ML23151A442

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Enclosure 2, Volume 4: Turkey Point Nuclear Plant, Units 3 and 4, Improved Technical Specifications Conversion, Chapter 2, Safety Limits, Revision 1
ML23151A442
Person / Time
Site: Turkey Point  NextEra Energy icon.png
Issue date: 05/24/2023
From:
Florida Power & Light Co
To:
Office of Nuclear Reactor Regulation
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ML23151A435 List:
References
L-2023-063
Download: ML23151A442 (1)
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Text

ENCLOSURE 2

VOLUME 4

TURKEY POINT NUCLEAR GENERATI NG STATION UNIT 3 AND UNIT 4

IMPROVED TECHNICAL SPECIFICATIONS CONVERSION

ITS CHAPTER 2.0 SAFETY LIMITS (SLs)

Revision 1

LIST OF ATTACHMENTS

1. ITS Chapter 2.0 - Safety Limits (SLs)

ATTACHMENT 1

ITS Chapter 2.0 - SAFETY LIMITS (SLs)

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

ITS Chapter 2.0

SECTION 2.0

SAFETY LIMITS

AND A01

LIMITING SAFETY SYSTEM SETTINGS

TURKEY POINT - UNITS 3 & 4 2-0 AMENDMENT NOS. 137 AND 132 Page 1 of 10 ITS Chapter 2.0

A01 ITS 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

2.1 2.1 SAFETY LIMITS SLs

2.1.1 REACTOR CORE

2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tavg) shall not exceed the limits specified in the COLR, for 3 loop operation; and the following S afety Limits shall not be exceeded: ReactoolaystRCS) hitp In MODES 1 and 2, aper and essiz prsure 2.1.1.1 a. The departure from nucleate boiling ratio (DNBR) shall be maintained > 1.17 for the WRB-1 DNB correlation.

2.1.1.2

b. The peak fuel centerline temperature shall be maintained < 5080°F, decreasing by 9°F per AFETY LIMTIONS 10,000 MWD/MTU of burnup.

APPLICABILITY: MODES 1 and 2, If1 isietciancn MODE 3 ACTION:

1 Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. A01 be maintained

2.1.2 REACTOR COOLANT SYSTEM PRESSURE

2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

If 2 ted ACTION:

2 In MODE 1 or 2,te ciae ae in MODE MODES 1 and 2:

1 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

MODES 3, 4 and 5:

2 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

In MODE 3, 4, 5,te comia

TURKEY POINT - UNIT 3 & 4 2-1 AMENDMENT NOS. 288 AND 282 P 2 ITS Chapter 2.0

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

2.2 LIMITING SAFETY SYSTEM SETTINGS

REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS

2.2.1 The Reactor Trip System Instrumentation and Interlock Setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

See ITS APPLICABILITY: As shown for each channel in Table 3.3-1. 3.3.1

Action:

a. With a Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 2.2-1, adjust the setpoint consistent with the Trip setpoint value within permissible calibration tolerance.
b. With the Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, either:
1. Adjust the Setpoint consistent with the Trip Setpoint value of Table 2.2-1 and determine within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that the affected channel is OPERABLE; or
2. 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 setpoint adjusted consistent with the Trip Setpoint value.

TURKEY POINT - UNITS 3 & 4 2-2 AMENDMENT NOS. 260 AND 255 Page 3 of 10 ITS Chapter 2.0

TABLE 2.2-1

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

FUNCTIONAL UNIT ALLOWABLE VALUE TRIP SETPOINT

1. Manual Reactor Trip N.A. N.A. See ITS 3.3.1
2. Power Range, Neutron Flux
a. High Setpoint 108.6 % of RTP** 108.0% of RTP**
b. Low Setpoint 28.0% of RTP** 25% of RTP**
3. Intermediate Range, Neutron Flux 31.0% of RTP** d 25% of RTP**
4. Source Range, Neutron Flux d 1.4 X 105 cps d 105 cps
5. Overtemperature ' T See Note 2 See Note 1
6. Overpower 'T See Note 4 See Note 3
7. Pressurizer Pressure-Low t 1817 psig t 1835 psig
8. Pressurizer Pressure-High d 2403 psig d 2385 psig
9. Pressurizer Water Level-High d 92.2% of instrument span d 92% of instrument span
10. Reactor Coolant Flow-Low 89.6% of loop design flow* 90% of loop design flow*
11. Steam Generator Water Level Low-Low 15.5% narr rang 16% rrow rge irumen sp irumen sp
  • Loop design flow = 86,900 gpm
    • RTP = Rated Thermal Power

Page 4 of 10 ITS Chapter 2.0

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

FUNCTIONAL UNIT ALLOWABLE VALUE TRIP SETPOINT

12. Steam/Feedwater Flow Mismatch Feed Flow 20.7% l Feed Flow 20% below rated Coincident with re Seam Fl Steam Flow See ITS Steam Generator Water Level-Low 15.5% of narrow range 16% of narrow range 3.3.1 instrument span instrument span
13. Undervoltage - 4.16 kV Busses A and B % s volage % s volage
14. Ufre°uy Tri Reaor Coa 55.9 Hz 56.1 Hz PBrker(s) Open
15. Turbine Trip
a. Emergency Trip Header Pressure 901 psig 1000 psig
b. Turbine Stop Valve Closure Fully Closed*** Fully Closed ***
16. Safety Injection Input from ESF N.A. N.A.
17. Reactor Trip System Interlocks
a. Intermediate Range Neutron Flux, P -6 6.0 X 10 -11 amps Nominal 1 X 10 -10 amps
      • Limit switch is set when Turbine Stop Valves are fully closed.

Page 5 of 10 ITS Chapter 2.0

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

FUNCTIONAL UNIT ALLOWABLE VALUE TRIP SETPOINT See ITS

b. Low Power Reactor Trips Block, P-7 3.3.1
1) P-10 input 13.0% RTP** Nominal 10% of RTP**
2) Turbine Inlet Pressure 13.0% Turbine Power Nominal 10% Turbine Power
c. Power Range Neutron Flux, P-8 48.0% RTP** Nominal 45% of RTP**
d. Power Range Neutron Flux, P-10 7.0% RTP** Nominal 10% of RTP**
18. Reactor Coolant Pump Breaker Position N.A. N.A.

Trip

19. Reactor Trip Breakers N.A. N.A.
20. Automatic Trip and Interlock Logic N.A. N.A.
    • RTP = RATED THERMAL POWER

Page 6 of 10 ITS Chapter 2.0

TABLE 2.2-1 (Continued)

TABLE NOTATIONS

NOTE 1: OVERTEMPERATURE T (Those values denoted with [*] are specified in the COLR.)

(1+ 1S) 1 (1+ 4S) 1 T T0 {K1 - K 2 [T - T' ] + K3(P - P' ) - f 1(I)}

(1+ 2S) 1 + S (1+ S) (1 + 6S) 3 5

See ITS Where: T = Measured T by RTD Instrumentation 3.3.1

1 + 1S = Lead/Lag compensator on measured T; 1 = [*]s, 2 = [*]s 1 + 2S

1 1 + = Lag compensator on measured T; 3 = [*]s 3S

T0 = Indicated T at RATED THERMAL POWER

K1 = [*];

K2 = [*]/°F;

1+ 4S = The function generated by the lead-lag compensator for T avg dynamic compensation; 1+ 5S

4, 5 = Time constants utilized in the lead-lag compensator for Tavg, 4 = [*]s, 5 = [*]s;

T = Average temperature, °F;

1 1 + = Lag compensator on measured Tavg; 6 = [*]s 6S

T [*]°F (Indicated Loop T avg at RATED THERMAL POWER);

K3 = [*]/psi;

P = Pressurizer pressure, psig;

Page 7 of 10 ITS Chapter 2.0

TABLE 2.2-1 (Continued)

TABLE NOTATIONS (Continued)

NOTE 1: (Continued) See ITS 3.3.1 P [*] psig (Nominal RCS operating pressure);

S = Laplace transform operator, s-1;

And f1(I) is a function of the indicated difference between top and bottom detectors of the power -range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(1) For qt - q b between - [ *]% and + [*]%, f1(I) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + qb is total THERMAL POWER in percent of RATED THERMAL POWER;

(2) For each percent that the magnitude of qt - q b exceeds - [ *]%, the T Trip Setpoint shall be automatically reduced by [*]% of its value at RATED THERMAL POWER; and

(3) For each percent that the magnitude of qt - q b exceeds + [*]%, the T Trip Setpoint shall be automatically reduced by [*]% of its value at RATED THERMAL POWER.

NOTE 2: The Overtemperature T function Allowable Value shall not exceed the nominal trip setpoint by more than 0.5% T span for the T channel, 0.2% T span for the Pressurizer Pressure channel, and 0.4% T span for the f(I) channel. No separate Allowable Value is provided for Tavg because this function is part of the T value.

Page 8 of 10 ITS Chapter 2.0

TABLE 2.2-1 (Continued)

TABLE NOTATIONS (Continued)

NOTE 3: OVERPOWER T (Those values denoted with [*] are specified in the COLR.)

(1+ 1S) 1 7S 1 1 T T0 {K4 - K 5 T - K 6 [T - T ] - f2(I)}

(1+ 2S) 1 + S 1+ 7S 1 + S 1+ 6S See ITS 3 6 3.3.1

Where: T = As defined in Note 1,

1+ 1S = As defined in Note 1, 1+ 2S

1 1 + = As defined in Note 1, 3S

T0 = As defined in Note 1,

K4 = [*],

K5 [*]/°F for increasing average temperature and [*]/°F for decreasing average temperature,

7s -lag compensator for T avg dynamic compensation; 1+ = The function generated by the lead7S

7 = Time constants utilized in the lead-lag compensator for Tavg, 7 [*] s,

1 1 + = As defined in Note 1, 6S

Page 9 of 10 ITS Chapter 2.0

TABLE 2.2-1 (Continued)

TABLE NOTATIONS (Continued)

See ITS NOTE 3: (Continued) 3.3.1

K6 = [*]/°F for T > T

= [*] for T T,

T = As defined in Note 1,

T [*]°F (Indicated Loop T avg at RATED THERMAL POWER)

S = As defined in Note 1, and

f2 (I) = [*]

NOTE 4: The Overpower T function Allowable Value shall not exceed the nominal trip setpoint by more than 0.5% T span for the T channel.

No separate Allowable Value is provided for Tavg because this function is part of the T value.

Page 10 of 10 DISCUSSION OF CHANGES ITS CHAPTER 2.0, SAFETY LIMITS (SL s)

ADMINISTRATIVE CHANGES

A01 In the conversion of the Turkey Point Nuclear Generating Station (PTN ) Unit 3 and Unit 4, 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

None

Turkey Point Unit 3 and Unit 4 Page 1 of 1 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

CTS SLs 2.0

2.0SAFETY LIMITS (SLs)

2.1 2.1 SLs

2.1.1, 2.1.1 Reactor Core SLs

2.1.1 Applicability In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in the COLR; and the following SLs shall not be exceeded:

2.1.1.a 2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained 1

[1.17 for the WRB-1/WRB-2 DNB correlations]. 2

2.1.1.b 2.1.1.2 The peak fuel centerline temperature shall be maintained < [5080°F, decreasing by 58°F per 10,000 MWD/MTU of burnup]. 2 9

2.1.2 2.1.2 Reactor Coolant System Pressure SL 2.1.2 Applicability In MODES 1, 2, 3, 4, and5, the RCS pressure shall be maintained [2735] psig. 2

2.2 SAFETY LIMIT VIOLATIONS

2.1.1 Action 2.2.1 If SL2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.1.2 Action 2.2.2 If SL 2.1.2 is violated:

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

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

Turkey Point Unit 3 and Unit 4 Amendment Nos. XXX and YYY

Westinghouse STS 2.0-1 Rev. 5.0 1 JUSTIFICATION FOR DEVIATIONS ITS CHAPTER 2.0, SAFETY LIMITS (SL s)

1. Changes are made (additions, deletions, and/or changes) to the Improved Standard Technical Specifications (ISTS ) that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
2. T he 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 provided. This is acceptable since the information/value is changed 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)

Reactor Core SLs B 2.1.1

B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core The 1967 proposed draft GDC 6

BASES the minimum

BACKGROUND GDC 10 (Ref.1) requires that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs). This is accomplished by having a departure from nucleate boiling (DNB) design 1 basis, which corresponds to a 95% probability at a 95% confidence level (the 95/95 DNB criterion) that DNB will not occur and by requiring that fuel centerline temperature stays below the melting temperature.

(Ref. 1)

The restrictions of this SL prevent overheating of the fuel and cladding, as well as 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) 1 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 1 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 proper functioning of the Reactor Protection System (RPS) and steam generator safety valves prevents violation of the reactor core SLs.

Turkey Point Unit 3 and Unit 4 Revision XXX

Westinghouse STS B 2.1.1-1 Rev. 5.0 1 Reactor Core SLs 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 preclude ANALYSES 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 DNB and
b. The hot fuel pellet in the core must not experience centerline fuel melting.

The Reactor Trip System setpoints (Ref. 2), in combination with all the LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System (RCS) temperature, pressure, RCS Flow, I, 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 appropriate operation of the RPS and the steam generator safety valves.

The SLs represent a design requirement for establishing the RPS trip setpoints identified previously. LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits," or the assumed initial conditions of the safety analyses (as indicated in the FSAR, Ref. 2) provide more restrictive limits to ensure that the SLs are not exceeded.

SAFETY LIMITS The figure provided in the COLR shows the loci of points of THERMAL POWER, RCS pressure, and average temperature for which the minimum DNBR is not less than the safety analyses limit, that fuel centerline temperature remains below melting, that the average enthalpy in the hot leg is less than or equal to the enthalpy of saturated liquid, or that the exit quality is within the limits defined by the DNBR correlation.

The reactor core SLs are established to preclude violation of the following fuel design criteria:

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

Turkey Point Unit 3 and Unit 4 Revision XXX

Westinghouse STS B 2.1.1-2 Rev. 5.0 1 Reactor Core SLs B 2.1.1

BASES

SAFETY LIMITS (continued)

The reactor core SLs are used to define the various RPS functions such that the above criteria are satisfied during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs).

To ensure that the RPS precludes the violation of the above criteria, additional criteria are applied to the Overtemperature and Overpower T reactor trip functions. That is, it must be demonstrated that the average enthalpy in the hot leg is less than or equal to the saturation enthalpy and that the core exit quality is within the limits defined by the DNBR correlation. Appropriate functioning of the RPS ensures that for variations in the THERMAL POWER, RCS Pressure, RCS average temperature, RCS flow rate, and I that the reactor core SLs will be satisfied during steady state operation, normal operational transients, and AOOs.

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 unit into MODE 3. Setpoints for the reactor trip functions are specified in LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation." 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. If SL2.1.1 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 <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10. 1

2. FSAR, Section [ 7.2]. 1 2

UFSAR, Sections 3.1.2 and 3.1.3.

U

Turkey Point Unit 3 and Unit 4 Revision XXX

Westinghouse STS B 2.1.1-3 Rev. 5.0 1 RCS Pressure SL B 2.1.2

B 2.0 SAFETY LIMITS (SLs)

B 2.1.2 Reactor Coolant System (RCS) Pressure SL 1967 Proposed GDC 32, Maximum Reactivity Worth of Control Rods and 1967 Proposed GDC 33, Reactor Coolant Pressure Boundary Capability 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 1967 Proposed GDC 9 atmosphere. By establishing an upper limit on RCS pressure, the continued integrity of the RCS is ensured. According to 10 CFR 50, Appendix A, GDC 14, "Reactor Coolant Pressure Boundary," and 1 1967 Proposed GDC 4 GDC 15, "Reactor Coolant System Design" (Ref. 1), the reactor pressure coolant boundary (RCPB) design conditions are not to be exceeded during normal operation and anticipated operational occurrences (AOOs).

Also, in accordance with GDC 28, "Reactivity Limits" (Ref.1), 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, RCS pressure is limited 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 such a breach 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 high pressure 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%, as specified in Section III of the ASME Code for Nuclear Power Plant Components (Ref.2). The transient that establishes the required relief capacity, and hence valve size requirements and lift settings, is a complete loss of

Turkey Point Unit 3 and Unit 4 Revision XXX

Westinghouse STS B 2.1.2-1 Rev. 5.0 1 RCS Pressure SL B 2.1.2

BASES

APPLICABLE SAFETY ANALYSES (continued)

external load without a direct reactor trip. During the transient, no control 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 Trip System setpoints (Ref. 5), together with the settings of the MSSVs, provide pressure protection for normal operation and AOOs.

The reactor high pressure trip setpoint is specifically set to provide protection against overpressurization (Ref. 5). The safety analyses for both the high pressure 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 line relief valve,
c. Steam Dump System,
d. Reactor Control System,
e. Pressurizer Level Control System, or
f. Pressurizer spray valve.

SAFETY LIMITS The maximum transient pressure allowed in the RCS pressure vessel under the ASME Code,Section III, is 110% of design pressure. The maximum transient pressure allowed 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 2735 psig.

Turkey Point Unit 3 and Unit 4 Revision XXX

Westinghouse STS B 2.1.2-2 Rev. 5.0 1 RCS Pressure SL B 2.1.2

BASES

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.

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

Exceeding the RCS pressure SL may cause immediate RCS failure and create a potential for radioactive releases in excess of 10 CFR 100, "Reactor Site Criteria," limits (Ref. 4).

The allowable Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> recognizes the importance of reducing power level to a MODE of operation where the potential for challenges to safety systems is minimized.

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 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. The 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. 1

2. ASME, Boiler and Pressure Vessel Code,Section III, Article NB-7000.
3. ASME, Boiler and Pressure Vessel Code,Section XI, Article IWX-5000.
4. 10 CFR 100.3.2 and U
5. FSAR, Section [7.2]. 1 2
6. USAS B31.1, Standard Code for Pressure Piping, American Society of Mechanical Engineers, 1967.

UFSAR 3.1.2, 4.1.2, and 4.1.3

Turkey Point Unit 3 and Unit 4 Revision XXX

Westinghouse STS B 2.1.2-3 Rev. 5.0 1 JUSTIFICATION FOR DEVIATIONS ITS CHAPTER 2.0 BASES, SAFETY LIMITS (SL s)

1. Changes are made (additions, deletions, and/or changes) to the Improved Standard Technical Specifications (ISTS ) Bases that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
2. T he ISTS Bases 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 provided. This is acceptable since the information/value is changed to reflect the current licensing basis.

Turkey Point Unit 3 and Unit 4 Page 1 of 1 Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 2.0, SAFETY LIMITS (SLs)

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

Turkey Point Unit 3 and Unit 4 Page 1 of 1

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