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Improved Technical Specifications Conversion, Chapter 2.0, Safety Limits, Revision 1 - Volume 4
	ML23275A114
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Site:	Saint Lucie
Issue date:	10/02/2023
From:	; Florida Power & Light Co
To:	; Office of Nuclear Reactor Regulation
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ENCLOSURE 2 VOLUME 4 ST. LUCIE PLANT UNIT 1 AND UNIT 2 IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS CHAPTER 2.0 SAFETY LIMITS Revision 1

LIST OF ATTACHMENTS

1. ITS Chapter 2.0, Safety Limits

ATTACHMENT 1 ITS Chapter 2.0, Safety Limits

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

ITS Chapter 2.0 SECTION 2.0 A01 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

ITS ITS Chapter 2.0 A01 (SLs) 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SLs 2.1 2.1 SAFETY LIMITS 2.1.1 SLs 2.1.1 REACTOR CORE

.1 the highest operating loop 2.1.1.1 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and maxi-mum cold leg coolant temperature shall not exceed the limits shown on Figure 2.1-1. in

.1 In ,

2.1.1.1 APPLICABILITY: MODES 1 and 2.

2.1.1.2 2.1.1.2 In MODES 1 and 2, peak fuel centerline temperature shall be maintained within fuel melt ACTION: temperature limits specified in the COLR.

M01 When the point defined by the combination of maximum cold leg temper-ature and THERMAL POWER has exceeded the appropriate pressurizer pressure 2.2.1 line, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

MODE 3 SL 2.1.2 REACTOR COOLANT SYSTEM PRESSURE be maintained 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.

RCS In ,

2.1.2 APPLICABILITY

MODES 1, 2, 3, 4 and 5.

2.2 2.2 SAFETY LIMIT VIOLATIONS ACTION:

2.2.1 If SL 2.1.1 is violated, restore compliance and 2.2.2 If SL 2.1.2 is violated:

2.2.2.1 MODES 1 and 2 In or , restore compliance and Whenever the Reactor Coolant System pressure has exceeded 2750 psia, 2.2.2.1 be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . MODE 3 2.2.2.2 MODES 3, 4 and 5 In or , restore compliance Whenever the Reactor Coolant System pressure has exceeded 2750 psia, reduce the Reactor Coolant System pressure to within its limit 2.2.2.2 within 5 minutes.

ST. LUCIE - UNIT 1 2-1 Amendment No. 145, 151

ITS A01 ITS Chapter 2.0 LA01 A02 LA01 LA01 Figure 2.1.1-1 FIGURE 2.1-1: REACTOR CORE THERMAL MARGIN SAFETY LIMIT - A03

.1 FOUR REACTOR COOLANT PUMPS OPERATING (page 1 of 1)

ST. LUCIE - UNIT 1 2-2 Amendment No. 48, 130, 145, 163, 213

ITS Chapter 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS See 2.2 LIMITING SAFETY SYSTEM SETTINGS ITS 3.3.1 REACTOR TRIP SETPOINTS 2.2.1 The reactor protective instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY: AS SHOWN FOR EACH CHANNEL IN TABLE 3.3-1.

ACTION:

With a reactor protective instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement require-ment of Specification 3.3.1.1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

ST. LUCIE - UNIT 1 2-3

ITS Chapter 2.0 TABLE 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES See ITS 3.3.1

1. Manual Reactor Trip Not Applicable Not Applicable

2. Power Level - High (1)

Four Reactor Coolant Pumps Operating < 9.61% above THERMAL POWER, with < 9.61% above THERMAL POWER, and a minimum setpoint of 15% of RATED a minimum setpoint of 15% of RATED THERMAL POWER, and a maximum of THERMAL POWER and a maximum of

< 107.0% of RATED THERMAL < 107.0% of RATED THERMAL POWER. POWER.

3. Reactor Coolant Flow - Low (1)

Four Reactor Coolant Pumps Operating > 95% of minimum reactor coolant flow > 95% of minimum reactor coolant flow with 4 pumps operating

with 4 pumps operating *

4. Pressurizer Pressure - High < 2400 psia < 2400 psia

5. Containment Pressure - High < 3.3 psig < 3.3 psig

6. Steam Generator Pressure - Low (2) > 600 psia > 600 psia

7. Steam Generator Water Level - Low > 35.0% Water Level - each steam > 35.0% Water Level - each steam generator generator

8. Local Power Density - High (3) Trip setpoint adjusted to not exceed the Trip set point adjusted to not exceed the limit lines of Figures 2.2-1 and 2.2-2. limit lines of Figures 2.2-1 and 2.2-2.

See

For minimum reactor coolant flow with 4 pumps operating, refer to Technical Specification LCO 3.2.5. ITS 3.3.1 ST. LUCIE - UNIT 1 2-4 Amendment No. 3, 27, 32, 45, 105, 130, 145, 163, 213

ITS Chapter 2.0 See ITS 3.3.1 TABLE 2.2-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

9. Thermal Margin/Low Pressure (1)

Trip setpoint adjusted to not exceed the limit Trip setpoint adjusted to not exceed the limit Four Reactor Coolant Pumps Operating lines of Figures 2.2-3 and 2.2-4. lines of Figures 2.2-3 and 2.2-4 9a. Steam Generator Pressure Difference High (1) (logic < 135 psid < 135 psid in TM/LP)

10. Loss of Turbine - Hydraulic Fluid Pressure - Low (3) > 800 psig > 800 psig

11. Rate of Change of Power - High (4) < 2.49 decades per minute < 2.49 decades per minute TABLE NOTATION (1) Trip may be bypassed below 1% of RATED THERMAL POWER; bypass shall be automatically removed when Wide Range Logarithmic Neutron Flux power is > 1% of RATED THERMAL POWER.

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

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

(4) Trip may be bypassed below 10-4 % and above 15% of RATED THERMAL POWER.

ST. LUCIE - UNIT 1 2-5 Amendment No. 27, 43, 45, 159

ITS Chapter 2.0 See ITS 3.3.1 FIGURE 2.2-1 Local Power Density - High Trip Setpoint Part 1 (Fraction of RATED THERMAL POWER Versus QR2)

ST. LUCIE - UNIT 1 2-6

ITS Chapter 2.0 See ITS 3.3.1 FIGURE 2.2-2 Local Power Density - High Trip Setpoint Part 2 (QR2 Versus Y1)

ST. LUCIE - UNIT 1 2-7 Amendment No. 27, 32, 48, 106

ITS Chapter 2.0 See ITS 3.3.1 FIGURE 2.2-3 Thermal Margin/Low Pressure Trip Setpoint ST. LUCIE - UNIT 1 2-8 Amendment No. 27, 48

ITS Chapter 2.0 See ITS 3.3.1 FIGURE 2.2-4 Thermal Margin/Low Pressure Trip Setpoint Part 2 (Fraction of RATED THERMAL POWER Versus QR1)

ST. LUCIE - UNIT 1 2-9 Amendment No. 27, 48

ITS ITS Chapter 2.0 A01 (SLs) 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SLs 2.1 2.1 SAFETY LIMITS SLs 2.1.1 2.1.1 REACTOR CORE DNBR the highest operating loop 2.1.1.1 2.1.1.1 The combination of THERMAL POWER, pressurizer pressure, and maximum cold leg coolant temperature shall not exceed the limits shown on Figure 2.1-1.

In , in .1 2.1.1.1 APPLICABILITY: MODES 1 and 2.

2.1.1.2 2.1.1.2 In MODES 1 and 2, peak fuel centerline temperature shall be maintained below fuel melt M01 ACTION: temperature limits specified in the COLR.

Whenever the combination of THERMAL POWER, pressurizer pressure and maximum cold leg 2.2.1 coolant temperature has exceeded the limits shown on Figure 2.1-1, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . MODE 3 SL 2.1.2 REACTOR COOLANT SYSTEM PRESSURE be maintained 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.

RCS In ,

2.1.2 APPLICABILITY

MODES 1, 2, 3, 4 and 5.

2.2 2.2 SAFETY LIMIT VIOLATIONS ACTION: 2.2.1 If SL 2.1.1 is violated, restore compliance and 2.2.2 If SL 2.1.2 is violated:

2.2.2.1 MODES 1 and 2 In , restore compliance and MODE 3 or Whenever the Reactor Coolant System pressure has exceeded 2750 psia, be in HOT 2.2.2.1 STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2.2 MODES 3, 4 and 5 In or , restore compliance Whenever the Reactor Coolant System pressure has exceeded 2750 psia, reduce the 2.2.2.2 Reactor Coolant System pressure to within its limit within 5 minutes.

ST. LUCIE - UNIT 2 2-1 Amendment No. 8, 105, 146

ITS Chapter 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS See 2.2 LIMITING SAFETY SYSTEM SETTINGS ITS 3.3.1 REACTOR TRIP SETPOINTS 2.2.1 The reactor protective instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY: As shown for each channel in Table 3.3-1.

ACTION:

With a reactor protective instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

ST. LUCIE - UNIT 2 2-2

ITS A01 ITS Chapter 2.0 A03 LA01 FIGURE 2.1-1: REACTOR CORE THERMAL MARGIN SAFETY LIMIT LINES Figure 2.1.1-1 A03

.1 FOUR REACTOR COOLANT PUMPS OPERATING (page 1 of 1)

ST. LUCIE - UNIT 2 2-3 Amendment No. 8, 138, 163

ITS Chapter 2.0 TABLE 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES See ITS 3.3.1

1. Manual Reactor Trip Not Applicable Not Applicable

2. Variable Power Level - High(1)

Four Reactor Coolant Pumps < 9.61% above THERMAL POWER, < 9.61% above THERMAL POWER, and Operating with a minimum setpoint of a minimum setpoint of 15% of 15% of RATED THERMAL POWER, RATED THERMAL POWER and a maximum and a maximum of < 107.0% of of < 107.0% of RATED THERMAL POWER.

RATED THERMAL POWER.

3. Pressurizer Pressure - High < 2370 psia < 2374 psia

4. Thermal Margin/Low Pressure(1)

Four Reactor Coolant Pumps Trip setpoint adjusted to not Trip setpoint adjusted to not Operating exceed the limit lines of exceed the limit lines of Figures 2.2-3 and 2.2-4. Figures 2.2-3 and 2.2-4.

Minimum value of 1900 psia. Minimum value of 1900 psia.

5. Containment Pressure - High < 3.0 psig < 3.1 psig

6. Steam Generator Pressure - Low > 626.0 psia(2) (2) > 621.0 psia(2) (2)

7. Steam Generator Pressure(1) < 120.0 psid < 132.0 psid Difference - High (Logic in TM/LP Trip Unit)

8. Steam Generator Level - LowLow > 20.535.0%(3) (3) > 19.535.0%(3) (3)

ST. LUCIE - UNIT 2 2-4 Amendment No. 8, 23, 60, 163

ITS Chapter 2.0 TABLE 2.2-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES See

9. Local Power Density - High(5) Trip setpoint adjusted to Trip setpoint adjusted to ITS 3.3.1 Operating not exceed the limit lines not exceed the limit lines of Figures 2.2-1 and 2.2-2 of Figures 2.2-1 and 2.2-2.

10. Loss of Component Cooling Water > 636 gpm** > 636 gpm to Reactor Coolant Pumps - Low

11. Reactor Protection System Logic Not Applicable Not Applicable

12. Reactor Trip Breakers Not Applicable Not Applicable

13. Rate of Change of Power - High(4) < 2.49 decades per minute < 2.49 decades per minute

14. Reactor Coolant Flow - Low(1) > 95.4% of minimum Reactor > 94.9% of minimum Reactor Coolant flow with four Coolant flow with four pumps operating* pumps operating*

15. Loss of Load (Turbine) > 800 psig > 800 psig Hydraulic Fluid Pressure - Low(5)

For minimum reactor coolant flow with four pumps operating, refer to Technical Specification LCO 3.2.5.

- 10-minute time delay after relay actuation.

ST. LUCIE - UNIT 2 2-5 Amendment No. 8, 60, 131, 138, 163

ITS Chapter 2.0 TABLE 2.2-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS See ITS 3.3.1 TABLE NOTATION (1) Trip may be manually bypassed below 0.5% of RATED THERMAL POWER during testing pursuant to Special Test Exception 3.10.3; bypass shall be automatically removed when Wide Range Logarithmic Neutron Flux power is greater than or equal to 0.5% of RATED THERMAL POWER.

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

(3) % of the narrow range steam generator level indication.

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

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

ST. LUCIE - UNIT 2 2-6 Amendment No. 98

ITS Chapter 2.0 See ITS 3.3.1 FIGURE 2.2-1 Local power density - High trip setpoint Part 1 (Fraction of RATED THERMAL POWER versus QR2)

ST. LUCIE - UNIT 2 2-7

ITS Chapter 2.0 See ITS 3.3.1 FIGURE 2.2-2 LOCAL POWER DENSITY - HIGH TRIP SETPOINT PART 2 (QR2 versus Y1)

ST. LUCIE - UNIT 2 2-8 Amendment No. 42

ITS Chapter 2.0 See ITS 3.3.1 FIGURE 2.2-3 THERMAL MARGIN/LOW PRESSURE TRIP SETPOINT PART 1 (Y1 Versus A1)

ST. LUCIE - UNIT 2 2-9 Amendment No. 8, 138

ITS Chapter 2.0 See ITS 3.3.1 FIGURE 2.2-4 THERMAL MARGIN/LOW PRESSURE TRIP SETPOINT PART 2 (FRACTION OF RATED THERMAL POWER VERSUS QR1)

ST. LUCIE - UNIT 2 2-10 Amendment No. 8

DISCUSSION OF CHANGES ITS CHAPTER 2.0, SAFETY LIMITS (SLs)

ADMINISTRATIVE CHANGES A01 In the conversion of the St. Lucie Plant (PSL) Unit 1 and Unit 2, 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-1432, Rev. 5.0, "Standard Technical Specifications-Combustion Engineering Plants" (ISTS).

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

A02 Unit 1 only: CTS Figure 2.1-1 states "Reactor Coolant System total flow rate -

specified in LCO 3.2.5." ITS Figure 2.1.1-1 does not contain the statement. This changes the CTS by not including a cross reference to the equivalent reference ITS 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits," which contains the Reactor Coolant System total flowrate.

This change is acceptable because the CTS Figure 2.1-1 detail used to alert the user that the Reactor Coolant System total flowrate is covered under a separate LCO requirement is an unnecessary presentation preference. ITS does not include information only type of cross references to other Specifications. ITS Figure 2.1.1-1, Reactor Core Thermal Margin Safety Limit, is retained without the cross reference detail.

This change is designated as administrative as it incorporates an ITS convention with no technical change.

A03 PSL Unit 1 CTS Figure 2.1-1, Reactor Core Thermal Margin Safety Limit - Four Reactor Coolant Pumps Operating contains an operational limit line showing thermal power limited to a maximum of 112% of RATED THERMAL POWER by the High Power Trip. PSL Unit 2 CTS Figure 2.1-1, Reactor Core Thermal Margin Safety Limit - Four Reactor Coolant Pumps Operating contains an operational limit line showing thermal power limited to a maximum of 107% of RATED THERMAL POWER by the High Power Level Trip, and contains an operational limit line showing limits on maximum cold leg temperature due to secondary safety valve actuation. Additionally, the Unit 1 and Unit 2 figure titles include the statement "Four Reactor Coolant Pumps Operating."

The RPS operational limit line is removed from the Unit 1 and Unit 2 figures. This operational limit could be violated during a transient; however, it does not necessarily mean the reactor core SL was violated. Removing the operational limit line shifts the unacceptable area to the actual SL limits based on RCS pressure lines. The RPS operational limit is provided in the ITS Bases. The figure title text Four Reactor Coolant Pumps Operating is deleted from the Unit 1 and Unit 2 figures. Unit 1 and Unit 2 only operate with four reactor coolant pumps. Therefore, this text is deleted for consistency with ISTS Figure 2.1.1-1 title. The operational limit line showing limits on maximum cold leg temperature due to secondary safety valve actuation is removed from the Unit 2 figure. This operational limit could be violated during a transient; however, it does not necessarily mean the reactor core SL was violated. Removing this operational limit line shifts the unacceptable area to the actual SL limits based on RCS pressure lines.

St. Lucie Unit 1 and Unit 2 Page 1 of 3

DISCUSSION OF CHANGES ITS CHAPTER 2.0, SAFETY LIMITS (SLs)

This change is designated as administrative as it incorporates an ITS convention to remove superfluous lines not directly related to violation of the safety limit with no technical change.

MORE RESTRICTIVE CHANGES M01 CTS 2.1.1 does not contain a Safety Limit to ensure fuel centerline temperature remains below the fuel melt temperature during normal operating conditions or design anticipated operational occurrences with adjustments for burnup and burnable poison. ITS requires Reactor Core Safety Limit 2.1.1.2, to ensure fuel centerline temperature remains below the fuel melt temperature during normal operating conditions or design anticipated operational occurrences with adjustments for burnup and burnable poison.

The purpose of ITS 2.1.1.2 is to provide a Safety Limit to ensure fuel centerline temperature remains below the fuel melt temperature during normal operating conditions or design anticipated operational occurrences with adjustments for burnup and burnable poison. This change is acceptable because it adds a Reactor Core Safety Limit that is not in the CTS.

This change is designated as more restrictive because it adds a Reactor Core Safety Limit to the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 2.1.1 states "The combination of THERMAL POWER, pressurizer pressure, and maximum cold leg coolant temperature shall not exceed the limits shown on Figure 2.1-1." ITS states "The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop cold leg coolant temperature shall not exceed the limits shown in Figure 2.1.1-1."

This changes the CTS by moving details shown on the CTS Figure 2.1-1 to the ITS Bases.

PSL Unit 1 CTS Figure 2.1-1 statements "Reactor operation limited to less than 580°F by actuation of Main Steam Safety Valves," "Limits contain no allowance for instruments error or fluctuations," "Based on the Axial Shape Index on Figure B 2.1-1," and "THERMAL POWER limited to a maximum of 112% of RATED THERMAL POWER by the high power trip level" are moved from the Technical Specifications to the ITS Bases. The line showing THERMAL POWER at 112% is deleted. The Reactor Core Safety Limits are retained in Technical Specification Chapter 2.0.

St. Lucie Unit 1 and Unit 2 Page 2 of 3

DISCUSSION OF CHANGES ITS CHAPTER 2.0, SAFETY LIMITS (SLs)

PSL Unit 2 CTS Figure 2.1-1 statements "Reactor Conditions limited to less than the temperatures defined by the actuation of the secondary safety valves,"

"Limits contain no allowance for instruments error or fluctuations," "Valid for Total Integrated Radial Peaking Factors less than specified in Tech Spec 3.2.3" and "THERMAL POWER limited to a maximum of 107% of RATED THERMAL POWER by the high power level trip" are moved from the Technical Specifications to the ITS Bases. The Reactor Core Safety Limits are retained in Technical Specification Chapter 2.0.

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. ITS 2.1.1.1 retains the Reactor Core Safety Limit and Figure 2.1.1-1, Reactor Core Thermal Margin Safety Limit.

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 None St. Lucie Unit 1 and Unit 2 Page 3 of 3

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) 186

CTS SLs (Analog) 3 2.0 2.0 SAFETY LIMITS (SLs) (Analog) 2.1 2.1 SLs 2.1.1 2.1.1 Reactor Core SLs 2.1.1 and 2.1.1.1 In MODES 1 and 2, the combination of THERMAL POWER, Applicability pressurizer pressure, and the highest operating loop cold leg coolant temperature shall not exceed the limits shown in Figure 2.1.1-1.

DOC M01 2.1.1.2 In MODES 1 and 2, peak fuel centerline temperature shall be maintained at < [5080]°F, decreasing by [58°F per 10,000 MWD/MTU] 2 and adjusted for burnable poison per [CENPD-275-P, Revision 1-P-A or CENPD-382-P-A].

below fuel melt temperature limits specified in the COLR 2.1.2 2.1.2 Reactor Coolant System Pressure SL 2.1.2 and In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained [2750] psia.

Applicability 2.2 SAFETY LIMIT VIOLATIONS 2.1.1 ACTION 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.1.2 ACTION 2.2.2 If SL 2.1.2 is violated:

2.1.2 ACTION 2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.1.2 ACTION 2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

Combustion Engineering STS 2.0-1 Rev. 5.0 1 ST. LUCIE - UNIT 1 Amendment XXX

SLs (Analog) 3 2.0 1

Insert 1 1 Combustion Engineering STS 2.0-2 Rev. 5.0 1 ST. LUCIE - UNIT 1 Amendment XXX

Section 2.0 1

INSERT 1 Unit 1 Only Insert Page 2.0-2

CTS SLs (Analog) 3 2.0 2.0 SAFETY LIMITS (SLs) (Analog) 2.1 2.1 SLs 2.1.1 2.1.1 Reactor Core SLs 2.1.1 and 2.1.1.1 In MODES 1 and 2, the combination of THERMAL POWER, Applicability pressurizer pressure, and the highest operating loop cold leg coolant temperature shall not exceed the limits shown in Figure 2.1.1-1.

DOC M01 2.1.1.2 In MODES 1 and 2, peak fuel centerline temperature shall be maintained at < [5080]°F, decreasing by [58°F per 10,000 MWD/MTU] 2 and adjusted for burnable poison per [CENPD-275-P, Revision 1-P-A or CENPD-382-P-A].

below fuel melt temperature limits specified in the COLR 2.1.2 2.1.2 Reactor Coolant System Pressure SL 2.1.2 and In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained [2750] psia.

Applicability 2.2 SAFETY LIMIT VIOLATIONS 2.1.1 ACTION 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.1.2 ACTION 2.2.2 If SL 2.1.2 is violated:

2.1.2 ACTION 2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.1.2 ACTION 2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

Combustion Engineering STS 2.0-1 Rev. 5.0 1 ST. LUCIE - UNIT 2 Amendment XXX

SLs (Analog) 3 2.0 1

Insert 1 1 Combustion Engineering STS 2.0-2 Rev. 5.0 1 ST. LUCIE - UNIT 2 Amendment XXX

Section 2.0 1

INSERT 1 Unit 2 Only Insert Page 2.0-2

JUSTIFICATION FOR DEVIATIONS ISTS CHAPTER 2.0, SAFETY LIMITS (SLs)

1. 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.

2. The ISTS contains bracketed information and/or values that are generic to all Combustion Engineering vintage plants. The brackets are removed, and the proper plant specific information/value is provided. This is acceptable since the information/value is changed to reflect the current licensing basis. PSL does not currently require an SL regarding peak fuel centerline temperature. However, peak fuel centerline temperature is maintained below the fuel melt temperature specified in the COLR. The fuel melt temperature is determined using the NRC approved methodology in Siemens Topical Report XN-NF-79-56(P)(A) Revision 1 and Supplement 1, Gadolinia Fuel Properties for LWR Fuel Safety Evaluation, dated October 1981.

3. The type of plant (Analog) is deleted since it is unnecessary. This information is provided in NUREG-1432, Rev. 5.0, to assist in identifying the appropriate Specification to be used as a model for the plant specific ITS conversion but serves no purpose in a plant specific implementation.

St. Lucie Unit 1 and Unit 2 Page 1 of 1

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

RCS Core SLs (Analog) 5 B 2.1.1 B 2.0 SAFETY LIMITS (SLs) (Analog) 5 B 2.1.1 Reactor Core SLs (Analog)

BASES BACKGROUND GDC 10 (Ref. 1) requires and SLs ensure that specified acceptable fuel design limits are not exceeded during steady state operation, normal operation transients, and anticipated operational occurrences (AOOs).

This is accomplished by having a departure from nucleate boiling (DNB) design basis, which corresponds to a 95% probability at a 95%

confidence level (95/95 DNB criterion) that DNB will not occur and by requiring that fuel centerline temperature stays below the melting temperature.

The restrictions of this SL prevent overheating of the fuel and cladding and possible cladding perforation that would result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate (LHR) below the level at which fuel centerline melting occurs. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime, where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant.

Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of DNB and the resultant sharp reduction in heat transfer coefficient. Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

The Reactor Protective System (RPS), in combination with the LCOs, is designed to prevent any anticipated combination of transient conditions for Reactor Coolant System (RCS) temperature, pressure, and THERMAL POWER level that would result in a violation of the reactor core SLs.

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RCS Core SLs (Analog) 5 B 2.1.1 BASES APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY operation and AOOs. The reactor core SLs are established to ANALYSES preclude violation of the following fuel design criteria:

a. There must be at least 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience a DNB and

b. The hot fuel pellet in the core must not experience fuel centerline melting.

The RPS setpoints, LCO 3.3.1, "Reactor Protective System (RPS)

Instrumentation," in combination with all the LCOs, are designed to prevent any anticipated combination of transient conditions for RCS temperature, pressure, and THERMAL POWER level that would result in a departure from nucleate boiling ratio (DNBR) of less than the DNBR limit and preclude the existence of flow instabilities.

Automatic enforcement of these reactor core SLs is provided by the following functions:

a. Pressurizer Pressure High trip,

b. Variable High Power trip, 1 Power Level - High

c. Power Rate of Change - High trip,

d. Reactor Coolant Flow - Low trip,

e. Steam Generator Pressure - Low trip,

f. Steam Generator Level - Low trip,

g. Axial Power Distribution - High trip, 1

Local Power Density

h. Thermal Margin/Low Pressure trip,

i. Steam Generator Pressure Difference trip, and

j. Steam Generator Safety Valves.

The SL represents a design requirement for establishing the RPS trip setpoints identified previously. LCO 3.2.1, "Linear Heat Rate (LHR)," and LCO 3.2.5, "AXIAL SHAPE INDEX (ASI)," or the assumed initial 1 conditions of the safety analyses (as indicated in the FSAR, Ref. 2) provide more restrictive limits to ensure that the SLs are not exceeded.

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RCS Core SLs (Analog) 5 B 2.1.1 The SL contains no allowance for instrument error. Actuation of the main 4

BASES steam safety valves limits reactor coolant temperature.

SAFETY LIMITS The curves provided in Figure B 2.1.1-1 show the loci of points of THERMAL POWER, pressurizer pressure, and highest operating loop cold leg temperature, for which the minimum DNBR is not less than the 3 safety analysis limit. SL 2.1.1.2 ensures that fuel centerline temperature remains below melting.

SL 2.1.1.2 ensures that fuel centerline temperature remains below the limits specified in the COLR fuel melt temperature of [5080]F during normal operating conditions or 2 design AOOs with adjustments for burnup and burnable poison. An adjustment of [58F per 10,000 MWD/MTU] has been established in

[Topical Report CEN-386-P-A] (Ref. 3) and adjustments for burnable 2

poisons are established based on [Topical Reports CENPD-275-P] (Ref.

4) and [CENPD-382-P-A] (Ref.5).

APPLICABILITY SL 2.1.1 only applies in MODES 1 and 2 because these are the only MODES in which the reactor is critical. Automatic protection functions are required to be OPERABLE during MODES 1 and 2 to ensure operation within the reactor core SLs. The steam generator safety valves or automatic protection actions serve to prevent RCS heatup to the reactor core SL conditions or to initiate a reactor trip function, which forces the 3 unit into MODE 3. Setpoints for the reactor trip functions are specified in LCO 3.3.1.

In MODES 3, 4, 5, and 6, Applicability is not required, since the reactor is not generating significant THERMAL POWER.

SAFETY LIMIT The following SL violation responses are applicable to the reactor core VIOLATIONS SLs.

2.2.1 If SL 2.1.1 is violated, the requirement to go to MODE 3 places the unit in a MODE in which this SL is not applicable.

The allowed Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recognizes the importance of bringing the unit to a MODE of operation where this SL is not applicable and reduces the probability of fuel damage.

Fuel melt temperature is determined using the NRC approved methodology in Reference 3. The fuel melt temperature limit is 2 defined as [(2790 - 17.9 x P - 3.2 x B) x 1.8 + 32] °F, where P is the maximum weight percent of Gadolinia (%) and B is the maximum pin burnup (GWD/MTU).

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RCS Core SLs (Analog) 5 B 2.1.1 BASES REFERENCES 1. 10 CFR 50, Appendix A, GDC 10, 1988.

2. FSAR, Section [7.2 ]. 2 U

3. [Topical Report CEN-386-P-A, Verification of the Acceptability of a 2

1-Pin Burnup Limit of 60 MWD/kgU for Combustion Engineering 16x16 PWR Fuel, August 1992.]

4. [Topical Report CENPD-275-P, Revision 1-P-A, CE Methodology for 2 Core Designs Containing Gadolini-Urania Burnable Absorbers, May 1988.]

5. [Topical Report CENPD-382-P- A, Methodology for Core Designs 2 Containing Erbium Burnable Absorbers, August 1993.]

3. Siemens Power Corporation Topical Report XN-NF-79-56(P)(A) Revision 1 and Supplement 1, Gadolinia Fuel Properties for LWR Fuel Safety Evaluation, October 1981. 2 Combustion Engineering STS B 2.1.1-4 Rev. 5.0 1

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RCS Core SLs (Analog) 5 B 2.1.1 1

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RCS Pressure SL (Analog) 5 B 2.1.2 B 2.0 SAFETY LIMITS (SLs) (Analog) 5 B 2.1.2 Reactor Coolant System (RCS) Pressure SL (Analog)

BASES BACKGROUND The SL on RCS pressure protects the integrity of the RCS against overpressurization. In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. By establishing an upper limit on RCS pressure, continued RCS integrity is ensured. According to 10 CFR 50, Appendix A, GDC 14, "Reactor Coolant Pressure Boundary," and GDC 15, "Reactor Coolant System Design" (Ref. 1), the reactor coolant pressure boundary (RCPB) design conditions are not to be exceeded during normal operation and anticipated operational occurrences (AOOs). Also, according to GDC 28 (Ref. 1), "Reactivity Limits," reactivity accidents, including rod ejection, do not result in damage to the RCPB greater than limited local yielding.

The design pressure of the RCS is 2500 psia. During normal operation and AOOs, the RCS pressure is kept from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code (Ref. 2). To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, according to the ASME Code requirements prior to initial operation, when there is no fuel in the core. Following inception of unit operation, RCS components shall be pressure tested, in accordance with the requirements of ASME Code,Section XI (Ref. 3).

Overpressurization of the RCS could result in a breach of the RCPB. If this occurs in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere, raising concerns relative to limits on radioactive releases specified in 10 CFR 100, "Reactor Site Criteria" (Ref. 4).

APPLICABLE The RCS pressurizer safety valves, the main steam safety valves SAFETY (MSSVs), and the Reactor Pressure - High trip have settings established ANALYSES to ensure that the RCS pressure SL will not be exceeded.

The RCS pressurizer safety valves are sized to prevent system pressure from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code for Nuclear Power Plant Components (Ref. 2). The transient that establishes the required relief capacity, and Combustion Engineering STS B 2.1.2-1 Rev. 5.0 1

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RCS Pressure SL (Analog) 5 B 2.1.2 BASES APPLICABLE SAFETY ANALYSES (continued) hence the valve size requirements and lift settings, is a [complete loss of turbine-generator external load without a direct reactor trip]. During the transient, no control 2 actions are assumed except that the safety valves on the secondary plant are assumed to open when the steam pressure reaches the secondary plant safety valve settings, and nominal feedwater supply is maintained.

The Reactor Protective System (RPS) trip setpoints (LCO 3.3.1, "Reactor Protective System (RPS) Instrumentation"), together with the settings of the MSSVs (LCO 3.7.1, "Main Steam Safety Valves (MSSVs)") and the pressurizer safety valves, provide pressure protection for normal operation and AOOs. In particular, the Pressurizer Pressure - High trip setpoint is specifically set to provide protection against overpressurization (Ref. 5). Safety analyses for both the Pressure - High trip and the RCS pressurizer safety valves are performed, using conservative assumptions relative to pressure control devices.

More specifically, no credit is taken for operation of any of the following:

a. Pressurizer power operated relief valves (PORVs),

b. Steam Bypass Control System,

c. Pressurizer Level Control System, or

d. Pressurizer Pressure Control System.

SAFETY LIMITS The maximum transient pressure allowable in the RCS pressure vessel under the ASME Code,Section III, is 110% of design pressure. The maximum transient pressure allowable in the RCS piping, valves, and fittings under [USAS, Section B31.1 (Ref. 6)], is 120% of design pressure. 2 The most limiting of these two allowances is the 110% of design pressure; therefore, the SL on maximum allowable RCS pressure is established at 2750 psia.

APPLICABILITY SL 2.1.2 applies in MODES 1, 2, 3, 4, and 5 because this SL could be approached or exceeded in these MODES due to overpressurization events. The SL is not applicable in MODE 6 because the reactor vessel head closure bolts are not fully tightened, making it unlikely that the RCS can be pressurized.

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RCS Pressure SL (Analog) 5 B 2.1.2 BASES SAFETY LIMIT The following SL violation responses are applicable to the RCS pressure VIOLATIONS SL.

2.2.2.1 If the RCS pressure SL is violated when the reactor is in MODE 1 or 2, the requirement is to restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

With RCS pressure greater than the value specified in SL 2.1.2 in MODE 1 or 2, the pressure must be reduced to below this value. A pressure greater than the value specified in SL 2.1.2 exceeds 110% of the RCS design pressure and may challenge system integrity.

The allowed Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provides the operator time to complete the necessary actions to reduce RCS pressure by terminating the cause of the pressure increase, removing mass or energy from the RCS, or a combination of these actions, and to establish MODE 3 conditions.

2.2.2.2 If the RCS pressure SL is exceeded in MODE 3, 4, or 5, RCS pressure must be restored to within the SL value within 5 minutes.

Exceeding the RCS pressure SL in MODE 3, 4, or 5 is potentially more severe than exceeding this SL in MODE 1 or 2, since the reactor vessel temperature may be lower and the vessel material, consequently, less ductile. As such, pressure must be reduced to less than the SL within 5 minutes. This action does not require reducing MODES, since this would require reducing temperature, which would compound the problem by adding thermal gradient stresses to the existing pressure stress.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC 28.

2. ASME, Boiler and Pressure Vessel Code,Section III, Article NB-7000. 1

3. ASME, Boiler and Pressure Vessel Code,Section XI, Article IWX-5000.

4. 10 CFR 100.

5. FSAR, Section [15.2.7]. 2 U

[ 6. ASME, USAS B31.1, Standard Code for Pressure Piping, 1967. ] 2 Combustion Engineering STS B 2.1.2-3 Rev. 5.0 ST. LUCIE - UNIT 1 Revision XXX