ML17334B595

Proposed Tech Specs,Supporting Operation of Facility at Increased Core Rated Thermal Power of 3,588 Mwt

ML17334B595
Person / Time
Site:	Cook
Issue date:	07/11/1996
From:	INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:
Shared Package
ML17333A496	List: ML17333A494 ML17334B594 ML17334B595
References
NUDOCS 9607150009
Download: ML17334B595 (220)

Text

PROPOSED CHANGES TO THE COOK NUCLEAR PLANT UNIT NO. 1 TECHNICAL SPECIFICATIONS VSO7i50009 9SO7ia PDR ADOCK 050003i5 P PDR

)f' 3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM SAFETY VALVES - SHUTDOWN UMITING CONDITION FOR OPERATION 3.4.2 A minimum of one pressurizer code safety valve shall be OPERABLE with a lift setting of 2485 PSIG +3%.

APPLICABILITY: MODES 4 and 5.

ACTION:

With no pressurizer code safety valve OPERABLE:

'8

a. Immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode.

b. Immediately render all Safety Injection pumps and all but one charging pump inoperable by removing the applicable motor circuit breakers from the electric power circuit within one hour.

SURVEILLANCE RE UIREMENTS 4.4.2 The pressurizer code safety valve shall be demonstrated OPERABLE per Surveillance Requirement 4.4.3.

The pressurizer code safety valve shall be reset to the nominal value J1% whenever found outside the

+1% tolerance.

The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

For purposes of this specification, addition of water from the RWST does not constitute a positive reactivity addition provided the boron concentration in the RWST is greater than the minimum required by Specification 3.1.2.8.b.2 (MODE 4) or 3.1.2.7.b.2 (MODE 5).

COOK NUCLEAR PLANT-UNIT 1 Page 3/4 44 AMENDMENT&, 4A,

3/4 LIMI'HNGCONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM SAFETY VALVES - OPERATING LIMITINGCONDITION FOR OPERATION 3.4.3 All pressurizer code safety valves shall be OPERABLE with a lift setting of 2485 PSIG + 3%.>>

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE RE UIREMENTS 4.4.3 No additional surveillance requirements other than those required by SpeciTication 4.0.5.

The pressurizer code safety valve shall be reset to the nominal value J1% whenever found outside the

+1% tolerance.

The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

COOK NUCLEAR PLANT-UNIT I Page 3/4 4-5 AMENDMENT440, 444,

3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.7 PLANT SYSTEMS CONDENSATE STORAGE TANK LIMITINGCONDITION FOR OPERATION 3.7.1.3 The condensate storage tank (CST) shall be OPERABLE with a minimum useable volume of 175,000 gallons of water.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With the condensate storage tank inoperable, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either:

Restore the CST to OPERABLE status or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, or Demonstrate the OPERABILITY of the Essential Service Water System as a backup supply to the auxiliary feedwater pumps and restore the condensate storage tank to OPERABLE status within 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE RE UIREMENTS 4.7.1.3.1 The condensate storage tank shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying the useable water volume is within its limits when the tank is the supply source for the auxiliary feedwater pumps.

4.7.1.3.2 The Essential Service Water System shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the Essential Service Water System is in operation whenever the Essential Service Water System is the supply source ior the auxiliary feedwater pumps.

COOK NUCLEAR PLANT-UNIT 1 Page 3/4 7-7 AIKNDMENT

3/4 BASES 3/4.6 CONTAINMENTSYSTEMS 3/4.6.1.4 INTERNALPRESSURE The limitations on containment internal pressure ensure that 1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 8 psig and 2) the containment peak pressure does not exceed the design pressure of 12 psig during LOCA conditions.

The maximum peak pressure resulting from a LOCA event is calculated to be less than the design limit of 12 psig, which includes 0.3 psig for initial positive containment pressure.

3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that 1) the containment air mass is limited to an initial mass sufficiently low to prevent exceeding the design pressure during LOCA conditions and 2) the ambient air temperature does not exceed that temperature allowable for the continuous duty rating specified for equipment and instrumentation located within containment.

The containment pressure transient is sensitive to the initially contained air mass during a LOCA. The contained air mass increases with decreasing temperature. The lower temperature limit of 60'F will limit the peak pressure to less than the containment design pressure of 12 psig. The upper temperature limit influences the peak accident temperature slightly during a LOCA; however, this limit is based primarily upon equipment protection and anticipated operating conditions. Both the upper and lower temperature limits are consistent with the parameters used in the accident analyses.

3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment steel vessel will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that (1) the steel liner remains leak tight and (2) the concrete surrounding the steel liner remains capable of providing external missile protection for the steel liner and radiation shielding in the event of a LOCA. A visual inspection in conjunction with Type A leakage tests is sufficient to demonstrate this capability.

COOK NUCLEAR PLANT-UNIT 1 Page B 3/4 6-2

3/4 BASES 3/4.7 PLANT SYSTEMS 3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> with steam discharge to the atmosphere concurrent with total loss of off-site power. The useable water volume limit reflects the volume of water above the centerline of the discharge pipe. An allowance for water not useable because of tank discharge line location or other physical characteristics is not required.

3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose also includes the effects of a coincident 1.0 gpm primary to secondary tube leak in the steam generator of the affected steam line.

These values are consistent with the assumptions used in the accident analyses.

3/4.7.1.5 STEAM GENERATOR STOP VALVES The OPERABILITY of the steam generator stop valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture. This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the steam generator stop valves within the closure times of the surveillance requirements are consistent with the assumptions in the accident analyses 'sed With one steam generator stop valve inoperable in MODE I, action must be taken to restore OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Some repairs to the valves can be made with the unit hot. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> completion time is reasonable, considering the low probability of an accident occurring during this time period that would require a closure of the steam generator stop valves. If the steam generator stop valve cannot be restored to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, 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 MODE 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and the MODES 2 and 3 action statement entered. The completion times are reasonable, based on operating experience, to reach MODE 2 and to close the steam generator stop valves in an orderly manner and without challenging unit systems.

Since the steam generator stop valves are required to be OPERABLE in MODES 2 and 3, the inoperable valves may either be restored to OPERABLE status or closed. When closed, the valves are already in the position required by the assumptions in the safety analysis. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> completion time is consistent with the MODE 1 action statement requirement. For inoperable steam generator stop valves that cannot be restored to OPERABLE status within the specified completion time, but are closed, the inoperable valves 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 judgement, in view of steam generator stop valve status indications available in the control room, and other administrative controls, to ensure that these valves are in the closed position.

If in MODES 2 or 3 the steam generator stop valves 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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.

COOK NUCLEAR PLANT-UNIT 1 Page B 3/4 7-3 AMENDMENT$ Q, 48$

This page intentionally left blank COOK NUCLEAR PLANT-UNIT I Page 8 3/4 7-3a

f PROPOSED CHANGES TO THE COOK NUCLEAR PLANT UNIT NO. 2 TECHNICAL SPECIFICATIONS

f h

1.0 DEFINITIONS DEFINED TERMS 1.1 The DEFINED TERMS of this section appear in capitalized type and are applicable throughout these Technical Specifications.

THERMAL POWER 1.2 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

RATED THERMAL POWER 1.3 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3588 MWt.

OPERATIONAL MODE 1.4 An OPERATIONAL MODE shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.

ACTION 1.5 ACTION shall be those additional requirements specified as corollary statements to each principle specification and shall be part of the specifications.

OPERABLE - OPERABILITY 1.6 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s). Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency elecuical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s).

COOK NUCLEAR PLANT-UNIT2 Page 1-1

0',

2.0 SAFETY LIMITS AND LIMITINGSAFETY SYSTEM SETI1NGS

'400 660 650 PSIA

~ 640 2250 PSI 8O 630:

2100 PSI Q> 620 I 2000 610 600- 1775 PSIA" 590 580 .

570 0,2 0.4 0.6 0.8 1.2 FRACTION OF THERMAI POWER t8588MWI 1.0)

Descri tion of Safe Limits Pressure Power av~ Power Tav~ Power Tav~ Power Tave (psia) (frac) ('FJ (frac) ( FII (frac) ('Fj (frac) ('Fj 1775 0.00 615.1 1.10 580.0 1.18 577.4 1.2, 576.4 2000 0.00 632.2 1.12 597.6 1.14 596.0 1.2 589A 2100 0.00 639.2 1.08 606.5 1.10 604.8 1.2 593.5 2250 0.00 649.4 1.02 619.5 1.10 610.9 1.2 599.7 2400 0.00 659.0 0.96 631.9 1.1 616.7 1.2 605.7

=

Flow Rate 91,600 gpm/loop Figure 2.1-1 Reactor Core Safety Limits Four Loops in Operations COOK NUCLEAR PLANT-UNIT2 Page 2-2 AMENDMENT2, 4Ã, 434,

2.0 SAggY LIMITS AND LHVIrrINGSAFEIY SYSTEM SETTINGS TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATIONTRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLEVALUES

l. Manual Reactor Trip Not Applicable Not Applicable

2. Power Range, Neutron Low Setpoint - Less thaa or equal Low Setpoint - Less than or equal Flux to 25% of RATED THERMAL to 26% of RATED THERMAL POWER POWER High Setpoint - Less than or equal High Setpoint - Less than or equal to 109% of RATED THERMAL to 110% of RATED THERMAL POWER POWER

3. Power Range, Neutron Less than or equal to 5% of Less than or equal to 5.5% of Flux, High Positive Rate RATED THERMAL POWER RATED THERMAL POWER with a time constant greater than with a time constant greater than or equal to 2 seconds or equal to 2 seconds

4. Power Raagc, Neutroa Less than or equal to 5% of,. Less than or equal to 5.5% of Flux, High Negative Rate RATED THERMAL POWER RATED THERMAL POWER with a time constaat greater thaa with a time constant greater than or equal to 2 seconds or equal to 2 seconds

5. Intermediate Range, Less thaa or equal to 25% of Less than or equal to 30% of Neutron Flux RATED THERMAL POWER RATED THERMAL POWER

6. Source Range, Neutron Less thaa or equal to 10 couats Less thaa or equal to 1.3 x 10 Flux per second counts pcr second

7. Overtemperature Sec Note 1 See Note 3 Delta T

8. Overpower Delta T See Note 2 See Note 4 Pressurizer Pressure- Greater than or equal to 1950 psig Greater than or equal to 1940 psig Low

10. Pressurizer Pressure Less than or. equal to 2385 psig Less than or equal to 2395 psig High

11. Pressurizer Water Lcvel- Less than or equal to 92% of Less than or equal to 93% of

- High Instrument span instrument span

12. Loss of Flow Greater than or equal to 90% of Greater than or equal to 89.1% of design flow per loop~ design flow per loop~

Design flow is I/4 Reactor Coolant System total flow rate from LCO 3.2.5.

COOK NUCLEAR PLANT-UNIT2 Page 2-5 AMENDMENTA, 434,

2.0 SAFETY LIMITS AND LIMI'IING SAFETY SYSTEM SEXI'INGS TABLE 2.2-1 Continued REACTOR TRIP SYSTEM INSTRUMENTATIONTRIP SETPOINTS NOTATION Note 1: Overtemperature

'+~

hT 6 hTo [K -Ka [

I +%'s 2

s

] (T-T') + Ks (P-P') -ft (~01 where: ETo Indtcated 6T at RATED THERMAL POWER T = Average temperature, 'F T = Indicated T>>g at RATED THERMALPOWER less than or equal to 581.3'F.

P = Pressurizer Pressure, psig P = Indicated RCS nominal operating pressure (2235 psig or 2085 psig).

I+t s The function generated by the lead-lag controller for Tavg dynamic I+~as compensation s~a ~ Time constants utilized in the lead-Iag controller for Tavg,'t 22 secs, T> 4 secs S = Laplace transform operator COOK NUCLEAR PLANT-UNIT2 Page 2-7 AMENDMENTA, 434,

2.0 SAFEIY LIMITS AND LIMITINGSAFETY SYSTEM SETTINGS k

TABLE 2.2-1 Continued REACTOR TRIP SYSTEM INSTRUMENTATIONTRIP SETPOINTS NOTATIONS Continued KI = 1.17 K2 = 0.0268 K3 = 0.00111 and fi(d I) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(i) For qt - qb between -16 percent and +6 percent, fl(hl)=0 (where q, and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and q, + qb is total THERMAL POWER in percent of RATED THERMAL POWER).

(ii) For each percent that the magnitude of (q, - qb) exceeds -16 percent, the IT trip setpoint shall be automatically reduced by 2.05 percent of its value at RATED THERMAL POWER.

(iii) For each percent that the magnitude of (qt - qb) exceeds +6 percent, the hT trip setpoint shall be automatically reduced by 2.7 percent of its value at RATED THERMAL POWER.

COOK NUCLEAR PLANT-UNIT2 Page 24 AMENDMENT82, 434,

2.0 SAFETY LIKGTS AND LIMI'IONGSAFETY SYSTEM SEI PIGS f TABLE 2.2-1 Continued REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS Note 2: Overpower 4T 6 4To PC,-Ks [

NOTATIONS Continued c>S

'"3 ] T-Ks (T-T") -f (41)]

where: 4To Indicated 4T at rated power Average temperature, 'F hdicated T>>g at RATED THERMAL POWER less than or equal to 576'F.

1.08 K5 0.02/'F for increasing average temperature and 0 for decreasing average temperature 0.00197 for T greater than T; K6=0 for T less than or equal to T v>S The haction generate by the rate lag conuoller for Tavg dy 'c comp nsanon (I+s>S)

+3 Time constant utilized in the rate lag controller for Ta<<,

T3 10 secs, S = Laplace transform operator f2 (4I) = 0.0 Note 3: The channel's maximum trip point shall not exceed its computed trip point by more than 3.75 percent 4T span.

Note 4: The channel's maximum trip point shall not exceed its computed trip point by more than 2.59 percent 4T span.

COOK NUCLEAR PLANT-UNIT2 Page 2-9 AMENDMENTSl, 434,

t f

d

3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.2 POWER DISTRIBUTION LIMITS DNB AND Tav OPERATING PARAMETERS LIMITINGCONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the following operational indicated limits:

1. Reactor Coolant System T>>g Less than or equal to 583.3'F

2. Pressurizer Pressure Greater than or equal to 2200 psig (for nominal pressure of 2235 psig) /

Greater than or equal to 2050 psig (for nominal pressure of 2085 psig) /

3. Reactor Coolant System Greater than or equal to 366,400 gpm Total Flow Rate APPLICABILITY: MODE 1 ACTION:

With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE RE UIREMENTS 4.2.5.1 Each of the above parameters shall be verified to be within their limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.2.5.2 The indicators used to determine RCS total flow shall be subjected to a CHANNEL CALIBRATIONat least once per 18 months.

4.2.5.3 The RCS total flow rate shall be determined by a power balance around the steam generators at least once per 18 months.

4.2.5A The provisions of Specification 4.0.4 shall not apply to primary flow surveillances.

Indicated average of at least three OPERABLE instrument loops.

Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of RTP Indicated value COOK NUCLEAR PLANT-UNIT2 Page 3/4 2-15 AMENDMENTA, 434,

3/4 LIMlTINGCONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 3.34 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATIONTRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINTS ALLOWABLEVALUES

l. SAFETY INJECTION, TURBINE TRIP, FEEDWATER ISOLATION, AND MOTOR DRIVEN AUXILIARYFEEDWATER PUMPS

a. Manual Initiation Sec Functional Unit 9

b. Automatic Actuation Not Applicable Not Applicable Logic

c. Containment Pressure Less than or equal to 1.1 psig Less than or equal to 1.2 psig High

d. Pressurizer Pressure- Greater than or equal to 1815 psig Greater than or equal to 1805 psig Low

e. Differential Pressure Less than or equal to 100 psi Less than or equal to to 112 psi Between Steam Lines High

f. Stcam Line Pressure- Greater than or equal to 500 psig Greater than or equal to 480 psig Low steam linc pressure steam line prcssure COOK NUCLEAR PLANT-UNIT2 Page 3/4 3-23 AMENDMENT4; 44, &, 434, ~,

3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 3.3C Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATIONTRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINTS ALLOWABLEVALUES

4. STEAM LINE ISOLATION

a. Manual See Functional Unit 9

b. Automatic Actuation Logic Not Applicable Not Applicable

c. Containment Pressure Less than or equal to 2.9 psig Less than or equal to 3.0 psig High-High

d. Steam Flow in Two Steam Less than or equal to a function Less than or equal to a Lines High Coincident with defined as follows: A Delta-p function defined as follows: A Tavg Low-Low corresponding to 1.6 x 106 lbs/hr Delta-p corresponding to 1.75 steam flow between 0% and 20% x 10 lbs/hr steam flow load and then a Delta-p increasing between 0% and 20% load and linearly to a Delta-p then a Delta-p increasing corresponding to 4.5 x 10 lbs/hr linearly to a Delta-p at full load. corresponding to 4.55 x 10 Ibs/hr at full load.

T>> greater than or equal to Tay greater than or equal to 541 F 539 F

e. Steam Line Pressure Low Greater than or equal to 500 psig Greater than or equal to 480 steam line pressure psig steam line pressure

5. TURBINE TRIP AND FEEDWATER ISOLATION

a. Steam Generator Water Level Less than or equal to 67% of Less than or equal to 68% of High-High narrow range mtrument span narrow range instrument span each steam generator each steam generator COOK NUCLEAR PLANT-UNIT2 Page 3/4 3-25 AMENDMENTA, 408, 434, ~,

,t 3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM SAFETY VALVES - SHUTDOWN LIMITINGCONDITION FOR OPERATION 3.4.2 A minimum of one pressurizer code safety valve shall be OPERABLE with a lift setting of 2485 PSIG 23%.

APPLICABILITY: MODES 4 and 5.

ACTION:

With no pressurizer code safety valve OPERABLE:

a.,Immediately suspend all operations involving positive reactivity changes*~ and place an OPERABLE RHR loop into operation in the shutdown cooling mode.

b. Immediately render all Safety Injection pumps and all but one charging pump inoperable by removing the applicable motor circuit breakers from the electrical power circuit within one hour.

SURVEILLANCE RE UIREMENTS 4.4.2 No additional Surveillance Requirements other than those required by Specification 4.0.5.

'I The pressurizer code safety valve shall be reset to the nominal value J 1% whenever found outside the +1% tolerance.

The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

For purposes of this specification, addition of water from the RWST does not constitute a dilution activity provided the boron concentration in the RWST is greater than or equal to the minimum required by specification 3.1.2.8.b.2 (MODE 4) or 3.1.2.7.b.2 (MODE 5).

COOK NUCLEAR PLANT-UNIT2 Page 3/4 44 AMENDMENT&, 4P,

3/4 LIMINGCONDITIONS FOR OPERATION AND SURVEILLANCE REQUDKMENTS 3/4,4 REACTOR COOLANT SYSTEM SAFETY VALVES - OPERATING LIMITINGCONDITION FOR OPERATION 3.4.3 All pressurizer code safety valves shall be OPERABLE with a lift setting of 2485 PSIG + 3%.~>>

APPLICABILITY: MODES I, 2 and 3.

ACTION:

With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE RE UIREMENTS 4.4.3 No additional Surveillance Requirements other than those required by Specification 4.0.5.

The pressurizer code safety valve shall be reset to the nominal value J1% whenever found outside the J1% tolerance.

The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

COOK NUCLEAR PLANT-UNIT2 Page 3/4 4-5

3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCE REQUIIKMENTS 3/4.4 REACTOR COOLANT SYSTEM LIMITINGCONDITIONS FOR OPERATION Continued SURVEILLANCE RE UIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:

Monitoring the containment atmosphere particulate radioactivity monitor at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. Monitoring the containment sump inventory and discharge at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

C. Determining the seal line resistance at least once per 31 days when the average pressurizer pressure is within 20 psi of its nominal full pressure value. The seal line resistance measured during the surveillance must be greater than or equal to 2.27 E-1 ft/gpm'. The seal line resistance, R<<, is determined from the following expression:

2.31(P,p P<<)

st qs where: Pcia, = charging pump header pressure, psig P<< = 2112 psig (Iow pressure operation)

= 2262 psig (high pressure operation) 2.31 = conversion factor (12 in/ft)t/(62.3 Ib/fthm)

Q = the total seal injection flow, gpm The provisions of Specification 4.0.4 are not applicable for entry into MODES 3 and 4.

d. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during steady state operation, and

e. Monitoring the reactor head flange leakoff system at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4.4.6.2.2 Each reactor coolant system pressure isolation valve specified in Table 3.4-0 shall be demonstrated OPERABLE pursuant to Specification 4.0.5.

COOK NUCLEAR PLANT-UNIT2 Page 3/4 4-16 AMENDMENT446, 474,

tI li

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIRElVQMIS 3/4.4 REACTOR COOLANT SYSTEM 2600 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE FOR FIRST 14.5 EFFECTIVE FULL POWER 2400 YEARS (MARGINS OF 60 PSIG AND 10'F ARE INCLUDED FOR POSSIBLE INSTRUMENT ERROR) 2200 LEAK lEST uMrr 2000 (9 MATERIALPROPERTY BASIS INTERMEDIATEPLATE, 2 C &.15%i 1800 &.57%

INITNL RTIIor 'F 14.5 EFPY RTII0T(1/4T) 178'F 1600 (3/47) 150'F g 1400 UNACCEPTABLE ACCEPTABLE I OPERATION OPERATION 1200 0 1000 o

0 PRESSURE-TEMPERATURE o~ 800 UMITFOR HEATUP RATES UP TO 60'/HR 50 100 150 200 250 300 350 400 450 AVERAGE REACTOR COOLANT SYSTEM TEMPERATURE (deg. F)

Figure 3.4-2 Reactor Coolant System Pressure - Temperature Limits for 60'F/hr Rate, Criticality Limit and Hydrostatic Test Limit COOK NUCLEAR PLANT-UNIT2 Page 3/4 4-25 AMENDMENT4Q, 42$ , XV',

3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM 2400 REACTOR COOLANT SYSTEM COOLDOWN UMITATIONS 2200 APPLICABLE FOR FIRST 14,5 EFFECTIVE FULL POW YEARS (MARGINS OF 60 PSIG AND 10'F ARE INCLUDED FOR POSSIBLE INSTRUMENTATIONERROR.)

2000 UNACCEPTABLE OPERATION 1800 I C9 ACCEPTABLE 0 OPERATION I

1600 ttj 1400 g 1200 1000 PRESSURE-TEMPERATURE LIMITS I

oO 800 I o~ 600 COOLDOWN RATE 'F/HR MATERIALPROPERTY BASIS INTERMEDIATE PLATE, C5556-2 400 Cu -.15%, Ni .57%

INITIALRTNDT 58 F 14.5 EFPY RT NDT(1/4I) 178'F 200 (3/4T)-150.F I I 50 100 150 200 250 300 350 400 450 AVERAGE REACTOR COOLANT SYSTEM TEMPERATURE (deg. F)

Figurc 3.4-3 Reactor Coolant System, Pressure - Temperature, Limits for Various Cooldown Rates COOK NUCLEAR PLANT-UNIT2 Page 3/4 4-26 AMENDMENT6Q, ~, 474,

3/4 LIMI'IONGCONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

ECCS SUBSYSTEMS - T 330'F 3 B LIMITINGCONDITION FOR OPERATION 3.5.2 Two independent ECCS subsystems shall be OPERABLE with each subsystem comprised of:

a. One OPERABLE centrifugal charging pump,

b. One OPERABLE safety injection pump, One OPERABLE residual heat removal heat exchanger, One OPERABLE residual heat removal pump, An OPERABLE flow path capable of taking suction from the refueling water storage tank on a safety injection signal and transferring suction to the containment sump during the recirculation phase of operation.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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

COOK NUCLEAR PLANT-UNIT2 Page 3/4 5-3 AMENDMENT447,

3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.7 PLANT SKI'EBLIS TABLE 3.7-1 MAXIMUMALLOWABLEPOWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 4 LOOP OPERATION Maximum Number of Inoperable Safety Maximum Allowable Power Range Valves on Any Operating Steam Generator Neuttcn Flux High Setpoint (Percent of RATED THERMAL POWER) 58.1 41.2 24.5 COOK NUCLEAR PLANT-UNIT2 Page 3/4 7-2

3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.7 PLANT SYSTEMS CONDENSATE STORAGE TANK LIMITINGCONDITION FOR OPERATION 3.7.1.3 The condensate storage tank (CST) shall be OPERABLE with a minimum useable volume of 175,000 gallons of water.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With the condensate storage tank inoperable, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either:

a. Restore the CST to OPERABLE status or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, or

b. Demonstrate the OPERABILITY of the Essential Service Water System as a backup supply to the auxiliary feedwater pumps and restore the condensate storage tank to OPERABLE status within 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE RE UIREMENTS 4.7.1.3.1 The condensate storage tank shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying the useable water volume is within its limits when the tank is the supply source for the auxiliary feedwater pumps.

4.7.1.3.2 The Essential Service Water System shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the Essential Service Water System is in operation whenever the Essential Service Water System is the supply source for the auxiliary feedwater pumps.

COOK NUCLEAR PLANT-UNIT2 Page 3/4 7-7

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BASES 2.0 SAFEIY LIMITS AND LIMITINGSAH'Y SYSTEM SKI'INGS 2.1 SAFETY LIMITS 2:1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. 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.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and reactor coolant temperature and pressure have been related to DNB through the WRB-2 correlation and W-3 correlation for conditions outside the range of WRB-2. The DNB correlations have been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, and is indicative of the margin to DNB.

The DNB design basis is as follows: there must be at least a 95 percent probability that the minimum DNBR of the limiting tod during Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-2 correlation for Vantage-5 fuel, and the W-3 conelation for conditions which fall outside the range of applicability of the WRB-2). The correlation DNBR limits are established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the DNBR limit (1.17 for WRB-2 and 1.3 for the W-3).

In meeting the DNB design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are statistically combined with the DNBR correlation statistics such that there is at least a 95 percent probability with a 95 percent confidence level that the minimum DNBR for the limiting rod is greater than or equal to a calculated design limit DNBR. The uncertainties in the above plant parameters are used to determine the plant DNBR uncertainty. This DNBR uncertainty, combined with the DNBR correlation statistics, establishes a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties. For Cook Nuclear Plant Unit 2, the design DNBR values are 1.23 and 1.22 for Vantage-5 fuel typical and thimble cells, respectively. In addition, margin has been maintained by performing safety analyses to a safety analysis limit DNBR. The margin between the design and safety analysis limit DNBR is used to offset known DNBR penalties (i.e., transition core penalties, rod bow, etc,) and provide DNBR margin for operating and design flexibility.

The curves of Figure 2.1-1 show the loci of points of THERMALPOWER, Reactor Coolant System pressure, and average temperature below which the calculated DNBR is no less than the design DNBR limit value or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid.

COOK NUCLEAR PLANT-UNIT2 Page B 2-1 AMENDMENT84, 434,

BASES 2.0 SAFEIY LIMITS AND LIIUKI1NGSAIK'IYSYSTEM SETHNGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)

Ove wer Delta T The Overpower Delta T reactor trip provides assurance of fuel integrity, e.g., no melting, under all possible overpower conditions, limits the required range for Overtemperature Delta T protection, and provides a backup to the High Neutron Flux trip. The setpoint includes corrections for changes in density and heat capacity of water with temperature, and dynamic compensation for piping delays from the core to the loop temperature detectors. The overpower delta T reactor trip provides protection or back-up protection for at-power steam line break events.

Credit was taken for operation of this trip in the steam line break mass/energy releases outside containment analysis.

In addition, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.

Pressurizer Pressure The Pressurizer High and Low Pressure trips are provided to limit the pressure range in which reactor operation is permitted. The High Pressure trip is backed up by the pressurizer code safety valves for RCS overpressure protection, and is therefore set lower than the set pressure for these valves (2485 psig). The High Pressure trip provides protection for a Loss of External Load event. The Low Pressure trip provides protection by tripping the reactor in the event of a loss of reactor coolant pressure.

Pressurizer Water Level The Pressurizer High Water Level trip ensures protection against Reactor Coolant System overpressurization by limiting the water level to a volume sufficient to retain a steam bubble and prevent water relief through the pressurizer safety valves. The pressurizer high water level trip precludes water relief for the uncontrolled control rod assembly bank withdrawal at-power event.

COOK NUCLEAR PLANT-UNIT2 Page B 2-5 AMENDMENT82, 434, 442,

3/4 BASES 3/4.2 POKER DIS'GUBUTION LIMITS 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR Continued When RCS flow rate and PAH are measured, no additional allowances are necessary prior to comparison with the limits of Specification 3.2.3. Measurement errors of 2.1% for RCS flow total flow rate and 4% for F>H have been allowed for in determination of the design DNBR value and in the determination of the LOCA/ECCS limit.

Margin between the safety analysis DNBRs and the design limit DNBRs is maintained. (Safety analyses DNBRs:

1.69 and 1.61 for the Vantage 5 typical and thimble cells, respectively. Design limit DNBRs: 1.23 and 1.22 for the Vantage 5 typical and thimble cells, respectively.) A fraction of this margin is utilized to accommodate applicable transition core penalties and the appropriate fuel rod bow DNBR penalty for the Vantage 5 fuel (equal to 1.3% per %CAP-8691, Rev. 1). The remainder of the margin between design and safety analysis DNBR limits can be used for plant design flexibility.

COOK NUCLEAR PLANT-UNIT2 Page B 3/4 24a AMENDMENT434,

3/4 BASES 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.4 UADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during power operation.

The limit of 1.02 at which corrective action is required provides DNB and linear heat generation rate protection with x-y plane power tilts.

The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned rod. In the event such action does not correct the tilt, the margin for uncertainty on F~ is reinstated by reducing the power by 3 percent from RATED THERMAL POWER for each percent of tilt in excess of 1.0.

3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters ensure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The Tav< less than or equal to 583.3'F and pressurizer pressure greater than or equal to 2200 psig (for nominal pressurizer operating pressure of 2235 psig) or greater than or equal to 2050 psig (for nominal pressurizer operating pressure of 2085 psig) are consistent with the UFSAR assumptions and have been analytically demonstrated adequate to maintain the core at or above the design DNBR thoughout each analyzed transient with allowance for measurement uncertainty.

Pressurizer pressure is limited to either of two nominal operating pressures of 2235 psig or 2085 psig, with the corresponding indicated limits set forth in the specifications. The limits are consistent with the UFSAR assumptions and have been analytically demonstrated to be adequate to maintain the core at or above the applicable design limit DNBR value for the current fuel type throughout each analyzed transient.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. The 12-hour surveillance of the RCS flow measurement is adequate to detect flow degradation. The CHANNEL CALIBRATIONperformed after refueling ensures the accuracy of the shiftly flow measurement. The total flow is measured after each refueling based on a secondary side calorimetric and measurements of primary loop temperatures.

COOK NUCLEAR PLANT-UNIT2 Page B 3/4 2-5 AMENDMENTS2, 434,

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3/4 BASES 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE/TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 4.1.4 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

An ID or OD one-quarter thickness surface flaw is postulated at the location in the vessel which is found to be the limiting case. There are several factors which influence the postulated location. The thermal induced bending stress during heatup is compressive on the inner surface while tensile on the outer surface of the vessel wall. During cooldown the bending stress profile is reversed. In addition, the material toughness is dependent upon irradiation and temperature and therefore, the fluence profile through the reactor vessel wall, the rate of heatup and also the rate of cooldown influence the postulated flaw location.

The heatup limit curve, Figure 3.4.2, is a composite curve which was prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup rate up to 60'F per hour. The cooldown limit curves of Figure 3.4-3 are composite curves which were prepared based upon the same type analysis with the exception that the controlling location is always the inside wall where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stresses at the outside wall. The heatup and cooldown curves were prepared based on the most limiting value of the predicted adjusted reference temperature at the end of 14.5 EFPY.

The reactor vessel materials have been tested to determine their initial RTNDT. The results of these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (E ) 1 MeV) irradiation will cause an increase in the RTNDT. Therefore, an adjusted reference temperature must be predicted in accordance with Regulatory Guide 1.99, Revision 2. This prediction is based on the fluence and a chemistry factor determined from one of two Positions presented in the Regulatory Guide. Position (1) determines the chemistry factor from the copper and nickel content of the material. Position (2) utilizes surveillance data sets which relate the shift in reference temperature of surveillance specimens to the fluence. The selection of Position (1) or (2) is made based on the availability of credible surveillance data, and the results achieved in applying the two Positions.

COOK NUCLEAR PLANT-UNIT2 Page 8 3/4 46 AMENDMENT6Q, 4', 474,

3/4 BASES 3/4.4 REACTOR COOLANT SYSTEM 3/4.4 9 PRESSURE/TEMPERATURE LIMITS Continued The actual shift in the reference temperature of surveillance specimens and neutron fluence is established periodically by removing and evaluating reactor vessel material irradiation surveillance specimens and dosimetry installed near the inside wall of the reactor vessel in the core area.

The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RT>>r at the end of 14.5 EFPY, as well as adjustments for possible errors in the pressure and temperature sensing instruments.

The 14.5 EFPY heatup and cooldown curves were developed based on the following:

1. The intermediate shellplate, C5556-2, is the limiting material as determined by position I of Regulatory Guide 1.99, Revision 2, with a Cu and Ni content of 0.15% and 0.57%, respectively.

2. The fluence values contained in Table 6-14 of Westinghouse WCAP-13515 report, "Analysis of Capsule U From the Indiana Michigan Power Company D. C. Cook Unit 2 Reactor Vessel Radiation Surveillance Program", dated February 1993.

The RT>>r shift of the reactor vessel material has been established by removing and evaluating the reactor material surveillance capsules in accordance with the removal schedule in Table 4.4-5. Per this schedule, Capsule U is the last capsule to be removed until Capsule S is to be removed after 32 EFPY (EOL). Capsules V, W, and Z will remain in the reactor vessel and will be removed to address industry reactor vessel embrittlement concerns, if required.

The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50.

The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4-5 to assure compliance with the requirements of Appendix H to 10 CFR Part 50.

The limitations imposed on pressurizer heatup and cooldown and spray water temperature differential are provided to assure tliat the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.

The OPERABILITY of two PORVs, or of one PORV and the RHR safety valve ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 152'F. Either PORV or RHR safety valve has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the stem generator less than or equal to 50'F above the RCS cold leg temperatures of (2) the start of a charging pump and its injection into a water solid RCS. Therefore, any one of the three blocked open PORVs constitutes an acceptable RCS vent to preclude APPLICABILITYof Specification 3.4.9.3.

3/4.4.10 STRU TURAL INTEGRITY The inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant. To the extent applicable, the inspection program for these components is in compliance with Section XI of the ASME Boiler and Pressure Vessel Code.

COOK NUCLEAR PLANT-UNIT2 Page B 3/4 4-10 AMENDMENTBQ, 423, 4A, 474

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3/4 BASES 3/4,5 EMERGENCY CORE COOLING SYSTEMS

-3/4.5.1 ACCUMULATORS Continued allowed completion times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

If more than one accumulator is inoperable, the plant is in a condition outside the accident analyses; therefore, LCO 3.0.3 must be entered immediately.

3/4.5.2 311d 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITYof two independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration. Either subsystem operating in conjunction with the accumulators is capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward. In addition, each ECCS subsystem provides long term core cooling capability in the recirculation mode during the accident recovery period.

COOK NUCLEAR PLANT-UNIT2 Page B 3/4 5-la AMENDMENTBQ, 447, 440,

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3/4 BASES 3/4.6 CONTAINMENTSYSTEMS 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that 1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 8 psig and 2) the containment peak pressure does not exceed the design pressure of 12 psig during LOCA conditions.

The maximum peak pressure resulting from a LOCA event is calculated to be less than the design limit of 12 psig, which includes 0.3 psig for initial positive containment pressure.

3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that I) the containment air mass is limited to an initial mass sufficiently low to prevent exceeding the design pressure during LOCA conditions and 2) the ambient air temperature does not exceed that temperature allowable for the continuous duty rating specified for equipment and instrumentation located within containment.

The containment pressure transient is sensitive to the initially contained air mass during a LOCA. The contained air mass increases with decreasing temperature. The lower temperature limit of 60'F will limit the peak pressure to less than the containment design pressure of 12 psig. The upper temperature limit influences the peak accident temperature slightly during a LOCA; however, this limit is based primarily upon equipment protection and anticipated operating conditions. Both the upper and lower temperature limits are consistent with the parameters used in the accident analyses.

3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that (1) the steel liner remains leak tight and (2) the concrete surrounding the steel liner remains capable of providing external missile protection for the steel liner and radiation shielding in the event of a LOCA. A visual inspection in conjunction with Type A leakage tests is sufficient to demonstrate this capability.

COOK NUCLEAR PLANT-UNIT2 Page B 3/4 6-2

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3/4 BASES 3/4.7 PLANT SYSTEMS 3/4.7.1.3 CONDENSATE STORAGE TANK I

The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> with steam discharge to the atmosphere concurrent with total loss of off-site power. The useable water volume limit reflects the volume of water above the centerline of the discharge pipe. An allowance for water not useable because of tank discharge line location or other physical characteristics is not required.

3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose also includes the effects of a coincident 1.0 gpm primary to secondary tube leak in the steam generator of the affected steam line.

These values are consistent with the assumptions used in the accident analyses.

3/4.7.1.5 STEAM GENERATOR STOP VALVES The OPERABILITY of the steam generator stop valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture. This restriction is required to I) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the steam generator stop valves within the closure times of the surveillance requirements are consistent with the assumptions used in the. accident analyses.

With one steam generator sto p valve inoperable in MODE I, action must be taken to restore OPERABLE status withm 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Some repairs to the valves can be made with the unit hot. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> completion time is reasonable, considering the low probability of an accident occurring during this time period that would require a closure of the steam generator stop valves. If the steam generator stop valve cannot be restored to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, 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 MODE 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and the MODES 2 and 3 action statement entered. The completion times are reasonable, based on operating experience, to reach MODE 2 and to close the steam generator stop valves in an orderly manner and without challenging unit systems.

COOK NUCLEAR PLANT-UNIT 2 Page B 3/4 7-3 AMENDMENTW, 470,

I ATTACHMENT 3 TO AEP:NRC:1223 EXISTI'NG TECHNICAL SPECIFICATION PAGES MARKED TO REFLECT PROPOSED CHANGES

CURRENT PAGES MARKED-UP TO REFLECT PROPOSED CHANGES TO THE COOK NUCLEAR PLANT UNIT NO. 1 TECHNICAL SPECIFICATIONS

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R ACTOR COOLANT SYSTEM SAFETY VALVES - SHU DOWN LIMITING CO'.:DITION OR OPE T ON 3.4.2 minimum of one pressurizer code safety valve shall with a lift Asetting of 2485 PSIG + *4 be OPERABLE

~CTION:

With no pressurizer code safety valve OPERABLE:

a. 'Immediately'uspend all operations involving positive reactivity changes++ and place an OPERABLE RHR loop into operation in the shutdown cooling mode.

b. Immediately render all Safety Injection pumps and all but one charging pump inoperable by removing the applicable motor circuit breakers from the electric power circuit within one hour.

SURVEILLANCE RE UIREMENTS 4.4.2 The pressurizer code safety valve shall be demonstrated OPERABLE per Surveillance Requirement 4.4. 3.

• The lift setting pressure shall correspo'nd to ambient conditions of the valve at nominal operating temperature and pressure.

~For purposes of this specification, addition of ~ater from the RWST does not constitute a positive reactivity addition provided the boron concentration in the RWST is greater than the minimum required by cification 3.1.2.8.b.2 (MODE 4) or 3.1.2.7.b.2 (MODE 5).

s KS+Pt D. C. COOK - 3JNIT 1 3/4 4-4 AMENDMENT N0.53. lan

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C 0 TEM SAF TY VhLVES - 0 ERATING G CO ION FOR OPERAT ON 3.4.3 hll pressurizer code safety valves shall be OPERhBLE vith a lift setting of 2485 PSZG g . 4 goy JRhRVJZt:

Pith one pressuri.zer code safety valve inoperable, either restore the inoperable valve to OPERABLE status vt.thin 15 minutes or be in HOT SHUTDOWN vithin 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE RE UI~NTS

(

4.4.3 No additional surveillance requirements other than those required by Specification 4.0.5.

The lift setting pressure shall correspond to ambient conditions nominal operating temperature and pressure.

of the valve at

~ Sce. 'wsvt COOK NUCLEAR PLQFZ - UNIT 1 3/la a l VOllh%P.WI 4th ~

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Insert 4-4 for 0 footnote on tech spec page 3/4 4-4 and 3/4 4-5 The pressurizer code safety valve shall be reset to the nominal value +1% whenever found outside the kl% tolerance.

PLANT SYSTB1S CONDENSATE STORAGE TANK LIMITING CONDITION FOR OPERATION 3.7.1.3 The condensate storage tank (CST) shall be OPERABLE with a minimum sea%~ volume of 175,000 gallons of water.

U5eabW APPLICABILITY: MODES 1, 2 and 3.

ACTION:

Mith the condensate storage tank inoperable, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either:

a. Restore the CST to OPERABLE status or be in HOT SHUTDOMN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, or

b. Demonstrate the OPERABILITY of the Essential Service Mater System as a backup supply to the auxiliary feedwater pumps and restore the condensate storage tank to OPERABLE status within 7 days or be in HOT SHUTDOMN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE RE UIREMENTS uQ49c 4.7.1.3.1 The condensate storage tank s all be d onstrated OPERABLE at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying the water is within its limits when the tank is the supply source for the auxiliary feedwater pumps.

4.7.1.3.2 The Essential Service Mater Syst'm shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the Essential Service Mater System is in operation whenever the Essential Service Mater System is .the supply source for the auxiliary feedwater pumps.

3/4 7-7

0 CONTAIN~

BASLC 3 4.6.1.4.. ~RÃaL PRESSURE The ]Mtatiaas oa containmaat internal pressure ensure that negative 1) th+

containmant structure is prevented from exceeding its design pressure differential vith respect co the outside atmosphere of 8 psig does not exceed*the design pressure and 2) the containmant peak pressure of 12 psig during LOCA conditions.

o

~,

The maximum peak pressure.

pressure resulting from a LOCA event is calculated to ba I~S vhich includes 0.3 psig ior inicial positive concainment

)a, ps(g, 3 4.6.1.5 AIR T~~PERATLRE

.he limicat: .".s on containment average air cemperaru.e ensure that 1) the contair en":r "ass is '.'..ite "o an initial mass sufficiently lov co p"event exceeding ='.".e C s'"n ".ress "e Cur'n" LOCA c rdi.c'ons and 2) che

~=bienc a r "e""er = re ==as "..:= x=eed ":".at te.:perarure allovable or che cont nuous C ty rac.".g s"eci"'ed ior e,uipment and instrumentation locaced vithin con"a ."..enc.

The containmenc pressure =rans'enc 's se..sitive to -'e 'ni.tially contained air =ass 'uring a L"CA.:.".e con"a'.".ed air gpss i.".c"eases vi=h ecreasi.nr. rem~erat re.;.e "ver =e=perac re I'sit og 60 r 4ill limit

":".e peak pressure =o .. ' ' 'ess chan 5g. concainmenc desi.gn pressure oi 12 ".sip. 7:".e u"per cemperacure limit inQcaences the peak acc'Cene cemoera"'e sl'>htly C r'ng a LOCA: hovever =his limit is based pr'.=a ' upon equipment procec" on and antici.paced c"crating conditions.

Borh the upper and lover temperacure limits are consistent vith che paramecers . sed in che acci.dent analvses.

3 4.6.1.6 COVTAIN~EVT VESSEL STRUCTURAL INTEGRITY This li-itacion ensures that che structural integrity of rhe containmenc steal vessel vill be maintained comparable to the ori-;nal design standards for rha li.fa of the facility. Structural integrity is requ'"ed co ensure that (1} the steel liner remains leak eight s..d (2) the concrete surrounding the steel li.ner remains capable c= providing excernal missile protection for the sceel liner and radiation shield;.-.g in the evenc of a LOCA. A visual inspection in candu-..scion vt.th Type A leakage tests is sufficient ro demonstrate chis capabi icy.

COOK NUCLEAR PLANT PitIT 1 B 3/4 6-2 AHENDMENT NO.

S S EHS

]USE S 3 CONDENSA E STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient ~ater is available to maintain the RCS at HOT

'4.

cL'~~ ~~ wctu~ 4mb mA~ ~

STANDBY conditions for 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> with steam discharge to the atmosphere concurrent

~~~ ~

with Cotal loss of off-site power. Voh~

+ ~

QsccLS&

d4bQc cf

~~'b~ Kkt1u osc~Q~

pipe ~ M o.Lhcl~ 4c A~v4+

Qsc4~c fi~ locchcoA N't4crp pg<icaJ clAa~~>~pcs ts 4 CT UI n~ tegoi~ ~

The limitations on secondary system specific actfivity ensure that the resultant off-site radiation dose vill be limited co a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose also includes the effects of a coincident 1.0 gpm primary to secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the accident analyses.

3 4 STEAN GENERATOR STOP UALUES e

The OPERABILITY of the steam generator stop valves ensures that no more than one steam generator vill blowdown in the event of a steam line rupture.

This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown,. and 2) limit the pressure rise within containment in the event Che steam line rupture occurs within containment. The OPERABILITY of the steam generator stop valises within the closure times of the surveillance requirements aze consistent w5.th the assumptions used in the accident analyses.

With one steam generator stop valve inoperable in MODE I., action must be Caken to restore OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Some repairs to the valves can be made with the unit hot. The 8 houx completion.time is reasonable, considering the low probability of an accident occurring during this time. period that would require a closure of'he steam generator stop valves. If the steam generator scop valve cannot be restoxed to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, 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 MODE 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and the MODES 2 and 3 acCion statement entered. The completion times are reasonable, based on operaCing experience, Co reach MODE 2 and to close the steam generator stop valves in an orderly manner and without challenging unit systems.

Since the steam generator stop valves are required to be OPERABLE in MODES 2 and 3, the inoperable valves may either be restored to OPERABLE status or closed. When closed, the valves are already in the posiCion required by the assumptions in che safety analysis. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> completion time is consiscenC COOK NUCLEAR PLANT - UNIT 1 B 3/4 7-3 AMENDMENT NO. W, 185,

IANT SYSTEMS

~AS~K 4 ST GENERATOR STOP VALVES co tinued

.with the MODE 1 action statement requirement. For inoperable steam generator stop valves that cannot be restored to OPERABLE status within the specified completion time, but are closed, the inoperable valves 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 Judgement, in view of steam generator stop valve status indications available in the control zoom, 'and other adlai.stzative controls, to ensure that these valves are in the closed position.

Xf in NODES 2 or 3 the steam generator stop valves cannot be restored to OPERABLE status or aze 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in NODE 4 within 12 hours. The allowed completion times aze reasonable, based on operating experience, to reach the required'nit conditions from MODE 2 conditions in an orderly manner and without challenging unit systems.

COOK NUCLEAR PLANT - UNXT 1 B 3/4 7-3a AMENDMENT NO. 185,

CURRENT PAGES MARKED-UP TO REFLECT PROPOSED CHANGES TO THE COOK NUCLEAR PLANT UNIT NO. 2 TECHNICAL SPECIFICATIONS

C~lvlTTCMS OEFlhH TERS 1.i The OEFTHEQ TERS of this sect<on ayaear <n cayftall~ type ancl are ayylfc-aole througnoict these Technical Soedflcat<ons.

>a~AL Hi~cR Z.Z TB60tN. tSCR shall be the total tor cool ant.

rsac~ core beat transfer rata to m reac-Ui-9 %Exalt. NsB 3 3 .RlicC THSNAL reactor coolant of

~ Shal1 be a total reaC~ COre heat tran!far rata tO the SzrS'PQATTGM~L, HMK 4 M QPH H'QL~K shall corresyonC to reac:lvlty concitjon, ~r le>el aniS cny one (nc4s<vt ~fnatfon of ~

ererage reactor coolant tacyerature syicif$ ect fn Taole D~

AC .0H L.5 ~QN snal1 be Ceaa acyl tf one recpsfreausss syecl flail aa corol1ary stata-oentx m earn yrfnc!yl~ syecf fccatfon anc shall be part of the soedNcaclons.

CPHXZt e CP~~i i mrs~.~Nc~sm, 096Nll 'TT lt ls cayaole train. ~onent er Cence shall of yerfornlag laa soedfleis be 0$ 694LE

~ce{s). <~

or bert balldt fn cled s oefln<t{on shall Oe tw ass~ten that al1 necessary ~n4ant eentat1on, controls, noroal aniS ~rllency elecMcal oner sourcas, coolly er seal star, luurtcac<on or'ther ecatlfcty eydyaenc that are ncpdres for the

~tan, suusystae. train, ~onenc oc Cerlca to perform its func=$ oo(s) are also cayaul>> of yerfo~og their relateis eeyort funccloaEs).

4 CÃ 9L 40 ES ?T 8 0 fA tag ta Vlt Vg  ? et T g l tt ps 4C ( ( Je F F O. 615 .9 . S.a .02 4.9 .2 cQOO O.O .SS 6 .t 0 6 59 .5 5 .5 QO .QO 6 .K 0 2 61< 0. 6 60.6 37 0 0 6'2 0.7 2$ . .9 45 340 :

f ~

0 0.0 39. .62 2.0 .2 S6 65 C<0 FSS 40 54 lA 0- 144 44 Stk 1 0 A 5 0-

, 57 e -.

so .

0 ffguiI 2.1 1

Ffer l4ta 91,600 gpafloop (Furl Vaa'Aia, 5 ~)

P~t Tave Tavg

~Q ~o Cfracl L2'fadL.44k Qmrl ~d D75 0.00 613.1 1.10 5ea.a I.le Sn.a 1.2 SZSA 2000 0.00 632.2 1.12 597.5 I. 14 596.0 1.2 589.4 2100 0.00 639.2 1.08 605.5 1.10 6N.8 1.2 593.5 2250 0.00 Q9.4 1.02 &19.5 1.10 610.9 1.2 599.7 2400 0.00 639.0 0.96 631.9 I.l , 616.7 605;F

~ ~

cg Lal CS CI 4SI 5

am

~ i+ V75 tQA

~ ~

M, "to FRACfJN Of llIEWQL NIKR (95OOWt;=La)

Figure 2.1-1 Reactor Con Safety t.fata Fen Loops 3a Operatfoa

~TUL 2.T.

TRI? fYSTDf TN~iQfQfTATI0N WI? S~~

0NAJ- UNIT WIT SETPOIÃZ AIICOASI? VAL~

l ~ Nanuel Reaocor TrSy Nec Applicable Nec ~Lteabl<<

2. fever Range, Neutron Lov Secyoknc ~ Laaa than Lov Seeyetnt ~ L<<<<a than Flux ~ r equal to 25'f RATED or equaL to 2Ct <<C RAIiD Q)ER)aU. NaX TERNAL?m EX Rtgh Secpotnt ~ Loaa than Qgh 5<<gotnc ~ L<<aa ~4 er equal co 10%% oC RATED or equaL to LL0I ef RATED CREEDAL BRED 'DER))AL 20QEX 5 ~ lover Range ~ Neutron Lkaa than er equal co 51 ef Laaa than er <<quaL co 5.$ i Flux, Mt'oatctvo RATED %QRNAL?CNXk vtch a of RATED QKlXAL 8ÃEX viA Race tive eonacant greater than a cthe cenacant g.<<at<<r

~ r equal to 2 aeeoaCa chan er <<qual ta 2 a<<coats

4. ?over Range, Neutron Laaa than er equaL to 5l ef Laaa than er <<quaL to 5.5i flux. &Sh Negative RATED THER)fAL NE'ER vtth a ~ C RAtED THECAL?CQXL vtW kate tioe conacant ggeatar than 4 test coclacaot +eater

~ r equaL to 2 aeeoeuta than er <<quaL ca 2 aecoack 5o Encawedtate Range ~ Leaa than er equaL to 25% Laaa than er <<qual to 20i Naut=on flux ~ f RATED 5)mQL NQXX oC RATED T))ERAL ROE'a)a C. Source Range, Laaa chan er equaL to LO than er <<quaL to L.l x Neucr ~flux ~ oounta p4r a4cooL 10 ceonca per <<<<cod

~ 7~ Overten)eeracuro DeLca T See Nota l See Nota 3 C. Ovegover Delta 7 See Nota 2 See Noca 4

%. Preaau=Qar Crutar than er equaL to Creacer than er equaL to

&eaaure Lov L$ 50 yatg U40 patg L0.freaaurtrer Laoa chan er equaL to Loaa than er <<qoaL co tzeaauca ee QQ 2545 yatg 25S5 yakg LL.?teaaurtaar Voter Leaa chan er equal to N% Leaa than or'qual co Sls Lovel >> Qgh ~ f tna~nc apaa ~C tna~nc tp<<n L2.Loaa of flov Creator than er <<qual to Creacer than <<r equaL to Nt ef 4aatgn flov per 4%.ll ef CeatpL .'L<<v y<<c'~

Death flov ta

<,~- ""'

~

)'(~~c.~-.+

TABLE 2. 2-1 Contfnued REA.CTOR TRIP SV~ atSmmeeaTIOS TRIP SETPOmrS NOTATION Hote 1:

Over temperature hT < hT'K -K f(1 + mls)/(1 + g2S)](T-T')+(P-P')-fl(hI) J

@here: hT 0

IndLcated hT at RhTED 5KQQL PSKR

~ hverage temperature, 0 F

~ Inttcatid T at RhTED THHQQLL tSKR less than or equal t,o

~ surfer Pressure.

l.3'F psfg P'+

~

+95 ~g p$ $ $ gfxldicRcId RCS DoEf041 opozs'cog pL' 55llzEf or QoP$ '@sic

~ S The function pnerate4 by the lead-lag controller for T4V j dynanic coctpensa&on

~ Tive constants utilized h the lead-lai controller, Tlo T2 for T; rl ~ secs, v2 4 secs.

Laplace trans fo rater COOK HUCLZhR TLhBT QHIT 2 2 7 mxmazr so. 82 ~34

II f

c 2.2-l Continued REhCTOR TRIP 5YSTBf INSTRUMBKlTICN TRIP ~c,;

TNT'ORTATION Cantiaaad) 4 Loops fn Oped'scion Xl ~ ke09 I.l7 a ~am o.oQ.47

+ ~ +41@54 0. 04 t1 t and f>(4I) ia a function of the indicated difference becveen toy snd bottoa detectors of the yavet-range nucleat ion chsnbez's; Kth gains to be selected based on aeaeued inst@ment response dutiag y4nc scatty tests esca cham:

(i) foz' and q

- qb becveen ate percent

~ yezcent RhTZD QKQQL and+6 pezcant, f>(hI)W (vheze NVKR in the top and bottoa halves q

of the coze respectively, and q + q is total OKRA B%ER in yet'cent of RhTED BANAL POUER).

(ii) fox'ach percent that the magnitude of {q - q ) exceeds yetcanc, che 4T eely setyoint shaQ be autocatically t'echoed peccant of its vatue at RAZED TH&LALESKER.

ok (iii) For each percent that the axed.ate of (g g) exceeds

+5 percent, the ht crLy setyoint sha11 be axcoeatically itdnced Q LrO pattens oi its valQ4 at RhTZD 5KRQL PSKR c4?

COOK HUCLEhR P?hHT VHZT g 2-4 AKHDt62fT 80.82 ~ <'3~

TABLE 2.2-1 Continued REACTOR TRIP SYSTBt INSTRUMENTATION TRIP SETPOINTS NOTATIONS Continued Overpo<<r AT < hT (K4-K5(t3S/(1+r3S)]T-K6(T-T"]-j (<I)]

%here:

Indicated hT at rated pover 0

Average temperature, F C

Indicated Tavg at RATED THERMAL POVER less than or equal to 576.0 F K4 1.08 K5 0.02/ 0 F for increasing average temperature and 0 for decreasing average temperature K6 - 0.0019? for T greater than T', K6 0 for T less than or equal to T" AS/(1+z 3 S) The function generated by the rate lag controller for T dynamic compensation avg T3 Time constant utilized in the rate lag controller for T y r3 ~ 10 sees'vg

.S= Laplace transform operator f2(g) - 0 0 Note 3: The channel's maximum trip point shall not exceed its computed trip point by more than (Qt percent Q'pan. I 3.'lS Note 4: The channel's maximum trip point shall not exceed its computed trip point by more than QP percent Q span.

BS'] .

COOK NUCLEAR PLANT - UNIT 2 2-9 AMamMENT NO. 82, 134

i POSE OI 5 aRI5UTION LINITS DN5 hND Tav OPERATINC PARAMZEERS LL9ITINC CONDITION FOR OPERATION 3.2.5 The folloving D?8 related parameters shall be maintain<<vichin che follovtng operational indicaced limits'.

a. DN Zo sz~k 0

1. Reactor Coolanc System T Less chan or equal co 57$ .7 W

2. Pressurizer Pressure

3. Reaccor Coolant Syscem Total Flov Rate Creacer or aqua to, Creacer than or equal co 2200 paid+/~

gp~

b. T avg

1. Reactor Coolant System T Creater chan or equal to 543,9 F>>

gg/~ APPLICA5ILITY: MODE 1 hCTION:

><ch any of the above parameters exceeding ics limit, rescore the parameter to vithin ics limit vithin 2 hours or reduce THZ@gZ, pyrZR. co less than 5g of RATED THERMAL PCQER vichin the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILvVNCE RE RBfQiTS 4.2.5.1 Each of the above parameters shall'be verified to be vithin their

.limits ac lease once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.2.5.2 Th>> indicators used co determine RCS tocal flov shall be subjected co a CHANNEL CALXBRATIOH ac, least once per 1$ months.

4.2.5.3 The RCS tocal'lov rata shall be determined by a pover balance around the steam generacors ac lease once per 1$ months.

4.2.5.4 The provisions of Spec&cation 4.0.4 shall not apply to primary flov surve illancas.

Indicacad average of at least three OPERA$ LE instrument loops.

Limit noc applicable during. either a QKBHhL PQMER tamp in excess of 5%

of RATED TKDtMAL PCKk pet'inute or a THEWhL POMEiL seep in excess of

~ 10'fIndicated RTP value COOK NUCLEAR PLANT UNIT 2 3/4 2-U muamEHT HO.S7.. f 34

I INSERT 3I4 2-<5

1. Reactor Coolant System T Less than or equal to 583.3'F

2. Pressurizer Pressure Greater than or equal to 2200 psig (for nominal pressure of 2235 psig) t Greater than or equal to 2050 psig(for nominal pressure of 2085 psig) I

TABLE 3.3-4 ENGViEKLED SAFETY FEATURE hCTUATXOH STUD( XHSTRUHEHThTIOH TRXP ~OXHTS hLLOVhBLE VALUES

l. SAFETT XHJECTXOH, TURBXHE TRIP, FEEDVhTER ISOLATION, hHD HOTOR DRIVEH hGZXLZART FEEDVhTER PUMPS

a. Eanual Initiation aaammaeaumma See 'Functional Unit 9 amaaw~a~~~~

b. hucoaacic Actuation Hot Applicable Hoc hpplicable Logic

c. Containmenc Pressure-- Less than oc'qual to Less than or equal to High l.l psig l.2 psig

d. Pressuriser Pressure- Creater than or equal Creater than or equal Lov to~ IE i8 psig to ~psig gras

e. Differential Pressure Becveen Steant Lines Less than or equal 100 psi to ll2 pai chan Lessassure or e~ to Hf.gh

f. Stean Line Prassure- Crea r than or equal Crea r chan or equal Lov - to psig steam line to psig stean line ressure ~

COOK NUCLEAR ~ >> IXT 2 3/4 3 23 f34, is7

TABLE 3. 3-4 Continued EHCZHZZEED SAFErr FEATmz hCTUATZOH STD( EHSTRmmTATZOH TRIP SETPOiHTS STIr~i LZHE ZSOLATZOH

a. manual ----------- See Puaccianal Quit 9----

b. Automatic Actuation Hoc Applicable Hoc Appli,cab].e Logic

c. Containment Pressure-- Lass than, or equal co Less chan or equa]. to High-High 2.9 psig 3.0 psig

d. Steam Flov tn Tvo Steam Less than or aqua!. to Less than or equal to Lines--High Coincident a function defined as a function defined as vith Tang--Lov-Lov follovs: A Delta-p follovs: A Delta-p corresponding to corresponding co 1.6 x 10,lbs/hr 1.75 x 10 lbs/hr steam flov becveen 0% sceam flov becveea 0%

and 20% load and chen aad 20% I.oad and thea a Delta p increasing a Delta-p increasing to a Delta-p 'inearly linearly co a Delca-p corresponding to co corresponding 4.5 x 10 I.bs/hr ac to 4.55 x 10 lbs/h fuLL load. ec full load.

T greater thea or T greata>> chan br eQI. to 541 7 equal to 539 P

e. Steam Line Pressure--Lov Crea er than oz equal Crea r chan oz equal to psig steam line t sig sceam line pressure. pressure

5. TQREIBE TRIP AHD PKKDVATEL ISOLATXOS

a. Steam Ceaeracor Vacer Less than or equal co Less than or equal co Level--Egh-Egh 67% of aarrov range 68\ of narrov range instrument span each instrument span each steam generator steam generator COOK HUCLFJLR PLEX'HZX 2 3/4 3-25 AHRHDHEHT Ho- 82, f88, f3{1, I 37

3.4.2 alaiaua of one preaaurizer coda aafety valve ahall vf.th a lift Asetting of 2415 PS' be OPERhlLZ s/

rKGK ULth ao preaaurirer code aafecy valve OPG4ULLE:

ao Zanediataly suspend all operation! invoking poaitive reactivity changes~ aad place an OPERhSLE RHR loop into operation M the ahutdova cooling aode.

b. Tauaediataly reader all Safety Infection puapa aad all bat one chary~ poap iaoperabLe by reaoviag the appLicable aotor circuit breakarx fcoo, the eleccricaL pover circuit vithin oae boar.

4.4.2 No additional Surveillance Reqnkreaeata other thea phoae required by Spec'cacion 4;0.5.

~ %he U.fc aecciag preset'halL correapond to ambient coaditiona of tba vaLve ac noaixaL operachg teaperacure aad preaanre.

~ For pnrpoiea of thea apecification, addition of eater free the ~

aoc conscience a dilution activity pcmridad che boroa coooeacracion La the doea MST ia ~eater than or equal co tha aiaiama required by specificacioa 3.L.2.8.b.2 (WOE 4) or 3.1 2.l.b.2 (~K 5).

~ 5am ~s~+ V~

3/4 4+4 ~me SO. SZ,

Insert 4-4 for 0 footnote on tech spec page 3/4 4-4 and 3/4 4-5 The pressurizer code safety valve shall be reset to the nominaL value 41% vhenever found outside the i1% tolerance.

REACT"R CMlANT mar VALVeS - OmmT1~

LINlTLM CON)ETTAN FOR OPGQTION 3.4.3 All pressurfzer code sat valves shall be OPUS@ fifth a lff't Iai:In't'485 PSN APPLKABtLITY: NOCHES 1. i and i%ON:

Mfa one presmrfzer code saf'e.y valve fnoperable, efther restore the fnooerable valve to OP64LM suNs rfthfn 15 Nfnucos or be fn %T QQTQSN ~i&fn 1Z hours.

SUR"KILU' R NENCfTS

<.<.3 Nc addftfonal Survefilance RCqufteeel& other than Mse ~ufrtd by 8".ccff'fcatfon 4.0.5.

'BT' valve ac, anal ogera fhg C4$ gcrare and pressuree 9.C. COOX - LifftT Z ~Inc 5

4.4.6.2.L Reactor Coolant System Leakages shaLL be demonstrated to be within 1 each af the abave Limits by; Monitoring the conadaaeac.atmosphere parciculate xadioaccivicy monicor ac least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Eonicoxing the concainmenc sump imrencoxy and discharge ac Lease once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

DetezminQ~ the seal Line resistance ac Least once per 31 days ~hen the average pxesstrizer pressure is @%thin 20 psi of its nominal fuLL pressure value. The seal Line xesWcmce measured during the surveiLLance muse.be gx'eater than ox eyzal to 2.27 E-l fc/gpm . The seaL line resistance, RSl, is determined fram the foLLocring expression:

R 2.31 (P - P )

i PCHP>> charging pump header pxessure, psig

%81exe

~ +giig,p~~

2262 (t~

psig (high PvesSam pressure

~&~/

operation)

PSZ 2.31 - cexversioa factor (12 in/fc)2/(62.3 Lb/Gc )--

the total seal infection floe, cpm .

The provisions af Specificacioa 4.0.4 are noc appUrsble for en~

into KODES 3 ancL 4.

d- Performance af c Reaccor Coolanc System vacar Qxveatoxy balince ac lease ance per 72 hcs during steady scaca aperacioa, aud.

e. Eoaicoring the raaccor head. flange breakoff system ac Lease once per 24 ~xzs.

4.4.6 2.2. Each reactor coolant system pressure isolation valv>> specified. in Table 3 4-0 shaLL be demoascxaced OPHUUKK pursuant to SpecQKcacioa 4.0.5.

COOK HUCZZaR PLgrZ - aaron 2 3/4 4-16

0 iy,S

'n PQ00 flEACTOA COOLAHT SYSTEH IIE LIHITATIOHS 2400 APPLICAQLE FOA FIRST EFFECTIVE FLLL POIIEA !t ij:tti!

YEAAs IttAAoltts oF ao Palo Ate IooF AAE lttcLUAEA Ii 0

FOA POSSISLE IttSTAUHEllfhfrdtl EAAOA.) It g4 3200 ~ ~

ti

~

! ill.l Ii ']

2000 t ~

LEAK MST LIHll" f000 lt 1600 IklACCEP fABLE ACCEPTASLE OPEAAl IOH OPEAATlott i400 ttt

~ ~

1800 Jid I

1000 800 UP la daof/IN PAE4 SNIE-TEHVAATLNE LIHll FOA ILATN'ATES CA 1l ICAL11Y LIHIT 600 TEAIAL PAOPEAlY SA414 ij lt  ;}:'j I IHTEAttEOIATE PLATE, Cdddd-tt 400 cu - -

.ra x. Ht .av 4 200 i}.)i'i)t IHIrrAL Al - aaof EfPY AT~< (I/ATI l74of I' 3/1TI tdoof 06 0 100 160 200 260 300 60 400 460 0

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I 5.5,2 Tvo independent ECCS cubcyctcac chall be OHXQlLK <<ith each cuhcyceaa coapriced of:

a. Ona OPKRASLK ccntribzgal charging puap,

b. Ona OPERASLK safety inf ection puap

c. Oaa OPEMLK residual heat rclotal heat exchanger,

d. Ona OPKMLK racidual beat rcaoval puap,

~. ha OPZRASLK flov path capable oC'atcLng cue@Lan froa thc ra5aaliag vaear storage ctck on a c4faty injection cigoa1 cad transferring auction eo the coneaiaacnt cusp daring the recirculation phase of I operation+

+6 I

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hSZZQK:

a. lith oae ECCL cubcyscca iaopercb4, raceoea the iaopeesb4 cubcyseea to OPERhSLK aeacus vithin 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in MZ STEROLÃ <<ithin the next L2 hours.

'<itis a ately ta] ~ aau-cue aaaea, naPa.cartSss-laca ~E~

~.cLa val<<e eo doe+.nac apply.

open poci eo or sochaoe tha-cora po<<er lave

<<iehin oem 'hoar. 5 ~ 4 ~ 4, Zn eha event tha KCCS ia aceuatad cnd in) acta <<aecr into tha Reactor

b. Coo4ne Syseaa, a Special Iapore chalI. ba prepared cad cubccLeaid eo the Coeaisaion pursuant eo gpecificatioa 4.9.2. <<ithfn. K days daacribing ehe circmaecncaa ot tha actuation cad the total acuaaalaead acacacion cyc4a co data COOK NUCLEAR PLANT UNIT I 5/4 5 5 ZmramtT NO. t67

3/4 LIMZ71NG CONDITIONS FOR OPERATION AND SURVIHLLANCE REQUGEKENTS 3/4.7 PLANT SYSTEMS TABLE 3.7-1 MAXIMUMALLOWABLEPOWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 4 LOOP OPERATION Maxinuun Allowable Power Range Maximum Number of Inoperable Safety Neutron Flux High Setpoint Valves on Any g Steam Generator (Petcent of RATED THERMALPOWER)

COOK NUCLEAR PLANT-UNlT2 Page 3/4 7-2

PLANT SYSTEMS CONOKNSATK STORAGE TANK gcen4lu LIMITING CONOITION FOR'PKRATION 3.7.1.3 The condensate storage tank (CST) shall be OPKRASLE- with a minimum conta>ne volume of 175,000 gallons of water.

APPLICASILITY: MOOES 1, 2 and 3.

ACTION:

Mith the condensate storage tank inoperable, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either:

a. Restore >e CST to OPKRASLE status or be fn HOT SHUTOOMN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, or

b. Oemonstrate the OPERASILITY of the Essentfal Service Mater System as a backup supply to the auxiliary feedwater nutnps and restore the condensate storage tank to OPERASLE status within 7 days or be fn HOT SHUTOOMN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE RK UIRKMKNTS uses bit 4.7.1.3.1 The condensate storage tank sha11 be emonstrated OPERASLE at 1east once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verffyfng the contain water volume fs within its lfmits wnen the tank.fs the suppIy source or the auxiliary feedwater pumps ~

4.7.1.3.2 The Essential Service Mater System shall be demonstrated OPKRASLE at, least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifyfng that the Essentfal Servfce Mater System fs fn operatfan whenever the Essential Service Mater Systea fs the supply source for the auxiliary feedwater pumps.

O. C. COOK - UNIT 2 3/4 7-7

2.

yo ai She:

~

ro jericeiona of chic 4 yor lafte Lime yt~nc oration &tch voul4 reaule in

~theat'f eho cha r~L roZIaso og fia i ne by ream'cew4 fuel oyoraeion eo vSthpn eha escleata boU,Q ri i +

boat c=anafar ar coefficient co ~ ta Lac'go aa4 @ho claLCtng aurfaco eNxporaeurl alightly chere eha coolant saturation eaayoraeura.

yoraeion abm tho uyyor beauhry of the nucleate.boSUng re!~ coulg f

reaule in cxcoaaSve cladding eaeyoraeuraa bocauaa o eha oaaat of day~~

froa nuclaaea botany (RQ) an4 eho raauleant chary reduction Sn boat e-anagoge coefficient. M ia not a 4tracely aeaaurabla yaraaeear 4u'Sag oyoration a~

therefore QHXM 2%SR anl reactor coolant eaayoraeuro an4 yreaauzo ~ b<<<

rolatal to QO through the 4'

~

2 corrolaeion anl 1 3 corralation for con4itioc outaida the range ot M-2. Iho NS corre4tioea boon Moloyol to predict the QQ flux an4 eho location of CA fer axiaLLy unifora aacl aonuni for heat flux Meribueiona. The Local NO heat flux ratio (MR) ia dafina4 ag eho ratio of cho hoat flux chat voQL4 cauao OtO ae a yareicular cora Location eo eho Local heat flux, an4 ia indicative of the aargin eo DI3.

The DQ doaiga baaia ia aa follows: ehaee mt be at Laaae a S5 yorconc yrobabiliey chit ehe akaSaua DHN, of ehe Lixiciag ro4 @ac~ Coaction an@

II evonta ia gcaaear than or equal ee the DIIR L~t of eha CA correla oa NX 2 correlation for Vaneago 5 fse1, cat ehe 1-3 cor:a eione Aich faLL oueaiko ehe raga of ayylicab of the - . correlation ONt Quite are oatabliahe4 baaol on the ayylicabla ixyorisoncaL data les Hach ehac chemi i$ 4 $ 5 yercacc yrobahilfty arith $ 5 yorcant confidence that QQ %11 tet occur vhea the aSniem ECQk ia at the QSt Liait (L.U fot QR$ -2 anl L.5 fm eho V-3).

In aeoCag eh>> 0% daaign baal, uneareaintiae in ylant oyoraeing yaralatars, escloar an4 ther,laL yaraaoeare, anl fuel fabrication yaraaoeara a-e atatiaticaLIy coehinal vith the %OR correlation, aeatiatics each that thoro ia

~~ ~ for ehe LidiCQ1$ rol ik gz04eat chas os ~

at Laaat a N yorcanc yrohabiliey Web a S5 yoraant confi4aaco L~L that eha tS a calculatal 4eaiya Liaie CAR. The uacereatneiae io the abele ylant yaraaeeaca aro uao4 e-

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4ata~La4 ehe jlan! MRR uncareainey. shia RR uncertaiaey, coabine4 vith cerrelaC4lc acaCL4CLcs, catahliahee a laa~ OAR ~So v5$ ch QJst be Ioc in plant aafacy analyses meggy ~uea of igygg yarg~geerg ~theet uncortaineiaa

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051R aaron for oyorath~ anl Caaiyi flaxiMLf.<.

The c~ of fiystoyraaeae, coactor Coolant 5ystaa 2.1 L ahev eho Loci of yoinee of 5RRttaL 8$ ZR, anl cveraia saeycraeuza melee Mch ehe calculatat DIIR Q ne leae than ehe 4aiiya NRL Usit ~aluN et the areraSe anthal jy at ehe veaae1 ad,t ia Laaa than the onehalyy of aaeuratal Liqs44-coec mmuc nrut - asia 2 j 20L gmma2C Ho. $ 2, >>

LIMIT:NO S~ Sermon SETTINCS hASES Overlayer Delta T The Overpover Delta T reactor trip provides assurance of fuel Lnteg:Lg, e.g..

no aelcing, under all possible overpover conditions. linits the zequLred range I

for Overtaarpetasuze Delta proceccion, and provides a backup to tha High Neutron Flux criy. The secpoint includaa corrections fot changes Ln density and heat capac'Lty af vatez vith temperature an4 47tL1lic caepensatLan far piping delays ftea the core to the loo te erasure detectors., re ace

~ Qpe (I' ~ 4-u f ~ . dur candisi fo ~ ful ove r e cures ~ safe yais. avezpaver delta T reactor t- p provides procacc on o ace uy prateo for at-paver stean L&e break events. Czedic vas taken for oyetation of chis ttip Ln the staaa LLne break ness/energy releases outside cantaimaent analysis. In addition. Lts functional capability ac the specified trap sassing Ls squired by this specif'cacion ta enhance she overall reliability of the Reactor Prataosian Sys ten.

Pressuri er Pressure The Pressurizer High and Lov Pressure stiya ate ytavi4e4 co Linis the yressuze range in vhich reactor operation Ls permitted. The High Pressure t=ip Ls backed up by che pressurizer code safety valves fot LC5 avezytessuze procection. and Ls therefore sac lover chan che sec ptessuze fot these valves (24~5 psig). The HLgh Pzaasut ~ ttiy provides protection fot a Loss of External Load event. The Lev Pteaauze czip provides protsc 'an by c:Lpping che reactor'n the evens of a loss of reactot coolant pressure.

Pressur'=er Peter Level The Pressurizer High Qatar Level trip ensures protection against ~actor Coolant Syatee ovetyreasuzizasion by limiting the vatat laveL co a volune suf Lciens co retain a ac44$ bubble an4 prevent vasar relief thtough che pressurizer safety valves. The pteasuzizar high vaser level cr'y pteclu4es vates relief fot the uncontrolled control rod assumably bank vithdzaval at ~ paver evens CaeC NUCme, tOet - UNIT 2 a 2-5 ~g)gyre HO. jZ. LJS

&~XX GT a CÃ SASK/: (Canthus)

~ ~ fl~ raca M4 aeceaa~ prior co coeyartaoc+~

2 are Seaigget, no gtQg+aaaL aLLovagceg a a

~ limital of gpectftcacSyc 3.2.3 Neaaureaet errora of 2.1% for RCX CIec tocaL flee gaea ancL 44 for g beea allow@ for La daeazakaactoa ef the deatga lMR ralua 404 ta this

~e wca~ctoa of aha LOC~CCf Ltmtc..

IaQlcaL044 becveeo ~ Safest 454%7etm Qgggk CREE the Skates Ltitc (Safe'cg 48417$ ea CAR@: L-Ct m4 L.CL fot the Vaacage MS'a 5 typical and Le calla, reapeccSmL Death Limit 411ca jo eh Vantage 5 reapecttveL 4 iTacKoc IC fueL (equai co L.3%

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death's flexibtltty.

3 4.2 POWER DISTIBUTION LIMITS BASES 3 4.2.4 UADRANT POWER TILT RATIO The quadx'anc paver cflc ratio limit assures that the radial pover distribution sacfsfies che design values used in che povar capability analysis. Radial povex distribution measurements are made during startup testing and periodically during pover operatfon.

The limit of 1.02 at vhfch corrective action is required provides DNB and linear heat generation rate protection vith x-y plane paver tilts.

The tvo hour time allovance for operation vith a tilt condition greater than 1.02 but less than 1.09 fs provfded to allov identfffcatfan and correctfon of a dropped or mfsalfgned x'od. In the event such action does not cozrect the tilt, the margin for uncertainty on F fs reinstated by zeducfng tilt the pover by 3 percent fzom RATED THERMAL POWER fear each percenc of fn excess of 1.0.

3 4.2.5 DNB PARAMETERS e limits an the DNB-xelated parametezs ensure that each of the paramete are maintained vichfn the normal steady-state envelope opex'ation a ed in che transient and accident analyses. Th . less than avg or equal to 57 . 0 F and pressurizer pxessure greater than equaY. to 2200 psfg are consfstenc th the UFSAR assumptions and ha een analytically demonstrated adequate maintain the core at or ve the design DNBR thraughout each analyzed t ienc vfth allov e foz'easurement 0

uncertainty. The T greater an or eq co 543.9 F is conservative to a safety analysis pezfoRned to demon that the plant may operate on a lineaz contz'ol program vhere the a y cal limit of T at 1008 RATED THERMAL POWER may xange fxom 5 .4 F to 5 1 F. The 1Y5it of 543.9 F

'contains a margin of 1.1 F The core may be o rated vith indicated vessel average tcmperacuze at y value betveen the uppe nd lover limits.

Pressurizer pressu is limited co a single nominal s oint, vith the lover limit of che f caced value setpofnt set forth in the sp ficacfons. The T/S value selected for consistency vith Unit 1 and conta a margin of 6 psi. e limits are consistent vfth the UFSAR assumptions and ve been an tically demonstxaced to be adequate to maintain the coxe at or ve the pplfcable design limit DNBR values for each fuel type,(which are lfste che bases for Sectfan 2.1.1) throughout each analyzed transient.

The 12 hour periodic surveillance of these parameters through instrumenc is sufficient to ensure chac the pazameters are restored vithin their 'eadout lfmitI folloving load changes and ocher expected transient operation. The 12-hour surveillance of the RCS flow measuremenc fs adequate to decect flov degradation. The CHANNEL CALIBRATION perfarmed after refueling ensures the accuracy af the shfftly flow measurement. The total flov fs measured after each refueling based on a secondary side calorimetric and measurements of primary loop temperacures.

1 COOK NUCLEAR PLANT - UNIT 2 B 3/4 2-5 AMENDMENT NO. 82. 134

INSERT A The limits on the DNB-related parameters ensur that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The T, less than or equal to 583.3 F d pressurizer pressure greater than or equal to 2200 psig ( for nominal pressurizer operating pressure of 2235 psig) or 2050 psig (for nominal pressurizer operating pressure of 2085 psig) are consistent with the USFAR assumptions and have been analytically demonstrated adequate to maintain the core at or above the'esign DNBR throughout each analyzed transient with allowance for measurement uncertainty. Pressurizer pressure is limited to either of two nominal operating pressures of 2235 psig or 2085 psig, with the corresponding indicated limits set forth in the specifications.

The limits are consistent with the UFSAR assumptions and have been analytically demonstrated to be adequate to maintain the core at or above the applicable design limit DNBR value for the current fuel type throughout each analyzed transient.

J ACTOR COOLANT SYSTEM BASES I 3 4.4.9 PRESSURE TEMPERATURE LIMITS All components in the Reactor Co olant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 4.1.4 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

An ZD or OD one~arter thickness surface flaw is postulated at the location in the vessel which is found to be the limiting case. There are several factors which influence the postulated location. The thermal induced bending stress during heatup is compressive on the inner surface while tensile on the outer surface of the vessel wall. During cooldown, the bending stress profile is reversed. Zn addition, the material toughness is dependent upon irradiation and temperature and therefore, the fluence profile through the reactor vessel wall, the rate of heatup and also the rate of cooldown influence the postulated flaw location.

The heatup limit curve, Figure 3.4.2, is a composite curve which= was prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup rate up to 60'F per hour. The cooldown limit curves of Figure 3.4-3 are composite curves which were prepared based upon the same type analysis with the exception that the controlling location is always the inside wall where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stresses at the outside wall. The heatup and cooldown curves were prepared based on the most imiting value of the predicted adjusted reference temperature at the end of EFPY.

reactor vessel materials have been tested to determine their initial t9.8'he RTmn.'he results of these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (E > 1 MeV) irradiation will cause an increase in the RT~. Therefore, an adjusted reference temperature must be predicted in accordance with Regulatory Guide 1.99, Revision 2. This prediction is based on the fluence and a chemistry -factor determined from one of two Positions presented in the Regulatory Guide. position (1) determines the chemistry factor from the copper and nickel content of the material. Position (2) utilizes surveillance data sets which relate the shift in reference temperature of surveillance specimens to the fluence. The selection of Position (1) or (2) is made based on the availability of credible surveillance data, and the results achieved in applying the two Positions.

COOR NUCLEAR PLANT - UNIT 2 0 3/4 4-6 AMENDMENT NO. 69 i,~, 17),

REACTOR COOLANT SYSTEM BASES The actual shift in the reference temperature of surveillance specimens and reactor neutron fluence is established periodically by removing and evaluating vessel material irradiation surveillance specimens and dosimetry installed near the inside wall of the reactor vessel in the core area.

The heatup and cooldown limit curves of Figures 3.4- d 3.4-3 include predicted adjustments for this shift in RT~ at the end of FPY, as well as adjustments f'r possible errors in the pressure and emperature sensing instruments.

The EFPY heatup and cooldown curves were developed based on the following:

1. The intermediate shellplate, C5556-2, is the limiting material as determined by position 1 of Regulatory Guide 1.99, Revision 2, with a Cu and Ni content of 0.15% and 0.57%, respectively.

2~ The fluence values contained in Table 6-14 of Westinghouse WCAP-13515 report, "Analysis of Capsule U From the Indiana Michi,gan Po~er Company D. C. Cook Unit 2 Reactor Vessel Radiation Surveillance Program", dated February 1993.

The RT~ shift of the reactor vessel material has been established by removing and evaluating the reactor material surveillance capsules in accordance with the removal schedule in Table 4.4-5. Per this schedule, Capsule U is the last capsule to be removed until Capsule S is to be removed after 32 EFPY (EOL).

Capsules V, W, and 2 will remain in the reactor vessel and will be removed to address industry reactor vessel embrittlement concerns, if required.

The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50.

The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4-5 to assure compliance with the requirements of Appendix H to 10 CFR Part 50.

The limitations imposed on pressurizer heatup and cooldown and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed accordance with the ASME Code requirements.

The OPERABILITY of two PORVs, or of one PORV and the RHR safety valve ensures that the Rcs will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 1524F. Either PORV or RHR safety valve has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the stem generator less than or equal to 50'F above the RCS cold leg temperatures of (2) the start of a charging pump and injection into a water solid RCS Therefore, any one of the three blocked

~

open PORVs constitutes an acceptable RCS vent to preclude APPLICABILITY of Specification 3.4.9.3.

3 4.4.10 STRUCTURAL INTEGRITY The inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant. To the extent applicable, the inspection program for these components is in compliance with Section XZ of the ASME Boiler and Pressure Vessel Code.

COOK NUCLEAR PLANT - UNIT 2 B 3/4 4-10 AMENDMENT NO ~

171 89, ~i 15k i

(Conabu~d) allovect ccnnpla&on tinea are caasonabla, bued on operactng experience, to reach che required plant conditions &m fuLL power coaditioaa in an orderly earner and vichouc chaLLcagiag plane aystaaa.

If nor ~ than one aocuasulator Sa inoperable, the plant is fn a condition outside

. tha accident analyau; therefore, LCO 3.0.3 axsc be entered Saaediataly.

The OPD4QZLITY of aco independent ECCS aubaystaas ensuru that sufhcient

~nergency cot'a coolLag capability vtLL be arailabla in tha crane of a QKA assuming the Loaa of one aubsyctaa tbr~h any aisle failure consideration.

Either aubayssea operaetng in conjunction vith the atamxlacncs ia dapabla of supplying audient cora cooling to Limit the peak claddhrg ~ieracuras &chin acceptable Xfsica for aLL poatulatad break aixaa ragtag froa the double ended break of the Largest PCS cold Leg pipe doenaar4. In addi~, each ECCS I I!

doaCag the accLdeac recovery period.

4 5EEis ~oo c?M$ clE 0 1$ , ~~ad b tvo lines the Loss cf one saf su a @le THERMhL failure co 'tion.

Quoit the The resnl taspezstnra es a decrease acceptable Lhxj ts in the erenc a postulated aaall break QÃl COOK NQC~~ PLhHT UHlT 2 5 5/4 5-4 CT HO. ~

I69,

CONTA IHMEHT S YSTEMS BASES 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that 1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 8 psig and

2) the containment peak pressure does not exceed the design pressure of 12 psig during LOCA conditions.

resvl 6n The maximum peak pressure from a LOCA event is

~lcolaWd +

+ +as f4~ M dcS\qm 4vQK + I psgq 0+i'-lv~ itic'Ivies 0.$ pstg gee lacuna(.

3/4.6.1.5 C.IR TEMPERATURE The 'limitations on containment average air temperature ensure that 1) (

the containment air mass is limited to an initial mass sufficiently low to prevent exceeding the design pressure during LOCA 'conditions and 2) the ambient air temperature does not exceed that temperature allowable for the

~t continuous duty rating .specified for equipment and instrumentation located within containment.

The containment pressure transient is sensitive to the initially contained air mass during a LOCA. The contained .air mass increases with decreasing temperature. The lower temperature limit of 60'F will limit k P PP I tk t d tg pressure of 12 psig. The'upper temperature limit influences the peak accident temperature slightly during a LOCA; however, this limit is based primarily upon equipment protection and anticipated operating conditions.

Both the upper and lower temperature limits are consistent with the para-meters used in the accident analyses.

3/4.6.1.6 CONTAIHMEHT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the con-tainment w'ill be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that (1) the steel liner remains leak tight and (2) the concrete surround-ing the steel liner remains capable of providing external missile protec-tion for the steel liner and radiation shielding in the event of a LOCA.

A visual inspection in conjunction with Type A leakage tests is sufficient to demonstrate this capability.

qO y

(

4 3 CONDENSATE STO GE TANK u5~ .

The OPERABILITY of the condensate st age tank with the m nimum water volume ensures that sufficient water is av ilable to maintain the RCS at HOT STANDBY conditions for 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> with steam di charge to the atmosphex concurrent with total loss of off-site power. The A~ allowance for water not usable because of tank discharge line location or other physical characteristicsis m+ cello'i~.

4 CT VITY The limitations on secondary system specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose also includes the effects of a coincident 1.0 gpm primary to secondary tube leak in the steam generator of the affected steam line. These values axe consistent with the assumptions used in the accident analyses.

4 5 S GENERATOR STOP VALVES The OPERABILITY of the steam generator stop valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture.

This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the steam generator stop valves within the closure times of the surveillance requirements .are consistent with the assumptions used in the accident analyses.

With one steam'enerator stop valve inoperable in MODE 1, action must be taken to restore OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.'ome repaizs to the valves can be made with the unit. hot. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> completion time is reasonable, considering the low pxobability of an accident occuzring during this time period that would require a closure of the steam generator stop valves. If the steam generator stop valve. cannot be restored to OPERABLE status within & hours, 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 MODE 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and the MODES 2 and 3 action statement entered. The completion times are reasonable, based on operating experience, to reach MODE 2 and to close the steam generator stop valves in an orderly manner and without challenging unit systems.

COOK NUCLEAR PLANT - UNIT 2 B 3/4 7-3 AMENDMENT NO. m, 170

ATTACHMENT 4 TO AEP:NRC:1223

SUMMARY

DESCRIPTION OF PROPOSED UNIT 2 POWER UPRATE TECHNICAL SPECIFICATIONS

I I l

to AEP:NRC:1223 Page 1 Key for Summary Table Page Technical Specification Page Section Technical Specification Group Related Groups Discussed in Attachment 1, Description of Proposed Changes and 10 CFR 50.92 Significant Hazards Consideration Analysis Uprate Group 1, Changes Directly Related to Increased rated thermal power HHSI Group 2, Change to Remove Power Restriction for High Head Safety Injection Cross Ties Closed Operation Margin Group 3, Changes Proposed to Increase Unit 2 Operating Margin Transition Group 4, Changes Related to Transition Core or Transition to Temperature Window/Dual Pressure Technical Specifications Both'roup 5, Changes Proposed for both units.

Admin Group 6, Administrative Change Description A Brief Description of Each Proposed Change Remarks Brief Comments with a Cross Reference to the Analyses Note that all changes are ~onl for unit 2 unless they are included in the "both" group. The changes in this group are proposed for both unit 1 and unit 2 of Cook Nuclear Plant.

(

i

Attachment 4 to AEP:NRC:1223 Page 2 Page Section Group. Description Remarks 1.3 Uprate Increase rated The support for this proposed change consists of analyses thermal power to that have been performed over a period of years. Including 3588 MNt. the new analyses, which are described in Attachment 6, NCAP 14489, and the evaluations described in Attachment 7, Balance of Plant Evaluations and Miscellaneous Safety Evaluations, all the necessary analyses and evaluations have been completed to support an uprate of Unit 2 to a core power of 3588 MNt.

The new analyses and summaries of earlier analyses and evaluations performed by Westinghouse Electric Corporation for the nuclear steam supply system (NSSS) are described in NCAP 14489. The impact of recent model changes on. the new analyses is discussed in Attachment 1 under Group 1 changes as well as in Attachment 6. Attachment 7 describes balance of plant evaluations and miscellaneous safety evaluations.

Since the analyses which support the uprated power have been performed over a period of years, Attachment 5 is provided to describe the history of earlier analyses and to identify the submittal of earlier work and the associated SER's. The review status of the analyses supporting the uprated core power is discussed in Attachment 1 under Group 1 changes and in Attachment 5.

2-2 Figure 2.1- Margin Revise Reactor The Safety Limit Figure currently in the Unit 2 Technical 1 Core Safety Limits Specifications was designed for a mixed core of Westinghouse and Advanced Nuclear Fuel. The Unit 2 core now consists totally of Westinghouse Vantage 5 fuel. The new thermal design is discussed in Section 3.3.2.1 of Attachment 6 WCAP 14489 'he proposed Safety Limit Figure is consistent with a rated thermal power of 3588 MWt and an all Vantage 5 core.

Attachment 4 to AEP:NRC:1223 Page 3 Page Section Group Description Remarks 2-5 Table 2.2-1 Admin Redefine design Minimum Measured Flow (MMF) is Reactor Coolant System Total Footnote flow in footnote Flow Rate of T.S. 3.2.5.

of Table 2.2-1 to be 1/4 MMF. MMF is used directly in the DNB analyses as discussed in Section 3.3.2.1 of Attachment 6, WCAP 14489. MMF is 1 '35 times thermal design flow (TDF). Therefore, the MMF employed in the DNB analysis is 1.035 times TDF. TDF is specified in Section 3.3.3.1 of WCAP 14489. TDF is generally used in no- DNB analyses. See, for example, WCAP 14489 tables 3.1-7 and 3.1-13. A lower (loop) TDF is indicated in Table 3.5-1 because the containment analysis bounds both units and Unit 1 is analyzed for a St flow asymmetry.

Design flow in current technical specification Table 2.2-1 is loop MMF or total MMF/4.

2-7 Table 2.2-1 Margin The upper limit on New OTDT and OPDT setpoints have been calculated to support T'ncreased to operation at a rated thermal power of 3588 MWt with all 581.34F to reflect Westinghouse Vantage 5 cores which are currently being used analyses. in Unit 2. K1 was increased from its current value of 1.09 to 1.17 thereby signi,ficantly increasing load rejection margin. The value 1. 17 was selected to allocate some margin to instrumentation, increased allowance for core burnup effects such as changes in hot leg streaming, and an increase in the positive ~I break point for the f(~I) penalty. Increased load rejection margin was also obtained by reducing tau 1 from 28 seconds to 22 seconds. The analysis value of K4 was also evaluated to obtain increased margin for burnup effects. The OTDT and OPDT trips are discussed in Section 3.3.2.1 of WCAP 14489. Details, including T', T', and time constants of the analyzed setpoint, are in Table 3.3-4 of WCAP 14489.

2-7 Table 2.2-1 Margin Both analyzed, Due to the analysis performed for transition cores of both nominal RCS Westinghouse and Advanced Nuclear Fuel, low pressure pressures, 2235 operation was not permitted for operation with mixed cores psig and 2085 of Westinghouse and Advanced Nuclear Fuel. The basis for psig, are this limitation is discussed in Section "Group 4" of indicated. Attachment 1. Unit 2 is currently operated with cores of all Westinghouse Vantage 5 fuel. Therefore, low pressure operation is acceptable.

2-7 Table 2.2-1 Margin Tau 1 reduced from See remark on increase of T'pper limit on page 2-7.

28 secs to 22 secs.

Attachment 4 to AEP:NRC:1223 Page 4 Page Section Group Description Remarks 2-8 Table 2.2-1 Margin Increase Kl from See remark on increase of T'pper limit on page 2-7.

1.09 to 1.17.

2-8 Table 2.2-1 Margin Increase K2 from See remark on increase of T'pper limit on page 2-7.

0.01331 to 0.0268 2-8 Table 2.2-1 Margin Increase K3 from See remark on increase of T'pper limit on page 2-7.

0.00058 to 0.00111 2-8 Table 2.2-1 Margin Change f(al) See remark on increase of T'pper limit on page 2-7.

penalty.

2-9 Table 2.2-1 Margin Maintain the upper This item is not a change. However, due to the fact that limit .on T at 5764F is not the maximum analyzed temperature of the 576oF to reflect temperature window, it is appropriate to identify the fact analyses. that the upper limit on T is being deliberately maintained at its current value.

Cook Nuclear Plant Unit 2 is operated a temperature significantly lower than the maximum analyzed temperature.

Therefore, the OPDT setpoint was analyzed with a low upper limit on T to convert unused margin to operating margin.

The OPDT trip is discussed in Section 3.3.2.1 of WCAP 14489.

Details, including T, of the analyzed setpoint are in Table 3 .3-4 of WCAP 14489.

2-9 Table 2.2-1 Margin Change the The values indicated in the markups of 'Attachment 3 and in allowable values the proposed technical specifications of Attachment 2 were in notes 3 and 4. calculated by our organization using the Westinghouse "stepit" methodology described in WCAP-12741. This is the same methodology used for the calculation of all existing Reactor Trip and Engineered Safety Feature Actuation Setpoints. This methodology is consistant with the requirements of ISA Standard S67.04.

Attachment 4 to AEP:NRC:1223 Page 5 Page Section Group Description Remarks 3/4 2-15 Section Transit Increase DNB Due to the analysis performed for transition cores 3.2.5 ion temperature limit consisting of both Westinghouse fuel and Advanced Nuclear from 578.74F to Fuel, the maximum nominal Tavg was limited to 576 F for 583.30F. operation with mixed cores of Westinghouse and Advanced Nuclear Fuel. Unit 2 is currently operated with cores of all Westinghouse Vantage 5 fuel. Therefore, the full analyzed temperature window analyzed for an all Vantage 5 core is acceptable.

The DNB temperature limit is obtained by adding the controller allowance to the high nominal Tavg used in the analysis and then subtracting the readability allowance.

The high nominal Tavg for a full Vantage 5 core is 581.34F and the controller allowance is 4.1oF. These values are found in Table 3 .3-1 and Section 3 .3 .3 .1 of WCAP 14489, respectively. The readability allowance, calculated by AEPSC, is 2.1 F. The resulting DNB temperature limit is 583.3OF.

The high nominal Tavg for a full Vantage 5 core is identified in the Vantage 5 Reload Transition Safety Report for Cook Nuclear Plant Unit 2 (RTSR), reference 11 of Attachment 5. RTSR was submitted via reference 13 of Attachment 13 of Attachment 5. Operation of Unit 2 with Westinghouse fuel was approved in reference 17 of Attachment 5.

Attachment 4 to AEP:NRC:1223 Page 6 Page Section Group Description Remarks 3/4 2-15 Section Transit DNB pressure Due to the analysis performed for transition cores 3.2.5 ion limits for both consisting of both Westinghouse fuel and Advanced Nuclear analyzed nominal Fuel, low pressure operation was not permitted for operation RCS pressures, with mixed cores of Westinghouse and Advanced Nuclear Fuel.

2200 psig and 2050 The basis for this limitation is discussed in Section "Group psig, are 4" of Attachment 1. Unit 2 is currently operated with cores indicated. of all Westinghouse Vantage 5 fuel. Therefore, low pressure operation is acceptable.

The DNB pressure limit is obtained by subtracting the total pressure allowance used in the analysis from the nominal operating pressure used in the analysis and then adding the readability allowance. The nominal pressures and the total allowance are found in Section 3.3.1 and 3.3.3.1 of WCAP 14489, respectively. The readability allowance, calculated by AEPSC, is 18.9 psi. The pressure limit currently in the T/Ss for high pressure operation is conservatively higher than the calculated value of 2191 psig. The proposed limit of 2050 psig for low pressure operation is an addition. Zt is conservatively higher than the calculated value of 2041 psig. The proposed value for the low pressure limit is the same as the unit 1 limit.

3/4 2-15 Section Transit Remove reference The proposed change converts the DNB speci.fication back to a 3.2.5 ion to low temperature purely DNB specification. This proposal is discussed limit in order to further in Attachment 1.

return the specification to a purely DNB specification.

3/4 3-23 Table 3.3-4 Transit Reduce safety Additional margin to trip for safety injection on low ion injection pressurizer pressure was needed as discussed in the actuation setpoint description of changes for Group 4 of Attachment 1. The on low pressurizer required evaluations to lower the safety analysis value for pressure to 1815 the setpoint have been performed as discussed in Sections psig . 3.3.4.5 and 3.3.4.6 of WCAP 14489. The nominal Technical Specification setpoint proposed is the same as the corresponding setpoint for Unit 1. Th'e instrument allowances calculated by our organization support this nominal setpoint.

Attachment 4 to AEP:NRC:1223 Page 7 Page Section Group Description Remarks 3/4 3-23 Table 3.3-4 Transit Change allowable The value 1805 psig, indicated in the markups of Attachment ion value for safety 3 and in the proposed technical specifications of Attachment injection 2, is conservative to the value that was calculated by our actuation setpoint organization. Et is consistent with value for unit 1 and is on low pressurizer proposed in order to make the Technical Specifications of pressure to 1805 the two units more similar.

psig.

3/4 3-23 Table 3.3-4 Transit Reduce safety At the time of the transition from Advanced Nuclear Fuel to ion injection Westinghouse fuel, the reanalysis of mass and energy release actuation setpoint (M&E) outside containment for rerating and reduced on low steam line temperature/reduced pressure operation was not complete. The pressure to 500 evaluation of the then applicable analysis assumed an NSSS peig. power of 3425 MWt and a nominal setpoint for low steam line pressure no less than 520 psig.

The AEPSC analysis of the impact of the steam line mass and energy release (SM&E) outside containment on the operabilit of equipment in the main steam enclosures, at 3588 MWt core power, was described and submitted in reference 24 of Attachment 5. Reference 24 was our proposal to operate both units with 0 ppm boron concentration in the boron injection tank (BZT) . The SHEE was analyzed consistently with the proposed safety injection setpoint on low steam line pressure. The mass and energy release portion of this analysis is discussed in Section 3.3.4.7 of WCAP 14489.

The core response steamline and feedwater line breaks submitted with the RTSR support the proposed setpoint. The references for the RTSR, its submittal, and approval are 11, 13, and 17 of Attachment 5. Evaluations of these analyses are discussed in Sections 3.3.4.6 and 3.3.4.7 of WCAP 14489.

The revised SM&E release analysis to containment also supports the proposed setpoint. This analysis is discussed in WCAP 14285, reference 29 of Attachment 5. WCAP 14285 was submitted via reference 30 of Attachment 5. Zt is not yet approved. This analysis is summarized, in Section 3.5.4 of WCAP 14489.

Attachment 4 to AEP:NRC:1223 Page 8 Page Section Group Description Remarks 3/4 3-23 Table 3.3-4 Transit Change allowable The value 480 psig, indicated in the markups of Attachment 3 ion value for safety and in the proposed technical specifications of Attachment injection 2, is conservative to the value that was calculated by our actuation setpoint organization. It is consistent with value for unit 1 and is on low steam line proposed in order to make the Technical Specifications of pressure to 480 the two units more similar.

psig.

3/4 3-25 Table 3.3-4 Transit Reduce steam line See remark on reduction of safety injection actuation on lo ion isolation steam line pressure on page 3/4 3-23.

actuation setpoint on low steam line pressure to 500 psig.

3/4 3-25 Table 3.3-4 Transit Change allowable See remark on the change of allowable value for safety ion value fox steam injection actuation setpoint on low steam line pressure on line isolation page 3/4 3-23.

actuation setpoint on low steam line pressure to 480 psig.

3/4 4-4 Section Margin Increase The Non-LOCA accidents were reanalyzed or reevaluated based 3.4.2 Pressuxizer Valve on a pressurizer valve setpoint tolerance of 3%. This is Setpoint Tolerance noted in section 1.1 and 3.3.2.2 of WCAP 14489. The to 3%.. analyses affected are discussed in Sections 3.3.4.3, 3.3.4.4, 3.3.4.6, 3.3.5.1, and 3.3.5.2 of WCAP 14489.

3/4 4-4 Section Both Add footnote This requirement is consistent with a similar requirement 3.4.2 requiring an as which was approved for the main steam safety valves. It is left tolerance of being submitted for both units because it was inadvertently 1%. omitted in our submittal AEP:NRC:1207, dated May 26, 1995, which included the analytical justification for an increase in pressurizer safety valve setpoint tolerance for Unit 1.

3/4 4-5 Section Margin Increase See remark on setpoint tolerance magnitude on page 3/4 4-4.

3.4.3 Pressurizer Valve Setpoint Tolerance to 3%.

3/4 4-5 Section Both Add footnote See remark on as left setpoint tolerance on page 3/4 4-4.

3.4.3 requiring an as left tolerance of 1%.

j',

Attachment 4 to AEP:NRC:1223 Page 9 Page Section Group Description Remarks 3/4 4-16 Section Transit Pressure criteria Due to the analysis performed for transition cores of both 4.4.6.2.1 ion for both analyzed Westinghouse and Advanced Nuclear Fuel, low pressure nominal RCS operation was not permitted for operation with mixed cores pressux'es are of Westinghouse and Advanced Nuclear Fuel. The basis for indicated. this limitation is discussed in Section "Group 4" of

l. Unit 2 is all Westinghouse Vantage currently Attachment operated with cores of 5 fuel. Therefore, low pressure operation is acceptable.

3/4 4-25 Figure 3.4- Uprate Reduce the The applicability of the current heatup and cooldown curves 2 applicability of is discussed in Section 3.11.2.1 of Attachment 6, WCAP the heatup curve 14489.

from 15 to 14.5 EFPY's.

3/4 4-26 F igu re 3 . 4- Uprate Reduce the See the xemark on the applicability of the heatup and 3 applicability of cooldown curves on page 3/4 4-25 the cooldown curve from 15 to 14.5 EFPY's.

3/4 5-3 Section HHSZ Remove power When the SBLOCA analysis performed to support the main stea 3.5.2 reduction safety valve tolerance relaxation to 3% was carried out, it currently required was found that a power reduction was required to obtain for operation with satisfactory results with the high head safety injection HHSZ cross tie (HHSZ) cross ties closed. Since then, improvements have valves closed. been made to the Westinghouse NOTRUMP SBLOCA model. As indicated in the cover letter to this submittal, the SBLO analysis performed for this submittal was performed using the new model. The results of this analysis show that an acceptable PCT results with the HHSZ cross tie valves closed at a core power of 3588 MWt. The SBLOCA analyses are described in Section 3.1.2.4 of WCAP 14489.

3/4 7-2 Table 3.7-1 Uprate Lower the maximum These setpoints are calculated in accordance with the Allowable Power prescription in Westinghouse Nuclear Safety Advisory Letter Range Setpoint.94-001. Zn this prescription, nominal NSSS power appears in the denominator of the equation for the setpoint.

Therefore, the proposed setpoints were lowered accordingly.

Attachment 4 to AEP:NRC:1223 Page 10 Page Section Group Description Remarks UNIT 1 UNIT 1 Both Change contained The analysis of 3 .10 .2.5 of WCAP-14489 shows that 174, 500 3/4 7-7 Sections volume to useable gallons of water are required to maintain the RCS at hot 3.7.1.3 and volume. standby for 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />. The Technical Specifications currentl 4.7.1.3.1 require a minimum contained volume of 175,000 gallons of water. However, due to the fact that the zero of the level UNIT 2 UNIT 2 instrumentation is located at the centerline of the 3/4 7-7 Sections discharge pipe, above the level for required NPSH, all the 3.7.1.3 and indicated volume is useable. Therefore, the proposed 4.7.1.3.1 Technical Specifications are revised to address useable volume.

B 2-1 Bases Transit Remove references The Unit 2 core now consists totally of Westinghouse Vantage Section ion to Advanced 5 fuel.

2.1.1 Nuclear Fuel.

B 2-5 Bases Margin Remove detail from The discussion of the proper normalization of T is being Section the discussion of removed. This information is documented in Section 3.3.2.1 2.2.1 the OPDT protection trip.

of WCAP 14489 and will be controlled administratively.

B3/4 2-4a Bases. Transit Remove references The Unit 2 core now consists totally of Westinghouse Vantage Section ion to Advanced 5 fuel.

3/4.2.2 and Nuclear Fuel.

3/4.2.3 B 3/4 2-5 Bases; Transit Remove reference See remark on low temperature limit on page 3/4 2-15.

Section ion to low temperature 3/4.2.5 limit in order to return the specification to a purely DNB specification.

B 3/4 4-6 Bases Uprate Reduce the See remark on the applicability of the heatup curve on page Section applicability of 3/4 4-25.

3/4.4.9 the heatup and cooldown curves from 15 to 14.5 EFPY's.

B 3/4 4- Bases Uprate Reduce the See remark on the applicability of the heatup curve on page 10 Section applicability of 3/4 4-25.

3/4.4.9 the heatup and cooldown curves from 15 to 14.5 EFPX's.

Attachment 4 to AEP:NRC:1223 Page 11 Page Section Group Description Remarks B 3/4 5- Bases HHSZ Remove power See remark on power reduction required by SBLOCA on page 3/4 1a Sections reduction 5-3.

3/4.5.2 and currently required 3/4.5.3 for operation with HHSZ cross tie valves closed.

UNIT 1 UNZT 1 Both Change peak Discussion of maximum calculated containment pressure is B 3/4 6-2 Bases containment given in Sections 1.2 and 3.5.2 of WCAP 14489. Zt is noted Sections pressure to that the analysis bounds both Unit 1 and 2 in Section 3/4.6.1.4 reflect analysis 3.5.2.1 and the peak containment pressure is documented in 3/4.6.1.5 result. Section 3.5.3.6.

UNIT 2 UNIT 2 B 3/4 6-2 Bases Sections 3/4.6.1.4 3/4.6.1.5 UNIT 1 UNIT 1 Admin Change contained See remark on changing contained volume to useable volume B 3/4 7-3 Bases volume to useable page 3/4 7-7.

Section volume.

3/4.7.1.3 UNIT 1 UNIT 2 B 3/4 7-3 Bases Section 3/4.7.1.3

I ATTACHMENT 5 TO AEP:NRC:1223 DISCUSSION OF PREVIOUS RELATED SUBMISSIONS

to AEP:NRC:1223 Page 1 Introduction The analyses that support the proposed uprating of Donald C. Cook Nuclear Plant unit 2 have been performed over a period of years in several contexts. The analysis of the nuclear steam supply system (NSSS) for an NSSS power of 3600 MWt was performed in conjunction with analyses to operate unit 1 at reduced temperature and pressure (the "Rerating Program" ). Most of the core response analyses were performed at an uprated core thermal power of 3588 MWt as a part of the transition from Advanced Nuclear Fuel to Westinghouse Vantage 5 fuel. The recently submitted analyses, AEP:NRC:1223, to support an increase in the permitted level of steam generator tube plugging (SGTP) for unit 1 includes a steam mass and energy release (SM&E) analysis to the containment which bounds both units at 3600 MWt. For this submittal, previous NSSS analyses and core response analyses have been reviewed, new analyses have been performed where necessary, and the balance of plant (BOP) evaluated, as described within this submittal, to support the proposal to increase the core rated thermal power to 3588 MWt. to this submittal is WCAP 14489. Xt describes the analyses, evaluations, and reviews performed by Westinghouse Electric Corporation and summarizes earlier work performed by Westinghouse Electric Corporation to support an increased core rated thermal power for unit 2. WCAP 14489 also describes analyses and evaluations performed simultaneously to support certain increases in operating margin such as increased setpoint tolerance for the pressurizer safety valves. Attachment 7 discusses balance of plant evaluations that have been performed by AEPSC to assess the impact of increased core power.

Section 2.0 of WCAP 14489 discusses the previous work performed by Westinghouse Electric Corporation to support the uprated core power for unit 2. The evaluations described in WCAP 14489 are based on these earlier analyses. The earlier analyses are described in Rerating Program WCAP's 11902 and 11902 Supplement 1, references 3 and 10, and in the Vantage 5 Reload Transition Safety Report for Donald C. Cook Nuclear Plant Unit 2, Revision 1, March 1990 (RTSR), reference 11. The SGTP SMEE analysis is described in WCAP 14285, reference 29.

WCAP 11902 and its supplement are referred to as the "Rerating Program" in WCAP 14489. The reload transition safety report is referred to as "RTSR" in WCAP 14489. The increase in the permitted level of steam generator tube plugging program is referred to as "SGTP Program" in WCAP 14489. The rerating Program, RTSR, SGTP Program, WCAP 14489 (the unit 2 Uprating Program), and the BOP evaluations provide the support for this submittal.

Pu ose of Attachment 5 The purposes of. this attachment are to:

indicate those aspects of earlier analyses which have been submitted for NRC review and approved,

2. indicate those aspects of earlier analyses which have been submitted for NRC review but are not yet approved (This category is comprised of the SGTP Program.),

Attachment 5 to AEP:NRC:1223 Page 2

3. briefly describe the earlier analyses, and 4 ~ provide references for previous submittals and NRC SER's for the convenience of the reviewer.

The discussion of this attachment describes submittals for both units because much of the supporting analysis for unit 2 was performed to bound both units. Submittals primarily for unit 1 are sometimes supported by analyses bounding both units and/or include Technical Specifications modifications for unit 2.

The following list summarizes the status of analysis features of earlier analyses:

Princi al Features of the Earlier Anal ses Which Have Been Reviewed and A roved b NRC 1~ Reduced temperature operation for unit 2.

2. 10% degradation for the RHR and HHSI pumps for both units.

3. Increased MSIV response time for both units.

4. BIT 0 ppm boric acid concentration for both units.

5. Reduced temperature and pressure operation for unit 1.

6. Reduced minimum measured flow for unit 1.

Princi al Features of the Earlier Anal ses Which Have Been Submitted for Review But Have Not Been A roved (These features are proposed in the unit 1 increased steam generator tube plugging limit submittal, reference 30.)

10% degradation for the centrifugal charging pumps for both units.

2. Minimum RWST temperature of 70~F for both units.

3. Shutdown Margin requirement of 1.3% for both units.

4, Proposals to increase operating margin for unit 1.

5. Analysis to support operation of the spent fuel pool with one unit. operating at 3588 MWt Princi al Features of This unit 2 U rate Submittal

1. Unit 2 rerate to 3588 MWt.

2. Increase the tolerance of pressurizer safety valve setpoint to 3% for unit 2.

3. Increase OTaT/OP~T operating margin.

4. Remove mixed core penalties.

5. 15% degradation for safety injection and RHR pumps.

Pu ose of the Earlier Anal ses Reratin Pro ram The earlier analyses were performed to accomplish a number of goals. The first of these was to permit operation of unit 1 at reduced primary temperature and pressure. The benefit of

.operating in a reduced primary temperature and pressure mode was to slow the degradation of the unit 1 steam generators. In addition, since essentially all of the analytic basis of the Cook Nuclear Plant units had to be reviewed or revised, analyses were performed to position unit 1 for subsequent uprating to 3413 MWt core power and unit 2 to 3588 MWt core power. The margin formerly intended to be allocated to a potential unit 1 uprate was subsequently allocated to an increased steam generator tube

to AEP:NRC:1223 Page 3 plugging limit in the unit 1 increased steam generator tube plugging limit submittal, reference 30. The earlier analyses also supported increased operating margins in selected areas.

Among these were increased allowable ECCS pump degradation, reduction of required shutdown margin (SDM), a reduction in the minimum temperature of the refueling water storage tanks (RWST),

reduction to zero of the boron concentration in the boron injection tanks (BIT), and slower response times for certain components and systems which applied to unit 2.

Descri tion and Review Histo of Prior Submittals This section describes the prior analyses for the Cook units in essentially chronological order with an emphasis on unit 2.

The first of the earlier analyses is described in reference 1, WCAP-11908, "Containment Integrity Analysis for Cook Nuclear Plant units 1 and 2." It was submitted for NRC review by reference 2. Reference 1 presented a long term containment analysis which bounded both units at a core power of 3413 MWt, operation at a reduced temperature and pressure, and operation of the ECCS with residual heat removal (RHR) crossties closed.

Since the analysis described in reference 1 was performed, two new long term containment integrity analyses have been performed for the Cook Nuclear Plant units. One was performed in conjunction with the proposal to increase the unit 1 steam generator tube plugging limit. It was performed at an NSSS power of 3425 MWt and is described in reference 29. Therefore, a new analysis was performed at an NSSS power of 3600 MWt. It is described in WCAP-14489 which is Attachment 6 to this submittal and reference 31. Neither of these new analyses has been reviewed at this time by the NRC.

The next group of analyses is described in reference 3, WCAP-11902, "Reduced Temperature and Pressure Operation for Cook Nuclear Plant Unit 1 Licensing Report." Reference 3 presented the remainder of the analyses and evaluations necessary to support operation of unit 1 at reduced temperature and pressure.

The NSSS systems and components analysis was performed for an NSSS power level of 3600 MWt. Reference 3 was submitted for NRC review by reference 4.

The letters of references 5, 6, 7, and 8 provided supplementary information to the staff related to the request for approval (references 2 and 4) to operate unit 1 at reduced temperature and pressure. The request to operate unit 1 in this manner was approved by reference 9.

Reference 10, WCAP 11902, Supplement 1, "Rerated Power and Revised Temperature and Pressure Operation for Cook Nuclear Plant Units 1 &. 2 Licensing Report", describes the balance of the analyses which were performed by Westinghouse Electric Corporation to support the operation of unit 1 at uprated power and reduced temperature and pressure. This report describes analyses and evaluations which were performed to bound both units at an uprated NSSS power of 3600 MWt. In particular, NSSS systems and components were evaluated for an uprated NSSS power of 3600 MWt for both units. The report also describes an analysis of the steam mass and energy release (SM&E) to containment, the associated containment analysis, and the SMRE

Attachment 5 to AEP:NRC:1223 Page 4 release outside containment. These two analyses assumed a shutdown margin of 1.3%, an increased time response for main steam isolation, 0 ppm boron concentration in the boron injection tank (BIT), and were performed to bound both units at the unit 2 uprated core power of 3588 MWt.

Since the SMGE to containment analysis described in reference 10 was performed, the SMRE to containment has been reanalyzed. The new analysis also bounds both units at an NSSS power of 3600 MWt.

Zt was performed in conjunction with the proposal to increase the unit 1 reference steam generator tube plugging limit.

29 and was submitted with reference 30.

It is described in Reference 11, "Vantage 5 Reload Transition Safety Report for Cook Nuclear Plant Unit 2 (RTSR)", together with reference 1, "Containment Integrity Analysis for Cook Nuclear Plant units 1 and 2", reference 3, WCAP 11902, Reduced Temperature and Pressure Operation, and reference 10, WCAP 11902, Supplement 1, Rerated Power and Revised Temperature and Pressure Operation, support reduced temperature and pressure operation for unit 2 at an uprated core power of 3588 MWt. However, reference 1 and the RHR and HHSZ crosstie closed LOCA cases of reference 11 only support a unit 2 core power of 3413 MWt. The analyses reported in references 1, 10, and 11 support operation with Westinghouse fuel, 10% degradation of the CCP's, HHSZ pumps, and RHR pumps, an increase of 3 seconds in MSIV response time for unit 2, 0 ppm boric acid concentration in the BIT for unit 2, a minimum RWST temperature of 70'F for unit 2, and a SDM of 1.3% for unit 2.

The letter of reference 13 submitted reference 11, RTSR, and the portions of reference 10, WCAP 11902, Supplement 1, which addressed the SMEE to the containment.

The letters of references 14, 15, and 16 provided supplementary information to the staff related to reference 13. Operation of unit 2 with Westinghouse fuel at reduced temperature, with 10%

degradation of the RHR and HHSZ pumps, was approved by reference

17. Some changes to both the unit 1 and unit 2 Technical Specifications which returned certain activities to administrative control were also made.

The letters of references 18 and 19 for unit 1 and unit 2 respectively proposed technical specifications that implemented an increase of 3 seconds in the MSZV response times. These proposals were supported by reference 3, WCAP-11902, for both units, reference 10, WCAP-11902, Supplement 1, for both units, reference 11, RTSR, for unit 2, and evaluations performed by us.

The letters in references 18 and 19 submitted the portions of reference 10, WCAP 11902, Supplement 1, which addressed the SMEE to the containment. The propo'sais to increase the MSZV response times by 3 seconds were approved by references 20 and 21.

The letter of reference 22 proposed to reduce the primary system minimum measured flow (MMF) for unit 1. This proposal was approved by reference 23.

The letter of reference 24 proposed to reduce the boron concentration in the BIT's of both units to 0 ppm. This proposal was supported by reference 3, WCAP-11902, reference 10, WCAP-11902, Supplement 1, reference 11, RTSR, and analyses performed by us. The AEPSC analysis of the impact of the steam line mass and energy release (SM&E) outside containment on the operability

4 fl to AEP:NRC:1223 Page 5 of equipment in the main steam enclosures was described in reference 24. Reference 10, WCAP 11902, Supplement 1, was submitted in its entirety in support of this proposal. The proposal was approved by reference 25.

The letters of references 26 and 27 proposed to relax the tolerance of the main steam safety valve (MSSV) setpoints for both Cook Nuclear Plant units. The proposal was based on new analyses and on evaluations performed by Westinghouse Electric Corporation. The evaluations were based on the analyses described in reference 1, WCAP-11908, "Containment Integrity Analysis for Cook Nuclear Plant units 1 and 2", reference 3, WCAP-11902, reference 10, WCAP-11902, Supplement 1, and reference 11, RTSR.

The descriptions of the new analyses and evaluations were included as attachments to these letters, references 26 and 27.

The unit 2 SBLOCA analysis described in the MSSV submittal was performed assuming that the high head safety injection (HHSI) crossties were closed. It was performed at a core power of 3250 MWt. As a result provisions were included in the Technical Specifications which required a power reduction when the HHSI crossties are closed. In this submittal, a proposal to remove this provision is made based on a new SBLOCA analysis using new models. The new analysis is de'scribed in Attachment 6 of this submittal which is reference 31.

The MSSV setpoint relaxation proposal was approved by reference 28.

A recent submittal which impacts the proposal to uprate unit 2 is the proposal to increase the limit of unit 1 steam generator tube plugging. This submittal is reference 30. It has not yet been approved by the NRC. Reference 30 includes a revised steam mass and energy release to containment analysis which bounds both units at an NSSS power of 3600 MWt and proposals to increase operational margin for unit 2. The increases in operating margin include 10% head degradation for the centrifugal charging pumps, a reduction in the minimum refueling water storage tank temperature to 70oF, and a reduction in the required shutdown margin to 1. 3't. Reference 30 also proposes to increase the pressurizer safety valve tolerance from 1% to 3% for unit 1 only.

Two of the proposed changes in this submittal, AEP:NRC: 1223, is the addition of footnotes which requires that the as left magnitude of the pressurizer safety valves be 1%. This proposal was inadvertently omitted from Reference 30. Attachment 6 to reference 30 is reference 29, WCAP 14285, Revision 1, "Donald C.

Cook Nuclear Plant Unit 1 Steam Generator Tube Plugging Program Licensing Report".

In addition, this submittal r'equires the approval of previous submittals (references 32 and 33) in order to be implemented.

They have not yet been approved by the NRC. These submittals contain a "Refueling Operations Decay Time Technical Specif ication Amendment Request" permitting core of fload 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after subcriticality. The analyses for this proposal support the operation of one unit, Donald C. Cook Nuclear Plant unit 2, at a core power of 3588 MWt from a spent fuel pool thermal hydraulic point of view.

The" final submittal for unit 2 is this submittal. It proposes to uprate the core rated thermal power to 3588 MWt. It also proposes increases in unit 2 operating margin by increasing as

Attachment 5 to AEP:NRC:1223 Page 6 found pressurizer safety valve setpoint tolerance to 3%,

increasing the OT~T/OP~T operating margin, and increasing the analyzed safety injection and residual heat removal pump head degradation to 15%. Attachment 6 to this submittal is reference 31, WCAP 14489, Revision 1, "Donald C. Cook Nuclear Plant Unit 2, 3600 MWt Uprating Program Licensing Report".

References WCAP-11908, Containment Integrity Analysis for Cook Nuclear Plant Units 1 and 2, M. E. Wills, July 1988.

2. Letter AEP:NRC:1024D, Containment Long Term Pressure Analysis to Support RHR Crosstie Closure, from M. P.

Alexich to T. E. Murley, August 22, 1988.

WCAP-11902, Reduced Temperature and Pressure Operation for Cook Nuclear Plant Unit 1 Licensing Report, D. L. Cecchett, and D. B. Augustine, October 1988.

4, Letter AEP:NRC:1067, Reduced Temperature and Pressure Program Analyses and Technical Specification Changes, from M. P. Alexich to T. E. Murley, October 14, 1988.

5. Letter AEP:NRC:1067A, Supplemental Technical Specification Changes for Reduced Temperature and Pressure Program, from M. P. Alexich to T. E. Murley, December 30, 1988.

6. Letter AEP:NRC:1067B, Additional Information on Reduced Temperature and Pressure Submittal: Boron Dilution Accident, from M. P. Alexich to T. E. Murley, February, 6, 1989.

7. Letter AEP:NRC:1067C, Unit 1 RTP Program: Additional Information on Containment Structural Analysis, from M. P.

Alexich to T. E. Murley, March 14, 1989.

8 ~ Letter AEP:NRC:1067D, Modification of Reduced Temperature and Pressure Program Technical Specification Changes, from M. P. Alexich to T. ED Murley, June 5, 1989.

9. Amendment No. 126 to Facility Operating License No. DPR-58.

10. WCAP 11902, Supplement 1, Rerated Power and Revised Temperature and Pressure Operation for Cook Nuclear Plant Units 1 6 2 Licensing Report, September 1989.

11. Vantage 5 Reload Transition Safety Report for Cook Nuclear Plant Unit 2, B. W. Gergos, Editor, January 1990.

12. No reference 12.

13. Letter AEP:NRC:1071E, Unit 2 Cycle 8 Reload Licensing, Proposed Technical Specifications for Unit 2 Cycle 8, and Related Unit. 1 Proposals, from M. P. Alexich to T. E.

Murley, February 6, 1990.

14. Letter AEP:NRC:1071H, Modification to Our Previous Submittal AEP:NRC:1071E; Revised Figures for the Loss of Load Event, from M. P. Alexich to T. E. Murley, April 6,

to AEP:NRC:1223 Page 7 1990.

15. Letter AEP:NRC:1071I, Information to Supplement Our Previous Submittals AEP:NRC:1071E and 1071H, from M. P.

Alexich to T. E. Murley, May 29, 1990.

16. Letter AEP:NRC:1071K, Offsite Dose Calculation for the Reactor Coolant Pump Locked Rotor Event for Unit 2 Cycle 8, from M. P. Alexich to T. E. Murley, July 23, 1990.

17. Amendment No. 148 to Facility Operating License No. DPR-58 and Amendment No. 134 to Facility Operating License No.

DPR-74.

18. Letter AEP:NRC:1120, Expedited Technical Specification Change Request Steam Generator Stop Valves, from M. P, Alexich to T. E. Murley, January 31, 1990.

19. Letter AEP:NRC:1123, Technical Specification Change Request, Steam Generator Stop Valves, from M. P. Alexich to T. E. Murley, May 14, 1990.

20. Amendment No. 147 to Facility Operating License No. DPR-58.

21. Amendment No. 135 to Facility Operating License No. DPR-74.

22 Letter AEP:NRC:1130, Technical Specification Change for Unit 1 Cycle 11, from M. P. Alexich to T. E. Murley, July 23,1990.

'3.

Amendment No. 152 to Facility Operating License No. DPR-58.

Letter AEP:NRC:1140, Technical Specification Change Request, BIT Boron Concentration Reduction, from M. P.

Alexich to T. E. Murley, March 26, 1991.

25. Amendment No. 158 to Facility Operating License No. DPR-58 and Amendment No. 142 to Facility Operating License No.

DPR-74.

26 Letter AEP:NRC:1169, Technical Specifications Change to Increase the Allowable Tolerance for Main Steam Safety Valves, from E. E. Fitzpatrick to T. E. Murley, November 11, 1992.

27. Letter AEP:NRC:1169A, Update for Technical Specification Change to Increase the Allowable Tolerances for Main Steam Safety Valves, from E. E. Fitzpatrick to T. E. Murley, December 17, 1993.

28. Amendment No. 182 to Facility Operating License No. DPR-58 and Amendment No. 167 to Facility Operating License No.

DPR"74.

29. WCAP 14285, Revision 1, Donald C. Cook Nuclear Plant Unit 1, Steam Generator Tube Plugging Program Licensing Report, May 1995.

30. Letter AEP:NRC:1207, Technical Specification Changes Supported by Analyses to Increase Unit 1 Steam Generator Tube Plugging Limit and Certain Proposed Changes for Unit 2 Supported by Related Analyses, from E. E. Fitzpatrick to

I, to AEP:NRC:1223 Page 8 USNRC Document Control Desk, May 26, 1995.

31. WCAP 14489, Revision 1, Donald C. Cook Nuclear Plant Unit 2, 3600 MWt, Uprating Program Licensing Report, May 31, 1996.

32. Letter AEP:NRC:1202, "Refueling Operations Decay Time Technical Specif ication Amendment Request", from- E. E.

Fitzpatrick to W. T. Russell, November 16, 1994.

33. Letter AEP:NRC:1202A, "Refueling Operations Decay Time Technical Updated Analysis and Response to Request for Additional Information", from E. E. Fitzpatrick to Document Control Desk, February 1, 1996

Oditgdg ~~~5 TCPIHEO TE~ og thfs sectfon appear fn capftalfaa4 type ancI are applfc-aol ~ ~rougnout these Tecnnfcal Soecf I'Icatfons.

f'AL~icR Lg THP?QL KVQ shall ba the total reactor core heat transfer rata to the reac tor <<ool ant.

RA~~.  %~ERAL DOVER 3 . M-,c5 THQNAl. KVHt shall be a total reactor core heat transfer rata to %e reacar coolant of'm'PKVT.ORAL 8308 to any one fnclusf ve cMfnatf on of cars An Qp~'KATRINA$~5 shal reaCtivity Canoitien, fn Table LL

~r 1 correspond leVel anc aVerage reaCta~ COOlant taCperature SpeCiffeo AC.."N 1.5 ACHCN Shall be thOSe aelftfcnal reCuf~ntS SpeCfffed aa COrellary Stata-aents each prfnCfple specff'fcatfon anC shall be part of the speCftfcatfonS.

OycRlSt ~ - OPKRAEI ITr 1.d A

~ ft systae, sesysta, trafn, aaponent or cevfca shall be OPQNQ or have fs capable of'erfenfnfl Its <pecfffed tune.fon(s). belfcft

~r at~t OPKRASLLLTY In this 4effnftfon shall be the ass~tfon that all necessary Instru-aentatfon, controls, noreal ancI ~agency electNcal sourcas, coolfng or seaI qatar, luhrfcatfon or other oaflfary equfpeent that are recufee4 tor the system, subsystem, trafn, coeenent or avfca to perron Its I'unc:fon(s) are also capanle of perf'orafng thefr related support functfon(s).

0. C. CXK UN'

f

, II

yE5ICN ~V ~ 91.600 CP:I/'OP

-ON OF SAFE% I.~uj-.S

~

O'ESCA-'P Poua r Pouter Tavg Poua't TL'lg

p. capture ~t.scl (0

~

' ~ ~(. Ic ~)1

~C ~F

0. CO~~ 615. 4 0.9d 553.5 02 0.9 1.2 5j 631.d 605.d 0 597.5 1 56 .5 I 00 .8.00 d6

';00 0. 00 639. 1 062 6140 096 016 12 5.3..

.'. 50 0 CO 0.72 62 .6 095 6 2 12 Sjo.-

0.00 659. 0.62 642.0 1.1 599.0 1.2 Sdd.;

2 CQ 50- 24 )5IA 40 ~

250 ~ 5IA 30- 2IOO 10" 2OOO eelA

)0-

)0- 1775 P5IA 30-IO 0 0.2 e.4 0. 0.8 1 1.2 FRACTiON OF RA D THER POISE hL Figure 2.1-1 y Foucaccor Cora Safacy'kakcs Loops Q Oparacfon

~I NUCLEAR P ~ UNIT 2 2.3 ANBQMm NO. Sk,l87 >34

D/-

'desi

~- . Floe Rate ~ 91,600 9~)oop Pressur ~ Peer Tave Cfoul QQ Peer Lfraal ~

Tave Paver Lfml l.li

~

Tave Poser Qml Tavo 1775 0.00 C15.1 1.10 5.D 577.4 1.2 576.4 2000 0,00 C32.2 1, lk 597.C 1.14 594.0 l,2 589.4 2100 0.00 639.2 1,00 604.5 1.10 604.8 1.2 593. 5 2250 0.00 649.4 1.02 61%.$ 1.10 610.9 l,2 599.7 2400 0.00 659.0 0.9C 631.% 1,1 61C.7 1.2 605.7 Figure 2.1-1 Rector Core Safety Lte)tc Four Loops )e'perat)oa

(%P

REA~R TRIP SYS~ INSTRUKQfTATION TRIP SETPOINTS FCVC. IONAJ. CRIT TRIP SETPO!ÃT ALLOVaSLE VA~ "ZS

l. manual Reactor Trtp Noe APPltcable Net hppltcabl ~

2. 'Fever Range. Neucton Lov Seepotnt Less chan Lov Seepotne Lees chan Flux or equal eo 2Si of RATED or equal eo 2ii of RA~i I3uuaL POVER THENtAL POVER Htgh Setpotne ~ Leaa chan Htgh Setpotnt - Less chan or equal to 109i of RATED ot equal to 110i of RATED THERMAL ?OVER ntERHAL Penta Pove'r Range, Neutron Lees chan et equaL to Si of Leaa than ot equaL co 5.Si Flux. Htgh toatttve RATED THERQL AVER vtch a of RATED THER'. M'ER vtch Race ttae constant greater thaa a ttae constant greeter or equaL to 2 aecoeta chan ot equaL to 2 seconds

4. Pover Range, Neutron Lese chan et equal co Si of Leaa than or equal to S.Si Flux, Htgh Negactv>> RATED TIEM tOVER vtth a of RATED THERMAL tOVKX vtch Race ctae constant greatet chan cthe constant gteecet or equaL co 2 aecon4a ~ chan er equaL eo 2 second@

5. Ineernedtate Range, Lees than ot equaL to 25i Leaa than ot equaL to 30i Neuetoa Flux of RATED THESNhL ?OVER, of RATED THERMAL 8REX C. Source Range, Lese than er equaL to 10 La~a chan er equal eo L.3 x Neuer one FLux couacs per sacer! 10 counts per second

.7. Overeeapetatute See Note 1 See Noee 3 Delta T

l. Ovetpover Delta T See Note 2 See Note 4

9. Pressuriset Creator chan or equal co Createt than or equaL eo

?tessure ~~ Lov 1950 pstg 1940 patg 10.tressurtser Liaa chan or equal to Leaa than ot equaL eo

?ressure ~ Htgh 25g5 yatg 2395 pstg

11. Presauttser Vatat Lese than or equal co 92i Leaa chan or equal eo 93i Lave 1 Htgh of tnsetuleae epaa of tnsetuaeat span 12.Loss ot Flov Cteatet than ot equal ce Creator than ot equal co 90i of 4eatgn flov 49.Li of dostgn flov per pet'oop+

1oopr Destgn flov ts + Am<~ F,2~

l(y g~+w Cori ~ ~+ ( p COOK NUCLEAR NO.$ 2, 1

2. 1. 1 REACTOR CORE The resertctions oj this safety lkatt prevent overheating of the fuel and posstbl ~ <Laddtng perforation vhtch vould result kn ehe release of fission produces to the reactor coolant. Overheating of the fuel cladding ks prevented by rescrtcstng fuel operacLon to vtthkn the nucleate boiling regiae vhere ehe heac transfer coeffickenc Ls large'nd the cladding surface teaperature ks slightly above the coolant saturation teaperatura.

Op<<ratton above ehe uyper boundary of the nucleate boiling regiae could result tn excessLve clad4ing eeayeraeures because of the onset of departure froa nucleate boiling (DNb) and the resultant shary reduction tn heat transfer coeffLckenc. OHb Ls noc a dLrectly aeasurable paraaecer during operation and therefore TNERNAL. POVER and reactor coolant teaperaace and pressure have been eelaeed eo DNb through the Vkb 2 correlation and V 3 correlation for conditions outside che range of Vkb.2. The DNb correlations haw been developed eo predict the DHb flux and the locaeton of DNb for axially unifora and nonunkfora heac flux diatrtbutkons. The local DHb heae flux raeLo (DHbR) ts defined as the ratio of che heat flux that vould cause DNl at a particular core location eo the local heac flux, and ks kndkcatiw of ehe aargkn to DNb.

The DNb design basks ks as follovs: there aust b>> ae lease a 95 percenc yrobabilkty chat the akntam DNbk of ehe lkakting rod during Condition I and II evencs ks greaser than or equal to ehe OHbk ltakt of ehe OHb correlacion betn ~ Vkb-2 correlaeion for Vantage.5 fuaL, and the V.S correlacion for fess'on4ktkons vhkch fall outside the range of apylkcabklity of

~ correlaeion DNbk lkaits are established based on the entire aypltcable axperkaental daca sec such thas share ks ~ 95 percent probahklkey vich 95 percenc confi4ence that ONb vill not occur vhen ehe aknkauw ONbk is at the DNbk liat t (1. U foe'kb 2 and S for the V. S) .

l.

In seating ehe, DHb design basis, uncertainties in plane oyeraekng paraaesers, nuclear and charnel paraaeeers, and fuel fabrication paraaeters are statksttcally coabkned vieh the DNlk corralackon staeksekcs such that share Ls ac least a 9S yercenc probabklLey vkth ~ 95 percent confidence level that ehe aknkaua ONbk for the 1kakekng rod ks greater shan or equal to a calculated design liaic DHbk. The uncertainties Ln the above plant paraaetars are use4 so deceratne the plane DHbk uncertakncy. This DHbk uncertakncy, coabkned vteh she DNbk correlackon scaekstics, establishes a design ONbk value vhLch aust be aec in plane safety analyses ustng values ot tnyut paraaecers vtehouc uncereaknttes.

For Cook Huclear Plane UnLe 2, she design DHbk values are 1.2S and Vents ~ -5 fuel thkable cells~ res ectkvel

~ n a tton, aarg n a ~

aa nea ne n oratng safety analysea eo a safecy analysis ltate OHbk. The aargkn betvaen she design and safecy analysis ltake DNbR ks used to offset knovn DHbk penalttes (i. ~ ., sransttkon core penalties.

eod bov, etc.) and provtde DHbk aargtn for oyeracing an4 design flexkbklkcy.

The curves of figure 2.1-1 shov th>> lock of points of TNERNAL POVER, Reactor Coolant Syssea pressure, and average taaperature belov vhich the calculated DNbk is no less than the 4eskgn DHbk ltake value or ehe average

~ nehalpy ae ehe vessel exte ta Less than ehe enthalpy of saturated liquid.

COOK NUCLEAR RIANT UNIT 2 b 2~1 ~MBC NO SZ, ~ I

AbLE . 2-1 Cont tnued CTOR TRIt SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION Note 1:

Oyerteaperature hT < hT tK1 K2((1 + rlS)/(1 + f25)](T T )+Kg(P 'P

) f1 (hI) 1

@here: 4T IndIcated hT at RATED THMAL ?OVER T Averaje temperature, F T

r ZT!.P '~

'ndicated at T Tressuriser Tressure, RATED S~

QtHNAL Ig POVER leea than or equal to P' 3455 peij gpdtcated RCS noaknal operating pressure[

ZZ g5 PSST) aT gOeS PSTg) 1+tS e function jenerated by the lead-lag controller for T dynaakc coapenaatkon ay!

Togae constance utilized kn the lead laj controller for Tayge Tl ~ car T2 < a ceca ~

gz S Laplace traneforN operator COOK NUCLEAR PLANT UNIT 2 F 7 A)tBrDKNT NO. S2 ~

~ 1 Cont Lnued R RIP 5YS DtSTRUNXNT+ Of TRIP SIT?OI S ROTA 0Ã Cont Lnued C lA s Ln oration Ll-o- o g4'P O. OOI I l and f (41) is a function of the. Ln4icated difference betveen top and bottok detectors of the power. range nuclear Lon chaabors: vith gains to be selected base4 on aeasured Lnsttuaent response during plant atartup tests such chat: ~lb (L) for q -

qb betveen

- percent and +C percent, fl(4I)W (vhere q qt aro percent RATED tho top an4 bectoe halves an4 of thl core respectively.

percent of RA~i T)CRQL and q ROVER).

iq THESNhL POSER Ln Ls total THIS(AI.?CNKk La iC ~

~g(LL) for

-A ro each 'percent that the %aptitude of (q ~ ~) exceeds percent. the 4T trip setpoint shall lA <<ut5aatically c.os percent of ita value at RATED QCRHAl PCQZX (LLL) For percent that tho aagnLtuda ef (q - q ) exceeds

+4 reduced by ~

percent, the 4T trip setpoint shall be auteKatically percent of Lts value at RATED THERMAL POCk.

n Say~

2-'s COOL HVCLXAk NJ8T UNIT 2 2 4 A)mammer HO 52 134

Insert 2-8 for Note a for tech spec page 2-&

Note a T'hall be set to a value equal to or less than the Indicated T~ at RATED THEF22L POWER. Indicated T~'and T'an be set to any value within the range of 547 to 581.3 deg. F.

LX 2. I Continued REACTOR TRI? SYSTEM INSTRUMENTATION TRI? SET?OINTS NOTATIONS Continued Note 2: Overyover hT 4 hT ]X ~ K (t 5/(I+t $ )]T-f. (T T'] f (hI)]

4>ere:

hT Indicated hT at rated yover

~ Average teayerature, F

~ Indicated Tavg at kATED THECAL tNTX lees than or o

equal to 576.0 F

~ 1.0$

0.02/ 0 F for increaaing average teayerature and 0 for decreaaing average teayerature e 0.00197 for T greater chan T', X< ~ 0 for T leaa than or equal to The function generated by the rate lag controller tor T'35/(I+t35)

T dpuaic coarpenaati.on

~ Tine conacant utiliaed in the race lag controller for t3 T ~

t3 ~ 10 aeca ~

5= ~ Laplace tranafon oyerator f ( hl) 0.0 Note 3: The channel'a aaxieia trip poinc shall not exceed its coayuted trip yoinc by aore'than%i@ percent hT apan. 'I ita trip Noce 4: The poinc by sore than ~

channel'a aaxiaua cr point shall noc exceed p rcenc hT apan.

coeyuted COOK NUCLGR PlhÃt - UNIT 2 2 ' A)meme NQ. 82, ~

(l LIHITINC S~ SYSTEM( S~INCS EASES Overoover 'Delta T The Ovcrpover Delta T reactor trip provides assurance of fuel integrity. e.g..

no melting. under all possible overpover conditions, 1tatts the required range for Overtemperature Delta T procection, and provides a backup to the High Neutron Flux trip. The secpoint includes corrections for changes tn denstty and heat capacity of vater vith teayerature, and dynaatc coap<<nsatton for piping delays from the core to che loo te recure decectors. re nce

~ mp (T s e ~ u er su f e dur o conditi fo e e ful eve r e cures as uae n e safe lysis. e overpover delta T reactor tr p provides procecc on o ac .up procecc on for at.pover steam 1!ne break events. Credit vas taken for operation of this trtp in the steam line break mass/energy releases outside containment analysis. In addition, its

.unctional capability at thi specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.

Pressurizer Pressure The Pressurtter High and Lov Pressure trips are provtded to limit the pressure range in vhich reactor operation is permitted. The High Pressure trip is backed up by the pressurtser code safecy valves for RCS overpressure protection. and is therefore sec lover than the set pressure for these valves (2485 pstg). The Mtgh Pressur'e trip provides proceccton for a Loss of External Load event. The Lov Pressure trip provides protection by tripping t'e reactor in the event of a loss of reactor coolanc pressure.

Pressurtter abater Level The Pressurizer High Vater Level trtp ensures procectton against Reactor Coolant Syscea overpreasurttatton by ltitctng the vacer level co a volume sufficienc to retain a stems bubble and prevent v'ater relief through che pressurizer safety valves. The presauriser htgh vater level crip precludes vater relief tor che unconcrolled control tod asaeably bank vithdraval at~pover event ~

COOK NUCLEAR PLANT UNIT 2 g 2-5 AHENDHENT NO. Sl. D4

~

~ES:

DrSmZauZTOX (Consinued)

~nS Vhen kCS flov rate and F are aeaaured, no additional allovancea are neseaeary prior to soapariaoAith the liaise of Specificacion 3,2.3 Heaaureaenc errora of 2.1t for kCS tlov total flov race and 4e for F'< hav<<

been alloved for in determination of the daaign DNSk value hand in thP determination ot the LOCA/ECCS linis.

margin becveen che eafecy analyaia DNA and she death liait DNA ia aaintained. (Safecy analyaea Mka.'.C9 and 1.Sl for the Vantage 5 sypical and thbable cells reapeeckvel Deaigi IMHfHSka: 1;2$ an4 1.22 for cd Vantagi 5 ical and

~ ca ISSS856' core a, reapeccively,I penaltiea an4 the appropriate fuel rod

'tanaision bow DM penalty for she Vantage 5 fuel (e~l to 1.$ t per MCA? 4691. kev. 1). The reaaindar of the aargin becveen deaign and aafesy analyata DHR liaise can be uaed for plans design flexibility.

I Dc@

COOK 8UCLXAk ASLANT UNIT 2 I 3/4 2.44

rOVKR ISTHMI Ur ON ~ ~

DMS AND Ta OrEU.TINO Bma~

lDMITINC CONDITION Ot EQUATION 3.2.5 The folloving DHS related paraaecara ahall be maintained vichin che folloving operational indicated liaica:

l.

f 1. keacto ane Syate Leaa 6pii o 524.7 6

2. P auditor er chan OQ peigt/m

. Reactor ant Syetoe Rate Creator ~~a 3CC,400 gp~

~

av 1,'or

~ ant S a T .C er chan o 1 ee~N~ i ACT10!l:

Vith any of ehe abow paraaeeere exceeding ita liaic, restore the paraseeer to viehin tta liait viehin 2 hasta or reduce TSKNAL NvEk to leaa chan Se of kATXD DtXRQL NOEL viehin the next 4 haxz,a.

SURVEILLANCE 4.2.5.1 Earth of the abow paraaetara ahall be wrified to be vichin their

.Linica at leaat once per 12 ho%ra.

4.2.S.2 The indicatora used to determine kCS total flov shall b<<cub)ected eo a CHANEL CALIBRATION ae leaat once per 18 aontha.

4.2.5.3 The kCS total flov race ahall be determined by a pover balance around the aeeaa generators at leaae ence per 1! acth!.

4.2.5.4 The provtaiona of Speciftcaeien 4;0.4 ahall noe apply eo primary flov aurve illancea.

Indicaead average ot ae leaae three OPQAlLI insceuene loope,

~e of net applicable haring eieher RATED THDHAL tOQXX per siNkce or a THOL NVR rasp a TSXRQL NVKX ace'p in exceaa of Se in exc ~ ee of 10% of RX?

Indicaced YalU4 COOK NCLXAk KANT UNIT 2 3/4 2-15 ANE2mntt SO.S2,134

insert 1 to page 3/4 2-15 Reactor Coolant System T,,

T,, ( (581.3+5.1- Indication errortn) 'F*

2. Pressurizer Pressure - for normal Pressure Operation, Przr Pres > (2235-63+ Indication error"') psig /'~

at Reduced Pressure Operation, Ptzr Pres > (2085-63+ Indication error"') psig*/ "

3. Reactor Coolant System > 366,400 gpm*"'otal Flow Rate (1) Indication error to be provide d by AEPSC for these limits note that once AEPSC determines the indication errors for these DNB limits the absolute limit can be calculated and inserted into the LCO

ff

! ff If

TASLK 3.3 4 ENCIHZZkXD SAXXT? TKlTUkZ ACTUATIOH SYSTZH IMSTkUKEKfATION TkII'ZTTO~S FUR CTIOÃAL UNIT TkI? SE?20IHTS AILOVAMXVALUES

1. S~

TZI? .

IHJZCTIOH. TUkaIZX TZXDUATXk ISOIATIOtf, AND HOTOk DkIVXN AQXIIZJXT FXXDVATXX ZEROS

a. Manual Initiation o ro o s o o o o o o o See functional Unit $ oaoooe oa os a

b. Autoaatic Actuacion Hot Applicable Sot Applicable Logf.c

c. Concahaent heeaure- Lee! than oc'qual co than or etpkl to ~

High 1.1 paig i+2 peig

'aaa

d. ?zeeeuriaer ?reeaur ~- Crea or Cr or egal Loer t pe t fKsW

e. Dfdferencial teeaaure Leaa or equal to ae than or eqaal to le@teen Steaa Linee-- 100 paf. 12 pei High

t. Stean Line Pzeaaure--

Lcnr Creatar than or to 0 eig steaa p a aura ~: 1, .Creater than or equal inL to pr paig aceaa line e

gp PjC gc P~

COOK NUCLZAX fLAÃ7 - UHIT 2 3/4 3 23 AffXHD~ SO. f fICta7

l~

TLSLX S.S 4 Coachmed EHCDtEERED SAFETY tEATURX ACTUATION STSTEN IliST1~4TIOg ~ SET~~

FUNCTIONAL UETZ TRIP StTPOIÃTS

4. STEAM LTSX ZSOLLTXOS

a. Namali ---- ------ See tanctional Uaic 9 -<<--------

b. Aacoaatic ketaacioa Sot Acyl,icable Sot ipplicable Logic

c. Coacainaeat ?zessare-- Less chan or 2.9 ps+

e~ L4ss than or equal cp RLgh Rt$ h

~ paid 3~o

d. Stean tloe ia Two Stean Lines--Nigh Coincident L444 chan a

or bmccioa defined as co Less chan oC a

e~l bmctioa defined co as arith Tang--Lcm-bar folly: L Delta-p foliose: 4 Delta-p corre~dSng to correspond~ to 1.I x 10 lbs/hn L.75 a 10 lbs/hr stean flee between tA stean floe between 0\

and 20% load and then aad 20% load aad thea a Delta p iacreasiag a Delta-p iacrushg linearly to a Delta p linearly to 4 Delta-p corres~ding to corresponding to 4.5 n 10 Ebs/hr't co 4.55 x 10 lbs/hr fall load. at M. load.

pouter thea or T greater chan or

~ Ql co 541 P ~ Ql co 539 1

e. Steu Line hessare Let ~ r chan oc'tpLl Cteater chan ot egal to paid sceaa line to paid stean line

~ sears snre

5. TURJDC TXZ? i'ttÃELfZR EP.C ( egg

~~e ZSOLATIQN

a. Scwa Cenerator 1facer L444 than or co Less then or egad to Level--1K'-ltd, I71 of narrow IN of narrows tease instrnNenc span each insczuaeac span each stean generator acean jeaetacor COOK HUCLXLk HL?C USXT 2 3/4 S 25

3.4.2 A minimum of one pzessuzixer code safety valve shall be OPGLQLE vith a lift setting of 2485 PSIG +

3/

~ ith no pressurizer code safety valve OPERhSLE:

Immediately suspend all operations involving positive reac ivity changes~ and place an OPERhbLE RHR loop into operation in the shutdovn cooling mode.

b. Zmmediately zendez all Safety Zn]ection pumps and all. but one charging pump inoperable by removing the applicable motor circuit breakers from the electrical pover circuit vt.thin one hour.

'.4.2 'Ho additional Surveillance Requirements ocher than those required by Specification 4.0.5.

~ The lift setting pressur ~ shall correspond co ambient conditions of valve ac.nominal opezating temperature and pressure.

che For purposes of this specification, addition of vater from che RUST does not constitute a dilution activity provided th>> boron concentration 'in the RJST is gzeacer than oz equal to the minimum required by specification 3.1.2.8.b.2 (MODE 4) or 3.1.2.7.b'.2 (MODE 5).

ss g~ M5g~f 3/4 4-4

Insert 4-4 for 4 footnote on tech spec page 3/4 4-4 and 3/4 4-5 The pressurizer code safety valve shall be reset to the nominal value'l% whenever found outside the +1% tolerance.

~J l~

REACTOR CNLAMT SYSTEM SAf 5TY YALVKS OPERATING

ITIA% COJOITION'OR OPERATION l.4.3 All pressurfxer code saf'alves shall be OPBNHLE ith a 1 fi't setting of 2485 PSIG +

APPi. I&8ILITY: USES 1 2 and ACTION:

Nth one pressurfzer code safety valve inoperable, either restore the inoperable valve to OPERA8LK status within 15 efnutos or be in ifOT S:UTCOX( withfn 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> s.

SURVKILH.'CK RE JIREMBITS 4.4.3 Nc addftional Survef1'ance Roqufr e.ents other than those recufr ed by Specfffcatfon 4.0.5.

'TKT .

valve at emfna! operatfnq tcmperat re and pressure.

~v S Cr+

Q.C. COOK - UNIT 2 2/4 4-5

4 .4.6.2.1 Reactor Coolant Systaa leakages shall be oeaonscraced to be vi nin each of the above limits bye aa Nonitoring the containment aerosphere parcioukate radioaccivicy aonitor at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. Nonitoring the containnent amp inventory snd discharge at leuc once per 12 bours.

Co Detenaining the seal line resistance at least once per 31 days vhen the average pressuriier pressure is vithin 20 psi of its nominal full pressure value. The seal Una resistance meuured during che surveillance aust be greater than or equal to 2.27 E-1 ft/gpa . The seal line resistance, RSL, is deteaained fron the folloving

~ xpresaion:

R << 2.31 (P - P )

L 0 ~

vhere: PCHP charging paap header pressure, psig PSZ ~ 2262 psig (high pressure operation) 2.31 ~ conversion faccor (12 in/ft)2/(62.3 lb/fc3) the total seal ln] ection flovy gpss The prevtsiona ot Specification 4.0.4 are noc applicable for <<ncxy inco NODES 3 and 4.

d. Perfonsance of a keactor Coolant Syscea vater inventory balance ac least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during steady state operacion, and

~. Monitoring the reactor bead flange ledcoff systea at least once per 24 bours.

4.4.6.2.2. Each reactor coolant systea pressure isolation valve specified in Table 3.4 0 shall be deaanstcated OPERASLE pursuant to Specification 4.0.5.

COCK HUCLEQ PLAFT - UNIT 2 3/4 4-16

I

~IFAIIF~HH

~IPMU Sl'UU~~

IIIII

[ .

I

llllllW~~

II Hll I

II 8'5 I

Q l Qll I

I II I I I I I I ~ I

3.5.2 Tvo independent ECCS subsysteas shall be OPH4QLE vith each subsystacs coacprised of:

4. One OPERAILE CenRifugal Charging pCnCpe

b. One OPMLK safety injection pucap

c. Ona OPERAILK residual heat reaova1 heat exchanEer,

d. One OPERABLE residual heat reaeval puap, e~ An OPERAILK flov path capable of taking suction troa the refueling vater storage tank on 4 safety injection siEnal and transferring suction to the containsent suscp during the recirculation phase of operation.

b3r&Waf4 ecti cr ~

EZZQEc 4>> Vith one ECCS subsystea inoperable, restore the inoperable subsystaa to OPERAELK status vithin 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or be in HOT QNTDOQH vithin the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. arith.a safety inje cross-tie closed, restore th~oss-tia valve to open posit or whee tha-core pover lave less ~w equal to does, nnt. apply.

vithin eoa hour. Specifi .0,4 Zt che a>>>>ac tha Etta te a>>ca>>ted aad cafe>>ca eatat tace the React>>a Coolant Systea, a Special Report shall be prepared and subcLitted to the Coemisa ion pursuant to S pecif ication 6.9. 2 vithin 90 days describing the circuaatancea of the actuation and the total accucaulatad actuation cyclea to date.

COOK HUCLEAR PLAHT - lAGT 2, 3/4 5-3 AHmmme HO. 167

PLANT SYSTEM COHOEHSATE STORAGE TAHK LIHITING COHO ITIOH FOR OPERATION 3.7.1.3 The condensate storage tank (CST) shall be OPERABLK with a minimum contained volume of , allons of water.

Zdd APPLICABILITY: HOOKS 1 3.

ACT'.5N:

Mfth the condensate storage tank inoperable, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either:

a. Restore the CST to OPERABLE status or be in HOT SHUTDOWN wfthfn he next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, ot

b. Oemonstrate the OPERABILITY of the Essentfal Service Water System as a backup supply to the auxiliary feedwater pumps and restore the condensate storage tank to OPERABLE status within 7 days or be in HOT SHUTDOMN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURYKILLAHCE RE UIREMENTS 4.7.1.3.1 The condensate storage tank shall be demonstrated OPERABLK at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying the contained water volume is wi afn fts limits when the tank fs the supply source for the auxiliary feedwater

'umps A

4.7.1.3.2 The Essential Service Mater System shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifyfng that the Essential Service Mater System fs in operation whenever the Kssential Service Mater System fs the supply source for the auxiliary feedwatcr pumps.

0. C. COOK - UNIT 2 3/4 7-7

3 4.2 tCVEX DISTIbUTION UMIT 3 4.2.4 AMAHT POVXR TI T RA 0 Th>> quadrant pover distribution satisfies tilt ratio li<<ic assures that the radial pover the design values used Ln the power capability

~ nalysis. kadial pover distribucion <<easure<<ence are <<ade during startup testing and periodically during pover operacion.

The lL<<it of 1.02 at vhich corrective action Ls required provides DA and linear heat generation race procection vith x.y piano pover tilts.

The tvo hour tLae allovance for eperation vith a tilt condition greater than 1.02 but less than 1.09 Ls provided to allov identification and correction of a dropped er <<Lsaligaed zo4. In the event such action does not correct the cilt, che aarjin for uncertainty'n F Ls teinstata4 by reducing che pover by 3 percenc fro<< 1ATXD THQNL NVEL Br each percent of tilt Ln excess of 1.0:

3 4.2.5 DN3 PARAMETERS The lL<<its on ~ M.related pa ters ensure that each of the para<<scars are ntaine ichin nor<<al steady.staW envelope o& ~

operation as 4 in transi and ac nt ananias. The less than or equal 57I.7 F pres iser pr ura pre r than o quito 2200 paig ar consisc c vith VFSAR uaptions have n analytica

<<o raced quate Lntain cere et above ~ design shout anal 4 trans c vith evance t <<assure<<en ertai . The gree than er qual te .9 F La c rvative co a

'afecy lysis rPolaed de<<ons ce that plant aay race on a Lne contro rograa re the lytical t of T t 100% RApD aay r froa 54 .a F to,5 .1 t. The t ot 54).9 ntains aazjin 1.1 t. e cote, be operace vith Lndica d vessel averse ce<<pere ~ at any slue becv n th>> upper lover 1 ts.

pres Lter pr~sure is Lce4 to single no<< 1 setpoint vith the lover 1 t ot th'syndicated alw set nt sec for Ln the ticacions. e T/5 valu ~ yas select fer c scency vi Lt 1 an@'tains a <<a n o 4 psi. +e liaL ~ consL nt vith as 'ions an4 n analytically de crated to be a4aquate co <<ainta che cor o e the applicable des lLILt ML values for each fuel type (vhich Lated Ln

( che bases fo ed tzansLent.

The 12 hour periodic surveillance ot these para<<etezs through Lnstru<<ent readouc Ls sufficient to ensure that the parameters are restored vithin their ltaita folloving load changes and other expected transient operation. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> surveillance of the tCS flov <<assure<<ent Ls adequate to detect flov degradation. The C24QACL CAUQhTZOt perforaad af ter refueling ensures che accuracy of the shiftly tlov <<assure<<ent. The cocal flov Ls <<easured after each refueling based on a secondary side calori<<attic and <<assure<<ants of pri<<ary loop te<<peratuzes.

I t

COOK NJCLXAk PLhXT QGT 2 lS/42 ~ 5 rumexaa so. SZ. 134

The limits on the OHB-related parameters ensure that each of parameters are maintained within the normal steady-state e assu in the transient and accident analy Th T to 81.0 and pressurizer pressure greater than or equal psfig (for nomina pressurizer operating pressure of 2235 psig) or pressurizer operating pressure of 2085 ps g are consiste psig) wi h the UFSAR a ssumptions and have been analytically demonstrated adeq te o maintain the core at or above the design OHBR throughout each analyz tr sient with allowance for measurement uncertainty. pressurizer pres ure is limited to either of two nominal operating pressures of 2235 psig 2 5 psig, with the corresponding indicated limits set forth in the specifi at> s. The limits are consistent with the UFSAR assumptions and have been ana tically demonstrated to be adequate to maintain the core it or bove the applicable design limit OHBR value for the current fuel tyype th rou out'ach analyzed transient.

1 All coeponenta Ln the Reactor Coolant %yeti are CeaLgae4 to vLthatand the effects ot cyclic loada Cue to oyster teaperature an4 preaaure chang@a. These cyclic loa4a are introduced by noraal load traneienta, reactor tripe, an4 etartup and ahutdovn operationa. The rarioua catagoriea of loa4 cyclea uaed for Ceaign purpoaea are provided Ln tection 4.1.4 of the tNR. ()uriay etartup and ahutdovn, the ratea of temperature and preaaure change>> are liaited ao that the aaxismum apecified heatup and cooldovn ratee are conaiatent vith the Ceaiga aaauaptioaa and aatiafy the etreaa lignite for cyclic operation.

An ZD or CO o~uarter thickneee aurface flav La poetulated at the location Ln the veaael vhich La found to be the 1Laitiay caee. There are several factora vhich influence the poatulated location. The theraal induced bend~

atreaa during heatup ia coapreaaive on the inner aurface vhile tenaile on the outer eurface of the veeael vail. Durin(( cooldovn, tbe bendix etreaa profile La refereed. In addition, the aaterial taeyhneee Le dependent upon irradiation and teeperature and t?~ore, the fluence profile through the reactor veaael vali, the rat ~ of heatup and alao the rate of oooldovn influence the poetulated flav location.

The heatup liaLt curve, PLyue 3. 6 2, La a coepoaite never vhLch vaa prepared by Ceteahnlsg the coen consecrative caae, vith either the inaLde or outaLde vali controllkay, for any heatup rate up to CO% per hour. The cooldovn linit curvoe of PLgere 3.4-3 are coepoaite carvea vhich vere prepared baaed upon the cane type analyaia vLth the exception that the controlliny locatLon ia alvaya the Lnaide vali vhere the cooldovn theaaal gradianta tend to produce tenaile atreeaea vhLle produciay caepreaeim etreeaes at the outeide vali. The heatup anC cooldovn nevoa vere geapazed baaed on the noet 1 tiny val the predicted adjuated reference teaperature at the en4 of XT?I.

The reactor naael aateriala have been teated to deteraLne the tial RT~c The reeulte of theee teeta are ahovn Ln Table I 3(i.4-1. Reactor vill operation and reeultant fact neutron (I s 1 )6ev) irradiation cauae an Lncreaae Ln the RT~. Therefore, an ad)ueted reference temperature anat be predicted Ln accordance vith Noqalatory Cuide 1.0%, Reriaion 2. Thia prediction baaed on the fluence an4 a cheaiatry factor deterained fnxa on ~ of Poaitiona preaented Ln the hagulatory Cuide. Poaition (1) Ceterninea cheaietry factor froa the copper and nickel content of the aateriai. poaition (2) utilixea ecrnillance data acta vhich relate the ahift Ln reference t+8+:ature oC aaznillance apeckaena to the fluence. The eelectioa of poaition (1) or (2) ia nade baae4 on the avail. ability of credible earveillance data, and the reaulta achieved Ln applying the tvo Poaitiona.

8 3/4 6 6 NQ M,444, a 7 ~

I (Continued) alloved conplecton tinea are reasonable, based on operating experience, to reach

~ required plane conditions from full pover conditions in an orderly aannar and vithouc challenging plant systems.

Kf aoze chan one accuuulitor ia tnoperable, the plant ia in a condition outside tha accident analyaea; therefore, LCO 3.0.3 aust be entered taaadtately.

The OPEBhSILXTY of cvo independent ECCS subsystems anaur<<a that sufficient emergency coze cooling capability vill be available tn the event of a LOCh a!waxing the loss of one subaystei through any single failure conaideracion.

ither subsystem operating in con)unction vich the accumulators is capable of supplying sufficient coze cooling co ltatc cha peak cladding temperatures vithin acceptable 1tntca for all poaculated break sizes zangtng from the double ended break of che largest RCS cold leg pipe dovnvard. Xn addition, each ECCS subsystem provides long cena cora cooling capability tn che zacirculacion soda during the accident reccnrery period.

"~f a safec jectton cross-tie e is osed, safety in oa vould be limit o tvo lines ass the loss of one safe ec n subsy rou a gle failure co aration. The resul overed ov r es a decrease o limit the pe teaperaturc acceptable limits in I

'lKRNAL the event a postulated asall break URL

COHTAIHKEHT SYSTBtS SASES 3/4.6.1.4 IHTERVAL PRESSURE The limitations on conCainment internal pressure ensure that 1) the contafnmenC structure is prevented from exceeding fts design negative pressure differential with respect to che outside aCmosphere of 8 psfq and

2) the contafreent peak pressure does not exceed the design pressure of 12 sfg during LOCA conditions.

he maximum peak r' x ec ed to be obCained from a LOCA event 1 s psf e it .3 for ti tiv ta pre ur i press 1 c e is nsis t pre wfthM to . p 1 c es n~

wL'~ ~~1w ~g Q,g '/Xi'+t )wc4rwl p oSq~~ co~a 3/4.6.1.5 AIR e >< ssu~

The lfmf tatfons on containmenc average afr temperature ensure that 1) the COntafrvnent afr maSS 1S limited tO an inf fal maSS SufffCfently lOw CO prevent exceeding the design pressure during LOCA conditions and 2) the ambient, air temperature does not exceed that temperature allowable for the

~

continuous duCy rating specfffed for equipment and 1nstrumentatfon located wf Chin contaireent.

/IM The containment pressur r fs sensitive to the initially contained air mass during OCA. The contained af'r mass increases with conditions decreasing temperatur . he lower temperature limit of 60'f'fll limit the peak pressure to psfg which fs less than the containment desfgn pr essure of 12 psfg, he upper temperature lfmft fnfluences the peak accfdent temperature slightly durfng a LOCA; however, this lfmft fs based primer 11y upon equfpeent protection and antfcf pated operating .

8oth the upper and lower tanoeratur e limits are consistent wfth the para-meters used fn the accfdenC analyses.

3/4.6.1.6 CONTAIHM HT 9 RA IHT R TY Thfs lfmftac1on ensures that the sCruccural fntegrfty of the con-tainment will be mafntafned comparable to the orfgfnal design standards for che 1 ffe of the fac11ity. Scructural integrity fs required to ensure that (1) the steel liner remafns leak tight and (2) the concrete surround-ing the steel liner reefns capable of provfdfng external mfssfle protec-tion for the sceel lf<<er and.radfat1on shielding fn the evenc of a LOCA.

A v1sual inspection fn con)unction with Type A leakage tests fs suff1cfent to denonstrate Chfs capability.

0, I r tlhv <<srtw

l PLANT SYSTEMS CONOENSATE STORAGE TANK LIMIT.'NG CONOITION FOR'PERATION 3.7.1.3 The condensate storage tank (CST) shall be OPKRABLE with a minimum volume of 175,000 gallons of water.

u sa.gl A . AB:LITY: HOOKS 1, 2 and 3.

ACT:QM:

Mith the condensate storage tank inoperable, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either:

a. Restore the CST to OPERABLE status or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, or

b. Oemonstrate the OPERABILITY of the Essential Service Mater System as a backup supply to the auxiliary feecwater oumps and restore the condensate storage tank to OPERABLE status within 7 days or be in HOT SHUTGOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />..

SURVEILLANCE RE UIREMENTS us~/ lq 4.7.1.3.1 The CandenSate StOrage tank Shall e' demanStrated OPERABLE at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 'by verifying the wat r volume is within its limi ts wnen the tank is the supply source for the auxiliary feedwater pumps'.7.1.3.2 The Essential Service Mate~ System shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying .hat the Essential Service Mater System is in operation whe~ever the Essential Service Mater System is the supply source for the auxiliary feedwater pumps.

0. C. COOK - UNIT 2 3/4 7-7

o ~

0 ctf5c kn>ge PiPc ~

The OPERABILITY of the condensate storage tank with the ~num water volume ensures that su ficient water is available to maintain chai RCS at HOT STANDBY conditions for 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> with steam discharge to the atmosphere concurrent

'trh total loaa of off-site pover. The water volume lb'.t~haalteJaa-an i n allovm~eor vatar tot uaatli beoauea of tatf't aoEiirre characteristics i's ~+ <<elf, lit ~ looitfot or otiier wP 0 'hysical

~ ~

'Pv 4 4 uSa o 2 The limitations on secondary system specific activity ensure that the resultant off-site radiation dose will be l.imited to a small fraction of 10 CPR Part 100 limits in the event of a steam line rupture. This dose also includes the effects of a coincident 1.0 gpm primary to secondary tube leak in the steam generator of the affected steam line. These values are consistent with thc assumptions used in rhe accident analyses.

The OPERABILITY of the steam generator stop valves ensures that no more than one steam generator will blowdown in the event of a steam line rupcure.

This restriction is required to 1) minimire the positive reactivity effects of the Reactor Coolant System cooldown associated with rhe blowdown, and 2) limit rhe pressure rise within containment in the event the steam line rupture occurs within containment. The OPEMILITY of the steam generator stop valves within the closure times of the surveillance requirements are consistent with the assumptions used in the accident analyses.

Virh one steam generator stop valve inoperable in MODE 1, action must be taken to restore OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Some repairs to the valves can be made with the unit hot. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> completion time is reasonable, considering the low probability of an accident occurring during this time period that would require a closure of the steam generator stop valves. If the steam generator stop valve cannot be restored to OPERABLE status within 8 hours, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the p a~ad in MODE 2 within 6 heu.i ....~~u 2 and 3 action statement entered. The completion times are reasonable, based on operating experience, to reach MODE 2 and to close the steam generator stop valves in an orderly manner <<nd wirhout challenging unit systems.

COOK NUCLEAR PLANT - UNIT 2 B 3(4 7-3 AMENDHBlT NO. ~ ~

~

~ >

I

ATTACHMENT 6 TO AEP:NRC:1223 DESCRZPTZON OF ANALYSES PERFORMED BY WESTZNGHOUSE ELECTRZC CORPORATZON FOR COOK NUCLEAR PLANT UNZT 2

i to AEP:NRC:1207 Page 1 WCAP 14489

f II 6