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Enclosure 2, Volume 12: Turkey Point Nuclear Plant, Units 3 and 4, Improved Technical Specifications Conversion, Section 3.7, Plant Systems, Revision 2
	ML23151A450
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Site:	Turkey Point
Issue date:	05/24/2023
From:	; Florida Power & Light Co
To:	; Office of Nuclear Reactor Regulation
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ML23151A435	List: L-2023-063, Enclosure 2, Volume 1: Turkey Point Nuclear Plant, Units 3 and 4, Improved Technical Specifications Conversion, Application of Selection Criteria to Technical Specifications, Revision 2, Volume 1; ML23151A436; ML23151A437; ML23151A438; ML23151A440; ML23151A441; ML23151A442; ML23151A443; ML23151A444; ML23151A445; ML23151A446; ML23151A447; ML23151A448; ML23151A449; ML23151A450; ML23151A451; ML23151A452; ML23151A453; ML23151A454;
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ENCLOSURE 2 VOLUME 12 TURKEY POINT NUCLEAR GENERATING STATION UNIT 3 AND UNIT 4 IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS SECTION 3.7 PLANT SYSTEMS Revision 2 l R2

LIST OF ATTACHMENTS

1. ITS 3.7.1 - Main Steam Safety Valves (MSSVs)

2. ITS 3.7.2 - Main Steam Isolation Valves (MSIVs)

3. ITS 3.7.3 - Feedwater Isolation Valves (FIVs) and Feedwater Control Valves (FRVs) and Associated Bypass Valves

4. ITS 3.7.4 - Secondary Specific Activity

5. ITS 3.7.5 - Auxiliary Feedwater (AFW) System

6. ITS 3.7.6 - Condensate Storage Tank (CST)

7. ITS 3.7.7 - Component Cooling Water (CCW) System

8. ITS 3.7.8 - Intake Cooling Water (ICW) System

9. ITS 3.7.9 - Ultimate Heat Sink (UHS)

10. ITS 3.7.10 - Control Room Emergency Ventilation System (CREVS)

11. ITS 3.7.11 - Control Room Emergency Air Temperature Control System (CREATCS)

12. ITS 3.7.12 - Fuel Storage Pool Water Level

13. ITS 3.7.13 - Fuel Storage Pool Boron Concentration

14. ITS 3.7.14 - Spent Fuel Storage

15. Relocated/Deleted Current Technical Specifications (CTS)

16. ISTS Not Adopted

ATTACHMENT 1 ITS 3.7.1, MAIN STEAM SAFETY VALVES (MSSVs)

Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

A01 ITS ITS 3.7.1 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE A01 SAFETY VALVES Main Steam MSSVs LIMITING CONDITION FOR OPERATION Four MSSVs per A02 LCO 3.7.1 3.7.1.1 All main steam line Code safety valves associated with each steam generator shall be OPERABLE with SR 3.7.1.1 lift settings as specified in Table 3.7-2.

Applicability APPLICABILITY: MODES 1, 2, and 3.

A03 Add proposed ACTIONS Note ACTION:

L01 With (3) reactor coolant loops and associated steam generators in operation and with one or more main steam line Code safety valves inoperable, and A04 Add proposed Required Action B.1 ACTION B a. in MODES 1 and 2, with a positive Moderator Temperature Coefficient, operation may continue provided that, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either the inoperable valve(s) are restored to OPERABLE status or the Power Range A05 Neutron Flux High Trip Setpoint is reduced to the maximum allowable percent of RATED THERMAL POWER listed In 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 L02 ACTION C HOT SHUTDOWN within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , or within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ACTION A b. in MODES 1 and 2, with a negative or zero Moderator Temperature Coefficient; or in Mode 3, with a positive, negative or zero Moderator Temperature Coefficient, operation may continue provided that, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either the inoperable valve(s) are restored to OPERABLE status or reactor power is A05 reduced to less than or equal to the maximum allowable percent of RATED THERMAL POWER listed in 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 HOT SHUTDOWN ACTION C within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS Add proposed SR Note L03 SR 3.7.1.1 4.7.1.1 No additional requirements other than those required by the INSERVICE TESTING PROGRAM. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3. A06 TURKEY POINT - UNITS 3 & 4 3/4 7-1 AMENDMENT NOS. 274 AND 269

A01 ITS ITS 3.7.1 TABLE 3.7-1 MAXIMUM ALLOWABLE POWER LEVEL WITH INOPERABLE STEAM LINE SAFETY VALVES DURING THREE LOOP OPERATION MSSVs MAXIMUM NUMBER OF INOPERABLE PER SAFETY VALVES ON ANY MAXIMUM ALLOWABLE POWER LEVEL OPERATING STEAM GENERATOR (PERCENT OF RATED THERMAL POWER) A01 3

1 44 2 27 1

3 10 TABLE 3.7-2 Main Lift Settings STEAM LINE SAFETY VALVES PER LOOP A01 ORIFICE SIZE VALVE NUMBER LIFT SETTING (+/-3%)* ** SQUARE INCHES LA01 Loop A Loop B Loop C

1. RV1400 RV1405 RV1410 1085 psig 16 LA02

2. RV1401 RV1406 RV1411 1100 psig 16

3. RV1402 RV1407 RV1412 1105 psig 16

4. RV1403 RV1408 RV1413 1105 psig 16 SR 3.7.1.1 *The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature LA01 Note and pressure.

SR 3.7.1.1 **All valves tested must have as left lift setpoints that are within +/-1% of the lift setting value listed in Table 3.7-2.

TURKEY POINT - UNITS 3 & 4 3/4 7-2 AMENDMENT NOS. 249 AND 245

DISCUSSION OF CHANGES ITS 3.7.1, MAIN STEAM SAFETY VALVES (MSSVs)

ADMINISTRATIVE CHANGES A01 In the conversion of the Turkey Point Nuclear Generating Station (PTN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG1431, Rev. 5.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).

These changes are designated as administrative changes and are acceptable, because they do not result in technical changes to the CTS.

A02 CTS 3.7.1.1 states that all main steam safety valves (MSSVs) shall be OPERABLE with lift settings as specified in Table 3.7-2. CTS Table 3.7-2 lists lift setting pressures for four MSSVs in each of the three loops. ITS Limiting Condition for Operation (LCO) 3.7.1 requires four MSSVs per steam generator (SG) to be OPERABLE. This changes the CTS by combining the current LCO requirement and portions of CTS Table 3.7-2 into a single ITS LCO requirement.

This change is acceptable because the number of MSSVs required OPERABLE under the various conditions has not changed. This change results in a format change only to follow the manner in which the ISTS presents the MSSV requirements. This change is designated as an administrative change because it does not result is any technical changes to the CTS.

A03 CTS 3.7.1.1, ACTIONS a and b, provide compensatory actions for one or more inoperable MSSVs. CTS 3.7.1.1, ACTION a, requires that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the MSSV(s) be restored to OPERABLE status or the Power Range Neutron Flux High Setpoint trip be reduced in accordance with the requirements of CTS Table 3.7-1. CTS 3.7.1.1 ACTION b, requires that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the MSSV(s) be restored to OPERABLE status or reactor power is reduced to less than or equal to the maximum allowable percent of RATED THERMAL POWER (RTP) listed in Table 3.7-1. ITS 3.7.1 ACTIONS Note states "Separate Condition entry is allowed for each MSSV." This changes the CTS by explicitly specifying separate condition entry for each inoperable MSSV.

The purpose of the CTS ACTIONS is to allow separate condition entry for each inoperable MSSV. Each time it is discovered that an MSSV is inoperable, entry is required and the specified Completion Time is applied to complete the compensatory actions. The ITS 3.7.1 ACTIONS Note allows a separate Completion Time period for each MSSV that is inoperable. This change is acceptable, because it only provides clarification of the Completion Time when one valve is inoperable and, subsequently, a second valve becomes inoperable.

This change is designated as administrative, because it does not result in a technical change to the Specifications.

A04 CTS 3.7.1.1, ACTION a, states that the Power Range Neutron Flux - High Setpoint trip must be reduced per CTS Table 3.7-1 when one or more MSSVs are found to be inoperable. CTS Table 3.7-1 provides the maximum allowable Power Range Neutron Flux - High Setpoint corresponding to the maximum Turkey Point Unit 3 and Unit 4 Page 1 of 6

DISCUSSION OF CHANGES ITS 3.7.1, MAIN STEAM SAFETY VALVES (MSSVs) number of inoperable MSSVs on any operating SG. ITS 3.7.1, ACTION B, requires a reduction in THERMAL POWER and a reduction in the Power Range Neutron Flux - High reactor trip setpoint consistent with the requirements of ITS Table 3.7.1-1. The Table has been revised slightly to provide the associated maximum allowable power for the number of OPERABLE MSSVs. This changes the CTS by adding an additional explicit statement to reduce THERMAL POWER consistent with ITS Table 3.7.1-1 and by stating the maximum allowable power as a function of OPERABLE, instead of inoperable, MSSVs.

The purpose of CTS 3.7.1.1, ACTION a, is to reduce the Power Range Neutron Flux - High Setpoint to within the limits of the safety analyses. Current plant operation dictates that THERMAL POWER is reduced before reducing the setpoints to prevent a reactor trip. Explicitly stating this practice in ITS and stating the maximum power level in terms of OPERABLE instead of inoperable MSSVs does not change how the plant is operated. This change is considered administrative because it does not result in technical changes to the CTS.

A05 CTS 3.7.1.1, ACTIONS a and b, state that with one or more MSSVs inoperable, l R2 either restore the inoperable valves to OPERABLE status or reduce the Power Range Neutron Flux - High Setpoints. ITS 3.7.1, ACTION A, does not include the restoration requirement, only the alternate compensatory measure. This changes the CTS by eliminating the explicit statement to restore the MSSV(s) to OPERABLE status.

This change is acceptable, because it does not result in a technical change to the Technical Specifications. Restoration of compliance with the LCO is always an option in an ACTION, so eliminating the restoration ACTION from the CTS has no effect. In both the CTS and the ITS, if the inoperable MSSV(s) are not restored, actions are taken that result in reducing reactor power to within the relief capability of the OPERABLE MSSVs within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This change is designated as administrative, because it does not result in a technical change to the CTS.

A06 CTS 4.7.1.1 requires verification of each required MSSV lift setpoint, and states "The provisions of Specification 4.0.4 are not applicable for entry into MODE 3."

ITS Surveillance Requirement (SR) 3.7.1.1 does not contain this statement.

However, ITS SR 3.7.1.1 does contain a Note that states, "Only required to be performed in MODES 1 and 2." This changes the CTS by revising presentation of the CTS 4.0.4 exception.

The CTS 4.0.4 exception allows entry into MODE 3 to perform CTS Surveillance 4.7.1.1. This exception is not required in ITS SR 3.7.1.1 because the SR Note in the ITS only requires the SR to be performed in MODES 1 and 2; i.e., allows entry into an operation in MODE 3 to perform the SR as explained in ITS Section 1.4, Example 1.4-5. This change is designated as administrative as it results in no technical change to the CTS.

Turkey Point Unit 3 and Unit 4 Page 2 of 6

DISCUSSION OF CHANGES ITS 3.7.1, MAIN STEAM SAFETY VALVES (MSSVs)

MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS Table 3.7-2 is modified by a footnote (footnote*)

that states, "The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure." ITS 3.7.1 does not contain this information. This changes the CTS by moving details on setting the lift pressure to the ITS Bases.

The removal of these details for performing SRs from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the lift settings and the definition of OPERABLE states that the components must be capable of performing the specified safety function. It is understood that the MSSVs must be adjusted to lift at the settings given under the conditions that the safety analysis assumes the MSSVs will operate. This change is acceptable, because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change, because details for meeting Technical Specification requirements are being removed from the Technical Specifications to the ITS Bases.

LA02 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS Table 3.7-2 specifies the MSSV number and associated lift settings and nozzle size for each MSSV. ITS Table 3.7.1-2 only provides the MSSV number and associated lift setting. This changes the CTS by deleting the required nozzle size and relocating this detail to the Updated Final Safety Analysis Report (UFSAR).

The removal of details related to system design from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the valve numbers and corresponding lift setting.

The nozzle size does not normally vary, because it is a function of the design of the valve. The lift settings can vary and are adjustable. Therefore, this information is important to be retained in the Technical Specification. Also, this change is acceptable because the removed information will be adequately controlled in the UFSAR. The UFSAR is controlled under 10 CFR 50.59, which Turkey Point Unit 3 and Unit 4 Page 3 of 6

DISCUSSION OF CHANGES ITS 3.7.1, MAIN STEAM SAFETY VALVES (MSSVs) ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change, because information relating to system design is being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 4 - Relaxation of Required Action) CTS 3.7.1.1, ACTION a, states, in part, that with one or more MSSVs inoperable in MODE 1 or 2 with a positive moderator temperature coefficient (MTC), reduce the Power Range Neutron Flux - High Setpoint trip within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . ITS 3.7.1, ACTION A, requires only a reduction in THERMAL POWER to less than or equal to the Maximum Allowable

% RTP specified in Table 3.7.1-1 for the number of OPERABLE MSSVs when one or more MSSVs are inoperable in MODE 2, irrespective of MTC condition. This changes the CTS by only requiring the Power Range Neutron Flux - High Setpoint trip be reduced when in MODE 1 and requiring only a power reduction when one or more MSSVs are inoperable in MODE 2.

The purpose of CTS 3.7.1.1 is to ensure that the MSSVs are capable of relieving Main Steam System pressure. ISTS 3.7.1, Required Action B.2, requires the l R2 Power Range Neutron Flux - High trip setpoint to be reduced but is modified by a Note stating that this action is only required in MODE 1. Since this Action is relevant only during MODE 1 operation at PTN, the MTC condition is irrelevant with respect to operation in MODE 2. In MODES 2 and 3, THERMAL POWER is at or below 5% RTP and Reactor Trip System neutron flux trips specified in LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation," other than the Power Range Neutron Flux - High trip, provide the necessary protection, regardless of the MTC condition. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, considering the low probability of a Design Basis Accident (DBA) occurring during the repair period and RTS neutron flux trips other than the power range neutron flux trip continue to provide reactor protection. Additionally, a reactor power reduction is sufficient to limit primary side heat generation such that overpressurization of the secondary side is precluded for any Reactor Coolant System (RCS) heatup event. Furthermore, for this case there is sufficient total steam flow capacity provided by the turbine and remaining OPERABLE MSSVs to preclude overpressurization in the event of an increased reactor power due to reactivity insertion, such as in the event of an uncontrolled rod cluster control assembly bank withdrawal at power. This change is consistent with the ISTS ACTIONS applying presentation differences to support the PTN licensing basis.

This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS.

L02 (Category 3 - Relaxation of Completion Time) CTS 3.7.1.1, ACTION a, specifies the compensatory actions when one or more MSSVs are inoperable. The action allows operation to continue provided that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either the inoperable MSSV(s) are restored to OPERABLE status or the Power Range Neutron Turkey Point Unit 3 and Unit 4 Page 4 of 6

DISCUSSION OF CHANGES ITS 3.7.1, MAIN STEAM SAFETY VALVES (MSSVs)

Flux - High Setpoint trip is reduced per Table 3.7-1. ITS 3.7.1, Required Action B.2, requires the reduction of the Power Range Neutron Flux - High reactor trip setpoint to less than or equal to the Maximum Allowable percent RTP specified in Table 3.7.1-1 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This changes the CTS by extending the time allowed to reduce the Power Range Neutron Flux - High reactor trip setpoints.

The purpose of CTS 3.7.1.1, ACTION a, is to limit the time the unit can operate with inoperable MSSVs without reducing the Power Range Neutron Flux - High reactor trip setpoints. This change is acceptable, because the Completion Time is consistent with safe operation under the specified Condition, considering the OPERABLE status of the redundant systems or features and that a power reduction is required. This includes the capacity and capability of remaining systems or features, a reasonable time for repairs, and the low probability of a DBA occurring during the allowed Completion Time. This change extends the time allowed to reduce the Power Range Neutron Flux - High reactor trip setpoints when the MSSVs are inoperable. The time extension is from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . However, the time to reduce THERMAL POWER to within the same limits (four hours) is maintained in ITS 3.7.1 as Required Action B.1. This change is acceptable because the Completion Time of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is a reasonable time to reset the channels of a protective function, and on the low probability of the occurrence of a transient that could result in steam generator overpressure during this period. In addition, the actual reactor power level continues to be required to be reduced to within the same limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Thus, operation of the unit at a RTP above limits with one or more inoperable MSSV(s) is still only allowed for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , consistent with the current allowance. This change is designated as less restrictive, because additional time is allowed to restore parameters to within the LCO limits than was allowed in the CTS.

L03 (Category 7 - Relaxation of Surveillance Frequency) CTS 4.7.1.1 requires verification of each MSSV lift setpoint pursuant to the INSERVICE TESTING PROGRAM. CTS Table 3.7-2 Footnote

states that the lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure. ITS SR 3.7.1.1 requires the same testing; however, a Note has been included that requires the performance of the lift setpoint verification only in MODES 1 and 2, which corresponds to the valve's ambient conditions at normal operating temperature and pressure. This changes the CTS by adding a Note that requires performance of the MSSV lift setpoint verification only in MODES 1 and 2.

The purpose of CTS 4.7.1.1 is to perform the MSSV lift setpoint verification in accordance with the Frequency of the INSERVICE TEST PROGRAM. This change is acceptable, because the allowance has been evaluated to ensure that it provides an acceptable level of equipment reliability. The SR is modified by a Note that states the Surveillance is only required to be performed in MODES 1 and 2. The Note allows entry into and operation in MODE 3 prior to performing the SR as explained in ITS Section 1.4, Example 1.4-5. By allowing entry into MODE 3 prior to performing the SR, testing can be performed at ambient conditions of normal operating temperature and pressure. Otherwise, if the MSSVs are not tested at hot conditions, the lift setting pressure is corrected to Turkey Point Unit 3 and Unit 4 Page 5 of 6

DISCUSSION OF CHANGES ITS 3.7.1, MAIN STEAM SAFETY VALVES (MSSVs) ambient conditions of the valve at operating temperature and pressure. This change is designated as less restrictive because Surveillances will be performed in fewer operating Conditions than in the CTS.

Turkey Point Unit 3 and Unit 4 Page 6 of 6

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

MSSVs CTS 3.7.1 3.7 PLANT SYSTEMS 3.7.1 Main Steam Safety Valves (MSSVs)

Four 3.7.1.1 LCO 3.7.1 [ Five] MSSVs per steam generator shall be OPERABLE. 1 Applicability APPLICABILITY: MODES 1, 2, and 3.

ACTIONS

NOTE-----------------------------------------------------------

DOC A03 Separate Condition entry is allowed for each MSSV.

REVIEWER' S NOTE-------------------------------------------------

The

noted text is required for units that are licensed to operate at partial power with a positive 3

Moderator Temperature Coefficient (MTC).

CONDITION REQUIRED ACTION COMPLETION TIME INSERT 1 A. One or more steam A.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months generators with one POWER to [ 72] % RTP.

MSSV inoperable [ and the Moderator Temperature Coefficient (MTC) zero or negative at all power levels] * .

DOC L01 B. One or more steam B.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 1

generators with two or POWER to less than or more MSSVs equal to the Maximum inoperable. Allowable % RTP specified in Table 3.7.1-1 for the

[ OR number of OPERABLE MSSVs.

One or more steam generators with one AND MSSV inoperable and the MTC positive at any power level. ]

Westinghouse STS 3.7.1-1 Rev. 5.0 1 Turk ey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY

CTS 3.7.1 2

INSERT 1 Action b A. One or more steam A.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months generators with one or POWER to less than or more MSSV inoperable equal to the Maximum in MODE 1 with the Allowable % RTP specified moderator temperature in Table 3.7.1-1 for the coefficient (MTC) zero number of OPERABLE or negative. MSSVs.

OR Action b One or more steam generators with one or more MSSV inoperable in MODE 2 or 3.

Action a B. One or more steam B.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months DOC L01 generators with one or POWER to less than or more MSSV inoperable equal to the Maximum in MODE 1 with a Allowable % RTP specified positive MTC. in Table 3.7.1-1 for the number of OPERABLE MSSVs.

AND Action a B.2 Reduce the Power Range 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months DOC L02 Neutron Flux - High reactor trip setpoint to less than or equal to the Maximum Allowable % RTP specified in Table 3.7.1-1 for the number of OPERABLE MSSVs.

Insert Page 3.7.1-1

MSSVs CTS 3.7.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.2 --------------NOTE--------------

Only required in MODE 1.

Reduce the Power Range 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Neutron Flux - High reactor trip setpoint to less than or 1 equal to the Maximum Allowable % RTP specified in Table 3.7.1-1 for the number of OPERABLE MSSVs.

ACTION a C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 5 ACTION b associated Completion Time not met. AND 18 OR C.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 5 1 One or more steam all 2 generators with [ 4]

MSSVs inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.7.1.1 SR 3.7.1.1 -------------------------------NOTE------------------------------

DOC L03 Only required to be performed in MODES 1 and 2.

4.7.1.1 Verify each required MSSV lift setpoint per In accordance Table 3.7.1-2 in accordance with the INSERVICE with the TESTING PROGRAM. Following testing, lift setting INSERVICE shall be within +/-1%. TESTING PROGRAM Westinghouse STS 3.7.1-2 Rev. 5.0 1 Turk ey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY

MSSVs CTS 3.7.1 Table 3.7.1-1 (page 1 of 1)

Table 3.7-1 OPERABLE Main Steam Safety Valves versus Maximum Allowable Power NUMBER OF OPERABLE MSSVs PER STEAM MAXIMUM ALLOWABLE GENERATOR POWER (% RTP) 3 44

[ 4] [ 65]

2 27 2 3 [ 46]

1 10 1 2 [ 28]

Westinghouse STS 3.7.1-3 Rev. 5.0 1 Turk ey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY

MSSVs CTS 3.7.1 Table 3.7-2 Table 3.7.1-2 (page 1 of 1)

Main Steam Safety Valve Lift Settings VALVE NUMBER STEAM GENERATOR LIFT SETTING (psig +/- [ 3] %)

A# 1 B# 2 [#

C 3] [ # 4]

[ ] [ ] [ ] [ ] [ ]

2

[ ] [ ] [ ] [ ] [ ]

INSERT 2 Westinghouse STS 3.7.1-4 Rev. 5.0 1 Turk ey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY

CTS 3.7.1 2

INSERT 2 RV1400 RV1405 RV1410 1085 Table 3.7-2 RV1401 RV1406 RV1411 1100 RV1402 RV1407 RV1412 1105 RV1403 RV1408 RV1413 1105 Insert Page 3.7.1-4

JUSTIFICATION FOR DEVIATIONS ITS 3.7.1, MAIN STEAM SAFETY VALVES (MSSVs)

1. Changes are made (additions, deletions, and/or changes) to the Improved Standard Technical Specification (ISTS) that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

3. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.

4. Editorial changes made for consistency with Specification.

5. Changes have been made to add or delete Actions, the subsequent Actions and Required Actions have been renumbered to reflect the additions and deletions.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

MSSVs B 3.7.1 B 3.7 PLANT SYSTEMS B 3.7.1 Main Steam Safety Valves (MSSVs)

BASES BACKGROUND The primary purpose of the MSSVs is to provide overpressure protection for the secondary system. The MSSVs also provide protection against overpressurizing the reactor coolant pressure boundary (RCPB) by providing a heat sink for the removal of energy from the Reactor Coolant System (RCS) if the preferred heat sink , provided by the Condenser and Circulating Water System, is not available.

Four

[ Five] MSSVs are located on each main steam header, outside 1 containment, upstream of the main steam isolation valves, as described U

in the FSAR, Section [ 10.3.1] (Ref. 1). The MSSVs must have sufficient 2 1 10.2 t to tt o t to 110% of the steam generator design pressure in order to meet the requirements of the ASME Code,Section III (Ref. 2). The MSSV design includes staggered setpoints, according to Table 3.7.1-2 in the accompanying LCO, so that only the needed valves will actuate. Staggered setpoints reduce the potential for valve chattering that is due to steam pressure insufficient to fully open all valves following a turbine reactor trip.

APPLICABLE The design basis for the MSSVs comes from Reference 2 and its purpose SAFETY is to limit the se o t to 110% of design pressure ANALYSES for any anticipated operational occurrence (AOO) or accident considered in the Design Basis Accident (DBA) and transient analysis.

The events that challenge the relieving capacity of the MSSVs, and thus RCS pressure, are those characterized as decreased heat removal Chapter 14 U

events, which are presented in the FSAR, Section [ 15.2] (Ref. 3). Of 2 1 these, the full power turbine trip without steam dump is typically the limiting AOO. This event also terminates normal feedwater flow to the steam generators.

The safety analysis demonstrates that the transient response for turbine trip occurring from full power without a direct reactor trip presents no hazard to the integrity of the RCS or the Main Steam System. One turbine trip analysis is performed assuming primary system pressure control via operation of the pressurizer relief valves and spray. This analysis demonstrates that the DNB design basis is met. Another analysis is performed assuming no primary system pressure control, but crediting reactor trip on high pressurizer pressure and operation of the pressurizer safety valves. This analysis demonstrates that RCS integrity Turk ey Point Unit 3 and Unit 4 Revision XXX 2 Westinghouse STS B 3.7.1-1 Rev. 5.0

MSSVs B 3.7.1 BASES APPLICABLE SAFETY ANALYSES (continued) is maintained by showing that the maximum RCS pressure does not exceed 110% of the design pressure. All cases analyzed demonstrate that the MSSVs maintain Main Steam System integrity by limiting the maximum steam pressure to less than 110% of the steam generator design pressure.

In addition to the decreased heat removal events, reactivity insertion events may also challenge the relieving capacity of the MSSVs. The uncontrolled rod cluster control assembly (RCCA) bank withdrawal at power event is characterized by an increase in core power and steam generation rate until reactor trip occurs when either the Overtemperature T or Power Range Neutron Flux-High setpoint is reached. Steam flow to the turbine will not increase from its initial value for this event. The increased heat transfer to the secondary side causes an increase in steam pressure and may result in opening of the MSSVs prior to reactor trip, assuming no credit for operation of the atmospheric or condenser U

steam dump valves. The FSAR Section [ 15.4] safety analysis of the 2 1 Chapter 14 RCCA bank withdrawal at power event for a range of initial core power levels demonstrates that the MSSVs are capable of preventing secondary side overpressurization for this AOO.

U The FSAR safety analyses discussed above assume that all of the 2 MSSVs for each steam generator are OPERABLE. If there are inoperable MSSV(s), it is necessary to limit the primary system power during steady-state operation and AOOs to a value that does not result in exceeding the combined steam flow capacity of the turbine (if available) and the remaining OPERABLE MSSVs. The required limitation on primary system power necessary to prevent secondary system overpressurization may be determined by system transient analyses or conservatively arrived at by a simple heat balance calculation. In some circumstances it is necessary to limit the primary side heat generation that can be achieved during an AOO by reducing the setpoint of the Power Range Neutron Flux-High reactor trip function. For example, if 3 more than one MSSV on a single steam generator is inoperable, an uncontrolled RCCA bank withdrawal at power event occurring from a partial power level may result in an increase in reactor power that exceeds the combined steam flow capacity of the turbine and the remaining OPERABLE MSSVs. Thus, for multiple inoperable MSSVs on the same steam generator it is necessary to prevent this power increase by lowering the Power Range Neutron Flux-High setpoint to an appropriate value. [ When the Moderator Temperature Coefficient (MTC) 2 is positive, the reactor power may increase above the initial value during Turk ey Point Unit 3 and Unit 4 Revision XXX 2 Westinghouse STS B 3.7.1-2 Rev. 5.0

MSSVs B 3.7.1 BASES APPLICABLE SAFETY ANALYSES (continued) an RCS heatup event (e.g., turbine trip). Thus, for any number of inoperable MSSVs, it is necessary to reduce the trip setpoint if a positive MTC may exist at partial power conditions, unless it is demonstrated by 2 analysis that a specified reactor power reduction alone is sufficient to prevent overpressurization of the steam system.]

The MSSVs are assumed to have two active and one passive failure modes. The active failure modes are spurious opening, and failure to reclose once opened. The passive failure mode is failure to open upon demand.

The MSSVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

four LCO The accident analysis requires that [ five] MSSVs per steam generator be 1 OPERABLE to provide overpressure protection for design basis four transients occurring at 102% RTP. The LCO requires that [ five] MSSVs 1 per steam generator be OPERABLE in compliance with Reference 2, and the DBA analysis.

The OPERABILITY of the MSSVs is defined as the ability to open upon demand within the setpoint tolerances, to relieve steam generator overpressure, and reseat when pressure has been reduced. The OPERABILITY of the MSSVs is determined by periodic surveillance testing in accordance with the INSERVICE TESTING PROGRAM.

This LCO provides assurance that the MSSVs will perform their designed safety functions to mitigate the consequences of accidents that could result in a challenge to the RCPB, or Main Steam System integrity.

four APPLICABILITY In MODES 1, 2, and 3, [ five] MSSVs per steam generator are required to 1 be OPERABLE to prevent Main Steam System overpressurization.

In MODES 4 and 5, there are no credible transients requiring the MSSVs.

The steam generators are not normally used for heat removal in MODES 5 and 6, and thus cannot be overpressurized; there is no requirement for the MSSVs to be OPERABLE in these MODES.

Turk ey Point Unit 3 and Unit 4 Revision XXX 2 Westinghouse STS B 3.7.1-3 Rev. 5.0

MSSVs B 3.7.1 BASES ACTIONS The ACTIONS Table is modified by a Note indicating that separate Condition entry is allowed for each MSSV.

With one or more MSSVs inoperable, action must be tak en so that the available MSSV relieving capacity meets Reference 2 requirements.

four Operation with less than all [ five] MSSVs OPERABLE for each steam 1 generator is permissible, if THERMAL POWER is limited to the relief capacity of the remaining MSSVs. This is accomplished by restricting THERMAL POWER so that the energy transfer to the most limiting steam generator is not greater than the available relief capacity in that steam generator.

A.1 one or more in MODE 1 In the case of only a single inoperable MSSV on one or more steam 1 generators [ when the Moderator Temperature Coefficient is not positive] 2

, a reactor power reduction alone is sufficient to limit primary side heat generation such that overpressurization of the secondary side is precluded for any RCS heatup event. Furthermore, for this case there is sufficient total steam flow capacity provided by the turbine and remaining OPERABLE MSSVs to preclude overpressurization in the event of an increased reactor power due to reactivity insertion, such as in the event of an uncontrolled RCCA bank withdrawal at power. Therefore, Required Action A.1 requires an appropriate reduction in reactor power within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The maximum THERMAL POWER corresponding to the heat removal capacity of the remaining OPERABLE MSSVs is determined via a conservative heat balance calculation as described in the attachment to Reference 6, with an appropriate allowance for calorimetric power uncertainty.

REVIEWER S NOTE-----------------------------------

To determine the maximum THERMAL POWER corresponding to the heat removal capacity of the remaining OPERABLE MSSVs, the governing heat transfer relationship is the equation q = m the heat input from the primary side, m is the mass flow rate of the 4 t t t t to o t t secondary side water to steam. If it is conservatively assumed that the secondary side water is all saturated liquid (i.e., no subcooled feedwater),

then t t to o to fg) at the steam relief pressure.

The following equation is used to determine the maximum allowable power level for continued operation with inoperable MSSV(s):

Turk ey Point Unit 3 and Unit 4 Revision XXX 2 Westinghouse STS B 3.7.1-4 Rev. 5.0

MSSVs B 3.7.1 BASES ACTIONS (continued)

Nuclear Steam Supply System (NSSS)

Maximum NSSS Power (100/Q) (wshfgN) / K 5 where:

Q = Nominal NSSS power rating of the plant (including reactor coolant pump heat), MWt K = Conversion factor, 947.82 (Btu/sec)/MWt ws = Minimum total steam flow rate capability of the OPERABLE MSSVs on any one steam generator at the highest OPERABLE MSSV opening pressure, including tolerance and accumulation, as appropriate, lbm/sec.

hfg = Heat of vaporization at the highest MSSV opening pressure, including tolerance and accumulation as appropriate, Btu/lbm.

N = Number of steam generators in the plant.

For use in determining the %RTP in the Required Action statement A.1, the Maximum NSSS Power calculated above is reduced by [2]% RTP to 2 account for calorimetric power uncertainty.

4 B.1 and B.2 one or more In the case of multiple inoperable MSSVs on one or more steam in MODE 1 with a positive MTC, generators, with a reactor power reduction alone there may be insufficient 1 total steam flow capacity provided by the turbine and remaining OPERABLE MSSVs to preclude overpressurization in the event of an increased reactor power due to reactivity insertion, such as in the event of an uncontrolled RCCA bank withdrawal at power. [Furthermore, for a single inoperable MSSV on one or more steam generators when the 2 Moderator Temperature Coefficient is positive the reactor power may increase as a result of an RCS heatup event such that flow capacity of the remaining OPERABLE MSSVs is insufficient.] The 4-hour Completion Time for Required Action B.1 is consistent with A.1. An additional 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months is allowed in Required Action B.2 to reduce the setpoints. The Completion Time of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is based on a reasonable time to correct the MSSV inoperability, the time required to perform the power reduction, operating experience in resetting all channels of a protective function, and on the low probability of the occurrence of a transient that could result in steam generator overpressure during this period.

Turkey Point Unit 3 and Unit 4 Revision XXX 2 Westinghouse STS B 3.7.1-5 Rev. 5.0

MSSVs B 3.7.1 BASES ACTIONS (continued)

The maximum THERMAL POWER corresponding to the heat removal capacity of the remaining OPERABLE MSSVs is determined via a conservative heat balance calculation as described in the attachment to Reference 6, with an appropriate allowance for Nuclear Instrumentation System trip channel uncertainties.

REVIEWERS NOTE-----------------------------------

To determine the Table 3.7.1-1 Maximum Allowable Power for Required Actions B.1 and B.2 (%RTP), the Maximum NSSS Power calculated 4 using the equation in the Reviewer's Note above is reduced by [9]% RTP to account for Nuclear Instrumentation System trip channel uncertainties.

Required Action B.2 is modified by a Note, indicating that the Power Range Neutron Flux-High reactor trip setpoint reduction is only required in 2 MODE 1. In MODES 2 and 3 the reactor protection system trips specified in LCO 3.3.1, "Reactor Trip System Instrumentation," provide sufficient protection.

The allowed Completion Times are reasonable based on operating experience to accomplish the Required Actions in an orderly manner without challenging unit systems.

C.1 and C.2 If the Required Actions are not completed within the associated all Completion Time, or if one or more steam generators have [4] 6 inoperable inoperable MSSVs, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least 18 MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed 6 Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.1.1 REQUIREMENTS This SR verifies the OPERABILITY of the MSSVs by the verification of each MSSV lift setpoint in accordance with the INSERVICE TESTING PROGRAM. The ASME Code (Ref. 4), requires that safety and relief valve tests be performed in accordance with ANSI/ASME OM-1-1987 1 (Ref. 5). According to Reference 5, the following tests are required:

a. Visual examination, Turkey Point Unit 3 and Unit 4 Revision XXX 2 Westinghouse STS B 3.7.1-6 Rev. 5.0

MSSVs B 3.7.1 BASES SURVEILLANCE REQUIREMENTS (continued)

b. Seat tightness determination,

c. Setpoint pressure determination (lift setting),

d. Compliance with owner' s seat tightness criteria, and

e. Verification of the balancing device integrity on balanced valves.

The ANSI/ASME Standard requires that all valves be tested every 5 years, and a minimum of 20% of the valves be tested every 24 months.

The ASME Code specifies the activities and frequencies necessary to satisfy the requirements. Table 3.7.1-2 allows a +/- [ 3] % setpoint tolerance 1 for OPERABILITY; however, the valves are reset to +/- 1% during the Surveillance to allow for drift. The lift settings, according to Table 3.7.1-2, correspond to ambient conditions of the valve at nominal operating temperature and pressure.

This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. The MSSVs may be either bench tested or tested in situ at hot conditions using an assist device to simulate lift pressure. If the MSSVs are not tested at hot conditions, the lift setting pressure shall be corrected to ambient conditions of the valve at operating temperature and pressure.

10.2 REFERENCES

1. FSAR, Section [ 10.3.1] . 1 2 U

2. ASME, Boiler and Pressure Vessel Code,Section III, Article NC-7000, Class 2 Components.

Chapter 14 1 2

3. FSAR, Section [ 15.2] .

U

4. ASME Code for Operation and Maintenance of Nuclear Power Plants.

2004 Edition through 2006 Addenda 2

5. ANSI/ASME OM-1-1987.

6. NRC Information Notice 94-60, " Potential Overpressurization of the Main Steam System," August 22, 1994.

Turk ey Point Unit 3 and Unit 4 Revision XXX 2 Westinghouse STS B 3.7.1-7 Rev. 5.0

JUSTIFICATION FOR DEVIATIONS ITS 3.7.1 BASES, MAIN STEAM SAFETY VALVES (MSSVs)

1. The Improved Standard Technical Specification (ISTS) contain bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is changed to reflect the current licensing basis.

2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

3. This redundant example has been deleted.

4. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.

5. Editorial/grammatical changes made.

6. Changes are made to be consistent with changes made to the Specification.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.7.1, MAIN STEAM SAFETY VALVES (MSSVS)

There are no specific No Significant Hazards Considerations for this Specification.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

ATTACHMENT 2 ITS 3.7.2, MAIN STEAM ISOLATION VALVES (MSIVs)

Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

A01 ITS ITS 3.7.2 PLANT SYSTEMS (MSIVs)

A01 MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION Three s A01 LCO 3.7.2 3.7.1.5 Each main steam line isolation valve (MSIV) shall be OPERABLE.

, excepted when all MSIVs are closed and de-activated. L01 Applicability APPLICABILITY: MODES 1, 2, and 3.

ACTION:

ACTION A MODE 1:

With one MSIV inoperable but open, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; otherwise be in MODE 2 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION C MODES 2 and 3:

Add Action C Note A02 With one or more MSIVs inoperable, subsequent operation in MODE 2 or 3 may continue provided:

ACTION C 1. The inoperable MSIVs are closed within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and ACTION C 2. The inoperable MSIVs are verified closed once per 7 days.

ACTION D Otherwise, be in 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 .

SURVEI LLANCE REQUIREMENTS the isolation time of each MSIV is within limits LA01 SR 3.7.2.1 4.7.1.5 Each MSIV shall be demonstrated OPERABLE by verifying full closure within 5 seconds when tested in accordance with the INSERVICE TESTING PROGRAM. The provisions of Specification 4.0.4 are not A03 applicable for entry into MODE 3.

M01 Add proposed SR 3.7.2.2 TURKEY POINT - UNITS 3 & 4 3/4 7-10 AMENDMENT NOS. 284 AND 278

A01 ITS ITS 3.7.2 PLANT SYSTEMS A01 3/4 3.7.1.6 DELETED TURKEY POINT - UNITS 3 & 4 3/4 7-11 AMENDMENT NOS. 282 AND 276

A01 ITS ITS 3.7.2 PLANT SYSTEMS A01 4.7.1.6.4 DELETED TURKEY POINT - UNITS 3 & 4 3/4 7-12 AMENDMENT NOS. 282 AND 276

DISCUSSION OF CHANGES ITS 3.7.2, MAIN STEAM ISOLATION VALVES (MSIVs)

ADMINISTRATIVE CHANGES A01 In the conversion of the Turkey Point Generating Station (PTN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 5.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 3.7.1.5 ACTIONS (MODES 2 and 3) 1 and 2 provide compensatory actions for one or more inoperable Main Steam Isolation Valves (MSIVs). CTS 3.7.1.5 ACTION 1 (MODES 2 and 3) requires that within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , inoperable MSIV(s) are closed. CTS 3.7.1.5 ACTION 2 (MODES 2 and 3) requires that once per 7 days the inoperable MSIV(s) be verified closed. ITS 3.7.1 ACTION C Note states "Separate Condition entry is allowed for each MSIV." This changes the CTS by explicitly specifying separate condition entry for each inoperable MSIV.

The purpose of the CTS ACTIONS is to allow separate condition entry for each inoperable MSIV. Each time it is discovered that an MSIV is inoperable, entry is required and the specified Completion Time is allowed to complete the compensatory actions. The ITS 3.7.2 ACTIONS Note allows a separate Completion Time clock for each MSIV that is inoperable. This change is acceptable, because it only provides clarification of the Completion Time when one valve is inoperable and, subsequently, a second valve becomes inoperable.

This change is designated as administrative, because it does not result in a technical change to the Specifications.

A03 CTS 4.7.1.5 requires verification of each required MSIV lift setpoint, and states "The provisions of Specification 4.0.4 are not applicable for entry into MODE 3."

ITS Surveillance Requirement (SR) 3.7.2.1 does not contain this statement.

However, ITS SR 3.7.2.1 does contain a Note that states, "Only required to be performed in MODES 1 and 2." This changes the CTS by not adding the CTS 4.0.4 exception.

The CTS 4.0.4 exception allows entry into MODE 3 to perform CTS Surveillance 4.7.1.5. This exception is not required in ITS SR 3.7.2.1 because the SR Note in the ITS only requires the SR to be performed in MODES 1 and 2.

This change is designated as administrative as it results in no technical change to the CTS.

MORE RESTRICTIVE CHANGES M01 ITS SR 3.7.2.2 states, in part, "Verify each MSIV actuates to the isolation position on an actual or simulated actuation signal." The CTS does not contain this Surveillance Requirement. This changes the CTS by adding the ITS requirement of SR 3.7.2.2.

Turkey Point Unit 3 and Unit 4 Page 1 of 3

DISCUSSION OF CHANGES ITS 3.7.2, MAIN STEAM ISOLATION VALVES (MSIVs)

The purpose of ITS SR 3.7.2.2 is to verify the actuation system of the MSIVs.

This SR is normally performed upon returning the plant to operation following a refueling outage. This change is acceptable because it provides additional assurance that the actuation system will be capable of performing its specified safety function. This change is designated as more restrictive because it adds a SR to the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including l R2 Design Limits) CTS 3.7.1.5 requirement to verify the closure time of the MSIV l contains the actual closure time of 5 seconds. ITS 3.7.2 requires verification of the isolation time of the MSIV but does not include the isolation time. This changes the CTS moving the actual closure time acceptance criteria from the Technical Specification to the TS Bases.

The purpose of the SR is to verify the closure time of the MSIVs. The removal of MSIV closure time from the Technical Specifications and moving it to the Technical Specification Bases is acceptable because this type of information is not necessary to be included in the Technical Specifications in order to provide adequate protection of public health and safety. The ITS retains the requirement to verify that the isolation time of each MSIV is within limits. Also, this change is acceptable because this detail will be adequately controlled in the Technical Specification Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because MSIV closure times is being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 2 - Relaxation of Applicability) CTS 3.7.1.5 is applicable in MODES 1, 2, and 3. ITS Limiting Condition for Operation (LCO) 3.7.2 is applicable in MODE 1, and in MODES 2 and 3 except when all MSIVs are closed and de-activated. This changes the CTS by making the Specification not applicable in MODES 2 and 3 when all MSIVs are closed and de-activated.

The purpose of the ITS 3.7.2 Applicability exception is to clarify that an MSIV is not required to be OPERABLE when the valve(s) is in a position that supports the safety analyses. This change is acceptable, because the requirements continue to ensure that the structures, systems, and components are maintained in the MODES and other specified conditions assumed in the safety analyses. When in Turkey Point Unit 3 and Unit 4 Page 2 of 3

DISCUSSION OF CHANGES ITS 3.7.2, MAIN STEAM ISOLATION VALVES (MSIVs) the closed position, the valves are in the assumed accident position. This change is designated as less restrictive, because the ITS LCO requirements are applicable in fewer operating conditions than in the CTS.

Turkey Point Unit 3 and Unit 4 Page 3 of 3

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

MSIVs CTS 3.7.2 3.7 PLANT SYSTEMS 3.7.2 Main Steam Isolation Valves (MSIVs)

Three 2

3.7.1.5 LCO 3.7.2 [ Four] MSIVs shall be OPERABLE.

Applicability APPLICABILITY: MODE 1, MODES 2 and 3 except when all MSIVs are closed [ and de-activated] . 2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 24 ACTION A. One MSIV inoperable in A.1 Restore MSIV to [ 8] hours 2 MODE 1 MODE 1. OPERABLE status.

[ OR In accordance with the Risk Informed Completion Time Program]

ACTION - B. Required Action and B.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months MODE 1 associated Completion Time of Condition A not met.

C. ------------NOTE------------ C.1 Close MSIV. [ 8] hours 2 Separate Condition entry is allowed for each AND MSIV.

C.2 Verify MSIV is closed. Once per 7 days ACTION - MODE One or more MSIVs 2 and 3, 1 and 2 inoperable in MODE 2 or 3.

ACTION - D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months MODE 2 and 3 associated Completion Time of Condition C not AND met.

D.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Westinghouse STS 3.7.2-1 Rev. 5 1 Turk ey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY

MSIVs CTS 3.7.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.7.1.5 SR 3.7.2.1 -------------------------------NOTE------------------------------

DOC L02 Only required to be performed in MODES 1 and 2.

Verify the isolation time of each MSIV is within In accordance limits. with the INSERVICE TESTING PROGRAM DOC M01 SR 3.7.2.2 -------------------------------NOTE------------------------------

Only required to be performed in MODES 1 and 2.

Verify each MSIV actuates to the isolation position [ [ 18] months on an actual or simulated actuation signal. 2 OR In accordance with the Surveillance Frequency Control Program ]

Westinghouse STS 3.7.2-2 Rev. 5 1 Turk ey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY

JUSTIFICATION FOR DEVIATIONS ITS 3.7.2, MAIN STEAM ISOLATION VALVES (MSIVs)

1. Changes are made (additions, deletions, and/or changes) to the Improved Standard Technical Specification (ISTS) that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

MSIVs B 3.7.2 B 3.7 PLANT SYSTEMS B 3.7.2 Main Steam Isolation Valves (MSIVs)

BASES BACKGROUND The MSIVs isolate steam flow from the secondary side of the steam generators following a high energy line break (HELB). MSIV closure terminates flow from the unaffected (intact) steam generators.

One MSIV is located in each main steam line outside, but close to, containment. The MSIVs are downstream from the main steam safety valves (MSSVs) and auxiliary feedwater (AFW) pump turbine steam supply, to prevent MSSV and AFW isolation from the steam generators by MSIV closure. Closing the MSIVs isolates each steam generator from the others, and isolates the turbine, Steam Bypass System, and other auxiliary steam supplies from the steam generators.

The MSIVs close on a main steam isolation signal generated by either low steam generator pressure or high containment pressure. The MSIVs fail closed on loss of control or actuation power.

Each MSIV has an MSIV bypass valve. Although these bypass valves are normally closed, they receive the same emergency closure signal as do their associated MSIVs. The MSIVs may also be actuated manually.

10.2 U A description of the MSIVs is found in the FSAR, Section [ 10.3] (Ref. 1). 1 2 APPLICABLE The design basis of the MSIVs is established by the containment analysis SAFETY 14.3.4 for the large steam line break (SLB) inside containment, discussed in the ANALYSES FSAR, Section [ 6.2] (Ref. 2). It is also affected by the accident analysis 1 2 U

U of the SLB events presented in the FSAR, Section [ 15.1.5] (Ref. 3). The design precludes the blowdown of more than one steam generator, 14.2.5 assuming a single active component failure (e.g., the failure of one MSIV to close on demand).

The limiting case for the containment analysis is the SLB inside containment, with a loss of offsite power following turbine trip, and failure of the MSIV on the affected steam generator to close. At lower powers, the steam generator inventory and temperature are at their maximum, maximizing the analyzed mass and energy release to the containment.

Due to reverse flow and failure of the MSIV to close, the additional mass and energy in the steam headers downstream from the other MSIV contribute to the total release. With the most reactive rod cluster control assembly assumed stuck in the fully withdrawn position, there is an increased possibility that the core will become critical and return to power.

The core is ultimately shut down by the boric acid inj ection delivered by the Emergency Core Cooling System.

Turk ey Point Unit 3 and Unit 4 Revision XXX Westinghouse STS B 3.7.2-1 Rev. 5.0 1

MSIVs B 3.7.2 BASES APPLICABLE SAFETY ANALYSES (continued)

The accident analysis compares several different SLB events against different acceptance criteria. The large SLB outside containment upstream of the MSIV is limiting for offsite dose, although a break in this short section of main steam header has a very low probability. The large SLB inside containment at hot zero power is the limiting case for a post trip return to power. The analysis includes scenarios with offsite power available, and with a loss of offsite power following turbine trip. With 1

a main steam check offsite power available, the reactor coolant pumps continue to circulate valve (MSCV), a main coolant through the steam generators, maximizing the Reactor Coolant feedwater control valve (FCV), an System cooldown. With a loss of offsite power, the response of mitigating auxiliary feedwater systems is delayed. Significant single failures considered include failure pump flow control of an MSIV to close.

valve, and loss of a containment safeguards train due to The MSIVs serve only a safety function and remain open during power EDG sequencer failure operation. These valves operate under the following situations:

main steam a. An HELB inside containment. In order to maximize the mass and check valve 1 (MSCV),

energy release into containment, the analysis assumes that the MSIV in the affected steam generator remains open. For this accident scenario, steam is discharged into containment from all steam generators until the remaining MSIVs close. After MSIV closure, steam is discharged into containment only from the affected steam generator and from the residual steam in the main steam header downstream of the closed MSIVs in the unaffected loops. Closure of the MSIVs isolates the break from the unaffected steam generators.

b. A break outside of containment and upstream from the MSIVs is not a containment pressurization concern. The uncontrolled blowdown of more than one steam generator must be prevented to limit the potential for uncontrolled RCS cooldown and positive reactivity addition. Closure of the MSIVs isolates the break and limits the blowdown to a single steam generator.

c. A break downstream of the MSIVs will be isolated by the closure of the MSIVs.

d. Following a steam generator tube rupture, closure of the MSIVs isolates the ruptured steam generator from the intact steam generators to minimize radiological releases.

e. The MSIVs are also utilized during other events such as a feedwater line break . This event is less limiting so far as MSIV OPERABILITY is concerned.

The MSIVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

Turk ey Point Unit 3 and Unit 4 Revision XXX Westinghouse STS B 3.7.2-2 Rev. 5.0 1

MSIVs B 3.7.2 BASES three LCO This LCO requires that [four] MSIVs in the steam lines be OPERABLE. 1 The MSIVs are considered OPERABLE when the isolation times are While the Main Steam Bypass Valves within limits, and they close on an isolation actuation signal.

(MSBV) and Main Steam Check Valves (MSCV) support the Main Steam This LCO provides assurance that the MSIVs will perform their design Isolation function, no Technical Specification Limiting Condition for safety function to mitigate the consequences of accidents that could result Operation or Action applies to them. in offsite exposures comparable to the 10 CFR 100 (Ref. 4) limits or the 1 NRC staff approved licensing basis. 10 CFR 50.67 APPLICABILITY The MSIVs must be OPERABLE in MODE 1, and in MODES 2 and 3 except when closed and de-activated, when there is significant mass and energy in the RCS and steam generators. When the MSIVs are closed, they are already performing the safety function.

In MODE 4, normally most of the MSIVs are closed, and the steam generator energy is low.

In MODE 5 or 6, the steam generators do not contain much energy because their temperature is below the boiling point of water; therefore, the MSIVs are not required for isolation of potential high energy secondary system pipe breaks in these MODES.

ACTIONS A.1 24 l R2 With one MSIV inoperable in MODE 1, action must be taken to restore OPERABLE status within [8] hours [or in accordance with the Risk l R2 Informed Completion Time Program]. Some repairs to the MSIV can be 2 made with the unit hot. The [8] hour Completion Time is reasonable, l R2 considering the low probability of an accident occurring during this time period that would require a closure of the MSIVs.

The [8] hour Completion Time is greater than that normally allowed for 2 l R2 containment isolation valves because the MSIVs are valves that isolate a closed system penetrating containment. These valves differ from other containment isolation valves in that the closed system provides an additional means for containment isolation.

B.1 If the MSIV cannot be restored to OPERABLE status within [8] hours, the 2 l R2 unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and Condition C would be entered. The Completion Times are reasonable, based on operating experience, to reach MODE 2 and to close the MSIVs in an orderly manner and without challenging unit systems.

Turkey Point Unit 3 and Unit 4 Revision XXX Westinghouse STS B 3.7.2-3 Rev. 5.0 1

MSIVs B 3.7.2 BASES ACTIONS (continued)

C.1 and C.2 Condition C is modified by a Note indicating that separate Condition entry is allowed for each MSIV.

Since the MSIVs are required to be OPERABLE in MODES 2 and 3, the inoperable MSIVs may either be restored to OPERABLE status or closed.

When closed, the MSIVs are already in the position required by the assumptions in the safety analysis.

The [8] hour Completion Time is consistent with that allowed in 1 l R2 l

Condition A.

For inoperable MSIVs that cannot be restored to OPERABLE status within the specified Completion Time, but are closed, the inoperable MSIVs must be verified on a periodic basis to be closed. This is necessary to ensure that the assumptions in the safety analysis remain valid. The 7 day Completion Time is reasonable, based on engineering judgment, in view of MSIV status indications available in the control room, and other administrative controls, to ensure that these valves are in the closed position.

D.1 and D.2 If the MSIVs cannot be restored to OPERABLE status or are not closed within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed at least in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from MODE 2 conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.2.1 REQUIREMENTS This SR verifies that the closure time of each MSIV is within the limit given in Reference 5 and is within that assumed in the accident and containment analyses. This SR also verifies the valve closure time is in accordance with the INSERVICE TESTING PROGRAM. This SR is normally performed upon returning the unit to operation following a refueling outage. The MSIVs should not be tested at power, since even a part stroke exercise increases the risk of a valve closure when the unit is generating power. As the MSIVs are not tested at power, they are exempt from the ASME Code (Ref. 6), requirements during operation in MODE 1 or 2.

Turkey Point Unit 3 and Unit 4 Revision XXX Westinghouse STS B 3.7.2-4 Rev. 5.0 1

MSIVs B 3.7.2 BASES SURVEILLANCE REQUIREMENTS (continued)

The Frequency is in accordance with the INSERVICE TESTING PROGRAM.

This test is conducted in MODE 3 with the unit at operating temperature and pressure. This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. This allows a delay of testing until MODE 3, to establish conditions consistent with those under which the acceptance criterion was generated.

SR 3.7.2.2 This SR verifies that each MSIV can close on an actual or simulated actuation signal. This Surveillance is normally performed upon returning the plant to operation following a refueling outage. [ The Frequency of 2 MSIV testing is every [18] months. The [18] month Frequency for testing is based on the refueling cycle. Operating experience has shown that these components usually pass the Surveillance when performed at the

[18] month Frequency. Therefore, this Frequency is acceptable from a reliability standpoint.

OR The Surveillance Frequency is controlled under the Surveillance l R2 Frequency Control Program. l

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency 3 description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

10.2 REFERENCES

1. FSAR, Section [10.3].

U 14.3.4

2. FSAR, Section [6.2]. 1 2 U

14.2.5

3. FSAR, Section [15.1.5].

U 10 CFR 50.67 1

4. 10 CFR 100.11.

Inservice Testing Program 2

5. [Technical Requirements Manual.]

6. ASME Code for Operation and Maintenance of Nuclear Power Plants.

Turkey Point Unit 3 and Unit 4 Revision XXX Westinghouse STS B 3.7.2-5 Rev. 5.0 1

JUSTIFICATION FOR DEVIATIONS ITS 3.7.2 BASES, MAIN STEAM ISOLATION VALVES (MSIVs)

1. Changes are made (additions, deletions, and/or changes) to the Improved Standard Technical Specification (ISTS) Bases that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS Bases contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

3. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.7.2, MAIN STEAM ISOLATION VALVES (MSIVs)

There are no specific No Significant Hazards Considerations for this Specification.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

ATTACHMENT 3 ITS 3.7.3, FEEDWATER ISOLATION VALVES (FIVS) AND FEEDWATER CONTROL VALVES (FRVS) AND ASSOCIATED BYPASS VALVES

Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

A01 ITS ITS 3.7.3 PLANT SYSTEMS 3/4.7.1.7 FEEDWATER ISOLATION LIMITING CONDITION FOR OPERATION LCO 3.7.3 3.7.1.7 Six Feedwater Control Valves (FCVs) both main and bypass and six Feedwater Isolation Valves (FIVs) both main and bypass shall be OPERABLE.*

A03 Applicability APPLICABILITY: MODES 1, 2 and 3**

except when all FIVs, and FCVs are closed and de-activated or isolated by a manual valve. L01 ACTION:

ACTION B a. With one or more FCVs inoperable, restore operability, or close or isolate the inoperable FCVs A02 within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and verify that the inoperable valve(s) is closed or isolated at least once per 7 days 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 in HOT SHUTDOWN within the ACTION E following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

A02 ACTION A b. With one or more FIVs inoperable, restore operability, or close or isolate the inoperable FIV(s) within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and verify that the inoperable valve(s) is closed or isolated at least once per 7 days 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 in HOT SHUTDOWN within the ACTION E following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

A04 ACTION C c. With one or more bypass valves in different steam generator flow paths inoperable, restore A02 operability, or close or isolate the inoperable bypass valve(s) within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and verify that the inoperable valve(s) is closed or isolated at least once per 7 days or be in HOT STANDBY within ACTION E 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 .

ACTION D d. With two valves in the same steam generator flow paths inoperable, restore operability, or isolate A02 the affected flowpath within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 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 in HOT ACTION E SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.7.1.7 Each FCV, FIV and bypass valve shall be demonstrated OPERABLE:

SR 3.7.3.2 a. In accordance with the Surveillance Frequency Control Program by:

1) Verifying that each FCV, FIV and bypass valve actuates to the isolation position on an actual or simulated actuation signal.

SR 3.7.3.1 b. In accordance with the INSERVICE TESTING PROGRAM by:

1) Verifying that each FCV, FIV and bypass valve isolation time is within limits.

ACTIONS Note

Separate Condition entry is allowed for each valve.

SR 3.7.3.1 and **The provisions of specification 4.0.4 are not applicable. A03 SR 3.7.3.2 Note Only required to be performed in MODES 1 and 2.

TURKEY POINT - UNITS 3 & 4 3/4 7-13 AMENDMENT NOS. 278 AND 273 Page 1 of 1

DISCUSSION OF CHANGES ITS 3.7.3, FEEDWATER ISOLATION VALVES (FIVs) AND FEEDWATER CONTROL VALVES (FCVs) AND ASSOCIATED BYPASS VALVES ADMINISTRATIVE CHANGES A01 In the conversion of the Turkey Point Nuclear Generating Station (PTN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 5.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 3.7.1.7 ACTION a states, in part, to either restore OPERABILITY, or close or isolate the inoperable Feedwater Control Valve (FCV) within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . CTS 3.7.1.7 ACTION b states, in part, to either restore operability, or close or isolate the inoperable Feedwater Isolation Valve (FIV) within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . CTS 3.7.1.7 ACTION c states, in part, to either restore operability, or close or isolate the inoperable bypass valve(s) within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . CTS 3.7.1.7 ACTION d states, in part, to either restore operability, or isolate the affected flow path within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

ITS 3.7.3 ACTIONS A, B, C and D do not contain the statement to restore an inoperable FIV, FCV, bypass valve, or affected flow path to OPERABLE status.

This changes the CTS by not including the statement to restore an inoperable FIV, FCV, bypass valve, or affected flow path to OPERABLE status.

This change is acceptable because the technical requirements have not changed. Restoration to compliance with the Limiting Condition for Operation (LCO) is always an available Required Action and it is the convention in the ITS to not state "restore" options explicitly unless it is the only action or is required for clarity. This change is designated as administrative because it does not result in technical changes to the CTS.

A03 CTS 3.7.1.7 Applicability Footnote ** contains the provision that Specification 4.0.4 is not applicable for entry into MODE 3. CTS 4.0.4 states that entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified frequency, except as provided by Specification 4.0.3. A similar Note is added to ITS 3.7.3 that is associated with SR 3.7.3.1 and SR 3.7.3.2 stating these surveillances are only required to be performed in MODES 1 and 2. This changes the CTS by providing similar wording that the surveillances associated with ITS LCO 3.7.3 are not required to be met for entry into MODE 3.

The CTS Applicability footnote allows entry into MODE 3 without meeting the Surveillance Requirement (SR). This portion of the note allows continuous operation in MODE 3. Allowances provided by the CTS and ITS accomplish equivalent results; both allow entry into and operation in MODE 3. This change is designated as administrative because it does not result in a technical change to the CTS.

A04 CTS 3.7.1.7 Action c states, in part, With one or more bypass valves in different steam generator flow paths inoperable ITS 3.7.3 Condition C does not Turkey Point Unit 3 and Unit 4 Page 1 of 3

DISCUSSION OF CHANGES ITS 3.7.3, FEEDWATER ISOLATION VALVES (FIVs) AND FEEDWATER CONTROL VALVES (FCVs) AND ASSOCIATED BYPASS VALVES contain the phrase in different steam generator flow paths. This changes the CTS by not including the clarifying statement that the condition applies in different steam generator flow paths.

The purpose of CTS Action c is to ensure proper action is taken when one or more bypass valves associated with the FIVs and FCVs are inoperable. The clarifying phrase in different steam generator flow paths, is unnecessary because it is understood that bypass valves are in different steam generator flow paths, therefore, this information is not included in ITS 3.7.3 Condition C. This change is acceptable because the technical requirements have not changed.

This change is designated as administrative because it does not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01 (Category 2 - Relaxation of Applicability) CTS 3.7.1.7 requires three FIVs, three FCVs, three FIV bypass valves, or three FCV bypass valves to be OPERABLE in MODES 1, 2 and 3. ITS LCO 3.7.3 requires three FIVs, three FCVs, three FIV bypass valves, and three FCV bypass valves to be OPERABLE in MODES 1, 2 and 3 except when all FIVs, FCVs, and associated bypass valves are closed and de-activated or isolated by a closed manual valve. This changes the CTS by making the Specification not applicable in MODES 1, 2, and 3 when all FIVs, FCVs, and associated bypass valves are closed and de-activated or isolated by a closed manual valve.

The purpose of CTS 3.7.1.7 Applicability is to ensure that the FIVs, FCVs and associated bypass valves are OPERABLE and capable of closing when required to support the safety analysis. This change is acceptable because the requirements continue to ensure that the structures, systems, and components are maintained in the MODES and other specified conditions assumed in the safety analyses and licensing basis. When the FIVs, FCVs, and associated bypass valves are in the closed position or isolated, the valves are in the assumed accident position. This change is designated as less restrictive Turkey Point Unit 3 and Unit 4 Page 2 of 3

DISCUSSION OF CHANGES ITS 3.7.3, FEEDWATER ISOLATION VALVES (FIVs) AND FEEDWATER CONTROL VALVES (FCVs) AND ASSOCIATED BYPASS VALVES because the LCO requirements are applicable in fewer operating conditions than in the CTS.

Turkey Point Unit 3 and Unit 4 Page 3 of 3

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

MFIVs and MFRVs and [ Associated Bypass Valves] 2 1 CTS C 3.7.3 3.7 PLANT SYSTEMS 3.7.3 Main Feedwater Isolation Valves (MFIVs) and Main Feedwater Control Valves 2 1 (MFRVs) and [ Associated Bypass Valves]

C Three three 3.7.1.7 LCO 3.7.3 [ Four] MFIVs, [ four] MFRVs, [ and associated bypass valves] shall be 1 2 OPERABLE. C C

Applicability APPLICABILITY: MODES 1, [ and 2] [ 2, and 3] except when MFIV, MFRV, [ or associated bypass valve] is closed and [ de-activated] [ or isolated by a closed 1 manual valve] .

ACTIONS

NOTE-----------------------------------------------------------

LCO 3.7.1.7 Separate Condition entry is allowed for each valve.

Note

CONDITION REQUIRED ACTION COMPLETION TIME ACTION b A. One or more MFIVs A.1 Close or isolate MFIV. [ 72] hours 2 1 inoperable.

AND A.2 Verify MFIV is closed or Once per 7 days isolated.

C C ACTION a B. One or more MFRVs B.1 Close or isolate MFRV. [ 72] hours 2 1 inoperable.

AND C

B.2 Verify MFRV is closed or Once per 7 days 2 isolated.

ACTION c C. [ One or more [ MFRV or C.1 Close or isolate bypass [ 72] hours preheater] bypass valve.

valves inoperable.

AND C.2 Verify bypass valve is Once per 7 days ]

closed or isolated.

Westinghouse STS 3.7.3-1 Rev. 5.0 2 Turk ey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY

MFIVs and MFRVs and [ Associated Bypass Valves] 2 1 CTS C 3.7.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME ACTION d D. Two valves in the same D.1 Isolate affected flow path. 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months flow path inoperable.

ACTION a, E. Required Action and E.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ACTION b, ACTION c associated Completion Time not met. [ AND 1

E.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ]

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY C

4.7.1.7 b.1 SR 3.7.3.1 Verify the isolation time of each MFIV, MFRV[ , and In accordance 2 1 associated bypass valve] is within limits. with the

NOTE--------------------------

INSERVICE Only required to be performed in MODES 1 and 2. TESTING 3

PROGRAM C

4.7.1.7 a.1 SR 3.7.3.2 Verify each MFIV, MFRV[ , and associated bypass [ [ 18] months 2 1 valves] actuates to the isolation position on an actual or simulated actuation signal. OR

NOTE-------------------------

Only required to be performed in MODES 1 and 2. In accordance 3

with the Surveillance Frequency Control Program ]

Westinghouse STS 3.7.3-2 Rev. 5.0 2 Turk ey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY

JUSTIFICATION FOR DEVIATIONS ITS 3.7.3, FEEDWATER ISOLATION VALVES (FIVs), FEEDWATER CONTROL VALVES (FCVs) AND ASSOCIATED BYPASS VALVES

1. The Improved Standard Technical Specification (ISTS) contain bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

2. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

3. The Improved Technical Specifications (ITS) is adopting the Turkey Point Nuclear Generating Station (PTN) current licensing basis allowance to enter MODE 3 without having to perform the applicable Surveillance Requirements (SRs). Instead of incorporating an exception in the Applicability as with the PTN Current Technical Specifications (CTS), a Note will be added to the SRs that would allow entry into MODE 3 with the SR unperformed. This exception is required to ensure the proper test environment is available to perform the test.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

MFIVs and MFRVs [ and Associated Bypass Valves] 1 2 C

B 3.7.3 B 3.7 PLANT SYSTEMS B 3.7.3 Main Feedwater Isolation Valves (MFIVs) and Main Feedwater Regulation Valves 1 2 (MFRVs) [ and Associated Bypass Valves] Control C

BASES C

BACKGROUND The MFIVs isolate main feedwater (MFW) flow to the secondary side of the steam generators following a high energy line break (HELB). The I

safety related function of the MFRVs is to provide the second isolation of MFW flow to the secondary side of the steam generators following an INSERT 1 HELB. Closure of the MFIVs and associated bypass valves or MFRVs C

and associated bypass valves terminates flow to the steam generators, terminating the event for feedwater line break s (FWLBs) occurring C

upstream of the MFIVs or MFRVs. The consequences of events occurring in the main steam lines or in the MFW lines downstream from the MFIVs will be mitigated by their closure. Closure of the MFIVs and C

associated bypass valves, or MFRVs and associated bypass valves, effectively terminates the addition of feedwater to an affected steam generator, limiting the mass and energy release for steam line break s (SLBs) or FWLBs inside containment, and reducing the cooldown effects for SLBs.

C The MFIVs and associated bypass valves, or MFRVs and associated bypass valves, isolate the nonsafety related portions from the safety related portions of the system. In the event of a secondary side pipe 1 rupture inside containment, the valves limit the quantity of high energy fluid that enters containment through the break , and provide a pressure boundary for the controlled addition of auxiliary feedwater (AFW) to the intact loops.

C One MFIV and associated bypass valve, and one MFRV and its associated bypass valve, are located on each MFW line, outside but C

close to containment. The MFIVs and MFRVs are located upstream of the AFW inj ection point so that AFW may be supplied to the steam C

generators following MFIV or MFRV closure. The piping volume from these valves to the steam generators must be accounted for in calculating mass and energy releases, and refilled prior to AFW reaching the steam generator following either an SLB or FWLB.

C The MFIVs and associated bypass valves, and MFRVs and associated

, safety bypass valves, close on receipt of a Tavg - Low coincident with reactor inj ection, trip (P-4) or steam generator water level - high high signal. They may The FCVs also also be actuated manually. In addition to the MFIVs and associated close on C

bypass valves, and the MFRVs and associated bypass valves, a check receipt of a Tavg - Low valve inside containment is available. The check valve isolates the coincident with reactor trip.

feedwater line, penetrating containment, and ensures that the consequences of events do not exceed the capacity of the containment heat removal systems.

Westinghouse STS B 3.7.3-1 Rev. 5.0 Turk ey Point Unit 3 and Unit 4 Revision XXX

1 INSERT 1 The LCO requires three FCVs, three FIVs, three bypass line control valves, and three bypass line isolation valves to be OPERABLE.

Insert Page B 3.7.3-1

MFIVs and MFRVs [and Associated Bypass Valves] 1 2 C

B 3.7.3 BASES BACKGROUND (continued) U C

A description of the MFIVs and MFRVs is found in the FSAR, 10.2 Section [10.4.7] (Ref. 1).

2 APPLICABLE The design basis of the MFIVs and MFRVs is established by the C

SAFETY analyses for the large SLB. It is also influenced by the accident analysis ANALYSES for the large FWLB. Closure of the MFIVs and associated bypass valves, or MFRVs and associated bypass valves, may also be relied on to C

terminate an SLB for core response analysis and excess feedwater event upon the receipt of a steam generator water level - high high signal or a feedwater isolation signal on high steam generator level.

Feedwater isolation C

Failure of an MFIV, MFRV, or the associated bypass valves to close valves do not have a following an SLB or FWLB can result in additional mass and energy being credited function in the Feedwater Line Break delivered to the steam generators, contributing to cooldown. This failure analysis. also results in additional mass and energy releases following an SLB or FWLB event.

C The MFIVs and MFRVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

C 1 l R2 LCO This LCO ensures that the MFIVs, MFRVs, and their associated bypass valves will isolate MFW flow to the steam generators, following an FWLB or main steam line break. These valves will also isolate the nonsafety related portions from the safety related portions of the system.

three three This LCO requires that [four] MFIVs and associated bypass valves and l R2 C

[four] MFRVs [and associated bypass valves] be OPERABLE. The 2 C

MFIVs and MFRVs and the associated bypass valves are considered OPERABLE when isolation times are within limits and they close on an isolation actuation signal.

Failure to meet the LCO requirements can result in additional mass and energy being released to containment following an SLB or FWLB inside containment. If a feedwater isolation signal on high steam generator level is relied on to terminate an excess feedwater flow event, failure to meet the LCO may result in the introduction of water into the main steam lines.

APPLICABILITY C The MFIVs and MFRVs and the associated bypass valves must be OPERABLE whenever there is significant mass and energy in the Reactor Coolant System and steam generators. This ensures that, in the event of an HELB, a single failure cannot result in the blowdown of more than one steam generator. In MODES 1, 2, [and 3], the MFIVs and 2 C

MFRVs and the associated bypass valves are required to be OPERABLE to limit the amount of available fluid that could be added to containment in the case of a secondary system pipe break inside containment. When the valves are closed and de-activated or isolated by a closed manual valve, they are already performing their safety function.

Westinghouse STS B 3.7.3-2 Rev. 5.0 Turkey Point Unit 3 and Unit 4 Revision XXX

MFIVs and MFRVs [ and Associated Bypass Valves] 1 2 C

B 3.7.3 BASES APPLICABILITY (continued)

In MODES 4, 5, and 6, steam generator energy is low. Therefore, the MFIVs, MFRVs, and the associated bypass valves are normally closed 1 C

since MFW is not required.

ACTIONS The ACTIONS Table is modified by a Note indicating that separate Condition entry is allowed for each valve.

A.1 and A.2 With one MFIV in one or more flow paths inoperable, action must be 1 tak en to restore the affected valves to OPERABLE status, or to close or isolate inoperable affected valves within [ 72] hours. When these valves 2 are closed or isolated, they are performing their required safety function.

The [ 72] hour Completion Time tak es into account the redundancy 2 afforded by the remaining OPERABLE valves and the low probability of an event occurring during this time period that would require isolation of the MFW flow paths. The [ 72] hour Completion Time is reasonable, 2 based on operating experience.

Inoperable MFIVs that are closed or isolated must be verified on a 1 periodic basis that they are closed or isolated. This is necessary to ensure that the assumptions in the safety analysis remain valid. The 7 day Completion Time is reasonable, based on engineering j udgment, in view of valve status indications available in the control room, and other administrative controls, to ensure that these valves are closed or isolated.

B.1 and B.2 C

With one MFRV in one or more flow paths inoperable, action must be 1 tak en to restore the affected valves to OPERABLE status, or to close or isolate inoperable affected valves within [ 72] hours. When these valves 2 are closed or isolated, they are performing their required safety function.

The [ 72] hour Completion Time tak es into account the redundancy 2 afforded by the remaining OPERABLE valves and the low probability of an event occurring during this time period that would require isolation of the MFW flow paths. The [ 72] hour Completion Time is reasonable, 2 based on operating experience.

Westinghouse STS B 3.7.3-3 Rev. 5.0 Turk ey Point Unit 3 and Unit 4 Revision XXX

MFIVs and MFRVs [ and Associated Bypass Valves] 1 2 C

B 3.7.3 BASES ACTIONS (continued)

C Inoperable MFRVs, that are closed or isolated, must be verified on a 1 periodic basis that they are closed or isolated. This is necessary to ensure that the assumptions in the safety analysis remain valid. The 7 day Completion Time is reasonable, based on engineering j udgment, in view of valve status indications available in the control room, and other administrative controls to ensure that the valves are closed or isolated.

C.1 and C.2 With one associated bypass valve in one or more flow paths inoperable, action must be tak en to restore the affected valves to OPERABLE status, or to close or isolate inoperable affected valves within [ 72] hours. When these valves are closed or isolated, they are performing their required safety function.

The [ 72] hour Completion Time tak es into account the redundancy afforded by the remaining OPERABLE valves and the low probability of 1 an event occurring during this time period that would require isolation of the MFW flow paths. The [ 72] hour Completion Time is reasonable, based on operating experience.

Inoperable associated bypass valves that are closed or isolated must be verified on a periodic basis that they are closed or isolated. This is necessary to ensure that the assumptions in the safety analysis remain valid. The 7 day Completion Time is reasonable, based on engineering j udgment, in view of valve status indications available in the control room, and other administrative controls, to ensure that these valves are closed or isolated.

D.1 1 With two inoperable valves in the same flow path, there may be no redundant system to operate automatically and perform the required safety function. Although the containment can be isolated with the failure of two valves in parallel in the same flow path, the double failure can be an indication of a common mode failure in the valves of this flow path, and as such, is treated the same as a loss of the isolation capability of this flow path. Under these conditions, affected valves in each flow path must be restored to OPERABLE status, or the affected flow path isolated within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . This action returns the system to the condition where at least one valve in each flow path is performing the required safety function. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is reasonable, based on operating C

experience, to complete the actions required to close the MFIV or MFRV, 1 or otherwise isolate the affected flow path.

Westinghouse STS B 3.7.3-4 Rev. 5.0 Turk ey Point Unit 3 and Unit 4 Revision XXX

MFIVs and MFRVs [ and Associated Bypass Valves] 1 2 C

B 3.7.3 BASES ACTIONS (continued) 1 E.1 and E.2 If the MFIV(s) and MFRV(s) and the associated bypass valve(s) cannot 1 C

be restored to OPERABLE status, or closed, or isolated within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months [ , and in MODE 4 within 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s] . The 2 allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.3.1 REQUIREMENTS C

This SR verifies that the closure time of each MFIV, MFRV, and 1

[ associated bypass valve] is within the limit given in Reference 2 and is 2 within that assumed in the accident and containment analyses. This SR also verifies the valve closure time is in accordance with the INSERVICE TESTING PROGRAM. This SR is normally performed upon returning the This SR is modified by a Note that allows entry into and operation in unit to operation following a refueling outage. These valves should not be MODE 3 prior to the surveillance tested at power since even a part strok e exercise increases the risk of a being performed. This allows a valve closure with the unit generating power. This is consistent with the delay of testing until MODE 3, to establish conditions consistent with ASME Code (Ref. 3), quarterly strok e requirements during operation in those under which the acceptance MODES 1 and 2.

criterion was generated.

4 The Frequency for this SR is in accordance with the INSERVICE TESTING PROGRAM.

SR 3.7.3.2 C

This SR verifies that each MFIV, MFRV, and [ associated bypass valves] 1 2 This SR is modified by a Note that can close on an actual or simulated actuation signal. This Surveillance is allows entry into and operation in MODE 3 prior to the surveillance normally performed upon returning the plant to operation following a being performed. This allows a refueling outage.

delay of testing until MODE 3, to establish conditions consistent with 2 those under which the acceptance [ The Frequency for this SR is every [ 18] months. The [ 18] month criterion was generated. Frequency for testing is based on the refueling cycle. Operating experience has shown that these components usually pass the Surveillance when performed at the [ 18] month Frequency. Therefore, this Frequency is acceptable from a reliability standpoint.

OR Westinghouse STS B 3.7.3-5 Rev. 5.0 Turk ey Point Unit 3 and Unit 4 Revision XXX

MFIVs and MFRVs [and Associated Bypass Valves] 1 2 C

B 3.7.3 BASES SURVEILLANCE REQUIREMENTS (continued)

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency 3 description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

U REFERENCES 1. FSAR, Section [10.4.7]. 1 2 10.2

2. [Technical Requirements Manual.] 2 Inservice Testing Program l R2

3. ASME Code for Operation and Maintenance of Nuclear Power Plants.

Westinghouse STS B 3.7.3-6 Rev. 5.0 Turkey Point Unit 3 and Unit 4 Revision XXX

JUSTIFICATION FOR DEVIATIONS ITS 3.7.3 BASES, FEEDWATER ISOLATION VALVES (FIVs) AND FEEDWATER CONTROL VALVES (FCVs) AND ASSOCIATED BYPASS VALVES

1. Changes are made (additions, deletions, and/or changes) to the Improved Standard Technical Specification (ISTS) Bases that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS Bases contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

3. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.

4. Changes are being made to the Improved Technical Specifications (ITS) Bases to be consistent with changes made to the ITS.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.7.3, FEEDWATER ISOLATION VALVES (FIVs) AND FEEDWATER CONTROL VALVES (FCVs) AND ASSOCIATED BYPASS VALVES There are no specific No Significant Hazards Considerations for this Specification.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

ATTACHMENT 4 ITS 3.7.4, SECONDARY SPECIFIC ACTIVITY

Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

A01 ITS 3.7.4 ITS PLANT SYSTEMS A01 SECONDARY SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION LCO 3.7.4 3.7.1.4 The specific activity of the Secondary Coolant System shall be less than or equal to 0.10 microCurie/gram DOSE EQUIVALENT I-131.

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

ACTION:

ACTION A With the specific activity of the Secondary Coolant System greater than 0.10 microCurie/gram DOSE EQUIVALENT I-131, 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS SR 3.7.4.1 4.7.1.4 The specific activity of the Secondary Coolant System shall be determined to be within the limit by performing the sampling and analysis described in Table 4.7-1.

A02 TURKEY POINT - UNITS 3 & 4 3/4 7-8 AMENDMENT NOS. 263 AND 258 Page 1 of 2

A01 ITS 3.7.4 ITS TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS L01 TYPE OF MEASUREMENT SAMPLE AND ANALYSIS AND ANALYSIS FREQUENCY LA01

1. Gross Radioactivity SFCP Determination

2. Isotopic Analysis for DOSE a) Once per 31 days, when-EQUIVALENT I-131 Concentration ever the gross radio-activity determination M01 indicates concentrations greater than 10% of the allowable limit for radioiodines.

b) Once per 6 months, when-ever the gross radio-activity determination indicates concentrations less than or equal to 10%

of the allowable limit for radioiodines.

in accordance with the Surveillance LA02 Frequency Control Program.

TURKEY POINT - UNITS 3 & 4 3/4 7-9 AMENDMENT NOS. 263 AND 258 Page 2 of 2

DISCUSSION OF CHANGES ITS 3.7.4, SECONDARY SPECIFIC ACTIVITY ADMINISTRATIVE CHANGES A01 In the conversion of the Turkey Point Nuclear Generating Station (PTN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 5.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 4.7.1.4 requires the specific activity of the secondary coolant system to be verified within limit in accordance with Table 4.7-1. CTS Table 4.7-1 provides the secondary coolant system sample analysis measurement and frequency requirements. ITS Surveillance Requirement (SR) 3.7.4.1 requires verification that the secondary coolant specific activity is less than or equal to 0.10 microcuries per gram DOSE EQUIVALENT I-131 at a Frequency that is in accordance with the Surveillance Frequency Control Program. This changes the CTS by moving the secondary coolant system sample analysis measurement and frequency requirements from a table and placing the information within the Surveillance Requirement.

The purpose of CTS 4.7.1.4 and Table 4.7-1 is to conduct a specific activity analysis on the secondary coolant system to monitor primary to secondary leakage and confirm secondary activity is within the initial assumptions of the accident analyses. This change is acceptable because moving the secondary coolant sample analysis measurement and frequency requirements from a table to within the Surveillance Requirement, ITS SR 3.7.4.1, centralizes the requirements. This change is designated as administrative, because it does not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS Table 4.7-1, Item 2, requires the secondary coolant system DOSE EQUIVALENT I-131 sampling frequency to be once per 31 days, whenever the gross activity determination indicates iodine concentration greater than 10% of the allowable limit. CTS Table 4.7-1, Item 2, allows the sampling frequency for the secondary coolant system DOSE EQUIVALENT I131 to be extended to once per 6 months, whenever the gross activity determination indicates iodine concentrations below 10% of the allowable limits. ITS SR 3.7.4.1 does not provide this extended period for determining secondary coolant system DOSE EQUIVALENT I-131 and requires verification of secondary coolant system specific activity every 31 days. This changes the CTS by deleting the allowance to perform less frequent secondary coolant system specific activity analyses when gross activity determination indicates iodine concentrations below 10% of the allowable limits.

The purpose of CTS 4.7.1.4 and Table 4.7-1 is to conduct a specific activity analysis on the secondary coolant system to monitor primary to secondary Turkey Point Unit 3 and Unit 4 Page 1 of 3

DISCUSSION OF CHANGES ITS 3.7.4, SECONDARY SPECIFIC ACTIVITY leakage and confirm secondary activity is within the initial assumptions of the accident analyses. This change is acceptable because the 31 day Frequency is appropriate to detect trends in the level of DOSE EQUIVALENT I-131 and allows appropriate action to be taken to maintain levels below the LCO limit. This change is designated as more restrictive, because Surveillances will be performed more frequently under the ITS than under the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS Table 4.7-1, Item 2 requires an isotopic analysis to determine whether DOSE EQUIVALENT I-131 concentration is within limit.

ITS SR 3.7.4.1 requires the verification that specific activity of the secondary coolant is within limit. This changes the CTS by moving the detail that an isotopic analysis must be performed to satisfy the requirements of the Surveillance to the Bases.

The removal of this detail for performing a Surveillance Requirement from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. ITS SR 3.7.4.1 retains the requirement to verify secondary coolant DOSE EQUIVALENT I-131 is within limit. In addition, this change is acceptable because this type of procedural detail will be adequately controlled in the ITS Bases. The Technical Specification Bases Control Program in Chapter 5 controls changes to the Bases. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications.

LA02 (Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program) CTS 4.7.1.4 requires that the specific activity of the secondary coolant system shall be determined to be within the limit by performance of the sampling and analysis program of Table 4.7-1. CTS Table 4.7-1, Item 2, lists a sample frequency of once per 31 days that can be extended if the gross activity determination indicates iodine concentration is below 10% of the allowable limit.

ITS SR 3.7.4.1 requires a similar Surveillance and specifies the periodic Frequency as, "In accordance with the Surveillance Frequency Control Program."

This changes the CTS by moving the specified Frequency for this SR and associated Bases (once per 31 days) to the Surveillance Frequency Control Program.

The removal of these details related to Surveillance Requirement Frequencies from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide Turkey Point Unit 3 and Unit 4 Page 2 of 3

DISCUSSION OF CHANGES ITS 3.7.4, SECONDARY SPECIFIC ACTIVITY adequate protection of public health and safety. The existing Surveillance Frequencies are removed from Technical Specifications and placed under licensee control pursuant to the methodology described in NEI 04-10. A new program (Surveillance Frequency Control Program) is being added to the Administrative Controls section of the Technical Specifications describing the control of Surveillance Frequencies. The surveillance test requirements remain in the Technical Specifications. The control of changes to the Surveillance Frequencies will be in accordance with the Surveillance Frequency Control Program. The Program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met. This change is designated as a less restrictive removal of detail change, because the Surveillance Frequency is being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 5 - Deletion of Surveillance Requirement) CTS Table 4.7-1, Item 1, require the secondary coolant system be tested for gross activity at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . ITS 3.7.4 does not contain this Surveillance Requirement. This changes the CTS by deleting this Surveillance Requirement.

The purpose of CTS Table 4.7-1, Item 1, is to sample the secondary coolant on a 72-hour basis and measure the gross activity. The result of this analysis is used to determine the frequency for sampling the secondary coolant for DOSE EQUIVALENT I-131. This change is acceptable because the deleted Surveillance Requirement is not necessary to verify that the values used to meet the LCO are consistent with the safety analyses because the gross beta or gamma activity of the secondary coolant is not used in the accident analyses.

The accident analyses use dose equivalent iodine I-131 (DEI-131). ITS SR 3.7.4.1 requires the DOSE EQUIVALENT I-131 to be determined in accordance with the Surveillance Frequency Control Program. Therefore, appropriate values (i.e., the secondary coolant DOSE EQUIVALENT I-131) continue to be tested in a manner and at a Frequency necessary to give confidence that the assumptions in the safety analyses are protected since the secondary coolant DOSE EQUIVALENT I-131 is used in the accident analyses. This change is designated as less restrictive because a Surveillance that is required in the CTS will not be required in the ITS.

Turkey Point Unit 3 and Unit 4 Page 3 of 3

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

CTS Secondary Specific Activity 3.7.18 1 4

3.7 PLANT SYSTEMS 4

1 3.7.18 Secondary Specific Activity 4

3.7.4.1 LCO 3.7.18 The specific activity of the secondary coolant shall be [ 0.10] µ Ci/gm 1 2 DOSE EQUIVALENT I-131.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME ACTION A. Specific activity not A.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months within limit.

AND A.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.7.1.4 SR 3.7.18.1 Verify the specific activity of the secondary coolant [ 31 days 1 4 is [ 0.10] µ Ci/gm DOSE EQUIVALENT I-131. 2 OR In accordance with the Surveillance Frequency Control Program ] 1 Turk ey Point Unit 3 and Unit 4 4 Amendment Nos. XXX and YYY Westinghouse STS 3.7.18-1 Rev. 5.0 3 1

ITS 3.7.4 JUSTIFICATION FOR DEVIATIONS ITS 3.7.4, SECONDARY SPECIFIC ACTIVITY

1. Improved Standard Technical Specification (ISTS) 3.7.18, "Secondary Specific Activity" has been renumbered as ITS 3.7.4, "Secondary Specific Activity."

2. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

3. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Secondary Specific Activity 4 1 B 3.7.18 B 3.7 PLANT SYSTEMS 4

B 3.7.18 Secondary Specific Activity 1 BASES BACKGROUND Activity in the secondary coolant results from steam generator tube outleak age from the Reactor Coolant System (RCS). Under steady state conditions, the activity is primarily iodines with relatively short half lives and, thus, indicates current conditions. During transients, I-131 spik es have been observed as well as increased releases of some noble gases.

Other fission product isotopes, as well as activated corrosion products in lesser amounts, may also be found in the secondary coolant.

A limit on secondary coolant specific activity during power operation minimizes releases to the environment because of normal operation, anticipated operational occurrences, and accidents.

This limit is lower than the activity value that might be expected from a DOSE EQUIVALENT 1 gpm tube leak (LCO 3.4.13, " RCS Operational LEAKAGE" ) of primary I-131 0.25 coolant at the limit of [ 1.0] µ Ci/gm (LCO 3.4.16, " RCS Specific Activity" ). 2 The steam line failure is assumed to result in the release of the noble gas and iodine activity contained in the steam generator inventory, the feedwater, and the reactor coolant LEAKAGE. Most of the iodine isotopes have short half lives (i.e., < 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months ).

With the specified activity limit, the resultant 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid dose to a person at the exclusion area boundary (EAB) would be about 0.58 rem if 3 the main steam safety valves (MSSVs) open for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following a trip from full power.

10 CFR 50.67 Operating a unit at the allowable limits could result in a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months EAB exposure of a small fraction of the 10 CFR 100 (Ref. 1) limits, or the limits 3 established as the NRC staff approved licensing basis.

INSERT 1 APPLICABLE 14 The accident analysis of the main steam line break (MSLB), as SAFETY discussed in the FSAR, Chapter [ 15] (Ref. 2) assumes the initial 3 2 U

ANALYSES secondary coolant specific activity to have a radioactive isotope concentration of [ 0.10] µ Ci/gm DOSE EQUIVALENT I-131. This 2 assumption is used in the analysis for determining the radiological consequences of the postulated accident. The accident analysis, based on this and other assumptions, shows that the radiological consequences

, SGTR, or REA, of an MSLB do not exceed a small fraction of the unit EAB limits (Ref. 1) 3 5 for whole body and thyroid dose rates. exclusion area boundary (EAB)

With the loss of offsite power, the remaining steam generators are available for core decay heat dissipation by venting steam to the 3 atmosphere through the MSSVs and steam generator atmospheric dump valves (ADVs). The Auxiliary Feedwater System supplies the necessary Turk ey Point Unit 3 and Unit 4 4 Revision XXX 3 1 Westinghouse STS B 3.7.18-1 Rev. 5.0

ITS 3.7.4 3

INSERT 1 Steam Generator Tube Rupture (SGTR), and Rod Ejection Accident (REA),

Insert Page B 3.7.4-1

Secondary Specific Activity 4 1 B 3.7.18 BASES APPLICABLE SAFETY ANALYSES (continued) mak eup to the steam generators. Venting continues until the reactor coolant temperature and pressure have decreased sufficiently for the 3 Residual Heat Removal System to complete the cooldown.

In the evaluation of the radiological consequences of this accident, the activity released from the steam generator connected to the failed steam line is assumed to be released directly to the environment. The unaffected steam generator is assumed to discharge steam and any 3

entrained activity through the MSSVs and ADVs during the event. Since no credit is tak en in the analysis for activity plateout or retention, the resultant radiological consequences represent a conservative estimate of the potential integrated dose due to the postulated steam line failure.

Secondary specific activity limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO As indicated in the Applicable Safety Analyses, the specific activity of the o oo t to [ 0.10] µ Ci/gm DOSE 2 EQUIVALENT I-131 to limit the radiological consequences of a Design Basis Accident (DBA) to a small fraction of the required limit (Ref. 1).

Monitoring the specific activity of the secondary coolant ensures that when secondary specific activity limits are exceeded, appropriate actions are tak en in a timely manner to place the unit in an operational MODE that would minimize the radiological consequences of a DBA.

APPLICABILITY In MODES 1, 2, 3, and 4, the limits on secondary specific activity apply due to the potential for secondary steam releases to the atmosphere.

In MODES 5 and 6, the steam generators are not being used for heat removal. Both the RCS and steam generators are depressurized, and primary to secondary LEAKAGE is minimal. Therefore, monitoring of secondary specific activity is not required.

ACTIONS A.1 and A.2 DOSE EQUIVALENT I-131 exceeding the allowable value in the secondary coolant, is an indication of a problem in the RCS and contributes to increased post accident doses. If the secondary specific activity cannot be restored to within limits within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

Turk ey Point Unit 3 and Unit 4 4 Revision XXX 3 1 Westinghouse STS B 3.7.18-2 Rev. 5.0

Secondary Specific Activity 4 1 B 3.7.18 BASES 4

SURVEILLANCE SR 3.7.18.1 1 REQUIREMENTS This SR verifies that the secondary specific activity is within the limits of the accident analysis. A gamma isotopic analysis of the secondary coolant, which determines DOSE EQUIVALENT I-131, confirms the validity of the safety analysis assumptions as to the source terms in post accident releases. It also serves to identify and trend any unusual isotopic concentrations that might indicate changes in reactor coolant activity or LEAKAGE. [ The 31 day Frequency is based on the detection 2 of increasing trends of the level of DOSE EQUIVALENT I-131, and allows for appropriate action to be taken to maintain levels below the LCO limit.

3 OR The Surveillance Frequency is controlled under the Surveillance l R2 Frequency Control Program. l

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency 4 description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

10 CFR 50.67 REFERENCES 1. 10 CFR 100.11. 3 14

2. FSAR, Chapter [15]. 1 2 U

Turkey Point Unit 3 and Unit 4 4 Revision XXX 3 1 Westinghouse STS B 3.7.18-3 Rev. 5.0

ITS 3.7.4 JUSTIFICATION FOR DEVIATIONS ITS 3.7.4 BASES, SECONDARY SPECIFIC ACTIVITY

1. Improved Standard Technical Specification (ISTS) 3.7.18, "Secondary Specific Activity," has been renumbered as ITS 3.7.4, "Secondary Specific Activity."

2. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

3. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

4. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.

5. Editorial/grammatical changes made.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.7.4, SECONDARY SPECIFIC ACTIVITY There are no specific No Significant Hazards Considerations for this Specification.

Turkey Point Unit 3 and 4 Page 1 of 1

ATTACHMENT 5 ITS 3.7.5, AUXILIARY FEEDWATER (AFW) SYSTEM

Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

ITS A01 ITS 3.7.5 PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LA01 LIMITING CONDITION FOR OPERATION 2 L01 LCO 3.7.5 3.7.1.2 Two independent auxiliary feedwater trains including 3 steam supply flowpaths, 3 pumps and associated discharge water flowpaths shall be OPERABLE.(1)(2) LA01 Applicability APPLICABILITY: MODES 1, 2 and 3 ACTION:

ACTIONS Note NOTE: LCO 3.0.4.b is not applicable to the required auxiliary feedwater trains when entering Mode 1.

ACTION B 1) With one of the two required independent auxiliary feedwater trains inoperable, either restore the inoperable train to an OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or in accordance with the Risk Informed Completion Time Program, or place the affected unit(s) in at least HOT STANDBY within the next ACTION D 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s* and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION C, 2) With both required auxiliary feedwater trains inoperable, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either restore both trains A02 ACTION G to an OPERABLE status, or restore one train to an OPERABLE status and follow ACTION statement 1 above for the other train. If neither train can be restored to an OPERABLE status ACTION D within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , verify that both Standby Feedwater Pumps are capable of providing makeup flow to the steam generators and place the affected unit(s) in at least HOT STANDBY within the next ACTION D 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s* and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Otherwise, initiate corrective action to restore at least one auxiliary feedwater train to an OPERABLE status as soon as A02 ACTION F l R2 possible and follow ACTION statement 1 above for the other train.

L01

3) With a single auxiliary feedwater pump inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , verify OPERABILITY of two independent auxiliary feedwater trains, or follow ACTION statements 1 or 2 above as applicable.

Upon verification of the OPERABILITY of two independent auxiliary feedwater trains, restore the inoperable auxiliary feedwater pump to an OPERABLE status within 30 days, or place the operating unit(s) in at least HOT STANDBY within 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s* and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

A02 ACTION A 4) With a single steam supply flowpath inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months verify OPERABILITY of two independent steam supply flowpaths or follow ACTION statement 1 or 2 above as applicable.

Upon verification of the OPERABILITY of two independent steam supply flowpaths, restore the inoperable steam supply flowpath to OPERABLE status within 7 days of discovery, or place the affected Unit(s) in at least HOT STANDBY within 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s* and in HOT SHUTDOWN within the ACTION D following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

NOTES:

(1)

One steam supply flowpath shall be OPERABLE in each AFW train and the third steam supply flowpath (via MOV-3-1404 for Unit 3 and MOV-4-1404 for Unit 4) shall be OPERABLE and aligned to either AFW train but not both simultaneously.

LA01 (2)

During single and two unit operation, one pump shall be OPERABLE in each train and the third auxiliary feedwater pump shall be OPERABLE and capable of being powered from, and supplying water to either train, except as noted in ACTION 3 of Technical Specification 3.7.1.2. The third auxiliary feedwater pump (normally the C pump) can be aligned to either train to restore OPERABILITY in the event one of the required pumps is inoperable.

ACTION E

If this ACTION applies to both units simultaneously, be in at least HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-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 .

TURKEY POINT - UNITS 3 & 4 3/4 7-3 AMENDMENT NOS. 284 AND 278 Page 1 of 3

ITS A01 ITS 3.7.5 PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM SURVEILLANCE REQUIREMENTS 4.7.1.2.1 The required independent auxiliary feedwater trains shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by:

LA02 SR 3.7.5.2 1) Verifying by control panel indication and visual observation of equipment that each steam turbine-driven pump operates for 15 minutes or greater and develops a flow of greater Only required to be than or equal to 373 gpm to the entrance of the steam generators. The provisions of l R2 SR 3.7.5.2 Note, performed in MODE 1. A03 l Specification 4.0.4 are not applicable for entry into MODES 2 and 3; SR 3.7.5.6 Note A04 SR 3.7.5.6 2) Verifying by control panel indication and visual observation of equipment that the auxiliary feedwater discharge valves and the steam supply and turbine pressure valves operate as l R2 required to deliver the required flow during the pump performance test above; manual and power operated A01 SR 3.7.5.1 3) Verifying that each non-automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in its correct position; and L02

4) Verifying that power is available to those components which require power for flow path operability.

b. In accordance with the Surveillance Frequency Control Program by:

SR 3.7.5.3 1) Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of each Auxiliary Feedwater Actuation test signal, and on an actual or simulated actuation L03 SR 3.7.5.4 2) Verifying that each auxiliary feedwater pump receives a start signal as designed automatically upon receipt of each Auxiliary Feedwater Actuation test signal.

Add proposed SR 3.7.5.4 Note L04 SR 3.7.5.5 4.7.1.2.2 An auxiliary feedwater flow path to each steam generator shall be demonstrated OPERABLE following each COLD SHUTDOWN of greater than 30 days prior to entering MODE 1 by verifying normal flow to each steam generator.

TURKEY POINT - UNITS 3 & 4 3/4 7-4 AMENDMENT NOS. 273 AND 268 Page 2 of 3

ITS A01 ITS 3.7.5 TABLE 3.7-3 AUXILIARY FEEDWATER SYSTEM OPERABILITY DELETED TURKEY POINT - UNITS 3 & 4 3/4 7-5 AMENDMENT NOS. 273 AND 268 Page 3 of 3

DISCUSSION OF CHANGES ITS 3.7.5, AUXILIARY FEEDWATER (AFW) SYSTEM ADMINISTRATIVE CHANGES A01 In the conversion of the Turkey Point Nuclear Generating Station (PTN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 5.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal.

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 3.7.1.2 Actions 2) and 4) provide directions to follow other Actions if specified conditions are not met. Action 2) states, in part, to either restore both trains to OPERABLE status or restore one train to OPERABLE status and follow ACTION statement 1 above for the other train. Action 4) states, in part, to verify OPERABILITY of two independent steam supply flowpaths or follow ACTION statement 1 or 2 above as applicable. ITS 3.7.5 does not contain these directions to follow other ACTION statements if specified conditions are not met.

This changes the CTS by not including direction to follow other Actions.

The purpose of these CTS 3.7.1.2 statements is to ensure all applicable Actions are entered based on the specific plant conditions at the time. ITS Section 1.3, Completion Times, states that if situations are discovered that require entry into more than one Condition at a time within a single LCO (multiple Conditions), the Required Actions for each Condition must be performed within the associated Completion Time. This is reiterated in ITS LCO 3.0.2 which states that upon discovery a failure to meet an LCO, the Required Actions of the associated Conditions [multiple Conditions] shall be met. Therefore, in ITS if two trains are inoperable there is also one train inoperable and both Conditions are entered or if a steam supply is inoperable and the remaining two are not independent the unsupported train is considered inoperable. This change is acceptable because the Required Actions specified in CTS will continue to be taken in ITS in accordance with the ITS rules of usage. This change is considered administrative because no technical changes are being made to the CTS.

A03 CTS 4.7.1.2.1.a.1 requires verifying each steam turbine-driven pump operates for l R2 15 minutes or greater and develops a flow of greater than or equal to 373 gpm. l CTS 4.7.1.2.1.a.2 requires verifying that the auxiliary feedwater (AFW) discharge l l

valves and the steam supply and turbine pressure valves operate as required to l

deliver the required flow during the pump performance test required by l CTS 4.7.1.2.1.a.1. CTS 4.7.1.2.1.a.1 states that the provisions of l Specification 4.0.4 are not applicable for entry into MODES 2 and 3. ITS l SR 3.7.5.2 (CTS 4.7.1.2.1.a.1) and SR 3.7.5.6 (CTS 4.7.1.2.1.a.2) require pump l flow testing and valve operational testing, respectively, and contain a similar Note l that states, "Only required to be performed in MODE 1." This changes the CTS l by revising the presentation of the CTS 4.0.4 exception. l Turkey Point Unit 3 and Unit 4 Page 1 of 7

DISCUSSION OF CHANGES ITS 3.7.5, AUXILIARY FEEDWATER (AFW) SYSTEM CTS 4.0.4 (ITS SR 3.0.4) requires the normal periodic Surveillances to be l R2 l

performed and be current prior to entry into the applicable MODES. CTS 3.7.1.2 l (ITS LCO 3.7.5) is Applicable in MODES 1, 2, and 3. The CTS 4.0.4 exception l allows entry into MODES 2 and 3 to perform CTS 4.7.1.2.1.a.1(ITS SR 3.7.5.2). l l

The CTS 4.0.4 exception also applies to CTS 4.7.1.2.1.a.2 (ITS SR 3.7.5.6) l because CTS 4.7.1.2.1.a.2 is required to be performed during performance of l l

CTS 4.7.1.2.1.a.1. The Note to ITS SRs 3.7.5.2 and 3.7.5.6 is functionally l equivalent to the CTS 4.0.4 exception of CTS 4.7.1.2.1.a.1 and 4.7.1.2.1.a.2 and l does not represent a technical change to the CTS. The ITS Note presentation l l

that only requires the SR to be performed in MODE 1 is explained in ITS l Section 1.4, Example 1.4-5, and states, in part, "this Note allows entry into and l operation in MODES 2 and 3 to perform the Surveillance." This change in l l

presentation is designated as administrative as it results in no technical change l to the CTS. l l

l A04 CTS 4.7.1.2.1.a.2 requires verifying the AFW System discharge valves and the l steam supply and turbine pressure valves operate as required to deliver the l l

required flow during the pump performance test specified in CTS 4.7.1.2.1.a.1. l ITS SR 3.7.5.6 similarly requires verifying the AFW pump discharge valves, l steam supply valves, and turbine valves operate as required to deliver the l l

required flow during performance of SR 3.7.5.2. This changes the CTS by l replacing "turbine pressure valves" with "turbine valves." l l

l The purpose of CTS 4.7.1.2.1.a.2 is to ensure the valves associated with the l steam supply to each required AFW pump turbine operate as required to ensure l l

the AFW pump can develop the proper flow. The original AFW pump design l consisted of a low pressure unit and included a turbine pressure reducing valve l in the AFW pump steam line to limit pressure to within the pressure rating of the l l

turbine casing. In 1980, an AFW pump modification upgraded the AFW pump l turbine casings allowing removal of the turbine pressure reducing valves. The l l

turbine pressure reducing valves were replaced with trip/throttle valves and l governor valves. This change removes a specific reference to turbine "pressure" l valves to avoid confusion on the requirement and is acceptable because ITS l l

SR 3.7.5.6 will continue to verify the steam supply and turbine valves operate as l required to ensure each required AFW pump can deliver the required flow. This l change is designated as an administrative change because it does not result in a l l

technical change to the CTS. l MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None Turkey Point Unit 3 and Unit 4 Page 2 of 7

DISCUSSION OF CHANGES ITS 3.7.5, AUXILIARY FEEDWATER (AFW) SYSTEM REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS LCO 3.7.1.2 requires two independent auxiliary feedwater trains including 3 steam supply flowpaths, 3 pumps (changed to 2 under DOC L01) and associated discharge water flowpaths shall be OPERABLE. The CTS LCO also contains footnotes further describing specifics that are required in each AFW train. ITS LCO 3.7.5 states "Two AFW trains and three steam generator steam l R2 supplies shall be OPERABLE." The ITS does not include design details, l components and associated flow paths that comprise an OPERABLE AFW train.

This changes the CTS by moving the description of the AFW independence, trains, and components required for OPERABILITY to the Bases.

The removal of these details, which are related to system design, from the Technical Specifications is acceptable because this type of information is not necessary to be included to provide adequate protection of public health and safety. The ITS retains all necessary requirements in the LCO to ensure OPERABILITY for the AFW trains and specific details, such as, system composition, independence, etc., are located in the Bases. Also, this change is acceptable because the removed information will be controlled in the ITS Bases.

Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program directs the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications.

LA02 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 4.7.1.2.1.a.1) and CTS 4.7.1.2.1.a.2) state, in part, to verify equipment performance, "by control panel indication and visual observation of equipment". ITS 3.7.5 does not include this guidance on how to verify equipment performed as required. This changes the CTS by removing procedural details for meeting TS requirements.

The removal of these details for performing surveillance requirements from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirements to verify that the steam turbine-driven pumps and the auxiliary feedwater valves operate as required. Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program directs the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications.

Turkey Point Unit 3 and Unit 4 Page 3 of 7

DISCUSSION OF CHANGES ITS 3.7.5, AUXILIARY FEEDWATER (AFW) SYSTEM LESS RESTRICTIVE CHANGES L01 (Category 1 - Relaxation of LCO Requirements) CTS 3.7.1.2 states, in part, that two independent auxiliary feedwater trains including 3 pumps shall be OPERABLE. CTS 3.7.1.2 Action 3) states, that with a single AFW pump inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to verify OPERABILITY of two independent AFW trains or follow Action 1 or Action 2 for 1 or 2 inoperable AFW train(s), as applicable.

ITS 3.7.5 states that two AFW trains shall be operable. This changes the CTS by reducing the number of AFW pumps required to be OPERABLE from three to two and deleting Action 3 for one of three AFW pumps inoperable.

The purpose of CTS 3.7.1.2 is to define the lowest functional capability or performance levels of equipment required for safe operation of the facility and the purpose of the CTS action requirement is to provide appropriate remedial actions when the LCO is not met pursuant to 10 CFR 50.36(c)(2)(i). This includes an occurrence which results in the loss of capability of a system to perform its intended safety functions due to the failure of an active component (active single failure). Required Actions are used to establish remedial measures that must be taken in response to degraded conditions in order to minimize risk associated with continued plant operation while providing time to repair inoperable equipment. This change is acceptable because the LCO requirements continue to ensure that structures, systems, and components are maintained consistent with the safety analyses and licensing basis and the required actions are consistent with safe operation under the specified Condition, considering the OPERABILITY status of the remaining AFW equipment, a reasonable time for repairs or replacement of required equipment, and the low probability of a design basis accident occurring during the time period.. The PTN AFW System consists of three turbine driven AFW pumps shared between the two units. The three AFW pumps discharge through check valves to one of two redundant discharge headers. The AFW System is normally aligned such that Pump A discharges to the Train 1 AFW header and Pumps B and C normally discharge to the Train 2 AFW header. Each train can supply feedwater to each steam generator through a flow control valve, flow transmitters, and isolation valves. As stated in the NRC safety evaluation (SE) associated with PTN Unit 3 and Unit 4 extended power uprate (EPU), the limiting event in terms of total flow from a single AFW pump would be a dual unit loss of offsite power event and assumes the active single failure of one AFW pump concurrent with one additional AFW pump out-of-service (Section 2.5.4 (pgs. 186 through 188) of the PTN EPU SE, NRC ADAMS Accession No. ML11293A365). Thus, the limiting design basis event analysis assumes only two AFW pumps are available at event initiation with one failing to start at the onset of the event resulting in one AFW train available for event mitigation. Therefore, the PTN AFW System effectively consists of an installed spare turbine driven AFW pump. The NRC staff concluded in the SE associated with PTN Unit 3 and Unit 4 EPU that there is reasonable assurance the AFW System would remain capable of performing its licensing basis functions following EPU implementation because the design operation of the AFW pumps would remain within existing bounds and the inventory remains adequate to support natural circulation cooldown to RHR entry conditions. This change is acceptable because a single AFW pump provides the necessary flow assuming one pump does not start. Therefore, two AFW trains, with one pump per train, Turkey Point Unit 3 and Unit 4 Page 4 of 7

DISCUSSION OF CHANGES ITS 3.7.5, AUXILIARY FEEDWATER (AFW) SYSTEM ensures the lowest functional capability or performance levels of AFW equipment required for safe operation of the facility is met pursuant to 10 CFR 50.36(c)(2)(i).

Eliminating the TS requirement for the third AFW pump to be OPERABLE is acceptable because other regulatory requirements ensure the degradation of structures, systems, and components (SSCs) will not result in undue risk to the health and safety of the public. Pursuant to 10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants, PTN monitors the performance or condition of SSCs, against established Maintenance Rule goals, in a manner sufficient to provide reasonable assurance that SSCs are capable of fulfilling their intended functions in accordance with PTN's Configuration Risk Management Program (CRMP). When the performance or condition of an SSC does not meet established goals, appropriate corrective action or compensatory measures will be taken to ensure the risk to the health and safety of the public is minimized. Therefore, NRC regulations contain the necessary programmatic requirements to ensure SSCs are capable of fulfilling their intended functions and appropriate compensatory measures are taken to minimize risk to the health and safety of the public. This change is designated as less restrictive because less stringent LCO and action requirements are being applied in the ITS than were applied in the CTS.

L02 (Category 5 - Deletion of Surveillance Requirement) CTS 4.7.1.2.1.a.4 requires verifying that power is available to those components which require power for flow path operability at a frequency in accordance with the Surveillance Frequency Control Program. ITS 3.7.5 does not include this surveillance test requirement. This changes the CTS by deleting a Surveillance Requirement.

The purpose of the CTS Surveillance Requirement is to ensure power is available to components which require power for flow path Operability. This change is acceptable because the deleted Surveillance Requirement is not necessary to verify the equipment being used to meet the LCO can perform its required function. Appropriate equipment continues to be tested in a manner and at a Frequency necessary to give confidence the equipment can perform its assumed safety function. The definition of OPERABLE-OPERABILITY specified in ITS Section 1.1, Definitions, requires power to be available to power operated and automatic valves to perform their intended safety function. ITS SR 3.7.5.1 (CTS 4.7.1.2.1.a.3) continues to periodically require each AFW manual and power operated valve in the flow path that is not locked, sealed, or otherwise secured in position, is in the correct position. In addition, ITS SR 3.7.5.3 (CTS 4.7.1.2.1.b.1) requires each AFW automatic valve that is not locked, sealed, or otherwise secured in position, to actuate to the correct position on an actual or simulated actuation signal. These Surveillance Requirements, combined with the definition of OPERABLE-OPERABILITY, ensure components in the flow path that require power are maintained OPERABLE. The Surveillance Requirements retained in the ITS are similar to the ISTS Surveillance Requirements and are considered adequate to assure, pursuant to the requirements of 10 CFR 50.36(c)(3), that the necessary quality of systems and components is maintained, facility operation will be within safety limits, and that the limiting condition for operation associated with the AFW System will be met.

This change is designated as less restrictive because Surveillance Requirements that were required in the CTS will not be required in the ITS.

Turkey Point Unit 3 and Unit 4 Page 5 of 7

DISCUSSION OF CHANGES ITS 3.7.5, AUXILIARY FEEDWATER (AFW) SYSTEM L03 (Category 6 - Relaxation Of Surveillance Requirement Acceptance Criteria)

CTS 4.7.1.2.1.b.1) and 4.7.1.2.1.b.2) require verification of the automatic actuation of auxiliary feedwater components upon receipt of each Auxiliary Feedwater Actuation test signal. ITS SR 3.7.5.3 and SR 3.7.5.4 specify that the signal may be from either an "actual" or simulated (i.e., test) signal. This changes the CTS by explicitly allowing the use of either an actual or simulated signal for the test.

The purpose of CTS 4.7.1.2.1.b.1) and 4.7.1.2.1.b.2) is to ensure that the auxiliary feedwater components operate correctly upon receipt of an actuation signal. This change is acceptable because using only a simulated signal to verify the correct actuation of components is not necessary for verification that the equipment used to meet the LCO can perform its required functions. Equipment cannot discriminate between an "actual," "simulated," or "test" signal and, therefore, the results of the testing are unaffected by the type of signal used to initiate the test. This change allows taking credit for unplanned actuation if sufficient information is collected to satisfy the Surveillance test requirements.

The change also allows a simulated signal to be used, if necessary. This change is designated as less restrictive because less stringent Surveillance Requirements are being applied in the ITS than were applied in the CTS.

L04 (Category 7 - Relaxation of Surveillance Frequency) CTS 4.7.1.2.1.b.2) requires l R2 verifying that each auxiliary feedwater pump receives a start signal as designed automatically upon receipt of each Auxiliary Feedwater Actuation test signal. ITS SR 3.7.5.4 requires verifying that each AFW pump starts automatically on an actual or simulated actuation signal and includes a Note that states that this SR is only required to be performed in MODE 1. This changes the CTS by requiring the pump to start, instead of just receiving a start signal, and allowing the SR to only be required to be performed in MODE 1.

The purpose of CTS 4.7.1.2.1.b.2) is to ensure each AFW pump receives a start signal from each AFW Actuation test signal. This is performed while shutdown with steam isolated to the turbine driven AFW (TDAFW) pumps by closing normally locked open manual valves. The test then manually initiates each AFW actuation signal and verifies that the automatic steam supply valves open to supply steam to the TDAFW pumps. The TDAFW pumps are then confirmed to start when these steam supply valves are opened to verify flow in accordance with CTS 4.7.1.2.1.a.1). CTS 4.7.1.2.1.a.1) is modified stating that the provisions of [CTS] Specification 4.0.4 are not applicable for entry into MODES 2 and 3.

This change is acceptable because the new surveillance allowance has been evaluated to ensure that it provides an acceptable level of equipment reliability.

CTS Specification 4.0.4 states, in part, that entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified frequency. Thus, the CTS SR is modified indicating that the SR should be deferred until suitable test conditions are established. This deferral is required because there is insufficient steam pressure to perform the test verifying TDAFW pump OPERABILITY. By changing the deferral requirement from met, in CTS, to performed, in ITS, reasonable assurance of the OPERABILITY of the TDAFW pumps must be Turkey Point Unit 3 and Unit 4 Page 6 of 7

DISCUSSION OF CHANGES ITS 3.7.5, AUXILIARY FEEDWATER (AFW) SYSTEM established to enter MODES 1, 2, or 3. Therefore, to provide this reasonable assurance testing to ensure the TDAFW pump steam supply valves open should be completed before entering the MODES of Applicable for the AFW system.

This change is designated as less restrictive because Surveillances may be performed in less MODES under the ITS than under the CTS.

Turkey Point Unit 3 and Unit 4 Page 7 of 7

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

CTS AFW System 3.7.5 3.7 PLANT SYSTEMS 3.7.5 Auxiliary Feedwater (AFW) System Two and three steam generator steam supplies 3.7.1.2 LCO 3.7.5 [Three] AFW trains shall be OPERABLE. 1 2

NOTE--------------------------------------------

[ Only one AFW train, which includes a motor driven pump, is required to 1

be OPERABLE in MODE 4. ]

Applicability APPLICABILITY: MODES 1, 2, and 3, MODE 4 when steam generator is relied upon for heat removal. 2 ACTIONS

NOTE-----------------------------------------------------------

Actions Note LCO 3.0.4.b is not applicable [when entering MODE 1.] 1 CONDITION REQUIRED ACTION COMPLETION TIME INSERT 1 Action 4) A. [ Turbine driven AFW A.1 Restore affected equipment 7 days train inoperable due to to OPERABLE status.

one inoperable steam [OR supply.

In accordance with OR the Risk Informed Completion Time

NOTE------------ Program] ]

Only applicable if 3 MODE 2 has not been entered following refueling.

One turbine driven AFW pump inoperable in MODE 3 following refueling.

Turkey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY Westinghouse STS 3.7.5-1 Rev. 5.0 2

CTS ITS 3.7.5 3

INSERT 1 Action 4) A. One steam generator A.1 --------------NOTE--------------

steam supply Required Action A.1 is not inoperable. applicable if two or more steam generator steam supplies are inoperable or one or more AFW trains are inoperable.

Verify each OPERABLE 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months steam generator steam supply is aligned to an OPERABLE AFW train.

AND A.2 Restore steam generator 7 days steam supply to OPERABLE status.

Insert Page 3.7.5-1

CTS AFW System 3.7.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME Action 1) B. One AFW train B.1 Restore AFW train to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable in MODE 1, OPERABLE status.

2, or 3 [for reasons other [OR 3 1 than Condition A].

In accordance with the Risk Informed Completion Time Program] 1 C. Turbine driven AFW C.1 Restore the steam supply [24 or 48] hours train inoperable due to to the turbine driven train to one inoperable steam OPERABLE status.

supply.

OR 3 AND C.2 Restore the motor driven [24 or 48] hours One motor driven AFW AFW train to OPERABLE train inoperable. status.

INSERT 2 INSERT 3 2

Action 1), D. Required Action and D. 1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action 2),

Action 4) associated Completion Time of Condition A [, B, AND or C] not met. 12 D. 2 [Be in MODE 4. [18] hours ]

[ OR 2

Two AFW trains inoperable in MODE 1, 2, or 3 for reasons other than Condition C. ]

INSERT 4 Turkey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY Westinghouse STS 3.7.5-2 Rev. 5.0 2

CTS ITS 3.7.5 3

INSERT 2 Action 2) C. Two AFW trains C.1 Restore one AFW train to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 inoperable. OPERABLE status.

4 INSERT 3

NOTE---------------

Action 2) Not applicable when a dual unit 2 shutdown is required.

4 INSERT 4 E. ---------NOTE--------- E.1 Be in MODE 3 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action 1), Only applicable when Action 2),

Action 4) a dual unit shutdown AND is required.

E.2 Be in MODE 4. 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months Required Action and Associated Completion Time of Condition A, B, or C not met.

Insert Page 3.7.5-2

NOTE---------------

CTS Not applicable when both AFW System standby feedwater pumps are capable of providing makeup 3.7.5 flow to the steam generators.

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME F Two F Action 2) E. [ Three] AFW trains E. 1 --------------NOTE-------------- 2 5 4 1 inoperable in MODE 1, LCO 3.0.3 and all other 3 2, or 3. LCO Required Actions requiring MODE changes are suspended until one AFW train is restored to OPERABLE status.

Initiate action to restore Immediately ] 1 one AFW train to OPERABLE status.

F. Required AFW train F.1 Initiate action to restore Immediately inoperable in MODE 4. AFW train to OPERABLE 2 status.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.7.1.2.1.a.3) SR 3.7.5.1 -------------------------------NOTE------------------------------

[ AFW train(s) may be considered OPERABLE during alignment and operation for steam generator 1

level control, if it is capable of being manually realigned to the AFW mode of operation. ]

and Verify each AFW manual, power operated, and [ 31 days automatic valve in each water flow path, [and in both 1 2 the steam supply flow paths to the steam turbine driven OR pump,] that is not locked, sealed, or otherwise secured in position, is in the correct position. In accordance with the Surveillance Frequency Control Program ] 1 Turkey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY Westinghouse STS 3.7.5-3 Rev. 5.0 2

CTS AFW System 3.7.5 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY INSERT 5 4.7.1.2.1.a.1) SR 3.7.5.2 -------------------------------NOTE------------------------------

[ Not required to be performed for the turbine driven AFW pump until [24 hours] after [1000] psig in the 1 steam generator. ]

2 Verify the developed head of each AFW pump at the In accordance flow test point is greater than or equal to the with the required developed head. INSERVICE TESTING PROGRAM 4.7.1.2.1.b.1) SR 3.7.5.3 -------------------------------NOTE------------------------------

[ AFW train(s) may be considered OPERABLE during alignment and operation for steam generator 1 level control, if it is capable of being manually realigned to the AFW mode of operation. ]

Verify each AFW automatic valve that is not locked, [ [18] months sealed, or otherwise secured in position, actuates to 1 the correct position on an actual or simulated OR actuation signal.

In accordance with the Surveillance Frequency Control Program ] 1 Turkey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY Westinghouse STS 3.7.5-4 Rev. 5.0 2

CTS ITS 3.7.5 2

INSERT 5 4.7.1.2.1.a.1) SR 3.7.5.2 -------------------------------NOTE------------------------------

Only required to be performed in MODE 1.

Verify each AFW pump operates for 15 minutes In accordance and develops a flow of 373 gpm to the entrance of with the the steam generators. Surveillance Frequency Control Program Insert Page 3.7.5-4

CTS AFW System 3.7.5 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY 4.7.1.2.1.b.2) SR 3.7.5.4 ------------------------------NOTES-----------------------------

1. [ Not required to be performed for the turbine Only required to be DOC L04 performed in MODE 1.

driven AFW pump until [24 hours] after

[1000] psig in the steam generator. ]

2. [ AFW train(s) may be considered OPERABLE 1 during alignment and operation for steam generator level control, if it is capable of being manually realigned to the AFW mode of operation. ]

Verify each AFW pump starts automatically on an [ [18] months actual or simulated actuation signal. 1 OR In accordance with the Surveillance Frequency Control Program ] 1 4.7.1.2.2 SR 3.7.5.5 [ Verify proper alignment of the required AFW flow Prior to entering 1 1

paths by verifying flow from the condensate storage MODE 2 2 tank to each steam generator. whenever unit has been in MODE 5, MODE 6, or defueled for a cumulative period of > 30 days ] 1 INSERT 6 2 Turkey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY Westinghouse STS 3.7.5-5 Rev. 5.0 2

CTS ITS 3.7.5 2

INSERT 6 4.7.1.2.1.a.2) SR 3.7.5.6 -------------------------------NOTE------------------------------

Only required to be performed in MODE 1.

Verify the AFW pump discharge valves, steam In accordance supply valves, and turbine valves operate as with the l R2 required to deliver the required flow during Surveillance performance of SR 3.7.5.2. Frequency Control Program Insert Page 3.7.5-5

JUSTIFICATION FOR DEVIATIONS ITS 3.7.5, AUXILIARY FEEDWATER (AFW) SYSTEM

1. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

2. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

3. The ISTS is based on a standard Westinghouse AFW System design with two motor driven AFW trains and one turbine driven AFW train with two redundant steam supplies from the steam generators. The PTN design consists of three steam driven AFW pumps shared between the two units. Steam supply to the three pumps is supplied from each unit steam generator via redundant steam supply headers. Each pump turbine can be manually aligned to either steam supply header. The steam supply line from each steam generator to both steam headers consists of a check valve and motor operated steam supply valve. The three pumps discharge through check valves to one of two redundant discharge headers. The AFW System is normally configured with one turbine drive pump aligned to Train 1 steam and feedwater headers and two turbine drive pumps aligned to Train 2 steam and feedwater headers. The safety analysis requires only one turbine driven AFW train to be OPERABLE to support the safety function. The changes are made to align the PTN Technical Specifications, as reasonably practical, with the ISTS to meet the intent of the ISTS requirements. Subsequent Conditions and Required Actions have been relabeled to reflect the additions and deletions.

4. PTN CTS includes an ACTION to perform a shut down of one of the units where the Completion Time is extended so that the units are shut down sequentially. An ACTION and ACTION Notes are added to allow for this sequential shut down. Later ACTIONs are relabeled.

5. The PTN design includes two standby feedwater pumps, one motor driven pump and one diesel driven pump that are available to provide adequate makeup to the steam generators to safety conduct a plant cooldown in the event both AFW trains are inoperable. Therefore, proposed ACTION C is added to ITS 3.7.5 to allow 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to restored one AFW train to OPERABLE status if both AFW trains are inoperable consistent with the PTN current licensing basis provided both standby feedwater pumps are available to support a cooldown of the units. ITS ACTION D (ISTS ACTION D) and proposed ACTION E provide shutdown requirements. These ACTIONS apply when Required Action C.1 is not met within the required Completion Time. ITS 3.7.5 ACTION F (ISTS ACTION E) is included in the ITS for the condition when both AFW trains are inoperable and the standby feedwater pumps are not available to support cooldown of the units. Condition F (ISTS 3.7.5, Condition E) is modified a Note stating that the Condition is not applicable when both standby feedwater pumps can provide makeup flow to the steam generators.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

AFW System B 3.7.5 B 3.7 PLANT SYSTEMS B 3.7.5 Auxiliary Feedwater (AFW) System BASES BACKGROUND The AFW System automatically supplies feedwater to the steam generators to remove decay heat from the Reactor Coolant System upon the loss of normal feedwater supply. The AFW pumps take suction through separate and independent suction lines from the condensate 1 storage tank (CST) (LCO 3.7.6, "Condensate Storage Tank (CST)") and pump to the steam generator secondary side via separate and independent connections to the main feedwater (MFW) piping outside containment. The steam generators function as a heat sink for core decay heat. The heat load is dissipated by releasing steam to the atmosphere from the steam generators via the main steam safety valves (MSSVs) (LCO 3.7.1, "Main Steam Safety Valves (MSSVs)") or atmospheric dump valves (LCO 3.7.4, "Atmospheric Dump Valves (ADVs)"). If the main condenser is available, steam may be released via the steam bypass valves and recirculated to the CST.

INSERT 1 The AFW System consists of [two] motor driven AFW pumps and one steam turbine driven pump configured into [three] trains. Each motor driven pump provides [100]% of AFW flow capacity, and the turbine driven pump provides [200]% of the required capacity to the steam generators, as assumed in the accident analysis. The pumps are equipped with independent recirculation lines to prevent pump operation against a closed system. Each motor driven AFW pump is powered from an independent Class 1E power supply and feeds [two] steam generators, although each pump has the capability to be realigned from the control room to feed other steam generators. The steam turbine driven AFW pump receives steam from two main steam lines upstream of the main steam isolation valves. Each of the steam feed lines will supply 100% of the requirements of the turbine driven AFW pump. 1 The AFW System is capable of supplying feedwater to the steam generators during normal unit startup, shutdown, and hot standby conditions.

The turbine driven AFW pump supplies a common header capable of feeding all steam generators with DC powered control valves actuated to the appropriate steam generator by the Engineered Safety Feature Actuation System (ESFAS). One pump at full flow is sufficient to remove decay heat and cool the unit to residual heat removal (RHR) entry conditions. Thus, the requirement for diversity in motive power sources for the AFW System is met.

Turkey Point Unit 3 and Unit 4 Revision XXX Westinghouse STS B 3.7.5-1 Rev. 5.0 1

ITS 3.7.5 1

INSERT 1 The AFW System is a shared system between Units 3 and 4. The AFW System consists of three steam driven pumps configured into two trains. Each pump provides 100% of required flow capacity to the steam generators, as assumed in the accident safety analysis. The three pumps are configured such that each can supply auxiliary feedwater to either Unit 3 or 4, with any single pump supplying the total feedwater requirement to both units. Steam supply to the three pumps is supplied from each unit steam generator via redundant steam supply headers.

Each pump turbine can be manually aligned to either steam supply header. The steam supply line from each steam generator to both steam headers consists of a check valve and motor operated steam supply valve. The three pumps discharge through check valves to one of two redundant discharge headers. The AFW System is normally configured with one turbine drive pump aligned to Train 1 steam and feedwater headers and two turbine drive pumps aligned to Train 2 steam and feedwater headers. Auxiliary feedwater can be supplied through redundant lines to the safety-related portions of the main feedwater lines to each of the steam generators.

Each pump has sufficient capacity for single and two unit operation to ensure adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350°F when the Residual Heat Removal (RHR) System may be placed into operation.

Standby steam generator feedwater pumps are normally used to supply feedwater to the steam generators during normal start-up, shutdown, and hot standby conditions. In the event the AFW system does not function properly, the Standby Feedwater System can be used as a backup water supply.

Upon a loss of normal feedwater, steam is supplied to the AFW System is supplied from the unit which has lost feedwater. Steam can also be supplied from the opposite unit steam generators or from the units auxiliary steam system. The unit steam supply valves automatically open by any one of the following five signals.

a. Safety Injection,

b. Low-Low Level in any of the three steam generators,

c. Loss of both feedwater pumps under normal operating conditions,

d. Loss of AC electrical distribution 4.16 kV or 480 load center buses, and

e. Anticipated Transient Without Scram (ATWS) Mitigating Actuation Circuitry (AMSAC) signal.

Insert Page B 3.7.5-1

AFW System B 3.7.5 BASES BACKGROUND (continued)

The AFW System is designed to supply sufficient water to the steam generator(s) to remove decay heat with steam generator pressure at the setpoint of the MSSVs. Subsequently, the AFW System supplies sufficient water to cool the unit to RHR entry conditions, with steam released through the ADVs.

The AFW System actuates automatically on steam generator water level -

low-low by the ESFAS (LCO 3.3.2, "Engineered Safety Feature Actuation 6

System (ESFAS) Instrumentation"). The system also actuates on loss of offsite power, safety injection, and trip of all MFW pumps.

U 9.11 The AFW System is discussed in the FSAR, Section [10.4. 9] (Ref. 1). 1 2 APPLICABLE The AFW System mitigates the consequences of any event with loss of SAFETY normal feedwater.

ANALYSES The design basis of the AFW System is to supply water to the steam generator to remove decay heat and other residual heat by delivering at least the minimum required flow rate to the steam generators at pressures corresponding to the lowest steam generator safety valve set pressure plus 3%.

In addition, the AFW System must supply enough makeup water to replace steam generator secondary inventory lost as the unit cools to MODE 4 conditions. Sufficient AFW flow must also be available to account for flow losses such as pump recirculation and line breaks.

The limiting Design Basis Accidents (DBAs) and transients for the AFW System are as follows:

(limited scope analysis),

a. Feedwater Line Break (FWLB) and Normal Feedwater (LONF) , 1

b. Loss of MFW. c. Main Steam Line Break (MSLB), and

d. Dual Unit Loss of Offsite Power (LOOP).

In addition, the minimum available AFW flow and system characteristics are serious considerations in the analysis of a small break loss of coolant accident (LOCA).

INSERT 2 The AFW System design is such that it can perform its function following an FWLB between the MFW isolation valves and containment, combined with a loss of offsite power following turbine trip, and a single active failure of the steam turbine driven AFW pump. In such a case, the Turkey Point Unit 3 and Unit 4 Revision XXX Westinghouse STS B 3.7.5-2 Rev. 5.0 1

ITS 3.7.5 1

INSERT 2 The limiting single unit event for the AFW System is a LONF event. The minimum required flow during the limiting LONF event is 373 gpm. The limiting dual unit transient for the AFW System is a Dual Unit LOOP event. The minimum flow requirement for this case is a total flow of 624.8 gpm delivered to both units, or approximately 312.4 gpm per unit.

Insert Page B 3.7.5-2

AFW System B 3.7.5 BASES APPLICABLE SAFETY ANALYSES (continued)

ESFAS logic may not detect the affected steam generator if the backflow check valve to the affected MFW header worked properly. One motor driven AFW pump would deliver to the broken MFW header at the pump 1 runout flow until the problem was detected, and flow terminated by the operator. Sufficient flow would be delivered to the intact steam generator by the redundant AFW pump.

The ESFAS automatically actuates the AFW turbine driven pump and 1 associated power operated valves and controls when required to ensure an adequate feedwater supply to the steam generators during loss of power. DC power operated valves are provided for each AFW line to control the AFW flow to each steam generator.

The AFW System satisfies the requirements of Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO This LCO provides assurance that the AFW System will perform its design safety function to mitigate the consequences of accidents that could result in overpressurization of the reactor coolant pressure Two of three boundary. [Three] independent AFW pumps in [three] diverse trains are 2 two required to be OPERABLE to ensure the availability of RHR capability for all events accompanied by a loss of offsite power and a single failure.

This is accomplished by powering two of the pumps from independent 1 Each emergency buses. The third AFW pump is powered by a different means, a steam driven turbine supplied with steam from a source that is not isolated by closure of the MSIVs.

two The AFW System is configured into [three] trains. The AFW System is 2 considered OPERABLE when the components and flow paths required to provide redundant AFW flow to the steam generators are OPERABLE.

INSERT 3 This requires that the two motor driven AFW pumps be OPERABLE in

[two] diverse paths, each supplying AFW to separate steam generators.

The turbine driven AFW pump is required to be OPERABLE with 1 redundant steam supplies from each of [two] main steam lines upstream of the MSIVs, and shall be capable of supplying AFW to any of the steam generators. The piping, valves, instrumentation, and controls in the required flow paths also are required to be OPERABLE.

The LCO is modified by a Note indicating that one AFW train, which includes a motor driven pump, is required to be OPERABLE in MODE 4.

This is because of the reduced heat removal requirements and short 1 period of time in MODE 4 during which the AFW is required and the insufficient steam available in MODE 4 to power the turbine driven AFW pump.

Turkey Point Unit 3 and Unit 4 Revision XXX Westinghouse STS B 3.7.5-3 Rev. 5.0 1

ITS 3.7.5 1

INSERT 3 This requires two independent AFW trains, each consisting of a steam supply header, discharge header with flow path to all three steam generators, at least one steam generator steam supply aligned to the steam supply header, and at least one OPERABLE pump aligned to the steam supply header and AFW discharge header.

Two steam generator steam supply lines must be aligned to a separate AFW steam supply header and the third steam generator steam supply line may be aligned to either AFW steam supply line but not both simultaneously. This ensures that three AFW steam supplies to the AFW steam supply headers are OPERABLE.

To ensure redundancy and independence, a separate AFW pump must be aligned to each train.

A third AFW pump may be credited as an OPERABLE pump when aligned to the supply water header it is credited, independent of the other AFW pump. The third AFW pump can be aligned to either train to restore OPERABILITY in the event one of the other pumps is inoperable. The steam turbine, trip and throttle valve, and governor valve are required to support AFW pump OPERABILITY.

Insert Page B 3.7.5-3

AFW System B 3.7.5 BASES APPLICABILITY In MODES 1, 2, and 3, the AFW System is required to be OPERABLE in the event that it is called upon to function when the MFW is lost. In addition, the AFW System is required to supply enough makeup water to replace the steam generator secondary inventory, lost as the unit cools to MODE 4 conditions. Reactor Coolant System loops and the RHR System provide decay heat removal In MODE 4 the AFW System may be used for heat removal via the steam 1 generators.

In MODE 5 or 6, the steam generators are not normally used for heat removal, and the AFW System is not required.

ACTIONS -----------------------------------REVIEWERS NOTE-----------------------------------

The LCO 3.0.4.b Note prohibits application of the LCO 3.0.4.b exception when entering MODE 1 if the plant does not depend on AFW for startup. 3 If the plant does depend on AFW for startup, the Note should state, LCO 3.0.4.b is not applicable.

A Note prohibits the application of LCO 3.0.4.b to an inoperable AFW train [when entering MODE 1]. There is an increased risk associated with

[entering a MODE or other specified condition in the Applicability] 2

[entering MODE 1] with an AFW train inoperable and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.

[ A.1 4 INSERT 4 If the turbine driven AFW train is inoperable due to one inoperable steam supply, or if a turbine driven pump is inoperable for any reason while in MODE 3 immediately following refueling, action must be taken to restore the inoperable equipment to an OPERABLE status within 7 days [or in accordance with the Risk Informed Completion Time Program]. The 7 day Completion Time is reasonable, based on the following reasons:

a. For the inoperability of the turbine driven AFW pump due to one 1 inoperable steam supply, the 7 day Completion Time is reasonable since there is a redundant steam supply line for the turbine driven pump and the turbine driven train is still capable of performing its specified function for most postulated events.

b. For the inoperability of a turbine driven AFW pump while in MODE 3 immediately subsequent to a refueling, the 7 day Completion Time is reasonable due to the minimal decay heat levels in this situation.

Turkey Point Unit 3 and Unit 4 Revision XXX Westinghouse STS B 3.7.5-4 Rev. 5.0 1

ITS 3.7.5 1

INSERT 4 A.1 and A.2 Condition A describes the actions to be taken when one steam generator steam supply is inoperable. Three steam generator steam supplies must be OPERABLE to satisfy the design basis requirement that the AFW System meet the single failure criterion in response to a MSLB.

With one steam generator steam supply out of service not associated with the faulted steam generator during a MSLB, failure of the remaining steam supply would result in a loss of AFW System function. Therefore, Required Action A.1 requires verification that each OPERABLE steam generator steam supply is aligned to an OPERABLE AFW train (i.e., a separate steam supply aligned to each AFW train steam supply header). Required Action A.2 allows seven days to restore the steam generator steam supply to OPERABLE status provided each AFW train is supplied by an OPERABLE steam generator steam supply. If an OPERABLE steam generator steam supply cannot be aligned to an OPERABLE AFW train, then at least one AFW train is inoperable and Condition B applies.

Required Action A.1 has been modified by a Note that ensures the Required Action and associated Completion Time are only required if two OPERABLE steam generator steam supplies and two OPERABLE AFW trains are available. With two or more steam generator steam supplies or one or more AFW trains inoperable, Required Action A.1 cannot be performed and the Conditions associated with one or two AFW trains inoperable provide the appropriate remedial actions. The exception of Note to Required Action A.1 does not affect tracking the Completion Time from the initial entry into Condition A; only the requirement to comply with the Required Action.

The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time provides time to manually align, if necessary, an OPERABLE steam generator steam supply to the applicable AFW train. The 7 day Completion Time is reasonable since there is a redundant steam supply line for the turbine driven pumps and the turbine driven trains are still capable of performing their specified function for most postulated events.

Insert Page B 3.7.5-4

AFW System B 3.7.5 BASES ACTIONS (continued)

c. For both the inoperability of the turbine driven pump due to one inoperable steam supply and an inoperable turbine driven AFW pump while in MODE 3 immediately following a refueling outage, the 7 day Completion Time is reasonable due to the availability of redundant OPERABLE motor driven AFW pumps, and due to the low probability of an event requiring the use of the turbine driven AFW pump.

1 Condition A is modified by a Note which limits the applicability of the Condition for an inoperable turbine driven AFW pump in MODE 3 to when the unit has not entered MODE 2 following a refueling. Condition A allows one AFW train to be inoperable for 7 days vice the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time in Condition B. This longer Completion Time is based on the reduced decay heat following refueling and prior to the reactor being critical. ]

B.1 With one of the required AFW trains (pump or flow path) inoperable in 1 MODE 1, 2, or 3 [for reasons other than Condition A], action must be 5 the train to taken to restore OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months [or in accordance with 2 steam generator the Risk Informed Completion Time Program]. This Condition includes the loss of two steam supply lines to the turbine driven AFW pump. The steam supply headers 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is reasonable, based on the redundant capabilities afforded by the AFW System, the time needed for repairs, and the low probability of a DBA occurring during this time period.

C.1 and C.2 With one of the required motor driven AFW trains (pump or flow path) inoperable and the turbine driven AFW train inoperable due to one inoperable steam supply, action must be taken to restore the affected equipment to OPERABLE status within [24] [48] hours. Assuming no single active failures when in this condition, the accident (a feedline break (FLB) or main steam line break (MSLB) could result in the loss of the 1 remaining steam supply to the turbine driven AFW pump due to the faulted steam generator (SG). In this condition, the AFW System may no longer be able to meet the required flow to the SGs assumed in the safety analysis, [either due to the analysis requiring flow from two AFW pumps or due to the remaining AFW pump having to feed a faulted SG].

Turkey Point Unit 3 and Unit 4 Revision XXX Westinghouse STS B 3.7.5-5 Rev. 5.0 1

AFW System B 3.7.5 BASES ACTIONS (continued)

REVIEWERS NOTE----------------------------------

Licensees should adopt the appropriate Completion Time based on their plant design. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is applicable to plants that can no longer meet the safety analysis requirement of 100% AFW flow to the SG(s) assuming no single active failure and a FLB or MSLB resulting in the loss of the remaining steam supply to the turbine driven AFW pump. The 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months Completion Time is applicable to plants that can still meet the safety analysis requirement of 100% AFW flow to the SG(s) assuming no single active failure and a FLB or MSLB resulting in the loss of the remaining steam supply to the turbine driven AFW pump.

1

[ The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is reasonable based on the remaining OPERABLE steam supply to the turbine driven AFW pump, the availability of the remaining OPERABLE motor driven AFW pump, and the low probability of an event occurring that would require the inoperable steam supply to be available for the turbine driven AFW pump. ]

[ The 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months Completion Time is reasonable based on the fact that the remaining motor driven AFW train is capable of providing 100% of the AFW flow requirements, and the low probability of an event occurring that would challenge the AFW system. ]

INSERT 5 D.1 and D.2 of Condition A, B, or C When Required Action A.1 [B.1, C.1, or C.2] cannot be completed within 2 the required Completion Time, or if two AFW trains are inoperable in 1 MODE 1, 2, or 3 for reasons other than Condition C, the unit must be 5 placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in 12 MODE 4 within [18] hours. 2 The allowed Completion Times are reasonable, based on operating Condition D is modified by a Note stating that experience, to reach the required unit conditions from full power this Condition is not conditions in an orderly manner and without challenging unit systems.

applicable when a dual unit shutdown is required. In the case of In MODE 4 with two AFW trains inoperable, operation is allowed to dual unit shutdown continue because only one motor driven pump AFW train is required in 1 Condition E is the accordance with the Note that modifies the LCO. Although not required, applicable condition.

the unit may continue to cool down and initiate RHR.

INSERT 6 Turkey Point Unit 3 and Unit 4 Revision XXX Westinghouse STS B 3.7.5-6 Rev. 5.0 1

ITS 3.7.5 1

INSERT 5 C.1 When two AFW trains are inoperable, there are no remaining safety related means for conducting a plant cooldown. Thus, with an assumed loss of power, insufficient safety related steam generator makeup sources are available. This Condition includes the loss of three steam generator steam supply lines to the AFW steam supply headers. This action is based on the consideration that both standby feedwater pumps (one motor driven pump and one diesel driven pump) can provide adequate makeup to the steam generators to safety conduct a plant cooldown. Therefore, continued operation for a very short time is acceptable with this level of degradation. If the standby feedwater pumps cannot provide makeup flow to the steam generators, Condition F also applies.

The Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is reasonable, based on the redundant capabilities afforded by the standby feedwater pumps, the time needed for repairs, and the low probability of a DBA occurring during this time period.

4 INSERT 6 E.1 and E.2 When Required Actions of Condition A, B, C, or D, cannot be completed within the required Completion Time, the unit must be placed in a MODE in which the LCO does not apply. If these Required Actions affect both units and a dual unit shutdown is required, one unit should be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months while the second unit should be placed in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and in MODE 4 within 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months .

Condition E is modified by a Note that states that this Condition is only applicable when a dual unit shutdown is required. This allows for the orderly shutdown of one unit at a time to not jeopardize the stability of the electrical grid by imposing a simultaneous dual unit shutdown.

The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

Insert Page B 3.7.5-6

AFW System B 3.7.5 BASES ACTIONS (continued)

F E.1 4 two

. The Condition is modified a If all [three] AFW trains are inoperable in MODE 1, 2, or 3, the unit is in a 5 2 Note stating that the Condition is seriously degraded condition with no safety related means for conducting not applicable when both a cooldown, and only limited means for conducting a cooldown with standby feedwater pumps are 1 capable of providing makeup nonsafety related equipment. In such a condition, the unit should not be flow to the steam generators. perturbed by any action, including a power change, that might result in a The standby feedwater pumps can provide adequate makeup to trip. The seriousness of this condition requires that action be started the steam generators to safety immediately to restore one AFW train to OPERABLE status.

conduct a plant cooldown. If F both AFW trains are inoperable and both standby feedwater Required Action E.1 is modified by a Note indicating that all required 4 pumps are not available MODE changes are suspended until one AFW train is restored to OPERABLE status. In this case, LCO 3.0.3 is not applicable because it could force the unit into a less safe condition.

F.1 1 In MODE 4, either the reactor coolant pumps or the RHR loops can be used to provide forced circulation. This is addressed in LCO 3.4.6, "RCS Loops - MODE 4." With one required AFW train inoperable, action must 1 be taken to immediately restore the inoperable train to OPERABLE status. The immediate Completion Time is consistent with LCO 3.4.6.

SURVEILLANCE SR 3.7.5.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the AFW System water and steam supply flow paths provides assurance that the proper flow paths will exist for AFW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since they are verified to be in the correct position prior to locking, sealing, or securing. This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This Surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position.

[ The SR is modified by a Note that states one or more AFW trains may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually (i.e., remotely or locally, as appropriate) realigned to the AFW mode of operation, provided 2 it is not otherwise inoperable. This exception allows the system to be out of its normal standby alignment and temporarily incapable of automatic initiation without declaring the train(s) inoperable. Since AFW may be Turkey Point Unit 3 and Unit 4 Revision XXX Westinghouse STS B 3.7.5-7 Rev. 5.0 1

AFW System B 3.7.5 BASES SURVEILLANCE REQUIREMENTS (continued) used during startup, shutdown, hot standby operations, and hot shutdown operations for steam generator level control, and these manual operations are an accepted function of the AFW System, OPERABILITY (i.e., the intended safety function) continues to be maintained. ]

2

[ The 31 day Frequency is based on engineering judgment, is consistent with the procedural controls governing valve operation, and ensures correct valve positions.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency 3 description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

SR 3.7.5.2 INSERT 7 Verifying that each AFW pump's developed head at the flow test point is greater than or equal to the required developed head ensures that AFW pump performance has not degraded during the cycle. Flow and differential head are normal tests of centrifugal pump performance required by the ASME Code (Ref 2). Because it is undesirable to introduce cold AFW into the steam generators while they are operating, this testing is performed on recirculation flow. This test confirms one point on the pump design curve and is indicative of overall performance.

Such inservice tests confirm component OPERABILITY, trend 1 performance, and detect incipient failures by indicating abnormal performance. Performance of inservice testing as discussed in the ASME Code (Ref. 2) and the INSERVICE TESTING PROGRAM satisfies this requirement.

[ This SR is modified by a Note indicating that the SR should be deferred until suitable test conditions are established. This deferral is required because there is insufficient steam pressure to perform the test. ]

Turkey Point Unit 3 and Unit 4 Revision XXX Westinghouse STS B 3.7.5-8 Rev. 5.0 1

ITS 3.7.5 1

INSERT 7 This SR verifies that each turbine driven pump operates for 15 minutes and develops a flow of 373 gpm to the entrance of the steam generators. The specified flow rate conservatively bounds the limiting AFW flow rate modeled in the single unit Loss of Normal Feedwater analysis. Dual unit events, such as a Dual Unit LOOP event, require a higher pump flow rate, but it is not practical to test both units simultaneously. The flow surveillance test specified ensures the minimum flow requirements for a single unit event are met and is considered to be a general performance test for the AFW System. This test does not represent the minimum AFW System flow requirement for both units. If total flowrate is determined to be less than the minimum required flowrate for the limiting dual unit event, an evaluation should be performed to determine the AFW pump capability to support a Dual Unit LOOP event. Verification of correct operation will be made both from instrumentation within the Control Room and direct visual observation of the pumps.

This SR is modified by a Note that allows entry into and operation in MODES 2 and 3 prior to the surveillance being performed. This allows establishment of adequate steam generator pressure and flow to perform the test.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Insert Page B 3.7.5-8

AFW System B 3.7.5 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.7.5.3 This SR verifies that AFW can be delivered to the appropriate steam generator in the event of any accident or transient that generates an ESFAS, by demonstrating that each automatic valve in the flow path actuates to its correct position on an actual or simulated actuation signal.

This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.

[ The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. The [18] month Frequency is acceptable based 2 on operating experience and the design reliability of the equipment.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency 3 description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

[ The SR is modified by a Note that states one or more AFW trains may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually (i.e., remotely or locally, as appropriate) realigned to the AFW mode of operation, provided it is not otherwise inoperable. This exception allows the system to be out of its normal standby alignment and temporarily incapable of automatic 2 initiation without declaring the train(s) inoperable. Since AFW may be used during startup, shutdown, hot standby operations, and hot shutdown operations for steam generator level control, and these manual operations are an accepted function of the AFW System, OPERABILITY (i.e., the intended safety function) continues to be maintained. ]

Turkey Point Unit 3 and Unit 4 Revision XXX Westinghouse STS B 3.7.5-9 Rev. 5.0 1

AFW System B 3.7.5 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.7.5.4 This SR verifies that the AFW pumps will start in the event of any accident or transient that generates an ESFAS by demonstrating that each AFW pump starts automatically on an actual or simulated actuation signal in MODES 1, 2, and 3. In MODE 4, the required pump is already 5 operating and the autostart function is not required. [ The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at 2 power.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency 3 description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

The This SR is modified by [a] [two] Note[s]. [Note 1 indicates that the SR be 2 deferred until suitable test conditions are established. This deferral is required because there is insufficient steam pressure to perform the test.] 2

[The Note [2] states that one or more AFW trains may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually (i.e., remotely or locally, as appropriate) realigned to the AFW mode of operation, provided it is not 1

otherwise inoperable. This exception allows the system to be out of its normal standby alignment and temporarily incapable of automatic initiation without declaring the train(s) inoperable. Since AFW may be used during startup, shutdown, hot standby operations, and hot shutdown operations for steam generator level control, and these manual operations are an accepted function of the AFW System. OPERABILITY (i.e., the intended safety function) continues to be maintained. ]

Turkey Point Unit 3 and Unit 4 Revision XXX Westinghouse STS B 3.7.5-10 Rev. 5.0 1

AFW System B 3.7.5 BASES SURVEILLANCE REQUIREMENTS (continued)

[ SR 3.7.5.5 2 This SR verifies that the AFW is properly aligned by verifying the flow 1 paths from the CST to each steam generator prior to entering MODE 2 1 after more than 30 days in any combination of MODE 5 or 6 or defueled.

OPERABILITY of AFW flow paths must be verified before sufficient core heat is generated that would require the operation of the AFW System during a subsequent shutdown. The Frequency is reasonable, based on engineering judgement and other administrative controls that ensure that flow paths remain OPERABLE. To further ensure AFW System alignment, flow path OPERABILITY is verified following extended outages to determine no misalignment of valves has occurred. This SR ensures that the flow path from the CST to the steam generators is properly aligned. ] 2 INSERT 8

REVIEWERS NOTE-----------------------------------

This SR is not required by those units that use AFW for normal startup 3 and shutdown.

U 9.11 REFERENCES 1. FSAR, Section [10.4.9]. 1

2. ASME Code for Operation and Maintenance of Nuclear Power 1 Plants.

Turkey Point Unit 3 and Unit 4 Revision XXX Westinghouse STS B 3.7.5-11 Rev. 5.0 1

ITS 3.7.5 1

INSERT 8 SR 3.7.5.6 This SR verifies that the AFW discharge valves and the steam supply and turbine valves l R2 operate as required. Check valves in the AFW System that require full stroke testing under limiting flow conditions are tested in accordance with the INSERVICE TESTING PROGRAM.

Proper functioning of the turbine admission valves and the operation of the pumps will l R2 demonstrate the integrity of the system. Verification of correct operation will be made both from instrumentation within the Control Room and direct visual observation of the pumps.

This SR is modified by a Note that allows entry into and operation in MODES 2 and 3 prior to l R2 the surveillance being performed. This allows establishment of adequate steam generator l l

pressure and flow to perform the test.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Insert Page B 3.7.5-11

JUSTIFICATION FOR DEVIATIONS ITS 3.7.5 BASES, AUXILIARY FEEDWATER (AFW) SYSTEM

1. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

3. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.

4. Changes are made to be consistent with the Specification.

5. PTN does not have a requirement to be OPERABLE in MODE 4, therefore identifying inoperabilities only in MODE 1, 2, or 3 is unnecessary and is deleted.

6. Duplicative information that was covered earlier.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.7.5, AUXILIARY FEEDWATER (AFW) SYSTEM There are no specific No Significant Hazards Considerations for this Specification.

Turkey Point Unit 3 and 4 Page 1 of 1

ATTACHMENT 6 ITS 3.7.6, CONDENSATE STORAGE TANK (CST)

Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

ITS A01 ITS 3.7.6 PLANT SYSTEMS CONDENSATE STORAGE TANK LIMITING CONDITION FOR OPERATION LCO 3.7.6 3.7.1.3 The condensate storage tanks (CST) system shall be OPERABLE with:

Opposite Unit in MODES 4, 5, 6, or defueled LA01 SR 3.7.6.2 A minimum indicated water volume of 210,000 gallons in either or both condensate storage tanks.

level A02 Opposite Unit in MODES 1, 2 or 3 SR 3.7.6.1 A minimum indicated water volume of 420,000 gallons.

Applicability APPLICABILITY: MODES 1, 2 and 3.

ACTION: LA01 Opposite Unit in MODES 4, 5, 6, or defueled Add proposed Required Actions A.1 and A.2 L01 Action A With the CST system inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months restore the CST system to OPERABLE status or be in at least Action B HOT STANDBY in 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 .

Opposite Unit in MODES 1, 2 or 3 LA01 Action A 1) With the CST system inoperable due to indicating less than 420,000 gallons, but greater than or equal to 210,000 gallons indicated, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months restore the inoperable CST system to OPERABLE status or Action B place one unit 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 . L01 Add proposed Required Actions A.1 and A.2 Action A 2) With the CST system inoperable with less than 210,000 gallons indicated, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore the CST system to OPERABLE status or be in at least HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in HOT Action C SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . This ACTION applies to both units simultaneously.

Add proposed Condition B and Condition C Notes LA02 A03 TURKEY POINT - UNITS 3 & 4 3/4 7-6 AMENDMENT NOS. 289 AND 283 Page 1 of 2

ITS A01 ITS 3.7.6 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

SR 3.7.6.1 4.7.1.3 The condensate storage tank (CST) system shall be demonstrated OPERABLE by verifying the SR 3.7.6.2 indicated water volume is within its limit when the tank is the supply source for the auxiliary feedwater A02 pumps in accordance with the Surveillance Frequency Control Program.

TURKEY POINT - UNITS 3 & 4 3/4 7-7 AMENDMENT NOS. 263 AND 258 Page 2 of 2

DISCUSSION OF CHANGES ITS 3.7.6, CONDENSATE STORAGE TANK (CST)

ADMINISTRATIVE CHANGES A01 In the conversion of the Turkey Point Nuclear Generating Station (PTN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 5.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal.

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 3.7.1.3 requires that the condensate storage tanks system be OPERABLE with a minimum indicated water volume of either 210,000 or 420,000 gallons based on the opposite unit's MODE. CTS 4.7.1.3 requires demonstration of the CST System OPERABILITY by verifying the indicated water volume is within limits. ITS 3.7.6 requires that condensate storage tank system be OPERABLE stating the required indicated water volume in surveillance requirements (SR) SR 3.7.6.2 and SR 3.7.6.1 as a level 210,000 or 420,000 gallons based on the opposite unit's MODE. This changes the CTS by describing the indicated water volume as a level.

This change is being made so that the ITS is as consistent as possible with NUREG 1431. This change is acceptable because there is not technical change to the specification. The determination of the acceptable amount of water in the condensate storage tanks will be the same based on the level of water indicated on the condensate storage tanks level indication. This change is considered administrative because no technical change is being made.

A03 CTS 3.7.1.3 provides OPERABILITY requirements for the CST System.

Because the CST System is shared between both Unit 3 and Unit 4, CTS 3.7.1.3 provides different requirements and ACTIONS based on the opposites unit's MODE. If the CST System's condition does not support operation of both units in MODES 1, 2, and 3 both units are required to shut down. CTS 3.7.1.3 ACTION 2), when the opposite unit is in MODE 1, 2, or 3, allows 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to be in at least HOT STANDBY and states, "This ACTION applies to both units simultaneously." The other CTS 3.7.1.3 ACTIONS allow 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to be in at least HOT STANDBY when their stated actions and allowed outage times are not met.

ITS 3.7.6 Condition B and Condition C contain a Note that states when the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Completion Time to shut down to MODE 3 is applicable. This changes the CTS replacing the statement associated with the ACTION that applies to both units simultaneously with Condition Note stating which Condition applies based on whether a single unit or both units are required to be shut down.

This change is being made so that the ITS is as consistent as possible with NUREG 1431. CTS provides separated shut down times based on whether a single unit or both units are required to be shut down. ITS provides similar requirements but changes how these requirements are presented. This change is acceptable because no technical requirements are being changed. This Turkey Point Unit 3 and Unit 4 Page 1 of 3

DISCUSSION OF CHANGES ITS 3.7.6, CONDENSATE STORAGE TANK (CST) changes is considered administrative because no technical requirements are being changed.

MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.7.1.3 provides OPERABILITY allowances (in either or both condensate storage tanks) and Actions based on the opposite unit's MODE and required indicated water volumes (210,000 gallons or 420,000 gallons). ITS 3.7.6 does not include these details of system design and limitations in the LCO or ACTIONS sections but provides the level requirements in the Surveillance Requirements section with the description of the system design and limitations included in the Bases for the Surveillance Requirements. This changes the CTS by describing the CST System alignment and associated interactions in the Bases.

The removal of these details, which are related to system design, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirements to maintain an acceptable volume of water in the CSTs to assure protection of public health and safety. Also, this change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications.

LA02 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.7.1.3 ACTIONS provides guidance on which ACTION applies to both units. ITS 3.7.6 does not include these details of system design and limitations in the ACTIONS but provides the details of which Conditions apply to one or both units in the Bases. This changes the CTS by removing the specific information on which ACTION applies to both units placing it in the Bases.

The removal of these details, which are related to system design, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate Turkey Point Unit 3 and Unit 4 Page 2 of 3

DISCUSSION OF CHANGES ITS 3.7.6, CONDENSATE STORAGE TANK (CST) protection of public health and safety. The ITS still retains the Conditions, Required Actions, and associated Competition Times to assure protection of public health and safety. Also, this change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 4 - Relaxation of Required Action) CTS 3.7.1.3 ACTION allows four hours, or 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months if < 210,000 gallons when the opposite unit is in MODES 1, 2, or 3, to restore the CST system to OPERABLE status or shut down one or both units. ITS 3.7.6 Required Action A.1 requires the verification of an available backup water supply within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter and Required Action A.2 requires the CST to be restored to OPERABLE status within 7 days. This changes the CTS by extending the Completion Time in the CTS 3.7.1.3 ACTION to restore the CST to OPERABLE status from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 7 days provided the backup water supply is verified available within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

The purpose of CTS 3.7.1.3 ACTION is to provide compensatory actions for when the CST system is found to be inoperable. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to degraded conditions to minimize risk associated with continued operation while providing time to repair inoperable features and the low probability of an event requiring the use of the water from the CST occurring during this period. The backup supply is the Unit 3 500,000 gallon demineralized water storage tank and the Unit 4 500,000 gallon demineralized water storage tank. The Completion Time to restore the CST to OPERABLE status considers that the backup supply may be performing this function in addition to its normal functions. Verification of a backup water supply with periodic re-verification provides adequate assurance that the CTS volume analysis assumptions are satisfied, and continued operation is justified for the 7 day allowed restoration period. In addition, the opportunity to restore the equipment to OPERABLE status is always available. ITS LCO 3.0.2 states that upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met. If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required unless otherwise stated. Therefore, based on ITS LCO 3.0.2 restoration is always an option. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS.

Turkey Point Unit 3 and Unit 4 Page 3 of 3

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

CTS CST 3.7.6 3.7 PLANT SYSTEMS 3.7.6 Condensate Storage Tank (CST)

System 3.7.1.3 LCO 3.7.6 The CST shall be OPERABLE. 1 Applicability APPLICABILITY: MODES 1, 2, and 3, MODE 4 when steam generator is relied upon for heat removal. 2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 1

A. CST inoperable. A.1 Verify by administrative 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 3

System availability means OPERABILITY of backup water supply. AND l R2 Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter AND l R2 System A.2 Restore CST to 7 days 1 OPERABLE status.

INSERT 1 1 Action B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND 12 B.2 Be in MODE 4, without [24] hours 4 reliance on steam 1 generator for heat removal.

INSERT 2 1 Westinghouse STS 3.7.6-1 Rev. 5.0 2 Turkey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY

1 INSERT 1

NOTE-------------

Action Not applicable when a dual unit shutdown is required.

1 INSERT 2 C. --------------NOTE---------- C.1 Be in MODE 3 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Only applicable when a Action dual unit shutdown is AND required.

C.2 Be in MODE 4. 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months Required Action and associated Completion Time of Condition A not met.

Insert Page 3.7.6-1

CTS CST 3.7.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY INSERT 3 4.7.1.3 SR 3.7.6.1 Verify the CST level is [110,000 gal]. [ 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 4

420,000 OR In accordance with the Surveillance Frequency Control Program ] 4 INSERT 4 Westinghouse STS 3.7.6-2 Rev. 5.0 2 Turkey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY

1 INSERT 3

NOTE---------------------------

3.7.1.3 Only required to be met when the opposite unit is in MODE 1, 2, or 3.

2 INSERT 4 3.7.1.3 SR 3.7.6.2 -------------------------------NOTE------------------------------

Only required to be met when the opposite unit is not in MODE 1, 2, or 3.

Verify the CST level is 210,000 gal. In accordance with the Surveillance Frequency Control Program Insert Page 3.7.6-2

JUSTIFICATION FOR DEVIATIONS ITS 3.7.6, CONDENSATE STORAGE TANK (CST)

1. Changes were made to reflect the Turkey Point Nuclear Generating Station (PTN) current licensing basis. The Condensate Storage Tank (CST) System is a shared system between the two units. There are two CSTs shared between the two units and it is called the CST System.

2. Changes are made (additions, deletions, and/or changes) to the Improved Standard Technical Specification (ISTS) that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

3. Improved Standard Technical Specifications (ISTS) Required Action A.1 requires verification of the OPERABILITY of a backup water supply. As discussed in NEI 18-03, the industrys OPERABILITY guidance, OPERABILITY only applies to SSCs that a TS LCO requires to be OPERABLE. Therefore the OPERABILITY has been changed to that availability of the backup water supply must be verified.

4. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant information/value is inserted to reflect the current licensing basis.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

CST B 3.7.6 B 3.7 PLANT SYSTEMS B 3.7.6 Condensate Storage Tank (CST)

BASES System BACKGROUND The CST provides a safety grade source of water to the steam generators 1 for removing decay and sensible heat from the Reactor Coolant System s

(RCS). The CST provides a passive flow of water, by gravity, to the Auxiliary Feedwater (AFW) System (LCO 3.7.5). The steam produced is released to the atmosphere by the main steam safety valves or the atmospheric dump valves. The AFW pumps operate with a continuous recirculation to the CST.

When the main steam isolation valves are open, the preferred means of heat removal is to discharge steam to the condenser by the nonsafety grade path of the steam bypass valves. The condensed steam is returned to the CST by the condensate transfer pump. This has the advantage of conserving condensate while minimizing releases to the 1 environment.

INSERT 1 Because the CST is a principal component in removing residual heat from the RCS, it is designed to withstand earthquak es and other natural phenomena, including missiles that might be generated by natural 1 phenomena. The CST is designed to Seismic Category I to ensure availability of the feedwater supply. Feedwater is also available from alternate sources.

U 9.11.3 s

A description of the CST is found in the FSAR, Section [ 9.2.6] (Ref. 1). 1 2 System APPLICABLE The CST provides cooling water to remove decay heat and to cool down SAFETY U the unit following all events in the accident analysis as discussed in the ANALYSES 14 FSAR, Chapters [ 6] and [ 15] (Refs. 2 and 3, respectively). For 2 anticipated operational occurrences and accidents that do not affect the OPERABILITY of the steam generators, the analysis assumption is 1 generally 30 minutes at MODE 3, steaming through the MSSVs, followed by a cooldown to residual heat removal (RHR) entry conditions at the design cooldown rate. INSERT 2 The limiting event for the condensate volume is the large feedwater line break coincident with a loss of offsite power. Single failures that also affect this event include the following: 1

a. Failure of the diesel generator powering the motor driven AFW pump to the unaffected steam generator (requiring additional steam to drive 1 the remaining AFW pump turbine) and a

b. Failure of the steam driven AFW pump (requiring a longer time for 1

cooldown using only one motor driven AFW pump).

Westinghouse STS B 3.7.6-1 Rev. 5.0 1 Turk ey Point Unit 3 and Unit 4 Revision XXX

ITS 3.7.6 1

INSERT 1 Normally the atmospheric dump valves are used, and CST levels are maintained with the demineralized water storage tank makeup. The normal makeup for the CSTs is the demineralized water storage tank.

The CST System consists of two seismically designed CSTs each with a capacity of 250,000 gallons. A minimum indicated volume of 210,000 gallons is maintained for each unit in MODE 1, 2, or 3.

The minimum indicated volume includes an allowance for instrument indication uncertainties and for water deemed unusable because of vortex formation and the configuration of the discharge line.

1 INSERT 2 Normal water supply to the AFW pumps is from the two 250,000-gallon (nominal) CSTs, through locked open gate valves and check valves. Each tank contains a 210,000 gallons minimum indicated volume which assures a minimum usable volume of 195,331 gallons of demineralized water for the auxiliary feedwater pumps. The CST design sizing is based on allowing each unit to be taken from full power to MODE 3 following a loss of offsite power, and:

a. Kept at MODE 3 for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and then cooled to 350°F in 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months , at which point the Residual Heat Removal System will be put in service, or

b. Kept at MODE 3 for 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months .

Insert Page B 3.7.6-1

CST B 3.7.6 BASES APPLICABLE SAFETY ANALYSES (continued)

These are not usually the limiting failures in terms of consequences for these events.

A nonlimiting event considered in CST inventory determinations is a break in either the main feedwater or AFW line near where the two join.

This break has the potential for dumping condensate until terminated by operator action, since the Emergency Feedwater Actuation System would not detect a difference in pressure between the steam generators for this break location. This loss of condensate inventory is partially compensated for by the retention of steam generator inventory.

The CST satisfies Criteria 2 and 3 of 10 CFR 50.36(c)(2)(ii).

LCO 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months To satisfy accident analysis assumptions, the CST must contain sufficient cooling water to remove decay heat for [30 minutes] following a reactor 2 trip from 102% RTP, and then to cool down the RCS to RHR entry 1 conditions, assuming a coincident loss of offsite power and the most adverse single failure. In doing this, it must retain sufficient water to ensure adequate net positive suction head for the AFW pumps during cooldown, as well as account for any losses from the steam driven AFW pump turbine, or before isolating AFW to a broken line.

The CST level required is equivalent to a usable volume of

[110,000 gallons], which is based on holding the unit in MODE 3 for

[2] hours, followed by a cooldown to RHR entry conditions at [75]°F/hour. 1 This basis is established in Reference 4 and exceeds the volume required by the accident analysis. INSERT 3 (s)

System The OPERABILITY of the CST is determined by maintaining the tank 1 level at or above the minimum required level.

APPLICABILITY In MODES 1, 2, and 3, and in MODE 4, when steam generator is being relied upon for heat removal, the CST is required to be OPERABLE. 1 4, System In MODE 5 or 6, the CST is not required because the AFW System is not 1 required. System ACTIONS A.1 and A.2 System availability If the CST is not OPERABLE, the OPERABILITY of the backup supply 1 5 should be verified by administrative means within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and once every Availability 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter. OPERABILITY of the backup feedwater supply must 5 include verification that the flow paths from the backup water supply to available l R2 the AFW pumps are OPERABLE, and that the backup supply has the 5 System required volume of water available. The CST must be restored to 1 OPERABLE status within 7 days, because the backup supply may be Westinghouse STS B 3.7.6-2 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 Revision XXX

ITS 3.7.6 1

INSERT 3 The OPERABILITY of the CST with the minimum indicated volume ensures that sufficient water is available to maintain the Reactor Coolant System at MODE 3 conditions for approximately 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months or maintain the Reactor Coolant System at MODE 3 conditions for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months to cool down the Reactor Coolant System to below 350°F at which point the Residual Heat Removal System may be placed in operation.

Insert Page B 3.7.6-2

CST B 3.7.6 BASES ACTIONS (continued) performing this function in addition to its normal functions. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months availability Completion Time is reasonable, based on operating experience, to verify the OPERABILITY of the back up water supply. Additionally, verifying the 5 back up water supply every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is adequate to ensure the back up water supply continues to be available. The 7 day Completion Time is available reasonable, based on an OPERABLE back up water supply being 5

available, and the low probability of an event occurring during this time period requiring the CST.

INSERT 4 B.1 and B.2 System If the CST cannot be restored to OPERABLE status within the associated 1 Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4, without reliance on the steam 1 12 generator for heat removal, within [ 24] hours. The allowed Completion 2 Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. INSERT 5 SURVEILLANCE SR 3.7.6.1 REQUIREMENTS System This SR verifies that the CST contains the required volume of cooling water. (The required CST volume may be single value or a function of 1 INSERT 6 RCS conditions.) [ The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is based on operating experience and the need for operator awareness of unit evolutions that may affect the CST inventory between check s. Also, the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of other indications in the 2

control room, including alarms, to alert the operator to abnormal deviations in the CST level.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER S NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency 4

description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

INSERT 7 1 Westinghouse STS B 3.7.6-3 Rev. 5.0 1 Turk ey Point Unit 3 and Unit 4 Revision XXX

ITS 3.7.6 1

INSERT 4 Condition A provides remedial actions if the CST System is not OPERABLE due to indicated l R2 water volume (level) < 420,000 gallons with the opposite unit in MODE 1, 2, or 3 and remedial l actions if the CST system is not OPERABLE due to indicated volume < 210,000 gallons with the l opposite unit not in MODE 1, 2, or 3. A backup, non-safety source of water is the demineralized l water storage tank (DWST). The DWST is a 500,000 gallon, non-safety related source of demineralized water that is considered part of the primary makeup demineralized water system.

1 INSERT 5 This condition is modified by a Note stating that this condition is not applicable during a dual unit shut down.

C.1 and C.2 If the CST cannot be restored to OPERABLE status within the associated Completion Time, the units must be placed in a MODE in which the LCO does not apply. To achieve this status, the units must be placed in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and in MODE 4, within 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months . The allowed Completion Times allow for the orderly shutdown of one unit at a time and not jeopardize the stability of the electrical grid by imposing a dual unit shutdown.

This condition is modified by a Note stating that this condition is only applicable during a dual unit shut down.

1 INSERT 6 when the opposite unit is also in MODE 1, 2, or 3. A minimum indicated volume of 210,000 gallons is maintained for each unit in MODE 1, 2 or 3 so with both units in MODE 1, 2, or 3, 420,000 gallons indicated water volume is required. This volume provides margin over the analysis minimum required volume and includes an allowance for instrument indication uncertainties and for water deemed unusable because of vortex formation and the configuration of the discharge line.

Insert Page B 3.7.6-3a

ITS 3.7.6 1

INSERT 7 SR 3.7.6.2 verifies that the CST system contains the required volume of cooling water when the opposite unit is not in MODE 1, 2, or 3. A minimum indicated water volume of 210,000 gallons is maintained for the unit in MODE 1, 2 or 3. This volume provides margin over the analysis minimum required volume and includes an allowance for instrument indication uncertainties and for water deemed unusable because of vortex formation and the configuration of the discharge line.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Insert Page B 3.7.6-3b

CST B 3.7.6 BASES U

9.

11.3 REFERENCES

1. FSAR, Section [ 9.2.6] . 1 2 U

2. FSAR, Chapter [ 6] . 1 2 U

14

3. FSAR, Chapter [ 15] . 1 2 Westinghouse STS B 3.7.6-4 Rev. 5.0 1 Turk ey Point Unit 3 and Unit 4 Revision XXX

JUSTIFICATION FOR DEVIATIONS ITS 3.7.6 BASES, CONDENSATE STORAGE TANK (CST)

1. Changes are made (additions, deletions, and/or changes) to the Improved Standard Technical Specification (ISTS) Bases that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

3. Changes have been made (additions, deletions, and/or changes) to the ISTS that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description. Where an addition/deletion has occurred, subsequent alphanumeric designators have been changed for any applicable affected Required Actions, Surveillance Requirements, Functions, or Footnotes.

4. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.

5. Changes are made to be consistent with changes made to the Specification.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.7.6, CONDENSATE STORAGE TANK (CST)

There are no specific No Significant Hazards Considerations for this Specification.

Turkey Point Unit 3 and 4 Page 1 of 1

ATTACHMENT 7 ITS 3.7.7, COMPONENT COOLING WATER (CCW) SYSTEM

Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

ITS A01 ITS 3.7.7 PLANT SYSTEMS (CCW) 3/4.7.2 COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION A02 Two trains LCO 3.7.7 3.7.2 The Component Cooling Water System (CCW) shall be OPERABLE with:

Two L01

a. Three CCW pumps, and LA01

b. Two CCW heat exchangers.

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

ACTION:

Required Action A.1 NOTE: Enter applicable ACTIONS of LCO 3.4.1.3, "Reactor Coolant System - Hot Shutdown," for residual heat R2 Note removal loops made inoperable by CCW. L01

a. With only two CCW pumps with independent power supplies OPERABLE, restore the inoperable CCW pump to OPERABLE status within 30 days 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 in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

LA01 l R2 Action A b. With only one CCW pump OPERABLE or with two CCW pumps OPERABLE but not from l independent power supplies, restore two pumps from independent power supplies to OPERABLE l status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or in accordance with the Risk Informed Completion Time Program, 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 in COLD SHUTDOWN within the following R2 Action B 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Add proposed Required Action B.2 and associated Note L02 Action C c. With less than two CCW heat exchangers OPERABLE, restore two heat exchangers to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 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 in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . R2 immediately enter LCO 3.0.3 M01 With two CCW trains inoperable Add proposed ACTION C SURVEILLANCE REQUIREMENTS A03 4.7.2 The Component Cooling Water System (CCW) shall be demonstrated OPERABLE:

SR 3.7.7.4 a. In accordance with the Surveillance Frequency Control Program, by verifying that two heat l R2 exchangers and one pump are capable of removing design basis heat loads. l TURKEY POINT - UNITS 3 & 4 3/4 7-14 AMENDMENT NOS. 287 AND 281 Page 1 of 2

ITS A01 ITS 3.7.7 SURVEILLANCE REQUIREMENTS (Continued)

Add proposed SR 3.7.7.1 Note A04 SR 3.7.7.1 b. 1) In accordance with the Surveillance Frequency Control Program verify that each valve (manual, power-operated, or automatic) servicing safety-related equipment A05 that is not locked, sealed, or otherwise secured in position is in its correct in the flowpath position.

SR 3.7.7.5 2) In accordance with the Surveillance Frequency Control Program verify by a L05 l R2 performance test the heat exchanger surveillance curves.* l l

A05

c. In accordance with the Surveillance Frequency Control Program during shutdown, by verifying that: that is not locked, sealed, or otherwise in the flowpath secured in position L03 SR 3.7.7.2 1) Each automatic valve servicing safety-related equipment actuates to its correct position on a SI test signal, and an actual or simulated actuation L04 SR 3.7.7.3 2) Each Component Cooling Water System pump starts automatically on a SI test signal.

SR 3.7.7.3 3) Interlocks required for CCW operability are OPERABLE.

A06

Technical specification 4.7.2.b.2 is not applicable for entry into MODE 4 or MODE 3, provided that: l R2 l

SR 3.7.7.5 1) Surveillance 4.7.2.b.2 is performed no later than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after reaching a Reactor l Note Coolant System Tavg of 547°F, and l

2) MODE 2 shall not be entered prior to satisfactory performance of this surveillance. l TURKEY POINT - UNITS 3 & 4 3/4 7-15 AMENDMENT NOS. 263 AND 258 Page 2 of 2

DISCUSSION OF CHANGES ITS 3.7.7, COMPONENT COOLING WATER (CCW) SYSTEM l R2 ADMINISTRATIVE CHANGES A01 In the conversion of the Turkey Point Nuclear Generating Station (PTN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 5.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal.

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 3.7.2 states that the Component Cooling Water System (CCW) shall be OPERABLE with: a. Three CCW pumps, and b. Two CCW Heat Exchangers.

ITS LCO 3.7.7 states that two CCW trains shall be OPERABLE. This changes the CTS by placing the CCW system components into separate trains.

The purpose of CTS 3.7.2 is to ensure that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single active failure, is consistent with the assumptions used in the safety analyses.

This change is acceptable because the LCO requirements continue to ensure that the structures, systems, and components are maintained consistent with the safety analyses and licensing basis. This is a change in presentation only and does not result in technical changes to the CTS.

A03 CTS 3.7.2 requires the Component Cooling Water System to be OPERABLE in MODES 1, 2, 3, and 4 and provide actions for various levels of system inoperability. However, CTS does not explicitly provide actions for all degraded conditions (e.g., three CCW pumps inoperable). ITS 3.7.7 ACTION C provides R2 similar actions for the condition when two CCW trains are inoperable. This changes the CTS by explicitly stating when entry into LCO 3.0.3 is required.

The purpose of CTS actions is to ensure applicable Actions are taken based on the level of CCW System degradation and when multiple system components are inoperable that results in a loss the required safety function, a plant shutdown is R2 required. CTS 3.7.2 Actions a and b provide remedial actions when one or two CCW pumps are inoperable or when at least two CCW pumps are not powered from independent power supplies.

Except for CTS 3.7.2 Action c, which provides actions for the condition of less than two OPERABLE CCW heat exchangers, CTS 3.7.2 does not provide explicit actions for conditions when multiple system components are inoperable (e.g.,

three inoperable CCW pumps) that represents a loss of cooling capability to both CCW cooling loops (i.e., trains). Therefore, CTS 3.0.3 would apply. Due to the R2 multiple CCW cooling loop configurations available, ITS 3.7.7 ACTION C ensures proper application of the ITS in a condition when neither CCW train can provide the required safety function, which requires at least one OPERABLE cooling pump and two OPERABLE heat exchangers aligned to at least one cooling loop.

Turkey Point Unit 3 and Unit 4 Page 1 of 8

DISCUSSION OF CHANGES ITS 3.7.7, COMPONENT COOLING WATER (CCW) SYSTEM l R2 ITS 3.7.7 ACTION C requires that when two CCW trains are inoperable (e.g.,

three inoperable CCW pumps or less than two CCW heat exchangers are OPERABLE) LCO 3.0.3 is entered immediately. This changes the CTS R2 requirements in presentation and format only. This change is designated as administrative because it does not result in technical changes to the CTS.

A04 CTS 4.7.2.b.1) does not contain explicit guidance concerning CCW system OPERABILITY when isolating CCW flow to individual components. ITS Surveillance Requirement (SR) 3.7.7.1 contains a Note, which states, "Isolation of CCW flow to individual components does not render the CCW system inoperable." This changes the CTS by adding an allowance that is not explicitly stated in the CTS.

The purpose of CTS 4.7.2.b.1) is to provide assurance that CCW is available to the appropriate plant components. This change is acceptable because by current use and application of the CTS, isolation of a component supplied with CCW does not necessarily result in a CCW system being considered inoperable, but the respective component may be declared inoperable for its system. This change clarifies this application.

This change is designated as administrative because it does not result in technical changes to the CTS.

A05 CTS 4.7.2.b.1) requires verification that each CCW valve (manual, power operated, or automatic) servicing safety related equipment that is not locked, sealed, or otherwise secured in position, is in its correct position. CTS 4.7.2.c.1) requires verification that each CCW automatic valve servicing safety related equipment actuates to its correct position on a Safety Injection test signal. ITS SR 3.7.7.1 requires verification that each CCW manual, power operated, and automatic valve in the flow path servicing safety related equipment that is not locked, sealed, or otherwise secured in position, is in the correct position. ITS SR 3.7.7.2 requires verification that each CCW automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal. This changes the CTS by adding the words "in the flow path" to CTS 4.7.2.b.1) (ITS SR 3.7.7.1) and replacing the words "servicing safety related equipment" with "in the flow path" in CTS 4.7.2.c.1) (ITS SR 3.7.7.2).

The purpose of CTS 4.7.2.b.1) is to ensure all valves in the CCW flow path are in the correct position. The purpose of CTS 4.7.2.c.1) is to provide assurance that each CCW automatic valve actuates to its correct position. The addition of the words "in the flow path" to CTS 4.7.2.b.1) (ITS SR 3.7.7.1) does not change the intent of the Surveillance Requirement. Each manual, power operated, and automatic valve servicing safety related equipment that is not locked, sealed, or otherwise secured in position will continue to be verified to be in the correct position. The removal of the words "servicing safety related equipment" in CTS 4.7.2.c.1) (ITS SR 3.7.7.2) does not change the intent of the Surveillance Requirement. Each CCW automatic valve in the flow path that is not locked, sealed or otherwise secured in position, will still be checked to ensure it actuates to the correct position on an actual or simulated Safety Injection actuation signal.

Turkey Point Unit 3 and Unit 4 Page 2 of 8

DISCUSSION OF CHANGES ITS 3.7.7, COMPONENT COOLING WATER (CCW) SYSTEM l R2 This change is designated as administrative because it does not result in technical changes to the CTS.

A06 CTS 4.7.2.c.2) requires verification that each CCW System pump starts automatically on a SI test signal. CTS 4.7.2.c.3) requires verification that the interlocks required for CCW OPERABILITY are OPERABLE. ITS SR 3.7.7.3 requires verification that each CCW pump starts automatically on an actual or simulated actuation signal. This changes the CTS by removing the duplicative requirement to verify that the interlocks required for OPERABILITY are OPERABLE, aligning those interlocks to the requirement to verify the CCW pumps start automatically on a SI test signal.

The purpose of CTS 4.7.2.c.3) is to verify the interlocks required for system OPERABILITY are OPERABLE. The associated interlock is the starting of the C CCW pump automatic start on an SI test signal when either CCW pump A or B breaker is racked out and the C CCW pump power supply is aligned to the racked out pump's power supply. The C CCW pump is interlocked with the A and B CCW pump such that the C CCW pump can only start automatically on an SI signal when aligned to replace either the A or B CCW pump. Because both of the CTS surveillances are associated with the automatic starting of the CCW pumps on an SI signal, the duplicative surveillance is being deleted. This change is acceptable because the test requirements and frequency regarding the CCW pumps remain the same. This change is designated as administrative because it does not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 3.7.2 Action c requires that with less than two CCW heat exchangers OPERABLE, restore two heat exchangers to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 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 in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . ITS 3.7.7 ACTION C requires that with two CCW trains R2 inoperable to enter LCO 3.0.3 immediately, which includes the condition of less than two CCW heat exchanges OPERABLE. This changes the CTS by explicitly stating when entry into LCO 3.0.3 is required which results in the additional requirement to be MODE 4 within a specified time period for the condition with less than two OPERABLE CCW heat exchangers.

The purpose of CTS actions is to ensure applicable Actions are taken based on the level of CCW System degradation and when multiple system components are inoperable that results in a loss of the required safety function, a plant shutdown is required. CTS 3.7.2 Action c provides actions for the condition where the number of CCW heat exchangers OPERABLE will not support the ability of the CCW System to perform its specified safety function. CTS 3.7.2 Action c R2 requires restoration of two CCW heat exchangers to OPERABLE status within one hour 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 in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months (37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months total). ITS 3.7.7 ACTION C provides a requirement to immediately enter LCO 3.0.3 when two CCW trains are inoperable, which includes the condition of less than two CCW heat exchanges OPERABLE. LCO 3.0.3 requires that within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action shall be initiated to Turkey Point Unit 3 and Unit 4 Page 3 of 8

DISCUSSION OF CHANGES ITS 3.7.7, COMPONENT COOLING WATER (CCW) SYSTEM l R2 place the unit, as applicable, to be in: a) MODE 3 in 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months (i.e., Hot Standby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , b) MODE 4 in 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months (i.e., Hot Shutdown within the R2 following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ), and c) MODE 5 in 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months (i.e., Cold Shutdown within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

In addition to the CTS 3.0.3 Actions that are equivalent to CTS 3.7.2 Action c, application of ITS LCO 3.0.3 (CTS 3.0.3) also requires the unit to be in MODE 4 in 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months , which is more restrictive than CTS 3.7.2 Action c actions and is R2 appropriate because this condition represents a loss of the required safety function. This change is designated as more restrictive because the remedial actions are more stringent in ITS than in CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.7.2 requires the Component Cooling Water System (CCW) to be OPERABLE with: a. Three CCW pumps, and b. Two CCW heat exchangers (Changed to two CCW pumps by DOC L01). CTS 3.7.2, Action b, states that with l R2 only one CCW pump OPERABLE or with two CCW pumps OPERABLE, but not l from independent power supplies, restore two pumps from independent power l supplies to OPERABLE status. ITS LCO 3.7.7 requires two CCW trains to be l OPERABLE but does not define the components, independence, and the l associated flow path that comprise an OPERABLE CCW train. This changes the CTS by moving the description of the CCW trains to the Bases.

The removal of these details which are related to system design from the Technical Specifications is acceptable because this type of information is not necessary to be included to provide adequate protection of public health and safety. The ITS still retains the requirement for both CCW trains to be OPERABLE. Also, this change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5.

This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications.

l R2 LESS RESTRICTIVE CHANGES L01 (Category 1 - Relaxation of LCO Requirements) CTS 3.7.2 states that the Component Cooling Water System (CCW) shall be OPERABLE with: a. Three CCW pumps, and b. Two CCW heat exchangers. In addition, CTS 3.7.2 Action

a. requires that with only two CCW pumps with independent power supplies Turkey Point Unit 3 and Unit 4 Page 4 of 8

DISCUSSION OF CHANGES ITS 3.7.7, COMPONENT COOLING WATER (CCW) SYSTEM l R2 OPERABLE to restore the inoperable CCW pump to OPERABLE status within 30 days. ITS 3.7.7 states that two CCW trains shall be OPERABLE. This changes the CTS by reducing the number of CCW pumps required to be OPERABLE from three to two and eliminating the associated action requirement to restore a single CCW pump to OPERABLE status within 30 days.

The purpose of CTS 3.7.2 is to provide the lowest functional capability or performance levels of equipment required for safe operation of the facility and the purpose of the CTS action requirement is to provide appropriate remedial actions when the LCO is not met consistent with the requirements of 10 CFR 50.36(c)(2)(i).

Required Actions are used to establish remedial measures that must be taken in response to degraded conditions in order to minimize risk associated with continued plant operation while providing time to repair inoperable equipment. The Required Actions are consistent with safe operation under the specified Condition, considering the OPERABILITY status of the remaining CCW equipment, a reasonable time for repairs or replacement of required equipment, and the low probability of a design basis accident occurring during the time period. The redundant cooling capacity of this system, assuming a single active failure, is consistent with the assumptions used in the safety analyses. The CCW System design includes an installed spare pump and heat exchanger that can be aligned to either CCW loop to provide adequate cooling to the required components. One pump and two heat exchangers provide the heat removal capability for accidents that have been analyzed. This change is acceptable because the LCO requirements continue to ensure that the number of CCW pumps required to be OPERABLE maintains the minimum cooling capacity of the system assuming a single active failure and maintained consistent with the safety analyses and licensing basis. In addition, eliminating the TS requirement for the third CCW pump to be OPERABLE is acceptable because other regulatory requirements ensure the degradation of structures, systems, and components (SSCs) will not result in undue risk to the health and safety of the public. Pursuant to 10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants, PTN monitors the performance or condition of SSCs, against established Maintenance Rule goals, in a manner sufficient to provide reasonable assurance that SSCs are capable of fulfilling their intended functions.

When the performance or condition of an SSC does not meet established goals, appropriate corrective action will be taken to ensure the risk to the health and safety of the public is minimized. Therefore, NRC regulations contain the necessary programmatic requirements to ensure SSCs are capable of fulfilling their intended functions and appropriate compensatory actions are taken to minimize risk to the health and safety of the public. This change is designated as less restrictive because less stringent LCO requirements and actions are being applied in the ITS than were applied in the CTS.

L02 (Category 4 - Relaxation of Required Action) CTS 3.7.2, ACTION b identifies a R2 degraded conditions of the CCW system and provides specific Completion Times to restore the degraded condition or commence a unit shutdown. If a unit shutdown is required, CTS 3.7.2 Action b requires the unit be in HOT STANDBY R2 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 .

ITS 3.7.7 ACTION B, similarly, states that if the Required Action and associated R2 Completion Time of Condition A not met, to be in MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE Turkey Point Unit 3 and Unit 4 Page 5 of 8

DISCUSSION OF CHANGES ITS 3.7.7, COMPONENT COOLING WATER (CCW) SYSTEM l R2 4 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and is modified by a Note stating that LCO 3.0.4.a is not applicable when entering MODE 4. This changes the CTS by permitting a Required Action end state of HOT SHUTDOWN (MODE 4) rather that an end state of COLD SHUTDOWN (MODE 5).

One purpose of CTS 3.7.2, ACTION b is to provide an end state, a condition that R2 the reactor must be placed in, if the Required Actions allowing remedial measures to be taken in response to the degraded conditions with continued operation are not met. End states are usually defined based on placing the unit into a MODE or condition in which the Technical Specification Limiting Condition for Operation (LCO) is not applicable. MODE 5 is the current end state for LCOs that are applicable in MODES 1 through 4. This change is acceptable because the risk of the transition from MODE 1 to MODES 4 or 5 depends on the availability of alternating current (AC) sources and the ability to remove decay heat such that remaining in MODE 4 may be safer. During the realignment from MODE 4 to MODE 5, there is an increased potential for loss of shutdown cooling and loss of inventory events. Decay heat removal following a loss-of-offsite power event in MODE 5 is dependent on AC power for shutdown cooling whereas, in MODE 4, the turbine driven auxiliary feedwater (AFW) pump will be available. Therefore, transitioning to MODE 5 is not always the appropriate end state from a risk perspective. Thus, for specific TS conditions, Westinghouse Topical Report WCAP-16294-A R1 (ADAMS Accession No. ML103430249) justifies MODE 4 as an acceptable alternate end state to Mode 5. The proposed change to the Technical Specifications will allow time to perform short-duration repairs, which currently necessitate exiting the original mode of applicability. The MODE 4 TS end state is applied, and risk is assessed and managed in accordance with Title 10 of the Code of Federal Regulations (10 CFR) Section 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants." This proposed change is consistent with NRC approved TSTF-432-A Revision 1 (ADAMS Accession No. ML103360003), noticed for availability by the NRC in the Federal Register (77 FR 27814) on May 11, 2012. The NRC's approval of WCAP-16294-A included four limitations and conditions on its use as identified in Section 4.0 of the NRC Safety Evaluation associated with WCAP-16294-A. Implementation of these stipulations were addressed in the Bases of TSTF-432-A. Florida Power & Light implemented these limitations and conditions at PTN in the adoption of the associated TSTF-432-A Bases. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS.

L03 (Category 6 - Relaxation of Surveillance Requirement Acceptance Criteria)

CTS 4.7.2.c.1 requires verification that each automatic valve servicing safety-related equipment actuates to its correct position on a safety injection (SI) test signal. ITS SR 3.7.7.2 requires verification that each CCW automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal. This changes the CTS by not requiring automatic valves that are locked, sealed or otherwise secured in position to be tested to verify that they automatically actuate to their correct position.

Turkey Point Unit 3 and Unit 4 Page 6 of 8

DISCUSSION OF CHANGES ITS 3.7.7, COMPONENT COOLING WATER (CCW) SYSTEM l R2 The purpose of CTS 4.7.2.c.1 is to provide assurance that the valves in the flowpath required to actuate in case of a design basis accident (DBA) actuate to the correct position. This change is acceptable because it has been determined that the relaxed SR acceptance criteria are not necessary for verification that the equipment used to meet the LCO can perform its required functions. Valves already in the correct position and are locked, sealed, or otherwise secured in position are not required to have the position verified or to be tested to automatically actuate because, in case of a DBA, the valves are already in the required position and secured to prevent changing from the required position.

This change is designated as less restrictive because less stringent SRs are being applied in the ITS than were applied in the CTS.

L04 (Category 6 - Relaxation of Surveillance Requirement Acceptance Criteria)

CTS 4.7.2.c.1 and CTS 4.7.2.c.2 require verifying that each automatic valve servicing safety-related equipment and each CCW system pump starts automatically on a SI test signal. ITS SR 3.7.7.2 and ITS SR 3.7.7.3 specify that the signal may be from either an "actual" or simulated (i.e., test) signal. This changes the CTS by explicitly allowing the use of either an actual or simulated signal for the test.

The purpose of CTS 4.7.2.c.1 and CTS 4.7.2.c.2 is to ensure that the automatic valves servicing safety related equipment and CCW system pumps operate correctly upon receipt of an actuation signal. This change is acceptable because it has been determined that the relaxed SR acceptance criteria are not necessary for verification that the equipment used to meet the LCO can perform its specified safety functions. Equipment cannot discriminate between an "actual," "simulated,"

or "test" signal and, therefore, the results of the testing are unaffected by the type of signal used to initiate the test. The change also allows a simulated signal to be used, if necessary. This change is designated as less restrictive because less stringent SRs are being applied in the ITS than were applied in the CTS.

L05 (Category 8 - Deletion of Surveillance Requirement Shutdown Performance l R2 Requirements) CTS 4.7.2.c requires testing of specified CCW equipment in l accordance with the Surveillance Frequency Control Program "during shutdown." l l

ITS SRs 3.7.7.2 and 3.7.7.3 do not contain a MODE restriction related to when l

this testing may be performed. This changes the CTS by removing the restriction l on surveillance performance during specific MODES. l l

The purpose of CTS 4.7.2.c is to demonstrate that, upon receipt of an associated l actuation signal, each CCW automatic valve in the flow path servicing safety l related equipment can actuate to its correct position and each CCW pump can l automatically start, which also verifies required interlocks are OPERABLE. The l control of unit conditions appropriate to perform the test is an issue for l procedures and scheduling and has been determined by the NRC to be l l

unnecessary as a Technical Specification restriction. As indicated in Generic l

Letter 91-04, removal of this specific restriction is consistent with the vast l majority of other Technical Specification Surveillances that do not dictate unit l conditions for the Surveillance. The proposed change is acceptable because it l does not change the method of test or frequency of the affected surveillances. l The proposed change only deletes the requirement to perform this testing during l Turkey Point Unit 3 and Unit 4 Page 7 of 8

DISCUSSION OF CHANGES ITS 3.7.7, COMPONENT COOLING WATER (CCW) SYSTEM l R2 shutdown conditions. In addition, allowing this testing to be performed either at l R2 refueling, shutdown, or at power does not affect the applicable safety analysis l conclusions and allows shutdown activities to be planned which will aid to reduce l risk and increase equipment availability during shutdowns. Thus, the proposed l change will continue to provide adequate assurance the required components l l

are routinely tested to ensure system OPERABILITY while providing some l

additional flexibility in planning and scheduling the required testing. In addition, l due to system designs that allow for safe testing at power, the proposed change l will not adversely affect the safe operation of the plant. The proposed change is l designated as less restrictive because the Surveillance may be performed during l plant conditions other than shutdown. l Turkey Point Unit 3 and Unit 4 Page 8 of 8

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

CTS CCW System 3.7.7 3.7 PLANT SYSTEMS 3.7.7 Component Cooling Water (CCW) System 3.7.2 LCO 3.7.7 Two CCW trains shall be OPERABLE.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.7.2 Action b A. One CCW train A.1 --------------NOTE-------------- R2 inoperable. Enter applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops -

MODE 4," for residual heat removal loops made inoperable by CCW.

Restore CCW train to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OPERABLE status.

[OR 3 In accordance with the Risk Informed Completion Time Program] 3 3.7.2 Action b, B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 3.7.2 Action c.

associated Completion Time of Condition A not AND R2 met.

B.2 --------------NOTE--------------

LCO 3.0.4.a is not DOC L02 applicable when entering MODE 4.

DOC L02 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months R2 INSERT 1 R2 Westinghouse STS 3.7.7-1 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 AMENDMENT Nos. XXX and YYY

CTS ITS 3.7.7 2

INSERT 1 DOC A03 C. Two CCW trains C.1 Enter LCO 3.0.3. Immediately R2 DOC M01 inoperable.

Insert Page 3.7.7-1

CTS CCW System 3.7.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.7.2.b.1) SR 3.7.7.1 -------------------------------NOTE------------------------------

Isolation of CCW flow to individual components does not render the CCW System inoperable.

Verify each CCW manual, power operated, and [ 31 days 3 automatic valve in the flow path servicing safety related equipment, that is not locked, sealed, or OR 3 otherwise secured in position, is in the correct position. In accordance with the Surveillance Frequency Control Program ] 3 4.7.2.c.1) SR 3.7.7.2 Verify each CCW automatic valve in the flow path [ [18] months 3 that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an OR 3 actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program ] 3 4.7.2.c.2) SR 3.7.7.3 Verify each CCW pump starts automatically on an [ [18] months 3 4.7.2.c.3) actual or simulated actuation signal.

OR 3 In accordance with the Surveillance Frequency Control Program ] 3 INSERT 2 4 l R2 Westinghouse STS 3.7.7-2 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 AMENDMENT Nos. XXX and YYY

CTS ITS 3.7.7 4

INSERT 2 l R2 l

l 4.7.2.a SR 3.7.7.4 Verify two CCW heat exchangers and one CCW In accordance l pump are capable of removing design basis heat with the l l

loads. Surveillance l Frequency l Control Program l l

l 4.7.2.b.2 and SR 3.7.7.5 -------------------------------NOTE------------------------------ l Footnote

l Not required to be performed until 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after reaching Reactor Coolant System Tavg 547°F but l prior to MODE 2. l l

l l

Verify the CCW heat exchanger performance curves In accordance l by performance test. with the l Surveillance l Frequency l Control Program l l

l Insert Page 3.7.7-2 l R2

JUSTIFICATION FOR DEVIATIONS ITS 3.7.7, COMPONENT COOLING WATER (CCW) SYSTEM

1. Changes are made (additions, deletions, and/or changes) to the Improved Standard Technical Specification (ISTS) that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS is based on a single plant design with two redundant CCW trains consisting of a CCW pump and associated heat exchanger, and a swing pump that can be aligned to either CCW train. The PTN design also includes three CCW pumps but, in addition, includes three CCW heat exchangers downstream of the CCW pumps via a common discharger header. The safety analysis requires two of three heat exchangers and one of three CCW pumps to be OPERABLE to support the safety function. Due to multiple CCW cooling loop configurations available, ITS 3.7.7 ACTION C is added to ensure proper application of the ITS in a condition when neither CCW train can provide the required safety function, which requires at least R2 one OPERABLE cooling pump and two OPERABLE heat exchangers aligned to at least one cooling loop.

3. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed, and the proper plant specific information/value is inserted to reflect the current licensing basis.

4. CTS 4.7.2.a and CTS 4.7.2.b.2, including footnote *, provide testing requirements for l R2 the CCW heat exchangers and CCW pumps. The purpose of CTS 4.7.2.a is to l ensure the CCW System is capable of removing design basis heat loads and the l purpose of CTS 4.7.2.b.2 is to quantify the effectiveness of the CCW heat l l

exchangers by performance testing. ITS SR 3.7.7.4 and SR 3.7.7.5 retain the CTS l

requirements. The testing required by CTS 4.7.2.a and CTS 4.7.2.b.2 and l associated frequency are the same as those required by ITS SR 3.7.7.4 and ITS l SR 3.7.7.5, respectively. Therefore, no technical change is being proposed. l Turkey Point Unit 3 and Unit 4 Page 1 of 1

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

CCW System B 3.7.7 B 3.7 PLANT SYSTEMS B 3.7.7 Component Cooling Water (CCW) System BASES BACKGROUND The CCW System provides a heat sink for the removal of process and operating heat from safety related components during a Design Basis Accident (DBA) or transient. During normal operation, the CCW System also provides this function for various nonessential components, as well as the spent fuel storage pool. The CCW System serves as a barrier to the release of radioactive byproducts between potentially radioactive systems and the Service Water System, and thus to the environment.

The A typical CCW System is arranged as two independent, full capacity cooling loops, and has isolatable nonsafety related components. Each The CCW System safety related train includes a full capacity pump, surge tank, heat 1 design also includes three heat exchanger, piping, valves, and instrumentation. Each safety related train exchangers that are is powered from a separate bus. An open surge tank in the system R2 common to both cooling loops.

provides pump trip protective functions to ensure that sufficient net positive suction head is available. The pump in each train is automatically started on receipt of a safety injection signal, and all nonessential components are isolated.

INSERT 1 Additional information on the design and operation of the system, along U

9.3.2 with a list of the components served, is presented in the FSAR, 1 Section [9.2.2] (Ref. 1). The principal safety related function of the CCW 2 System is the removal of decay heat from the reactor via the Residual Heat Removal (RHR) System. This may be during a normal or post accident cooldown and shutdown.

APPLICABLE The design basis of the CCW System is for one CCW train to remove the SAFETY post loss of coolant accident (LOCA) heat load from the containment ANALYSES sump during the recirculation phase, with a maximum CCW temperature of [120]°F (Ref. 2). The Emergency Core Cooling System (ECCS) LOCA and containment OPERABILITY LOCA each model the maximum and minimum performance of the CCW System, respectively. The normal 1

temperature of the CCW is [80]°F, and, during unit cooldown to MODE 5 (Tcold < [200]°F), a maximum temperature of 95°F is assumed. This prevents the containment sump fluid from increasing in temperature during the recirculation phase following a LOCA, and provides a gradual reduction in the temperature of this fluid as it is supplied to the Reactor Coolant System (RCS) by the ECCS pumps.

INSERT 2 The CCW System is designed to perform its function with a single failure of any active component, assuming a loss of offsite power.

Westinghouse STS B 3.7.7-1 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 Revision XXX

ITS 3.7.7 1

INSERT 1 In addition, the CCW System design includes an additional CCW pump that can swing from one train to the other, with interlocks to ensure the swing pump can serve as a backup to either CCW train. Each of the two standby pumps provides 100% backup, during normal operation.

The CCW pumps and heat exchangers are arranged such that any combination of pumps and R2 heat exchangers can supply either CCW cooling loop. The CCW trains are normally cross-tied at common suction and discharge headers and can be separated by closing crosstie valves between headers.

The CCW heat loads are transferred by the CCW System to the Intake Cooling Water (ICW)

System. The CCW System serves as an intermediate system to provide a barrier between the ICW and the CCW cooled components. This barrier prevents any potential leakage of radioactive fluid into the environment or any saltwater intrusion into the Reactor Coolant System.

The CCW head tank accommodates normal expansion and limited in-leakage of water. The head tank and surge tank combine to accommodate contraction and ensure a continuous component cooling water supply until the leak can be isolated.

Each pump is automatically started on receipt of a start signal from the emergency bus load sequencer, and all nonessential components are isolated. The emergency bus load sequencer is actuated by a loss of offsite power (LOOP), a safety injection (SI) signal on its associated unit, a SI from the opposite unit, or a combination LOOP/LOCA. CCW pumps A and B are sequenced on by their associated train sequencer for these actuation signals except for the opposite units SI signal without a LOOP. A swing 4.16 kV emergency bus provides power to the swing CCW pump and can be manually aligned to either the A or B train 4.16 kV bus. The swing CCW pump is sequenced based on which AC electrical power train the 4.16 kV swing bus is aligned to. In addition, the swing pump is interlocked with the A and B CCW pumps such that for a pump start signal to initiate the swing pump on a LOOP or SI signal, the supply breaker associated with the train to which the swing pump is aligned must be open and racked out.

With offsite power available the CCW pumps receive a start signal 25 seconds (nominally) after the sequencer load timing begins; 41 seconds (nominally) with offsite power unavailable.

Insert Page B 3.7.7-1a

ITS 3.7.7 1

INSERT 2 one pump and two heat exchangers to provide the heat removal capability for accidents that have been analyzed. Peak CCW System operating temperatures occur during post-accident operations due to elevated containment temperatures and unrestricted heat rejection into the CCW System. A calculated maximum CCW System supply temperature of 158.5°F is acceptable for post-accident operation.

Insert Page B 3.7.7-1b

CCW System B 3.7.7 BASES APPLICABLE SAFETY ANALYSES (continued)

The CCW System also functions to cool the unit from RHR entry conditions (Tcold < [350]°F), to MODE 5 (Tcold < [200]°F), during normal and post accident operations. The time required to cool from [350]°F to 2

[200]°F is a function of the number of CCW and RHR trains operating.

INSERT 3 One CCW train is sufficient to remove decay heat during subsequent operations with Tcold < [200]°F. This assumes a maximum service water 1 temperature of [95]°F occurring simultaneously with the maximum heat loads on the system.

The CCW System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO The CCW trains are independent of each other to the degree that each has separate controls and power supplies and the operation of one does not depend on the other. In the event of a DBA, one CCW train is required to provide the minimum heat removal capability assumed in the safety analysis for the systems to which it supplies cooling water. To ensure this requirement is met, two trains of CCW must be OPERABLE.

At least one CCW train will operate assuming the worst case single active failure occurs coincident with a loss of offsite power.

A CCW train is considered OPERABLE when:

An independent ;

a. The pump and associated surge tank are OPERABLE and

b. Two common heat exchanges are OPERABLE; and 1 c

b. The associated piping, valves, heat exchanger, and instrumentation and controls required to perform the safety related function are OPERABLE.

INSERT 4 The isolation of CCW from other components or systems not required for the individual 1 safety may render those components or systems inoperable but does not affect the OPERABILITY of the CCW System.

APPLICABILITY In MODES 1, 2, 3, and 4, the CCW System is a normally operating system, which must be prepared to perform its post-accident safety functions, primarily RCS heat removal, which is achieved by cooling the RHR heat exchanger.

Although the LCO for the CCW System is not applicable in MODES 5 and 6, the capability of the CCW System to perform its necessary related support functions may be required for OPERABILITY of supported systems.

Westinghouse STS B 3.7.7-2 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 Revision XXX

ITS 3.7.7 1

INSERT 3 One CCW pump, two CCW heat exchangers will remove design basis accident heat loads and maintain ICW heat exchangers 100°F. Additionally, cases were run to demonstrate cooldown times for normal operations and refueling operations.

1 INSERT 4 The swing CCW pump can be substituted for a normal CCW pump, provided the swing CCW pump is aligned to the applicable CCW loop and the power supply for the swing pump is aligned to the same AC electrical power distribution train as the pump it is replacing.

Insert Page B 3.7.7-2

CCW System B 3.7.7 BASES 1

ACTIONS A.1 Required Action A.1 is modified by a Note indicating that the applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops - MODE 4,"

be entered if an inoperable CCW train results in an inoperable RHR loop.

This is an exception to LCO 3.0.6 and ensures the proper actions are (e.g., an inoperable CCW pump taken for these components.

associated with the train or associated CCW pump not 1 capable of supplying at least two If one CCW train is inoperable, action must be taken to restore heat exchangers), OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months [or in accordance with the Risk 2

Informed Completion Time Program]. In this Condition, the remaining OPERABLE CCW train is adequate to perform the heat removal function. R2 The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is reasonable, based on the redundant capabilities afforded by the OPERABLE train, and the low probability of a DBA occurring during this period.

B.1 and B.2 If the CCW train cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which overall plant risk is reduced. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Remaining within the Applicability of the LCO is acceptable to accomplish short duration repairs to restore inoperable equipment because the plant risk in MODE 4 is similar to or lower than MODE 5 (Ref. 3). In MODE 4 the steam generators and Residual Heat Removal System are available to remove decay heat, which provides diversity and defense in depth. As stated in Reference 3, the steam turbine driven auxiliary feedwater pump must be available to remain in MODE 4. Should steam generator cooling be lost while relying on this Required Action, there are preplanned actions to ensure long-term decay heat removal. Voluntary entry into MODE 5 may be made as it is also acceptable from a risk perspective.

Required Action B.2 is modified by a Note that states that LCO 3.0.4.a is R2 not applicable when entering MODE 4. This Note prohibits the use of LCO 3.0.4.a to enter MODE 4 during startup with the LCO not met.

However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering MODE 4, and establishment of risk management actions, if appropriate. LCO 3.0.4 is not applicable to, and the Note does not preclude, 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.

Westinghouse STS B 3.7.7-3 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 Revision XXX

CCW System B 3.7.7 BASES ACTIONS (continued)

The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

INSERT 5 3 R2 SURVEILLANCE SR 3.7.7.1 REQUIREMENTS This SR is modified by a Note indicating that the isolation of the CCW flow to individual components may render those components inoperable but does not affect the OPERABILITY of the CCW System.

Verifying the correct alignment for manual, power operated, and automatic valves in the CCW flow path provides assurance that the proper flow paths exist for CCW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing. This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This Surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position.

[ The 31 day Frequency is based on engineering judgment, is consistent with the procedural controls governing valve operation, and ensures correct valve positions. 2 OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency 4 description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

SR 3.7.7.2 This SR verifies proper automatic operation of the CCW valves on an actual or simulated actuation signal. The CCW System is a normally operating system that cannot be fully actuated as part of routine testing during normal operation. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under Westinghouse STS B 3.7.7-4 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 Revision XXX

ITS 3.7.7 3

INSERT 5 R2 C.1 R2 If neither CCW train can provide the required safety function, which requires at least one OPERABLE cooling pump and two OPERABLE heat exchangers aligned to at least one cooling R2 loop, the plant is in a condition outside the accident analyses; therefore, LCO 3.0.3 must be entered immediately.

Insert Page B 3.7.7-4

CCW System B 3.7.7 BASES SURVEILLANCE REQUIREMENTS (continued) administrative controls. [ The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillance when 2 performed at the [18] month Frequency. Therefore, the Frequency is acceptable from a reliability standpoint.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency 4 description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

SR 3.7.7.3 This SR verifies proper automatic operation of the CCW pumps on an actual or simulated actuation signal. The CCW System is a normally operating system that cannot be fully actuated as part of routine testing during normal operation. [ The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillance when 2 performed at the [18] month Frequency. Therefore, the Frequency is acceptable from a reliability standpoint.

OR INSERT 6 1 R2 The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency 4 description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

INSERT 7 5 l R2 Westinghouse STS B 3.7.7-5 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 Revision XXX

ITS 3.7.7 1

INSERT 6 R2 The CCW swing pump (C pump) is interlocked to prevent starting if CCW pumps A and B are aligned for starting. For a start signal to initiate starting the swing pump on a LOOP or SI signal, the supply breaker for the CCW pump, associated with the AC electrical power distribution train to which it is aligned, must be open and racked out. Testing the automatic starting of the swing CCW pump includes testing this interlock.

5 l R2 INSERT 7 l l

l SR 3.7.7.4 l l

l This SR verifies CCW System OPERABILITY by ensuring that sufficient cooling capacity is l available for the continued operation of safety-related equipment during normal and accident l conditions by demonstrating the systems ability to remove the design basis heat loads l assumed in the safety analysis. The redundant cooling capacity of this system, assuming a l single active failure, is consistent with the assumptions used in the safety analyses. One pump l and two heat exchangers provide the heat removal capability for accidents that have been l analyzed. l l

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. l l

l l

SR 3.7.7.5 l l

This SR verifies CCW System OPERABILITY by specifically verifying CCW heat exchanger l performance is within the heat exchanger surveillance curve limits by performance testing. The l performance test allows for quantification of the effectiveness of the CCW heat exchangers and l provides assurance that the heat exchangers can provide sufficient cooling capacity during l normal and accident conditions. Because CCW heat exchanger effectiveness can be affected l over time by fouling (i.e., tube resistance), CCW heat exchanger performance is monitored to l l

ensure the CCW heat exchangers can perform their intended function and at their credited l

analysis capacity during normal and accident conditions. CCW heat exchanger performance l monitoring consists of performing SR 3.7.7.4 and SR 3.7.7.5 within their required Frequencies l and heat exchanger cleaning to evaluate CCW heat exchanger performance when fouling of the l heat exchangers is taken into consideration. l l

SR 3.7.7.5 is modified by a Note that specifies the surveillance is not required to be performed l until 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after Reactor Coolant System average temperature reaches 547° F but prior to l entering MODE 2. This allows a delay of testing to establish conditions consistent with those l under which the acceptance criterion was generated. However, the SR must be met within its l l

Frequency prior to reactor operation in MODE 2.

Insert Page B 3.7.7-5a l R2

ITS 3.7.7 5 l R2 INSERT 7 (continued) l l

l Data is collected at a Frequency determined by the Surveillance Frequency Control Program to l determine heat exchanger fouling at the corresponding conditions and a maximum ultimate heat l sink (UHS) temperature. The performance test conservatively allows for tube resistance above l l

the fouling factor used in the DBA safety analyses when the actual UHS temperature is lower l

than the maximum temperature allowed in the DBA safety analyses. A heat exchanger l performance software program computes an overall heat transfer coefficient with the l corresponding fouling (tube resistance) to determine a maximum allowed intake cooling water l temperature for those conditions. If the maximum allowable UHS temperature l (Specification 3.7.9, "Ultimate Heat Sink (UHS)") limit is maintained, the CCW heat exchanger l performance meets the heat transfer required by the DBA safety analysis. l l

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. l l

l SR 3.7.7.4 and SR 3.7.7.5 in conjunction with Specification 3.7.9 limit on UHS temperature l

ensures that sufficient cooling capacity is available to provide for normal cooldown of the facility, l or to mitigate the effects of accident conditions within acceptable limits. l Insert Page B 3.7.7-5b l R2

CCW System B 3.7.7 BASES U 9.

3.2 REFERENCES

1. FSAR, Section [9.2.2]. 1 2 U

2. FSAR, Section [6.2]. 1 2

3. WCAP-16294-NP-A, Rev. 1, "Risk-Informed Evaluation of Changes to Technical Specification Required Action Endstates for Westinghouse NSSS PWRs," June 2010.

Westinghouse STS B 3.7.7-6 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4) Revision XXX

JUSTIFICATION FOR DEVIATIONS ITS 3.7.7 BASES, COMPONENT COOLING WATER (CCW) SYSTEM

1. Changes are made (additions, deletions, and/or changes) to the Improved Standard Technical Specification (ISTS) Bases that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed, and the proper plant specific information/value is inserted to reflect the current licensing basis.

3. Changes made to the Bases reflect the changes made to the Specification. Due to the multiple CCW cooling loop configurations available, ITS 3.7.7 ACTION C is added to ensure proper application of the ITS in a condition when neither CCW train can provide the required safety function, which requires at least one OPERABLE R2 cooling pump and two OPERABLE heat exchangers aligned to at least one cooling loop.

4. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.

5. CTS 4.7.2.a requires CCW System testing that verifies the capability to remove the l R2 design basis heat loads assumed in the safety analyses. CTS 4.7.2.b.2 requires l verification by performance test that the CCW heat exchangers are capable of l performing within the established heat exchanger surveillance curves when l considering heat exchanger fouling (tube resistance). This is a periodic l quantification of CCW heat exchanger effectiveness to ensure the heat exchangers l can provide the required cooling during normal and accident conditions. l l

l Improved Technical Specification (ITS) 3.7.7 adds Surveillance Requirement l (SR) 3.7.7.4 and SR 3.7.7.5 to require testing of the CCW System (heat exchanges l and pumps) and testing of the effectiveness of the heat exchangers specifically when l fouling is considered. The testing required by CTS 4.7.2.a and CTS 4.7.2.b.2 are the l same as those required by ITS SR 3.7.7.4 and SR 3.7.7.5, respectively. Therefore, l current technical specification requirements are being retained and no technical l change is being proposed. l Turkey Point Unit 3 and Unit 4 Page 1 of 1

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.7.7, COMPONENT COOLING WATER (CCW) SYSTEM There are no specific No Significant Hazards Considerations for this Specification.

Turkey Point Unit 3 and 4 Page 1 of 1

ATTACHMENT 8 ITS 3.7.8, INTAKE COOLING WATER (ICW) SYSTEM

Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

ITS A01 ITS 3.7.8 PLANT SYSTEMS (ICW) 3/4.7.3 INTAKE COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION A02 Two trains LCO 3.7.8 3.7.3 The Intake Cooling Water System (ICW) shall be OPERABLE with:

Two L01

a. Three ICW pumps, and

b. Two ICW headers. LA01 Applicability APPLICABILITY: MODES 1, 2, 3, and 4.

Required Action A.1 NOTE: Enter applicable ACTIONS of LCO 3.4.1.3, "Reactor Coolant System - Hot Shutdown," for Note residual heat removal loops made inoperable by ICW.

ACTION: L01

a. With only two ICW pumps with independent power supplies OPERABLE, restore the inoperable ICW pump to OPERABLE status within 14 days or in accordance with the Risk Informed Completion Time Program, 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 in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

LA01 Action A b. With only one ICW pump OPERABLE or with two ICW pumps OPERABLE, but not from independent power supplies, restore two pumps from independent power supplies to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or in accordance with the Risk Informed Completion Time Program, 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 in COLD Action B SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Add proposed Required Action B.2 and associated Note L02 Action A c. With only one ICW header OPERABLE, restore two headers to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or in accordance with the Risk Informed Completion Time Program, 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 in COLD SHUTDOWN within the following Action B 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Add proposed Required Action B.2 and associated Note L02 SURVEILLANCE REQUIREMENTS 4.7.3 The Intake Cooling Water System (ICW) shall be demonstrated OPERABLE:

SR 3.7.8.1 a. In accordance with the Surveillance Frequency Control Program by verifying that each in the A03 valve (manual, power-operated, or automatic) servicing safety-related equipment that is flowpath not locked, sealed, or otherwise secured in position is in its correct position; and Add proposed SR 3.7.8.1 Note A04

b. In accordance with the Surveillance Frequency Control Program during shutdown, by verifying that: that is not locked, sealed, or otherwise L03 secured in position SR 3.7.8.2 1) Each automatic valve servicing safety-related equipment actuates to its correct position on a SI test signal, and in the flowpath A03 an actual or simulated actuation SR 3.7.8.3 2) Each Intake Cooling Water System pump starts automatically on a SI test signal.

L04 SR 3.7.8.2 3) Interlocks required for system operability are OPERABLE.

SR 3.7.8.3 A05 TURKEY POINT - UNITS 3 & 4 3/4 7-16 AMENDMENT NOS. 287 AND 281 Page 1 of 1

DISCUSSION OF CHANGES ITS 3.7.8, INTAKE COOLING WATER (ICW) SYSTEM ADMINISTRATIVE CHANGES A01 In the conversion of the Turkey Point Nuclear Generating Station (PTN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 5.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal.

These changes are designated as administrative changes and are acceptable, because they do not result in technical changes to the CTS.

A02 CTS 3.7.3 states that the Intake Cooling Water System (ICW) shall be OPERABLE with:

a. Three ICW pumps, and b. Two ICW headers. ITS LCO 3.7.8 states that two ICW trains shall be OPERABLE. This changes the CTS by placing the ICW system components into separate trains.

The purpose of CTS 3.7.3 is to ensure that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions.

The redundant cooling capacity of this system, assuming a single active failure, is consistent with the assumptions used in the safety analyses. This change is acceptable because the LCO requirements continue to ensure that the structures, systems, and components are maintained consistent with the safety analyses and licensing basis.

This is a change in presentation only and does not result in technical changes to the CTS.

A03 CTS 4.7.3.a requires verification that each valve (manual, power operated, or automatic) servicing safety related equipment that is not locked, sealed, or otherwise secured in position, is in its correct position. CTS 4.7.3.b.1) requires verification that each automatic valve servicing safety related equipment actuates to its correct position on a Safety Injection test signal. ITS SR 3.7.8.1 requires verification that each ICW manual, power operated, and automatic valve in the flow path servicing safety related equipment that is not locked, sealed, or otherwise secured in position, is in the correct position. ITS SR 3.7.8.2 requires verification that each ICW automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal. This changes the CTS by adding the words "in the flow path" to CTS 4.7.3.a (ITS SR 3.7.8.1) and replacing the words "servicing safety related equipment" with "in the flow path" in CTS 4.7.2.b.1) (ITS SR 3.7.8.2).

The purpose of CTS 4.7.3.a is to ensure all valves in the ICW flow path are in the correct position. The purpose of CTS 4.7.3.b.1) is to provide assurance that each ICW automatic valve actuates to its correct position. The addition of the words "in the flow path" to CTS 4.7.3.a (ITS SR 3.7.8.1) does not change the intent of the Surveillance Requirement. Each manual, power operated, and automatic valve servicing safety related equipment that is not locked, sealed, or otherwise secured in position will continue to be verified to be in the correct position. The removal of the words "servicing safety related equipment" in CTS 4.7.3.b.1) (ITS SR 3.7.8.2) does not change the intent of the Surveillance Requirement. Each ICW automatic valve in the flow path that is not locked, sealed or otherwise secured in position, will still be checked to ensure it actuates to the correct position on an actual or simulated Safety Injection actuation signal. This Turkey Point Unit 3 and Unit 4 Page 1 of 5

DISCUSSION OF CHANGES ITS 3.7.8, INTAKE COOLING WATER (ICW) SYSTEM change is designated as administrative because it does not result in technical changes to the CTS.

A04 CTS 4.7.3.a does not contain explicit guidance concerning Intake Cooling Water (ICW) system OPERABILITY when isolating ICW flow to individual components. ITS Surveillance Requirement (SR) 3.7.8.1 contains a Note, which states, "Isolation of ICW flow to individual components does not render the ICW system inoperable." This changes the CTS by adding an allowance that is not explicitly stated in the CTS.

The purpose of CTS 4.7.3.a is to provide assurance that ICW is available to the appropriate plant components. This change is acceptable because by current use and application of the CTS, isolation of a component supplied with ICW does not necessarily result in ICW system being considered inoperable, but the respective component may be declared inoperable for its system. This change clarifies this application. This change is designated as administrative because it does not result in technical changes to the CTS.

A05 CTS 4.7.3.b.2) requires verification that each ICW system pump starts automatically on a SI test signal. CTS 4.7.3.b.3) requires verification that the interlocks required for system OPERABILITY are OPERABLE. ITS SR 3.7.8.3 requires verification that each ICW pump starts automatically on an actual or simulated actuation signal. This changes the CTS by removing the duplicative requirement to verify that the interlocks required for OPERABILITY are OPERABLE, aligning those interlocks to the requirement to verify the ICW pumps start automatically on a SI test signal.

The purpose of CTS 4.7.3.b.3) is to verify the interlocks required for system OPERABILITY are OPERABLE. The associated interlock is the starting of the C ICW pump automatic start on an SI test signal when either ICW pump A or B breaker is racked out and the C ICW pump power supply is aligned to the racked-out pump's power supply. The C ICW pump is interlocked with the A and B ICW pump such that the C ICW pump can only start automatically on an SI signal when aligned to replace either the A or B ICW pump. Because both of the CTS surveillances are associated with the automatic starting of the ICW pumps on an SI signal, the duplicative surveillance is being deleted. This change is acceptable because the test requirements and frequency regarding the ICW pumps remain the same. This change is designated as administrative because it does not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.7.3 requires the Intake Cooling Water System (ICW) to be OPERABLE Turkey Point Unit 3 and Unit 4 Page 2 of 5

DISCUSSION OF CHANGES ITS 3.7.8, INTAKE COOLING WATER (ICW) SYSTEM with: a. Three ICW pumps, and b. Two ICW headers (Two ICW pumps by DOC L01).

CTS 3.7.3 Action b states that with only one ICW pump OPERABLE or with two ICW pumps OPERABLE, but not from independent power supplies, restore two pumps from independent power supplies to OPERABLE status. ITS LCO 3.7.8 requires two ICW trains to be OPERABLE but does not define the components the associated flow path and independence that comprise an OPERABLE ICW train while Condition A states with one ICW train inoperable. This changes the CTS by moving the description of an OPERABLE ICW train to the Bases.

The removal of these details which are related to system design from the Technical Specifications is acceptable because this type of information is not necessary to be included to provide adequate protection of public health and safety. The ITS still retains the requirement for two ICW trains to be OPERABLE. Also, this change is acceptable because the removed information will be adequately controlled in the ITS Bases.

Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 1 - Relaxation of LCO Requirements) CTS 3.7.3 states that the Intake Cooling Water System (ICW) shall be OPERABLE with: a. Three ICW pumps, and b.

Two ICW headers. In addition, CTS 3.7.3 Action a. requires that with only two ICW pumps with independent power supplies OPERABLE to restore the inoperable ICW pump to OPERABLE status within 14 days. ITS 3.7.7 states that two ICW trains shall be OPERABLE. This changes the CTS by reducing the number of ICW pumps required to be OPERABLE from three to two.

The purpose of CTS 3.7.3 is to provide the lowest functional capability or performance levels of equipment required for safe operation of the facility assuming a single active failure. The redundant cooling capacity of this system, assuming a single active failure, is consistent with the assumptions used in the safety analyses. One pump and one header provides the heat removal capability for accidents that have been analyzed. This change is acceptable because the LCO requirements continue to ensure that the number of ICW pumps required to be OPERABLE maintains the cooling capacity of the system assuming a single active failure and maintained consistent with the safety analyses and licensing basis. In addition, the risk associated with one inoperable ICW pump will be assessed and managed in accordance with the maintenance rule program.

This change is designated as less restrictive because less stringent LCO requirements are being applied in the ITS than were applied in the CTS.

L02 (Category 4 - Relaxation of Required Action) CTS 3.7.3, ACTIONS b and c, identify degraded conditions of the ICW system and provide specific Completion Times to restore the degraded condition or commence a unit shutdown. If a unit shutdown is required each CTS 3.7.3 Action b and c require the unit be in 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 . ITS 3.7.8 ACTION B states that if the Required Action and associated Completion Time of ICW degraded conditions are not met to be in MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 4 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Turkey Point Unit 3 and Unit 4 Page 3 of 5

DISCUSSION OF CHANGES ITS 3.7.8, INTAKE COOLING WATER (ICW) SYSTEM and is modified by a Note stating that Limiting Condition for Operation (LCO) 3.0.4.a is not applicable when entering MODE 4. This changes the CTS by allowing a Required Action end state of HOT SHUTDOWN (MODE 4) rather than an end state of COLD SHUTDOWN (MODE 5).

One purpose of CTS 3.7.3, ACTIONS b, and c is to provide an end state, a condition that the reactor must be placed in, if the Required Actions allowing remedial measures to be taken in response to the degraded conditions with continued operation are not met.

End states are usually defined based on placing the unit into a MODE or condition in which the Technical Specification Limiting Condition for Operation (LCO) is not applicable. MODE 5 is the current end state for LCOs that are applicable in MODES 1 through 4. This change is acceptable because the risk of the transition from MODE 1 to MODES 4 or 5 depends on the availability of alternating current (AC) sources and the ability to remove decay heat such that remaining in MODE 4 may be safer. During the realignment from MODE 4 to MODE 5, there is an increased potential for loss of shutdown cooling and loss of inventory events. Decay heat removal following a loss-of-offsite power event in MODE 5 is dependent on AC power for shutdown cooling whereas, in MODE 4, the turbine driven auxiliary feedwater (AFW) pump will be available. Therefore, transitioning to MODE 5 is not always the appropriate end state from a risk perspective. Thus, for specific TS conditions, Westinghouse Topical Report WCAP-16294-A R1 (ADAMS Accession No. ML103430249) justifies MODE 4 as an acceptable alternate end state to Mode 5. The proposed change to the Technical Specifications will allow time to perform short-duration repairs, which currently necessitate exiting the original mode of applicability. The MODE 4 end state is applied, and risk is assessed and managed in accordance with Title 10 of the Code of Federal Regulations (10 CFR) Section 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants." This proposed change is consistent with NRC approved TSTF-432-A Revision 1 (ADAMS Accession No. ML103360003), noticed for availability by the NRC in the Federal Register (77 FR 27814) on May 11, 2012. The NRC's approval of WCAP-16294-A included four limitations and conditions on its use as identified in Section 4.0 of the NRC Safety Evaluation associated with WCAP-16294-A.

Implementation of these stipulations were addressed in the Bases of TSTF-432-A.

Florida Power & Light implemented these limitations and conditions at PTN in the adoption of the associated TSTF-432-A Bases. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS.

L03 (Category 6 - Relaxation of Surveillance Requirement Acceptance Criteria)

CTS 4.7.3.b.1 requires verification that each automatic valve servicing safety-related equipment actuates to its correct position on a safety injection (SI) test signal. ITS SR 3.7.8.2 requires verification that each ICW System automatic valve servicing safety-related equipment that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal. This changes the CTS by exempting automatic valves that are locked, sealed, or otherwise secured in position from being tested to verify that they automatically actuate to their correct position.

The purpose of CTS 4.7.3.b.1 is to provide assurance that the valves in the flow path that are required to actuate in case of a design basis accident (DBA) actuate to the correct position. This change is acceptable because it has been determined that the relaxed SR acceptance criteria are not necessary for verification that the equipment used to meet the LCO can perform its specified safety functions. Valves already in the Turkey Point Unit 3 and Unit 4 Page 4 of 5

DISCUSSION OF CHANGES ITS 3.7.8, INTAKE COOLING WATER (ICW) SYSTEM correct position and are locked, sealed, or otherwise secured in position are not required have the position verified or to be tested to automatically actuate because, in case of a DBA, the valves are already in the required position and secured to prevent changing from the required position. This change is designated as less restrictive because less stringent SRs are being applied in the ITS than were applied in the CTS.

L04 (Category 6 - Relaxation of Surveillance Requirement Acceptance Criteria)

CTS 4.7.3.b.1 and CTS 4.7.3.b.2 require verifying that each automatic valve servicing safety-related equipment and each ICW system pump starts automatically on a SI test signal. ITS SR 3.7.8.2 and ITS SR 3.7.8.3 specify that the signal may be from either an "actual" or simulated (i.e., test) signal. This changes the CTS by explicitly allowing the use of either an actual or simulated signal for the test.

The purpose of CTS 4.7.3.b.1 and CTS 4.7.3.b.2 is to ensure that the automatic valves servicing safety related equipment and ICW system pumps operate correctly upon receipt of an actuation signal. This change is acceptable because it has been determined that the relaxed SR acceptance criteria are not necessary for verification that the equipment used to meet the LCO can perform its specified safety functions.

Equipment cannot discriminate between an "actual," "simulated," or "test" signal and, therefore, the results of the testing are unaffected by the type of signal used to initiate the test. This change is designated as less restrictive because less stringent SRs are being applied in the ITS than were applied in the CTS.

Turkey Point Unit 3 and Unit 4 Page 5 of 5

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

ICW System CTS SWS 1 3.7.8 3.7 PLANT SYSTEMS Intake Cooling Water (ICW) System 3.7.8 Service Water System (SWS) 1 ICW 3.7.3 LCO 3.7.8 Two SWS trains shall be OPERABLE. 1 Applicability APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Action b, A. One SWS train A.1 --------------NOTES------------- 1 Action c ICW inoperable. 1. Enter applicable Conditions and Required Actions of LCO 3.8.1, "AC 2 Sources - Operating,"

for emergency diesel generator made inoperable by SWS.

2. Enter applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops

- MODE 4," for residual heat removal loops made inoperable by SWS. ICW ICW Restore SWS train to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 1 OPERABLE status.

[OR 3 In accordance with the Risk Informed Completion Time Program] 3 Westinghouse STS 3.7.8-1 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY

ICW System CTS SWS 1 3.7.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME Action b, Action c B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A not AND 4 met.

B.2 --------------NOTE--------------

LCO 3.0.4.a is not applicable when entering MODE 4.

DOC L02 ------------------------------------- l R2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.7.3.a SR 3.7.8.1 -------------------------------NOTE------------------------------

ICW Isolation of SWS flow to individual components does 1 not render the SWS inoperable.

ICW Verify each SWS manual, power operated, and [ 31 days 1 automatic valve in the flow path servicing safety 3 related equipment, that is not locked, sealed, or OR otherwise secured in position, is in the correct position. In accordance with the Surveillance Frequency Control Program ] 3 Westinghouse STS 3.7.8-2 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY

ICW System CTS SWS 1 3.7.8 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY ICW 4.7.3.b.1), SR 3.7.8.2 Verify each SWS automatic valve in the flow path [ [18] months 1 4.7.3.b.3) that is not locked, sealed, or otherwise secured in 3 position, actuates to the correct position on an OR actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program ] 3 ICW 4.7.3.b.2) SR 3.7.8.3 Verify each SWS pump starts automatically on an [ [18] months 1 actual or simulated actuation signal. 3 OR In accordance with the Surveillance Frequency Control Program ] 3 Westinghouse STS 3.7.8-3 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY

JUSTIFICATION FOR DEVIATIONS ITS 3.7.8, INTAKE COOLING WATER (ICW) SYSTEM

1. Changes are made (additions, deletions, and/or changes) to the Improved Standard Technical Specification (ISTS) that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. PTN Unit 3 and Unit 4 emergency diesel generators are designed with a self-contained cooling system, which consists of a forced circulation cooling water loop, to cool the engine directly, rejecting heat through an air-cooled radiator and do not depend on any support plant cooling water system. Therefore, Note 1 to ISTS 3.7.8 Required Action A.1 is unnecessary and deleted.

3. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed, and the proper plant specific information/value is inserted to reflect the current licensing basis.

4. Corrections to the Required Action made consistent with the Writers Guide for the Improved Technical Specifications, TSTF-GG-05-01, Section 4.1.6.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

ICW System SWS 1 B 3.7.8 B 3.7 PLANT SYSTEMS Intak e Cooling Water (ICW) System B 3.7.8 Service Water System (SWS) 1 BASES ICW System BACKGROUND The SWS provides a heat sink for the removal of process and operating 1 heat from safety related components during a Design Basis Accident (DBA) or transient. During normal operation, and a normal shutdown, the ICW System SWS also provides this function for various safety related and nonsafety 1 related components. The safety related function is covered by this LCO.

ICW System The SWS consists of two separate, 100% capacity, safety related, cooling water trains. Each train consists of two 100% capacity pumps, one component cooling water (CCW) heat exchanger, piping, valving, INSERT 1 instrumentation, and two cyclone separators. The pumps and valves are remote and manually aligned, except in the unlik ely event of a loss of coolant accident (LOCA). The pumps aligned to the critical loops are 1 automatically started upon receipt of a safety inj ection signal, and all essential valves are aligned to their post accident positions. The SWS also provides emergency mak eup to the spent fuel pool and CCW System [ and is the back up water supply to the Auxiliary Feedwater System] .

ICWS Additional information about the design and operation of the SWS, along U

9.6 with a list of the components served, is presented in the FSAR, ICWS 1 Section [ 9.2.1] (Ref. 1). The principal safety related function of the SWS 2 is the removal of decay heat from the reactor via the CCW System.

ICW System ICW pump and one header APPLICABLE The design basis of the SWS is for one SWS train, in conj unction with the 1 SAFETY CCW System and a 100% capacity containment cooling system, to ANALYSES Chapter 6.0 remove core decay heat following a design basis LOCA as discussed in U

the FSAR, Section [ 6.2] (Ref. 2). This prevents the containment sump 1 2 fluid from increasing in temperature during the recirculation phase following a LOCA and provides for a gradual reduction in the temperature of this fluid as it is supplied to the Reactor Coolant System by the ECCS ICW System pumps. The SWS is designed to perform its function with a single failure 1 of any active component, assuming the loss of offsite power.

ICW System The SWS, in conj unction with the CCW System, also cools the unit from 1 9.3 residual heat removal (RHR), as discussed in the FSAR, Section [ 5.4.7] , 1 2 U

(Ref. 3) entry conditions to MODE 5 during normal and post accident operations. The time required for this evolution is a function of the ICW number of CCW and RHR System trains that are operating. One SWS train is sufficient to remove decay heat during subsequent operations in 1 ICW MODES 5 and 6. This assumes a maximum SWS temperature of [ 95] ° F 104 2 occurring simultaneously with maxium heat loads on the system.

maximum ICW System The SWS satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

Westinghouse STS B 3.7.8-1 Rev. 5.0 1 Turk ey Point Unit 3 and Unit 4 Revision XXX

ITS 3.7.8 1

INSERT 1 In addition, the ICW System design includes an additional ICW pump that can swing from one train to the other, with interlocks to ensure the swing pump can serve as a backup to either ICW train. The ICW System supplies salt water to the tube side of the component cooling water (CCW) heat exchangers and to cold side of the turbine area cooling water heat exchangers.

The redundant header system is provided with isolation valves that can be shut so that failure of one loop does not require immediate shutdown of the unit. The supply headers are redundant while the return merges to a discharge header that returns water to the discharge canal.

Three ICW pumps are provided for each unit. Only one pump is required following a maximum hypothetical accident (MHA). The A and B ICW pumps are powered by 4.16 kV buses A or B, which can be powered by each train's associated emergency diesel generator (EDG). A swing 4.16 kV emergency bus provides power to the swing ICW pump C and can be manually aligned to either the A or B train 4.16 kV bus. ICW pump C is interlocked, such that, it can only start on a loss of offsite power (LOOP) or safety injection (SI) signal if the supply breaker for the A or B ICW pump associated with train to which it is aligned is open and racked out.

The ICW system provides sufficient redundancy so that at least one ICW pump will continue to operate to handle heat loads from DBAs following a postulated single active failure. A single ICW pump, however, is limited in its ability to supply the required cooling water to the CCW water heat exchangers during an accident when flow is also allowed to continue through the turbine plant cooling water (TPCW) heat exchangers. OPERABILITY of the ICW header during an accident is maintained by isolation of the TPCW.

The most limiting single active failure considered was the loss of one emergency diesel generator, which results in only one ICW pump starting automatically to mitigate the consequences of the MHA.

Insert Page B 3.7.8-1

ICW System SWS 1 B 3.7.8 BASES ICW LCO Two SWS trains are required to be OPERABLE to provide the required 1 redundancy to ensure that the system functions to remove post accident heat loads, assuming that the worst case single active failure occurs coincident with the loss of offsite power.

ICW An SWS train is considered OPERABLE during MODES 1, 2, 3, and 4 1 when:

An ICW pump with an independent power supply a. The pump is OPERABLE and 1

b. The associated piping, valves, heat exchanger, and instrumentation and controls required to perform the safety related function are INSERT 2 OPERABLE.

ICW System APPLICABILITY In MODES 1, 2, 3, and 4, the SWS is a normally operating system that is 1 ICW System required to support the OPERABILITY of the equipment serviced by the SWS and required to be OPERABLE in these MODES. 1 Although the LCO for the SWS is not applicable in MODES 5 and 6, the 1 ICW System capability of the SWS to perform its necessary related support functions may be required for OPERABILITY of supported systems.

ACTIONS A.1 ICW If one SWS train is inoperable, action must be taken to restore 1 OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months [or in accordance with the Risk 2

ICW Informed Completion Time Program]. In this Condition, the remaining OPERABLE SWS train is adequate to perform the heat removal function. 1 However, the overall reliability is reduced because a single failure in the ICW OPERABLE SWS train could result in loss of SWS function. Required l R2 ICW System a

Action A.1 is modified by two Notes. The first Note indicates that the applicable Conditions and Required Actions of LCO 3.8.1, "AC Sources - 3 Operating," should be entered if an inoperable SWS train results in an inoperable emergency diesel generator. The second Note indicates that ICW System the applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops - MODE 4," should be entered if an inoperable SWS train results in 1 an inoperable decay heat removal train. This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is based on the redundant capabilities afforded by the OPERABLE train, and the low probability of a DBA occurring during this time period.

Westinghouse STS B 3.7.8-2 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 Revision XXX

ITS 3.7.8 1

INSERT 2 The swing ICW pump can be substituted for a normal ICW pump, provided the swing ICW pump is aligned to the applicable ICW loop and the power supply for the swing pump is aligned to the same AC electrical power distribution train as the pump it is replacing.

Insert Page B 3.7.8-2

ICW System SWS 1 B 3.7.8 BASES ACTIONS (continued)

B.1 and B.2 ICW If the SWS train cannot be restored to OPERABLE status within the 1 associated Completion Time, the unit must be placed in a MODE in which overall plant risk is reduced. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Remaining within the Applicability of the LCO is acceptable to accomplish short duration repairs to restore inoperable equipment because the plant risk in MODE 4 is similar to or lower than MODE 5 (Ref. 4). In MODE 4 the steam generators and Residual Heat Removal System are available to remove decay heat, which provides diversity and defense in depth. As stated in Reference 4, the steam turbine driven auxiliary feedwater pump must be available to remain in MODE 4. Should steam generator cooling be lost while relying on this Required Action, there are preplanned actions to ensure long-term decay heat removal. Voluntary entry into MODE 5 may be made as it is also acceptable from a risk perspective.

Required Action B.2 is modified by a Note that states that LCO 3.0.4.a is not applicable when entering MODE 4. This Note prohibits the use of LCO 3.0.4.a to enter MODE 4 during startup with the LCO not met.

However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering MODE 4, and establishment of risk management actions, if appropriate. LCO 3.0.4 is not applicable to, and the Note does not preclude, 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 allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.8.1 REQUIREMENTS ICW flow to This SR is modified by a Note indicating that the isolation of the SWS 1 components or systems may render those components inoperable, but ICW System does not affect the OPERABILITY of the SWS. 1 Verifying the correct alignment for manual, power operated, and automatic valves in the SWS flow path provides assurance that the ICW System 1 proper flow paths exist for SWS operation. This SR does not apply to Westinghouse STS B 3.7.8-3 Rev. 5.0 1 Turk ey Point Unit 3 and Unit 4 Revision XXX

ICW System SWS 1 B 3.7.8 BASES SURVEILLANCE REQUIREMENTS (continued) valves that are lock ed, sealed, or otherwise secured in position, since they are verified to be in the correct position prior to being lock ed, sealed, or secured. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

[ The 31 day Frequency is based on engineering j udgment, is consistent with the procedural controls governing valve operation, and ensures correct valve positions. 2 OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER S NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency 4 description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

SR 3.7.8.2 This SR verifies proper automatic operation of the SWS valves on an ICW System actual or simulated actuation signal. The SWS is a normally operating 1 power operation system that cannot be fully actuated as part of normal testing. This Surveillance is not required for valves that are lock ed, sealed, or otherwise secured in the required position under administrative controls.

[ The [ 18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillance when performed at the 2

[ 18] month Frequency. Therefore, the Frequency is acceptable from a reliability standpoint.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Westinghouse STS B 3.7.8-4 Rev. 5.0 1 Turk ey Point Unit 3 and Unit 4 Revision XXX

ICW System SWS 1 B 3.7.8 BASES SURVEILLANCE REQUIREMENTS (continued)

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency 4 description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

SR 3.7.8.3 ICW System This SR verifies proper automatic operation of the SWS pumps on an actual or simulated actuation signal. The SWS is a normally operating 1 system that cannot be fully actuated as part of normal testing during power normal operation. [ The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit The ICW swing pump (C pump) outage and the potential for an unplanned transient if the Surveillance l R2 is interlocked to prevent starting l if ICW pumps A and B are were performed with the reactor at power. Operating experience has l aligned for starting. For a start shown that these components usually pass the Surveillance when 2 l signal to initiate starting the swing pump on a LOOP or SI performed at the [18] month Frequency. Therefore, the Frequency is l signal, the supply breaker for the acceptable from a reliability standpoint. l ICW pump, associated with the AC electrical power distribution l train to which it is aligned, must OR l be open and racked out. Testing 1 l the automatic starting of the l swing ICW pump includes testing The Surveillance Frequency is controlled under the Surveillance this interlock. Frequency Control Program. l

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency 4 description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

U 9.6 1 2 REFERENCES 1. FSAR, Section [9.2.1].

Chapter 6.0 U 1 2

2. FSAR, Section [6.2].

U 9.3

3. FSAR, Section [5.4.7]. 1 2

4. WCAP-16294-NP-A, Rev. 1, "Risk-Informed Evaluation of Changes to Technical Specification Required Action Endstates for Westinghouse NSSS PWRs," June 2010.

Westinghouse STS B 3.7.8-5 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 Revision XXX

JUSTIFICATION FOR DEVIATIONS ITS 3.7.8 BASES, INTAKE COOLING WATER (ICW) SYSTEM

1. Changes are made (additions, deletions, and/or changes) to the Improved Standard Technical Specification (ISTS) Bases that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed, and the proper plant specific information/value is inserted to reflect the current licensing basis.

3. Changes are made to be consistent with changes made to the Specification.

4. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.7.8, INTAKE COOLING WATER (ICW) SYSTEM There are no specific No Significant Hazards Considerations for this Specification.

Turkey Point Unit 3 and 4 Page 1 of 1

ATTACHMENT 9 ITS 3.7.9, ULTIMATE HEAT SINK (UHS)

Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

ITS A01 ITS 3.7.9 PLANT SYSTEMS 3/4.7.4 ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION LCO 3.7.9 3.7.4 The ultimate heat sink shall be OPERABLE with an average supply water temperature less than or SR 3.7.9.1 equal to 104F.

l R2 Applicability APPLICABILITY: MODES 1, 2, 3, and 4. l l

ACTION: l Add proposed Action A and associated Note L01 l UHS inoperable Action B With the requirements of the above specification not satisfied, be in at least HOT STANDBY within l 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-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 . This ACTION shall be applicable to l both units simultaneously. A02 l MODE 4 within 18 l

Add proposed Action B Note l Add proposed Required Action B.2 Note L01 l SURVEILLANCE REQUIREMENTS 4.7.4 The ultimate heat sink shall be determined OPERABLE:

SR 3.7.9.1 a. In accordance with the Surveillance Frequency Control Program by verifying the average supply water temperature* is less than or equal to 104F.

SR 3.7.9.1 b. At least once per hour by verifying the average supply water temperature* is less than or l R2 1st Frequency l equal to 104F, when water temperature exceeds 100F.

Portable monitors may be used to measure the temperature.

LA01 TURKEY POINT - UNITS 3 & 4 3/4 7-17 AMENDMENT NOS. 263 AND 258 Page 1 of 1

DISCUSSION OF CHANGES ITS 3.7.9, ULTIMATE HEAT SINK (UHS)

ADMINISTRATIVE CHANGES A01 In the conversion of the Turkey Point Nuclear Generating Station (PTN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 5.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal.

These changes are designated as administrative changes and are acceptable, because they do not result in technical changes to the CTS.

A02 The CTS Action includes a statement that the action applies to both units l R2 simultaneously. ITS 3.7.9 ACTIONS do not explicitly state that the ACTIONS apply to l both units. This changes the CTS by deleting redundant detail. l l

l The purpose of the CTS statement is to ensure action requirements are performed l simultaneously for both units when required equipment shared between the units is l inoperable. The statement is redundant to the generic requirement provided in l CTS 3.0.5.a (ITS 3.0.10.a) and therefore is not necessary. ITS 3.0.10.a states, l "Whenever the LCO refers to systems or components which are shared by both units, l the Conditions and Required Actions will apply to both units simultaneously." UHS is a l system common to both units. As a result, applying Technical Specification l requirements to both units simultaneously is required of all requirements associated with l an Limiting Condition for Operation (LCO) that refers to systems and components l l

shared by both units, irrespective of whether an explicit statement is provided or not.

l Technical Specification ACTIONS associated with shared systems will continue to be l performed for both units simultaneously per the requirements of ITS LCO 3.0.10. A Note l to ITS Condition B clarifies that the condition applies when a dual unit shutdown is l required. l l

This change is solely a presentation preference and is designated as administrative l because it does not result in a technical change to the CTS. l MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 4.7.4 provides the requirements for verifying the UHS average supply water temperature is within limits. CTS 4.7.4 includes a footnote that states, Turkey Point Unit 3 and Unit 4 Page 1 of 3

DISCUSSION OF CHANGES ITS 3.7.9, ULTIMATE HEAT SINK (UHS)

"Portable monitors may be used to measure the temperature." ITS 3.7.9 does not include this footnote. This changes the CTS by moving an allowable method of measuring the supply temperature to the Bases.

The removal of these details for performing Surveillance Requirements from the Technical Specifications, is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement for verifying the average supply temperature with no change in frequency. This change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 4 - Relaxation of Required Action) CTS 3.7.4, Action, identifies a degraded condition of the Ultimate Heat Sink (UHS) and provides specific Completion Times to restore the degraded condition or commence a unit shutdown. If a unit shutdown is required, the CTS 3.7.4 Action requires the unit be in HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-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 . ITS 3.7.9, ACTION A, requires that with the UHS inoperable and a dual unit shutdown is not l R2 required, to be in MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 4 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and is modified by a Note l stating that LCO 3.0.4.a is not applicable when entering MODE 4. ITS 3.7.9, ACTION B, l is applicable when the UHS is inoperable and a dual unit shutdown is required, and l requires both units to be in MODE 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and MODE 4 in 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months and is l modified by a Note stating that LCO 3.0.4.a is not applicable when entering MODE 4.

This changes the CTS by allowing a Required Action end state of HOT SHUTDOWN (MODE 4) rather than an end state of COLD SHUTDOWN (MODE 5).

One purpose of CTS 3.7.4, Action is to provide an end state, a condition that the reactor must be placed in, if the Required Actions allowing remedial measures to be taken in response to the degraded conditions with continued operation are not met. End states are usually defined based on placing the unit into a MODE or condition in which the Technical Specification Limiting Condition for Operation (LCO) is not applicable.

MODE 5 is the current end state for LCOs that are applicable in MODES 1 through 4.

This change is acceptable because the risk of the transition from MODE 1 to MODES 4 or 5 depends on the availability of alternating current (AC) sources and the ability to remove decay heat such that remaining in MODE 4 may be safer. During the realignment from MODE 4 to MODE 5, there is an increased potential for loss of shutdown cooling and loss of inventory events. Decay heat removal following a loss-of-offsite power event in MODE 5 is dependent on AC power for shutdown cooling whereas, in MODE 4, the turbine driven auxiliary feedwater (AFW) pump will be available. Therefore, transitioning to MODE 5 is not always the appropriate end state from a risk perspective. Thus, for specific TS conditions, Westinghouse Topical Report WCAP-16294-A R1 (ADAMS Accession No. ML103430249) justifies MODE 4 as an acceptable alternate end state to Mode 5. The proposed change to the Technical Turkey Point Unit 3 and Unit 4 Page 2 of 3

DISCUSSION OF CHANGES ITS 3.7.9, ULTIMATE HEAT SINK (UHS)

Specifications will allow time to perform short-duration repairs, which currently necessitate exiting the original mode of applicability. The MODE 4 TS end state is applied, and risk is assessed and managed in accordance with Title 10 of the Code of Federal Regulations (10 CFR) Section 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants." This proposed change is consistent with NRC approved TSTF-432-A Revision 1 (ADAMS Accession No. ML103360003), noticed for availability by the NRC in the Federal Register (77 FR 27814) on May 11, 2012. The NRC's approval of WCAP-16294-A included four limitations and conditions on its use as identified in Section 4.0 of the NRC Safety Evaluation associated with WCAP-16294-A. Implementation of these stipulations were addressed in the Bases of TSTF-432-A. Florida Power & Light implemented these limitations and conditions at PTN in the adoption of the associated TSTF-432-A Bases.

This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS.

Turkey Point Unit 3 and Unit 4 Page 3 of 3

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

CTS UHS 3.7.9 3.7 PLANT SYSTEMS 3.7.9 Ultimate Heat Sink (UHS) 3.7.4 LCO 3.7.9 The UHS shall be OPERABLE.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. [ One or more cooling A.1 Restore cooling tower 7 days towers with one cooling fan(s) to OPERABLE tower fan inoperable. status. [OR In accordance with the Risk Informed Completion Time Program] ]

REVIEWER'S NOTE----- B.1 Verify water temperature of Once per hour] l R2 The [ ]°F is the maximum the UHS is [90]°F l 1

allowed UHS temperature averaged over the previous l value and is based on 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period. l temperature limitations of l l

the equipment that is relied l

upon for accident mitigation l and safe shutdown of the l unit. l

l l

B. [ Water temperature of l the UHS > [90]°F and l

[ ]°F. l l

Westinghouse STS 3.7.9-1 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY

CTS UHS 3.7.9 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME A INSERT 1 A 3 l R2 DOC L01 C. [ Required Action C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months l and associated l Completion Time of AND l Condition A or B not A 1 l

met. C.2 --------------NOTE-------------- l l

LCO 3.0.4.a is not l OR ] applicable when entering l MODE 4. l UHS inoperable [for ------------------------------------- l reasons other than 1 l

Condition A or B]. Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months l INSERT 2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.9.1 [ Verify water level of UHS is [562] ft [mean sea [ [24] hours level].

OR In accordance 1 with the Surveillance Frequency Control Program ] ]

1 3.7.4 SR 3.7.9.2 [ Verify average water temperature of UHS is [ 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 6 l R2 4.7.4.a l 4.7.4.b [90]°F. 1 104 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months when UHS OR l temperature is > 100°F l 4

l AND In accordance l with the Surveillance Frequency Control Program ] ] 1 Westinghouse STS 3.7.9-2 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY

4 5 l R2 INSERT 1

NOTE-------------

Action Not applicable when a dual DOC L01 unit shutdown is required. l R2 4 5 l R2 INSERT 2 CONDITION REQUIRED ACTION COMPLETION TIME Action B. ------------NOTE----------- B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 3 l R2 Only applicable when a l DOC A02 DOC L01 dual unit shutdown is AND l required.

B.2 --------------NOTE-------------- l R2 LCO 3.0.4.a is not UHS inoperable. applicable when entering l R2 MODE 4.

Be in MODE 4. 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months Insert Page 3.7.9-2

CTS UHS 3.7.9 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.7.9.3 [ O t oo to o [ 15] minutes. [ 31 days OR In accordance 1

with the Surveillance Frequency Control Program ] ]

SR 3.7.9.4 [ Verify each cooling tower fan starts automatically [ [ 18] months on an actual or simulated actuation signal.

OR In accordance 1

with the Surveillance Frequency Control Program ] ]

Westinghouse STS 3.7.9-3 Rev. 5.0 1 Turk ey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY

JUSTIFICATION FOR DEVIATIONS ITS 3.7.9, ULTIMATE HEAT SINK (UHS)

1. The Improved Standard Technical Specification (ISTS) contain bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed, and the proper plant specific information/value is inserted to reflect the current licensing basis.

2. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.

3. Changes have been made to delete Actions; the subsequent Actions and Required Actions have been renumbered to reflect the deletions.

4. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

5. The Turkey Point Nuclear Generating Station (PTN) Current Technical Specifications (CTS) 3.7.9 (Ultimate Heat Sink) allows for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to be in MODE 3 if the inoperability applies to both units simultaneously. The Completion Time allows for an orderly sequential shutdown of both units when the inoperability of a component affects both units with equal severity and is reasonable based on operating experience to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. Another 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is allowed to reach MODE 4 and is reasonable based on operating experience to reach the required unit conditions from MODE 3 in an orderly manner.

6. Changes have been made that deleted Surveillance Requirements (SRs)' the subsequent SRs have been renumbered to reflect the deletions.

l R2 Turkey Point Unit 3 and Unit 4 Page 1 of 1

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

UHS B 3.7.9 B 3.7 PLANT SYSTEMS B 3.7.9 Ultimate Heat Sink (UHS)

BASES BACKGROUND The UHS provides a heat sink for processing and operating heat from safety related components during a transient or accident, as well as Intake Cooling Water (ICW) System during normal operation. This is done by utilizing the Service Water 1 System (SWS) and the Component Cooling Water (CCW) System.

INSERT 1 The UHS has been defined as that complex of water sources, including necessary retaining structures (e.g., a pond with its dam, or a river with its dam), and the canals or conduits connecting the sources with, but not including, the cooling water system intake structures as discussed in the FSAR, Section [9.2.5] (Ref. 1). If cooling towers or portions thereof are required to accomplish the UHS safety functions, they should meet the same requirements as the sink. The two principal functions of the UHS 1 are the dissipation of residual heat after reactor shutdown, and dissipation of residual heat after an accident.

A variety of complexes is used to meet the requirements for a UHS. A lake or an ocean may qualify as a single source. If the complex includes a water source contained by a structure, it is likely that a second source will be required.

The basic performance requirements are that a 30 day supply of water be available, and that the design basis temperatures of safety related INSERT 2 equipment not be exceeded. Basins of cooling towers generally include less than a 30 day supply of water, typically 7 days or less. A 30 day supply would be dependent on other source(s) and makeup system(s) for replenishing the source in the cooling tower basin. For smaller basin sources, which may be as small as a 1 day supply, the systems for 1 replenishing the basin and the backup source(s) become of sufficient importance that the makeup system itself may be required to meet the same design criteria as an Engineered Safety Feature (e.g., single failure considerations), and multiple makeup water sources may be required.

Additional information on the design and operation of the system, along with a list of components served, can be found in Reference 1.

APPLICABLE The UHS is the sink for heat removed from the reactor core following all SAFETY accidents and anticipated operational occurrences in which the unit is ANALYSES cooled down and placed on residual heat removal (RHR) operation. For units that use UHS as the normal heat sink for condenser cooling via the Circulating Water System, unit operation at full power is its maximum heat load. Its maximum post accident heat load occurs 20 minutes after a 1 99 Westinghouse STS B 3.7.9-1 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 Revision XXX

1 INSERT 1 The heat generated by operation of Turkey Point is rejected to a closed cooling canal water system. The cooling canal water system occupies an area approximately 2 miles wide by 5 miles long and includes 168 miles of earthen canals covering approximately 4370 acres of water surface. The average canal depth is 2.8 feet. The entire circulation route from the plant discharge back to plant intake is 13.2 miles and takes approximately 44 hours5.092593e-4 days 0.0122 hours 7.275132e-5 weeks 1.6742e-5 months to complete.

The cooling canal system provides the coolant for the circulating water (CW) system and serves as the UHS for the safety-related intake cooling water (ICW) system. The CW system provides cooling water to the main plant condensers. The ICW system removes heat loads from the component cooling water (CCW) system during normal and accident conditions to support both reactor and containment heat removal requirements and spent fuel cooling requirements. The ICW system has three 100 percent capacity pumps. During normal operation, the ICW system provides cooling water to three 50 percent capacity CCW heat exchangers and two non-safety related turbine plant cooling water (TPCW) heat exchangers. The limit on Ultimate Heat Sink (UHS) temperature in conjunction with the surveillance requirements of LCO 3.7.7, Component Cooling Water System, will ensure that sufficient cooling capacity is available either: (1) To provide normal cooldown of the facility, or (2) To mitigate the effects of accident conditions within acceptable limits.

1 INSERT 2 Section 9.2.5 of U.S. Nuclear Regulatory Commission NUREG 0800, Standard Review Plan (SRP), "Ultimate Heat Sink," Revision 3, states that the UHS should be able to dissipate the maximum possible total heat load, including that of a loss of coolant accident (LOCA), under the worst combination of adverse environmental conditions, even freezing, and can cool the unit (or units, including a LOCA for one unit of a multi-unit station with one heat sink) for a minimum of 30 days without makeup. The CCW supply temperature profile to safety-related cooling loads during a design basis accident (DBA) is maintained by implementing stricter CCW cleanliness requirements, which meets the regulatory guidance in Section 9.2.5 of the SRP.

Insert Page B 3.7.9-1

UHS B 3.7.9 BASES APPLICABLE SAFETY ANALYSES design basis loss of coolant accident (LOCA). Near this time, the unit 5 switches from inj ection to recirculation and the containment cooling systems and RHR are required to remove the core decay heat.

The operating limits are based on conservative heat transfer analyses for the worst case LOCA. Reference 1 provides the details of the assumptions used in the analysis, which include worst expected 1 meteorological conditions, conservative uncertainties when calculating decay heat, and worst case single active failure (e.g., single failure of a manmade structure). The UHS is designed in accordance with Regulatory Guide 1.27 (Ref. 2), which requires a 30 day supply of cooling 1 water in the UHS.

The UHS satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO The UHS is required to be OPERABLE and is considered OPERABLE if it is contains a sufficient volume of water at or below the maximum 1 ICW System temperature that would allow the SWS to operate for at least 30 days following the design basis LOCA without the loss of net positive suction head (NPSH), and without exceeding the maximum design temperature ICW System of the equipment served by the SWS. To meet this condition, the UHS 1 104 temperature should not exceed [ 90° F] and the level should not fall below 2

[ 562 ft mean sea level] during normal unit operation.

APPLICABILITY In MODES 1, 2, 3, and 4, the UHS is required to support the OPERABILITY of the equipment serviced by the UHS and required to be OPERABLE in these MODES.

Although the LCO for the UHS is not applicable in MODES 5 and 6, the capability of the UHS to perform its necessary related support functions may be required for OPERABILITY of supported systems.

ACTIONS [ A.1 If one or more cooling towers have one fan inoperable (i.e., up to one fan per cooling tower inoperable), action must be tak en to restore the inoperable cooling tower fan(s) to OPERABLE status within 7 days [ or in accordance with the Risk Informed Completion Time Program] . 2 The 7 day Completion Time is reasonable based on the low probability of an accident occurring during the 7 days that one cooling tower fan is inoperable (in one or more cooling towers), the number of available systems, and the time required to reasonably complete the Required Action. ]

Westinghouse STS B 3.7.9-2 Rev. 5.0 1 Turk ey Point Unit 3 and Unit 4 Revision XXX

UHS B 3.7.9 BASES ACTIONS (continued)

[ B.1 1 l R2

REVIEWERS NOTE-----------------------------------

The [ ]°F is the maximum allowed UHS temperature value and is based on temperature limitations of the equipment that is relied upon for 4 accident mitigation and safe shutdown of the unit.

With water temperature of the UHS > [90]°F, the design basis assumption l R2 associated with initial UHS temperature are bounded provided the l temperature of the UHS averaged over the previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period is l

[90]°F. With the water temperature of the UHS > [90]°F, long term l cooling capability of the ECCS loads and DGs may be affected. 1 l Therefore, to ensure long term cooling capability is provided to the ECCS l l

loads when water temperature of the UHS is > [90]°F, Required Action l

B.1 is provided to more frequently monitor the water temperature of the l UHS and verify the temperature is [90]°F when averaged over the l previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period. The once per hour Completion Time takes into l consideration UHS temperature variations and the increased monitoring l frequency needed to ensure design basis assumptions and equipment l limitations are not exceeded in this condition. If the water temperature of l the UHS exceeds [90]°F when averaged over the previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period l or the water temperature of the UHS exceeds [ ]°F, Condition C must be l entered immediately.] l A

[ C.1 and C.2 3 l R2 If the Required Actions and Completion Times of Condition [A or B] are l R2 not met, or the UHS is inoperable for reasons other than Condition A [or 2 l B], the unit must be placed in a MODE in which overall plant risk is l reduced. To achieve this status, the unit must be placed in at least l 1

MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

INSERT 3 Remaining within the Applicability of the LCO is acceptable to accomplish short duration repairs to restore inoperable equipment because the plant 2

risk in MODE 4 is similar to or lower than MODE 5 (Ref. 3). In MODE 4 1 the steam generators and Residual Heat Removal System are available to remove decay heat, which provides diversity and defense in depth. As 2

stated in Reference 3, the steam turbine driven auxiliary feedwater pump 1 must be available to remain in MODE 4. Should steam generator cooling Westinghouse STS B 3.7.9-3 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 Revision XXX

1 INSERT 3 Condition A allows for one unit to be in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months if the inoperability applies to one l R2 unit. The Completion Time allows for an orderly shutdown of one unit when the inoperability l affects only one unit. The 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Completion Time is reasonable based on operating experience to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. Another 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is allowed to reach MODE 4 and is also reasonable based on operating experience to reach the required unit conditions in an orderly manner. Condition A is modified by a Note stating that this ACTION is not applicable when a l R2 dual unit shutdown is required.

Insert Page B 3.7.9-3

UHS B 3.7.9 BASES ACTIONS (continued) be lost while relying on this Required Action, there are preplanned actions to ensure long-term decay heat removal. Voluntary entry into MODE 5 may be made as it is also acceptable from a risk perspective.

A l R2 Required Action C.2 is modified by a Note that states that LCO 3.0.4.a is 3 not applicable when entering MODE 4. This Note prohibits the use of LCO 3.0.4.a to enter MODE 4 during startup with the LCO not met.

However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering MODE 4, and establishment of risk management actions, if appropriate. LCO 3.0.4 is not applicable to, and the Note does not preclude, 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 allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. ]

INSERT 4 SURVEILLANCE [ SR 3.7.9.1 REQUIREMENTS This SR verifies that adequate long term (30 day) cooling can be maintained. The specified level also ensures that sufficient NPSH is available to operate the SWS pumps. [ The [24] hour Frequency is based on operating experience related to trending of the parameter variations during the applicable MODES. This SR verifies that the UHS water level 2 l R2 is [562] ft [mean sea level].

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency 4 l R2 description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

] ]

Westinghouse STS B 3.7.9-4 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 Revision XXX

1 INSERT 4 B.1 and B.2 l R2 If the UHS is inoperable, the units must be placed in a MODE in which overall plant risk is l R2 reduced. Because this ACTION applies to both units simultaneously to achieve this status, the l second unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in MODE 4 within 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months .

Condition B is modified by a Note stating that this ACTION is only applicable when a dual unit l R2 shutdown is required. Allowing one unit to be shut down to MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and the second unit to be shut down to MODE 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , sequentially shutting down.

Remaining within the Applicability of the LCO is acceptable to accomplish short duration repairs to restore inoperable equipment because the plant risk in MODE 4 is similar to or lower than MODE 5 (Ref. 2). In MODE 4 the steam generators and RHR System are available to remove decay heat, which provides diversity and defense in depth. As stated in Reference 2, the steam turbine driven auxiliary feedwater pump must be available to remain in MODE 4. Should steam generator cooling be lost while relying on this Required Action, there are preplanned actions to ensure long-term decay heat removal. Voluntary entry into MODE 5 may be made as it is also acceptable from a risk perspective.

Required Action B.2 is modified by a Note that states that LCO 3.0.4.a is not applicable when l R2 entering MODE 4. This Note prohibits the use of LCO 3.0.4.a to enter MODE 4 during startup with the LCO not met. However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering MODE 4, and establishment of risk management actions, if appropriate. LCO 3.0.4 is not applicable to, and the Note does not preclude, 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 allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. This also allows the orderly shutdown of one unit at a time and not jeopardize the stability of the electrical grid by imposing a simultaneous dual unit shutdown.

Insert Page B 3.7.9-4

UHS B 3.7.9 BASES SURVEILLANCE REQUIREMENTS (continued) 1

[ SR 3.7.9.2 2 3 ICW System This SR verifies that the SWS is available to cool the CCW System to at 1 DBA least its maximum design temperature with the maximum accident or INSERT 5 normal design heat loads for 30 days following a Design Basis Accident. 1

[ The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is based on operating experience related to trending of the parameter variations during the applicable MODES. This SR verifies that the average water temperature of the UHS is [90°F]. 2 The hourly Frequency when the l R2 UHS average supply temperature OR is >100°F accounts for potential non-conditional l 1 l daily variations in temperature. The Surveillance Frequency is controlled under the Surveillance When the UHS average supply l water temperature exceeds 100°F Frequency Control Program.

l the water temperature must be verified at least hourly to ensure l

REVIEWERS NOTE----------------------------------- l that Cooling Canal System temperature variations are Plants controlling Surveillance Frequencies under a Surveillance l appropriately captured, thus Frequency Control Program should utilize the appropriate Frequency 4 l ensuring the limit is not exceeded.

(Ref. 3) description, given above, and the appropriate choice of Frequency in the l Surveillance Requirement. l

] ]

[ SR 3.7.9.3 Operating each cooling tower fan for [15] minutes ensures that all fans are OPERABLE and that all associated controls are functioning properly.

It also ensures that fan or motor failure, or excessive vibration, can be detected for corrective action. [ The 31 day Frequency is based on operating experience, the known reliability of the fan units, the redundancy available, and the low probability of significant degradation of the UHS cooling tower fans occurring between surveillances.

OR 2 The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

] ]

Westinghouse STS B 3.7.9-5 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 Revision XXX

1 INSERT 5 The frequency of verifying UHS water temperature to ensure the limit of 104ºF is not exceeded when the water temperature is less than 100ºF is controlled under the Surveillance Frequency Control Program as there is ample (greater than or equal to 4ºF) margin to the limit.

For the verification of UHS average supply water temperature a portable instrument may be used with an appropriate instrument uncertainty subtracted from the acceptance criteria to l R2 ensure the Technical Specification limit is not exceeded.

Insert Page B 3.7.9-5

UHS B 3.7.9 BASES SURVEILLANCE REQUIREMENTS (continued)

[ SR 3.7.9.4 This SR verifies that each cooling tower fan starts and operates on an actual or simulated actuation signal. [ The [18] month Frequency is consistent with the typical refueling cycle. Operating experience has shown that these components usually pass the Surveillance when performed at the [18] month Frequency. Therefore, the Frequency is acceptable from a reliability standpoint.

OR 2 The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

] ]

U 14.

3.4 REFERENCES

1. FSAR, Section [9.2.5]. 1 2

2. Regulatory Guide 1.27. 1 2

3. WCAP-16294-NP-A, Rev. 1, "Risk-Informed Evaluation of Changes 3 to Technical Specification Required Action Endstates for Westinghouse NSSS PWRs," June 2010.

3. Letter from A. L. Klett (NRC) to M. Nazar (NextEra) dated August 8, 2014, l R2 Turkey Point Nuclear Generating Unit Nos. 3 and 4- Issuance of Amendments l Under Exigent Circumstances Regarding Ultimate Heat Sink and Component 3 l Cooling Water Technical Specifications (TAC Nos. MF4392 and MF4393) l (NRC ADAMS Accession No. ML14199A107).

l Westinghouse STS B 3.7.9-6 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 Revision XXX

JUSTIFICATION FOR DEVIATIONS ITS 3.7.9 BASES, ULTIMATE HEAT SINK (UHS)

1. Changes are made (additions, deletions, and/or changes) to the Improved Standard Technical Specification (ISTS) Bases that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed, and the proper plant specific information/value is inserted to reflect the current licensing basis.

3. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description. Where a deletion has occurred, subsequent alphanumeric designators have been changed for any applicable affected Required Actions, Surveillance Requirements, Functions, and Footnotes.

4. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.

5. Editorial/grammatical changes made.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.7.9, ULTIMATE HEAT SINK (UHS)

There are no specific No Significant Hazards Considerations for this Specification.

Turkey Point Unit 3 and 4 Page 1 of 1

ATTACHMENT 10 ITS SECTION 3.7.10 - CONTROL ROOM EMERGENCY VENTILATION SYSTEM (CREVS)

Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

ITS A01 ITS 3.7.10 PLANT SYSTEMS (CREVS) 3.7.10 3/4.7.5 CONTROL ROOM EMERGENCY VENTILATION SYSTEM A01 LIMITING CONDITION FOR OPERATION Two CREVS trains LCO 3.7.10 3.7.5 The Control Room Emergency Ventilation System shall be OPERABLE* with:

A02 AND control room

a. Three air handling units, (AHUs) shall be OPERABLE See ITS 3.7.11

b. Two condensing units,

c. Two control room recirculation fans,

d. Two recirculation dampers,

e. One filter train, LA01

f. Two isolation dampers in the normal outside air intake duct,

g. Two isolation dampers in the emergency outside air intake duct,

h. Control Room Envelope APPLICABILITY APPLICABILITY: MODES 1, 2, 3, 4, 5 and 6 or during movement of irradiated fuel assemblies.

ACTION:`

ACTION A a.1 With one air handling unit inoperable, within 7 days, restore the inoperable air handling unit to ACTION F.2 OPERABLE status or, immediately suspend all movement of irradiated fuel and be in at least HOT ACTION D & Condition Note STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for one Unit, or 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for both Units, and in COLD ACTION E Condition Note SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Add proposed Required Actions D.2, E.2 and Note L01 a.2 With only one OPERABLE condensing unit, within 30 days, restore at least one of the inoperable See ITS condensing units to OPERABLE status or, immediately suspend all movement of irradiated fuel and 3.7.11 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 for one Unit, or 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for both Units, and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION B a.3 With one recirculation fan inoperable, within 7 days, restore the inoperable fan to OPERABLE ACTION F.2 status or, immediately suspend all movement of irradiated fuel and be in at least HOT STANDBY ACTION D & Condition Note within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for one Unit, or 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for both Units, and in COLD SHUTDOWN within ACTION E & Condition Note the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Add proposed Required Actions D.2, E.2 and Note L01 Add proposed Required Actions F.1.1 and F.1.2 L06 l R2 LCO 3.7.10 Note *The Control Room Envelope (CRE) boundary may be opened intermittently under administrative control.

TURKEY POINT - UNITS 3 & 4 3/4 7-18 AMENDMENT NOS. 275 AND 270 Page 1 of 4

ITS A01 ITS 3.7.10 PLANT SYSTEMS 3/4.7.5 CONTROL ROOM EMERGENCY VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION (continued)

L05 a.4 With one recirculation damper inoperable, within 7 days, restore the inoperable damper to OPERABLE status or, place and maintain at least one of the recirculation dampers in the open position and place the system in recirculation mode or, immediately suspend all movement of irradiated fuel and 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 for one Unit, or 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for both Units, and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

a.5 With the filter train inoperable, e.g., an inoperable filter, and/or two inoperable recirculation fans, A04 l R2 ACTION H l ACTION L and/or two inoperable recirculation dampers, immediately suspend all movement of irradiated fuel l and immediately initiate action to place the compensatory filtration unit in service and verify proper l ACTION H operation within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , following which movement of irradiated fuel may resume, and within 7 L02 l ACTION B l days, restore the filter train to OPERABLE status.

ACTION L.2 lR With the above requirements not met, immediately suspend all movement of irradiated fuel and be l ACTION I & Condition Note in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for one Unit, or 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for both Units, and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . lR A01 l MODE 5 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months for one unit, or 42 hours4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months for both units ACTION J & Condition Note l a.6 With an inoperable damper in the normal outside air intake, within 7 days, restore the inoperable damper to OPERABLE status or, place and maintain at least one of the normal outside air intake isolation dampers in the closed position and place the system in recirculation mode or, immediately suspend all movement of irradiated fuel and 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 for one Unit, or 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for both Units, and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

L05 a.7 With an inoperable damper in the emergency outside air intake, within 7 days, restore the inoperable damper to OPERABLE status or, place and maintain at least one of the emergency outside air intake isolation dampers in the open position or, immediately suspend all movement of irradiated fuel and 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 for one Unit, or 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for both Units, and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

L05 Add proposed ACTION G M01 TURKEY POINT - UNITS 3 & 4 3/4 7-19 AMENDMENT NOS. 275 AND 270 Page 2 of 4

ITS A01 ITS 3.7.10 PLANT SYSTEMS 3/4.7.5 CONTROL ROOM EMERGENCY VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION (continued)

ACTION C b. With the Control Room Emergency Ventilation System inoperable due to an inoperable CRE boundary during MODES 1, 2, 3 or 4, immediately initiate action to implement mitigating actions.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , verify mitigating actions ensure CRE occupant radiological and chemical hazards will not exceed limits, and CRE occupants are protected from smoke hazards, and restore CRE boundary to OPERABLE status within 90 days.

ACTION D& Condition Note With the above requirements not met, 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 for L01 ACTION E & Condition Note one Unit, or 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for both Units, and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

MODE 4 6 A03 With the Control Room Emergency Ventilation System inoperable due to an inoperable CRE ACTION K boundary during MODES 5, 6 or during the movement of irradiated fuel assemblies, immediately l R2 or two or more l AHUs inoperable suspend all movement of irradiated fuel. M02 l SURVEILLANCE REQUIREMENTS 4.7.5 The Control Room Emergency Ventilation System shall be demonstrated OPERABLE:

See ITS a. In accordance with the Surveillance Frequency Control Program by verifying that the control room 3.7.11 air temperature is less than or equal to 120F;

b. In accordance with the Surveillance Frequency Control Program by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at SR 3.7.10.1 least 15 minutes**;

c. In accordance with the Surveillance Frequency Control Program or (1) after 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation, or (2) after any structural maintenance on the HEPA filter or charcoal adsorber housings, See ITS or (3) following exposure of the filters to effluents from painting, fire, or chemical release in any 5.5.8 ventilation zone communicating with the system that may have an adverse effect on the functional SR 3.7.10.2 capability of the system, or (4) after complete or partial replacement of a filter bank by:

- As the mitigation actions of TS 3.7.5 Action a.5 include the use of the compensatory filtration unit, the unit shall meet the surveillance requirements of TS 4.7.5.b, by manual initiation from outside the control room and TS LA02 4.7.5.c, d and f.

TURKEY POINT - UNITS 3 & 4 3/4 7-20 AMENDMENT NOS. 275 AND 270 Page 3 of 4

ITS A01 ITS 3.7.10 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

1) Verifying that the air cleanup system satisfies the in-place penetration and bypass leak age testing acceptance criteria of greater than or equal to 99.95% DOP and 99%

halogenated hydrocarbon removal at a system flow rate of 1000 cfm +/-10%**.

2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory SR 3.7.10.2 Guide 1.52, Revision 2, March 1978, and analyzed per ASTM D3803 - 1989 at 30 C and See ITS 95% relative humidity, meets the methyl iodide penetration criteria of less than 2.5% or 5.5.8 the charcoal be replaced with charcoal that meets or exceeds the stated performance requirement**, and

3) Verifying by a visual inspection the absence of foreign materials and gask et deterioration**.

d.1 In accordance with the Surveillance Frequency Control Program by verifying that the pressure drop across the combined HEPA filters and charcoal adsorber bank s is less than 6 inches Water Gauge while operating the system at a flow rate of 1000 cfm +/-10%**;

See ITS 5.5.15 d.2 In accordance with the Surveillance Frequency Control Program, test the supply fans (trains A and B) and measure CRE pressure relative to external areas adj acent to the CRE boundary.**

SR 3.7.10.3 e. In accordance with the Surveillance Frequency Control Program by verifying that on a Containment L03 Phase A Isolation test signal the system automatically switches into the recirculation mode of operation, an actual or except for dampers and valves that are lock ed, simulated sealed, or otherwise secured in the actuated position SR 3.7.10.4 f. By performing required CRE unfiltered air inleak age testing in accordance with the Control Room Envelope Habitability Program.**

L04

- As the mitigation actions of TS 3.7.5 Action a.5 include the use of the compensatory filtration unit, the unit shall meet the surveillance requirements of TS 4.7.5.b, by manual initiation from outside the control room and TS LA02 4.7.5.c, d and f.

TURKEY POINT - UNITS 3 & 4 3/4 7-21 AMENDMENT NOS. 275 AND 270 Page 4 of 4

DISCUSSION OF CHANGES ITS 3.7.10, CONTROL ROOM EMERGENCY VENTILATION SYSTEM (CREVS)

ADMINISTRATIVE CHANGES A01 In the conversion of the Turkey Point Nuclear Generating Station (PTN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 5.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 3.7.5 requires the Control Room Emergency Ventilation System (CREVS) to l R2 l

be OPERABLE and explicitly lists the components required to meet the Limiting l Condition for Operation (LCO). ITS LCO 3.7.10 requires two CREVS trains and l l

three control room air handling units (AHUs) to be OPERABLE. ITS LCO 3.7.11 l requires two Control Room Emergency Air Temperature Control System l (CREATCS) trains to be OPERABLE. This changes the CTS by providing l l

separate Specifications for the CREVS and the CREATCS. l l

The purpose of CTS 3.7.5 is to provide control room emergency filtration l l

(recirculation) and control room cooling requirements and includes heating, l ventilation, and air conditioning (HVAC) units. Each HVAC unit consists of a l l

condenser cooling unit and an AHU. ITS separates the requirements based on l safety function. ITS 3.7.10 contains the requirements for the emergency filtration l function while ITS 3.7.11 contains the requirements for the control room cooling l l

function. Since the AHUs support both the filtration function and the cooling l function, the AHUs are required to support OPERABILITY of CREVS and l CREATCS. Additionally, ITS LCO 3.7.10 requires two CREVS trains and three l l

control room AHUs to be OPERABLE with the details of the components required l for a CREV train to be OPERABLE, which includes the control room envelop and l l

the common CREV filter train, described in the ITS 3.7.10 Bases (refer to l DOC LA01). l l

l This change represents a change in presentation of the existing requirements l and is designated as administrative because no technical changes are being l made to the CTS. l A03 The CTS 3.7.5, ACTION b, states, in part, that with the above requirements (CREVS Actions) not satisfied to 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 for one Unit, or 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for both Units, and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . In addition, CTS Bases states that that when an ACTION statement requires a dual unit shutdown, the time to be in HOT STANDBY is 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This is to allow the orderly shutdown of one unit at a time without jeopardizing the stability of the electrical grid by imposing a simultaneous dual unit shutdown. ITS 3.7.10, Condition D (be in MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ), is modified by a Note that states that Condition D is not applicable when a dual unit shutdown is required. Proposed ITS 3.7.10, Condition E (be in MODE 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ), is added and modified by a Note stating that Condition E is only applicable when a dual unit shutdown is required. This modifies the CTS to clarify what Completion Times should be followed when requirements are not met.

Turkey Point Unit 3 and Unit 4 Page 1 of 9

DISCUSSION OF CHANGES ITS 3.7.10, CONTROL ROOM EMERGENCY VENTILATION SYSTEM (CREVS)

The purpose of the CTS 3.7.4, ACTION b, is to ensure prompt Action is taken to restore the CREVS to an OPERABLE status within the specified Completion Times. The Notes described above conservatively allow only one unit in a dual unit shutdown event to apply the 12-hour Completion Time to reach MODE 3.

This change is intended to provide clarity with respect to the different CTS Completion Times which depend on the CREVS. The described Notes and addition of proposed Condition E is consistent with the CTS in that the 12-hour Completion Time is only applicable during dual unit shutdown conditions. This change is considered Administrative as no technical changes are being made to the CTS.

A04 CTS 3.7.5, ACTION a.5, contains the remedial actions for all MODES and l R2 conditions listed in the Applicability when both recirculation fans or the required l l

CREVS filter is inoperable (MODES 1, 2, 3, 4, 5, 6, and during the movement of l irradiated fuel assemblies). The ITS proposes to create separate Conditions, l l

one applicable to operation in MODES 1, 2, 3, and 4 (ACTION H) and one l applicable in MODES 5 and 6, and during the movement of irradiated fuel l assemblies (ACTION L) consistent with the ISTS presentation approach. This l l

changes the CTS by creating separate ACTIONS to differentiate between l specific MODES of operation. l l

l The purpose of the CTS Action is to ensure appropriate remedial measures are l taken upon the loss of both recirculation fans or the common CREVS filter is l l

inoperable. ITS 3.7.10, ACTION H, requires placing the compensatory filtration l unit in service within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while CTS 3.7.5, ACTION a.5, states to initiate l action immediately to place the compensatory filtration unit in service and verify l l

its operation within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . In either case, the compensatory filtration unit is l required to be operational and in service within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . A statement to "initiate l action" immediately does not imply that physical realignments can begin at time l l

zero. Action to initiate would include, but is not limited to, verification of plant l conditions, ongoing activities, personnel availability, component configurations, l l

and verification of filter unit testing requirements are met within the required l frequency. These activities must be taken into account prior to changing the l status of a system or train. Also, verification of the compensatory filter unit's l l

proper operation is part of the action of placing the compensatory unit in service. l Therefore, a single Required Action to place the compensatory filtration unit in l service within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months fully encompasses the CTS requirements to initiate action l l

immediately (i.e., without delay and in a controlled manner) to ensure the l compensatory filtration unit is in service and proper operation verified within the l l

24-hour Completion Time. l l

ITS 3.7.10, ACTION L, provides two options: 1) place the compensatory filtration l l

unit in service immediately, or 2) immediately suspend the movement of l irradiated fuel assemblies. This is equivalent to the CTS 3.7.5, ACTION a.5, l requirement to immediately suspend the movement of irradiated fuel in that the l l

CTS Action further states the movement of irradiated fuel may recommence once l the compensatory filtration unit is placed in service. Therefore, if action is taken l l

to suspend the movement of irradiated fuel, the compensatory filtration unit need l not be placed in service. This is acceptable because the radiological l consequences to Control Room operators in MODES 5, 6, and during the l Turkey Point Unit 3 and Unit 4 Page 2 of 9

DISCUSSION OF CHANGES ITS 3.7.10, CONTROL ROOM EMERGENCY VENTILATION SYSTEM (CREVS) movement of irradiated fuel assemblies are associated with a fuel handling l R2 accident (i.e., other design basis accidents are addressed by the ACTIONS l l

applicable to operation in MODES 1, 2, 3, and 4). With the unit operating in l MODES 1, 2, 3, or 4 while also moving irradiated fuel, both ACTION H and l l

ACTION L would be applicable, ensuring the movement of irradiated fuel is l suspended until the compensatory filtration unit is placed in service. Based on l the above, these changes are considered administrative in that the movement of l l

irradiated fuel will continue to be suspended immediately until the compensatory l filtration unit is in service and the compensatory filtration unit will be placed in l service within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , when required. l MORE RESTRICTIVE CHANGES M01 CTS 3.7.5, Actions a.4, one recirculation damper inoperable and a.6, an inoperable damper in the normal outside air intake require, in part, to place the CREVS in the recirculation mode. ITS 3.7.10, ACTION G, requires that with two CREVS trains inoperable due to normal outside air intake isolated to immediately place one CREVS train in the recirculation mode. If ITS 3.7.10, Required l R2 l

Action G.1, cannot be performed within its associated Completion Time, l ITS 3.7.10, ACTION I, will apply when the unit is in MODE 1, 2, 3, or 4 and l l

ITS 3.7.10, ACTION K, will apply when the unit is in MODE 5 or 6 or during l movement of irradiated fuel assemblies. This changes the CTS by replacing l specific action requirements associated with an inoperable recirculation or intake damper in the CREVS with a requirement to place the CREVS in the recirculation mode if both CREVS trains are inoperable due to normal outside air intake isolated.

The purpose of the CTS actions is to ensure the control room envelop boundary and filtration function are maintained when redundant isolation or actuation capability is lost on the CREVS. The CREVS design includes dual dampers on the normal intake, the emergency intake, and the recirculation ducting to ensure these penetrations are isolated and proper actuation in the event of an accident requiring control room habitability. The required dampers are powered and actuated from redundant trains. When one required damper in the normal air intake duct is closed both intake radiation monitors are isolated unable to detect an increase in external radiation levels, potentially preventing CREVS actuation when needed. An appropriate action for this level of degradation is to declare both CREVS intake radiation monitors inoperable causing the CREVS to be placed in the recirculation mode in accordance with LCO 3.3.4, "Control Room Emergency Ventilation System (CREVS) Actuation Instrumentation." This change is acceptable because actions retained in the ITS continue to ensure redundant capability is restored or the safety related function is performed.

These remedial actions are consistent with safe operation under the specified Condition, considering the OPERABLE status of the redundant dampers and the low probability of a DBA occurring during this condition. This change is designated as more restrictive because the CREVS will be placed in the recirculation mode sooner in ITS (immediately) than in CTS (7 days).

Turkey Point Unit 3 and Unit 4 Page 3 of 9

DISCUSSION OF CHANGES ITS 3.7.10, CONTROL ROOM EMERGENCY VENTILATION SYSTEM (CREVS)

M02 CTS 3.7.5 does not contain an Action addressing two or more inoperable AHUs l R2 l

in MODE 5, or 6, or during movement of irradiated fuel assemblies. Therefore, l the first condition described in ITS 3.7.10, ACTION K, is modified to address this l configuration. This changes the CTS by providing specific action requirements l l

associated with two or more inoperable AHUs inoperable in MODE 5, or 6, or l during movement of irradiated fuel assemblies. l l

l The purpose of the CTS actions is to ensure the control room envelop boundary l and filtration functions are maintained when redundant CREVS components, l l

isolation, or actuation capability is lost. The CREVS design requires two of the l three AHUs to be in service to support the CREVS recirculation mode of l operation. With two or more AHUs inoperable, the recirculation function cannot l l

be met. An appropriate action for this level of degradation in MODE 5, or 6, or l during movement of irradiated fuel assemblies is immediately suspend the l movement of irradiated fuel assemblies, since LCO 3.0.3 is not applicable in l l

MODES 5 or 6. This change is acceptable because actions in the ITS continue l to ensure appropriate measures are taken until the safety related function is l l

restored. l l

Because the only DBA associated with operation in MODE 5, or 6, or during l l

movement of irradiated fuel assemblies is a fuel handling accident (FHA), the l proposed ACTION, requiring the suspension of the movement of irradiated fuel l assemblies removes the potential for an FHA. This change is designated as l l

more restrictive because the movement of irradiated fuel assemblies will now be l suspended upon the loss of two or more AHUs in MODE 5, or 6, or during l l

movement of irradiated fuel assemblies.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including l R2 l

Design Limits) CTS LCO 3.7.5 lists the components that comprise the CREVS which are required to be OPERABLE. ITS 3.7.10 does not list the components in the LCO. This changes the CTS by removing the specific components that comprise the CREVS from the Technical Specifications to the Technical Specification Bases.

The removal of CREVS components from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications in order to provide adequate protection of public health and safety. The ITS requires the CREVS to be OPERABLE and lists the components in the Bases. This change is acceptable because this type of detail will be adequately controlled in the Technical Specification Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Turkey Point Unit 3 and Unit 4 Page 4 of 9

DISCUSSION OF CHANGES ITS 3.7.10, CONTROL ROOM EMERGENCY VENTILATION SYSTEM (CREVS)

Bases are properly controlled. This change is designated as a less restrictive removal of detail change because CREVS components are being removed from the Technical Specifications.

LA02 (Type 4 - - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, QAP, CLRT Program, IST Program, ISI Program, or Surveillance Frequency Control Program) A footnote is associated with CTS Surveillance Requirements (SRs) 4.7.5.b, 4.7.5.c.1, 4.7.5.c.2, 4.7.5.c.3, 4.7.5.d.1, 4.7.5.d.2, and 4.7.5.f that require these SRs to be performed on the compensatory filtration unit. ITS 3.7.10 does not contain any requirements to perform SRs on the compensatory filtration unit. This changes the CTS by removing the requirements to perform SRs on the compensatory filtration unit out of Technical Specifications to the Technical Requirements Manual (TRM).

The compensatory filtration unit is not required by the CTS LCO for CREVS but is mentioned in the ACTIONS when the primary filtration unit is inoperable and mentioned in the SRs as a footnote. The ITS will only mention the compensatory filtration unit in the ACTIONS for the inoperable filtration unit. The compensatory filtration unit is a manual, safety-related, Seismic Class I backup to the installed system with the same functional and operational capabilities as the installed filter train. The unit is surveillance tested in accordance with the same requirements as those imposed on the installed filter train and will continue to be tested as such with the removal of the SR requirements to the TRM. The removal of these requirements to the TRM is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of the public health and safety. These requirements can be adequately controlled in the TRM and changes to the TRM are controlled via 10 CFR 50.59. The 10 CFR 50.59 program provides for the evaluation of changes to ensure SRs are properly controlled. This change is designated as a less restrictive removal of detail change because the compensatory filtration unit SRs are being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 4 - Relaxation of Required Action) CTS 3.7.5 ACTIONS identify degraded conditions of the CREVS and provides specific completion times to restore the degraded conditions or commence a unit shutdown. If a unit shutdown is required, CTS 3.7.5 ACTIONS require the unit be in 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 .

ITS 3.7.10, ACTION D (associated with single unit shutdown), states that if the Required Action and associated Completion Time associated with the CREVS degraded conditions are not met to be in MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 4 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and is modified by a Note stating that LCO 3.0.4.a is not applicable when entering MODE 4. ITS 3.7.10, ACTION E (associated with a dual unit shutdown), states that if the Required Action and associated Completion Time associated with the CREVS degraded conditions are not met to be in MODE 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and MODE 4 in 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months and is modified by a Note stating that LCO 3.0.4.a is not applicable when entering MODE 4. This changes the CTS by allowing a Required Action end state of HOT SHUTDOWN (MODE 4) rather than an end state of COLD SHUTDOWN (MODE 5).

Turkey Point Unit 3 and Unit 4 Page 5 of 9

DISCUSSION OF CHANGES ITS 3.7.10, CONTROL ROOM EMERGENCY VENTILATION SYSTEM (CREVS)

One purpose of the CTS 3.7.5, ACTIONS is to provide an end state, a condition that the reactor must be placed in, if the Required Actions allowing remedial measures to be taken in response to the degraded conditions with continued operation are not met. End states are usually defined based on placing the unit into a MODE or condition in which the Technical Specification LCO is not applicable. MODE 5 is the current end state for LCOs that are applicable in MODES 1 through 4. This change is acceptable because the risk of the transition from MODE 1 to MODES 4 or 5 depends on the availability of alternating current (AC) sources and the ability to remove decay heat such that remaining in MODE 4 may be safer. During the realignment from MODE 4 to MODE 5, there is an increased potential for loss of shutdown cooling and loss of inventory events. Decay heat removal following a loss-of-offsite power event in MODE 5 is dependent on AC power for shutdown cooling whereas, in MODE 4, the turbine driven auxiliary feedwater (AFW) pump will be available. Therefore, transitioning to MODE 5 is not always the appropriate end state from a risk perspective. Thus, for specific TS conditions, Westinghouse Topical Report WCAP-16294-A, Revision 1 (ADAMS Accession No. ML103430249), justifies MODE 4 as an acceptable alternate end state to Mode 5. The proposed change to the Technical Specifications will allow time to perform short-duration repairs, which currently necessitate exiting the original mode of applicability. The MODE 4 Technical Specification end state is applied, and risk is assessed and managed in accordance with Title 10 of the Code of Federal Regulations (10 CFR) Section 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants." This proposed change is consistent with NRC approved Technical Specification Task Force (TSTF) traveler TSTF-432-A, Revision 1 (ADAMS Accession No. ML103360003), noticed for availability by the NRC in the Federal Register (77 FR 27814) on May 11, 2012. The NRC's approval of WCAP-16294-A included four limitations and conditions on its use as identified in Section 4.0 of the NRC Safety Evaluation associated with WCAP-16294-A. Implementation of these stipulations were addressed in the Bases of TSTF-432-A. Florida Power & Light implemented these limitations and conditions at PTN in the adoption of the associated TSTF-432-A Bases. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS.

L02 (Category 4 - Relaxation of Required Action) CTS 3.7.5, ACTION a.5, contains l R2 Action for when two recirculation dampers are inoperable. The ACTION requires l l

placing the compensatory filtration unit in service and suspending the movement l of irradiated fuel. ITS 3.7.10 ACTIONS do not include specific actions for an l inoperable CREVS damper. This changes the CTS by removing specific action to placing the compensatory filtration unit in service and suspending the movement of irradiated fuel when two recirculation dampers are inoperable l R2 l

provided the CREVS safety function can be performed.

The purpose of the CTS ACTION is to ensure the CREVS safety function is l R2 maintained when both recirculation dampers are inoperable. The CREVS design l includes dampers on the common intake, exhaust, and recirculation ducts to ensure these penetrations are correctly aligned in the event of accident requiring control room isolation. With a recirculation, normal air intake, or emergency air intake damper inoperable, deactivating the damper is the position assumed in the Turkey Point Unit 3 and Unit 4 Page 6 of 9

DISCUSSION OF CHANGES ITS 3.7.10, CONTROL ROOM EMERGENCY VENTILATION SYSTEM (CREVS) accident analysis ensures the safety function can be performed and restores the LCO. CTS 4.7.5.e requires CREVS components to actuate on a containment l R2 isolation signal, which includes the required automatic dampers. ITS SR 3.7.10.3 also requires similar testing with an exception for dampers and valves that are locked, sealed, or otherwise secured in the actuated position (see DOC L04 for the addition of this exception). Therefore, the exception in the SR effectively allows continued operation when one or more required CREVS dampers are inoperable, provided the dampers are locked, sealed, or otherwise secured in the actuated position. This change is acceptable because actions retained in the ITS continue to provide remedial action when the CREVS safety function cannot be l R2 performed. These remedial actions, including the exception provided in ITS SR 3.7.10.3, are consistent with safe operation under the specified Condition, considering the secured position of the affected damper(s) and the low probability of a Design Basis Accident (DBA) occurring during this condition.

This change is designated as less restrictive because removal of the CTS action allows one or more CREVS dampers to be inoperable provided the CREVS l R2 l

safety function can be performed.

L03 (Category 6 - Relaxation of Surveillance Requirement Acceptance Criteria)

CTS 4.7.5.e requires verification that on a containment isolation signal the system automatically isolates the control room and switches into a recirculation mode of operation. ITS SR 3.7.10.3 specifies that the signal may be either an "actual" or "simulated" actuation signal. This changes the CTS by allowing the use of either an actual or a simulated signal for the test.

The purpose of CTS SR is to ensure that each CREVS train actuates (switches to the recirculation mode) upon receipt of an actuation signal. This change is acceptable because it has been determined that the relaxed SR acceptance criteria are not necessary for verification that the equipment used to meet the LCO can perform its specified safety functions. Equipment cannot discriminate between an "actual," "simulated," or "test" signal and, therefore, the results of the testing are unaffected by the type of signal used to initiate the test. This change is designated as less restrictive because less stringent SRs are being applied in the ITS than were applied in the CTS.

L04 (Category 6 - Relaxation of Surveillance Requirement Acceptance Criteria)

CTS 4.7.5.e requires verification that on a containment isolation signal the system automatically isolates the control room and switches into a recirculation mode of operation. ITS SR 3.7.10.3 requires verification that each CREVS train actuates and additionally states, except for dampers and valves that are locked, l R2 sealed, or otherwise secured in the actuated position. This changes the CTS by not requiring dampers and valves locked, sealed, or otherwise secured in position to be tested.

The purpose of the CTS SR is to provide assurance that the dampers and valves required to actuate in event of a DBA isolate the control room and shift the CREVS into the recirculation mode. This change is acceptable because it has been determined that the relaxed SR acceptance criteria are not necessary for verification that the equipment used to meet the LCO can perform the specified safety functions. Dampers and valves secured in the actuated position are not required to be tested to automatically actuate because, in the event of a DBA, the Turkey Point Unit 3 and Unit 4 Page 7 of 9

DISCUSSION OF CHANGES ITS 3.7.10, CONTROL ROOM EMERGENCY VENTILATION SYSTEM (CREVS) components are already in the required position and secured to prevent changing from the required position. This change is designated as less restrictive because less stringent SRs are being applied in the ITS than were applied in the CTS.

L05 (Category 4 - Relaxation of Required Action) CTS 3.7.5 Action a.4, one recirculation damper inoperable; a.6 , an inoperable damper in the normal outside air intake; a.7, an inoperable damper in the emergency outside air intake provide actions when a required damper is inoperable. ITS 3.7.10 ACTIONS do not include specific actions for inoperable required dampers. This changes the CTS by removing specific action requirements associated with an inoperable damper in the CREVS.

The purpose of the CTS actions is to ensure the control room envelop boundary and filtration function are maintained when redundant isolation or actuation capability is lost on the CREVS. The CREVS design includes dual dampers on the normal intake, the emergency intake, and the recirculation ducting to ensure these penetrations are isolated and proper actuation in the event of accident requiring control room habitability. The required dampers are powered and actuated from redundant trains. Therefore, when one required damper in an air duct is inoperable, appropriate action for that level of degradation is to declare the CREVS train associated with the inoperable damper inoperable. Additionally, closing and de-activating either valve in the required safety function position ensures the safety function is performed and restores the LCO. CTS 4.7.5.e requires CREVS components to actuate on a containment isolation signal, which includes automatic isolation valves. ITS SR 3.7.10.3 requires similar testing with an exception for dampers and valves that are locked, sealed, or otherwise secured in the actuated position. Therefore, the exception in the SR effectively allows continued operation when one or more dampers inoperable, provided the dampers are locked, sealed, or otherwise secured in the actuated position. This change is acceptable because actions retained in the ITS continue to ensure redundant capability is restored or the safety related function is performed.

These remedial actions, including the exception provided in ITS SR 3.7.10.3, are consistent with safe operation under the specified Condition, considering the OPERABLE status of the redundant dampers and the low probability of a DBA occurring during this condition. This change is designated as less restrictive because removal of the CTS actions allows restoration of the damper's capability to be controlled by other ITS 3.7.10 actions, which are less restrictive and allows more than one isolation damper per air duct to be inoperable provided the safety related function is performed.

L06 (Category 4 - Relaxation of Required Action) CTS 3.7.5, Actions a.1 and a.3, l R2 require the suspension of the movement of irradiated fuel assemblies l l

immediately if required AHUs or recirculation fans are not restored to l OPERABLE status within 7 days. ITS 3.7.10, Required Actions F.1.1 and F.1.2, l l

permit placing the control room in the recirculation mode of operation with at l least two AHUs operating as an option to suspending the movement of irradiated l fuel (ITS 3.7.10, Required Action F.2). This changes the CTS by providing l l

optional action requirements when required AHUs or recirculation fans are not l restored to OPERABLE status within 7 days during operation in MODES 5 or 6, l or during the movement of irradiated fuel assemblies. l Turkey Point Unit 3 and Unit 4 Page 8 of 9

DISCUSSION OF CHANGES ITS 3.7.10, CONTROL ROOM EMERGENCY VENTILATION SYSTEM (CREVS)

The purpose of the CTS actions is to ensure the control room envelop boundary l R2 and filtration function are maintained when redundant CREVS components, l l

isolation, or actuation capability is lost. The CREVS safety function is based on l limiting radiological dose to the control room operators following a DBA. The l l

safety function is accomplished by placing the control room in the recirculation l mode of operation with the appropriate number of AHUs in operation. In l MODES 5 and 6, and during the movement of irradiated fuel assemblies, the only l l

associated DBA is the fuel handling accident (FHA). CTS 3.7.5, Actions a.1 l and a.3, remove the potential of an FHA by requiring the suspension of irradiated l fuel assemblies. However, placing the control in the recirculation mode of l l

operation with at least two AHUs in operation will also ensure the potential l radiological dose to the control room operators will remain within safety analyses l l

limits should an FHA occur. Therefore, the addition of the ITS 3.7.10, Required l Actions F.1.1 and F.1.2, options of placing the control in the recirculation mode of l operation with at least two AHUs in operation provides the required radiological l l

protection of the control room operators assumed in the accident analyses. This l change is acceptable because the safety function is met by performance of l proposed ITS 3.7.10, Required Actions F.1.1 and F.1.2, or Required Action F.2, l l

during operation in MODES 5 or 6, or during the movement of irradiated fuel l assemblies. This change is designated as less restrictive because optional l l

actions are proposed in lieu of the CTS actions.

Turkey Point Unit 3 and Unit 4 Page 9 of 9

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

V CREFS 1 3.7.10 CTS 3.7 PLANT SYSTEMS Ventilation V 1 3.7.5 3.7.10 Control Room Emergency Filtration System (CREFS) 3 AND LCO 3.7.5 LCO 3.7.10 Two CREFS trains shall be OPERABLE. 1 V Three control room air handling units (AHUs) shall be OPERABLE.

NOTE--------------------------------------------

The control room envelope (CRE) boundary may be opened intermittently LCO 3.7.5 Footnote

under administrative control.

APPLICABILITY APPLICABILITY: MODES 1, 2, 3, 4, [5, and 6], 2 During movement of [recently] irradiated fuel assemblies.

INSERT 1, ACTION A 3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B V A. One CREFS train C A.1 Restore CREFS train to 7 days V

inoperable for reasons OPERABLE status.

ACTIONs a.3, a.4, l R2 a.6, & a.7 other than Condition B.

B. One or more CREFS B.1 Initiate action to implement Immediately trains inoperable due to mitigating actions.

ACTION b inoperable CRE boundary in MODE 1, 2, AND 3, or 4.

1 B.2 Verify mitigating actions 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ensure CRE occupant C

exposures to radiological, chemical, and smoke hazards will not exceed limits.

AND B.3 Restore CRE boundary to 90 days OPERABLE status.

Westinghouse STS 3.7.10-1 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY

ITS 3.7.10 CTS 3

INSERT 1 ACTION a.1 A. One control room AHU A.1 Restore control room AHU 7 days l R2 inoperable. to OPERABLE status.

Insert Page 3.7.10-1

V

NOTE-------------

CREFS 1 CTS Not applicable when a dual unit 3.7.10 l R2 shutdown is required.

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

, B, or C ACTIONs a.1, a.3, C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months l R2 a.4, a.6, & a.7, b associated Completion 3

Time of Condition A or B AND not met in MODE 1, 2, 3, or 4. C.2 --------------NOTE--------------

D LCO 3.0.4.a is not applicable when entering MODE 4.

DOC A03 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months INSERT 2, ACTION E 3 F F.1.1 ACTIONs a.3, a.4, a.6 D. Required Action and D.1 --------------NOTE-------------- l R2 or B associated Completion [ Place in toxic gas l Time of Condition A not protection mode if 2

met [in MODE 5 or 6, or] automatic transfer to toxic during movement of gas protection mode is

[recently] irradiated fuel inoperable. ]

assemblies. -------------------------------------

V Place OPERABLE CREFS Immediately 1 AND train in emergency mode.

F.1.2 Place two OPERABLE recirculation 3

control room AHUs in Immediately l R2 service. OR F

D.2 Suspend movement of Immediately 2

[recently] irradiated fuel assemblies.

INSERT 3, ACTIONS G & H Westinghouse STS 3.7.10-2 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY

ITS 3.7.10 CTS INSERT 2 3 E. ------------NOTES----------- E.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Only applicable when a dual unit shutdown is AND required.

ACTIONs a.1, a.3, -------------------------------- E.2 -------------NOTE----------------

a.4, a.6, a.7, & b Required Action and LCO 3.0.4.a is not associated Completion applicable when entering Time of Condition A, B, MODE 4.

or C not met in MODE 1, ------------------------------------- l R2 2, 3, or 4. Be in MODE 4. 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months l 1

INSERT 3 G. Two CREVS trains G.1 Place one CREVS train in Immediately 1 l R2 DOC M01 inoperable due to normal recirculation mode.

outside air intake isolated.

H. Two CREVS trains H.1 Place compensatory 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1 ACTION a.5 inoperable in MODE 1, filtration unit in service.

DOC A04 2, 3, or 4 for reasons ----------------NOTE------------- l R2 l

other than Condition C Not applicable when a dual unit l

shutdown is required.

or G. -------------------------------------- l 4

I. Two or more control I.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months l R2 ACTION a.5 room AHUs inoperable l l

in MODE 1, 2, 3, or 4. AND l l

OR I.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months 1 l Required Action and associated Completion Time of Condition G or H not met in MODE 1, 2, 3, l R2 or 4. l Insert Page 3.7.10-2a l R2

ITS 3.7.10 CTS 1

INSERT 3 (continued) l R2 l

l l

4 ACTION a.5 ----------------NOTE------------- J.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months l l

Only applicable when a dual l unit shutdown is required. AND l

l 1 l J.2 Be in MODE 5. 42 hours4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months l J. Two or more control l l

room AHUs inoperable l in MODE 1, 2, 3, or 4. l l

l OR l l

Required Action and l l

associated Completion l Time of Condition G or l l

H not met in MODE 1, 2, l 3, or 4. l l

Insert Page 3.7.10-2b l R2

V CREFS 1 CTS 3.7.10 l R2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME K K l R2 DOC M02 E. Two CREFS trains E.1 Suspend movement of Immediately l 3 l inoperable [in MODE 5 [recently] irradiated fuel or 6, or] during assemblies.

movement of [recently]

Two or more control room irradiated fuel AHUs assemblies.

l R2 OR 2 V

1 One or more CREFS ACTION b trains inoperable due to an inoperable CRE boundary [in MODE 5 ACTIONs a.4, or 6, or] during OR l R2 a.6, a.7 movement of [recently] l L.2 Suspend movement DOC L05 irradiated fuel of irradiated fuel l Place compensatory assemblies. filtration unit in service.

assemblies.

ACTION a.5 l R2 L l 1

V F. Two CREFS trains F.1 Enter LCO 3.0.3. Immediately inoperable in MODE 1, 3 Immediately 5 or 6, or during 2, 3, or 4 for reasons K l R2 movement of irradiated fuel assemblies, other than Condition B.

OR DOC M01 3 l R2 Required Action and associated Completion Time of Condition G not met in MODE 5 or 6, or during movement of irradiated fuel assembles.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY V

SR 4.7.5.b SR 3.7.10.1 Operate each CREFS train for 15 continuous [ 31 days minutes [with heaters operating]. 2 OR In accordance with the Surveillance Frequency Control Program ]

Westinghouse STS 3.7.10-3 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY

V CREFS 1 3.7.10 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY V

SR 4.7.5.c SR 3.7.10.2 Perform required CREFS filter testing in accordance In accordance SR 4.7.5.d.1 with the [Ventilation Filter Testing Program (VFTP)]. with the [VFTP] 2 V

SR 4.7.5.e SR 3.7.10.3 Verify each CREFS train actuates on an actual or [ [18] months simulated actuation signal, except for dampers and 2 valves that are locked, sealed, or otherwise secured OR in the actuated position.

In accordance with the Surveillance Frequency Control Program ]

SR 4.7.5.f SR 3.7.10.4 Perform required CRE unfiltered air inleakage In accordance testing in accordance with the Control Room with the Control Envelope Habitability Program. Room Envelope Habitability Program Westinghouse STS 3.7.10-4 Rev. 5.0 1 Amendment Nos. XXX and YYY Turkey Point Unit 3 and Unit 4

JUSTIFICATION FOR DEVIATIONS ITS 3.7.10, CONTROL ROOM EMERGENCY VENTILATION SYSTEM (CREVS)

1. Changes are made (additions, deletions, and/or changes) to the Improved Standard Technical Specification (ISTS) that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

3. Changes made to reflect the current licensing basis. The Control Room Emergency Ventilation System (CREVS) Current Technical Specifications (CTS) for the Turkey Point Nuclear Generating Station (PTN) is shared between Units 3 and 4 Control Rooms. The ISTS is based on a single plant design with two redundant trains/subsystems. Due to the shared systems between the PTN units, changes are made (additions, deletions, and/or changes) to the ISTS presentation that reflect the plant-specific multi-unit and shared systems design on a unit basis. In addition, renumbering is required due to added Actions. The changes do not represent a change to the plant design or safety analysis basis.

4. CTS 3.7.5, Action a.5, includes shutdown requirements if the other actions contained l R2 l

in CTS 3.7.5, Action a.5, are not completed within the designated time. For these l conditions, ISTS 3.7.10, ACTION I, requires entry into LCO 3.0.3. CTS 3.7.5, l l

Action a.5, provides different times to complete a shutdown, depending on whether a l single or dual unit shutdown is necessary. ISTS 3.7.10, Condition I, is written for l single unit, in which case entry into LCO 3.0.3 would be appropriate upon failure to l l

meet the requirements. However, to avoid unnecessary risk to the offsite power grid l and station operation, simultaneous shutdown of both PTN units is not required by l CTS when a dual unit shutdown is necessary. An extra six hours is provided beyond l l

the standard 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to MODE 3 and 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months to MODE 5 Completion Times to l allow the unit shutdowns to be staggered. Therefore, the ISTS 3.7.10, ACTION I, l l

requirement to immediately enter LCO 3.0.3 is replaced with ITS 3.7.10, ACTIONS I l and J, consistent with the current licensing basis requirements associated with dual l unit shutdowns. l Turkey Point Unit 3 and Unit 4 Page 1 of 1

Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

V CREFS 1 B 3.7.10 B 3.7 PLANT SYSTEMS Ventilation V B 3.7.10 Control Room Emergency Filtration System (CREFS) 1 BASES V

BACKGROUND The CREFS provides a protected environment from which occupants can 1 control the unit following an uncontrolled release of radioactivity, hazardous chemicals, or smoke.

The CREFS consists of two independent, redundant trains that recirculate and filter the air in the control room envelope (CRE) and a CRE boundary that limits the inleakage of unfiltered air. Each CREFS train consists of a prefilter or demister, a high efficiency particulate air (HEPA) filter, an INSERT 1 activated charcoal adsorber section for removal of gaseous activity 1 (principally iodines), and a fan. Ductwork, valves or dampers, doors, barriers, and instrumentation also form part of the system, as well as demisters to remove water droplets from the air stream. A second bank of HEPA filters follows the adsorber section to collect carbon fines and provides backup in case of failure of the main HEPA filter bank.

The CRE is the area within the confines of the CRE boundary that contains the spaces that control room occupants inhabit to control the unit during normal and accident conditions. This area encompasses the control room, and may encompass other non-critical areas to which frequent personnel access or continuous occupancy is not necessary in the event of an accident. The CRE is protected during normal operation, natural events, and accident conditions. The CRE boundary is the combination of walls, floor, roof, ducting, doors, penetrations and equipment that physically form the CRE. The OPERABILITY of the CRE boundary must be maintained to ensure that the inleakage of unfiltered air into the CRE will not exceed the inleakage assumed in the licensing basis analysis of design basis accident (DBA) consequences to CRE occupants. The CRE and its boundary are defined in the Control Room Envelope Habitability Program.

The CREFS is an emergency system, parts of which may also operate during normal unit operations in the standby mode of operation. Upon receipt of the actuating signal(s), normal air supply to the CRE is isolated, and the stream of ventilation air is recirculated through the system filter trains. The prefilters or demisters remove any large particles in the air, and any entrained water droplets present, to prevent excessive loading of the HEPA filters and charcoal adsorbers. Both the demister and heater are important to the effectiveness of the charcoal adsorbers.

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ITS 3.7.10 INSERT 1 1 The CREVS is a subsystem of the Control Building Ventilation System and consists of the l R2 following components:

l R2

a. Air handling units (AHUs),

l R2

b. Recirculation fans, l R2

c. Recirculation dampers, l R2

d. Recirculation filter unit, l R2

e. Normal outside air intake dampers, and l R2

f. Emergency outside air intake dampers.

The AHUs also support the OPERABILITY requirements of LCO 3.7.11, "Control Room l R2 l

Emergency Air Temperature Control System (CREATCS)." Refer to LCO 3.7.11 for AHU l requirements associated with CREATCS.

Units 3 and 4 share a common control room envelope (CRE). The CRE is the area within the confines of the CRE boundary that contains the spaces that control room occupants inhabit to control the units during normal and accident conditions. This area encompasses the main control room area, control room offices, rack area, kitchen, lavatory, and mechanical equipment room located below the control room which contains the CREVS equipment. The CRE is protected during normal operation, natural events, and accident conditions. The CRE boundary is the combination of walls, floor, roof, ducting, doors, penetrations, and equipment that physically form the CRE. The OPERABILITY of the CRE boundary must be maintained to ensure that the inleakage of unfiltered air into the CRE will not exceed the inleakage assumed in the radiological dose consequence analyses and that CRE occupants are protected from hazardous chemicals and smoke. The CRE and its boundary are defined in Control Room Envelope Habitability Program.

During normal operation, fresh makeup air is admitted to this system through an intake louver and two dampers in series located in the west wall of the control building. This system maintains a positive pressure in the CRE greater than that in the cable spreading room in order to prevent smoke from a hypothesized fire in the cable spreading room from entering the control room. All control room penetrations are designed for leak tightness. Since the control room is maintained at slightly more than atmospheric pressure, the infiltration of contaminated air into the control room is negligible.

Two radiation monitors located in the normal air intake ducting continuously monitor for radiation in the incoming air. In the unlikely event of a maximum hypothetical accident (MHA), the control room ventilation will automatically be placed in a recirculation mode.

INSERT Page B 3.7.10-1a

ITS 3.7.10 1

INSERT 1 (continued)

The control room recirculation mode is initiated by a containment Phase A signal, a high l R2 radiation signal from the containment air radiation monitors (R-11 and R-12), the manual initiation from a test switch, or a high radiation signal from the redundant monitors in the control room normal air intake. Following initiation, all exhaust fans shut off, and the redundant series exhaust isolation dampers close. Redundant normal air intake isolation dampers in series close. Redundant parallel emergency air intake dampers open. Likewise, the recirculation air path opens. A single air supply fan is energized to move the appropriate mixture of recirculation control room air and fresh outdoor air through the HEPA and charcoal filter system.

INSERT Page B 3.7.10-1b

V CREFS 1 B 3.7.10 BASES BACKGROUND (continued)

Actuation of the CREFS places the system in either of two separate states (emergency radiation state or toxic gas isolation state) of the emergency mode of operation, depending on the initiation signal. Actuation of the system to the emergency radiation state of the emergency mode of operation, closes the unfiltered outside air intake and unfiltered exhaust dampers, and aligns the system for recirculation of the air within the CRE through the redundant trains of HEPA and the charcoal filters. The emergency radiation state also initiates pressurization and 1 filtered ventilation of the air supply to the CRE.

Outside air is filtered, diluted with building air from the electrical equipment and cable spreading rooms, and added to the air being recirculated from the CRE. Pressurization of the CRE minimizes infiltration of unfiltered air through the CRE boundary from all the surrounding areas adjacent to the CRE boundary. The actions taken in the toxic gas isolation state are the same, except that the signal switches the CREFS to an isolation alignment to minimize any outside air from entering the CRE through the CRE boundary.

The air entering the CRE is continuously monitored by radiation and toxic gas detectors. One detector output above the setpoint will cause actuation of the emergency radiation state or toxic gas isolation state, as required. The actions of the toxic gas isolation state are more restrictive, and will override the actions of the emergency radiation state.

A single CREFS train operating at a flow rate of < [3000] cfm will pressurize the CRE to about [0.125] inches water gauge relative to external areas adjacent to the CRE boundary. The CREFS operation in maintaining the CRE habitable is discussed in the FSAR, Section [9.4]

(Ref. 1).

Redundant supply and recirculation trains provide the required filtration should an excessive pressure drop develop across the other filter train.

Normally open isolation dampers are arranged in series pairs so that the failure of one damper to shut will not result in a breach of isolation. The CREFS is designed in accordance with Seismic Category I requirements.

The CREFS is designed to maintain a habitable environment in the CRE for 30 days of continuous occupancy after a DBA without exceeding a

[5 rem whole body dose or its equivalent to any part of the body] [5 rem total effective does equivalent (TEDE)].

Westinghouse STS B 3.7.10-2 Rev. 5.0 1

Turkey Point Unit 3 and Unit 4 Revision XXX

V CREFS 1 B 3.7.10 BASES APPLICABLE The CREFS components are arranged in redundant, safety related SAFETY ventilation trains. The location of components and ducting within the ANALYSES CRE ensures an adequate supply of filtered air to all areas requiring access. The CREFS provides airborne radiological protection for the 1 CRE occupants, as demonstrated by the CRE occupant dose analyses INSERT 2 for the most limiting design basis accident, fission product release presented in the FSAR, Chapter [15] (Ref. 2).

The CREFS provides protection from smoke and hazardous chemicals to the CRE occupants. The analysis of hazardous chemical releases demonstrates that the toxicity limits are not exceeded in the CRE following a hazardous chemical release (Ref. 3). The evaluation of a smoke challenge demonstrates that it will not result in the inability of the CRE occupants to control the reactor either from the control room or from the remote shutdown panels (Ref. 4).

The worst case single active failure of a component of the CREFS, assuming a loss of offsite power, does not impair the ability of the system to perform its design function. 1 V

The CREFS satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Two independent and redundant CREFS trains are required to be OPERABLE to ensure that at least one is available if a single active failure disables the other train. Total system failure, such as from a loss of both ventilation trains or from an inoperable CRE boundary, could result in exceeding a dose of [5 rem whole body or its equivalent to any part of the body] [5 rem TEDE] to the CRE occupants in the event of a large radioactive release.

1 Each CREFS train is considered OPERABLE when the individual INSERT 3 components necessary to limit CRE occupant exposure are OPERABLE.

A CREFS train is OPERABLE when the associated:

a. Fan is OPERABLE,

b. HEPA filters and charcoal adsorbers are not excessively restricting flow, and are capable of performing their filtration functions, and

c. Heater, demister, ductwork, valves, and dampers are OPERABLE, and air circulation can be maintained.

In order for the CREFS trains to be considered OPERABLE, the CRE boundary must be maintained such that the CRE occupant dose from a large radioactive release does not exceed the calculated dose in the licensing basis consequence analyses for DBAs, and that CRE occupants are protected from hazardous chemicals and smoke.

Westinghouse STS B 3.7.10-3 Rev. 5.0 1

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ITS 3.7.10 1

INSERT 2 The location of CREVS components and ducting within the CRE ensures an adequate supply of filtered air to all areas requiring access. CREVS provides airborne radiological protection for the CRE occupants, as demonstrated by radiological dose consequence analyses for the most limiting Design Basis Accident (DBA) presented in UFSAR Chapter 14. CREVS also provides protection from chemical hazards and smoke hazards. CREVS pressurizes the CRE relative to external areas adjacent to the CRE boundary. The analysis of hazardous chemical releases for NUREG-0737 Item III.D.3.4, "Control Room Habitability Requirement," and the subsequent reanalysis for new chemical release hazards, as a result of the addition of PTN Unit 5, demonstrate that the toxicity limits of Regulatory Guide 1.78 are not exceeded in the CRE following a hazardous chemical release. Thus, neither automatic nor manual actuation of CREVS is required for an analyzed hazardous chemical release. Analysis of a smoke challenge demonstrates that it will not result in the inability of the CRE occupants to control the reactors either from the control room or alternate shutdown panels.

The alternative source term radiological analyses assume both emergency outside air intake flow paths are available with parallel dampers ensuring outside makeup air can be drawn through both intake locations during a DBA and a single active failure. These analyses rely on a provision in Regulatory Guide 1.194 Section 3.3.2 that allows a reduction in the atmospheric dispersion factors (X/Qs) for dual intake arrangements with balanced flow rates to one half of the more limiting X/Q value provided the two intakes are not within the same wind direction window for each release / receptor location.

1 INSERT 3 The OPERABILITY of the CREVS ensures that the CRE will remain habitable for occupants during and following an uncontrolled release of radioactivity. The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the CRE to 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The radiological limits are consistent with the requirements of 10 CFR 50.67.

Two CREVS trains and three AHUs are required to meet the LCO.

OPERABILITY of three AHUs ensure at least two AHU fans are available to maintain positive pressure in the control room during accident conditions considering a single active failure.

Each CREVS train is considered OPERABLE when the individual components necessary to limit CRE occupant exposure are OPERABLE. A CREVS train is OPERABLE when the associated:

a. Fan is OPERABLE,

b. Common HEPA filter and charcoal adsorber unit is not excessively restricting flow, and is capable of performing its filtration function, and

c. Heater, demister, ductwork, valves, and dampers are OPERABLE, and air circulation can be maintained.

INSERT Page B 3.7.10-3a

ITS 3.7.10 INSERT 3 (cont.)

In order for the CREVS trains to be considered OPERABLE, the CRE boundary must be maintained such that the CRE occupant dose from a large radioactive release does not exceed the calculated dose in the licensing basis consequence analyses for DBAs. Additionally, placing a damper in the required safety function position and de-activating it ensures the safety function is performed and restores the LCO.

INSERT Page B 3.7.10-3b

V CREFS 1 B 3.7.10 BASES LCO (continued)

The LCO is modified by a Note allowing the CRE boundary to be opened intermittently under administrative controls. This Note only applies to openings in the CRE boundary that can be rapidly restored to the design condition, such as doors, hatches, floor plugs, and access panels. For entry and exit through doors, the administrative control of the opening is performed by the person(s) entering or exiting the area. For other openings, these controls should be proceduralized and consist of stationing a dedicated individual at the opening who is in continuous communication with the operators in the CRE. This individual will have a method to rapidly close the opening and to restore the CRE boundary to a condition equivalent to the design condition when a need for CRE isolation is indicated.

APPLICABILITY V In MODES 1, 2, 3, 4, [5, and 6,] and during movement of [recently] 2 irradiated fuel assemblies, the CREFS must be OPERABLE to ensure that the CRE will remain habitable during and following a DBA.

V In [MODES 5 and 6], the CREFS is required to cope with the release from 2 the rupture of an outside waste gas tank.

decay V

During movement of [recently] irradiated fuel assemblies, the CREFS 2 or an inoperable filter must be OPERABLE to cope with the release from a fuel handling train replaced by the compensatory filter unit accident [involving handling recently irradiated fuel]. [The CREFS is only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor 2 core within the previous [X] days), due to radioactive decay.]

ACTIONS A.1 When one CREFS train is inoperable, for reasons other than an inoperable CRE boundary, action must be taken to restore OPERABLE status within 7 days. In this Condition, the remaining OPERABLE CREFS INSERT 4 train is adequate to perform the CRE occupant protection function.

However, the overall reliability is reduced because a failure in the 4 OPERABLE CREFS train could result in loss of CREFS function. The 1 7 day Completion Time is based on the low probability of a DBA occurring during this time period, and ability of the remaining train to provide the required capability.

B.1, B.2, and B.3 If the unfiltered inleakage of potentially contaminated air past the CRE boundary and into the CRE can result in CRE occupant radiological dose 4 greater than the calculated dose of the licensing basis analyses of DBA consequences (allowed to be up to [5 rem whole body or its equivalent to Westinghouse STS B 3.7.10-4 Rev. 5.0 1

Turkey Point Unit 3 and Unit 4 Revision XXX

ITS 3.7.10 1

INSERT 4 A.1 With one AHU inoperable, action must be taken to restore the AHU to OPERABLE status within 7 days. In this Condition, the remaining OPERABLE AHUs are adequate to support the CREVS specified safety function. However, the overall reliability is reduced because a single failure of 4 the remaining AHUs could result in loss of CREVS function. The 7-day Completion Time is based on the low probability of an event requiring control room isolation and the consideration that the remaining AHUs can provide the required protection.

B.1 When one CREVS train is inoperable, for reasons other than an inoperable CRE boundary, action must be taken to restore the train to OPERABLE status. This action allows inoperability of the redundant active CREVS components (one recirculation fan, one recirculation damper, one normal outside air intake damper, and one emergency outside air intake damper) for a period of up to 7 days based on the low probability of occurrence of a DBA challenging control 1 room habitability during this time period and the continued capability of the remaining OPERABLE system components to perform the required CREVS safety function. When a redundant recirculation damper or redundant emergency outside air intake damper is inoperable, at least one damper in the associated flowpath(s) may be locked, sealed, or otherwise secured in the accident position to restore system OPERABILITY. When a redundant normal outside air intake damper is inoperable, isolation of the associated flowpath renders both CREVS trains inoperable and Condition G applies.

C.1, C.2, and C.3 In order for the CREVS to be considered OPERABLE, the CRE boundary must be maintained such that the CRE occupant radiological dose does not exceed that calculated in the DBA radiological dose consequence analyses. Since the CREVS and CRE are common to both units, the ACTION requirements are applicable to both units simultaneously, and must be applied according to each units operational MODE.

INSERT Page B 3.7.10-4

V CREFS 1 B 3.7.10 BASES ACTIONS (continued) any part of the body] [5 rem TEDE]), or inadequate protection of CRE occupants from hazardous chemicals or smoke, the CRE boundary is 1 inoperable. Actions must be taken to restore an OPERABLE CRE boundary within 90 days.

During the period that the CRE boundary is considered inoperable, action must be initiated to implement mitigating actions to lessen the effect on CRE occupants from the potential hazards of a radiological or chemical event or a challenge from smoke. Actions must be taken within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to verify that in the event of a DBA, the mitigating actions will ensure that CRE occupant radiological exposures will not exceed the calculated dose of the licensing basis analyses of DBA consequences, and that CRE occupants are protected from hazardous chemicals and smoke. These mitigating actions (i.e., actions that are taken to offset the consequences of the inoperable CRE boundary) should be preplanned for implementation upon entry into the condition, regardless of whether entry is intentional or unintentional. The 24-hour Completion Time is reasonable based on the low probability of a DBA occurring during this time period, and the use of mitigating actions. The 90-day Completion Time is reasonable based on the determination that the mitigating actions will ensure protection of CRE occupants within analyzed limits while limiting the probability that CRE occupants will have to implement protective measures that may adversely affect their ability to control the reactor and maintain it in a safe shutdown condition in the event of a DBA. In addition, the 90-day Completion Time is a reasonable time to diagnose, plan and possibly repair, and test most problems with the CRE boundary.

D.1 and D.2 C.1 and C.2 V 4

, AHU, In MODE 1, 2, 3, or 4, if the inoperable CREFS train or the CRE boundary Condition D is modified by a Note stating that Condition D is not cannot be restored to OPERABLE status within the required Completion 1 applicable when a dual unit Time, the unit must be placed in a MODE in which the overall plant risk is shutdown is required. Since both reduced. To achieve this status, the unit must be placed in at least 4 units share the same Control Room, both units are affected by CREVS MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

inoperabilities. When inoperable CREVS components result in both units being required to shutdown, Remaining within the Applicability of the LCO is acceptable to accomplish Condition E is applicable. short duration repairs to restore inoperable equipment because the plant risk in MODE 4 is similar to or lower than MODE 5 (Ref. 5). In MODE 4 1 (RHR) the steam generators and Residual Heat Removal System are available to remove decay heat, which provides diversity and defense in depth. As 1

Westinghouse STS B 3.7.10-5 Rev. 5.0 1

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V CREFS 1 B 3.7.10 BASES ACTIONS (continued) 1 (AFW) stated in Reference 5, the steam turbine driven auxiliary feedwater pump 1 must be available to remain in MODE 4. Should steam generator cooling be lost while relying on this Required Action, there are preplanned actions to ensure long-term decay heat removal. Voluntary entry into MODE 5 may be made as it is also acceptable from a risk perspective.

D.2 Required Action C.2 is modified by a Note that states that LCO 3.0.4.a is 1

not applicable when entering MODE 4. This Note prohibits the use of LCO 3.0.4.a to enter MODE 4 during startup with the LCO not met.

However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering MODE 4, and establishment of risk management actions, if appropriate. LCO 3.0.4 is not applicable to, and the Note does not preclude 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 allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power INSERT 5 conditions in an orderly manner and without challenging unit systems.

D.1 and D.2 4 F.1.1, F.1.2, and F.2

[In MODE 5 or 6, or] during movement of [recently] irradiated fuel 2 V

assemblies, if the inoperable CREFS train cannot be restored to 1

OPERABLE status within the required Completion Time, action must be V

taken to immediately place the OPERABLE CREFS train in the emergency mode. This action ensures that the remaining train is with two recirculation OPERABLE 4 OPERABLE, that no failures preventing automatic actuation will occur, AHUs in service and that any active failure would be readily detected.

F.1.1 l R2 An alternative to Required Action D.1 is to immediately suspend activities 1 that could result in a release of radioactivity that might require isolation of the CRE. This places the unit in a condition that minimizes the accident risk. This does not preclude the movement of fuel to a safe position.

[Required Action D.1 is modified by a Note indicating to place the system 2 INSERT 6 in the toxic gas protection mode if automatic transfer to the toxic gas protection mode is inoperable. ]

INSERT 7 Westinghouse STS B 3.7.10-6 Rev. 5.0 1

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ITS 3.7.10 INSERT 5 E.1 and E.2 Condition E is modified by a Note stating that Condition E is only applicable when a dual unit shutdown is required. Since both units share the same control room, both units can be affected 4

by CREVS inoperabilities. With both units operating in MODE 1, 2, 3, or 4, if the inoperable CREVS or the CRE boundary cannot be restored to OPERABLE status within the required Completion Time, the units must be placed in a MODE in which the overall plant risk is reduced.

To achieve this status, the units must be placed in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and in MODE 4 within 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months . The extra 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is to allow for a dual unit shutdown such that the units can be shut down sequentially.

Remaining within the Applicability of the LCO is acceptable to accomplish short duration repairs to restore inoperable equipment because the plant risk in MODE 4 is similar to or lower than MODE 5 (Ref. 1). In MODE 4 the steam generators and RHR System are available to remove decay heat, which provides diversity and defense in depth. As stated in Reference 1, the steam 1 turbine driven AFW pump must be available to remain in MODE 4. Should steam generator cooling be lost while relying on this Required Action, there are preplanned actions to ensure long-term decay heat removal. Voluntary entry into MODE 5 may be made as it is also acceptable from a risk perspective.

Required Action E.2 is modified by a Note that states that LCO 3.0.4.a is not applicable when entering MODE 4. This Note prohibits the use of LCO 3.0.4.a to enter MODE 4 during startup with the LCO not met. However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable 1

systems and components, consideration of the results, determination of the acceptability of entering MODE 4, and establishment of risk management actions, if appropriate. LCO 3.0.4 is not applicable to, and the Note does not preclude 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 allowed Completion Times are reasonable, based on operating experience, to reach the 1 required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

1 INSERT 6 G.1 With both CREVS trains inoperable due to the normal outside air intake flowpath isolated, action must be taken immediately to place at least one CREVS train in the recirculation mode of operation. This action is necessary since the radiation monitors that automatically actuate the CREVS trains can no longer detect radioactivity in the outside atmosphere.

INSERT Page B 3.7.10-6a

ITS 3.7.10 1

INSERT 6 (cont.)

H.1 With both CREVS trains inoperable for reasons other than an inoperable CRE boundary (e.g.,

the common filter unit is inoperable, two recirculation fans are inoperable, or required dampers are inoperable and not secured in the accident position) or the normal outside air intake flowpath is isolated, the compensatory filtration unit is required to be placed in service within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This remedial action ensures control room occupant radiological exposures will not exceed limits. The 24-hour allowance to place the compensatory filtration unit in service is reasonable based on the low probability of a DBA occurring during this time period.

The compensatory filtration unit is designed as a manual, safety-related, Seismic Class I backup to the installed system with the same functional and operational capabilities as the installed CREVS filter unit. The requirements for the compensatory filtration unit are located in the Technical Requirements Manual.

4 INSERT 7 I.1 and I.2 l R2 In MODES 1, 2, 3, or 4 if two or more AHUs are inoperable, the CREVS can no longer provide l R2 its specified safety function. Additionally, if the Required Action of Condition G or H cannot be performed within the required Completion Time, no additional time is justified for continued operation. Therefore, the unit must be placed in a MODE in which the overall plant risk is reduced. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and l R2 l

in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . l l

Condition I is modified by a Note stating that the Condition is not applicable when a dual unit l l

shutdown is required. Condition J is entered if a dual unit shutdown is required. The allowed l Completion Times are reasonable, based on operating experience, to reach the required unit l conditions from full power conditions in an orderly manner and without challenging unit systems. l l

l J.1 and J.2 l l

l With both units operating in MODE 1, 2, 3, or 4, if two or more AHUs are inoperable or if the l Required Action of Condition G or H cannot be performed within the required Completion Time, the l l

units must be placed in a MODE in which the overall plant risk is reduced. To achieve this status, l the units must be placed in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and in MODE 5 within 42 hours4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months . l l

l Condition J is modified by a Note stating that the Condition is only applicable when a dual unit l shutdown is required. To achieve a consecutive unit shutdown, the first unit should be in MODE 3 l l

and MODE 5 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , respectively, to allow time for the second unit to reach l MODE 3 and MODE 5 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and 42 hours4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months , respectively. This is to allow for the orderly l shutdown of one unit at a time and not jeopardize the stability of the electrical grid by imposing a l l

simultaneous dual unit shutdown. Condition I is entered if a single unit shutdown is required. l l

The allowed Completion Times are reasonable, based on operating experience, to reach the l l

required unit conditions from full power conditions in an orderly manner and without challenging l unit systems. l INSERT Page B 3.7.10-6b

V CREFS 1 B 3.7.10 BASES ACTIONS (continued)

E.1 4 K.1 l R2 or more Control Room [In MODE 5 or 6, or] during movement of [recently] irradiated fuel 2 AHUs inoperable or with assemblies, with two CREFS trains inoperable or with one or more 4

CREVS trains inoperable due to an inoperable CRE boundary, action must be taken immediately to suspend activities that could result in a release of radioactivity that might require isolation of the CRE. This places the unit in a condition that minimizes the accident risk. This does not preclude the movement of fuel to a safe position.

INSERT 8 F.1 4 If both CREFS trains are inoperable in MODE 1, 2, 3, or 4 for reasons other than an inoperable CRE boundary (i.e., Condition B), the CREFS may not be capable of performing the intended function and the unit is in a condition outside the accident analyses. Therefore, LCO 3.0.3 must be entered immediately.

SURVEILLANCE SR 3.7.10.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly. As the environment and normal operating conditions on this system are not too severe, testing each train once every month provides an adequate check of this system. Operation [with the heaters on] for 15 continuous minutes demonstrates OPERABILITY of the 3 system. Periodic operation ensures that [heater failure,] blockage, fan or motor failure, or excessive vibration can be detected for corrective action.

[ The 31 day Frequency is based on the reliability of the equipment and 3 the two train redundancy.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the 2 Surveillance Requirement.

]

Westinghouse STS B 3.7.10-7 Rev. 5.0 1

Turkey Point Unit 3 and Unit 4 Revision XXX

ITS 3.7.10 4

INSERT 8 L.1 and L.2 l R2 In MODE 5 or 6, or during movement of irradiated fuel assemblies, if both CREVS trains are inoperable for reasons other than Condition K or the Required Action and associated l R2 Completion Time of Condition G cannot be met, action must be taken to immediately to place the compensatory filtration unit in service, ensuring control room occupant radiological exposures will not exceed limits. The compensatory filtration unit is designed as a manual, safety-related, Seismic Class I backup to the installed system with the same functional and operational capabilities as the installed CREVS filter unit. The requirements for the compensatory filtration unit are located in the Technical Requirements Manual.

Because the compensatory filter may not be immediately available to place in service, an l R2 l

alternative action is provided to immediately suspend activities that could result in a release of radioactivity that might require isolation of the CRE. This places the unit in a condition that minimizes the accident risk. This does not preclude the movement of fuel to a safe position.

Once the compensatory filter is placed in service, movement of irradiated fuel assemblies may l R2 resume and fuel handling operation continued with the compensatory filter operating. l INSERT Page B 3.7.10-7

V CREFS 1 B 3.7.10 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.7.10.2 V

This SR verifies that the required CREFS testing is performed in 1 accordance with the [Ventilation Filter Testing Program (VFTP)]. The 3

[VFTP] includes testing the performance of the HEPA filter, charcoal adsorber efficiency, minimum flow rate, and the physical properties of the activated charcoal. Specific test Frequencies and additional information are discussed in detail in the [VFTP]. 3 SR 3.7.10.3 V

This SR verifies that each CREFS train starts and operates on an actual 1 or simulated actuation signal. The SR excludes automatic dampers and valves that are locked, sealed, or otherwise secured in the actuated position. The SR does not apply to dampers or valves that are locked, sealed, or otherwise secured in the actuated position since the affected dampers or valves were verified to be in the actuated position prior to being locked, sealed, or otherwise secured. Placing an automatic valve or damper in a locked, sealed, or otherwise secured position requires an assessment of the OPERABILITY of the system or any supported systems, including whether it is necessary for the valve or damper to be repositioned to the non-actuated position to support the accident analysis.

Restoration of an automatic valve or damper to the non-actuated position requires verification that the SR has been met within its required Frequency. [ The Frequency of [18] months is based on industry operating experience and is consistent with the typical refueling cycle.

3 OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the 2 Surveillance Requirement.

]

Westinghouse STS B 3.7.10-8 Rev. 5.0 1

Turkey Point Unit 3 and Unit 4 Revision XXX

V CREFS 1 B 3.7.10 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.7.10.4 This SR verifies the OPERABILITY of the CRE boundary by testing for unfiltered air inleakage past the CRE boundary and into the CRE. The details of the testing are specified in the Control Room Envelope Habitability Program.

The CRE is considered habitable when the radiological dose to CRE occupants calculated in the licensing basis analyses of DBA consequences is no more than [5 rem whole body or its equivalent to any 3

part of the body] [5 rem TEDE] and the CRE occupants are protected from hazardous chemicals and smoke. This SR verifies that the unfiltered air inleakage into the CRE is no greater than the flow rate assumed in the licensing basis analyses of DBA consequences. When unfiltered air C.3 inleakage is greater than the assumed flow rate, Condition B must be C l R2 entered. Required Action B.3 allows time to restore the CRE boundary to 4 OPERABLE status provided mitigating actions can ensure that the CRE remains within the licensing basis habitability limits for the occupants 2 following an accident. Compensatory measures are discussed in Regulatory Guide 1.196, Section C.2.7.3, (Ref. 6) which endorses, with 1 exceptions, NEI 99-03, Section 8.4 and Appendix F (Ref. 7). These 3

C.2 compensatory measures may also be used as mitigating actions as required by Required Action B.2. Temporary analytical methods may also 4 be used as compensatory measures to restore OPERABILITY (Ref. 8). 1 Options for restoring the CRE boundary to OPERABLE status include 4 changing the licensing basis DBA consequence analysis, repairing the CRE boundary, or a combination of these actions. Depending upon the nature of the problem and the corrective action, a full scope inleakage test may not be necessary to establish that the CRE boundary has been restored to OPERABLE status.

REFERENCES 1. FSAR, Section [9.4].

2. FSAR, Chapter [15].

3. FSAR, Section [6.4]. 1

4. FSAR, Section [9.5]

5. WCAP-16294-NP-A, Rev. 1, "Risk-Informed Evaluation of Changes 1 to Technical Specification Required Action Endstates for Westinghouse NSSS PWRs," June 2010.

Westinghouse STS B 3.7.10-9 Rev. 5.0 1

Turkey Point Unit 3 and Unit 4 Revision XXX

V CREFS 1 B 3.7.10 BASES REFERENCES (continued)

6. Regulatory Guide 1.196. 1 2

7 NEI 99-03, " Control Room Habitability Assessment," J une 2001. 1 3

8. Letter from Eric J . Leeds (NRC) to J ames W. Davis (NEI) dated 1 4 J anuary 30, 2004, " NEI Draft White Paper, Use of Generic Letter 91-18 Process and Alternative Source Terms in the Context of Control Room Habitability," (ADAMS Accession No. ML040300694).

Westinghouse STS B 3.7.10-10 Rev. 5.0 Turk ey Point Unit 3 and Unit 4 Revision XXX

JUSTIFICATION FOR DEVIATIONS ITS 3.7.10 BASES, CONTROL ROOM EMERGENCY VENTILATION SYSTEM (CREVS)

1. Changes are made (additions, deletions, and/or changes) to the Improved Standard Technical Specification (ISTS) Bases that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.

3. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

4. Changes are made to the Improved Technical Specifications (ITS) Bases to reflect changes made to the ITS.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.7.10, CONTROL ROOM EMERGENCY VENTILATION SYSTEM (CREVS)

There are no specific No Significant Hazards Considerations for this Specification.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

ATTACHMENT 11 ITS SECTION 3.7.11 - CONTROL ROOM EMERGENCY AIR TEMPERATURE CONTROL SYSTEM (CREATCS)

Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

ITS ITS 3.7.11 A01 PLANT SYSTEMS Air Temperature Control System (CREATCS)

A01 3/4.7.5 CONTROL ROOM EMERGENCY VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION See ITS 3.7.10 LCO 3.7.11 3.7.5 The Control Room Emergency Ventilation System shall be OPERABLE* with:

Two CREATCS trains A02

a. Three air handling units, See ITS b. Two condensing units, LA01 3.7.10

c. Two control room recirculation fans,

d. Two recirculation dampers,

e. One filter train,

f. Two isolation dampers in the normal outside air intak e duct,

g. Two isolation dampers in the emergency outside air intak e duct,

h. Control Room Envelope Applicability APPLICABILITY: MODES 1, 2, 3, 4, 5 and 6 or during movement of irradiated fuel assemblies.

ACTION:

a.1 With one air handling unit inoperable, within 7 days, restore the inoperable air handling unit to See ITS OPERABLE status or, immediately suspend all movement of irradiated fuel and be in at least HOT 3.7.10 STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for one Unit, or 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for both Units, and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Action A a.2 With only one OPERABLE condensing unit, within 30 days, restore at least one of the inoperable Action D condensing units to OPERABLE status or, immediately suspend all movement of irradiated fuel and A01 Action B 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 for one Unit, or 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for both Units, and in Action C COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Add proposed Required Actions B.2, C.2 and Note L01 Action B & C a.3 With one recirculation fan inoperable, within 7 days, restore the inoperable fan to OPERABLE status or, immediately suspend all movement of irradiated fuel and 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 for one Unit, or 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for both Units, and in COLD SHUTDOWN within See ITS the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3.7.10

The Control Room Envelope (CRE) boundary may be opened intermittently under administrative control.

Add proposed Required Action D.1 L02 Add proposed Condition E and Required Action E.1 A03 Add proposed Condition F and Required Action F.1 TURKEY POINT - UNITS 3 & 4 3/4 7-18 AMENDMENT NOS. 275 AND 270

ITS ITS 3.7.11 A01 PLANT SYSTEMS 3/4.7.5 CONTROL ROOM EMERGENCY VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION (continued) a.4 With one recirculation damper inoperable, within 7 days, restore the inoperable damper to OPERABLE status or, place and maintain at least one of the recirculation dampers in the open position and place the system in recirculation mode or, immediately suspend all movement of irradiated fuel and 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 for one Unit, or 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for both Units, and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

a.5 With the filter train inoperable, e.g., an inoperable filter, and/or two inoperable recirculation fans, and/or two inoperable recirculation dampers, immediately suspend all movement of irradiated fuel and immediately initiate action to place the compensatory filtration unit in service and verify proper See ITS 3.7.10 operation within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , following which movement of irradiated fuel may resume, and within 7 days, restore the filter train to OPERABLE status.

With the above requirements not met, immediately suspend all movement of irradiated fuel and 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 for one Unit, or 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for both Units, and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

a.6 With an inoperable damper in the normal outside air intak e, within 7 days, restore the inoperable damper to OPERABLE status or, place and maintain at least one of the normal outside air intak e isolation dampers in the closed position and place the system in recirculation mode or, immediately suspend all movement of irradiated fuel and 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 for one Unit, or 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for both Units, and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

a.7 With an inoperable damper in the emergency outside air intak e, within 7 days, restore the inoperable damper to OPERABLE status or, place and maintain at least one of the emergency outside air intak e isolation dampers in the open position or, immediately suspend all movement of irradiated fuel and 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 for one Unit, or 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for both Units, and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

TURKEY POINT - UNITS 3 & 4 3/4 7-19 AMENDMENT NOS. 275 AND 270

ITS ITS 3.7.11 A01 PLANT SYSTEMS 3/4.7.5 CONTROL ROOM EMERGENCY VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION (continued)

b. With the Control Room Emergency Ventilation System inoperable due to an inoperable CRE boundary during MODES 1, 2, 3 or 4, immediately initiate action to implement mitigating actions.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , verify mitigating actions ensure CRE occupant radiological and chemical hazards will not exceed limits, and CRE occupants are protected from smok e hazards, and restore CRE boundary to OPERABLE status within 90 days.

See ITS 3.7.10 With the above requirements not met, 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 for one Unit, or 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for both Units, and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

With the Control Room Emergency Ventilation System inoperable due to an inoperable CRE boundary during MODES 5, 6 or during the movement of irradiated fuel assemblies, immediately suspend all movement of irradiated fuel.

SURVEILLANCE REQUIREMENTS SR 3.7.11.1 4.7.5 The Control Room Emergency Ventilation System shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying that the control room air temperature is less than or equal to 120 F; See ITS b. In accordance with the Surveillance Frequency Control Program by initiating, from the control room, 3.7.10 flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 15 minutes**;

c. In accordance with the Surveillance Frequency Control Program or (1) after 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system See ITS operation, or (2) after any structural maintenance on the HEPA filter or charcoal adsorber housings, 5.5.8 or (3) following exposure of the filters to effluents from painting, fire, or chemical release in any ventilation zone communicating with the system that may have an adverse effect on the functional capability of the system, or (4) after complete or partial replacement of a filter bank by:

See ITS **As the mitigation actions of TS 3.7.5 Action a.5 include the use of the compensatory filtration unit, the unit shall 3.7.10 meet the surveillance requirements of TS 4.7.5.b, by manual initiation from outside the control room and TS 4.7.5.c, d and f.

TURKEY POINT - UNITS 3 & 4 3/4 7-20 AMENDMENT NOS. 275 AND 270

ITS ITS 3.7.11 A01 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

1) Verifying that the air cleanup system satisfies the in-place penetration and bypass leak age testing acceptance criteria of greater than or equal to 99.95% DOP and 99%

halogenated hydrocarbon removal at a system flow rate of 1000 cfm +/-10%**.

See ITS 5.5.8 2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, and analyzed per ASTM D3803 - 1989 at 30 C and 95% relative humidity, meets the methyl iodide penetration criteria of less than 2.5% or the charcoal be replaced with charcoal that meets or exceeds the stated performance requirement**, and

3) Verifying by a visual inspection the absence of foreign materials and gask et deterioration**.

d.1 In accordance with the Surveillance Frequency Control Program by verifying that the pressure drop across the combined HEPA filters and charcoal adsorber bank s is less than 6 inches Water Gauge while operating the system at a flow rate of 1000 cfm +/-10%**;

See ITS 5.5.15 d.2 In accordance with the Surveillance Frequency Control Program, test the supply fans (trains A and B) and measure CRE pressure relative to external areas adj acent to the CRE boundary.**

e. In accordance with the Surveillance Frequency Control Program by verifying that on a Containment Phase A Isolation test signal the system automatically switches into the recirculation mode of operation,

f. By performing required CRE unfiltered air inleak age testing in accordance with the Control Room See ITS Envelope Habitability Program.**

3.7.10

- As the mitigation actions of TS 3.7.5 Action a.5 include the use of the compensatory filtration unit, the unit shall meet the surveillance requirements of TS 4.7.5.b, by manual initiation from outside the control room and TS 4.7.5.c, d and f.

TURKEY POINT - UNITS 3 & 4 3/4 7-21 AMENDMENT NOS. 275 AND 270

DISCUSSION OF CHANGES ITS 3.7.11, CONTROL ROOM EMERGENCY AIR TEMPERATURE CONTROL SYSTEM (CREATCS)

ADMINISTRATIVE CHANGES A01 In the conversion of the Turkey Point Nuclear Generating Station (PTN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 5.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 3.7.5 requires, in part, that the Control Room Emergency Ventilation System shall be OPERABLE* with: a. Three air handling units, and b. Two condensing units. ITS Limiting Condition for Operation (LCO) 3.7.11 states that two Control Room Emergency Air Temperature Control System (CREATCS) trains shall be OPERABLE. This changes the CTS by providing a separate LCO for the CREATCS.

The purpose of CTS 3.7.5 is to provide control room emergency filtration l R2 l

(recirculation) and control room cooling requirements and includes heating, l ventilation, and air conditioning (HVAC) units. Each HVAC unit consists of a l condenser cooling unit and an AHU. ITS separates the requirements based on l l

safety function. ITS 3.7.10 contains the requirements for the emergency filtration l function while ITS 3.7.11 contains the requirements for the control room cooling l function. Since the AHUs support both the filtration function and the cooling l l

function, the AHUs are required to support OPERABILITY of CREVS and l CREATCS. l l

l ITS LCO 3.7.11 requires two of three CREATCS trains (i.e., HVAC units), with l each CREATCS train consisting of a condenser cooling unit and an AHU, to be l l

OPERABLE for control room temperature and humidity control. As such, two l condensing units and two AHUs are required to meet ITS LCO 3.7.11. Other l components listed in CTS 3.7.5, including the requirement to maintain three l l

AHUs OPERABLE for the control room emergency filtration function, are retained l in ITS 3.7.10, CREVS. This change represents a change in presentation of the l l

existing requirements and is designated as administrative because no technical l changes are being made to the CTS.

A03 CTS 3.7.5 does not provide any actions to take if two CREATCS trains are inoperable. ITS 3.7.11, Condition E and Required Action E.1, provide Action for two inoperable CREATCS trains when in MODES 5 or 6, or during the movement of irradiated fuel assemblies. ITS 3.7.11, Condition F and Required Action F.1, provide Action for two inoperable CREATCS trains when in MODES 1, 2, 3, or 4.

This changes the CTS by explicitly providing action to be taken when two CREATCS trains are inoperable.

Turkey Point Unit 3 and Unit 4 Page 1 of 4

DISCUSSION OF CHANGES ITS 3.7.11, CONTROL ROOM EMERGENCY AIR TEMPERATURE CONTROL SYSTEM (CREATCS)

CTS 3.7.5 does not provide any actions to be taken if two CREATCS trains are inoperable. CTS 3.0.3 states (applicable in MODES 1, 2, 3, or 4), in part, that when an 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, as applicable, in the listed conditions within the specified times. ITS 3.7.11, Condition F states that if two CREATCS trains are inoperable to immediately enter LCO 3.0.3 (also applicable in MODES 1, 2, 3, or 4), providing similar direction and in CTS. ITS 3.7.11, Condition E, requires the suspension of the movement of irradiated fuel assemblies when operating in MODES 5 or 6, or during the movement of irradiated fuel assemblies. This Action is the same as CTS 3.7.5, ACTION a.2, for one inoperable CREATCS train. Therefore, the adoption of ITS 3.7.11, ACTIONS E and F, are designated as Administrative because no technical changes are being made to the CTS.

MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including l R2 Design Limits) CTS 3.7.5 lists the components that comprise the CREATCS l which are required to be OPERABLE in the LCO. ITS 3.7.11 does not list the components in the LCO. This changes the CTS by removing the specific components that comprise the CREATCS from the Technical Specifications to the Technical Specification Bases.

The removal of CREATCS components from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications in order to provide adequate protection of public health and safety. The ITS requires the CREATCS to be OPERABLE and lists the components in the Bases. This change is acceptable because this type of detail will be adequately controlled in the Technical Specification Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because CREATCS components are being removed from the Technical Specifications.

Turkey Point Unit 3 and Unit 4 Page 2 of 4

DISCUSSION OF CHANGES ITS 3.7.11, CONTROL ROOM EMERGENCY AIR TEMPERATURE CONTROL SYSTEM (CREATCS)

LESS RESTRICTIVE CHANGES L01 (Category 4 - Relaxation of Required Action) CTS 3.7.5, ACTION a.2, identifies a degraded condition of the CREATCS and provides specific completion times to restore the degraded condition or commence a unit shutdown. If a unit shutdown is required, CTS 3.7.5, ACTION a.2, requires the unit be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months if a dual unit shutdown) and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . ITS 3.7.11 ACTION B states that if the Required Action and associated Completion Time of CREATCS degraded conditions are not met to be in MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 4 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and is modified by a Note stating that LCO 3.0.4.a is not applicable when entering MODE 4. ITS 3.7.11, ACTION C, states that if the Required Action and associated Completion Time of CREATCS degraded conditions are not met to be in MODE 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and MODE 4 in 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months and is modified by a Note stating that LCO 3.0.4.a is not applicable when entering MODE 4. This changes the CTS by allowing a Required Action end state of HOT SHUTDOWN (MODE 4) rather than an end state of COLD SHUTDOWN (MODE 5).

One purpose of CTS 3.7.5, ACTION a.2, is to provide an end state, a condition that the reactor must be placed in, if the Required Actions allowing remedial measures to be taken in response to the degraded conditions with continued operation are not met. End states are usually defined based on placing the unit into a MODE or condition in which the Technical Specification LCO is not applicable. MODE 5 is the current end state for LCOs that are applicable in MODES 1 through 4. This change is acceptable because the risk of the transition from MODE 1 to MODES 4 or 5 depends on the availability of alternating current (AC) sources and the ability to remove decay heat such that remaining in MODE 4 may be safer. During the realignment from MODE 4 to MODE 5, there is an increased potential for loss of shutdown cooling and loss of inventory events. Decay heat removal following a loss-of-offsite power event in MODE 5 is dependent on AC power for shutdown cooling whereas, in MODE 4, the turbine driven auxiliary feedwater (AFW) pump will be available. Therefore, transitioning to MODE 5 is not always the appropriate end state from a risk perspective. Thus, for specific TS conditions, Westinghouse Topical Report WCAP-16294-A, Revision 1 (ADAMS Accession No. ML103430249), justifies MODE 4 as an acceptable alternate end state to Mode 5. The proposed change to the Technical Specifications will allow time to perform short-duration repairs, which currently necessitate exiting the original mode of applicability. The MODE 4 TS end state is applied, and risk is assessed and managed in accordance with Title 10 of the Code of Federal Regulations (10 CFR)

Section 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants." This proposed change is consistent with NRC approved Technical Specification Task Force (TSTF) traveler TSTF-432-A, Revision 1 (ADAMS Accession No. ML103360003), noticed for availability by the NRC in the Federal Register (77 FR 27814) on May 11, 2012. The NRC's approval of WCAP-16294-A included four limitations and conditions on its use as identified in Section 4.0 of the NRC Safety Evaluation associated with WCAP-16294-A.

Implementation of these stipulations were addressed in the Bases of TSTF-432-A. Florida Power & Light implemented these limitations and Turkey Point Unit 3 and Unit 4 Page 3 of 4

DISCUSSION OF CHANGES ITS 3.7.11, CONTROL ROOM EMERGENCY AIR TEMPERATURE CONTROL SYSTEM (CREATCS) conditions at PTN in the adoption of the associated TSTF-432-A Bases. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS.

L02 (Category 4 - Relaxation of Required Action) ISTS 3.7.11, Required Action C.1 (ITS 3.7.11, Required Action D.1), provides an option of placing the OPERABLE CREATCS train in service in lieu of suspending the movement of irradiated fuel assemblies when one CREATCS train is inoperable in MODES 5 or 6, or during the movement of irradiated fuel assemblies. CTS 3.7.5, ACTION a.2, does not include this option. This changes the CTS by explicitly providing an optional Action to be taken when one CREATCS train is inoperable in MODES 5 or 6, or during the movement of irradiated fuel assemblies.

When one CREATCS train is inoperable in MODES 5 or 6, or during the movement of irradiated fuel assemblies, CTS 3.7.5, ACTION a.2, does not provide an option to place the remaining OPERABLE CREATCS train in service.

This option, provided by ISTS 3.7.11, Required Action C.1 (ITS 3.7.11, Required Action D.1), is appropriate because this Action ensures that the remaining train is OPERABLE, that no failures preventing automatic actuation will occur, and that active failures will be readily detected. In lieu of placing the OPERABLE CREATCS train in service, the unit can opt to suspend the movement of irradiated fuel assemblies, consistent with current requirements. This change is acceptable because the Required Actions establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS.

Turkey Point Unit 3 and Unit 4 Page 4 of 4

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

CREATCS 3.7.11 3.7 PLANT SYSTEMS 3.7.11 Control Room Emergency Air Temperature Control System (CREATCS)

LCO 3.7.11 Two CREATCS trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, 4, [5, and 6], 1 During movement of [recently] irradiated fuel assemblies.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME required Action a.2 A. One CREATCS train A.1 Restore CREATCS train to 30 days 3 l R2 inoperable. OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action a.2 associated Completion Time of Condition A not AND met in MODE 1, 2, 3, or 4. B.2 --------------NOTE--------------

NOTE------------ LCO 3.0.4.a is not Not applicable when a dual 2 applicable when entering unit shutdown is required.

MODE 4.

Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 2 INSERT 1 D

C. Required Action and C.1 Place OPERABLE Immediately Action a.2 associated Completion CREATCS train in Time of Condition A not operation.

met [in MODE 5 or 6, or] 1 during movement of OR l R2

[recently] irradiated fuel 1 assemblies. C.2 Suspend movement of Immediately

[recently] irradiated fuel 1 assemblies.

Westinghouse STS 3.7.11-1 Rev. 5.0 3 Turkey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY

2 CTS INSERT 1 C. ------------NOTE------------ C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months l R2 Only applicable when a dual unit shutdown is AND required.

C.2 -------------NOTE----------------

ACTION a.1, Required Action and LCO 3.0.4.a is not ACTION a.2 associated Completion applicable when entering Time of Condition A not MODE 4.

met in MODE 1, 2, 3, or 4. -------------------------------------

Be in MODE 4. 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months Insert Page 3.7.11-1

CREATCS 3.7.11 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E required E 3

D. Two CREATCS trains D.1 Suspend movement of Immediately DOC A03 inoperable [in MODE 5 [recently] irradiated fuel or 6, or] during assemblies. 1 movement of [recently]

irradiated fuel assemblies.

F required F 3

DOC A03 E. Two CREATCS trains E.1 Enter LCO 3.0.3. Immediately inoperable in MODE 1, 2, 3, or 4.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.7.5.a SR 3.7.11.1 Verify each CREATCS train has the capability to [ [18] months 2 remove the assumed heat load. 1 OR control room air temperature is 120ºF.

In accordance with the Surveillance Frequency Control Program ]

Westinghouse STS 3.7.11-2 Rev. 5.0 3 Turkey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY

JUSTIFICATION FOR DEVIATIONS ITS 3.7.11, CONTROL ROOM EMERGENCY AIR TEMPERATURE CONTROL SYSTEM (CREATCS)

1. The Improved Standard Technical Specifications (ISTS) contain bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

2. Changes made to reflect the current licensing basis. In addition, renumbering is required due to added Actions.

3. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

CREATCS B 3.7.11 B 3.7 PLANT SYSTEMS B 3.7.11 Control Room Emergency Air Temperature Control System (CREATCS)

BASES BACKGROUND The CREATCS provides temperature control for the control room following isolation of the control room.

three air conditioning units The CREATCS consists of two independent and redundant trains that provide cooling and heating of recirculated control room air. Each train INSERT 1 consists of heating coils, cooling coils, instrumentation, and controls to 1 provide for control room temperature control. The CREATCS is a the Control Building subsystem providing air temperature control for the control room. l R2 Ventilation System s l l

CRAC unit The CREATCS is an emergency system, parts of which may also operate 1 l 120ºF during normal unit operations. A single train will provide the required temperature control to maintain the control room between [70]° and [85]°. 2 The CREATCS operation in maintaining the control room temperature is U 9.9 1 2 discussed in the FSAR, Section [6.4] (Ref. 1).

APPLICABLE The design basis of the CREATCS is to maintain the control room SAFETY temperature for 30 days of continuous occupancy.

ANALYSES units The CREATCS components are arranged in redundant, safety related 120ºF trains. During emergency operation, the CREATCS maintains the 1 temperature between [70]° and [85]°. A single active failure of a 2 component of the CREATCS, with a loss of offsite power, does not impair the ability of the system to perform its design function. Redundant detectors and controls are provided for control room temperature control.

The CREATCS is designed in accordance with Seismic Category I requirements. The CREATCS is capable of removing sensible and latent heat loads from the control room, which include consideration of equipment heat loads and personnel occupancy requirements, to ensure equipment OPERABILITY.

The CREATCS satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Two independent and redundant trains of the CREATCS are required to be OPERABLE to ensure that at least one is available, assuming a single failure disabling the other train. Total system failure could result in the equipment operating temperature exceeding limits in the event of an accident.

Westinghouse STS B 3.7.11-1 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 Revision XXX

ITS 3.7.11 1

INSERT 1 The control room is maintained at the personnel comfort level of (70 + 10)°F. Protective equipment inside the room is designed to operate within design tolerance over this temperature range and will perform its protective function in an ambient of 120°F and 95% relative humidity (i.e., there will be no loss-of-function in an ambient temperature of 120°F).

Each air conditioning unit consists of a condensing unit, an air handling unit (AHU), l R2 l

instrumentation, and controls to provide for control room temperature control. The AHUs also l support the OPERABILITY requirements of LCO 3.7.10, "Control Room Emergency Ventilation l l

System (CREVS)." Refer to LCO 3.7.10 for AHU requirements associated with CREVS. All l three control room air condition (CRAC) units (AHU and condensing unit) are powered by swing l power sources, each of which can be powered by the emergency diesel generators. One CRAC unit is powered by MCC 3D, one unit by MCC 4D, and the third unit is powered via a transfer switch which automatically transfers between MCCs 3B and 4B. This configuration precludes the loss of more than one CRAC unit for any postulated single failure. Control room equipment is designed to operate in an environment of 120°F and 95% relative humidity. If two of three units were inoperable, the third would maintain the environment within these limits. l R2 INSERT B 3.7.11-1

CREATCS B 3.7.11 BASES LCO (continued)

The CREATCS is considered to be OPERABLE when the individual components necessary to maintain the control room temperature are OPERABLE in both trains. These components include the heating and 1 cooling coils and associated temperature control instrumentation. In addition, the CREATCS must be operable to the extent that air circulation can be maintained.

APPLICABILITY In MODES 1, 2, 3, 4, [5, and 6,] and during movement of [recently] 2 irradiated fuel assemblies, the CREATCS must be OPERABLE to ensure that the control room temperature will not exceed equipment operational requirements following isolation of the control room. [The CREATCS is only required to be OPERABLE during fuel handling involving handling 2

recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous [X] days), due to radioactive decay.]

[In MODE 5 or 6,] CREATCS may not be required for those facilities that 2 1 do not require automatic control room isolation.

ACTIONS A.1 With one CREATCS train inoperable, action must be taken to restore OPERABLE status within 30 days. In this Condition, the remaining OPERABLE CREATCS train is adequate to maintain the control room temperature within limits. However, the overall reliability is reduced because a single failure in the OPERABLE CREATCS train could result in loss of CREATCS function. The 30-day Completion Time is based on the low probability of an event requiring control room isolation, the consideration that the remaining train can provide the required protection, and that alternate safety or non-safety related cooling means are available.

B.1 and B.2 Condition B is modified by a Note stating that Condition B is not In MODE 1, 2, 3, or 4, if the inoperable CREATCS train cannot be applicable when a dual unit restored to OPERABLE status within the required Completion Time, the shutdown is required. Since both units share the same Control unit must be placed in a MODE in which the overall plant risk is reduced.

Room, both units can be affected by CREATCS inoperabilities, To achieve this status, the unit must be placed in at least MODE 3 within 1

which may require a shutdown of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

both units. When a dual unit shutdown is required, Condition C l R2 is applicable. Remaining within the Applicability of the LCO is acceptable to accomplish short duration repairs to restore inoperable equipment because the plant risk in MODE 4 is similar to or lower than MODE 5 (Ref. 2). In MODE 4 the steam generators and Residual Heat Removal System are available to remove decay heat, which provides diversity and defense in depth. As Westinghouse STS B 3.7.11-2 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 Revision XXX

CREATCS B 3.7.11 BASES ACTIONS (continued) stated in Reference 2, the steam turbine driven auxiliary feedwater pump must be available to remain in MODE 4. Should steam generator cooling be lost while relying on this Required Action, there are preplanned actions to ensure long-term decay heat removal. Voluntary entry into MODE 5 may be made as it is also acceptable from a risk perspective.

Required Action B.2 is modified by a Note that states that LCO 3.0.4.a is not applicable when entering MODE 4. This Note prohibits the use of LCO 3.0.4.a to enter MODE 4 during startup with the LCO not met.

However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering MODE 4, and establishment of risk management actions, if appropriate. LCO 3.0.4 is not applicable to, and the Note does not preclude, 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 allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power Insert 2 conditions in an orderly manner and without challenging unit systems.

C.1 and C.2 1 D.1 and D.2

[In MODE 5 or 6, or] during movement of [recently] irradiated fuel, if the 2 inoperable CREATCS train cannot be restored to OPERABLE status within the required Completion Time, the OPERABLE CREATCS train must be placed in operation immediately. This action ensures that the remaining train is OPERABLE, that no failures preventing automatic actuation will occur, and that active failures will be readily detected.

D An alternative to Required Action C.1 is to immediately suspend activities 1 that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes accident risk. This does not preclude the movement of fuel to a safe position.

Westinghouse STS B 3.7.11-3 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 Revision XXX

ITS 3.7.11 1

INSERT 2 C.1 and C.2 Condition C is modified by a Note stating that Condition C is only applicable when a dual unit shutdown is required. Since both units share the same Control Room, both units can be affected by CREATCS inoperabilities. Condition B is applicable if only one unit is required to shut down.

In MODE 1, 2, 3, or 4, if the inoperable CREATCS is not restored to OPERABLE status within the required Completion Time the units must be placed in a MODE in which the overall plant risk is reduced. To achieve this status, the second unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and in MODE 4 within 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months . The extra 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is to facilitate a sequential dual unit shutdown.

Remaining within the Applicability of the LCO is acceptable to accomplish short duration repairs to restore inoperable equipment because the plant risk in MODE 4 is similar to or lower than MODE 5 (Ref. 2). In MODE 4 the steam generators and Residual Heat Removal System are available to remove decay heat, which provides diversity and defense in depth. As stated in Reference 2, the steam turbine driven auxiliary feedwater pump must be available to remain in MODE 4. Should steam generator cooling be lost while relying on this Required Action, there are preplanned actions to ensure long-term decay heat removal. Voluntary entry into MODE 5 may be made as it is also acceptable from a risk perspective.

Required Action C.2 is modified by a Note that states that LCO 3.0.4.a is not applicable when l R2 entering MODE 4. This Note prohibits the use of LCO 3.0.4.a to enter MODE 4 during startup with the LCO not met. However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering MODE 4, and establishment of risk management actions, if appropriate. LCO 3.0.4 is not applicable to, and the Note does not preclude, 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 allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

INSERT B 3.7.11-3

CREATCS B 3.7.11 BASES ACTIONS (continued)

E D.1 1

[In MODE 5 or 6, or] during movement of [recently] irradiated fuel 2 assemblies, with two CREATCS trains inoperable, action must be taken immediately to suspend activities that could result in a release of radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk. This does not preclude the movement of fuel to a safe position.

F 1

E.1 If both CREATCS trains are inoperable in MODE 1, 2, 3, or 4, the control room CREATCS may not be capable of performing its intended function.

Therefore, LCO 3.0.3 must be entered immediately.

control room air temperature is 120 ºF. This ensures that 1

SURVEILLANCE SR 3.7.11.1 the temperature in the control room is below the design REQUIREMENTS temperatures for equipment in the control room.

This SR verifies that the heat removal capability of the system is sufficient to remove the heat load assumed in the [safety analyses] in the control 2

room. This SR consists of a combination of testing and calculations.

[ The [18] month Frequency is appropriate since significant degradation of the CREATCS is slow and is not expected over this time period.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency 3

description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

U REFERENCES 1. FSAR, Section [6.4]. 1 2 9.9

2. WCAP-16294-NP-A, Rev. 1, "Risk-Informed Evaluation of Changes to Technical Specification Required Action Endstates for Westinghouse NSSS PWRs," June 2010.

Westinghouse STS B 3.7.11-4 Rev. 5.0 1 Turkey Point Unit 3 and Unit 4 Revision XXX

JUSTIFICATION FOR DEVIATIONS ITS 3.7.11 BASES, CONTROL ROOM EMERGENCY AIR TEMPERATURE CONTROL SYSTEM (CREATCS)

1. Changes are made (additions, deletions, and/or changes) to the Improved Standard Technical Specification (ISTS) Bases that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

3. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.

4. Changes are made to the Improved Technical Specifications (ITS) Bases to reflect changes made to the ITS.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.7.11, CONTROL ROOM EMERGENCY AIR TEMPERATURE CONTROL SYSTEM (CREATCS)

There are no specific No Significant Hazards Considerations for this Specification.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

ATTACHMENT 12 ITS 3.7.12, FUEL STORAGE POOL WATER LEVEL

Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

ITS A01 ITS 3.7.12 REFUELING OPERATIONS 3/4.9.11 WATER LEVEL STORAGE POOL LIMITING CONDITION FOR OPERATION LCO 3.7.12 3.9.11 The water level shall be maintained greater than or equal to elevation 56 - 10 the spent fuel storage l R2 pool.* l During Movement of irradiated fuel assemblies in the fuel storage pool.

L01 Applicability APPLICABILITY: Whenever irradiated fuel assemblies are in the storage pool.

ACTION:

irradiated L01 ACTION A a. With the requirements of the above specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas. A02 L02 l R2 immediately ACTION A.1 b. The provisions of Specification 3.0.3 are not applicable.

Note SURVEILLANCE REQUIREMENTS SR 3.7.12.1 4.9.11 The water level in the storage pool shall be determined to be at least its minimum required depth in L01 accordance with the Surveillance Frequency Control Program when irradiated fuel assemblies are in the fuel storage pool.

3/4.9.12 DELETED TURKEY POINT - UNITS 3 & 4 3/4 9-11 AMENDMENT NOS. 279 AND 274

DISCUSSION OF CHANGES ITS 3.7.12, FUEL STORAGE POOL WATER LEVEL ADMINISTRATIVE CHANGES A01 In the conversion of the Turkey Point Nuclear Generating Station (PTN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 5.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).

These changes are designated as administrative changes and are acceptable, because they do not result in technical changes to the CTS.

A02 CTS 3.9.11 ACTION states, in part, that with the requirements of the l R2 Specification not satisfied, to suspend all movement of fuel assemblies.

ITS 3.7.12 Required Action A.1 requires the immediate suspension of movement of irradiated fuel assemblies in the spent fuel pool. This changes the CTS by explicitly specifying that the compensatory action to suspend all movement of fuel assemblies requires an immediate response, not to preclude movement of a fuel assembly to a safe position.

The purpose of the CTS 3.9.11 ACTION a is to preclude a fuel handling accident from occurring. The current action does not specify a time; however, it implies that the action is immediate. This change is acceptable because it only provides clarification that the compensatory action requires an immediate response. This change is designated as administrative because it does not result in a technical change to the CTS.

MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01 (Category 2 - Relaxation of Applicability) CTS 3.9.11 Applicability states "Whenever irradiated fuel assemblies are in the spent fuel pool." CTS Surveillance Requirement (SR) 4.9.11 also contains a statement that it is required to be performed when, "irradiated fuel assemblies are in the fuel storage pool." CTS ACTION a requires suspension of "movement of fuel assemblies" Turkey Point Unit 3 and Unit 4 Page 1 of 2

DISCUSSION OF CHANGES ITS 3.7.12, FUEL STORAGE POOL WATER LEVEL when the Limiting Condition for Operation (LCO) is not satisfied. ITS 3.7.12 LCO is applicable "During movement of irradiated fuel assemblies in the spent fuel pool." In addition, ITS ACTION A requires the suspension of "movement of irradiated fuel assemblies" when the minimum water level requirements cannot be met. This changes the CTS by relaxing the Applicability and performance of the SR to only during movement of irradiated fuel assemblies, as well as relaxing the Action to suspending movement of irradiated fuel assemblies. These changes are consistent with the assumptions of the fuel handling accident analysis.

The purpose of CTS 3.9.11 is to ensure that sufficient shielding will be available during fuel movement and for removal of iodine in the event of a fuel handling accident. The spent fuel minimum water depth is consistent with the assumptions of the safety analysis. This change is acceptable because the proposed changes are consistent with the fuel handling accident assumptions.

The proposed changes will continue to ensure that the conditions assumed in the safety analyses and licensing basis are maintained. This change is designated as less restrictive because less stringent Technical Specification requirements are allowed in the ITS than are allowed in the CTS.

L02 (Category 4 - Relaxation of Required Action) CTS 3.9.11 ACTION a states, in part, that when the spent fuel pool water level is not met, suspend all crane operations with loads in the fuel storage areas. ITS 3.7.12 Required Action A.1 states that when spent fuel pool water level is not within limits, immediately suspend movement of irradiated fuel assemblies in the spent fuel pool. This changes the CTS by deleting the requirements to suspend crane operations over the spent fuel storage areas.

The purpose of the CTS 3.9.11 ACTION is to preclude a fuel handling accident from occurring; however, the fuel handling accident assumes a spent fuel assembly is dropped and not loads carried by the crane. In 2011, the spent fuel cask handling crane was upgraded to the single-failure-proof criteria of NUREG-0554, "Single-Failure-Proof Cranes for Nuclear Power Plants." The NRC approved the use of the upgraded cask crane as single-failure-proof in License Amendments 243 and 239 which authorized the deletion of the cask drop accident from the safety analysis. Therefore, a fuel handling accident is only assumed to occur when an irradiated fuel assembly is being moved over the fuel storage pool. ITS 3.7.12 ACTION A requires suspending movement of irradiated fuel. Therefore, requiring the suspension of crane operations over the fuel storage pool is not required per Technical Specifications; however, PTN has administrative controls in place to prevent handling heavy loads over the fuel storage pool. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS.

Turkey Point Unit 3 and Unit 4 Page 2 of 2

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

CTS Fuel Storage Pool Water Level 3.7.15 1 12 3.7 PLANT SYSTEMS 1

12 3.7.15 Fuel Storage Pool Water Level 56 ft 10 inches elevation 12 1 l R2 3.9.11 LCO 3.7.15 The fuel storage pool water level shall be 23 ft over the top of irradiated l 2 l fuel assemblies seated in the storage racks.

Applicability APPLICABILITY: During movement of irradiated fuel assemblies in the fuel storage pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME ACTION a A. Fuel storage pool water A.1 --------------NOTE--------------

ACTION b Note level not within limit. LCO 3.0.3 is not applicable.

Suspend movement of Immediately irradiated fuel assemblies in the fuel storage pool.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 12 56 ft 10 inches elevation l R2 3

SR 4.9.11 SR 3.7.15. 1 Verify the fuel storage pool water level is 23 ft [ 7 days 1 l above the top of the irradiated fuel assemblies l 2 l seated in the storage racks. OR In accordance with the Surveillance Frequency Control Program ] 3 Turkey Point Unit 3 and Unit 4 12 Amendment Nos. XXX and YYY Westinghouse STS 3.7.15-1 Rev. 5.0 2 1

JUSTIFICATION FOR DEVIATIONS ITS 3.7.12, FUEL STORAGE POOL WATER LEVEL

1. Improved Standard Technical Specification (ISTS) 3.7.15 has been renumbered as Improved Technical Specifications (ITS) 3.7.12.

2. Changes are made (additions, deletions, and/or changes) to the ISTS which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

3. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Fuel Storage Pool Water Level 1

B 3.7.15 12 B 3.7 PLANT SYSTEMS 12 2

B 3.7.15 Fuel Storage Pool Water Level BASES BACKGROUND The minimum water level in the fuel storage pool meets the assumptions of iodine decontamination factors following a fuel handling accident. The specified water level shields and minimizes the general area dose when the storage racks are filled to their maximum capacity. The water also provides shielding during the movement of spent fuel.

U A general description of the fuel storage pool design is given in the FSAR, 2 9.5.2 Section [9.1.2] (Ref. 1). A description of the Spent Fuel Pool Cooling and Cleanup System is given in the FSAR, Section [9.1.3] (Ref. 2). The 9.5.3 2 3 U

assumptions of the fuel handling accident are given in the FSAR, 14.2.1 Section [15.7.4] (Ref. 3).

APPLICABLE The minimum water level in the fuel storage pool meets the SAFETY 183 assumptions of the fuel handling accident described in Regulatory 2 ANALYSES Guide 1.25 (Ref. 4). The resultant 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid dose per person at the exclusion area boundary is a small fraction of the 10 CFR 100 (Ref. 5) 50.67 limits.

According to Reference 4, there is 23 ft of water between the top of the damaged fuel bundle and the fuel pool surface during a fuel handling accident. With 23 ft of water, the assumptions of Reference 4 can be used directly. In practice, this LCO preserves this assumption for the bulk the minimum of the fuel in the storage racks. In the case of a single bundle dropped l R2 storage pool level at l an elevation of 56 ft and lying horizontally on top of the spent fuel racks, however, there may l 10 inches ensures be < 23 ft of water above the top of the fuel bundle and the surface, 6 l that at least 23 ft of water remains above indicated by the width of the bundle. To offset this small l l

the damaged fuel nonconservatism, the analysis assumes that all fuel rods fail, although l assembly. The analysis shows that only the first few rows fail from a hypothetical l maximum drop.

The fuel storage pool water level satisfies Criteria 2 and 3 of 10 CFR 50.36(c)(2)(ii). 56 ft 10 inches elevation l R2 l

l LCO The fuel storage pool water level is required to be 23 ft over the top of l equates to 25 feet irradiated fuel assemblies seated in the storage racks. The specified 6 l above the top of the water level preserves the assumptions of the fuel handling accident l spent fuel storage l racks and analysis (Ref. 3). As such, it is the minimum required for fuel storage and l movement within the fuel storage pool.

Turkey Point Unit 3 and Unit 4 12 Revision XXX Westinghouse STS B 3.7.15-1 Rev. 5.0 2 1

Fuel Storage Pool Water Level 1

B 3.7.15 12 BASES APPLICABILITY This LCO applies during movement of irradiated fuel assemblies in the fuel storage pool, since the potential for a release of fission products exists.

ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.

When the initial conditions for prevention of an accident cannot be met, steps should be tak en to preclude the accident from occurring. When the fuel storage pool water level is lower than the required level, the movement of irradiated fuel assemblies in the fuel storage pool is immediately suspended to a safe position. This action effectively precludes the occurrence of a fuel handling accident. This does not preclude movement of a fuel assembly to a safe position.

If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODES 1, 2, 3, and 4, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.

12 SURVEILLANCE SR 3.7.15.1 1 REQUIREMENTS This SR verifies sufficient fuel storage pool water is available in the event of a fuel handling accident. The water level in the fuel storage pool must be check ed periodically. [ The 7 day Frequency is appropriate because the volume in the pool is normally stable. Water level changes are 3 controlled by plant procedures and are acceptable based on operating experience.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER S NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency 4 description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

During refueling operations, the level in the fuel storage pool is in equilibrium with the refueling canal, and the level in the refueling canal is check ed daily in accordance with SR 3.9.6.1. 5 2

Turk ey Point Unit 3 and Unit 4 12 Revision XXX Westinghouse STS B 3.7.15-2 Rev. 5.0 2 1

Fuel Storage Pool Water Level 1

B 3.7.15 12 BASES U 5 REFERENCES 1. FSAR, Section [ 9.1.2] .

U 5

2. FSAR, Section [ 9.1.3] .

2 3 U 14.2.1

3. FSAR, Section [ 15.7.4] .

183

4. Regulatory Guide 1.25, [ Rev. 0] .

50.67 2

5. 10 CFR 100.11.

Turk ey Point Unit 3 and Unit 4 12 Revision XXX Westinghouse STS B 3.7.15-3 Rev. 5.0 2 1

JUSTIFICATION FOR DEVIATIONS ITS 3.7.12 BASES, FUEL STORAGE POOL WATER LEVEL

1. Improved Standard Technical Specification (ISTS) B 3.7.15, "Fuel Storage Pool Water Level" has been renumbered as Improved Technical Specifications (ITS) B 3.7.12, "Fuel Storage Pool Water Level."

2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

3. The ISTS Bases contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

4. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.

5. ITS 3.7.12 Bases references SR 3.6.9.1 as verifying canal water level. Canal water level is verified by SR 3.9.2.1.

6. Information is added to the ITS 3.7.12 Bases associated with the current licensing basis l R2 l

minimum storage pool level based on a plant elevation of 56 ft 10 inches. A water level at a l plant elevation of 56 ft 10 inches is equivalent to 25 feet of water above the top of the fuel l storage racks in the storage pool and ensures the fuel handling accident analysis l l

assumption of 23 feet of water above a damaged fuel assembly is maintained. l Turkey Point Unit 3 and Unit 4 Page 1 of 1

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.7.12, FUEL STORAGE POOL WATER LEVEL There are no specific No Significant Hazards Considerations for this Specification.

Turkey Point Unit 3 and 4 Page 1 of 1

ATTACHMENT 13 ITS 3.7.13, FUEL STORAGE POOL BORON CONCENTRATION

Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

A01 ITS ITS 3.7.13 REFUELING OPERATIONS Fuel Storage Pool Boron Concentration A01 3/4.9.14 SPENT FUEL STORAGE LIMITING CONDITION FOR OPERATION LCO 3.7.13 3.9.14 The following conditions shall apply to spent fuel storage:

storage pool

a. The minimum boron concentration in the Spent Fuel Pit shall be 2300 ppm. A01

b. The combination of initial enrichment, burnup, and cooling time of each fuel assembly stored in See ITS the Spent Fuel Pit shall be in accordance with Specification 5.5.1. 3.7.14 Applicability APPLICABILITY: At all times when fuel is stored in the Spent Fuel Pit. A01 assemblies are storage pool and a fuel storage pool verification ACTION: storage pool has not been performed since the last movement L01 of fuel assemblies in the fuel storage pool ACTION A.1 a. With boron concentration in the Spent Fuel Pit less than 2300 ppm, suspend movement of spent ACTION A.2 A01 fuel in the Spent Fuel Pit and initiate action to restore boron concentration to 2300 ppm or greater. storage pool or Initiate action to perform a fuel storage pool verification, immediately L01

b. With condition b not satisfied, suspend movement of additional fuel assemblies into the Spent Fuel Pit and restore the spent fuel storage configuration to within the specified conditions.

ACTION A c. The provisions of Specification 3.0.3 are not applicable. See ITS Note 3.7.14 SURVEI LLANCE REQUIREMENTS SR 3.7.13.1 4.9.14.1 The boron concentration of the Spent Fuel Pit shall be verified to be 2300 ppm or greater in accordance with the Surveillance Frequency Control Program.

4.9.14.2 A representative sample of inservice Metamic inserts shall be visually inspected in accordance with See ITS the Metamic Surveillance Program described in UFSAR Section 16.2. The surveillance program 3.7.14 ensures that the performance requirements of Metamic are met over the surveillance interval.

TURKEY POINT - UNITS 3 & 4 3/4 9-13 AMENDMENT NOS. 263 AND 258

DISCUSSION OF CHANGES ITS 3.7.13, FUEL STORAGE POOL BORON CONCENTRATION ADMINISTRATIVE CHANGES A01 In the conversion of the Turkey Point Nuclear Generating Station (PTN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 5.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01 (Category 2 - Relaxation of Applicability) CTS 3.9.14 Limiting Condition for Operation (LCO) a states "The minimum boron concentration in the fuel storage pool shall be 2300 ppm." CTS 3.9.14 ACTION a states "With boron concentration in the fuel storage pool less than 2300 ppm, suspend movement of spent fuel in the fuel storage pool and initiate action to restore boron concentration to 2300 ppm or greater." ITS 3.7.13 requires that the fuel pool boron concentration shall be 2300 ppm when fuel assemblies are stored in the fuel pool and a fuel pool verification has not been performed since the last movement of fuel assemblies in the spent fuel pool. Additionally, a new Required Action (Required Action A.2.2) has been added to allow the initiation of an action to perform a fuel pool verification as an option to initiating an action to restore fuel pool boron concentration to within limit. This changes the CTS by changing the Applicability and by adding an option for restoring the fuel pool boron concentration to within its limit.

The purpose of CTS 3.9.14 is to ensure adequate dissolved boron is in spent fuel storage water to maintain the required subcriticality margin in the event of a fuel handling accident. This change is acceptable because the requirements continue to ensure that the boron concentration is maintained during the specified conditions assumed in the safety analyses and licensing basis (i.e.,

during fuel movement). Performing a spent fuel pool verification provides Turkey Point Unit 3 and Unit 4 Page 1 of 2

DISCUSSION OF CHANGES ITS 3.7.13, FUEL STORAGE POOL BORON CONCENTRATION assurance that no fuel assemblies have been inadvertently misplaced in the spent fuel storage pool. This change is designated as less restrictive because the LCO requirements in the ITS are more relaxed than the requirements in the CTS.

Turkey Point Unit 3 and Unit 4 Page 2 of 2

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

CTS [ Fuel Storage Pool Boron Concentration] 2 3.7.16 1 13 3.7 PLANT SYSTEMS 13 1 2 3.7.16 [ Fuel Storage Pool Boron Concentration ]

13 3.9.14 LCO 3.7.16 to oo oo o t to [ 2300] ppm. 1 2 Applicability APPLICABILITY: When fuel assemblies are stored in the fuel storage pool and a fuel storage pool verification has not been performed since the last movement of fuel assemblies in the fuel storage pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME ACTION a A. Fuel storage pool boron --------------------NOTE-------------------

ACTION c concentration not within LCO 3.0.3 is not applicable.

limit. ------------------------------------------------

A.1 Suspend movement of fuel Immediately assemblies in the fuel storage pool.

AND A.2.1 Initiate action to restore fuel Immediately storage pool boron concentration to within limit.

OR DOC L01 A.2.2 Initiate action to perform a Immediately fuel storage pool verification.

Turk ey Point Unit 3 and Unit 4 13 Amendment Nos. XXX and YYY Westinghouse STS 3.7.16-1 Rev. 5.0 3 1

CTS [ Fuel Storage Pool Boron Concentration] 2 3.7.16 1 13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 13 SR 4.9.14.1 SR 3.7.16.1 Verify the fuel storage pool boron concentration is [ 7 days 2 within limit.

OR In accordance with the Surveillance Frequency Control Program ]

Turk ey Point Unit 3 and Unit 4 13 Amendment Nos. XXX and YYY Westinghouse STS 3.7.16-2 Rev. 5.0 3 1

JUSTIFICATION FOR DEVIATIONS ITS 3.7.13, FUEL STORAGE POOL BORON CONCENTRATION

1. Improved Standard Technical Specification (ISTS) 3.7.16 has been renumbered as Improved Technical Specifications (ITS) 3.7.13.

2. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

3. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

[Fuel Storage Pool Boron Concentration] 1 B 3.7.16 2 13 B 3.7 PLANT SYSTEMS B 3.7.16 [ Fuel Storage Pool Boron Concentration ] 2 1 13 BASES BACKGROUND In the Maximum Density Rack (MDR) [(Refs. 1 and 2)] design, the spent fuel storage pool is divided into two separate and distinct regions which, for the purpose of criticality considerations, are considered as separate pools. [Region 1], with [336] storage positions, is designed to accommodate new fuel with a maximum enrichment of [4.65] wt% U-235, or spent fuel regardless of the discharge fuel burnup. [Region 2], with 3

[2670] storage positions, is designed to accommodate fuel of various initial enrichments which have accumulated minimum burnups within the acceptable domain according to Figure [3.7.17-1], in the accompanying LCO. Fuel assemblies not meeting the criteria of Figure [3.7.17-1] shall be stored in accordance with paragraph 4.3.1.1 in Section 4.3, Fuel Storage. INSERT 1 3 The water in the spent fuel storage pool normally contains soluble boron, which results in large subcriticality margins under actual operating conditions. However, the NRC guidelines, based upon the accident condition in which all soluble poison is assumed to have been lost, specify that the limiting keff of 0.95 be evaluated in the absence of soluble boron. Hence, the design of both regions is based on the use of unborated water, which maintains each region in a subcritical condition during normal operation with the regions fully loaded. The double /ANS-8.1-1983 1

contingency principle discussed in ANSI N-16.1-1975 and the April 1978 3

NRC letter (Ref. 3) allows credit for soluble boron under other abnormal or accident conditions, since only a single accident need be considered at one time. For example, the most severe accident scenario is associated with the movement of fuel from [Region 1 to Region 2], and accidental misloading of a fuel assembly in [Region 2]. This could potentially 3 increase the criticality of [Region 2]. To mitigate these postulated criticality related accidents, boron is dissolved in the pool water. Safe operation of the MDR with no movement of assemblies may therefore be achieved by controlling the location of each assembly in accordance with 13 LCO 3.7.17, "Spent Fuel Assembly Storage." Prior to movement of an 3 l R2 assembly, it is necessary to perform SR 3.7.16.1.

APPLICABLE Most accident conditions do not result in an increase in the activity of SAFETY either of the two regions. Examples of these accident conditions are the ANALYSES loss of cooling (reactivity increase with decreasing water density) and the dropping of a fuel assembly on the top of the rack. However, accidents 3 can be postulated that could increase the reactivity. This increase in reactivity is unacceptable with unborated water in the storage pool. Thus, for these accident occurrences, the presence of soluble boron in the Turkey Point Unit 3 and Unit 4 13 Revision XXX 2 Westinghouse STS B 3.7.16-1 Rev. 5.0 3

3 INSERT 1 The Spent Fuel Storage racks provide safe subcritical storage of fuel assemblies by providing sufficient center-to-center spacing or a combination of spacing and poison to assure: a) Keff less than or equal to 0.95 with a minimum soluble boron concentration of 500 ppm present, and b) Keff less than 1.0 when flooded with unborated water for normal operations and postulated accidents. The 500 ppm value is needed to assure keff less than 0.95 for normal operating conditions. The criticality analysis needs 1700 ppm to assure keff less than 0.95 under the worst case accident condition. There is significant margin between the calculated ppm requirement and the spent fuel boron concentration requirement of 2300 ppm. The higher boron concentration value is chosen because, during refueling, the water volume in the spent fuel pool, the transfer canal, the refueling canal, the refueling cavity, and the reactor vessel form a single mass.

The spent fuel racks are divided into two regions, Region I and Region II. The Region I permanent racks have a 10.6 inch center-to-center spacing. The Region lI racks have a 9.0 inch center-to-center spacing. The cask area storage rack has a nominal 10.1 inch center to center spacing in the east-west direction and a nominal 10.7 inch center-to-center spacing in the north-south direction.

Any fuel for use at Turkey Point, and enriched to less than or equal to 5.0 wt % U-235, may be stored in the Cask Area Storage Rack. Fresh or irradiated fuel assemblies not stored in the Cask Area Storage Rack shall be stored in accordance with LCO 3.7.14.

Fresh unirradiated fuel may be placed in the permanent Region I racks in accordance with the restrictions of Figure 3.7.14-1. Prior to placement of irradiated fuel in Region I or II spent fuel storage rack cell locations, strict controls are employed to evaluate burnup of the fuel assembly. Upon determination that the fuel assembly meets the nominal burnup and associated post-irradiation cooling time requirements of Table 3.7.14-1 or Table 3.7.14-2, it may be placed in a Region I or II cell in accordance with the restrictions of Figures 3.7.14-1 through 3.7.14-3, respectively.

For all assemblies with blanketed fuel, the initial enrichment is based on the central zone enrichment (i.e., between the axial blankets) consistent with the assumptions of the analysis. These positive controls assure that the fuel enrichment limits, burnup, and post irradiation cooling time requirements assumed in the safety analyses will NOT be violated.

Insert Page B 3.7.13-1

[ Fuel Storage Pool Boron Concentration] 1 B 3.7.16 2 13 BASES APPLICABLE SAFETY ANALYSES (continued) storage pool prevents criticality in both regions. The postulated accidents are basically of two types. A fuel assembly could be incorrectly transferred from [ Region 1 to Region 2] (e.g., an unirradiated fuel assembly or an insufficiently depleted fuel assembly). The second type of postulated accidents is associated with a fuel assembly which is dropped adj acent to 3 the fully loaded [ Region 2] storage rack . This could have a small positive reactivity effect on [ Region 2] . However, the negative reactivity effect of the soluble boron compensates for the increased reactivity caused by either one of the two postulated accident scenarios. The accident 2

analyses is provided in the FSAR, Section [ 15.7.4] (Ref. 4).

INSERT 2 3 The concentration of dissolved boron in the fuel storage pool satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO to oo o o o t to to [ 2300] ppm. 1 The specified concentration of dissolved boron in the fuel storage pool preserves the assumptions used in the analyses of the potential critical 2

accident scenarios as described in Reference 4. This concentration of 3 dissolved boron is the minimum required concentration for fuel assembly storage and movement within the fuel storage pool.

APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel storage pool, until a complete spent fuel storage pool verification has been performed following the last movement of fuel assemblies in the spent fuel storage pool. This LCO does not apply following the verification, since the verification would confirm that there are no misloaded fuel assemblies. With no further fuel assembly movements in progress, there is no potential for a misloaded fuel assembly or a dropped fuel assembly.

ACTIONS A.1, A.2.1, and A.2.2 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.

When the concentration of boron in the fuel storage pool is less than required, immediate action must be tak en to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress.

This is most efficiently achieved by immediately suspending the movement of fuel assemblies. The concentration of boron is restored simultaneously with suspending movement of fuel assemblies.

Alternatively, beginning a verification of the fuel storage pool fuel locations, to ensure proper locations of the fuel, can be performed.

However, prior to resuming movement of fuel assemblies, the concentration of boron must be restored. This does not preclude movement of a fuel assembly to a safe position.

Turk ey Point Unit 3 and Unit 4 13 Revision XXX 2 Westinghouse STS B 3.7.16-2 Rev. 5.0 3

3 INSERT 2 The following fuel handling accidents are evaluated to ensure that no hazards are created: a) A fuel assembly is dropped in containment. b) A spent fuel cask is dropped in the passage between the spent fuel pits of Units 3 & 4 while transferring a fuel element between the spent fuel pits. The consideration of a cask drop accident is historical and is retained as discussed in UFSAR Section 14.2.1.3. (Ref.2)

Since the spent fuel cask will not be handled over or in the vicinity of spent fuel except as provided for in UFSAR Section 14.2.1.3.1, the re-racking does not result in a significant increase in the probability of the cask drop accident previously evaluated in the Turkey Point Updated UFSAR. Furthermore, as shown in UFSAR Section 14.2.1.3.2, by requiring the decay time of spent fuel to be a minimum of 1525 hours0.0177 days 0.424 hours 0.00252 weeks 5.802625e-4 months prior to moving a spent fuel cask into the spent fuel pit, the potential offsite doses will be well within 10 CFR Part 100 limits should a dropped cask strike the stored fuel assemblies. The proposed spent fuel pit modifications will not increase the radiological consequences of a cask drop accident previously evaluated. (Ref.2)

Insert Page B 3.7.13-2

[ Fuel Storage Pool Boron Concentration] 1 B 3.7.16 2 13 BASES ACTIONS (continued)

If the LCO is not met while moving irradiated fuel assemblies in MODE 5 or 6, LCO 3.0.3 would not be applicable. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation. Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.

13 SURVEILLANCE SR 3.7.16.1 3 REQUIREMENTS This SR verifies that the concentration of boron in the fuel storage pool is within the required limit. As long as this SR is met, the analyzed accidents are fully addressed. [ The 7 day Frequency is appropriate 1 because no maj or replenishment of pool water is expected to tak e place over such a short period of time.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWER S NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the 4 Surveillance Requirement.

]

REFERENCES [ 1. Callaway FSAR, Appendix 9.1A, " The Maximum Density Rack (MDR)

Design Concept."

2. Description and Evaluation for Proposed Changes to Facility Operating Licenses DPR-39 and DPR-48 (Zion Power Station). ]

/ANS-8.1-1983 1

3. Double contingency principle of ANSI N16.1-1975, as specified in the 3 April 14, 1978 NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).

U 14.2.1 2

3 1

4. FSAR, Section [ 15.7.4] .

Turk ey Point Unit 3 and Unit 4 13 Revision XXX 2 Westinghouse STS B 3.7.16-3 Rev. 5.0 3

JUSTIFICATION FOR DEVIATIONS ITS 3.7.13 BASES, FUEL STORAGE POOL BORON CONCENTRATION

1. The Improved Standard Technical Specification (ISTS) contain bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

2. ISTS B 3.7.16 has been renumbered as Improved Technical Specifications (ITS) B 3.7.13. "Fuel Storage Pool Boron Concentration" does not change.

3. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

4. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.7.13, Fuel Storage Pool Boron Concentration There are no specific No Significant Hazards Considerations for this Specification.

Turkey Point Unit 3 and 4 Page 1 of 1

ATTACHMENT 14 ITS 3.7.14, SPENT FUEL STORAGE

Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

ITS ITS 3.7.14 A01 REFUELING OPERATIONS 3/4.9.14 SPENT FUEL STORAGE LIMITING CONDITION FOR OPERATION LCO 3.7.14 3.9.14 The following conditions shall apply to spent fuel storage:

a. The minimum boron concentration in the Spent Fuel Pit shall be 2300 ppm. See ITS 3.7.13

b. The combination of initial enrichment, burnup, and cooling time of each fuel assembly stored in A02 the Spent Fuel Pit shall be in accordance with Specification 5.5.1. the following:

Insert CTS Specification 5.5.1.1.e and 5.5.1.3 l R2 Applicability APPLICABILITY: At all times when fuel is stored in the Spent Fuel Pit.

See ITS ACTION: 3.7.13

a. With boron concentration in the Spent Fuel Pit less than 2300 ppm, suspend movement of spent fuel in the Spent Fuel Pit and initiate action to restore boron concentration to 2300 ppm or greater.

ACTION A b. With condition b not satisfied, suspend movement of additional fuel assemblies into the Spent Fuel Pit and restore the spent fuel storage configuration to within the specified conditions.

ACTION A.1 c. The provisions of Specification 3.0.3 are not applicable.

Note SURVEI LLANCE REQUIREMENTS 4.9.14.1 The boron concentration of the Spent Fuel Pit shall be verified to be 2300 ppm or greater in See ITS accordance with the Surveillance Frequency Control Program. 3.7.13 4.9.14.2 A representative sample of inservice Metamic inserts shall be visually inspected in accordance with the Metamic Surveillance Program described in UFSAR Section 16.2. The surveillance program LA01 ensures that the performance requirements of Metamic are met over the surveillance interval.

Add proposed SR 3.7.14.1 M01 TURKEY POINT - UNITS 3 & 4 3/4 9-13 AMENDMENT NOS. 263 AND 258

ITS ITS 3.7.14 A01 DESIGN FEATURES 5.5 FUEL STORAGE 5.5.1 CRITICALITY LCO 3.7.14 5.5.1.1 The spent fuel storage rack s are designed and shall be maintained with:

a. A k eff less than 1.0 when flooded with unborated water, which includes an allowance for biases and uncertainties as described in UFSAR Chapter 9.

b. A k eff less than or equal to 0.95 when flooded with water borated to 500 ppm, which includes an See ITS allowance for biases and uncertainties as described in UFSAR Chapter 9.

4.0

c. A nominal 10.6 inch center-to-center distance for Region I and 9.0 inch center-to-center distance for Region II for the two region spent fuel pool storage rack s. A nominal 10.1 inch center-to-center distance in the east-west direction and a nominal 10.7 inch center-to-center distance in the north-south direction for the cask area storage rack .

d. A maximum enrichment loading for fuel assemblies of 5.0 weight percent of U-235.

LCO 3.7.14 e. No restriction on storage of fresh or irradiated fuel assemblies in the cask area storage rack .

f. Fresh or irradiated fuel assemblies not stored in the cask area storage rack shall be stored in accordance with Specification 5.5.1.3.

g. The Metamic neutron absorber inserts shall have a minimum certified 10B areal density greater than or equal to 0.015 grams 10B/cm2.

See ITS 5.5.1.2 The rack s for new fuel storage are designed to store fuel in a safe subcritical array and shall be 4.0 maintained with:

a. A nominal 21 inch center-to-center spacing to assure k eff equal to or less than 0.98 for optimum moderation conditions and equal to or less than 0.95 for fully flooded conditions.

b. Fuel assemblies placed in the New Fuel Storage Area shall contain no more than a nominal 4.5 weight percent of U-235 if the assembly contains no burnable absorber rods and no more than 5.0 weight percent of U-235 if the assembly contains at least 16 IFBA rods.

TURKEY POINT - UNITS 3 & 4 5-2 AMENDMENT NOS. 260 AND 255

ITS 3.7.14 A01 ITS DESIGN FEATURES 5.5.1.3 Credit for burnup and cooling time is tak en in determining acceptable placement locations for spent fuel in the two-region spent fuel rack s. Fresh or irradiated fuel assemblies in the Region I or Region II rack s shall be stored in compliance with the following:

a. Any 2x2 array of Region I storage cells containing fuel shall comply with the storage patterns in Figure 5.5-1 and the requirements of Tables 5.5-1 and 5.5-2, as applicable. The reactivity rank of fuel assemblies in the 2x2 array (rank determined using Table 5.5-3) shall be equal to or less reactive than that shown for the 2x2 array.

A02 LCO 3.7.14 b. Any 2x2 array of Region II storage cells containing fuel shall:

i. Comply with the storage patterns in Figure 5.5-2 and the requirements of Tables 5.5-1 and 5.5-2, as applicable. The reactivity rank of fuel assemblies in the 2x2 array (rank determined using Table 5.5-3) shall be equal to or less reactive than that shown for the 2x2 array, ii. Have the same directional orientation for Metamic inserts in a contiguous group of 2x2 arrays where Metamic inserts are required, and iii. Comply with the requirements of 5.5.1.3.c for cells adj acent to Region I rack s.

c. Any 2x2 array of Region II storage cells that interface with Region I storage cells shall comply with the rules of Figure 5.5-3.

d. Any fuel assembly may be replaced with a fuel rod storage bask et or non-fuel hardware.

e. Storage of Metamic inserts or RCCAs is acceptable in locations designated as empty (water-filled) cells.

DRAINAGE 5.5.2 The spent fuel storage pit is designed and shall be maintained to prevent inadvertent draining of the pool See ITS below a level of 6 feet above the fuel assemblies in the storage rack s.

4.0 CAPACITY 5.5.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1535 fuel assemblies.

TURKEY POINT - UNITS 3 & 4 5-3 AMENDMENT NOS. 260 AND 255

A01 ITS 3.7.14 A02 T a b le 5 .5 -1 TURKEY POINT - UNITS 3 & 4 B la n k e te d F u e l - C o e ffic ie n ts to C a lc u la te th e M in im u m R e q u ir e d F u e l A s s e m b ly B u r n u p (B u ) a s a F u n c tio n o f E n r ic h m e n t (E n ) a n d C o o lin g T im e (C t)

See notes 1-6 for use of Table 5.5-1 F u e l C a te g o ry C o e ff.

I-3 I-4 II-1 II-2 II-3 II-4 II-5 A 1 5.66439153 -14.7363682 -7.74060457 -7.63345029 24.4656526 8.5452608 26.2860949 A 2 -7.22610116 11.0284547 5.13978237 10.7798957 -20.3141124 -4.47257395 -18.0738662 A 3 2.98646188 -1.80672781 -0.360186309 -2.81231555 6.53101471 2.09078914 5.8330891 A 4 -0.287945644 0.119516492 0.0021681285 0.29284474 -0.581826027 -0.188280562 -0.517434342 A 5 -0.558098618 0.0620559676 -0.0304713673 0.0795058096 -0.16567492 0.157548739 -0.0614152031 A 6 0.476169245 0.0236575787 0.098844889 -0.0676341983 0.243843226 -0.0593584027 0.134626308 A 7 -0.117591963 -0.0088144551 -0.0277584786 0.0335130877 -0.0712130368 0.0154678626 -0.0383060399 A 8 0.0095165354 0.0008957348 0.0024057185 -0.0040803875 0.0063998706 -0.0014068318 0.0033419846 A 9 -47.1782783 -20.2890089 -21.424984 14.6716317 -41.1150 -0.881964768 -12.1780 A 1 0 33.4270029 14.7485847 16.255208 -10.0312224 43.9149156 9.69128392 23.6179517 A 1 1 -6.11257501 -1.22889103 -1.77941882 5.62580894 -9.6599923 -0.18740168 -4.10815592 A 1 2 0.490064351 0.0807808548 0.127321203 -0.539361868 0.836931842 0.0123398618 0.363908736 5-4 Notes:

1. All relevant uncertainties are explicitly included in the criticality analysis. For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required. For a fuel assembly to meet the requirements of a Fuel Category, the assembly burnup must exceed the minimum burnup (GWd/MTU) given by the curve fit for the assembly cooling time and initial enrichment. The specific minimum burnup required for each fuel assembly is calculated from the following equation:

AMENDMENT NOS. 260 AND 255 B u = (A 1 + A 2 *E n + A 3 *E n 2

+ A 4 *E n 3 )

e x p [ - ( A 5 + A 6 *E n + A 7 *E n 2

+ A 8 *E n 3 ) *C t ] + A 9 + A 1 0 *E n + A 1 1 *E n 2

+ A 1 2 *E n 3

2. Initial enrichment, En, is the nominal central zone U-235 enrichment. Axial blank et material is not considered when determining enrichment. Any enrichment between 2.0 and 5.0 may be used.

3. Cooling time, Ct, is in years. Any cooling time between 0 years and 25 years may be used. An assembly with a cooling time greater than 25 years must use 25 years.

4. DELETED

5. DELETED

6. This Table applies for any blank eted fuel assembly.

A01 ITS 3.7.14 T a b le 5 .5 -2 A02 TURKEY POINT - UNITS 3 & 4 N o n -B la n k e te d F u e l - C o e ffic ie n ts to C a lc u la te th e M in im u m R e q u ir e d F u e l A s s e m b ly B u r n u p (B u ) a s a F u n c tio n o f E n r ic h m e n t (E n ) a n d C o o lin g T im e (C t)

See notes 1-4 for use of Table 5.5-2 F u e l C a te g o ry C o e ff.

I-3 I-4 II-1 II-2 II-3 II-4 II-5 A 1 2.04088171 -27.6637884 -11.2686777 20.7284208 29.8862876 -83.5409405 35.5058622 A 2 -4.83684164 26.1997193 2.0659501 11.9673275 -37.0771132 94.7973724 -30.1986997 A 3 2.59801889 -7.2982252 2.66204924 -14.4072388 16.3986049 -31.9583373 11.0102438 A 4 -0.300597247 0.723731768 -0.513334362 2.83623963 -2.1571669 3.55898487 -1.27269125 A 5 -0.610041808 0.401332891 -0.0987986108 -1.49118695 1.02330848 0.299948492 1.34723758 A 6 0.640497159 -0.418616707 -0.0724198633 1.75361041 -1.21889631 -0.312341996 -1.19871392 A 7 -0.219000712 0.144304039 0.106248806 -0.659046438 0.467440882 0.107463895 0.352920811 A 8 0.0252870451 -0.0154239536 -0.0197359109 0.080884618 -0.0560129443 -0.0108814287 -0.0325155213 A 9 -4.48207836 -5.54507376 -1.34620551 -245.825283 12.1549 39.4975573 -5.2576 A 1 0 -2.12118634 -5.76555416 -10.1728821 243.59979 -22.7755385 -50.5818253 10.1733379 A 1 1 2.91619317 6.29118025 8.71968815 -75.7805818 14.3755458 23.3093829 0.369083041 5-5 A 1 2 -0.196645176 -0.732079719 -1.14461356 8.10936356 -1.80803352 -2.69466612 0.0443577624 Notes:

1. All relevant uncertainties are explicitly included in the criticality analysis. For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required. For a fuel assembly to meet the requirements of a Fuel Category, the assembly burnup must exceed the minimum burnup (GWd/MTU) given by the curve fit for the assembly cooling time and initial enrichment. The specific minimum burnup required for each fuel assembly is calculated from the following equation:

AMENDMENT NOS. 260 AND 255 B u = (A 1 + A 2 *E n + A 3 *E n 2

+ A 4 *E n 3 )

e x p [ - ( A 5 + A 6 *E n + A 7 *E n 2

+ A 8 *E n 3 ) *C t ] + A 9 + A 1 0 *E n + A 1 1 *E n 2

+ A 1 2 *E n 3

2. Initial enrichment, En, is the nominal U-235 enrichment. Any enrichment between 1.8 and 4.0 may be used.

3. Cooling time, Ct, is in years. Any cooling time between 15 years and 25 years may be used. An assembly with a cooling time greater than 25 years must use 25 years.

4. This Table applies only for pre-EPU non-blank eted fuel assemblies. If a non-blank eted assembly is depleted at EPU conditions, none of the burnup accrued at EPU conditions can be credited (i.e., only burnup accrued at pre-EPU conditions may be used as burnup credit).

A01 ITS 3.7.14 A02 T a b le 5 .5 -3 F u e l C a te g o r ie s R a n k e d b y R e a c tiv ity See notes 1-5 for use of Table 5.5-3 I-1 H ig h R e a c tiv ity I-2 R e g io n I I-3 I-4 L o w R e a c tiv ity II-1 H ig h R e a c tiv ity II-2 R e g io n II II-3 II-4 II-5 L o w R e a c tiv ity Notes:

1. Fuel Category is rank ed by decreasing order of reactivity without regard for any reactivity-reducing mechanisms, e.g., Category I-2 is less reactive than Category I-1, etc. The more reactive fuel categories require compensatory measures to be placed in Regions I and II of the SFP, e.g., use of water filled cells, Metamic inserts, or full length RCCAs.

2. Any higher numbered fuel category can be used in place of a lower numbered fuel category from the same Region.

3. Category I-1 is fresh unburned fuel up to 5.0 wt% U-235 enrichment.

4. Category I-2 is fresh unburned fuel that obeys the IFBA requirements of Table 5.5-4.

5. All Categories except I-1 and I-2 are determined from Tables 5.5-1 and 5.5-2.

A02 T a b le 5 .5 -4 IF B A R e q u ir e m e n ts fo r F u e l C a te g o r y I-2 N o m in a l E n r ic h m e n t M in im u m R e q u ir e d N u m b e r (w t% U -2 3 5 ) o f IF B A P in s Enr. 4.3 0 4.3 < Enr. 4.4 3 2 4.4 < Enr. 4.7 6 4 4.7 < Enr. 5.0 8 0 TURKEY POINT - UNITS 3 & 4 5-6 AMENDMENT NOS. 260 AND 255

A01 ITS 3.7.14 A02 F IG U R E 5 .5 -1 A L L O W A B L E R E G IO N I S T O R A G E A R R A Y S See notes 1-8 for use of Figure 5.5-1 D E F IN IT IO N IL L U S T R A T IO N A r r a y I-A Check erboard pattern of Category I-1 assemblies I-1 X and empty (water-filled) cells.

X I-1 A r r a y I-B Category I-4 assembly in every cell. I-4 I-4 I-4 I-4 A r r a y I-C Combination of Category I-2 and I-4 assemblies. Each Category I-2 assembly shall contain a full length RCCA. I-2 I-4 I-2 I-2 I-4 I-4 I-4 I-4 I-2 I-2 I-2 I-2 I-2 I-4 I-2 I-2 A r r a y I-D Category I-3 assembly in every cell. One of every four I-3 I-3 assemblies contains a full length RCCA.

I-3 I-3 Notes:

1. In all arrays, an assembly of lower reactivity can replace an assembly of higher reactivity.

2. Category I-1 is fresh unburned fuel up to 5.0 wt% U-235 enrichment.

3. Category I-2 is fresh unburned fuel that obeys the IFBA requirements in Table 5.5-4.

4. Categories I-3 and I-4 are determined from Tables 5.5-1 and 5.5-2.

5. Shaded cells indicate that the fuel assembly contains a full length RCCA.

6. X indicates an empty (water-filled) cell.

7. Attributes for each 2x2 array are as stated in the definition. Diagram is for illustrative purposes only.

8. An empty (water-filled) cell may be substituted for any fuel containing cell in all storage arrays.

TURKEY POINT - UNITS 3 & 4 5-7 AMENDMENT NOS. 260 AND 255

A01 ITS 3.7.14 A02 F IG U R E 5 .5 -2 A L L O W A B L E R E G IO N II S T O R A G E A R R A Y S See notes 1-6 for use of Figure 5.5-2 D E F IN IT IO N IL L U S T R A T IO N A r r a y II-A Category II-1 assembly in three of every four cells; II-1 II-1 X II-1 one of every four cells is empty (water-filled); X II-1 II-1 II-1 the cell diagonal from the empty cell contains a Metamic insert or full length RCCA.

A r r a y II-B II-3 II-5 II-3 II-5 Check erboard pattern of Category II-3 and II-5 assemblies with two of every four cells containing a Metamic insert or full length II-5 II-3 II-5 II-3 RCCA.

A r r a y II-C II-4 II-4 II-4 II-4 Category II-4 assembly in every cell with two of every four cells containing a Metamic insert or full length RCCA. II-4 II-4 II-4 II-4 A r r a y II-D II-2 II-2 Category II-2 assembly in every cell with three of every four cells containing a Metamic insert or full length RCCA. II-2 II-2 Notes:

1. In all arrays, an assembly of lower reactivity can replace an assembly of higher reactivity.

2. Fuel categories are determined from Tables 5.5-1 and 5.5-2.

3. Shaded cells indicate that the cell contains a Metamic insert or the fuel assembly contains a full length RCCA.

4. X indicates an empty (water-filled) cell.

5. Attributes for each 2x2 array are as stated in the definition. Diagram is for illustrative purposes only.

6. An empty (water-filled) cell may be substituted for any fuel containing cell in all storage arrays.

TURKEY POINT - UNITS 3 & 4 5-8 AMENDMENT NOS. 260 AND 255

A01 ITS 3.7.14 A02 F IG U R E 5 .5 -3 IN T E R F A C E R E S T R IC T IO N S B E T W E E N R E G IO N I A N D R E G IO N II A R R A Y S See notes 1-8 for use of Figure 5.5-3 D E F IN IT IO N IL L U S T R A T IO N R e g io n I R a c k I-4 I-4 I-4 I-4 Array II-A, as defined in Figure 5.5-2, I-4 I-4 I-4 I-4 when placed on the interface with Region I shall have the empty cell in II-1 X II-1 X the row adj acent to the Region I rack . II-1 II-1 II-1 II-1 A r r a y II-A R e g io n I R a c k R e g io n I R a c k R e g io n I R a c k I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 Arrays II-B, II-C and II-D, as defined in I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 Figure 5.5-2, when placed on the interface with Region I shall have an II-3 II-5 II-3 II-5 II-4 II-4 II-4 II-4 II-2 II-2 II-2 II-2 insert in every cell in the row adj acent II-5 II-3 II-5 II-3 II-4 II-4 II-4 II-4 II-2 II-2 II-2 II-2 to the Region I rack . A r r a y II-B A r r a y II-C A r r a y II-D Notes:

1. In all arrays, an assembly of lower reactivity can replace an assembly of higher reactivity.

2. Fuel categories are determined from Tables 5.5-1 and 5.5-2.

3. Shaded cells indicate that the cell contains a Metamic insert or the fuel assembly contains a full length RCCA.

4. X indicates an empty (water-filled) cell.

5. Attributes for each 2x2 array are as stated in the definition. Diagram is for illustrative purposes only.

Region I Array I-B is depicted as the example; however, any Region I array is allowed provided that

a. For Array I-D, the RCCA shall be in the row adj acent to the Region II rack , and

b. Array I-A shall not interface with Array II-D.

6. If no fuel is stored adj acent to Region II in Region I, then the interface restrictions are not applicable.

7. Figure 5.5-3 is applicable only to the Region I - Region II interface. There are no restrictions for the interfaces with the cask area rack .

8. An empty (water-filled) cell may be substituted for any fuel containing cell in all storage arrays.

TURKEY POINT - UNITS 3 & 4 5-9 AMENDMENT NOS. 260 AND 255

DISCUSSION OF CHANGES ITS 3.7.14, SPENT FUEL STORAGE ADMINISTRATIVE CHANGES A01 In the conversion of the Turkey Point Nuclear Generating Station (PTN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 5.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 3.9.14 LCO states "The combination of initial enrichment, burnup, and l R2 cooling time of each fuel assembly stored in the Spent Fuel Pit shall be in accordance with Specification 5.5.1." ITS 3.7.14 Limiting Condition for Operation (LCO) states something similar; however, instead of referencing the sections from the Design Features Section (ITS 4.0), the applicable Section is being moved to ITS 3.7.14, including a portion that is added to the LCO, and the associated Tables and Figures. This changes the CTS by moving the items referenced in the Design Features Section of CTS to ITS Section 3.7.14.

The purpose of the CTS is to ensure there is proper loading of fuel in the fuel storage racks. Whether the requirements for doing so are in the Design Features or in the Plant Systems section of Technical Specifications will not impact accomplishing this goal. Thus, moving the requirements from the one section of Technical Specifications to another will continue to ensure the fuel loading requirements are met and that appropriate action is taken if not. These changes are designated as administrative because the changes do not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES M01 ITS Surveillance Requirement (SR) 3.7.14.1 verifies by administrative means the fuel assemblies stored in Regions I and II are stored in accordance with the requirements of Figures 3.7.14-1 through Figure 3.7.14-3 and Tables 3.7.14-1 through 3.7.14.3 with credited for burnup and cooling time taken in determining acceptable placement locations for spent fuel in the two-region spent fuel racks.

The CTS do not contain this SR. This changes the CTS by adding a new SR to Technical Specifications.

The purpose of the ITS SR is to verify by administrative means that the fuel in the fuel storage pool is in accordance with the requirements contained in the figures and tables in the Technical Specifications. While the CTS did not explicitly contain a requirement to perform this verification, these requirements were verified per the figure and tables that were located in the Design Features section of Technical Specifications. This change is designated as more restrictive because an SR is being added to the ITS that was not in the CTS.

Turkey Point Unit 3 and Unit 4 Page 1 of 2

DISCUSSION OF CHANGES ITS 3.7.14, SPENT FUEL STORAGE RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 4 - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, QAP, CLRT Program, IST Program, ISI Program, or Surveillance Frequency Control Program) CTS SR 4.9.14.2 states, "A representative sample of inservice metamic inserts shall be visually inspected in accordance with the Metamic Surveillance Program described in UFSAR Section 16.2. The surveillance program ensures that the performance requirements of Metamic are met over the surveillance interval." ITS 3.7.14 does not contain this SR. This changes the CTS by moving a SR out of Technical l R2 Specifications and into the Technical Requirements Manual (TRM).

The purpose of this SR is to ensure the performance requirements of the Metamic are met in accordance with the Metamic Surveillance Program located in the Updated Safety Analysis Report (UFSAR). The CTS LCO does not discuss the Metamic inserts. The ITS is proposing to move this SR to the TRM.

This change is acceptable because neither the LCO nor the Actions discuss the Metamic inserts. The Metamic inserts will still be required to be inspected if moved to the TRM in accordance with the UFSAR program and can be adequately controlled via the TRM. Any changes to the TRM are made under 10 CFR 50.59, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail because an SR is being moved from the Technical Specifications to the TRM.

LESS RESTRICTIVE CHANGES None Turkey Point Unit 3 and Unit 4 Page 2 of 2

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

CTS [ Spent Fuel Pool Storage] 3 3.7.17 2 14 3.7 PLANT SYSTEMS 14 2 3 3.7.17 [ Spent Fuel Pool Storage ]

, 1 14 , and cooling time 2

3.9.14 LCO 3.7.17 The combination of initial enrichment and burnup of each fuel assembly the spent fuel pit stored in [ Region 2] shall be within the Acceptable [ Burnup Domain] of Figure 3.7.17-1 or in accordance with Specification 4.3.1.1. 2 4 the following:

INSERT 1 Applicability APPLICABILITY: Whenever any fuel assembly is stored in [ Region 2] of the spent fuel 1 storage pool.

3 pit ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME ACTION b, c A. Requirements of the A.1 --------------NOTE--------------

LCO not met. LCO 3.0.3 is not applicable.

Initiate action to move the Immediately noncomplying fuel to an acceptable location assembly from [ Region 2] . 1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 14 SR 3.7.17.1 Verify by administrative means the initial enrichment Prior to storing the 2

, burnup, and cooling time and burnup of the fuel assembly is in accordance fuel assembly in with Figure 3.7.17-1 or Specification 4.3.1.1. [ Region 2] 3 2 3 Figure 3.7.14-1 through Figure 3.7.14-3 Region I or II 1 Turk ey Point Unit 3 and Unit 4 14 Amendment Nos. XXX and YYY Westinghouse STS 3.7.17-1 Rev. 5.0 3 2

ITS 3.7.14 4

INSERT 1 CTS 5.5.1.1.e a. No restrictions on storage of fresh or irradiated fuel assemblies in the cask area storage rack are applicable.

5.5.1.3 b. Fuel assemblies stored in Region I and II shall be stored in accordance with the DOC A02 requirements of Figures 3.7.14-1 through 3.7.14-3 with credit for burnup and cooling l R2 time taken in determining acceptable placement locations for spent fuel in the two-region spent fuel racks. Fresh and irradiated fuel assemblies in the Region I or Region II racks shall be stored in compliance with the following:

1. any 2x2 array of Region I storage cells containing fuel shall comply with the storage patterns in Figure 3.7.14-1 and the requirements of Tables 3.7.14-1 and 3.7.14-2, as applicable. The reactivity rank of fuel assemblies in the 2x2 array (rank determined using Table 3.7.14-3) shall be equal to or less reactive than that shown for the 2x2 array.

2. any 2x2 array of Region II storage cells containing fuel shall:

i. comply with the storage patterns in Figure 3.7.14-2 and the requirements of Tables 3.7.14-1 and 3.7.14-2, as applicable. The reactivity rank of fuel assemblies in the 2x2 array (rank determined using Table 3.7.14-3) shall be equal to or less reactive than that shown for the 2x2 array, ii. have the same directional orientation for Metamic inserts in a contiguous group of 2x2 arrays where Metamic inserts are required, and iii. comply with the requirements of LCO 3.7.14.b.3. for cells adjacent to Region I racks.

3. Any 2x2 array of Region II storage cells that interface with Region I storage cells shall comply with the rules of Figure 3.7.14-3.

4. Any fuel assembly may be replaced with a fuel rod storage basket or non-fuel hardware.

5. Storage of Metamic inserts or rod cluster control assemblies (RCCAs) is acceptable in locations designated as empty (water-filled) cells.

Insert Page 3.7.14-1

CTS [ Spent Fuel Pool Storage] 3 3.7.17 2 14 INSERT 2 4 Turk ey Point Unit 3 and Unit 4 14 Amendment Nos. XXX and YYY Westinghouse STS 3.7.17-2 Rev. 5.0 3 2

ITS 3.7.14 4

INSERT 2 CTS Table 3.7.14-1 Blanketed Fuel - Coefficients to Calculate the Minimum Required Fuel Assembly Burnup (Bu) as a Function of Enrichment (En) and Cooling Time (Ct)

Table 5.5-1 See notes 1-4 for use of Table 3.7.14-1 DOC A02 Fuel Category Coeff.

I-3 I-4 II-1 II-2 II-3 II-4 II-5 A1 5.66439153 -14.7363682 -7.74060457 -7.63345029 24.4656526 8.5452608 26.2860949 A2 -7.22610116 11.0284547 5.13978237 10.7798957 -20.3141124 -4.47257395 -18.0738662 A3 2.98646188 -1.80672781 -0.360186309 -2.81231555 6.53101471 2.09078914 5.8330891 A4 -0.287945644 0.119516492 0.0021681285 0.29284474 -0.581826027 -0.188280562 -0.517434342 A5 -0.558098618 0.0620559676 -0.0304713673 0.0795058096 -0.16567492 0.157548739 -0.0614152031 A6 0.476169245 0.0236575787 0.098844889 -0.0676341983 0.243843226 -0.0593584027 0.134626308 A7 -0.117591963 -0.0088144551 -0.0277584786 0.0335130877 -0.0712130368 0.0154678626 -0.0383060399 A8 0.0095165354 0.0008957348 0.0024057185 -0.0040803875 0.0063998706 -0.0014068318 0.0033419846 A9 -47.1782783 -20.2890089 -21.424984 14.6716317 -41.1150 -0.881964768 -12.1780 A10 33.4270029 14.7485847 16.255208 -10.0312224 43.9149156 9.69128392 23.6179517 A11 -6.11257501 -1.22889103 -1.77941882 5.62580894 -9.6599923 -0.18740168 -4.10815592 A12 0.490064351 0.0807808548 0.127321203 -0.539361868 0.836931842 0.0123398618 0.363908736 Notes:

1. All relevant uncertainties are explicitly included in the criticality analysis. For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required. For a fuel assembly to meet the requirements of a Fuel Category, the assembly burnup must exceed the "minimum burnup" (GWd/MTU) given by the curve fit for the assembly "cooling time" and "initial enrichment." The specific minimum burnup required for each fuel assembly is calculated from the following equation:

Bu = (A1 + A2*En + A3*En2 + A4*En3)* exp [ - (A5 + A6*En + A7*En2 + A8*En3)*Ct ] + A9 + A10*En + A11*En2 + A12*En3

2. Initial enrichment, En, is the nominal central zone U-235 enrichment. Axial blanket material is not considered when determining enrichment.

Any enrichment between 2.0 and 5.0 may be used.

3. Cooling time, Ct, is in years. Any cooling time between 0 years and 25 years may be used. An assembly with a cooling time greater than 25 years must use 25 years.

4. This Table applies for any blanketed fuel assembly.

Insert Page 3.7.14-2a

4 ITS 3.7.14 CTS INSERT 2 (Continued)

Table 3.7.14-2 Non-Blanketed Fuel - Coefficients to Calculate the Minimum Required Fuel Assembly Burnup (Bu) as a Function of Table 5.5-2 Enrichment (En) and Cooling Time (Ct)

DOC A02 See notes 1-4 for use of Table 3.7.14-2 Fuel Category Coeff.

I-3 I-4 II-1 II-2 II-3 II-4 II-5 A1 2.04088171 -27.6637884 -11.2686777 20.7284208 29.8862876 -83.5409405 35.5058622 A2 -4.83684164 26.1997193 2.0659501 11.9673275 -37.0771132 94.7973724 -30.1986997 A3 2.59801889 -7.2982252 2.66204924 -14.4072388 16.3986049 -31.9583373 11.0102438 A4 -0.300597247 0.723731768 -0.513334362 2.83623963 -2.1571669 3.55898487 -1.27269125 A5 -0.610041808 0.401332891 -0.0987986108 -1.49118695 1.02330848 0.299948492 1.34723758 A6 0.640497159 -0.418616707 -0.0724198633 1.75361041 -1.21889631 -0.312341996 -1.19871392 A7 -0.219000712 0.144304039 0.106248806 -0.659046438 0.467440882 0.107463895 0.352920811 A8 0.0252870451 -0.0154239536 -0.0197359109 0.080884618 -0.0560129443 -0.0108814287 -0.0325155213 A9 -4.48207836 -5.54507376 -1.34620551 -245.825283 12.1549 39.4975573 -5.2576 A10 -2.12118634 -5.76555416 -10.1728821 243.59979 -22.7755385 -50.5818253 10.1733379 A11 2.91619317 6.29118025 8.71968815 -75.7805818 14.3755458 23.3093829 0.369083041 A12 -0.196645176 -0.732079719 -1.14461356 8.10936356 -1.80803352 -2.69466612 0.0443577624 Notes:

1. All relevant uncertainties are explicitly included in the criticality analysis. For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required. For a fuel assembly to meet the requirements of a Fuel Category, the assembly burnup must exceed the "minimum burnup" (GWd/MTU) given by the curve fit for the assembly "cooling time" and "initial enrichment." The specific minimum burnup required for each fuel assembly is calculated from the following equation:

Bu = (A1 + A2*En + A3*En2 + A4*En3)* exp [ - (A5 + A6*En + A7*En2 + A8*En3)*Ct ] + A9 + A10*En + A11*En2 + A12*En3

2. Initial enrichment, En, is the nominal U-235 enrichment. Any enrichment between 1.8 and 4.0 may be used.

3. Cooling time, Ct, is in years. Any cooling time between 15 years and 25 years may be used. An assembly with a cooling time greater than 25 years must use 25 years.

4. This Table applies only for pre-extended power uprate (EPU) non-blanketed fuel assemblies. If a non-blanketed assembly is depleted at EPU conditions, none of the burnup accrued at EPU conditions can be credited (i.e., only burnup accrued at pre-EPU conditions may be used as burnup credit).

Insert Page 3.7.14-2b

ITS 3.7.14 4

INSERT 2 (Continued)

CTS Table 3.7.14-3 Table 5.5-3 Fuel Categories Ranked by Reactivity DOC A02 See notes 1-5 for use of Table 3.7.14-3 I-1 High Reactivity I-2 Region I I-3 I-4 Low Reactivity II-1 High Reactivity II-2 Region II II-3 II-4 II-5 Low Reactivity Notes:

1. Fuel Category is ranked by decreasing order of reactivity without regard for any reactivity-reducing mechanisms, e.g., Category I-2 is less reactive than Category I-1, etc. The more reactive fuel categories require compensatory measures to be placed in Regions I and II of the spent fuel pit, e.g., use of water filled cells, Metamic inserts, or full length RCCAs.

2. Any higher numbered fuel category can be used in place of a lower numbered fuel category from the same Region.

3. Category I-1 is fresh unburned fuel up to 5.0 wt% U-235 enrichment.

4. Category I-2 is fresh unburned fuel that obeys the Integral Fuel Burnable Absorber (IFBA) requirements of Table 3.7.14-4.

5. All Categories except I-1 and I-2 are determined from Tables 3.7.14-1 and 3.7.14-2.

Table 3.7.14-4 Table 5.5-4 DOC A02 IFBA Requirements for Fuel Category I-2 Nominal Enrichment Minimum Required Number (wt% U-235) of IFBA Pins Enr. 4.3 0 4.3 < Enr. 4.4 32 4.4 < Enr. 4.7 64 4.7 < Enr. 5.0 80 Insert Page 3.7.14-2c

ITS 3.7.14 4

INSERT 2 (Continued)

CTS FIGURE 3.7.14-1 ALLOWABLE REGION I STORAGE ARRAYS Figure 5.5-1 See notes 1-8 for use of Figure 3.7.14-1 DOC A02 DEFINITION ILLUSTRATION Array I-A Checkerboard pattern of Category I-1 assemblies I-1 X and empty (water-filled) cells.

X I-1 Array I-B Category I-4 assembly in every cell. I-4 I-4 I-4 I-4 Array I-C Combination of Category I-2 and I-4 assemblies. Each I-2 I-4 I-2 I-2 Category I-2 assembly shall contain a full length RCCA.

I-4 I-4 I-4 I-4 I-2 I-2 I-2 I-2 I-2 I-4 I-2 I-2 Array I-D Category I-3 assembly in every cell. One of every four I-3 I-3 assemblies contains a full length RCCA. I-3 I-3 Notes:

1. In all arrays, an assembly of lower reactivity can replace an assembly of higher reactivity.

2. Category I-1 is fresh unburned fuel up to 5.0 wt% U-235 enrichment.

3. Category I-2 is fresh unburned fuel that obeys the IFBA requirements in Table 3.7.14-4.

4. Categories I-3 and I-4 are determined from Tables 3.7.14-1 and 3.7.14-2.

5. Shaded cells indicate that the fuel assembly contains a full length RCCA.

6. X indicates an empty (water-filled) cell.

7. Attributes for each 2x2 array are as stated in the definition. Diagram is for illustrative purposes only.

8. An empty (water-filled) cell may be substituted for any fuel containing cell in all storage arrays.

Insert Page 3.7.14-2d

ITS 3.7.14 4

CTS INSERT 2 (Continued)

FIGURE 3.7.14-2 Figure 5.5-2 DOC A02 ALLOWABLE REGION II STORAGE ARRAYS See notes 1-6 for use of Figure 3.7.14-2 DEFINITION ILLUSTRATION Array II-A Category II-1 assembly in three of every four cells; II-1 II-1 X II-1 one of every four cells is empty (water-filled); X II-1 II-1 II-1 the cell diagonal from the empty cell contains a Metamic insert or full length RCCA.

Array II-B II-3 II-5 II-3 II-5 Checkerboard pattern of Category II-3 and II-5 assemblies with two of every four cells containing a Metamic insert or full II-5 II-3 II-5 II-3 length RCCA.

Array II-C II-4 II-4 II-4 II-4 Category II-4 assembly in every cell with two of every four cells containing a Metamic insert or full length RCCA. II-4 II-4 II-4 II-4 Array II-D II-2 II-2 Category II-2 assembly in every cell with three of every four cells containing a Metamic insert or full length RCCA. II-2 II-2 Notes:

1. In all arrays, an assembly of lower reactivity can replace an assembly of higher reactivity.

2. Fuel categories are determined from Tables 3.7.14-1 and 3.7.14-2.

3. Shaded cells indicate that the cell contains a Metamic insert or the fuel assembly contains a full length RCCA.

4. X indicates an empty (water-filled) cell.

5. Attributes for each 2x2 array are as stated in the definition. Diagram is for illustrative purposes only.

6. An empty (water-filled) cell may be substituted for any fuel containing cell in all storage arrays.

Insert Page 3.7.14-2e

ITS 3.7.14 4

IN S E R T 2 (C o n tin u e d )

CTS F IG U R E 3 .7 .1 4 -3 IN T E R F A C E R E S T R IC T IO N S B E T W E E N R E G IO N I A N D R E G IO N II A R R A Y S Figure 5.5-3 See notes 1-8 for use of Figure 3.7.14-3 DOC A02 D E F IN IT IO N IL L U S T R A T IO N R e g io n I R a c k I-4 I-4 I-4 I-4 Array II-A, as defined in Figure 3.7.14-2, when placed on the interface with I-4 I-4 I-4 I-4 2 Region I shall have the empty cell in II-1 X II-1 X the row adj acent to the Region I rack . II-1 II-1 II-1 II-1 A r r a y II-A R e g io n I R a c k R e g io n I R a c k R e g io n I R a c k I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 Arrays II-B, II-C and II-D, as defined in I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 Figure 3.7.14-2, when placed on the interface with Region I shall have an II-3 II-5 II-3 II-5 II-4 II-4 II-4 II-4 II-2 II-2 II-2 II-2 insert in every cell in the row adj acent II-5 II-3 II-5 II-3 II-4 II-4 II-4 II-4 II-2 II-2 II-2 II-2 to the Region I rack . A r r a y II-B A r r a y II-C A r r a y II-D Notes:

1. In all arrays, an assembly of lower reactivity can replace an assembly of higher reactivity.

2. Fuel categories are determined from Tables 3.7.14-1 and 3.7.14-2.

3. Shaded cells indicate that the cell contains a Metamic insert or the fuel assembly contains a full length RCCA.

4. X indicates an empty (water-filled) cell.

5. Attributes for each 2x2 array are as stated in the definition. Diagram is for illustrative purposes only.

Region I Array I-B is depicted as the example; however, any Region I array is allowed provided that

a. For Array I-D, the RCCA shall be in the row adj acent to the Region II rack , and

b. Array I-A shall not interface with Array II-D.

6. If no fuel is stored adj acent to Region II in Region I, then the interface restrictions are not applicable.

7. Figure 3.7.14-3 is applicable only to the Region I - Region II interface. There are no restrictions for the interfaces with the cask area rack .

8. An empty (water-filled) cell may be substituted for any fuel containing cell in all storage arrays.

Insert Page 3.7.14-2f

JUSTIFICATION FOR DEVIATIONS ITS 3.7.14, SPENT FUEL STORAGE

1. The Improved Standard Technical Specification (ISTS) contain bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

2. ISTS 3.7.17, "Spent Fuel Pool Storage" has been renumbered as Improved Technical Specifications (ITS) 3.7.14, "Spent Fuel Storage."

3. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

4. The ITS incorporates the spent fuel storage section of the Turkey Point Nuclear Generating Station Current Technical Specifications (CTS) Design Features Chapter (CTS 5.0) which consists of requirements, and tables and figures. The fuel storage section requirements are incorporated in the Limiting Condition for Operation (LCO).

The LCO also references the tables and figures which are being added to the end of the Technical Specifications.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

[Spent Fuel Pool Storage] 1 B 3.7.17 2 14 B 3.7 PLANT SYSTEMS 14 2 1 B 3.7.17 [ Spent Fuel Pool Storage ]

BASES BACKGROUND In the Maximum Density Rack (MDR) [(Refs. 1 and 2)] design, the spent fuel storage pool is divided into two separate and distinct regions which, for the purpose of criticality considerations, are considered as separate pools. [Region 1], with [336] storage positions, is designed to accommodate new fuel with a maximum enrichment of [4.65] wt% U-235, 3

or spent fuel regardless of the discharge fuel burnup. [Region 2], with

[2670] storage positions, is designed to accommodate fuel of various initial enrichments which have accumulated minimum burnups within the acceptable domain according to Figure 3.7.17-1, in the accompanying LCO. Fuel assemblies not meeting the criteria of Figure [3.7.17-1] shall be stored in accordance with paragraph 4.3.1.1 in Section 4.3, Fuel Storage.

INSERT 1 3 pit The water in the spent fuel storage pool normally contains soluble boron, l R2 which results in large subcriticality margins under actual operating conditions. However, the NRC guidelines, based upon the accident condition in which all soluble poison is assumed to have been lost, 1.0 specify that the limiting keff of 0.95 be evaluated in the absence of soluble 3 boron. Hence, the design of both regions is based on the use of unborated water, which maintains each region in a subcritical condition during normal operation with the regions fully loaded. The double /ANS-8.1-1983 contingency principle discussed in ANSI N-16.1-1975 and the April 1978 3 1

NRC letter (Ref. 3) allows credit for soluble boron under other abnormal or accident conditions, since only a single accident need be considered at one time. For example, the most severe accident scenario is associated with the movement of fuel from [Region 1 to Region 2], and accidental misloading of a fuel assembly in [Region 2]. This could potentially 3

increase the criticality of [Region 2]. To mitigate these postulated criticality related accidents, boron is dissolved in the pool water. Safe operation of the MDR with no movement of assemblies may therefore be achieved by controlling the location of each assembly in accordance with the accompanying LCO. Prior to movement of an assembly, it is necessary to perform SR 3.7.16.1. 3 13 Turkey Point Unit 3 and Unit 4 14 Revision XXX Westinghouse STS B 3.7.17-1 Rev. 5.0

ITS 3.7.14 3

INSERT 1 The spent fuel pit storage racks provide safe subcritical storage of fuel assemblies by providing sufficient l R2 center-to-center spacing or a combination of spacing and poison to assure: a) Keff less than or equal to 0.95 with a minimum soluble boron concentration of 500 ppm present, and b) Keff less than 1.0 when flooded with unborated water for normal operations and postulated accidents. The 500 ppm value is needed to assure keff less than 0.95 for normal operating conditions. The criticality analysis needs 1700 ppm to assure keff less than 0.95 under the worst case accident condition. There is significant margin between the calculated ppm requirement and the spent fuel boron concentration requirement of 2300 ppm. The higher boron concentration value is chosen because, during refueling, the water volume in the spent fuel pit, the transfer l R2 canal, the refueling canal, the refueling cavity, and the reactor vessel form a single mass.

The spent fuel pit storage racks are divided into two regions, Region I and Region II. The Region I l R2 permanent racks have a 10.6 inch center-to-center spacing. The Region lI racks have a 9.0 inch center-to-center spacing. The cask area storage rack has a nominal 10.1 inch center to center spacing in the east-west direction and a nominal 10.7 inch center-to-center spacing in the north-south direction.

Any fuel for use at Turkey Point, and enriched to less than or equal to 5.0 wt % U-235, may be stored in the cask area storage rack. Fresh or irradiated fuel assemblies not stored in the cask area storage rack shall be l R2 stored in accordance with ITS 3.7.14.

Fresh unirradiated fuel may be placed in the permanent Region I racks in accordance with the restrictions of Figure 3.7.14-1. Fresh unirradiated fuel may be placed in the permanent Region II racks in accordance with the restrictions of Figure 3.7.14-2. Prior to placement of irradiated fuel in Region I or II spent fuel pit storage l R2 rack cell locations, strict controls are employed to evaluate burnup of the fuel assembly. Upon determination that the fuel assembly meets the nominal burnup and associated post-irradiation cooling time requirements of Table 3.7.14-1 or Table 3.7.14-2, it may be placed in a Region I or II cell in accordance with the restrictions of Figures 3.7.14-1 through 3.7.14-3, respectively.

For all assemblies with blanketed fuel, the initial enrichment is based on the central zone enrichment (i.e.,

between the axial blankets) consistent with the assumptions of the analysis. These positive controls assure that the fuel enrichment limits, burnup, and post irradiation cooling time requirements assumed in the safety analyses will not be violated. l R2 Insert Page B 3.7.14-1

[Spent Fuel Pool Storage] 1 B 3.7.17 2 14 BASES APPLICABLE 2 The hypothetical accidents can only take place during or as a result of SAFETY pit the movement of an assembly (Ref. 4). For these accident occurrences, 3 ANALYSES 13 the presence of soluble boron in the spent fuel storage pool (controlled by l R2 LCO 3.7.16, "Fuel Storage Pool Boron Concentration") prevents criticality 3 in both regions. By closely controlling the movement of each assembly and by checking the location of each assembly after movement, the time period for potential accidents may be limited to a small fraction of the total operating time. During the remaining time period with no potential for accidents, the operation may be under the auspices of the accompanying LCO.

The configuration of fuel assemblies in the fuel storage pool satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

2 LCO The restrictions on the placement of fuel assemblies within the spent fuel pit pool, in accordance with Figure 3.7.17-1, in the accompanying LCO, 14 1.0 3 l R2 through 3.7.14-3 and Tables ensures the keff of the spent fuel storage pool will always remain < 0.95, pit l R2 3.7.14-1 through 3.7.14-4 assuming the pool to be flooded with unborated water. Fuel assemblies l not meeting the criteria of Figure [3.7.17-1] shall be stored in accordance 3 INSERT 2 with Specification 4.3.1.1 in Section 4.3.

Region I or Region II 1 APPLICABILITY This LCO applies whenever any fuel assembly is stored in [Region 2] of 3 pit spent the fuel storage pool. 3 l R2 ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.

pit When the configuration of fuel assemblies stored in [Region 2] the spent fuel storage pool is not in accordance with Figure 3.7.17-1, or l R2 Figure 3.7.14-1, paragraph 4.3.1.1, the immediate action is to initiate action to make the Figure 3.7.14-2, or Figure 3.7.14-3 necessary fuel assembly movement(s) to bring the configuration into compliance with Figure 3.7.17-1 or Specification 4.3.1.1.

If unable to move irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not be applicable. If unable to move irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the action is independent of reactor operation. Therefore, inability to move fuel assemblies is not l R2 sufficient reason to require a reactor shutdown.

Turkey Point Unit 3 and Unit 4 14 Revision XXX Westinghouse STS B 3.7.17-2 Rev. 5.0

ITS 3.7.14 3

INSERT 2 The LCO states that there are no restrictions on storage of fresh or irradiated fuel assemblies in l R2 the cask area storage rack. This is because in the cask area rack criticality is prevented by the design of the rack which limits fuel assembly interaction by fixing the separation distance between stored assemblies and/or by placing a neutron absorber panel between storage cells.

Insert Page B 3.7.14-2

[ Spent Fuel Pool Storage] 1 B 3.7.17 2 14 BASES fuel assemblies stored in Regions I and II are stored in accordance with the requirements of Figure 3.7.14-1 through Figure 3.7.14-3 with credit for burnup 14 and cooling time tak en in determining acceptable placement locations for spent SURVEILLANCE SR 3.7.17.1 fuel in the two-region spent fuel rack s.

REQUIREMENTS This SR verifies by administrative means that the initial enrichment and burnup of the fuel assembly is in accordance with Figure [ 3.7.17-1] in the 3 accompanying LCO. For fuel assemblies in the unacceptable range of Figure 3.7.17-1, performance of this SR will ensure compliance with Specification 4.3.1.1. not meeting requirements of Figures 3.7.14-1 through 3.7.14-3 REFERENCES [ 1. Callaway FSAR, Appendix 9.1A, " The Maximum Density Rack (MDR)

Design Concept."

2. Description and Evaluation for Proposed Changes to Facility Operating Licenses DPR-39 and DPR-48 (Zion Power Station). ]

/ANS-8.1-1983

3. Double contingency principle of ANSI N16.1-1975, as specified in the 3 1 April 14, 1978 NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).

2 14.2.1 3 1

4. FSAR, Section [ 15.7.4] .

U Turk ey Point Unit 3 and Unit 4 14 Revision XXX Westinghouse STS B 3.7.17-3 Rev. 5.0

JUSTIFICATION FOR DEVIATIONS ITS 3.7.14 BASES, SPENT FUEL STORAGE

1. The Improved Standard Technical Specification (ISTS) contain bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

2. ISTS 3.7.17, "Spent Fuel Pool Storage" has been renumbered as Improved Technical Specifications (ITS) 3.7.14, "Spent Fuel Storage."

3. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.7.14, SPENT FUEL STORAGE There are no specific No Significant Hazards Considerations for this Specification.

Turkey Point Unit 3 and 4 Page 1 of 1

ATTACHMENT 15 Relocated/Deleted Current Technical Specifications (CTS)

CTS 3.7.6 - Snubbers l R2 CTS 3.7.7 - Sealed Source Contamination l

CTS 3.7.6 SNUBBERS Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

PLANT SYSTEMS R01 3/4.7.6 SNUBBERS LIMITING CONDITION FOR OPERATION 3.7.6 All snubbers shall be OPERABLE. The only snubbers excluded from the requirements are those installed on nonsafety-related systems and then only if their failure or failure of the system on which they are installed would have no adverse effect on any safety-related system.

APPLICABILITY: MODES 1, 2, 3, and 4. MODES 5 and 6 for snubbers located on systems required OPERABLE in those MODES.

ACTION:

With one or more snubbers inoperable on any system, within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months replace or restore the inoperable snubber(s) to OPERABLE status and determine the impact on the attached component by evaluation in accordance with Specification 4.7.6, or declare the attached system inoperable and follow the appropriate ACTION statement for that system.

SURVEILLANCE REQUIREMENTS 4.7.6 Each snubber shall be demonstrated OPERABLE by performance of the Snubber Testing Program in Specification 6.8.4.m.

TURKEY POINT - UNITS 3 & 4 3/4 7-22 AMENDMENT NOS. 272 AND 267

DISCUSSION OF CHANGES CTS 3.7.6, SNUBBERS ADMINISTRATIVE CHANGES NONE MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS R01 In the conversion of the Turkey Point Nuclear Generating Station (PTN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), CTS 3.7.6 provides requirements for all safety-related snubbers. This specification, with the exception of the Action to restore an inoperable snubber within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , is not included in the ITS. This changes the CTS by moving the explicit snubber requirements from the Technical Specifications to the Technical Requirements Manual (TRM).

The removal of these details from the Technical Specification is acceptable because this type of information is not necessary to provide adequate protection of public health and safety. The purpose of the snubber requirements is to ensure that the structural integrity of the Reactor Coolant System and other safety related systems is maintained during and following a seismic or other event initiating dynamic loads.

This change is acceptable because the Limiting Condition for Operation (LCO) requirements continue to ensure that the structures, systems, and components are maintained consistent with the safety analyses and licensing basis.

Specifically, ITS LCO 3.0.8 continues to require, in part, action to assess and manage risk during the period one or more required snubbers are unable to perform the associated support function(s). ITS LCO 3.0.8 also places limits on the delay period when one or more required snubbers are unable to perform the associated support function(s) based on the affect the degraded snubber has on the support system (e.g., affects one train or multiple trains). Refer to Section 3.0 Discussion of Changes related to the addition of ITS LCO 3.0.8. The requirement to perform snubber inspections is specified in 10 CFR 50.55a and the requirement to perform snubber inspections and testing is specified in ASME Section XI, as modified by approved relief requests. Therefore, both PTN commitments and NRC regulations or generic guidance contain the necessary programmatic requirements for the inspection and testing of safety related snubbers without repeating them in the ITS. ASME code requirements associated with snubber inspections and testing will continue to be controlled pursuant 10 CFR 50.55a. This change is acceptable because the removed information will be adequately controlled in the TRM. Any changes to the TRM are made under 10 CFR 50.59, which ensures changes are properly evaluated.

This change is designated as a less restrictive removal of detail change because a requirement is being removed from the Technical Specifications.

Turkey Point Unit 3 and Unit 4 Page 1 of 2

DISCUSSION OF CHANGES CTS 3.7.6, SNUBBERS 10 CFR 50.36(c)(2)(ii) Criteria Evaluation:

1. Snubber limitations do not constitute an instrumentation system that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

2. Snubber limitations are not a process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or challenge to the integrity of a fission product barrier. This Technical Specification specifies limits on process variables consistent with the structural analysis results. These limits, however, do not reflect initial condition assumptions in the Design Basis Accident (DBA).

3. Snubber limitations are not a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

4. Snubber limitations were found to be non-significant risk contributor to core damage frequency and offsite releases. These indications are not structures, systems, or components that operating experience or probabilistic safety assessment has shown to be significant to the public health and safety.

Because 10 CFR 50.36(c)(2)(ii) criteria have not been satisfied, the Snubber Specification may be relocated to a licensee-controlled document outside the Technical Specifications.

REMOVED DETAIL CHANGES NONE LESS RESTRICTIVE CHANGES NONE Turkey Point Unit 3 and Unit 4 Page 2 of 2

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.7.6, SNUBBERS There are no specific No Significant Hazards Considerations for this Specification.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

CTS 3.7.7, SEALED SOURCE CONTAMINATION Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

PLANT SYSTEMS 3/4.7.7 SEALED SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION 3.7.7 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 microCurie of removable contamination.

APPLICABILITY: At all times.

ACTION:

a. With a sealed source having removable contamination in excess of the above limits, 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 R01 4.7.7.1 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 microCurie per test sample.

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

a. Sources in use - in accordance with the Surveillance Frequency Control Program for all sealed sources containing radioactive materials:

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

2) In any form other than gas.

TURKEY POINT - UNITS 3 & 4 3/4 7-23 AMENDMENT NOS. 272 AND 267

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. Stored sources not in use - Each sealed source and fission detector shall be tested prior to use or transfer to another licensee unless tested within the previous 6 months. Sealed sources and fission detectors transferred without a certificate indicating the last test date shall be tested prior to being placed into use; and

c. Startup sources and fission detectors - Each sealed startup source and fission detector shall be R01 tested within 31 days prior to being subjected to core flux or installed in the core and following repair or maintenance to the source.

4.7.7.3 Deleted 4.7.7.4 A complete inventory of licensed radioactive materials in possession shall be maintained current at all times.

TURKEY POINT - UNITS 3 & 4 3/4 7-24 AMENDMENT NOS. 279 AND 274

PLANT SYSTEMS 3/4 3.7.8 DELETED TURKEY POINT - UNITS 3 & 4 3/4 7-25 AMENDMENT NOS. 282 AND 276

PLANT SYSTEMS 3/4 3.7.9 DELETED TURKEY POINT - UNITS 3 & 4 3/4 7-26 AMENDMENT NOS. 282 AND 276

DISCUSSION OF CHANGES CTS 3.7.7, SEALED SOURCE CONTAMINATION ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS R01 CTS 3.7.7 provides the requirements that 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 microCurie of removable contamination.

The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(a)(3) limits for plutonium. This limitation will ensure that leakage from Byproduct, Source, and Special Nuclear Material sources will not exceed allowable intake values.

This requirement and the associated surveillance requirements bear no relation to the conditions or limitations that are necessary to ensure safe reactor operation.

10 CFR 50.36(c)(2)(ii) Criteria Evaluation:

1. Sealed Source Contamination limitations do not constitute an instrumentation system that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

2. Sealed Source Contamination limitations are not a process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or challenge to the integrity of a fission product barrier. This Technical Specification specifies limits on process variables consistent with the structural analysis results. These limits, however, do not reflect initial condition assumptions in the DBA.

3. Sealed Source Contamination limitations are not a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

4. Sealed Source Contamination limitations were found to be non-significant risk contributor to core damage frequency and offsite releases. These indications are not structures, systems, or components that operating experience or Turkey Point Unit 3 and Unit 4 Page 1 of 2

DISCUSSION OF CHANGES CTS 3.7.7, SEALED SOURCE CONTAMINATION probabilistic safety assessment has shown to be significant to the public health and safety.

Because 10 CFR 50.36(c)(2)(ii) criteria have not been satisfied, the Sealed Source Contamination Specification may be relocated to a licensee-controlled document outside the Technical Specifications. Requirements associated with the sealed sources are governed by 10 CFR Part 70. Compliance with applicable portions of 10 CFR Part 70 is required by the operating licenses of PTN Units 3 and 4. Changes to the TRM will be controlled by the provisions of 10 CFR 50.59. This change is designated as relocation because the Specification does not meet the criteria in 10 CFR 50.36(c)(2)(ii) and will be relocated to the TRM.

REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Turkey Point Unit 3 and Unit 4 Page 2 of 2

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.7.7, SEALED SOURCE CONTAMINATION There are no specific No Significant Hazards Considerations for this Specification.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

ATTACHMENT 16 Improved Standard Technical Specifications (ISTS)

Not Adopted in the Turkey Point ITS ISTS 3.7.4 - Atmospheric Dump Valves (ADVs) l R2 l

ISTS 3.7.12 - Emergency Core Cooling System (ECCS) Pump Room l Exhaust Air Cleanup System (PREACS) l l

ISTS 3.7.13 - Fuel Building Air Cleanup System (FBACS) l ISTS 3.7.14 - Penetration Room Exhaust Air Cleanup System (PREACS) l

ISTS 3.7.4, ATMOSPHERIC DUMP VALVES (ADVS)

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

ADVs 3.7.4 1

3.7 PLANT SYSTEMS 3.7.4 Atmospheric Dump Valves (ADVs)

LCO 3.7.4 [Three] ADV lines shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, MODE 4 when steam generator is relied upon for heat removal.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required ADV line A.1 Restore required ADV line 7 days inoperable. to OPERABLE status.

[OR In accordance with the Risk Informed Completion Time Program]

B. Two or more required B.1 Restore all but one ADV 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ADV lines inoperable. line to OPERABLE status.

[OR In accordance with the Risk Informed Completion Time Program]

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND C.2 Be in MODE 4 without [24] hours reliance upon steam generator for heat removal.

Westinghouse STS 3.7.4-1 Rev. 5.0

ADVs 1 3.7.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 Verify one complete cycle of each ADV. [ [18] months OR In accordance with the Surveillance Frequency Control Program ]

SR 3.7.4.2 [ Verify one complete cycle of each ADV block [ [18] months valve.

OR In accordance with the Surveillance Frequency Control Program ] ]

Westinghouse STS 3.7.4-2 Rev. 5.0

JUSTIFICATION FOR DEVIATIONS ISTS 3.7.4, ATMOSPHERIC DUMP VALVES (ADVS)

1. Improved Standard Technical Specification (ISTS) 3.7.4, Atmospheric Dump Valves (ADVs), are not being included in the Turkey Point Nuclear Generating Station (PTN) Improved Technical Specifications (ITS). The PTN Current Technical Specifications (CTS) does not contain ADVs nor does the safety analysis credit ADVs for accident mitigation. However, the accident analysis describes the use of the ADVs if available.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

ADVs B 3.7.4 B 3.7 PLANT SYSTEMS 1

B 3.7.4 Atmospheric Dump Valves (ADVs)

BASES BACKGROUND The ADVs provide a method for cooling the unit to residual heat removal (RHR) entry conditions should the preferred heat sink via the Steam Bypass System to the condenser not be available, as discussed in the FSAR, Section [10.3] (Ref. 1). This is done in conjunction with the Auxiliary Feedwater System providing cooling water from the condensate storage tank (CST). The ADVs may also be required to meet the design cooldown rate during a normal cooldown when steam pressure drops too low for maintenance of a vacuum in the condenser to permit use of the Steam Dump System.

One ADV line for each of the [four] steam generators is provided. Each ADV line consists of one ADV and an associated block valve.

The ADVs are provided with upstream block valves to permit their being tested at power, and to provide an alternate means of isolation. The ADVs are equipped with pneumatic controllers to permit control of the cooldown rate.

The ADVs are usually provided with a pressurized gas supply of bottled nitrogen that, on a loss of pressure in the normal instrument air supply, automatically supplies nitrogen to operate the ADVs. The nitrogen supply is sized to provide the sufficient pressurized gas to operate the ADVs for the time required for Reactor Coolant System cooldown to RHR entry conditions.

A description of the ADVs is found in Reference 1. The ADVs are OPERABLE with only a DC power source available. In addition, handwheels are provided for local manual operation.

APPLICABLE The design basis of the ADVs is established by the capability to cool the SAFETY unit to RHR entry conditions. The design rate of [75]°F per hour is ANALYSES applicable for two steam generators, each with one ADV. This rate is adequate to cool the unit to RHR entry conditions with only one steam generator and one ADV, utilizing the cooling water supply available in the CST.

In the accident analysis presented in Reference 1, the ADVs are assumed to be used by the operator to cool down the unit to RHR entry conditions for accidents accompanied by a loss of offsite power. Prior to operator actions to cool down the unit, the ADVs and main steam safety valves (MSSVs) are assumed to operate automatically to relieve steam and maintain the steam generator pressure below the design value. For the recovery from a steam generator tube rupture (SGTR) event, the Westinghouse STS B 3.7.4-1 Rev. 5.0

1 ADVs B 3.7.4 BASES APPLICABLE SAFETY ANALYSES (continued) operator is also required to perform a limited cooldown to establish adequate subcooling as a necessary step to terminate the primary to secondary break flow into the ruptured steam generator. The time required to terminate the primary to secondary break flow for an SGTR is more critical than the time required to cool down to RHR conditions for this event and also for other accidents. Thus, the SGTR is the limiting event for the ADVs. The number of ADVs required to be OPERABLE to satisfy the SGTR accident analysis requirements depends upon the number of unit loops and consideration of any single failure assumptions regarding the failure of one ADV to open on demand.

The ADVs are equipped with block valves in the event an ADV spuriously fails to open or fails to close during use.

The ADVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO [Three] ADV lines are required to be OPERABLE. One ADV line is required from each of [three] steam generators to ensure that at least one ADV line is available to conduct a unit cooldown following an SGTR, in which one steam generator becomes unavailable, accompanied by a single, active failure of a second ADV line on an unaffected steam generator. The block valves must be OPERABLE to isolate a failed open ADV line. A closed block valve does not render it or its ADV line inoperable if operator action time to open the block valve is supported in the accident analysis.

Failure to meet the LCO can result in the inability to cool the unit to RHR entry conditions following an event in which the condenser is unavailable for use with the Steam Bypass System.

An ADV is considered OPERABLE when it is capable of providing controlled relief of the main steam flow and capable of fully opening and closing on demand.

APPLICABILITY In MODES 1, 2, and 3, and in MODE 4, when a steam generator is being relied upon for heat removal, the ADVs are required to be OPERABLE.

In MODE 5 or 6, an SGTR is not a credible event.

ACTIONS A.1 With one required ADV line inoperable, action must be taken to restore OPERABLE status within 7 days [or in accordance with the Risk Informed Completion Time Program]. The 7 day Completion Time allows for the redundant capability afforded by the remaining OPERABLE ADV lines, a nonsafety grade backup in the Steam Bypass System, and MSSVs.

Westinghouse STS B 3.7.4-2 Rev. 5.0

ADVs 1 B 3.7.4 BASES ACTIONS (continued)

B.1 With two or more ADV lines inoperable, action must be taken to restore all but one ADV line to OPERABLE status. Since the block valve can be closed to isolate an ADV, some repairs may be possible with the unit at power. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is reasonable to repair inoperable ADV lines, based on the availability of the Steam Bypass System and MSSVs, and the low probability of an event occurring during this period that would require the ADV lines. [Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program.]

C.1 and C.2 If the ADV lines cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4, without reliance upon steam generator for heat removal, within [24] hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.4.1 REQUIREMENTS To perform a controlled cooldown of the RCS, the ADVs must be able to be opened either remotely or locally and throttled through their full range.

This SR ensures that the ADVs are tested through a full control cycle at least once per fuel cycle. Performance of inservice testing or use of an ADV during a unit cooldown may satisfy this requirement. [ Operating experience has shown that these components usually pass the Surveillance when performed at the [18] month Frequency. The Frequency is acceptable from a reliability standpoint.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

Westinghouse STS B 3.7.4-3 Rev. 5.0

ADVs 1

B 3.7.4 BASES SURVEILLANCE REQUIREMENTS (continued)

[ SR 3.7.4.2 The function of the block valve is to isolate a failed open ADV. Cycling the block valve both closed and open demonstrates its capability to perform this function. Performance of inservice testing or use of the block valve during unit cooldown may satisfy this requirement. [ Operating experience has shown that these components usually pass the Surveillance when performed at the [18] month Frequency. The Frequency is acceptable from a reliability standpoint.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

] ]

REFERENCES 1. FSAR, Section [10.3].

Westinghouse STS B 3.7.4-4 Rev. 5.0

JUSTIFICATION FOR DEVIATIONS ITS 3.7.4 BASES, ATMOSPHERIC DUMP VALVES (ADVS)

1. Improved Standard Technical Specification (ISTS) 3.7.4 Bases are being deleted to be consistent with the deletion of ISTS 3.7.4.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

ISTS 3.7.12, EMERGENCY CORE COOLING SYSTEM (ECCS)

PUMP ROOM EXHAUST AIR CLEANUP SYSTEM (PREACS)

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

ECCS PREACS 3.7.12 1

3.7 PLANT SYSTEMS 3.7.12 Emergency Core Cooling System (ECCS) Pump Room Exhaust Air Cleanup System (PREACS)

LCO 3.7.12 Two ECCS PREACS trains shall be OPERABLE.

NOTE--------------------------------------------

The ECCS pump room boundary may be opened intermittently under administrative control.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One ECCS PREACS A.1 Restore ECCS PREACS 7 days train inoperable. train to OPERABLE status.

B. Two ECCS PREACS B.1 Restore ECCS pump room 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months trains inoperable due to boundary to OPERABLE inoperable ECCS pump status.

room boundary.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND C.2 --------------NOTE--------------

LCO 3.0.4.a is not applicable when entering MODE 4.

Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Westinghouse STS 3.7.12-1 Rev. 5.0

ECCS PREACS 3.7.12 1

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.12.1 Operate each ECCS PREACS train for [ 31 days 15 continuous minutes [with heaters operating].

OR In accordance with the Surveillance Frequency Control Program ]

SR 3.7.12.2 Perform required ECCS PREACS filter testing in In accordance accordance with the [Ventilation Filter Testing with the [VFTP]

Program (VFTP)].

SR 3.7.12.3 Verify each ECCS PREACS train actuates on an [ [18] months actual or simulated actuation signal, except for dampers and valves that are locked, sealed, or OR otherwise secured in the actuated position.

In accordance with the Surveillance Frequency Control Program ]

SR 3.7.12.4 Verify one ECCS PREACS train can maintain a [ [18] months on a pressure [-0.125] inches water gauge relative to STAGGERED atmospheric pressure during the [post accident] TEST BASIS mode of operation at a flow rate of [3000] cfm.

OR In accordance with the Surveillance Frequency Control Program ]

Westinghouse STS 3.7.12-2 Rev. 5.0

ECCS PREACS 1 3.7.12 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.7.12.5 [ Verify each ECCS PREACS filter bypass damper [ [18] months can be closed, except for dampers that are locked, sealed, or otherwise secured in the closed position. OR In accordance with the Surveillance Frequency Control Program ] ]

Westinghouse STS 3.7.12-3 Rev. 5.0

JUSTIFICATION FOR DEVIATIONS ISTS 3.7.12, EMERGENCY CORE COOLING SYSTEM (ECCS) PUMP ROOM EXHAUST AIR CLEANUP SYSTEM (PREACS)

1. Improved Standard Technical Specification (ISTS) 3.7.12, Emergency Core Cooling System (ECCS) Pump Room Exhaust Air Cleanup System (PREACS), is not being included in the Turkey Point Nuclear Generating Station (PTN) Improved Technical Specifications (ITS). The PTN Current Technical Specifications (CTS) does not contain ECCS PREACS nor does the safety analysis credit ECCS PREACS.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

ECCS PREACS B 3.7.12 1

B 3.7 PLANT SYSTEMS B 3.7.12 Emergency Core Cooling System (ECCS) Pump Room Exhaust Air Cleanup System (PREACS)

BASES BACKGROUND The ECCS PREACS filters air from the area of the active ECCS components during the recirculation phase of a loss of coolant accident (LOCA). The ECCS PREACS, in conjunction with other normally operating systems, also provides environmental control of temperature and humidity in the ECCS pump room area and the lower reaches of the Auxiliary Building.

The ECCS PREACS consists of two independent and redundant trains.

Each train consists of a heater, a prefilter or demister, a high efficiency particulate air (HEPA) filter, an activated charcoal adsorber section for removal of gaseous activity (principally iodines), and a fan. Ductwork, valves or dampers, and instrumentation also form part of the system, as well as demisters functioning to reduce the relative humidity of the air stream. A second bank of HEPA filters follows the adsorber section to collect carbon fines and provide backup in case the main HEPA filter bank fails. The downstream HEPA filter is not credited in the accident analysis, but serves to collect charcoal fines, and to back up the upstream HEPA filter should it develop a leak. The system initiates filtered ventilation of the pump room following receipt of a safety injection (SI) signal.

The ECCS PREACS is a standby system, aligned to bypass the system HEPA filters and charcoal adsorbers. During emergency operations, the ECCS PREACS dampers are realigned, and fans are started to begin filtration. Upon receipt of the actuating Engineered Safety Feature Actuation System signal(s), normal air discharges from the ECCS pump room isolate, and the stream of ventilation air discharges through the system filter trains. The prefilters remove any large particles in the air, and any entrained water droplets present, to prevent excessive loading of the HEPA filters and charcoal adsorbers.

The ECCS PREACS is discussed in the FSAR, Sections [6.5.1], [9.4.5],

and [15.6.5] (Refs. 1, 2, and 3, respectively) since it may be used for normal, as well as post accident, atmospheric cleanup functions. The primary purpose of the heaters is to maintain the relative humidity at an acceptable level, consistent with iodine removal efficiencies per Regulatory Guide 1.52 (Ref. 4).

Westinghouse STS B 3.7.12-1 Rev. 5.0

ECCS PREACS B 3.7.12 1

BASES APPLICABLE The design basis of the ECCS PREACS is established by the large break SAFETY LOCA. The system evaluation assumes a passive failure of the ECCS ANALYSES outside containment, such as an SI pump seal failure, during the recirculation mode. In such a case, the system limits radioactive release to within the 10 CFR 100 (Ref. 5) limits, or the NRC staff approved licensing basis (e.g., a specified fraction of Reference 5 limits). The analysis of the effects and consequences of a large break LOCA is presented in Reference 3. The ECCS PREACS also actuates following a small break LOCA, in those cases where the ECCS goes into the recirculation mode of long term cooling, to clean up releases of smaller leaks, such as from valve stem packing.

Two types of system failures are considered in the accident analysis:

complete loss of function, and excessive LEAKAGE. Either type of failure may result in a lower efficiency of removal for any gaseous and particulate activity released to the ECCS pump rooms following a LOCA.

The ECCS PREACS satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Two independent and redundant trains of the ECCS PREACS are required to be OPERABLE to ensure that at least one is available, assuming that a single failure disables the other train coincident with loss of offsite power. Total system failure could result in the atmospheric release from the ECCS pump room exceeding 10 CFR 100 limits in the event of a Design Basis Accident (DBA).

ECCS PREACS is considered OPERABLE when the individual components necessary to maintain the ECCS pump room filtration are OPERABLE in both trains.

An ECCS PREACS train is considered OPERABLE when its associated:

a. Fan is OPERABLE,

b. HEPA filter and charcoal adsorbers are not excessively restricting flow, and are capable of performing their filtration functions, and

c. Heater, demister, ductwork, valves, and dampers are OPERABLE and air circulation can be maintained.

The LCO is modified by a Note allowing the ECCS pump room boundary to be opened intermittently under administrative controls. For entry and exit through doors, the administrative control of the opening is performed by the person(s) entering or exiting the area. For other openings, these controls consist of stationing a dedicated individual at the opening who is in continuous communication with the control room. This individual will have a method to rapidly close the opening when a need for ECCS pump room isolation is indicated.

Westinghouse STS B 3.7.12-2 Rev. 5.0

ECCS PREACS B 3.7.12 1

BASES APPLICABILITY In MODES 1, 2, 3, and 4, the ECCS PREACS is required to be OPERABLE consistent with the OPERABILITY requirements of the ECCS.

In MODE 5 or 6, the ECCS PREACS is not required to be OPERABLE since the ECCS is not required to be OPERABLE.

ACTIONS A.1 With one ECCS PREACS train inoperable, action must be taken to restore OPERABLE status within 7 days. During this time, the remaining OPERABLE train is adequate to perform the ECCS PREACS function.

The 7 day Completion Time is appropriate because the risk contribution is less than that for the ECCS (72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time), and this system is not a direct support system for the ECCS. The 7 day Completion Time is based on the low probability of a DBA occurring during this time period, and ability of the remaining train to provide the required capability.

Concurrent failure of two ECCS PREACS trains would result in the loss of functional capability; therefore, LCO 3.0.3 must be entered immediately.

B.1

REVIEWERS NOTE-----------------------------------

Adoption of Condition B is dependent on a commitment from the licensee to have written procedures available describing compensatory measures to be taken in the event of an intentional or unintentional entry into Condition B.

If the ECCS pump room boundary is inoperable, the ECCS PREACS trains cannot perform their intended functions. Actions must be taken to restore an OPERABLE ECCS pump room boundary within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

During the period that the ECCS pump room boundary is inoperable, appropriate compensatory measures [consistent with the intent, as applicable, of GDC 19, 60, 64 and 10 CFR Part 100] should be utilized to protect plant personnel from potential hazards such as radioactive contamination, toxic chemicals, smoke, temperature and relative humidity, and physical security. Preplanned measures should be available to address these concerns for intentional and unintentional entry into the condition. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is reasonable based on the low probability of a DBA occurring during this time period, and the use of compensatory measures. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is a typically reasonable time to diagnose, plan and possibly repair, and test most problems with the ECCS pump room boundary.

Westinghouse STS B 3.7.12-3 Rev. 5.0

ECCS PREACS B 3.7.12 1

BASES ACTIONS (continued)

C.1 and C.2 If the ECCS PREACS train or ECCS pump room boundary cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which overall plant risk is reduced.

To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Remaining within the Applicability of the LCO is acceptable to accomplish short duration repairs to restore inoperable equipment because the plant risk in MODE 4 is similar to or lower than MODE 5 (Ref. 6). In MODE 4 the steam generators and Residual Heat Removal System are available to remove decay heat, which provides diversity and defense in depth. As stated in Reference 6, the steam turbine driven auxiliary feedwater pump must be available to remain in MODE 4. Should steam generator cooling be lost while relying on this Required Action, there are preplanned actions to ensure long-term decay heat removal. Voluntary entry into MODE 5 may be made as it is also acceptable from a risk perspective.

Required Action C.2 is modified by a Note that states that LCO 3.0.4.a is not applicable when entering MODE 4. This Note prohibits the use of LCO 3.0.4.a to enter MODE 4 during startup with the LCO not met.

However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering MODE 4, and establishment of risk management actions, if appropriate. LCO 3.0.4 is not applicable to, and the Note does not preclude, 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 allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.12.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly. As the environment and normal operating conditions on this system are not severe, testing each train once a month provides an adequate check on this system. Operation [with the heaters on] for 15 continuous minutes demonstrates OPERABILITY of the system.

Periodic operation ensures that [heater failure,] blockage, fan or motor failure, or excessive vibration can be detected for corrective action. [ The 31 day Frequency is based on the known reliability of equipment and the two train redundancy available.

Westinghouse STS B 3.7.12-4 Rev. 5.0

ECCS PREACS B 3.7.12 1

BASES SURVEILLANCE REQUIREMENTS (continued)

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

SR 3.7.12.2 This SR verifies that the required ECCS PREACS testing is performed in accordance with the [Ventilation Filter Testing Program (VFTP)]. The

[VFTP] includes testing HEPA filter performance, charcoal adsorbers efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations).

Specific test Frequencies and additional information are discussed in detail in the [VFTP].

SR 3.7.12.3 This SR verifies that each ECCS PREACS train starts and operates on an actual or simulated actuation signal. The SR excludes automatic dampers and valves that are locked, sealed, or otherwise secured in the actuated position. The SR does not apply to dampers or valves that are locked, sealed, or otherwise secured in the actuated position since the affected dampers or valves were verified to be in the actuated position prior to being locked, sealed, or otherwise secured. Placing an automatic valve or damper in a locked, sealed, or otherwise secured position requires an assessment of the OPERABILITY of the system or any supported systems, including whether it is necessary for the valve or damper to be repositioned to the non-actuated position to support the accident analysis. Restoration of an automatic valve or damper to the non-actuated position requires verification that the SR has been met within its required Frequency. [ The [18] month Frequency is consistent with that specified in Reference 4.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Westinghouse STS B 3.7.12-5 Rev. 5.0

ECCS PREACS 1 B 3.7.12 BASES SURVEILLANCE REQUIREMENTS (continued)

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

SR 3.7.12.4 This SR verifies the integrity of the ECCS pump room enclosure. The ability of the ECCS pump room to maintain a negative pressure, with respect to potentially uncontaminated adjacent areas, is periodically tested to verify proper functioning of the ECCS PREACS. During the

[post accident] mode of operation, the ECCS PREACS is designed to maintain a slight negative pressure in the ECCS pump room, with respect to adjacent areas, to prevent unfiltered LEAKAGE. The ECCS PREACS is designed to maintain a [-0.125] inches water gauge relative to atmospheric pressure at a flow rate of [3000] cfm from the ECCS pump room. [ The Frequency of [18] months is consistent with the guidance provided in NUREG-0800, Section 6.5.1 (Ref. 7).

This test is conducted with the tests for filter penetration; thus, an

[18] month Frequency on a STAGGERED TEST BASIS is consistent with that specified in Reference 4.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

Westinghouse STS B 3.7.12-6 Rev. 5.0

ECCS PREACS B 3.7.12 1

BASES SURVEILLANCE REQUIREMENTS (continued)

[ SR 3.7.12.5 Operating the ECCS PREACS bypass damper is necessary to ensure that the system functions properly. The OPERABILITY of the ECCS PREACS bypass damper is verified if it can be specified in Reference 4.

The SR excludes automatic dampers that are locked, sealed, or otherwise secured in the closed position. The SR does not apply to dampers that are locked, sealed, or otherwise secured in the closed position since the affected dampers were verified to be in the closed position prior to being locked, sealed, or otherwise secured. Placing an automatic damper in a locked, sealed, or otherwise secured position requires an assessment of the OPERABILITY of the system or any supported systems, including whether it is necessary for the damper to be opened to support the accident analysis. Restoration of an automatic damper to the opened position requires verification that the SR has been met within its required Frequency. [ An [18] month Frequency is consistent with that specified in Reference 4.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

] ]

REFERENCES 1. FSAR, Section [6.5.1].

2. FSAR, Section [9.4.5].

3. FSAR, Section [15.6.5].

4. Regulatory Guide 1.52 (Rev. 2).

5. 10 CFR 100.11.

6. WCAP-16294-NP-A, Rev. 1, "Risk-Informed Evaluation of Changes to Technical Specification Required Action Endstates for Westinghouse NSSS PWRs," June 2010.

7. NUREG-0800, Section 6.5.1, Rev. 2, July 1981.

Westinghouse STS B 3.7.12-7 Rev. 5.0

JUSTIFICATION FOR DEVIATIONS ITS 3.7.12 BASES, EMERGENCY CORE COOLING SYSTEM (ECCS) PUMP ROOM EXHAUST AIR CLEANUP SYSTEM (PREACS)

1. Improved Standard Technical Specification (ISTS) 3.7.12 Bases are being deleted to be consistent with the deletion of ISTS 3.7.12.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

ISTS 3.7.13, FUEL BUILDING AIR CLEANUP SYSTEM (FBACS)

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

FBACS 3.7.13 1

3.7 PLANT SYSTEMS 3.7.13 Fuel Building Air Cleanup System (FBACS)

LCO 3.7.13 Two FBACS trains shall be OPERABLE.

NOTE--------------------------------------------

The fuel building boundary may be opened intermittently under administrative control.

APPLICABILITY: [MODES 1, 2, 3, and 4, ]

During movement of [recently] irradiated fuel assemblies in the fuel building.

ACTIONS

NOTE-----------------------------------------------------------

LCO 3.0.3 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. One FBACS train A.1 Restore FBACS train to 7 days inoperable. OPERABLE status.

B. Two FBACS trains B.1 Restore fuel building 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months inoperable due to boundary to OPERABLE inoperable fuel building status.

boundary in MODE 1, 2, 3, or 4.

Westinghouse STS 3.7.13-1 Rev. 5.0

FBACS 3.7.13 1

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. [ Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A or B AND not met in MODE 1, 2, 3, or 4. C.2 --------------NOTE--------------

LCO 3.0.4.a is not OR applicable when entering MODE 4.

Two FBACS trains -------------------------------------

inoperable in MODE 1, 2, 3, or 4 for reasons Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ]

other than Condition B.

D. Required Action and D.1 Place OPERABLE FBACS Immediately associated Completion train in operation.

Time [of Condition A] not met during movement of OR

[recently] irradiated fuel assemblies in the fuel D.2 Suspend movement of Immediately building. [recently] irradiated fuel assemblies in the fuel building.

E. Two FBACS trains E.1 Suspend movement of Immediately inoperable during [recently] irradiated fuel movement of [recently] assemblies in the fuel irradiated fuel building.

assemblies in the fuel building.

Westinghouse STS 3.7.13-2 Rev. 5.0

FBACS 1

3.7.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.13.1 Operate each FBACS train for 15 continuous [ 31 days minutes [with heaters operating].

OR In accordance with the Surveillance Frequency Control Program ]

SR 3.7.13.2 Perform required FBACS filter testing in accordance In accordance with the [Ventilation Filter Testing Program (VFTP)]. with the [VFTP]

SR 3.7.13.3 [ Verify each FBACS train actuates on an actual or [ [18] months simulated actuation signal, except for dampers and valves that are locked, sealed, or otherwise secured OR in the actuated position.

In accordance with the Surveillance Frequency Control Program ] ]

SR 3.7.13.4 Verify one FBACS train can maintain a pressure [ [18] months on a

[-0.125] inches water gauge with respect to STAGGERED atmospheric pressure during the [post accident] TEST BASIS mode of operation at a flow rate [20,000] cfm.

OR In accordance with the Surveillance Frequency Control Program ]

Westinghouse STS 3.7.13-3 Rev. 5.0

FBACS 3.7.13 1

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.7.13.5 [ Verify each FBACS filter bypass damper can be [ [18] months closed, except for dampers that are locked, sealed, or otherwise secured in the closed position. OR In accordance with the Surveillance Frequency Control Program ] ]

Westinghouse STS 3.7.13-4 Rev. 5.0

JUSTIFICATION FOR DEVIATIONS ISTS 3.7.13, FUEL BUILDING AIR CLEANUP SYSTEM (FBACS)

1. Improved Standard Technical Specification (ISTS) 3.7.13, Fuel Building Air Cleanup System (FBACS), is not being included in the Turkey Point Nuclear Generating Station (PTN) Improved Technical Specifications (ITS). The PTN Current Technical Specifications (CTS) does not contain FBACS nor does the safety analysis credit FBACS.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

FBACS B 3.7.13 B 3.7 PLANT SYSTEMS B 3.7.13 Fuel Building Air Cleanup System (FBACS) 1 BASES BACKGROUND The FBACS filters airborne radioactive particulates from the area of the fuel pool following a fuel handling accident or loss of coolant accident (LOCA). The FBACS, in conjunction with other normally operating systems, also provides environmental control of temperature and humidity in the fuel pool area.

The FBACS consists of two independent and redundant trains. Each train consists of a heater, a prefilter or demister, a high efficiency particulate air (HEPA) filter, an activated charcoal adsorber section for removal of gaseous activity (principally iodines), and a fan. Ductwork, valves or dampers, and instrumentation also form part of the system, as well as demisters, functioning to reduce the relative humidity of the airstream. A second bank of HEPA filters follows the adsorber section to collect carbon fines and provide backup in case the main HEPA filter bank fails. The downstream HEPA filter is not credited in the analysis, but serves to collect charcoal fines, and to back up the upstream HEPA filter should it develop a leak. The system initiates filtered ventilation of the fuel handling building following receipt of a high radiation signal.

The FBACS is a standby system, parts of which may also be operated during normal plant operations. Upon receipt of the actuating signal, normal air discharges from the building, the fuel handling building is isolated, and the stream of ventilation air discharges through the system filter trains. The prefilters or demisters remove any large particles in the air, and any entrained water droplets present, to prevent excessive loading of the HEPA filters and charcoal adsorbers.

The FBACS is discussed in the FSAR, Sections [6.5.1], [9.4.5],

and [15.7.4] (Refs. 1, 2, and 3, respectively) because it may be used for normal, as well as post accident, atmospheric cleanup functions.

APPLICABLE The FBACS design basis is established by the consequences of the SAFETY limiting Design Basis Accident (DBA), which is a fuel handling accident ANALYSES [involving handling recently irradiated fuel]. The analysis of the fuel handling accident, given in Reference 3, assumes that all fuel rods in an assembly are damaged. The analysis of the LOCA assumes that radioactive materials leaked from the Emergency Core Cooling System (ECCS) are filtered and adsorbed by the FBACS. The DBA analysis of the fuel handling accident assumes that only one train of the FBACS is functional due to a single failure that disables the other train. The accident analysis accounts for the reduction in airborne radioactive Westinghouse STS B 3.7.13-1 Rev. 5.0

FBACS B 3.7.13 1

BASES APPLICABLE SAFETY ANALYSES (continued) material provided by the one remaining train of this filtration system. The amount of fission products available for release from the fuel handling building is determined for a fuel handling accident and for a LOCA. [Due to radioactive decay, FBACS is only required to isolate during fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous [X] days).]

These assumptions and the analysis follow the guidance provided in Regulatory Guide 1.25 (Ref. 4).

The FBACS satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Two independent and redundant trains of the FBACS are required to be OPERABLE to ensure that at least one train is available, assuming a single failure that disables the other train, coincident with a loss of offsite power. Total system failure could result in the atmospheric release from the fuel handling building exceeding the 10 CFR 100 (Ref. 5) limits in the event of a fuel handling accident [involving handling recently irradiated fuel].

The FBACS is considered OPERABLE when the individual components necessary to control exposure in the fuel handling building are OPERABLE in both trains. An FBACS train is considered OPERABLE when its associated:

a. Fan is OPERABLE,

b. HEPA filter and charcoal adsorber are not excessively restricting flow, and are capable of performing their filtration function, and

c. Heater, demister, ductwork, valves, and dampers are OPERABLE, and air circulation can be maintained.

The LCO is modified by a Note allowing the fuel building boundary to be opened intermittently under administrative controls. For entry and exit through doors the administrative control of the opening is performed by the person(s) entering or exiting the area. For other openings, these controls consist of stationing a dedicated individual at the opening who is in continuous communication with the control room. This individual will have a method to rapidly close the opening when a need for fuel building isolation is indicated.

Westinghouse STS B 3.7.13-2 Rev. 5.0

FBACS B 3.7.13 1

BASES APPLICABILITY In MODE 1, 2, 3, or 4, the FBACS is required to be OPERABLE to provide fission product removal associated with ECCS leaks due to a LOCA and leakage from containment and annulus.

In MODE 5 or 6, the FBACS is not required to be OPERABLE since the ECCS is not required to be OPERABLE.

During movement of [recently] irradiated fuel in the fuel handling area, the FBACS is required to be OPERABLE to alleviate the consequences of a fuel handling accident.

ACTIONS LCO 3.0.3 is not applicable while in MODE 5 or 6. However, since irradiated fuel assembly movement can occur in MODE 1, 2, 3, or 4, the ACTIONS have been modified by a Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operations. Entering LCO 3.0.3, while in MODE 1, 2, 3, or 4 would require the unit to be shutdown unnecessarily.

A.1 With one FBACS train inoperable, action must be taken to restore OPERABLE status within 7 days. During this period, the remaining OPERABLE train is adequate to perform the FBACS function. The 7 day Completion Time is based on the risk from an event occurring requiring the inoperable FBACS train, and the remaining FBACS train providing the required protection.

B.1

REVIEWERS NOTE-------------------------

Adoption of Condition B is dependent on a commitment from the licensee to have guidance available describing compensatory measures to be taken in the event of an intentional and unintentional entry into Condition B.

If the fuel building boundary is inoperable in MODE 1, 2, 3, or 4, the FBACS trains cannot perform their intended functions. Actions must be taken to restore an OPERABLE fuel building boundary within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

During the period that the fuel building boundary is inoperable, appropriate compensatory measures [consistent with the intent, as Westinghouse STS B 3.7.13-3 Rev. 5.0

FBACS B 3.7.13 1

BASES ACTIONS (continued) applicable, of GDC 19, 60, 61, 63, 64 and 10 CFR Part 100] should be utilized to protect plant personnel from potential hazards such as radioactive contamination, toxic chemicals, smoke, temperature and relative humidity, and physical security. Preplanned measures should be available to address these concerns for intentional and unintentional entry into the condition. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is reasonable based on the low probability of a DBA occurring during this time period, and the use of compensatory measures. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is a typically reasonable time to diagnose, plan and possibly repair, and test most problems with the fuel building boundary.

[ C.1 and C.2 In MODE 1, 2, 3, or 4, when Required Action A.1 or B.1 cannot be completed within the associated Completion Time, or when both FBACS trains are inoperable for reasons other than an inoperable fuel building boundary (i.e., Condition B), the unit must be placed in a MODE in which overall plant risk is reduced. To achieve this status, the unit must be placed in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Remaining within the Applicability of the LCO is acceptable to accomplish short duration repairs to restore inoperable equipment because the plant risk in MODE 4 is similar to or lower than MODE 5 (Ref. 6). In MODE 4 the steam generators and Residual Heat Removal System are available to remove decay heat, which provides diversity and defense in depth. As stated in Reference 6, the steam turbine driven auxiliary feedwater pump must be available to remain in MODE 4. Should steam generator cooling be lost while relying on this Required Action, there are preplanned actions to ensure long-term decay heat removal. Voluntary entry into MODE 5 may be made as it is also acceptable from a risk perspective.

Required Action C.2 is modified by a Note that states that LCO 3.0.4.a is not applicable when entering MODE 4. This Note prohibits the use of LCO 3.0.4.a to enter MODE 4 during startup with the LCO not met.

However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering MODE 4, and establishment of risk management actions, if appropriate. LCO 3.0.4 is not applicable to, and the Note does not preclude, 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.

Westinghouse STS B 3.7.13-4 Rev. 5.0

FBACS B 3.7.13 1

BASES ACTIONS (continued)

The Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. ]

D.1 and D.2 When Required Action A.1 cannot be completed within the required Completion Time, during movement of [recently] irradiated fuel assemblies in the fuel building, the OPERABLE FBACS train must be started immediately or [recently] irradiated fuel movement suspended.

This action ensures that the remaining train is OPERABLE, that no undetected failures preventing system operation will occur, and that any active failure will be readily detected.

If the system is not placed in operation, this action requires suspension of

[recently] irradiated fuel movement, which precludes a fuel handling accident [involving handling recently irradiated fuel]. This does not preclude the movement of fuel assemblies to a safe position.

E.1 When two trains of the FBACS are inoperable during movement of

[recently] irradiated fuel assemblies in the fuel building, action must be taken to place the unit in a condition in which the LCO does not apply.

Action must be taken immediately to suspend movement of [recently]

irradiated fuel assemblies in the fuel building. This does not preclude the movement of fuel to a safe position.

SURVEILLANCE SR 3.7.13.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly. As the environmental and normal operating conditions on this system are not severe, testing each train once every month provides an adequate check on this system.

Operation [with the heaters on] for 15 continuous minutes demonstrates OPERABILITY of the system. Periodic operation ensures that [heater failure,] blockage, fan or motor failure, or excessive vibration can be detected for corrective action. [ The 31 day Frequency is based on the known reliability of the equipment and the two train redundancy available.

OR Westinghouse STS B 3.7.13-5 Rev. 5.0

FBACS B 3.7.13 1

BASES SURVEILLANCE REQUIREMENTS (continued)

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

[ SR 3.7.13.2 This SR verifies that the required FBACS testing is performed in accordance with the [Ventilation Filter Testing Program (VFTP)]. The

[VFTP] includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations).

Specific test frequencies and additional information are discussed in detail in the [VFTP]. ]

[ SR 3.7.13.3 This SR verifies that each FBACS train starts and operates on an actual or simulated actuation signal. The SR excludes automatic dampers and valves that are locked, sealed, or otherwise secured in the actuated position. The SR does not apply to dampers or valves that are locked, sealed, or otherwise secured in the actuated position since the affected dampers or valves were verified to be in the actuated position prior to being locked, sealed, or otherwise secured. Placing an automatic valve or damper in a locked, sealed, or otherwise secured position requires an assessment of the OPERABILITY of the system or any supported systems, including whether it is necessary for the valve or damper to be repositioned to the non-actuated position to support the accident analysis.

Restoration of an automatic valve or damper to the non-actuated position requires verification that the SR has been met within its required Frequency. [ The [18] month Frequency is consistent with Reference 7. ]

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Westinghouse STS B 3.7.13-6 Rev. 5.0

FBACS B 3.7.13 1

BASES SURVEILLANCE REQUIREMENTS (continued)

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

SR 3.7.13.4 This SR verifies the integrity of the fuel building enclosure. The ability of the fuel building to maintain negative pressure with respect to potentially uncontaminated adjacent areas is periodically tested to verify proper function of the FBACS. During the [post accident] mode of operation, the FBACS is designed to maintain a slight negative pressure in the fuel building, to prevent unfiltered LEAKAGE. The FBACS is designed to maintain a [-0.125] inches water gauge with respect to atmospheric pressure at a flow rate of [20,000] cfm to the fuel building. [ The Frequency of [18] months is consistent with the guidance provided in NUREG-0800, Section 6.5.1 (Ref. 8).

An [18] month Frequency (on a STAGGERED TEST BASIS) is consistent with Reference 7.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

[ SR 3.7.13.5 Operating the FBACS filter bypass damper is necessary to ensure that the system functions properly. The OPERABILITY of the FBACS filter bypass damper is verified if it can be closed. The SR excludes automatic dampers that are locked, sealed, or otherwise secured in the closed position. The SR does not apply to dampers that are locked, sealed, or Westinghouse STS B 3.7.13-7 Rev. 5.0

FBACS B 3.7.13 1

BASES SURVEILLANCE REQUIREMENTS (continued) otherwise secured in the closed position since the affected dampers were verified to be in the closed position prior to being locked, sealed, or otherwise secured. Placing an automatic damper in a locked, sealed, or otherwise secured position requires an assessment of the OPERABILITY of the system or any supported systems, including whether it is necessary for the damper to be opened to support the accident analysis.

Restoration of an automatic damper to the opened position requires verification that the SR has been met within its required Frequency. [ An

[18] month Frequency is consistent with Reference 7.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

] ]

REFERENCES 1. FSAR, Section [6.5.1].

2. FSAR, Section [9.4.5].

3. FSAR, Section [15.7.4].

4. Regulatory Guide 1.25.

5. 10 CFR 100.

6. WCAP-16294-NP-A, Rev. 1, "Risk-Informed Evaluation of Changes to Technical Specification Required Action Endstates for Westinghouse NSSS PWRs," June 2010.

7. Regulatory Guide 1.52, Rev. [2].

8. NUREG-0800, Section 6.5.1, Rev. 2, July 1981.

Westinghouse STS B 3.7.13-8 Rev. 5.0

JUSTIFICATION FOR DEVIATIONS ITS 3.7.13 BASES, FUEL BUILDING AIR CLEANUP SYSTEM (FBACS)

1. Improved Standard Technical Specification (ISTS) 3.7.13 Bases are being deleted to be consistent with the deletion of ISTS 3.7.13.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

ISTS 3.7.14, PENETRATION ROOM EXHAUST AIR CLEANUP SYSTEM (PREACS)

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

PREACS 3.7.14 1

3.7 PLANT SYSTEMS 3.7.14 Penetration Room Exhaust Air Cleanup System (PREACS)

LCO 3.7.14 Two PREACS trains shall be OPERABLE.

NOTE--------------------------------------------

The penetration room boundary may be opened intermittently under administrative control.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One PREACS train A.1 Restore PREACS train to 7 days inoperable. OPERABLE status.

B. Two PREACS trains B.1 Restore penetration room 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months inoperable due to boundary to OPERABLE inoperable penetration status.

room boundary.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND C.2 --------------NOTE--------------

LCO 3.0.4.a is not applicable when entering MODE 4.

Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Westinghouse STS 3.7.14-1 Rev. 5.0

PREACS 3.7.14 1

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.14.1 Operate each PREACS train for 15 continuous [ 31 days minutes [with heaters operating].

OR In accordance with the Surveillance Frequency Control Program ]

SR 3.7.14.2 Perform required PREACS filter testing in In accordance accordance with the [Ventilation Filter Testing with the [VFTP]

Program (VFTP)].

SR 3.7.14.3 [ Verify each PREACS train actuates on an actual or [ [18] months simulated actuation signal, except for dampers and valves that are locked, sealed, or otherwise secured OR in the actuated position.

In accordance with the Surveillance Frequency Control Program ] ]

SR 3.7.14.4 [ Verify one PREACS train can maintain a pressure [ [18] months on a

[-0.125] inches water gauge relative to STAGGERED atmospheric pressure during the [post accident] TEST BASIS mode of operation at a flow rate of [3000] cfm.

OR In accordance with the Surveillance Frequency Control Program ] ]

Westinghouse STS 3.7.14-2 Rev. 5.0

PREACS 3.7.14 1

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.7.14.5 [ Verify each PREACS filter bypass damper can be [ [18] months closed, except for dampers that are locked, sealed, or otherwise secured in the closed position. OR In accordance with the Surveillance Frequency Control Program ] ]

Westinghouse STS 3.7.14-3 Rev. 5.0

JUSTIFICATION FOR DEVIATIONS ISTS 3.7.14, PENETRATION ROOM EXHAUST AIR CLEANUP SYSTEM (PREACS)

1. Improved Standard Technical Specification (ISTS) 3.7.14, Pump Room Exhaust Air Cleanup System (PREACS), is not being included in the Turkey Point Nuclear Generating Station (PTN) Improved Technical Specifications (ITS). The PTN Current Technical Specifications (CTS) does not contain PREACS nor does the safety analysis credit PREACS.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

PREACS B 3.7.14 B 3.7 PLANT SYSTEMS 1 B 3.7.14 Penetration Room Exhaust Air Cleanup System (PREACS)

BASES BACKGROUND The PREACS filters air from the penetration area between containment and the Auxiliary Building.

The PREACS consists of two independent and redundant trains. Each train consists of a heater, a prefilter or demister, a high efficiency particulate air (HEPA) filter, an activated charcoal adsorber section for removal of gaseous activity (principally iodines), and a fan. Ductwork, valves or dampers, and instrumentation, as well as demisters, functioning to reduce the relative humidity of the air stream, also form part of the system. A second bank of HEPA filters, which follows the adsorber section, collects carbon fines and provides backup in case of failure of the main HEPA filter bank. The downstream HEPA filter, although not credited in the accident analysis, collects charcoal fines and serves as a backup should the upstream HEPA filter develop a leak. The system initiates filtered ventilation following receipt of a safety injection signal.

The PREACS is a standby system, parts of which may also operate during normal unit operations. During emergency operations, the PREACS dampers are realigned and fans are started to initiate filtration.

Upon receipt of the actuating signal(s), normal air discharges from the penetration room, the penetration room is isolated, and the stream of ventilation air discharges through the system filter trains. The prefilters remove any large particles in the air, as well as any entrained water droplets, to prevent excessive loading of the HEPA filters and charcoal adsorbers.

The PREACS is discussed in the FSAR, Sections [6.5.1], [9.4.5],

and [15.6.5] (Refs. 1, 2, and 3, respectively) since it may be used for normal, as well as post accident, atmospheric cleanup functions. Heaters may be included for moisture removal on systems operating in high humidity conditions. The primary purpose of the heaters is to maintain the relative humidity at an acceptable level consistent with iodine removal efficiencies per Regulatory Guide 1.52 (Ref. 4).

APPLICABLE The PREACS design basis is established by the large break loss of SAFETY coolant accident (LOCA). The system evaluation assumes a passive ANALYSES failure outside containment, such as valve packing leakage during a Design Basis Accident (DBA). In such a case, the system restricts the failure outside containment, such as valve packing leakage during a Westinghouse STS B 3.7.14-1 Rev. 5.0

PREACS B 3.7.14 1

BASES APPLICABLE SAFETY ANALYSES (continued)

Design Basis Accident (DBA). In such a case, the system restricts the radioactive release to within the 10 CFR 100 (Ref. 4) limits, or the NRC staff approved licensing basis (e.g., a specified fraction of 10 CFR 100 limits). The analysis of the effects and consequences of a large break LOCA are presented in Reference 3.

Two types of system failures are considered in the accident analysis: a complete loss of function, and excessive LEAKAGE. Either type of failure may result in less efficient removal of any gaseous or particulate material released to the penetration room following a LOCA.

The PREACS satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Two independent and redundant trains of the PREACS are required to be OPERABLE to ensure that at least one train is available, assuming there is a single failure disabling the other train coincident with a loss of offsite power.

The PREACS is considered OPERABLE when the individual components necessary to control radioactive releases are OPERABLE in both trains.

A PREACS train is considered OPERABLE when its associated:

a. Fan is OPERABLE,

b. HEPA filter and charcoal adsorber are not excessively restricting flow, and are capable of performing their filtration functions, and

c. Heater, demister, ductwork, valves, and dampers are OPERABLE and air circulation can be maintained.

The LCO is modified by a Note allowing the penetration room boundary to be opened intermittently under administrative controls. For entry and exit through doors, the administrative control of the opening is performed by the person(s) entering or exiting the area. For other openings, these controls consist of stationing a dedicated individual at the opening who is in continuous communication with the control room. This individual will have a method to rapidly close the opening when a need for penetration room isolation is indicated.

APPLICABILITY In MODES 1, 2, 3, and 4, the PREACS is required to be OPERABLE, consistent with the OPERABILITY requirements of the Emergency Core Cooling System (ECCS).

In MODE 5 or 6, the PREACS is not required to be OPERABLE since the ECCS is not required to be OPERABLE.

Westinghouse STS B 3.7.14-2 Rev. 5.0

PREACS B 3.7.14 1

BASES ACTIONS A.1 With one PREACS train inoperable, the action must be taken to restore OPERABLE status within 7 days. During this period, the remaining OPERABLE train is adequate to perform the PREACS function. The 7 day Completion Time is appropriate because the risk contribution of the PREACS is less than that of the ECCS (72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time), and this system is not a direct support system for the ECCS. The 7 day Completion Time is based on the low probability of a DBA occurring during this period, and the remaining train providing the required capability.

B.1

REVIEWERS NOTE-----------------------------------

Adoption of Condition B is dependent on a commitment from the licensee to have guidance available describing compensatory measures to be taken in the event of an intentional and unintentional entry into Condition B.

If the penetration room boundary is inoperable, the PREACS trains cannot perform their intended functions. Actions must be taken to restore an OPERABLE penetration room boundary within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . During the period that the penetration room boundary is inoperable, appropriate compensatory measures [consistent with the intent, as applicable, of GDC 19, 60, 64 and 10 CFR Part 100] should be utilized to protect plant personnel from potential hazards such as radioactive contamination, toxic chemicals, smoke, temperature and relative humidity, and physical security. Preplanned measures should be available to address these concerns for intentional and unintentional entry into the condition. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is reasonable based on the low probability of a DBA occurring during this time period, and the use of compensatory measures. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is a typically reasonable time to diagnose, plan and possibly repair, and test most problems with the penetration room boundary.

C.1 and C.2 If the inoperable train or penetration room boundary cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which overall plant risk is reduced. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Westinghouse STS B 3.7.14-3 Rev. 5.0

PREACS B 3.7.14 1

BASES ACTIONS (continued)

Remaining within the Applicability of the LCO is acceptable to accomplish short duration repairs to restore inoperable equipment because the plant risk in MODE 4 is similar to or lower than MODE 5 (Ref. 5). In MODE 4 the steam generators and Residual Heat Removal System are available to remove decay heat, which provides diversity and defense in depth. As stated in Reference 5, the steam turbine driven auxiliary feedwater pump must be available to remain in MODE 4. Should steam generator cooling be lost while relying on this Required Action, there are preplanned actions to ensure long-term decay heat removal. Voluntary entry into MODE 5 may be made as it is also acceptable from a risk perspective.

Required Action C.2 is modified by a Note that states that LCO 3.0.4.a is not applicable when entering MODE 4. This Note prohibits the use of LCO 3.0.4.a to enter MODE 4 during startup with the LCO not met.

However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering MODE 4, and establishment of risk management actions, if appropriate. LCO 3.0.4 is not applicable to, and the Note does not preclude, 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 Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.14.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly. As the environmental and normal operating conditions on this system are not severe, testing each train once every month provides an adequate check on this system. Operation [with the heaters on] for 15 continuous minutes demonstrates OPERABILITY of the system. Periodic operation ensures that [heater failure,] blockage, fan or motor failure, or excessive vibration can be detected for corrective action.

[ The 31 day Frequency is based on the known reliability of equipment and the two train redundancy available.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Westinghouse STS B 3.7.14-4 Rev. 5.0

PREACS B 3.7.14 1

BASES SURVEILLANCE REQUIREMENTS (continued)

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

SR 3.7.14.2 This SR verifies that the required PREACS testing is performed in accordance with the [Ventilation Filter Testing Program (VFTP)]. The

[VFTP] includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations).

Specific test frequencies and additional information are discussed in detail in the [VFTP].

[ SR 3.7.14.3 This SR verifies that each PREACS starts and operates on an actual or simulated actuation signal. The SR excludes automatic dampers and valves that are locked, sealed, or otherwise secured in the actuated position. The SR does not apply to dampers or valves that are locked, sealed, or otherwise secured in the actuated position since the affected dampers or valves were verified to be in the actuated position prior to being locked, sealed, or otherwise secured. Placing an automatic valve or damper in a locked, sealed, or otherwise secured position requires an assessment of the OPERABILITY of the system or any supported systems, including whether it is necessary for the valve or damper to be repositioned to the non-actuated position to support the accident analysis.

Restoration of an automatic valve or damper to the non-actuated position requires verification that the SR has been met within its required Frequency. [ The [18] month Frequency is consistent with that specified in Reference 6.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Westinghouse STS B 3.7.14-5 Rev. 5.0

PREACS B 3.7.14 1

BASES SURVEILLANCE REQUIREMENTS (continued)

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

] ]

[ SR 3.7.14.4 This SR verifies the integrity of the penetration room enclosure. The ability of the penetration room to maintain a negative pressure, with respect to potentially uncontaminated adjacent areas, is periodically tested to verify proper function of PREACS. During the [post accident]

mode of operation, the PREACS is designed to maintain a

[-0.125] inches water gauge relative to atmospheric pressure at a flow rate of [3000] cfm in the penetration room, with respect to adjacent areas, to prevent unfiltered LEAKAGE.

The minimum system flow rate maintains a slight negative pressure in the penetration room area, and provides sufficient air velocity to transport particulate contaminants, assuming only one filter train is operating. The number of filter elements is selected to limit the flow rate through any individual element to about [3000] cfm. This may vary based on filter housing geometry. The maximum limit ensures that the flow through, and pressure drop across, each filter element are not excessive.

The number and depth of the adsorber elements ensure that, at the maximum flow rate, the residence time of the air stream in the charcoal bed achieves the desired adsorption rate. At least a [0.125] second residence time is necessary for an assumed [99]% efficiency.

The filters have a certain pressure drop at the design flow rate when clean. The magnitude of the pressure drop indicates acceptable performance, and is based on manufacturers' recommendations for the filter and adsorber elements at the design flow rate. An increase in pressure drop or a decrease in flow indicates that the filter is being loaded or that there are other problems with the system.

[ This test is conducted along with the tests for filter penetration; thus, the

[18] month Frequency is consistent with that specified in Reference 6.

The Frequency of [18] months is also consistent with the guidance provided in NUREG-0800 (Ref. 7).

OR Westinghouse STS B 3.7.14-6 Rev. 5.0

PREACS B 3.7.14 1

BASES SURVEILLANCE REQUIREMENTS (continued)

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

] ]

[ SR 3.7.14.5 It is necessary to operate the PREACS filter bypass damper to ensure that the system functions properly. The OPERABILITY of the PREACS filter bypass damper is verified if it can be closed. The SR excludes automatic dampers that are locked, sealed, or otherwise secured in the closed position. The SR does not apply to dampers that are locked, sealed, or otherwise secured in the closed position since the affected dampers were verified to be in the closed position prior to being locked, sealed, or otherwise secured. Placing an automatic damper in a locked, sealed, or otherwise secured position requires an assessment of the OPERABILITY of the system or any supported systems, including whether it is necessary for the damper to be opened to support the accident analysis. Restoration of an automatic damper to the opened position requires verification that the SR has been met within its required Frequency. [ An [18] month Frequency is consistent with that specified in Reference 6.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

] ]

Westinghouse STS B 3.7.14-7 Rev. 5.0

PREACS B 3.7.14 1

BASES REFERENCES 1. FSAR, Section [6.5.1].

2. FSAR, Section [9.4.5].

3. FSAR, Section [15.6.5].

4. 10 CFR 100.

5. WCAP-16294-NP-A, Rev. 1, "Risk-Informed Evaluation of Changes to Technical Specification Required Action Endstates for Westinghouse NSSS PWRs," June 2010.

6. Regulatory Guide 1.52, Rev. [2].

7. NUREG-0800, Section 6.5.1, Rev. 2, July 1981.

Westinghouse STS B 3.7.14-8 Rev. 5.0

JUSTIFICATION FOR DEVIATIONS ITS 3.7.14 BASES, PENETRATION ROOM EXHAUST AIR CLEANUP SYSTEM (PREACS)

1. Improved Standard Technical Specification (ISTS) 3.7.14 Bases are being deleted to be consistent with the deletion of ISTS 3.7.14.

Turkey Point Unit 3 and Unit 4 Page 1 of 1

ML23151A450

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10.2 REFERENCES

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