recirculation actuation signal.

b. Prior to decreasing the reactor coolant system temperature below 270°F a maximum of only one high pressure safety injection pump shall be OPERABLE with its associated header stop valve open.

c. Prior to decreasing the reactor coolant system temperature below 236°F all high pressure safety injection pumps shall be disabled and their associated header stop valves closed except as allowed by Specifications 3.1.2.1 and 3.1.2.3.

APPLICABILITY: MODES 3* and 4.

MODES 5 and 6 when the Pressurizer manway cover is in place and the reactor vessel head is on.

ACTION:

a. With no ECCS subsystems OPERABLE in MODES 3* and 4#, immediately restore one ECCS subsystem to OPERABLE status or be in COLD SHUTDOWN within 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b. With RCS temperature below 270°F and with more than the allowed high pressure safety injection pump OPERABLE or injection valves and header isolation valves open, immediately disable the high pressure safety injection pump(s) or close the header isolation valves.

c. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

d. LCO 3.0.4.b is not applicable to ECCS High Pressure Safety Injection subsystem when entering MODE 4.

SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2.

4.5.3.2 The high pressure safety injection pumps shall be verified inoperable and the associated header stop valves closed prior to decreasing below the above specified Reactor Coolant System temperature and once per month when the Reactor Coolant System is at refueling temperatures.

With pressurizer pressure < 1750 psia.

REACTOR COOLANT SYSTEM cold leg temperature above 250°F.

ST. LUCIE - UNIT 1 3/4 5-7 Amendment No. ~. W, 84-,

4-04, ~. 4-7+ 220

CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERABLE with:

a. Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and

b. An overall air lock leakage rate in accordance with the Containment Leakage Rate Testing Program.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With one containment air lock door inoperable*:

1. Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the OPERABLE air lock door closed.

2. Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be closed at least once per 31 days.

3. Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With the containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

If the inner air lock door is inoperable, passage through the OPERABLE outer air lock door is permitted to effect repairs to the inoperable inner air lock door. No more than one airlock door shall be open at any time.

ST. LUCIE - UNIT 1 3/4 6-10 Amendment No. ~. 449 220

3/4.7 PLANT SYSTEMS 3.4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line code safety valves shall be OPERABLE with lift settings as specified in Table 4.7-1.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a. With both reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves inoperable, operation in MODES 1, 2 and 3 may proceed provided that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either the inoperable valve is restored to OPERABLE status or the Power Level-High trip setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.1.1 Verify each main steam line code safety valve is OPERABLE in accordance with the lnservice Testing Program. Following testing, as-left lift settings shall be within+/- 1% of 1000 psia for valves 8201 through 8208, and within +/- 1% of 1040 psia for valves 8209 through 8216 specified in Table 4.7-1.

ST. LUCIE - UNIT 1 3/4 7-1 Amendment No. W, ~. 400 220

PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:

a. Two motor driven feedwater pumps, and

b. One feedwater pump capable of being powered from an OPERABLE steam supply system.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a. With one auxiliary feedwater pump inoperable, restore at least three auxiliary feedwater pumps (two motor driven pumps and one capable of being powered by an OPERABLE steam supply system) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. LCO 3.0.4.b is not applicable.

SURVEILLANCE REQUIREMENTS 4.7.1.2 Each auxiliary feedwater pump shall be demonstrated OPERABLE:

a. At least once per 31 days by:

ST. LUCIE - UNIT 1 3/4 7-4 Amendment No. ~. 00, 499 220

PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main steam line isolation valve shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

MODE 1 With one main steam line isolation valve inoperable, POWER OPERATION may continue provided the inoperable valve is either restored to OPERABLE status or closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; otherwise, be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

MODES 2 - With one or both main steam isolation valve(s) inoperable, subsequent operation in and 3 MODES 2 or 3 may proceed provided the isolation valve(s) is (are) maintained closed. Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.7.1.5 Each main steam line isolation valve that is open shall be demonstrated OPERABLE by verifying full closure within 6.0 seconds when tested pursuant to the lnservice Testing Program.

ST. LUCIE - UNIT 1 3/4 7-9 Amendment No. W, 4-tt, ~ 220

PLANT SYSTEMS 3/4.7.5 ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION 3.7.5.1 The ultimate heat sink shall be OPERABLE with:

a. Cooling water from the Atlantic Ocean providing a water level above

-10.5 feet elevation, Mean Low Water, at the plant intake structure, and

b. Two OPERABLE valves in the barrier dam between Big Mud Creek and the intake structure.

APPLICABILITY: At all times.

ACTION:

a. With the water level requirement of the above Specification not satisfied, be in at least HOT STANDBY within six hours and provide cooling water from Big Mud Creek within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With one isolation valve in the barrier dam between Big Mud Creek and the intake structure inoperable, restore the inoperable valve to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or, within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , install a temporary flow barrier and open the barrier dam isola-tion valve. The availability of the onsite equipment capable of removing the barrier shall be verified at least once per seven days thereafter.

c. With both of the isolation valves in the barrier dam between the in-take structure and Big Mud Creek inoperable, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months either:

1) Install both temporary flow barriers and manually open both barrier dam isolation valves. The availability of the onsite equipment capable of removing the barriers shall be verified at least once per seven days thereafter, or

2) Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.5.1.1 The ultimate heat sink shall be determined OPERABLE at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the average water level to be within the limits.

4. 7 .5.1.2 The isolation valves in the barrier dam between the intake structure and Big Mud Creek shall be demonstrated OPERABLE at least once per six months by cycling each valve through at least one complete cycle of full travel.

ST. LUCIE~ UNIT 1 3/47-18 Amendment No. e4 22.0

PLANT SYSTEMS 3/4.7.9 SEALED SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION 3.7.9.1 Each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material shall be free of~ 0.005 microcu-ries of removable contamination.

APPLICABILITY: At all times.

ACTION:

a. Each sealed source with removable contamination in excess of the above limit shall be immediately withdrawn from use and:

1. Either decontaminated and repaired, or

2. Disposed of in accordance with Commission Regulations.

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.9.1.1 Test Requirements - Each sealed source shall be tested for leakage and/or contamination by:

a. The licensee, or

b. Other persons specifically authorized by the Commission or an Agreement State.

The test method shall have a detection sensitivity of at least 0.005 microcuries per test sample.

4. 7.9.1.2 Test Frequencies - Each category of sealed sources shall be tested at the frequencies described below.

a. Sources in use (excluding startup sources previously subjected to core flux) - At least once per six months for all sealed sources containing radioactive material:

ST. LUCIE - UNIT 1 3/4 7-27 Amendment No. 220

ELECTRICAL POWER SYSTEMS ACTION (continued)

e. With two of the above required diesel generators inoperable, demonstrate the OPERABILITY of two offsite A.C. circuits by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore one of the inoperable diesel generators to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in the at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Following restoration of one diesel generator unit, follow ACTION Statement b. with the time requirement of that ACTION Statement based on the time of initial loss of the remaining inoperable diesel generator.

f. With one Unit 1 startup transformer (1A or 1B) inoperable and with a Unit 2 startup transformer (2A or 2B) connected to the same A or B offsite power circuit and administratively available to both units, then should Unit 2 require the use of the startup transformer administratively available to both units, Unit 1 shall demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter. Restore the inoperable startup transformer to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

g. LCO 3.0.4.b is not applicable to diesel generators.

SURVEILLANCE REQUIREMENTS 4.8.1.1.1 Each of the above required independent circuits between the offsite transmission network and the onsite Class 1E distribution system shall be:

a. Determined OPERABLE at least once per 7 days by verifying correct breaker alignments, indicated power availability; and

b. Demonstrated OPERABLE at least once per 18 months by transferring (manually and automatically) unit power supply from the auxiliary transformer to the startup transformer.

4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by:

1. Verifying fuel level in the engine-mounted fuel tank,

2. Verifying the fuel level in the fuel storage tank,

3. Verifying the fuel transfer pump can be started and transfers fuel from the storage system to the engine-mounted tank, ST. LUCIE - UNIT 1 3/4 8-3 Amendment No. 400, ~. ~ 220

REFUELING OPERATIONS FUEL POOL VENTILATION SYSTEM - FUEL STORAGE LIMITING CONDITION FOR OPERATION 3.9.12 At least one fuel pool ventilation system shall be OPERABLE.

APPLICABILITY: Whenever recently irradiated fuel is in the spent fuel pool.

ACTION:

a. With no fuel pool ventilation system OPERABLE, suspend all operations involving movement of recently irradiated fuel within the spent fuel pool or crane operation with loads over the recently irradiated spent fuel until at least one fuel pool ventilation system is restored to OPERABLE status.

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.12 The above required fuel pool ventilation system shall be demonstrated OPERABLE:

a. At least once per 31 days by initiating flow through the HEPA filter and charcoal adsorber train and verifying that the train operates for at least 15 minutes.

b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by:

ST. LUCIE - UNIT 1 3/4 9-12 Amendment No. 4-84 220

RADIOACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of oxygen in the waste gas decay tanks shall be limited to less than or equal to 2% by volume whenever the hydrogen concentration exceeds 4% by volume.

APPLICABILITY: At all times.

ACTION:

a. With the concentration of oxygen in the waste gas decay tank greater than 2% by volume but less than or equal to 4% by volume, reduce the oxygen concentration to the above limits within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

b. With the concentration of oxygen in the waste gas decay tank greater than 4% by volume and the hydrogen concentration greater than 2% by volume, immediately suspend all additions of waste gases to the system and immediately commence reduction of the concentration of oxygen to less than or equal to 2% by volume.

c. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.5.1 The concentration of oxygen in the waste gas decay tank shall be determined to be within the above limits by continuously* monitoring the waste gases in the on service waste gas decay tank.

4.11.2.5.2 With the oxygen concentration in the on service waste gas decay tank greater than 2% by volume as determined by Specification 4.11.2.5.1, the concentration of hydrogen in the waste gas decay tank shall be determined to be within the above limits by gas partitioner sample at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

When continuous monitoring capability is inoperable, waste gases shall be monitored in accordance with the actions specified for the Waste Gas Decay Tanks Explosive Gas Monitoring System in Chapter 13 of the Updated Final Safety Analysis Report.

ST. LUCIE - UNIT 1 3/411-14 Amendment No. W, 440, ~. 4-W 220

RADIOACTIVE EFFLUENTS GAS STORAGE TANKS LIMITING CONDITION FOR OPERATION 3.11.2.6 The quantity of radioactivity contained in each gas storage tank shall be limited to less than or equal to 165,000 curies noble gases (considered as Xe-133).

APPLICABILITY: At all times.

ACTION:

a. With the quantity of radioactive material in any gas storage tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank.

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.6 The quantity of radioactive material contained in each gas storage tank shall be determined to be within the above limit at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when radioactive materials are being added to the tank when reactor coolant system activity exceeds 518.9 µCi/gram DOSE EQUIVALENT XE-133.

ST. LUCIE - UNIT 1 3/411-15 Amendment No. W, W, 2-1-3 220

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. 170 Renewed License No. NPF-16

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

A. The application for amendment by Florida Power & Light Company (FPL, the licensee), dated February 26, 2014, as supplemented by letters dated May 29 and July 25, 2014, 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.

Enclosure 2

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 3.8 to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendices A and 8, as revised through Amendment No.170, 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.

FOR THE NUCLEAR REGULATORY COMMISSION Shana R. Helton, 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: February 27, 2015

ATTACHMENT TO LICENSE AMENDMENT NO. 170 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.

Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

REMOVE INSERT 3/4 0-1 3/4 0-1 3/4 0-2 3/4 0-2 3/4 3-2 3/4 3-2 3/4 3-3 3/4 3-3 3/4 3-12 3/4 3-12 3/4 3-13 3/4 3-13 3/4 3-14 3/4 3-14 3/4 3-15 3/4 3-15 3/4 3-24 3/4 3-24 3/4 3-38 3/4 3-38 3/4 3-41 3/4 3-41 3/4 4-10 3/4 4-10 3/4 4-18 3/4 4-18 3/4 4-25 3/4 4-25 3/4 4-36 3/4 4-36 3/4 5-7 3/4 5-7 3/4 6-9 3/4 6-9 3/4 7-1 3/4 7-1 3/4 7-4 3/4 7-4 3/4 7-9 3/4 7-9 3/4 7-15 3/4 7-15 3/4 7-28 3/4 7-28 3/4 8-3 3/4 8-3 3/4 11-14 3/4 11-14 3/411-15 3/4 11-15

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 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 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. 170 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. 170

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.

3.0.2 Noncompliance with a specification shall exist when the requirements of the Limiting Condition for Operation and/or associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.

3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , action shall be initiated to place the unit in a MODE in which specification does not apply by placing it, as applicable, in:

1. At least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

2. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

3. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications.

This specification is not applicable in MODE 5 or 6.

3.0.4 When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made:

a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time;

b. After performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate (exceptions to this Specification are stated in the individual Specifications); or

c. When an allowance is stated in the individual value, parameter, or other Specification.

This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

ST. LUCIE - UNIT 2 3/4 0-1 Amendment No. 33 corrected 170

APPLICABILITY SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be applicable during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement. Failure to meet a Surveillance Requirement, whether such failure is experienced during the performance of the Surveillance Requirement or between performances of the Surveillance Requirement, shall be failure to meet the LCO. Failure to perform a Surveillance Requirement within the allowed surveillance interval shall be failure to meet the LCO except as provided in SR 4.0.3.

Surveillance Requirements do not have to be performed on inoperable equipment or variables outside specified limits.

4.0.2 Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25% of the specified surveillance interval.

4.0.3 If it is discovered that a Surveillance was not performed within its specified frequency, then compliance with the requirement to declare the Limiting Condition for Operation not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and the risk impact shall be managed.

If the Surveillance is not performed within the delay period, the Limiting Condition for Operation must immediately be declared not met, and the applicable ACTION(s) must be taken.

When the Surveillance is performed within the delay period and the Surveillance is not met, the Limiting Condition for Operation must immediately be declared not met, and the applicable ACTION(s) must be taken.

4.0.4 Entry into a MODE or other specified condition in the Applicability of a Limiting Condition for Operation (LCO) shall only be made when the LCO's Surveillances have been met within their specified frequency, except as provided by Surveillance Requirement 4.0.3. When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4.

This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

4.0.5 Surveillance Requirements for inservice inspection of ASME Code Class 1, 2 and 3 components shall be applicable as follows:

a. lnservice inspection of ASME Code Class 1, 2 and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(g) (6) (i).

b. deleted

c. deleted ST. LUCIE - UNIT 2 3/4 0-2 Amendment No. dd, 48, 94-, 429 170

TABLE 3.3-1 REACTOR PROTECTIVE INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

1. Manual Reactor Trip 4 2 4 1, 2 1 4 2 4 3*, 4*, 5* 5

2. Variable Power Level - High 4 2(a)(d) 3 1, 2 2

3. Pressurizer Pressure - High 4 2 3 1, 2 2

4. Thermal Margin/Low Pressure 4 2(a)(d) 3 1, 2 2

5. Containment Pressure - High 4 2 3 1, 2 2

6. Steam Generator Pressure - Low 4/SG 2/SG(b) 3/SG 1, 2 2

7. Steam Generator Pressure 4 2(a)(d) 3 1, 2 2 Difference - High

8. Steam Generator Level - Low 4/SG 2/SG 3/SG 1, 2 2

9. Local Power Density - High 4 2(c)(d) 3 1 2

10. Loss of Component Cooling Water 4 2 3 1, 2 2 to Reactor Coolant Pumps

11. Reactor Protection System Logic 4 2 3 1, 2 2 3*, 4*, 5* 5

12. Reactor Trip Breakers 4 2(f) 4 1, 2 4 3*, 4*, 5* 5

13. Wide Range Logarithmic Neutron Flux Monitor

a. Startup and Operating -

Rate of Change of Power - 4 2(e)(g) 3 1, 2 2 High

b. Shutdown 4 0 2 3,4,5 3

14. Reactor Coolant Flow - Low 4/SG 2/SG(a)(d) 3/SG 1, 2 2

15. Loss of Load (Turbine 4 2(c) 3 1 2 Hydraulic Fluid Pressure - Low)

ST. LUCIE - UNIT 2 3/4 3-2 Amendment No. 00 170

TABLE 3.3-1 (Continued)

TABLE NOTATION

With the protective system trip breakers in the closed position, the CEA drive system capable of CEA withdrawal, and fuel in the reactor vessel.

(a) Trip may be manually bypassed below 0.5% of RATED THERMAL POWER in conjunction with (d) below; bypass shall be automatically removed when Wide Range Logarithmic Neutron Flux power is greater than or equal to 0.5% of RATED THERMAL POWER.

(b) Trip may be manually bypassed below 705 psig; bypass shall be automatically removed at or above 705 psig.

(c) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed when Power Range Neutron Flux power is greater than or equal to 15% of RATED THERMAL POWER.

(d) Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3.

4 (e) Trip may be bypassed below 10- 3 and above 15% of RATED THERMAL POWER; bypass shall be automatically removed when Wide Range Logarithmic Neutron Flux power is ~ 10- 3 4

and Power Range Neutron Flux power~ 15% of RATED THERMAL POWER.

(f) Each channel shall be comprised of two trip breakers; actual trip logic shall be one-out-of-two taken twice.

(g) There shall be at least two decades of overlap between the Wide Range Logarithmic Neutron Flux Monitoring Channels and the Power Range Neutron Flux Monitoring Channels.

ACTION STATEMENTS ACTION 1 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the protective system trip breakers.

ST. LUCIE - UNIT 2 3/4 3-3 Amendment No. 98170

TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

1. SAFETY INJECTION (SIAS)

a. Manual (Trip Buttons) 2 1 2 1,2, 3,4 12

b. Containment Pressure - 4 2 3 1, 2, 3 13, 14 High

c. Pressurizer Pressure - 4 2 3 1, 2, 3(a) 13, 14 Low

d. Automatic Actuation - 2 1 2 1,2,3,4 12 Logic

2. CONTAINMENT SPRAY (CSAS)

a. Manual (Trip Buttons) 2 1 2 1,2, 3,4 12

b. Containment Pressure - 4 2 3 1(b), 2(b), 3(b) 18a, 18b, 18c High-High

c. Automatic Actuation 2 1 2 1,2, 3,4 12 Logic

3. CONTAINMENT ISOLATION (CIAS)

a. Manual CIAS (Trip 2 1 2 1,2,3,4 12 Buttons)

b. Safety Injection (SIAS) See Functional Unit 1 for all Safety Injection Initiating Functions and Requirements

c. Containment Pressure - 4 2 3 1, 2, 3 13, 14 High

d. Containment Radiation - 4 2 3 1, 2, 3 13, 14 High

e. Automatic Actuation 2 1 2 1,2, 3,4 12 Logic ST. LUCIE - UNIT 2 3/4 3-12 Amendment No. ~ 170

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

4. MAIN STEAM LINE ISOLATION (MSIS)

a. Manual (Trip Buttons) 2 1 2 1,2,3 16

b. Steam Generator 4/steam 2/steam 3/steam 1, 2, 3(c) 13, 14 Pressure - Low generator generator generator

c. Containment Pressure -

4 2 3 1,2,3 13, 14 High

d. Automatic Actuation Logic 2 1 2 1,2,3 12

5. CONTAINMENT SUMP RECIRCULATION (RAS)

a. Manual RAS (Trip Buttons) 2 1 2 1,2,3,4 12

b. Refueling Water Storage 4 2 3 1,2,3 19 Tank- Low

c. Automatic Actuation Logic 2 1 2 1, 2, 3 12 ST. LUCIE - UNIT 2 3/4 3-13 Amendment No. eG, ~ 170

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

6. LOSS OF POWER (LOV)

a. (1) 4.16 kV Emergency Bus Undervoltage (Loss of Voltage) 2/Bus 2/Bus 1/Bus 1,2,3 17 (2) 480 V Emergency Bus 3/Bus 2/Bus 2/Bus 1, 2, 3 17 Undervoltage (Loss of Voltage)

b. (1) 4.16 kV Emergency Bus 3/Bus 2/Bus 2/Bus 1, 2, 3 17 Undervoltage (Degraded Voltage)

(2) 480 V Emergency Bus 3/Bus 2/Bus 2/Bus 1,2, 3 17 Undervoltage (Degraded Voltage)

7. AUXILIARY FEEDWATER (AFAS)

a. Manual (Trip Buttons) 4/SG 2/SG 4/SG 1, 2, 3 15

b. Automatic Actuation Logic 4/SG 2/SG 3/SG 1,2,3 15

c. SG Level (2A/2B) - Low 4/SG 2/SG 3/SG 1,2,3 20a,20b,20c

8. AUXILIARY FEEDWATER ISOLATION

a. SG 2A - SG 2B Differential Pressure 4/SG 2/SG 3/SG 1, 2, 3 20a,20b,20c

b. Feedwater Header 2A - 2B 4/SG 2/SG 3/SG 1,2, 3 20a,20c Differential Pressure ST. LUCIE - UNIT 2 3/4 3-14 Amendment No. 28, +9, ~ 170

TABLE 3.3-3 (Continued)

TABLE NOTATION (a) Trip function may be bypassed in this MODE when pressurizer pressure is less than 1836 psia; bypass shall be automatically removed when pressurizer pressure is greater than or equal to 1836 psia.

(b) An SIAS signal is first necessary to enable CSAS logic.

(c) Trip function may be bypassed in this MODE below 700 psia; bypass shall be automatically removed at or above 700 psia.

ACTION OF STATEMENTS ACTION 12 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 13 - With the number of channels OPERABLE one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may continue provided the inoperable channel is placed in the bypassed or tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The channel shall be returned to OPERABLE status no later than during the next COLD SHUTDOWN.

With a channel process measurement circuit that affects multiple functional units inoperable or in test, bypass or trip all associated functional units as listed below.

Process Measurement Circuit Functional Unit Bypassed

1. Containment Pressure - Containment Pressure - High (SIAS, CIAS, CSAS)

Containment Pressure - High (RPS)

2. Steam Generator Pressure - Steam Generator Pressure - Low (MSIS)

AFAS-1 and AFAS-2 (AFAS)

Thermal Margin/Low Pressure (RPS)

Steam Generator Pressure - Low (RPS)

3. Steam Generator Level - Steam Generator Level - Low (RPS)

If SG-2A, then AFAS-1 (AFAS)

If SG-2B, then AFAS-2 (AFAS)

4. Pressurizer Pressure - Pressurizer Pressure - High (RPS)

Pressurizer Pressure - Low (SIAS)

Thermal Margin/Low Pressure (RPS)

ST. LUCIE - UNIT 2 3/4 3-15 Amendment No. ~. ~. 449 170

INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels shown in Table 3.3-6 shall be OPERABLE with their alarm/trip setpoints within the specified limits.

APPLICABILITY: As shown in Table 3.3-6.

ACTION:

a. With a radiation monitoring channel alarm/trip setpoint exceeding the value shown in Table 3.3-6, adjust the setpoint to within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable.

b. With one or more radiation monitoring channels inoperable, take the ACTION shown in Table 3.3-6.

c. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.1 Each radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-3.

4.3.3.2 At least once per 18 months, each Control Room Isolation radiation monitoring instrumentation channel shall be demonstrated OPERABLE by verifying that the response time of the channel is within limits.

ST. LUCIE - UNIT 2 3/4 3-24 Amendment No. ~. 4-09 170

INSTRUMENTATION REMOTE SHUTDOWN SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.5 The remote shutdown system transfer switches, control and instrumentation channels shown in Table 3.3-9 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With the number of OPERABLE remote shutdown channels less than the Required Number of Channels shown in Table 3.3-9, either restore the inoperable channel to OPERABLE status within 30 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With the number of OPERABLE remote shutdown channels less than the Minimum Channels OPERABLE requirements of Table 3.3-9, either restore the inoperable channel to OPERABLE status within 7 days, or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.3.3.5.1 Each remote shutdown monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-6.

4.3.3.5.2 Each remote shutdown system instrumentation transfer switch and control circuit shall be demonstrated OPERABLE by verifying its capability to perform its intended function(s) at least once per 18 months.

ST. LUCIE - UNIT 2 3/4 3-38 Amendment No.170

INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a.* With the number of OPERABLE accident monitoring channels less than the Required Number of Channels shown in Table 3.3-10, either restore the inoperable channel to OPERABLE status within 7 days, or be in HOT STANDBY in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and HOT SHUTDOWN in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.* With the number of OPERABLE accident monitoring channels less than the Minimum Channels OPERABLE requirements of Table 3.3-10, either restore the inoperable channels to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and HOT SHUTDOWN in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c.** With the number of OPERABLE Channels one less than the Total Number of Channels shown in Table 3.3-10, either restore the inoperable channel to OPERABLE status within 7 days if repairs are feasible without shutting down or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

d.** With the number of OPERABLE Channels less than the Minimum Channels OPERABLE requirements of Table 3.3-10, either restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months if repairs are feasible without shutting down or:

1. Initiate an alternate method of monitoring the reactor vessel inventory; and

2. Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event out-lining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status, and

3. Restore the Channel to OPERABLE status at the next scheduled refueling.

Action statements do not apply to Reactor Vessel Level Monitoring System, Containment Sump Water Level (narrow range) and Containment Sump Water Level (wide range) instruments.

- Action statements apply only to Reactor Vessel Level Monitoring System, Containment Sump Water Level (narrow range) and Containment Sump Water Level (wide range) instruments.

ST. LUCIE - UNIT 2 3/4 3-41 Amendment No. 49, 49, 4-W 170

REACTOR COOLANT SYSTEM 3/4.4.4 PORV BLOCK VALVES LIMITING CONDITION FOR OPERATION 3.4.4 Each Power Operated Relief Valve (PORV) Block valve shall be OPERABLE.

No more than one block valve shall be open at any one time.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a. With one or more block valve(s) inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the block valve(s) to OPERABLE status or close the block valve(s) and remove power from the block valve(s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With both block valves open, close one block valve within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.4 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed with power removed in order to meet the requirements of Action a. or b. above.

ST. LUCIE - UNIT 2 3/4 4-10 Amendment No. ~. 4e 170

REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following RCS leakage detection systems will be OPERABLE:

a. The reactor cavity sump inlet flow monitoring system; and

b. One containment atmosphere radioactivity monitor (gaseous or particulate).

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With the reactor cavity sump inlet flow monitoring system inoperable with an operable containment particulate radioactivity monitor, perform a RCS water inventory balance at least once per 24* hours and restore the sump inlet flow monitoring system to OPERABLE status within 30 days; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With the reactor cavity sump inlet flow monitoring system inoperable with only the containment gaseous radioactivity monitor operable, perform an RCS water inventory balance at least once per 24* hours and analyze grab samples of the containment atmosphere at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and either restore the sump inlet flow monitoring system to OPERABLE status within 7 days or restore the containment particulate radioactivity monitor to OPERABLE status within 7 days and enter action a. above with the time in this action applied against the allowed outage time of action a.; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With the required radioactivity monitor inoperable, analyze grab samples of the containment atmosphere or perform a RCS water inventory balance at least once per 24* hours, and restore the required radioactivity monitor to OPERABLE status within 30 days; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d. With all required monitors inoperable, enter LCO 3.0.3 immediately.

SURVEILLANCE REQUIREMENTS 4.4.6.1 The RCS leakage detection instruments shall be demonstrated OPERABLE by:

a. Performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION of the required containment atmosphere radioactivity monitor at the frequencies specified in Table 4.3-3.

b. Performance of the CHANNEL CALIBRATION of the reactor cavity sump inlet flow monitoring system at least once per 18 months.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

ST. LUCIE - UNIT 2 3/4 4-18 Amendment No. 84, 44+, ~ 170

REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the primary coolant shall be limited to:

a. Less than or equal to 1.0 microcurie/gram DOSE EQUIVALENT 1-131, and

b. Less than or equal to 518.9 microcuries/gram DOSE EQUIVALENT XE-133.

APPLICABILITY: MODES 1, 2, 3, and 4 ACTION:

a. With the specific activity of the primary coolant > 1.0 µCi/gram DOSE EQUIVALENT 1-131, verify DOSE EQUIVALENT 1-131 is~ 60.0 µCi/gram once per four hours.

b. With the specific activity of the primary coolant > 1.0 µCi/gram DOSE EQUIVALENT 1-131, but~ 60.0 µCi/gram DOSE EQUIVALENT 1-131, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while efforts are made to restore DOSE EQUIVALENT 1-131 to within the 1.0 µCi/gram limit. LCO 3.0.4.c is applicable.

c. With the specific activity of the primary coolant > 1.0 µCi/gram DOSE EQUIVALENT 1-131 for greater than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval, or> 60.0 µCi/gram DOSE EQUIVALENT 1-131, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d. With the specific activity of the primary coolant> 518.9 µCi/gram DOSE EQUIVALENT XE-133, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while efforts are made to restore DOSE EQUIVALENT XE-133 to within the 518.9 µCi/gram DOSE EQUIVALENT XE-133 limit. LCO 3.0.4.c is applicable.

e. With the specific activity of the primary coolant> 518.9 µCi/gram DOSE EQUIVALENT XE-133 for greater than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.

ST. LUCIE - UNIT 2 3/4 4-25 Amendment No. 44, ~ l 70

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION ACTION: (Continued)

c. In the event either the PORVs, SDCRVs or the RCS vent(s) are used to mitigate a RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs, SDCRVs or vent(s) on the transient and any corrective action necessary to prevent recurrence.

d. LCO 3.0.4.b is not applicable to PORVs when entering MODE 4.

SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:

a. In addition to the requirements of the lnservice Testing Program, operating the PORV through one complete cycle of full travel at least once per 18 months.

ST. LUCIE - UNIT 2 3/4 4-36 Amendment No. 4B, ~. 94- 170

EMERGENCY CORE COOLING SYSTEMS 3/4.5.3 ECCS SUBSYSTEMS - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a. One OPERABLE high-pressure safety injection pump, and

b. An OPERABLE flow path capable of taking suction from the refueling water tank on a Safety Injection Actuation Signal and automatically transferring suction to the containment sump on a Sump Recirculation Actuation Signal.

APPLICABILITY: MODES 3* and 4#.

Footnote# shall remain applicable in MODES 5 and 6 when the Pressurizer manway cover is in place and the reactor vessel head is on.

ACTION:

a. With no ECCS subsystems OPERABLE, restore at least one ECCS subsystem to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

c. LCO 3.0.4.b is not applicable to ECCS High Pressure Safety Injection subsystem when entering MODE 4.

SURVEILLANCE REQUIREMENTS 4.5.3 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2.

With pressurizer pressure less than 1750 psia.

One HPSI shall be rendered inoperable prior to entering MODE 5.

ST. LUCIE - UNIT 2 3/4 5-7 Amendment No. 4e, ~. +w 170

CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERABLE with:

a. Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and

b. An overall air lock leakage rate in accordance with the Containment Leakage Rate Testing Program.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With one containment air lock door inoperable*:

1. Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the OPERABLE air lock door closed.

2. Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days.

3. Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With the containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

If the inner air lock door is inoperable, passage through the OPERABLE outer air lock door is permitted to effect repairs to the inoperable inner air lock door. No more than one airlock door shall be open at any time.

ST. LUCIE - UNIT 2 3/4 6-9 Amendment No. ~. 88 170

3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line code safety valves shall be OPERABLE with lift settings as shown in Table 3.7-2.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a. With both reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves inoperable, operation in MODES 1, 2 and 3 may proceed provided that, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either the inoperable valve is restored to OPERABLE status or the Power Level-High trip setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.1.1 Verify each main steam line code safety valve is OPERABLE in accordance with the lnservice Testing Program. Following testing, as-left lift settings shall be within +/- 1% of 1000 psia for valves 8201 through 8208, and within +/- 1% of 1040 psia for valves 8209 through 8216 specified in Table 3.7-2.

ST. LUCIE - UNIT 2 3/4 7-1 Amendment No. 8, e8, 94-, 4-1-0 170

PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:

a. Two feedwater pumps, each capable of being powered from separate OPERABLE emergency busses, and

b. One feedwater pump capable of being powered from an OPERABLE steam supply system.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With two auxiliary feedwater pumps inoperable be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status.

d. LCO 3.0.4.b is not applicable.

SURVEILLANCE REQUIREMENTS 4.7.1.2 Each auxiliary feedwater pump shall be demonstrated OPERABLE:

a. At least once per 31 days by:

1. Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

ST. LUCIE - UNIT 2 3/4 7-4 Amendment No. ~. 440 170

PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main steam line isolation valve shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

MODE 1 - With one main steam line isolation valve inoperable but open, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

MODES 2, 3 - With one or both main steam isolation valve(s) inoperable, subsequent and 4 operation in MODES 2, 3 or 4 may proceed provided the isolation valve(s) is (are) maintained closed. Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.7.1.5 Each main steam line isolation valve shall be demonstrated OPERABLE by verifying full closure within 6.75 seconds when tested pursuant to the lnservice Testing Program.

ST. LUCIE - UNIT 2 3/4 7-9 Amendment No. ~. ~. 9 170

PLANT SYSTEMS 3/4.7.5 ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION 3.7.5.1 The ultimate heat sink shall be OPERABLE with:

a. Cooling water from the Atlantic Ocean providing a water level above

-10.5 feet elevation, Mean Low Water, at the plant intake structure, and

b. Two OPERABLE valves in the barrier dam between Big Mud Creek and the intake structure.

APPLICABILITY: At all times.

ACTION:

a. With the water level requirement of the above specification not satisfied, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and provide cooling water from Big Mud Creek within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With one isolation valve in the barrier dam between Big Mud Creek and the intake structure inoperable, restore the inoperable valve to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , install a temporary flow barrier and open the barrier dam isola-tion valve. The availability of the onsite equipment capable of removing the barrier shall be verified at least once per 7 days thereafter.

c. With both of the isolation valves in the barrier dam between the in-take structure and Big Mud Creek inoperable, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , either:

1. Install both temporary flow barriers and manually open both barrier dam isolation valves. The availability of the onsite equipment capable of removing the barriers shall be verified at least once per 7 days thereafter, or

2. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.5.1.1 The ultimate heat sink shall be determined OPERABLE at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the average water level to be within the limits.

4. 7.5.1.2 The isolation valves in the barrier dam between the intake structure and Big Mud Creek shall be demonstrated OPERABLE at least once per six months by cycling each valve through at least one complete cycle of full travel.

ST. LUCIE - UNIT 2 314 7-15 Amendment No. 170

PLANT SYSTEMS 3/4. 7 .10 SEALED SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION 3.7.10 Each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material shall be free of greater than or equal to 0.005 microcuries of removable contamination.

APPLICABILITY: At all times.

ACTION:

a. With a sealed source having removable contamination in excess of the above limit, immediately withdraw the sealed source from use and either:

1. Decontaminate and repair the sealed source, or

2. Dispose of the sealed source in accordance with Commission Regulations.

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.10.1 Test Requirements - Each sealed source shall be tested for leakage and/or contamination by:

a. The licensee, or

b. Other persons specifically authorized by the Commission or an Agreement State.

The test method shall be a detection sensitivity of at least 0.005 microcuries per test sample.

4.7.10.2 Test Frequencies - Each category of sealed sources (excluding startup sources and fission detectors previously subjected to core flux) shall be tested at the frequencies described below.

a. Sources in use - At least once per 6 months for all sealed sources containing radioactive material:

1. With a half-life greater than 30 days (excluding Hydrogen 3),

and

2. In any form other than gas.

ST. LUCIE - UNIT 2 3/4 7-28 AmencDent No. 170

ELECTRICAL POWER SYSTEMS ACTION: (Continued)

d. With two of the required offsite A.C. circuits inoperable, restore one of the inoperable offsite sources to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Following restoration of one offsite source, follow ACTION Statement a. with the time requirement of that ACTION Statement based on the time of the initial loss of the remaining inoperable offsite A.C. circuit.

e. With two of the above required diesel generators inoperable, demonstrate the OPERABILITY of two offsite A.C. circuits by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore one of the inoperable diesel generators to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in the at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Following restoration of one diesel generator unit, follow ACTION Statement b. with the time requirement of that ACTION Statement based on the time of initial loss of the remaining inoperable diesel generator.

f. With one Unit 2 startup transformer (2A or 2B) inoperable and with a Unit 1 startup transformer (1A or 1B) connected to the same A or B offsite power circuit and administratively available to both units, then should Unit 1 require the use of the startup transformer administratively available to both units, Unit 2 shall demonstrate the operability of the remaining A.C. sources by performing Surveillance Requirement 4.8.1.1.1 a. within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter. Restore the inoperable startup transformer to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

g. LCO 3.0.4.b is not applicable to diesel generators.

SURVEILLANCE REQUIREMENTS 4.8.1.1.1 Each of the above required independent circuits between the offsite transmission network and the onsite Class 1E distribution system shall be:

a. Determined OPERABLE at least once per 7 days by verifying correct breaker alignments, indicated power availability; and

b. Demonstrated OPERABLE at least once per 18 months by transferring (manually and automatically) unit power supply from the normal circuit to the alternate circuit.

4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS BY:

ST. LUCIE - UNIT 2 3/4 8-3 Amendment No. 39, ~. +8 170

RADIOACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of oxygen in the waste gas decay tanks shall be limited to less than or equal to 2% by volume whenever the hydrogen concentration exceeds 4% by volume.

APPLICABILITY: At all times.

ACTION:

a. With the concentration of oxygen in the waste gas decay tank greater than 2% by volume but less than or equal to 4% by volume, reduce the oxygen concentration to the above limits within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

b. With the concentration of oxygen in the waste gas decay tank greater than 4% by volume and the hydrogen concentration greater than 2% by volume, immediately suspend all additions of waste gases to the system and immediately commence reduction of the concentration of oxygen to less than or equal to 2% by volume.

c. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.5.1 The concentration of oxygen in the waste gas decay tank shall be determined to be within the above limits by continuously* monitoring the waste gases in the on service waste gas decay tank.

4.11.2.5.2 With the oxygen concentration in the on service waste gas decay tank greater than 2% by volume as determined by Specification 4.11.2.5.1, the concentration of hydrogen in the waste gas decay tank shall be determined to be within the above limits by gas partitioner sample at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

When continuous monitoring capability is inoperable, waste gases shall be monitored in accordance with the actions specified for the Waste Gas Decay Tanks Explosive Gas Monitoring System in Chapter 13 of the Updated Final Safety Analysis Report.

ST. LUCIE - UNIT 2 3/4 11-14 Amendment No. a+, 89, Q4 1 70

RADIOACTIVE EFFLUENTS GAS STORAGE TANKS LIMITING CONDITION FOR OPERATION 3.11.2.6 The quantity of radioactivity contained in each gas storage tank shall be limited to less than or equal to 165,000 curies noble gases (considered as Xe-133).

APPLICABILITY: At all times.

ACTION:

a. With the quantity of radioactive material in any gas storage tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank.

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.6 The quantity of radioactive material contained in each gas storage tank shall be determined to be within the above limit at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when radioactive materials are being added to the tank when reactor coolant system activity exceeds 518.9 µCi/gram DOSE EQUIVALENT XE-133.

ST. LUCIE - UNIT 2 3/4 11-15 Amendment No. ~. ~ 170

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 220 AND 170 TO RENEWED FACILITY OPERATING LICENSE NOS. DPR-67 AND NPF-16 FLORIDA POWER AND LIGHT COMPANY. ET AL.

ST. LUCIE PLANT. UNIT NOS. 1 AND 2 DOCKET NOS. 50-335 AND 50-389

1.0 INTRODUCTION

By application dated February 26, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14077A265), as supplemented by letters dated May 29, 2014 (ADAMS Accession No. ML14161A145), and July 25, 2014 (ADAMS Accession No. ML14210A006), Florida Power and Light (the licensee), requested changes to Technical Specifications (TSs) for St. Lucie Plant, Unit Nos. 1 and 2 (SL-1 and 2). The proposed changes would modify TSs requirements for mode change limitations in Limiting Condition for Operation (LCO) 3.0.4 and Surveillance Requirement (SR) 4.0.4 to adopt the provisions of Technical Specification Task Force (TSTF)-359, "Increase Flexibility in MODE [Operational Mode]

Restraints." On April 4, 2003, the Commission announced in the Federal Register (68 FR 16579) that the staff of the U.S. Nuclear Regulatory Commission (NRC) had prepared a model application relating to the modification of requirements regarding TSs mode change limitations, and the availability of a model application concerning TS improvement to modify requirements regarding mode change limitations using the "Consolidated Line Item Improvement Process (CLllP)." The purpose of this model was to permit the NRC to efficiently process amendments that propose to modify requirements for TS mode change limitations as generically approved by this notice.

By letter dated April 28, 2003, the Nuclear Energy Institute (NEI) submitted to the NRC TSTF-359, Revision 9, "Increase Flexibility in Mode Restraints" (ADAMS Accession No. ML031190607). According to NEI, TSTF-359, Revision 9, incorporates the NRC's responses to comments in the April 4, 2003, Federal Register Notice of Availability, and clarifies some minor differences between the model safety evaluation and TSTF-359, Revision 8. The licensee is proposing to adopt Revision 9 of TSTF-359 in SL-1 and 2 TSs. Revision 9 of TSTF-359 has been incorporated in the improved vendor-specific Standard Technical Specifications (STSs) in the following NRC reports: NUREG-1430, "Standard Technical Specifications, Babcock and Wilcox Plants"; NUREG-1431, "Standard Technical Specifications, Westinghouse Plants"; NUREG-1432, "Standard Technical Specifications, Combustion Enclosure 3

Engineering [CE] Plants"; NUREG-1433, "Standard Technical Specifications, General Electric Plants, BWR/4"; and NUREG-1434, "Standard Technical Specifications, General Electric Plants, BWR/6." Therefore, FPL adoption of Revision 9 of TSTF-359 is acceptable and consistent with the NRC staff's current position.

TSTF-359 is one of the industry's initiatives under the risk-informed TSs program. These initiatives are intended to maintain or improve safety, while reducing unnecessary burden and to make TSs requirements consistent with the NRC's other risk-informed regulatory requirements, and in particular, the requirements of Title 10, Code of Federal Regulations (10 CFR)

Section 50.65 (Maintenance Rule).

The Commission published improved vendor-specific STSs in NUREG-1430 through NUREG-1434. NUREG-1432, Revision 4, specifies that a nuclear power plant cannot go to higher modes of operation 1 (i.e., move toward power operation) unless all TSs systems normally required for the higher mode are operable. This limitation is included (with several exceptions for some plants) in LCO 3.0.4 and SR 3.0.4. LCO 3.0.4 and SR 3.0.4 in the CE STSs, Revision 4, state, in part, that when an LCO or SR is not met, "entry into a MODE or other specified condition in the applicability shall not be made except when the associated actions to be entered permit continued operation in the MODE or other specified condition in the applicability for an unlimited period of time." The industry believes that this requirement is unnecessarily restrictive and can unduly delay plant startup, while considerable resources are being used to resolve startup issues that are risk insignificant or low risk. A maintenance activity that takes longer than planned can delay a mode change and adversely impact a utility's orderly plant startup and return to power operation. The objective of the proposed change is to provide additional operational flexibility without compromising plant safety.

SL-1 and 2 have not adopted CE STSs (NUREG-1432, Revision 4), which results in some numbering differences between the SL-1 and 2 TSs and TSTF-359, Revision 9 (ADAMS Accession No. ML031190607). Specifically:

1. STS SR 3.0.1 is SR 4.0.1 in the SL-1 and 2 TSs.

2. STS SR 3.0.4 is SR 4.0.4 in the SL-1 and 2 TSs.

3. STS LCO 3.4.7 is LCO 3.4.5 in the SL-1 and 2 TSs.

4. STS LCO 3.5.3 is LCO 3. 7.4 in the SL-1 and 2 TSs.

5. STS LCO 3.8.1 is 3.8.1.1 in the SL-1 and 2 TSs.

6. SL-1 and 2 administrative controls are located under TS 6.0, not 5.0.

7. The numberings of parts of LCO 3.0.4 in the current SL-1 and 2 TSs differ from those in STSs.

1 MODE numbers decrease in transition "up to a higher mode of operation"; power operation is MODE 1.

The licensee's proposed changes to LCO 3.0.4 and SR 4.0.4 for SL-1 and 2 would allow for systems and components mode changes into a TS condition that has a specific required action and completion time (CT). The licensee will utilize the LCO 3.0.4 and SR 4.0.4 allowances only when it determines that there is a high likelihood that the LCO will be satisfied within the LCO CT, after the mode change. In addition, the LCO 3.0.4 and SR 4.0.4 allowances can be applied to values and parameters in specifications when explicitly stated in the TSs (nonsystem/component TSs such as Reactor Coolant System Specific Activity). These changes are in addition to the current mode change allowance when a required action has an indefinite CT. The LCO 3.0.4 and SR 4.0.4 mode change allowances are not permitted for the systems and components (termed "higher risk") listed in Section 3.1.1, "Identification of Risk-Important Technical Specifications Systems and Components," forthe modes specified. Two examples are (1) the licensee cannot transition from Mode 5 to Mode 4 or to Mode 3 (below 1700 per square inch absolute (psia)) with the High Pressure Safety Injection System inoperable, and (2) the licensee cannot transition up into any mode with required Diesel Generators (including other Emergency/Shutdown Alternating Current (AC) Power Supplies) inoperable.

2.0 REGULATORY EVALUATION

TSs are required by 10 CFR 50.36 to include LCOs. LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met.

TSs are also required to include SRs, which are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCO will be met.

By convention, the LCOs are contained in 3/4.1 through 3/4.11 for SL-1 and 2. TSs Section 3/4.0 on "LCO and SR Applicability" provides details or usage rules for complying with the LCOs.

LCO 3.0.4 and SR 4.0.4 address requirements for LCO compliance when transitioning between modes.

TSs have taken advantage of risk technology as experience and capability have increased.

Since the mid-1980s, the NRC has been reviewing and granting improvements to TSs that are based, at least in part, on probabilistic risk assessment (PRA) insights. In its Final Policy Statement on Technical Specifications Improvements for Nuclear Power Plants, July 22, 1993 (58 FR 39 and 132), the Commission stated that it expects that licensees will utilize any plant-specific PRA or risk survey in preparing its TS-related submittals. In evaluating these submittals, the NRC staff applies the guidance in Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," dated July 1998, and in RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," dated August 1998. The NRC staff has appropriately adapted this guidance to assess the acceptability of upward mode changes with equipment inoperable. This review had the following objectives:

To ensure that the plant risk does not increase unacceptably during the actual implementation of the proposed change (e.g., when the plant enters a higher mode while an LCO is not met). This risk increase is referred to as "temporary."

To compare and assess the risk impact of the proposed change to the acceptance guidelines of the Commission's Safety Goal Policy Statement, as documented in RG 1.174. The risk impact, which is measured by the average yearly risk increase associated with the change, aims at minimizing the "cumulative" risk associated with the proposed change so that the plant's average baseline risk is maintained within a minimal range.

To assess the licensee's ability to identify risk-significant configurations resulting from maintenance or other operational activities and take appropriate compensatory measures to avoid such configurations.

The NRC staff reviewed licensee reliance on 10 CFR 50.65(a)(4) for the nonhigher-risk systems and components and related guidance to assess and manage the risk of upward mode changes. The Commission has found that following the industry guidance for implementation of 10 CFR 50.65(a)(4), as endorsed by RG 1.182, "Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants," and mandated by LCO 3.0.4, SR 4.0.4, and SR 4.0.3, satisfies the configuration risk management objectives of RG 1.177 for TSs surveillance interval and CT extensions. The licensee's reliance on 10 CFR 50.65(a)(4) processes that are consistent with the provisions of the NRG-endorsed industry guidance was also found to be adequate for managing the risk of missed surveillances as described in the Federal Register on September 28, 2001 (66 FR 49714).

The NRC staff review also had the objective of ensuring that existing NRC inspection programs have the necessary controls in place to allow the NRC staff to oversee the implementation of the proposed change, reliance on 10 CFR 50.65(a)(4) processes or programs, and the ability to adequately assess the licensee's performance associated with risk assessments. The review encompassed inspection procedures (i.e., NRC Inspection Procedure (IP) 62709 dated December 28, 2000, "Configuration Risk Assessment and Risk Management Process," and NRC IP 71111.13 dated January 17, 2002, "Maintenance Risk Assessments and Emergent Work Control"), the significance determination process (SOP) (i.e., "Maintenance Risk Assessment and Risk Management Significance Determination Process"), enforcement guidance (i.e., Enforcement Manual Section 7 .11, "Actions Involving 10 CFR 50.65, The Maintenance Rule"), and the associated Reactor Oversight Process (ROP).

2.1 Proposed Change to St. Lucie Plant. Unit Nos. 1 and 2. LCO 3.0.4. SR 4.0.1. and SR 4.0.4 Currently, SL-1 and 2 LCO 3.0.4 does not allow entrance into a higher mode (or other specified condition) in the applicability when an LCO is not met, except when the associated actions to be entered permit continued operation in that mode or condition indefinitely, or a specific exception is granted. Similarly, when an LCO's surveillances have not been met within their specified frequency, entry into a higher mode (or other specified condition) is not allowed by SR 4.0.4.

The current SL-1 and 2 TSs LCO 3.0.4 reads:

Entry into an OPERATIONAL MODE or other specified applicability condition shall not be made when the conditions of the Limiting Condition for Operation are not met and the associated ACTION requires a shutdown if they are not met

within a specified time interval. Entry into an OPERATIONAL MODE or specified condition may be made in accordance with ACTION requirements when conformance to them permits continued operation of the facility for an unlimited period of time. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION statements.

Exceptions to these requirements are stated in the individual specifications.

The licensee's proposed revision to LCO 3.0.4 will read:

When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made:

(a) When the associated ACTIONS to be entered permit continued operation in that MODE or other specified condition in the Applicability for an unlimited period of time; (b) After performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate (exceptions to this Specification are stated in the individual Specifications); or (c) When an allowance is stated in the individual value, parameter, or other Specification.

This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

The SL-1 and 2 current TSs SR 4.0.1 reads:

Surveillance Requirements shall be applicable during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement.

Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Specification 4.0.2, shall constitute noncompliance with the OPERABILITY requirements for a Limiting Condition for Operation. Surveillance Requirements do not have to be performed on inoperable equipment.

The licensee's proposed revision to SR 4.0.1 will conform to SR 3.0.1 in NUREG-1432, Revision 4, and read:

Surveillance Requirements shall be met during the OPERATIONAL MODES or other specified conditions in the Applicability for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement.

Failure to meet a Surveillance Requirement, whether such failure is experienced during the performance of the Surveillance Requirement or between

performances of the Surveillance Requirement, shall be failure to meet the LCO.

Failure to perform a Surveillance Requirement within the allowed surveillance interval shall be failure to meet the LCO except as provided in SR 4.0.3.

Surveillance Requirements do not have to be performed on inoperable equipment or variables outside specified limits.

Adopting STS LCO 3.0.4 as Specification 3.0.4 and STS SR 4.0.4 as Specification 4.0.4 in accordance with TSTF-359, Revision 9, without adopting STS SR 3.0.1 as Specification 4.0.1, would result in non-conservative implementation of TSs due to language incongruences between STS SR 3.0.1 and Specification 4.0.1. Therefore, to establish a consistent licensing basis that would ensure the allowances of TSTF-359, Revision 9, are correctly implemented, the licensee is proposing to revise SL-1 and 2 TSs SR 4.0.1 to be equivalent to STS SR 3.0.1.

The SL-1 and 2 current TSs SR 4.0.4 reads:

Entry into an OPERATIONAL MODE or other specified applicability condition shall not be made unless the Surveillances Requirement(s) associated with the Limiting Condition for Operation have been performed within the stated surveillance interval or as otherwise specified. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements.

The licensee's proposed revision to SR 4.0.4 will conform to the proposed changes to LCO 3.0.4 and read:

Entry into a MODE or other specified condition in the Applicability of a Limiting Condition for Operation (LCO) shall only be made when the LCO's Surveillances have been met within their specified frequency, except as provided by Surveillance Requirement 4.0.3. When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4.

This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

The proposed LCO 3.0.4(a) retains the current allowance for when the required actions allow indefinite operation. The proposed LCO 3.0.4(b) allows entering modes or other specified conditions in the applicability except when higher-risk systems and components (listed in Section 3.1.1) for the mode being entered are inoperable. When applying LCO 3.0.4(b), the decision for entering a higher mode or condition in the Applicability of the LCO will be made by plant management after the required risk assessment has been performed and requisite risk management actions established, through the program established to implement 10 CFR 50.65(a)(4). Entry into the modes or other specified conditions in the Applicability of the TSs shall be for no more than the duration of the applicable required actions CT, or until the LCO is met. The licensee has proposed to remove current notes in individual specifications that prohibit mode changes that are now encompassed by LCO 3.0.4(b). Similarly, the licensee has proposed to add notes that prohibit mode changes under LCO 3.0.4(b) for higher-risk systems

and components. The proposed LCO 3.0.4(b) allowance can involve multiple components in a single LCO or in multiple LCOs. However, use of the LCO 3.0.4(b) provisions are always contingent on an analysis demonstrating acceptable level of risk and implementation of risk management actions, if appropriate, as consistent with 10 CFR 50.65(a)(4)-based risk assessment.

The notes limiting the applicability to Modes 1, 2, 3, and 4 of the current TSs LCO 3.0.4 and SR 4.0.4 are holdovers from existing TSs that are no longer needed and are removed consistent with TSTF-359, Revision 9. Consideration was originally given to adding notes to various TS, as defined by the tables of higher-risk systems, precluding entry into Modes 5 and 6 for pressurized-water reactors (PWRs). However, the addition of notes in these cases is unnecessary since the action statements require immediate CTs, which means that entry into the mode or other specified condition in the Applicability is not allowed, and the notes would be superfluous.

LCO 3.0.4 allowances related to values and parameters of TSs are not typically addressed by LCO 3.0.4(b) risk assessments and are, therefore, addressed by a new LCO 3.0.4(c).

LCO 3.0.4(c) refers to allowances already in the TSs and annotated in the individual TSs.

LCO 3.0.4(c) also allows for entry into the modes or other specified conditions in the Applicability for TSs for no more than the duration of the applicable required actions CT or until the LCO is met or the unit is not within the applicability of the TSs.

3.0 TECHNICAL EVALUATION

During the development of the current STSs, improvements were made to LCO 3.0.4, such as clarifying its applicability with respect to plant shutdowns, cold shutdown mode, and refueling mode. In addition, during the STSs development, almost all the LCOs with CTs greater than or equal to 30 days, and many LCOs with CTs greater than or equal to 7 days, were given individual LCO 3.0.4 exceptions. The specific LCO 3.0.4 exceptions allowed entry into a mode or other specified condition in the TSs applicability while relying on the TSs required actions and associated CTs. The change proposed by the licensee for SL-1 and 2 would provide standardization and consistency to the use and application of LCO 3.0.4, both internal to and between each of the specifications, as well as with the STSs. This proposed change will also ensure consistency through the utilization of appropriate levels of risk assessment of plant configurations for application of LCO 3.0.4. However, nothing in this safety evaluation should be interpreted as encouraging upward mode transition with inoperable equipment.

Good practice should dictate that such transitions should normally be initiated only when all required equipment is operable and that mode transition with inoperable equipment should be the exception rather than the rule.

The current LCO 3.0.4(a) allowances are retained in the proposal and do not represent a change in risk from the current situation. The LCO 3.0.4(b) allowances apply to systems and components, and require a risk assessment and risk management actions, if appropriate, prior to utilization to ensure that an acceptable level of safety is maintained. The LCO 3.0.4(c) allowances apply to parameters and values that have been previously approved by the NRC in a plant-specific TS. The licensee provided in its TSs Bases a discussion and list of each NRG-approved LCO 3.0.4(c) specific value and parameter allowance, but as stated in 10 CFR 50.36(a)(1 }, the bases shall not become part of the TSs.

In its review of TSTF-359, the NRC staff did a generic qualitative assessment of the risk impact of the proposed change in LCO 3.0.4(b) allowances by evaluating how licensee implementation of the proposed risk-informed approach is expected to meet the requirements of the applicable RGs. The NRC staff referred to the guidance provided in RG 1.174 and RG 1.177. RG 1.177 provides the NRC staff's recommendations on using risk information to assess the impact of proposed changes to nuclear power plant TSs on the risk associated with plant operation.

Although RG 1.177 does not specifically address the type of generic change in this proposal, the NRC staff considered the approach documented in RG 1.177 in evaluating the risk information provided to support the proposed changes in LCO 3.0.4.

Section 3.1 below discusses the NRC staff's evaluation of how the implementation of the proposed risk-informed approach used to justify LCO 3.0.4(b) allowances agrees with the objectives of the guidance outlined in RG 1.177. Section 3.2 discusses oversight of the risk-informed approach associated with the LCO 3.0.4(b) allowances.

3.1 Evaluation of Risk Management Both the temporary and cumulative risks of the licensee's proposed change are adequately limited. The temporary risk is limited by the exclusion of higher-risk systems and components and CT limits contained in TSs (Section 3.1.1 ). The cumulative risk is limited by the temporary risk limitations and by the expected low frequency of the proposed mode changes with inoperable equipment (Section 3.1.2). The NRC's inspection of the licensee's implementation of 10 CFR 50.65(a)(4), as applied to the proposed change, will provide adequate NRC oversight of the licensee's ability to use the LCO 3.0.4(b) provisions under appropriate circumstances (i.e., to identify risk-significant configurations when entering a higher mode or condition in the applicability of an LCO (Section 3.1.3)).

3.1.1 Temporary Risk Increases RG 1.177 proposes the incremental conditional core damage probability (ICCDP) and the incremental conditional large early release probability (I CLE RP) as appropriate measures of the increase in probability of core damage and large early release, respectively, during the period of implementation of a proposed TS. In addition, RG 1.177 stresses the need to preclude potentially high-risk configurations introduced by the proposed change. The ICCDP associated with any specified plant condition, such as the condition introduced by entering a higher mode with plant equipment inoperable, is expressed by the following equation:

ICCDP =~Rd = (R1 - Ro) d (1)

Where:

~R = the conditional risk increase, in terms of core damage frequency (CDF), caused by the specified condition d = the duration of the specified plant condition R1 = the plant CDF with the specified condition permanently present

Ra = the plant CDF without the specified condition The same expression can be used for ICLERP by substituting the measure of risk (i.e., large early release frequency (LERF) for CDF). The magnitude of the ICCDP and ICLERP values associated with plant conditions applicable to LCO 3.0.4(b) allowances can be managed by controlling the conditional risk increase, ~R (in terms of both CDF and LERF), and the duration, d, of such conditions. The following sections discuss how the key elements of the proposed risk-informed approach used to justify LCO 3.0.4(b) allowances are expected to limit ~Rand d and, thus, prevent any significant temporary risk increases.

Identification of Risk-Important Technical Specifications Systems and Components A major element that limits the risk of the proposed mode change flexibility is the exclusion of certain systems and associated LCO for the mode change allowance. TSs allow operation in Mode 1 (power operation) with specified levels of inoperability for specified times. This provides a benchmark of currently acceptable risk against which to measure any incremental risk inherent in the proposed LCO 3.0.4(b). If a system inoperability accrues risk at a higher rate in one or more of the transition modes than it would in Mode 1, then an upward transition into that mode should not be allowed without demonstration of a high degree of experience and sophistication in risk management. However, the risk management process evaluated in Section 3.1.3 is adequate if higher-risk systems/components are excluded from the scope of LCO 3.0.4(b).

The importance of most TSs systems in mitigating accidents increases as power increases.

However, some TSs systems are relatively more important during lower power and shutdown operations, because:

Certain events are peculiar to modes of plant operation other than power operation.

Certain events are more probable at modes of plant operation other than power operation.

Some modes of plant operation have less mitigation system capability than power operation.

The risk information submitted in support of the proposed changes to LCO 3.0.4 and SR 4.0.4 in TSTF-359 included qualitative risk assessments performed by each owners group to identify higher-risk systems and components at the various modes of operation, including transitions between modes, as the plant moves upward from the refueling mode of operation toward power operation. The CE Owners Group (GEOG) Task 1181, "Qualitative Risk Assessment for Relaxation of Mode Entry Restraints," CE Nuclear Power LLC, CE NPSD-1207 (Revision 0) contains generic qualitative risk assessments. It discusses the technical approach used and the systems/components subsequently determined to be of higher-risk significance and the systems/components not to be granted the LCO 3.0.4 allowances for the various modes.

Following interactions with the NRC staff, all owners groups used the same systematic approach in their qualitative risk assessments to identify the higher-risk systems in the STSs, consisting of the following steps:

Identification of plant conditions (i.e., plant parameters and availability of key mitigation systems) associated with changes in plant modes while returning to power

Identification of key activities that have the potential to impact risk and which are in progress during transitions between modes while the plant is returning to power

Identification of applicable accident initiating events for each mode or other specified condition in the applicability

Identification of the higher-risk systems and components by combining the information in the first three steps (qualitative risk assessment)

The risk assessments properly used the results and insights from previous deterministic and probabilistic studies to systematically search for plant conditions in which certain key plant components are more important in mitigating accidents than during operation at power (Mode 1). This search took the following factors into account for the various stages of returning the plant to power:

The status of accident mitigation and normally operating systems

The status of key plant parameters such as reactor coolant system pressure

The key activities that are in progress during transitions between modes that have the potential to impact risk (e.g., the transfer from auxiliary to main feedwater at some PWR plants when Mode 1 is entered)

The applicable accident initiating events for each mode of plant operation

Design and operational differences among plants or groups of plants The following systems and components were identified by the CEOG as higher-risk systems and components when the plant is entering a new mode:

Combustion Engineering Owners Group Plants System Entering Mode(s)

Emergency Diesel Generators (EDGs) 5,4, 3, 2, 1 Auxiliary Feedwater/Emergency Feedwater System 4, 3, 2, 1 High Pressure Safety Injection (HPSI) System 4, 3 (below 1700 psia)

Low Temperature Overpressure Protection (LTOP) 5, 4 (below set temperature)

Power Operated Relief Valves (PORVs) (when used for LTOP)

Shutdown Cooling System (Low Pressure 5 Safety Injection (LPSI) pumps)

In adopting TSTF-359, if the licensee identifies any higher-risk system(s) and the higher-risk modes of applicability, the TSs for that system are modified by a note that reads, for example, "LCO 3.0.4(b) is not applicable." Systems identified as higher risk for Modes 5 and 6 for PWRs are also excluded from transitioning up to the mode of higher risk, and as previously discussed, notes for those transitions are superfluous. In addition, mode transitions for Modes 5 and 6 for PWRs will be addressed by administrative controls.

In summary, the NRG staff's review of the owners groups' qualitative risk assessments finds that they are of adequate quality to support the application (i.e., they identify the higher-risk systems and components) associated with entering higher modes of plant operation with equipment inoperable while returning to power.

The licensee is adopting the wording in TSTF-359 for LCO 3.0.4 and SR 4.0.4. LCO 3.0.4(c) has been referenced appropriately for parameters and values in LCO 3/4.4.8, "Specific Activity,"

applicable in Modes 1, 2, 3, and 4.

The licensee has, consistent with the above GEOG table, proposed that notes to the appropriate TSs be added to SL-1 and 2 to state LCO 3.0.4(b) applicability to LCOs, which have been identified as PWR higher-risk systems.

EDGs - not applicable to EDGs in Modes 1, 2, 3, and 4 for Unit No. 1 and Unit No. 2 TS 3/4.8.1, AC Sources

Auxiliary Feedwater System - not applicable in Modes 1, 2, and 3 for Unit No. 1 and Unit No. 2 TS 3/4. 7.1.2, Auxiliary Feedwater System

HPSI - applicable to HPSI when entering Mode 4 for Unit No. 1 TS 3/4.5.3, Emergency Core Cooling System (ECCS) Subsystems -Tavg [average temperature] less than 325 degrees Fahrenheit (°F), and for Unit No. 2 TS 3/4.5.3, ECCS Subsystems -

Shutdown

LTOP/PORVs - not applicable to PORVs when entering Mode 4 for Unit No. 1 TS 3/4.4.13, Power Operated Relief Valves, and for Unit 2 TS 3/4.4.9.2, Pressurizer Heatup/Cooldown Limits

Shutdown Cooling System LPSI pumps-TS 3/4.5.3, ECCS Subsystems-Tavg less than 325 °F (Unit No. 1), and TS 3/4.5.3, ECCS Subsystems - Shutdown (Unit No. 2)

SL-1 and 2 have plant-specific LCO 3.0.4 statements stating that, "The provisions of specification 3.0.4 are not applicable," in their TSs. In accordance with TSTF-359, the LCO 3.0.4 provisions are removed from the following SL-1 and 2 TSs:

St Lucie Plant, Unit No. 1 TS SYSTEM PAGE Table 3.3-1 Reactor Protective Instrumentation 3/4 3-2 3/4 3-3 3/4 3-4 Table 3.3-3 Engineered Safety Feature Actuation 3/4 3-10 System Instrumentation 3/4 3-11 3/4 3-12 3.3.3.1 Radiation Monitoring Instrumentation 3/4 3-21 3.3.3.8 Accident Monitoring Instrumentation 3/4 3-41 3.4.6.1 Reactor Coolant System Leakage Detection 3/4 4-12 Systems 3.6.1.3 Containment Air Locks 3/4 6-10 3.7.1.1 Turbine Cycle Safety Valves 3/4 7-1 3.7.1.5 Main Steam Line Isolation Valves 3/4 7-9 3.7.5.1 Ultimate Heat Sink 3/47-18 3.7.9.1 Sealed Source Contamination 3/4 7-27 3.9.12 Fuel Pool Ventilation System - Fuel Storage 3/4 9-12 3.11.2.5 Explosive Gas Mixture 3/4 11-14 3.11.2.6 Gas Storage Tanks 3/411-15 St. Lucie Plant, Unit No. 2 TS SYSTEM PAGE Table 3.3-1 Reactor Protective Instrumentation 3/4 3-2 3/4 3-3 Table 3.3-3 Engineered Safety Feature Actuation 3/4 3-12 System Instrumentation 3/4 3-13 3/4 3-14 3/4 3-15 3.3.3.1 Radiation Monitoring Instrumentation 3/4 3-24 3.3.3.5 Remote Shutdown System Instrumentation 3/4 3-38 3.3.3.6 Accident Monitoring Instrumentation 3/4 3-41 3.4.4 PORV Block Valves 3/4 4-10 3.4.6.1 Reactor Coolant System Leakage Detection 3/4 4-18 Systems 3.6.1.3 Containment Air Locks 3/4 6-9 3.7.1.1 Turbine Cycle Safety Valves 3/4 7-1 3.7.1.5 Main Steam Line Isolation Valves 3/4 7-9 3.7.5.1 Ultimate Heat Sink 3/4 7-15 3.7.10 Sealed Source Contamination 3/4 7-28 3.11.2.5 Explosive Gas Mixture 3/4 11-14 3.11.2.6 Gas Storage Tanks 3/411-15

Limited Time in Technical Specifications Required Actions Any temporary risk increase will be limited by, among other factors, duration constraints imposed by the TSs CTs of the inoperable systems. For the systems and components that are not higher risk, any temporary risk increase associated with the proposed allowance will be smaller than what is considered acceptable when the same systems and components are inoperable at power. This is due to the fact that CTs associated with the majority of TSs systems and components were developed for power operation and pose a smaller plant risk for action statement entries initiated or occurring at lower modes of operation as compared to power operation.

The LCO 3.0.4(b) allowance will be used only when the licensee determines that there is a high likelihood that the LCO will be satisfied following the mode change. This will minimize the likelihood of additional temporary risk increases associated with the need to exit a mode due to failure to restore the unavailable equipment within the CT. In most cases, licensees will enter into a higher mode with the intent to move up to Mode 1 (power operation). As discussed in Section 3.2, the revised ROP monitors unplanned power changes as a performance indicator.

Thus, the ROP discourages licensees from entering a mode or other specified condition in the applicability of an LCO and moving up in power when there is a likelihood that the mode would have to be subsequently exited due to failure to restore the unavailable equipment within the CT. Another disincentive for licensees to enter a higher mode when an LCO is not met is related to reporting requirements. It clearly states in 10 CFR 50. 72 and 50. 73 that a report is required when a nuclear plant shutdown or mode change is required by TSs. The NRC's oversight program will provide the framework for inspectors and other staff to follow the history at a specific plant of entering higher modes while an LCO is not met and will use such information in assessing the licensee's actions and performance.

3.1.2 Cumulative Risk Increases The cumulative risk impact of the change to allow the plant to enter a higher mode of operation with one or more safety-related components unavailable (as proposed here), is measured by the average yearly risk increase associated with the change. In general, this cumulative risk increase is assessed in terms of both CDF and LERF (i.e., LlCDF and LlLERF, respectively).

The increase in CDF, due to the proposed change, is expressed by the following equation, which integrates the risk impact from all expected specified conditions (i.e., all expected plant conditions caused by mode changes with various TSs systems and components unavailable):

LlCDF =LDLlCDFi) =L ICCDPi fi (2)

Where LlCDFi =the CDF increase due to specified condition i ICCDPi =the ICCDP associated with specified condition i fi =the average yearly frequency of occurrence of specified condition i

A similar expression can be used for 6LERF by substituting the measure of risk (i.e., LERF for CDF). The magnitude of the 6CDF and 6LERF values associated with plant conditions applicable to LCO 3.0.4(b) allowances can be managed by controlling the temporary risk increases, in terms of both CDF and LERF (i.e., ICCDP and ICLERP), and the frequency of each of such conditions. In addition to the points made in the previous section regarding temporary risk increases, the following points put into perspective how the key elements of the proposed risk-informed approach, used to justify an LCO 3.0.4(b) allowance, are expected to prevent significant cumulative risk increases by limiting the frequency of its use:

The frequency of risk-significant conditions will be limited by not providing the LCO 3.0.4(b) allowances to the higher-risk systems and components.

The frequency of risk-significant conditions will be limited by the requirement to assess the likelihood that the LCO will be satisfied following the mode change.

The frequency of risk-significant conditions is limited by the fact that such conditions can occur only when the plant is returning to power following shutdown (i.e., during a small fraction of time per year) (data over the past 5 years indicate that the plants are averaging 2.1 startups per year).

The addition of the proposed LCO 3.0.4(b) allowances to the plant maintenance activities are not expected to change the plant's average (cumulative) risk significantly.

Risk Assessment and Risk Management of Mode Changes With all safety systems and components operable, a plant can transition up in mode to power operation. With one or more system(s) or component(s) inoperable, this change permits a plant to transition up in mode to power operation if (1) the inoperable system(s) or component(s) are not in the pre-analyzed, higher-risk category; (2) a 10 CFR 50.65(a)(4)-based risk assessment is performed prior to the mode transition; and (3) the requisite risk management actions are taken. The proposed TSs Bases, which shall not become part of the TSs, state:

When an LCO is not met, LCO 3.0.4 also allows entering modes or other specified conditions in the Applicability following assessment of the risk impact and determination that the impact can be managed. The risk assessment may use quantitative, qualitative, or blended approaches, and the risk assessment will be conducted using the plant program, procedures, and criteria in place to implement 10 CFR 50.65(a)(4), which requires that risk impacts of maintenance activities to be assessed and managed.

It should be noted that the risk assessment, for the purposes of LCO 3.0.4(b), must take into account all inoperable equipment, regardless of whether the equipment is included in the licensee's normal 10 CFR 50.65(a)(4) risk assessment scope. The risk assessments will be conducted using the procedures and guidance endorsed by RG 1.182. The results of the risk assessment shall be considered in determining the acceptability of entering the mode or other specified condition in the Applicability and any corresponding risk management actions. A risk assessment and establishment of risk management actions, as appropriate, are required for

determination of acceptable risk for entering mode or other specified conditions in the Applicability when an LCO is not met.

While inoperability permitted by the CTs of TSs required actions take into consideration the safety significance and redundancy of the system or components within the scope of an LCO, the CTs generally do not address or consider concurrent system or component inoperability in multiple LCOs. Therefore, the performance of the 10 CFR 50.65(a)(4) risk assessment, which looks at the entire plant configuration is essential (and required) prior to changing the operational condition. The 10 CFR 50.65(a)(4)-based risk assessment will be used to confirm (or reject) the appropriateness of transitioning up in mode, given the actual status of plant safety equipment.

The risk impact on the plant condition of invoking an LCO 3.0.4(b) allowance will be assessed and managed through the program established to implement 10 CFR 50.65(a)(4). This program is consistent with RG 1.177 and RG 1.174 in its approach. The implementation guidance for paragraph (a)(4) of the Maintenance Rule addresses controlling temporary risk increases resulting from maintenance activities. This guidance, consistent with guidance in RG 1.177, establishes action thresholds based on qualitative and quantitative considerations and risk management actions. Significant temporary risk increases following an LCO 3.0.4(b) allowance are unlikely to occur unless:

High-risk configurations are allowed (e.g., certain combinations of multiple component outages), or

Risk management of plant operation activities is inadequate.

The requirements associated with the proposed change are established to ensure that such conditions will not occur.

The thresholds of the cumulative (aggregate) risk impacts, assessed pursuant to 10 CFR 50.65(a)(4) and the associated implementation guidance, are based on the permanent change guidelines in RG 1.174. Therefore, licensees will manage the risk exercising LCO 3.0.4 in conjunction with the risk from other concurrent plant activities to ensure that any increase, in terms of CDF and LERF, will be small and consistent with the Commission's Safety Goal Policy Statement.

3.2 Oversight The ROP provides a means for assessing the licensee's performance in the application of the proposed mode change flexibility. The adequacy of the licensee's assessment and management of maintenance-related risk are addressed by existing inspection programs and guidance for 10 CFR 50.65(a)(4). Although the current versions of that guidance do not specifically address application of the licensee's 10 CFR 50.65(a)(4) program to support risk-informed TSs, it is expected that, in most cases, risk assessment and management associated with risk-informed TSs would be required by 10 CFR 50.65(a)(4) anyway, because maintenance activities will be involved.

Adoption of the proposed change will make failure to assess and manage the risk of an upward mode change with inoperable equipment covered by TSs prior to commencing such a mode change a violation of TSs. Further, as explained above in general, under most foreseeable circumstances, such a change in configuration would also require a risk assessment under 10 CFR 50.65(a)(4). Inoperable systems or components will necessitate maintenance to restore them to operability, and hence a 10 CFR 50.65(a)(4) risk assessment would be performed prior to the performance of those maintenance actions (except for immediate plant stabilization and restoration actions if necessary). Further, before altering the plant's configuration, including plant configuration changes associated with mode changes, the licensee must update the existing (a)(4) risk assessment to reflect those changes.

The licensee is required by 10 CFR 50.65(a)(4) to assess and manage the increase in risk that may result from the proposed maintenance activities before performing maintenance activities (including but not limited to surveillance, post-maintenance testing, and corrective and preventive maintenance). Oversight of licensee performance in assessing and managing the risk of plant maintenance activities is conducted principally by inspection in accordance with Reactor Oversight Program Baseline IP 71111.13, "Maintenance Risk Assessment and Emergent Work Control." Supplemental IP 62709, "Configuration Risk Assessment and Risk Management Process," is utilized to evaluate the licensee's process, when necessary.

The ROP is described in overview in NUREG-1649, Revision 5, "Reactor Oversight Process,"

and in detail in the NRC Inspection Manual. IP 71111.13 requires verification of performance of risk assessments when they are required by 10 CFR 50.65(a)(4) and in accordance with licensee procedures. The procedure also requires verification of the adequacy of those risk assessments and verification of effective implementation of licensee-prescribed risk management actions. The rule itself requires such assessment and management of risk prior to maintenance activities, including preventive maintenance, surveillance, and testing (and promptly for emergent work) during all modes of plant operation. The guidance documents for both industry implementation of 10 CFR 50.65(a)(4) and NRC oversight of that implementation indicate that changes in plant configuration, which would include mode changes in support of maintenance activities, must be taken into account in the risk assessment and management process.

In addition, the NRC staff developed Inspector Manual Chapter 0609, Appendix K, "Maintenance Risk Assessment and Risk Management Significance Determination Process (SDP)" (ADAMS Accession No. ML051400244), for use in assessing inspection findings related to 10 CFR 50.65(a)(4). The ROP considers inspection findings and performance indicators in evaluating licensee ability to operate safely. The SDP is used to determine the significance of inspection findings related to licensee's assessment and management of the risk associated with performing maintenance activities under all plant operating or shutdown conditions.

Unplanned reactor scrams and unplanned power changes are two of the Reactor Safety Performance Indicators that the ROP utilizes to assess licensee performance and inform the public. The ROP will provide a disincentive to entering into power operation (Operational Condition 1) when there is a significant likelihood that the mode would have to be subsequently exited due to failure to restore the unavailable equipment within the CT.

3.3 Summary The licensee submitted proposed TSs changes to allow entry into a higher mode of operation, or other specified condition in the TSs Applicability, while relying on the TSs conditions and associated required actions and CTs, provided a risk assessment is performed to confirm the acceptability of that action for the existing plant configuration. The proposal revises SL-1 and 2 LCO 3.0.4 and SR 4.0.4. New paragraphs (a), (b), and (c) are proposed for LCO 3.0.4.

The proposed LCO 3.0.4(a) retains the current allowance, permitting the mode change when the TSs required actions allow indefinite operation.

The NRC staff finds that the process proposed by the licensee for assessing and managing risk during the implementation of the proposed LCO 3.0.4(b) allowances meets Commission guidance for TSs changes. Key elements of this process are listed below.

A risk assessment shall be performed before any LCO 3.0.4(b) allowance is invoked.

The risk impact on the plant condition when invoking an LCO 3.0.4(b) allowance will be assessed and managed through the program established to implement 10 CFR 50.65(a)(4) and the associated guidance in RG 1.182. Allowing entry into a higher mode or condition in the Applicability of an LCO after a 10 CFR 50.65(a)(4)-based risk assessment and appropriate risk management actions are taken for the existing plant configuration will ensure that plant safety is maintained.

The LCO 3.0.4(b) allowance will be used only when the licensee determines that there is a high likelihood that the LCO will be satisfied within the required action's CT.

TS systems and components that are of higher risk during mode changes have been identified generically by each owners group for each plant operational mode or condition.

In addition, the licensee identified higher-risk plant-specific systems and components in the individual plant TS. The proposed LCO 3.0.4(b) allowance does not apply to these systems and components for the mode or condition in the applicability of an LCO at which they are of higher risk.

In adopting LCO 3.0.4(b), the licensee will ensure that plant procedures in place to implement 10 CFR 50.65(a)(4) address the situation where entering a mode or other specified condition in the applicability is contemplated with plant equipment inoperable.

Such plant procedures will follow the guidance endorsed by RG 1.182.

The NRC's ROP provides the framework for inspectors and other staff to oversee the implementation of 10 CFR 50.65(a)(4) requirements at a specific plant and assess the licensee's actions and performance.

The LCO 3.0.4(b) allowance does not apply to values and parameters of the TSs that have their own respective LCO (e.g., Reactor Coolant System Specific Activity), but instead those values and parameters are addressed by LCO 3.0.4(c). The TSs values and parameters for which mode transition allowances apply will have a note that states LCO 3.0.4(c) is applicable.

The objective of the proposed change is to provide additional operational flexibility without compromising plant safety.

The NRC staff concludes that the licensee's proposed TS changes are acceptable, because the adopted key elements require the licensee to assess and manage risk and they are consistent with TSTF-359 and the Commission's regulations.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Florida State official was notified of the proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

These amendments change requirements with respect to installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and change SRs.

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 amendment involves no significant hazards consideration and there has been no public comment on such finding (79 FR 30187, dated May 27, 2014). 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) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: Caroline E. Tilton Date: February 27, 2015

M. Nazar The amendments also revise SL-1 and 2, SR 4.0.1 to be consistent with the model SR in NUREG-1432, "Standard Technical Specifications, Combustion Engineering Plants."

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

Sincerely,

/RAJ Farideh E. Saba, 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. 220 to Renewed Facility Operating License No. DPR-67

2. Amendment No. 170 to Renewed Facility Operating License No. NPF-16

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

PUBLIC RidsNrrDorlDpr Resource RidsRgn2MailCenter Resource LPL2-2 R/F RidsNrrDorllp12-2 Resource CTilton, NRR RidsAcrsAcnw_MailCTR Resource RidsNrrDssSrxb Resource PTam, NRR LRonewicz, NRR RidsNrrPMStlucie Resource RidsNrrLABClayton Resource RidsNrrDssStsb Resource AD A MSAccession No.: ML14343A918 *bv e-mail OFFICE DORULPL 1-1/PM DORULPL2-2/PM DORULPL2-2/LAiT DORULPL2-2/LA DSS/STSB/BC NAME PTam FSaba LRonewicz BClavton RElliott DATE 2/2/15 2/24/15 2/25/15 2/25/15 2/2/15 OFFICE OGC-NLO* DORULPL2-2/BC DORULPL2-2/PM NAME DRoth SHelton FSaba DATE 2/25/15 2/27/15 2/27/15 OFFICIAL RECORD COPY

v t e Letter Sequence Approval
TAC:MF3618, Increase Flexibility in Mode Restraints, TSTF-3 (Approved, Closed) TAC:MF3619, Increase Flexibility in Mode Restraints, TSTF-3 (Approved, Closed)
Initiation Request, Request Acceptance... Administration Questions Answers 29 May 2014 Results Approval Other:
MONTHYEAR2014-05-29 [Table View]

ML14343A918

Text

SUBJECT:

Dear Mr. Nazar:

Enclosures:

Attachment:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

Enclosures:

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