Revision as of 06:36, 26 April 2019

Text

Proposed Tech Specs Re Analyses Supported Increases in Unit 1 SG Tube Plugging Limit & Maintenance of Consistency of Unit 2 Acceptance Criteria
	ML17332A790
Person / Time
Site:	Cook
Issue date:	05/26/1995
From:	; INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	;
Shared Package
ML17332A791	List: ML17332A790; ML17334B548;
References
	NUDOCS 9506130053
	Download: ML17332A790 (889)
	v • d • e

2.0 SAHsTY LIMITS AND LIMITING SAIKIY SYSTEM SETTINGS 660 650 2400 a UNACCtIPrhBLE 640 630 21 pala U-620 6 i 0 600 590 580 570 0'.2 0.4 0.6 0.8 Power (fraction of rated thermal power)1.2 PRESSURE Q%hl 1840 2000 2100 2400 (0.02, 62086)>>(0.02, 632.?9), (0.02, 63985)>>(0.02, 64986)>>(0.02, 65982)>>(1.136, 586.17), (1.094, 60021), (1.068, 608.72), (1.031>>620 83)>>(0.996, 632.42), (1Q 57784)(12, 58682)(1+591.77)(1+599AO)(1Z, 606.63)Figure 2.1-1 Reactor Core Safety Limits page 2-2 COOK NUCLEAR PLANT-UNIT 1 9506l30053 950526 PDR ADQCK 050003l5'PDR AMENDMENT W>>4>>~>>~

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2.0 SMARTY

LIMITS AND LIMITING SAFETY SYSTEM SETTINGS TABL'E 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT 1.Manual Reactor Trip 2.Power Range, Neutron Flux 3.Power Range, Neutron Flux, High Positive Rate.4.Power Range, Neutron Flux, High Negative Rate TRIP SETPOINT Not Applicable Low Setpoint-less than or equal to 25%of RATED THERMAL POWER-High Setpoint-less than or equal.to 109%of RATED THERMAL POWER Less than or equal to 5%of RATED THERMAL POWER with a time constant greater than or equal to 2 seconds Less than or equal to 5%of RATED THERMAL POWER with a time constant greater than or equal to 2 seconds ALLOWABLE VALUES Not Applicable Low Setpoint-less than or equal to 26%of RATED THERMAL POWER High Setpoint-less than or equal to 110%of RATED THERMAL POWER Less than or equal to 5.5%of RATED THERMAL POWER with a time constant greater than or equal to 2 seconds Less than or equal to 5.5%of RATED THERMAL POWER with a time constant greater than or equal to 2 seconds 5.Intermediate Range, Neutron Flux Source Range, Neutron Flux Less than or equal to 25%of RATED THERMAL POWER Less than or equal to 10'ounts per second Less than or equal to 30%of RATED THERMAL POWER Less than or equal to 1.3 x 10'ounts per second 7.Overtemperature Delta T 8.Overpower Delta T See Note 1 See Note 2 See Note 3 See Note 4 9.Pressurizer Pressure-Low Greater than or equal to 1875 psig Greater than or equal to 1865 psig 10.Pressurizer Pressure-High 11.Pressurizer Water Level--High 12.Loss of Flow Less than or equal to 2385 psig Less than or equal to 92%of instrument span Greater than or equal to 90%of design flow per loop*Less than or equal to 2395 psig Less than or equal to 93%of instrument span Greater than or equal to 89.1%of design flow per loop**Design flow is I/4 Reactor Coolant System total fiow rate from Table 3.2-1.COOK NUCLEAR PLANT-UNIT I Page 2-5 AMENDMENT$4, 426, 483

O O TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION 1+~,s Note 1: OvertemperaturehT 6 dT,[K,-K-'T-T')+K (P-Pg-f, (41)]1+~p where: hT, Indicated hT at RATED THERMAL POWER p/1+,s 1+st Average temperature,'F Indicated T,~at RATED THERMAL POWER ((576.3'F)Pressurizer pressure, psig Indicated RCS nominal operating pressure (2235 psig or 2085 psig)The function generated by the lead-lag controller for T,~dynamic compensation Time constants utilized in the lead-lag controller for T,~Tt=22 secs.v>=4 secs.Laplace transform operator I

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATIONS Continued Operation with 4 Loops K,=1.17 Ki=0.0230 K3=0.00110 and f,(hl)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 q,-q, between-37 percent and+3 percent, f,(hl)=0 (where q, and q, are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and q,+q, is total THERMAL POWER in percent of RATED THERMAL POWER).(ii)For each percent that the magnitude of (q,-rg exceeds-37 percent, the d, T trip setpoint shall be automatically reduced by 0.33 percent of its value at RATED THERMAL POWER.(iii)For each percent that the magnitude of (q,-qQ exceeds+3 percent, the hT trip setpoint shall be automatically reduced by 2.34 percent of its value at RATED THERMAL POWER.

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION Continued Note2: Overpowers,T K hT,[K-K-T-K (T-T)-f (hi)]wsS I I+i S 3 where: Indicated d,T at RATED THERMAL POWER Average temperature,'F Indicated T,~at RATED THERMAL POWER ((563.0'F)1.083 Ks 0.0177/'F for increasing average temperature and 0 for decreasing average temperature 0.0015 for T>T";+=0 for T 6 T"~~S I+~qS~3 The function generated by the rate lag controller for T,~dynamic compensation Time constant utilized in the rate lag controller for T,~vs=10 secs.S=Laplace transform operator f, (BI)=0 s p Note 3: The channel's maximum trip point shall not exceed its computed trip point by more than 3.4 percent d, T span.Note 4: The channel's maximum trip point shall not exceed its computed trip point by more than 2.5 percent hT span.

In In I 0 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN-TAVG GREATER THAN 200'F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.3%Delta k/k.APPLICABILITY:

MODES 1, 2', 3, and 4.ACTION: With the SHUTDOWN MARGIN less than 1.3%Delta k/k, immediately initiate and continue boration at greater than or equal to 10 gpm of a solution containing greater than or equal to 20,000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.SURVEILLANCE RE UIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.3%Delta k/k: Within one hour after detection of an inoperable control rods(s)and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod(s)is inoperable.

If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s).b.When in MODE 1 or MODE 2 with Keff greater than or equal to 1.0, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that control bank withdrawal is within the limits of Specification 3.1.3.5.C.When in MODE 2 with Keff less than 1.0, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.5.Prior to initial operation above 5%RATED THERMAL POWER after each fuel loading, by consideration of the factors of e below, with the control banks at the maximum insertion limit of Specification 3.1.3.5.See Special Test Exception 3.10.1.COOK NUCLEAR PLANT-UNIT 1 Page 3/4 1-1 AMENDMENT&, 4A, 448

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.1 REACTIVITY CONTROL SYSTEMS CHARGING PUMP-SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 a.One charging pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency bus.b.One charging flowpath associated with support of Unit 2 shutdown functions shall be available.

APPLICABILITY:

Specification 3.1.2.3.a.

-MODES 5 and 6 Specification 3.1.2.3.b.

-At all times when Unit 2 is in MODES 1, 2, 3, or 4.ACTION: With no charging pump OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes." With more than one charging pump OPERABLE or with a safety injection pump(s)OPERABLE when the temperature of any RCS cold leg is less than or equal to 152'F, unless the reactor vessel head is removed, remove the additional charging pump(s)and the safety injection pump(s)motor circuit breakers from the electrical power circuit within one hour.The provisions of Specification 3.0.3 are not applicablc.

In addition to the above, when Specification 3.1.2.3.b is applicable and the required flow path is not available, return the required flow path to available status within 7 days, or provide equivalent shutdown capability in Unit 2 and return the required flow path to available status within the next 60 days, or have Unit 2 in HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and HOT SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .e.The requirements of Specification 3.0.4 are not applicable when Specification 3.1.2.3.b applies.SURVEILLANCE RE UIREMENTS 4.1.2.3.1 The above required charging pump shall be demonstrated OPERABLE by verifying, that on recirculation flow, the pump develops a differential pressure of greater than or equal to 2290 psid when tested pursuant to Specification 4.0.5.A maximum of one centrifugal charging pump shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 152'F."For purposes of this specification, addition of water from the RWST does not constitute a positive reactivity addition provided the boron concentration on the RWST is greater than the minimum required by Specification 3.1.2.7.b.2.

COOK NUCLEAR PLANT-UNIT 1 Page 3/4 1-11 AMENDMENT 98, 420, 434, 444, 447 0

3/4 LIMITING CONDITIONS FOR OPE'RATION AND SURVEILLANCE REQUIREMENTS 3/4.1 REACTIVITY CONTROL SYSTEMS CHARGING PUMPS-OPERATING LIMITING CONDITION FOR OPERATION 3.1.2A At least two charging pumps shall be OPERABLE.APPLICABILITY:

MODES 1, 2, 3 and 4.ACTION: With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ;restore at least two charging pumps to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .SURVEILLANCE RE UIREMENTS 4.1.2.4 At least two charging pumps shall be demonstrated OPERABLE by verifying that on recirculation flow, each pump develops a differential pressure of greater than or equal to 2290 psid when tested pursuant to Specification 4.0.5., COOK NUCLEAR PLANT-UNIT 1 Page 3/4 1-12 lf 0 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.1 REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES-SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.7 As a minimum, one of the following borated water sources shall be OPERABLE: A boric acid storage system and associated heat tracing with: 1.A minimum usable borated water volume of 4300 gallons, 2.Between 20,000 and 22,500 ppm of boron, and 3.A minimum solution temperature of 145'F.The refueling water storage tank with: 1.A minimum usable borated water volume of 90,000 gallons, 2.A minimum boron concentration of 2400 ppm, and 3.A minimum solution temperature of 70'F.APPLICABILITY:

ACTION: MODES 5 and 6.With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes'ntil at least one borated water source is restored to OPERABLE status.SURVEILLANCE RE UIREMENTS 4.1.2.7 The above required borated water source shall be demonstrated OPERABLE: At least once per 7 days by: 1.2.3.Verifying the boron concentration of the water, Verifying the water level volume of the tank, and Verifying the boric acid storage tank solution temperature when it is the source of borated water.At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when it is the source of borated water.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 3.1.2.7.b.2.

f COOK NUCLEAR PLANT-UNIT 1 Page 3/4 1-15 f:,!

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.1 REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES-OPERATIONS LIMITING CONDITION FOR OPERATION 3.1.2.8 Each of the following borated water sources shall be OPERABLE: b.A boric acid storage system and associated heat tracing with: 1.A minimum usable borated water volume of 5650 gallons, 2.Between 20,000 and 22,500 ppm of boron, and 3.A minimum solution temperature of 145'F.The refueling water storage tank with: 1.A minimum contained volume of 350,000 gallons of water, 2.Between 2400 and 2600 ppm of boron, and 3.A minimum solution temperature of 70'F.APPLICABILITY:

MODES 1, 2, 3 and 4.ACTION: With the boric acid storage system inoperable, restore the storage system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWN MARGIN equivalent to at least 1%d,k/k at 200'F;restore the boric acid storage system to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .b.With the refueling water storage tank inoperable, restore the tank to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .SURVEILLANCE RE UIREMENTS 4.1.2.8 Each borated water source shall be demonstrated OPERABLE: COOK NUCLEAR PLANT-UNIT 1 Page 3/4 1-16 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.2 POWER DISTRIBUTION LIMITS TABLE 3.2-1 DNB PARAMETERS PARAMETER Reactor Coolant System Tavg Pressurizer Pressure Reactor Coolant System Total Flow Rate LIMITS 4 Loops in Operation at RATED THERMAL POWER 6 579.3'F R 2050 psig">341,100 gpm" Indicated average of at least three OPERABLE instrument loops."Limit not applicable during either a THERMAL POWER ramp increase in excess of 5 percent RATED THERMAL POWER per minute or a THERMAL POWER step increase in excess of 10 percent RATED THERMAL POWER."'Indicated value.COOK NUCLEAR PLANT-UNIT 1 Page 3/4 2-14 AMENDMENT 94, 420, 426,~

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 3.3-3 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION FUNCTIONAL UNIT f.Steam Line Pressure-Low MINIMUM TOTAL NO.OF CHANNELS CHANNELS APPLICABLE CHANNELS TO TRIP OPERABLE MODES ACTION Four Loops Operating Three Loops Operating 1 pressure/loop 1 pressure/operating loop 2 pressures any loops 1"'ressure in any operating loop 1 pressure 1,2,3" any 3 loops 1 pressure in 3" QIly 2 operating loops 14 15 COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-17 AMENDMENT$4, 420,~

t 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRVMENTATION TABLE 3.3-3 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION FUNCTIONAL UNIT COINCIDENT WITH MINIMUM TOTAL NO.OF CHANNELS CHANNELS APPLICABLE CHANNELS TO TRIP OPERABLE MODES ACTION T,~-Low-Low Four Loops Operating 1-T,+oop 2 Ta<any loops 1 T,~any 3 1, 2, 3'~loops 14 Three Loops Operating 1 T,+operating loop 1" T,~in I T,~inany 3'ny operating two operating loop loops 15 e.Steam Line Pressure-Low Four Loops Operating 1 pressure/loop 2 pressures any loops 1 pressure 1, 2,'3" any 3 loops 14'hree Loops Operating 5.TURBINE TRIP 8c FEEDWATER ISOLATION 1 pressure/operating loop 1"'ressure ln any operating loop 1 pressure in 3" any 2 operating loops 15 a.Steam Generator Water Level-High-High 3/loop 2/loop in any 2/loop in operating each loop operating loop 1,2,3 14 COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-21 AMENDMENT$4, 440, 4' I'

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION ENGINEERED SAFETY FEATURES INTERLOCKS DESIGNATION P-11 P-12 CONDITION AND SETPOINT With 2 of 3 pressurizer pressure channels greater than or equal to 1915 psig.With 2 of 4 T,~channels less than or equal to Setpoint.Setpoint greater than or equal to 541'F FUNCTION P-11 prevents or defeats manual block of safety injection actuation on low pressurizer pressure.P-12 allows the manual block of safety injection actuation on low steam line pressure causes steam line isolation on high steam flow.Affects steam dump blocks.With 3 of 4 T,~channels above the reset point, prevents or defeats the manual block of safety injection actuation on low steam line prcssure.COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-23a 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4,3 INSTRUMENTATION TABLE 3.3C ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT 1.SAFETY INJECTION, TURBINE TRIP, FEEDWATER ISOLATION, AND MOTOR DRIVEN FEEDWATER PUMPS TRIP SETPOINT ALLOWABLE VALUES a.Manual Initiation

See Functional Unit 9-b.Automatic Actuation Logic c.Containment Pressure-High Not Applicable Less than or equal to 1.1 psig Not Applicable Less than or equal to 1.2 psig d.Pressurizer Pressure-Low Greater than or equal to 1815 psig Greater than or equal to 1805 psig e.Differential Pressure Between Steam Lines-High f.Steam Line Pressure-Low Less than or equal to 100 psi Greater than or equal to 500 psig steam line pressure Less than or equal to 112 psi Greater than or equal to 480 psig steam line pressure COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-24 AMENDMENT 40, 426,~

3/4 LIMI'HNG CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 3.3P Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT 2.Containment Radio-activity-High Train A (VRS-1101, ERS-1301, ERS-1305)3.Containment Radio-activity-High Train B (VRS-1201, ERS-1401, ERS-1405)4.STEAM LINE ISOLATION TRIP SETPOINT See Table 3.34 See Table 3.3-6 ALLOWABLE VALUES Not Applicable Not Applicable a.Manual---------See Functional Unit 9-b.Automatic Actuation Logic I c.Containment Pressure-High-High d.Steam Flow in Two Steam Lines-High Coincident with T,~-Low-Low e.Steam Line Pressure-Low 5.TURBINE TRIP AND FEEDWATER ISOLATION Not Applicable Less than or equal to 2.9 psig Less than or equal to 1.42 x 10~lbs/hr from 0%load to 20%load.Linear from 1.42 x 10'bs/hr at 20%load to 3.88 x 10'bs/hr at 100%load.T,~greater than or equal to 541'F Greater than or equal to 500 psig steam line pressure Not Applicable Less than or equal to 3 psig Less than or equal to 1.56 x 10 lbs/hr from 0%load to 20%load.Linear from 1.56 x 10'bs/hr at 20%load to 3.93 10'bs/hr at 100%load.T,~greater than or equal to 539'F Greater than or equal to 480 psig steam line pressure a.Steam Generator Water Level-High-High Less than or equal to 67%of narrow-range instrument span each steam generator Less than or equal to 68%of narrow-range instrument span each steam generator COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-26 AMENDMENT 94, 436,~

iI 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS FUNCTIONAL UNIT 1.SAFETY INJECTION, TURBINE TRIP, FEEDWATER ISOLATION, AND MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS TRIP ACTUATING MODES IN CHANNEL DEVICE WHICH CHANNEL CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE CHECK CALIBRATION TEST a.Manual Initiation b.Automatic Actuation Logic N.A.c.Containment Pressure-High d.Pressurizer Pressure-Low S e.Differential Pressure Between Steam Lines-'igh f.Steam Line Pressure-Low S 2.CONTAINMENT SPRAY a.Manual Initiation N.A.See Functional Unit 9 M(2)M(3)M See Functional Unit 9 N.A.N.A.N.A.N.A.N.A.1,2,3,4 1, 2, 3 1, 2, 3 1, 2, 3 1, 2, 3 b.Automatic Actuation Logic N.A.c.Containment Pressure-High-High N.A.M(2)M(3)N.A.N.A.I, 2, 3, 4 1,2,3 COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-31 AMENDMENT 400)420)XRk) 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4.3-2 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS FUNCTIONAL UNIT 4.STEAM LINE ISOLATION TRIP ACTUATING MODES IN CHANNEL DEVICE WHICH CHANNEL CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE CHECK CALIBRATION TEST a.Manual b.Automatic Actuation Logic N.A.c.Containment Pressure-High-High d.Steam Flow in Two Steam S Lines-High Coincident with T~-Low-Low e.Steam Line Pressure-Low S 5.TURBINE TRIP AND FEEDWATER ISOLATION N.A.See Functional Unit 9 M(2)M(3)M N.A.N.A.N.A.N.A.1, 2, 3, 1, 2, 3 1, 2, 3 I, 2, 3 a.Steam Generator Water Level-High-High 6.MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS N.A.1, 2, 3 c.Safety Injection N.A.d.Loss of Main Feed Pumps N.A.a.Steam Generator Water Level-Low-Low b.4 kv Bus Loss of Voltage S N.A.N.A.M(2)N.A.N.A.N.A.N.A.1, 2, 3 1, 2, 3 1, 2, 3 1,2 COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-33 AMENDMENT 400, 420, 424, 444, 4' t

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM SAFETY VALVES-SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2 A minimum of one pressurizer code safety valve shall be OPERABLE with a lift setting of 2485 PSIG J 3%.I APPLICABILITY:

ACTION: MODES 4 and 5.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 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 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 4Q

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM SAFETY VALVES-OPERATING LIMITING CONDITION FOR OPERATION 3.4.3 All pressurizer code safety valves shall be OPERABLE with a lift setting of%85 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .SURVEILLANCE RE UIREMENTS 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 of the valve at nominal operating temperature and pressure.COOK NUCLEAR PLANT-UNIT 1 Page 3/4 4-5

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)SURVEILLANCE RE UIREMENTS Continued d.At least once per 18 months by: 1.Verifying automatic isolation and interlock action of the RHR system from the Reactor Coolant System when the Reactor Coolant System pressure is above 600 psig.2.A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.)show no evidence of structural distress or abnormal corrosion.

e.At least once per 18 months, during shutdown, by: 1.Verifying that each automatic valve in the flow path actuates to its correct position on a Safety Injection test signal.2.Verifying that each of the following pumps start automatically upon receipt of a safety injection signal: a)Centrifugal charging pump b)Safety injection pump c)Residual heat removal pump f.By verifying that each of the following pumps develops the indicated differential pressure on recirculation flow when tested pursuant to Specification 4.0.5.1.Centrifugal charging pump greater than or equal to 2290 psid 2.Safety injection pump greater than or equal to 1326 psid 3.Residual heat removal pump greater than or equal to 150 psid By verifying the correct position of each mechanical stop for the following Emergency Core Cooling System throttle valves: 1.Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE.COOK NUCLEAR PLANT-UNIT 1 Page 3/4 5-5 AMENDMENT 4P, 426, 444, 448, 444

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4,5 EMERGENCY CORE COOLING SYSTEMS (ECCS)REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.5 The refueling water storage tank (RWST)shall be OPERABLE with: A minimum contained volume of 350,000 gallons of borated water.Between 2400 and 2600 ppm of boron, and A minimum water temperature of 70'F.APPLICABILITY:

ACTION: MODES 1, 2, 3 and 4.With the refueling water storage tank inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .SURVEILLANCE RE UIREMENTS 4.5.5 The RWST shall be demonstrated OPERABLE: a.At least once per 7 days by: 1.Verifying the contained borated water level in the tank, and Verifying the boron concentration of the water.b.At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature.

COOK NUCLEAR PLANT-UNIT 1 Page 3/4 5-11 AMENDMENT 5B, 444 f

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.7 PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued 4.7.1.2 Each auxiliary feedwater pump shall be demonstrated OPERABLE when tested pursuant to Specification 4.0.5 by: a.Verifying that each motor driven pump develops an equivalent discharge pressure of greater than or equal to 1240 psig at 60'F in recirculation flow.Verifying that the steam turbine driven pump develops an equivalent discharge pressure of greater than or equal to 1180 psig at 60'F and at a flow of greater than or equal to 700 gpm when the secondary steam supply pressure is greater than 310 psig.The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.Verifying that each non-automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in its correct position.Verifying that each automatic valve in the flow path is in the fully open position whenever the auxiliary feedwater system is placed in automatic control or when above 10%RATED THERMAL POWER.This requirement is not applicable for those portions of the auxiliary feedwater system being used intermittently to maintain steam generator water level.e.Verifying at least once per 18 months during shutdown that each automatic valve in the flow path actuates to its correct position upon receipt of the appropriate engineered safety features actuation test signal required by Specification 3/4.3.2.Verifying at least once per 18.months during shutdown that each auxiliary feedwater pump starts as designed automatically upon receipt of the appropriate engineered safety features actuation test signal required by Specification 3/4.3.2.Verifying at least once per 18 months during shutdown that the unit cross-tie valves can cycle full travel.Following cycling, the valves will be verified to be in their closed positions.

COOK NUCLEAR PLANT-UNIT 1 Page 3/4 7-6 AMENDMENT 400, 43k, 444, 464

5.0 DESIGN

FEATURES 5.4 REACTOR COOLANT SYSTEM Continued a.In accordance with the code requirements specified in Section 4.1.6 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.For a pressure of 2485 psig, and c.For a temperature of 650'F, except for the pressurizer which is 680'F.VOLUME 5.4.2 The total contained volume of the reactor coolant system is approximately 12,466 cubic feet at 0%steam generator tube plugging and 11,551 cubic feet at 30%steam generator tube plugging at a nominal T,, of 70'F.5.5 EMERGENCY CORE COOLING SYSTEMS 5.5.1 The emergency core cooling systems are designed and shall be maintained in accordance with the original design provisions contained in Section 6.2 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements, with one exception.

This exception is the CVCS boron makeup system and the BIT.5.6 FUEL STORAGE CRITICALITY

-SPENT FUEL 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with: ae A k,it equivalent to less than 0.95 when flooded with unborated water.A nominal 8.97 inch center-to~nter distance between fuel assemblies placed in the storage racks.The fuel assemblies will be classified as acceptable for Region 1, Region 2, or Region 3 storage based upon their assembly average burnup versus initial nominal enrichment.

Cells acceptable for Region 1, Region 2, and Region 3 assembly storage are indicated in Figures 5.6-1 and 5.6-2.Assemblies that are acceptable for storage in Region 1, Region 2, and Region 3 must meet the design criteria that define the regions as follows: COOK NUCLEAR PLANT-UNIT 1 Page 5-5 AMENDMENT 468, 44$, 469

BASES 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS BASES 4 Loop Operation Westinghouse Fuel (15x15 OFA)(WRB-1 Correlation)

Correlation Limit Design Limit DNBR Safety Analysis Limit DNBR Typical Cell 1.17 1.23 1.40 Thimble Cell" 1.17 1.22 1.42 The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than the applicable design DNBR limit, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.represents typical fuel rod"represents fuel rods near guide tube COOK NUCLEAR PLANT-UNIT 1 Page B 2-1(a)AMENDMENT 74, 4%, 4%

I'I'l, BASES 2.0 SALTY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SAFETY LIMITS BASES The Power Range Negative Rate Trip provides protection for control rod drop accidents.

At high power, a rod drop accident could cause local flux peaking which could cause an unconservative local DNBR to exist.The Power Range Negative Rate Trip will prevent this from occumng by tripping the reactor.No credit is taken for operation of the Power Range Negative Rate Trip for those control rod drop accidents for which the DNBRs will be greater than the applicable design limit DNBR-value for each fuel type.Intermediate and Source Ran e Nuclear Flux The Intermediate and Source Range, Nuclear Flux trips provide reactor core protection during reactor startup.These trips provide redundant protection to the low setpoint trip of the Power Range, Neutron Flux channels.The source Range Channels will initiate a reactor trip at about 10+'ounts per second, unless manually blocked when P-6 becomes active.The Intermediate Range Channels will initiate a reactor trip at a current level proportional to approximately 25 percent of RATED THERMAL POWER unless manually blocked when P-10 becomes active.No credit was taken for operation of the trips associated with either the Intermediate or Source Range Channels in the accident analyses;however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Protection System.Overtem erature Delta T The Overtemperature delta T trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds), and pressure is within the range between the High and Low Pressure reactor trips.This 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.

With normal axial power distribution, this reactor trip limit is always below the core safety limit as shown in Figure 2.1-1.If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the notations in Table 2.2-1.COOK NUCLEAR PLANT-UNIT I Page B 2Q AMENDMENT V4, 426

BASES 2.0 SAHH Y LIMITS AND LIMITING SAFETY SYSTEM SETTINGS LIMITING SAFETY SYSTEM SETTINGS BASES Ove ower 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 steamline break events.Credit was taken for operation of this trip in the steamline 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 RCCA Withdrawal at Power event.COOK NUCLEAR PLANT-UNIT I Page B 2-5 AMENDMENT 420, 4%i, 4$S 3/4 BASES 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1)the reactor can be made subcritical from all operating conditions, 2)the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3)the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T,~.The most restrictive condition occurs at EOL, with T,~at no load operating temperature, and is associated with a postulated steam line break accident and resulting uncontrolled RCS cooldown.In the analysis of this accident, a minimum SHUTDOWN MARGIN of 1.3%Delta k/k is initially required to control the reactivity transient and automatic ESF is assumed to be available.

With T,~less than 200'F, the'eactivity transients resulting from a postulated steam line break cooldown are minimal and a 1%Delta k/k SHUTDOWN MARGIN provides adequate protection for this event.The SHUTDOWN MARGIN requirements are based upon the limiting conditions described above and are consistent with FSAR safety analysis assumptions.

3/4.1.1.3 BORON DILUTION A minimum flow rate of at least 2000 GPM provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual during boron concentration reductions in the Reactor Coolant System.A flow rate of at least 2000 GPM will circulate an equivalent Reactor Coolant System volume of 12,612 plus or minus 100 cubic feet in approximately 30 minutes.The reactivity change rate associated with boron reductions will therefore be within the capability for operator recognition and control.3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT TC The limitations on MTC are provided to ensure that the assumptions used in the accident and transient analyses remain valid through each fuel cycle.The surveillance requirement for measurement of the MTC at the beginning, and near the end of each fuel cycle is adequate to confirm the MTC value since this coefficient changes slowly due principally to the reduction in RCS boron COOK NUCLEAR PLANT-UNIT 1 Page B 3/4 1-1 AMENDMENT V4, 4A, 448 3/4 BASES 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with all reactor coolant loops in operation, and maintain DNBR above the safety analysis limit during all normal operations and anticipated transients.

A loss of flow in two loops will cause a reactor trip if operating above P-7 (11 percent of RATED THERMAL POWER)while a loss of flow in one loop will cause a reactor trip if operating above P-8 (31 percent of RATED THERMAL POWER).In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat;however, single failure considerations require that two loops be OPERABLE.Three loops are required to be OPERABLE and to operate if the control rods are capable of withdrawal and the reactor trip breakers are closed.The requirement assures adequate DNBR margin in the event of an uncontrolled rod withdrawal in this mode.In MODES 4 and 5, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat;but single failure considerations require that at least two loops be OPERABLE.Thus, if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE.The operation of one Reactor Coolant Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.The restrictions on starting a Reactor Coolant Pump below P-7 with one or more RCS cold legs less than or equal to 152'F are provided to prevent RCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50.The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by either (1)restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into or (2)by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50'F above each of the RCS cold leg temperatures.

COOK NUCLEAR PLANT-UNIT I Page B 3/4 4-1 AMENDMENT 88, 420, 467.

I f 3/4 BASES 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.5 REFUELING WATER STORAGE TANK The OPERABILITY of the RWST as part of the ECCS ensures that sufficient negative reactivity is injected into the core to counteract any positive increase in reactivity caused by RCS system cooldown, and ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA.Reactor coolant system cooldown can be caused by inadvertent depressurization, a loss of coolant accident or a steam line rupture.The limits on RWST minimum volume and boron concentration ensure that 1)sufficient water is available within containment to permit recirculation cooling flow to the core, and 2)the reactor will remain subcritical in the cold condition following a LOCA assuming mixing of the RWST, RCS, ECCS water, and other sources of water that may eventually reside in the sump, with all control rods assumed to be out.These assumptions are consistent with the LOCA analyses.The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 7.6 and 9.5 for the solution recirculated within containment after a LOCA.This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The ECCS analyses to determine F<limits in Specifications 3.2.2 and 3.2.6 assumed a RWST water temperature of 70'F.This temperature value of the RWST water determines that of the spray water initially delivered to the containment following LOCA.It is one of the factors which determines the containment back-pressure in the ECCS analyses, performed in accordance with the provisions of 10 CFR 50.46 and Appendix K to 10 CFR 50.COOK NUCLEAR PLANT-UNIT 1 Page B 3/4 5-3 AMENDMENT$3, 420, 4$8 l,.0 3/4 BASES 3/4.6 CONTAINMENT SYSTEMS 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 11.49 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 11.49 psig which is 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

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

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN T v GREATER THAN 200 F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.3%Delta k/k.APPLICABILITY:

MODES 1, 2, 3, and 4.ACTION: With the SHUTDOWN MARGIN less than 1.3%Delta k/k, immediately initiate and continue boration at greater than or equal to 10 gpm of a solution containing greater than or equal to 20,000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.SURVEILLANCE RE UIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.3%Delta k/k: 4 a.Within one hour after detection of an inoperable control rod(s)and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod(s)is inoperable.

If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s).When in MODE 1 or MODE 2 with K,>>greater than or equal to 1.0, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6.C.When in MODE 2 with K,less than 1.0, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6.Prior to initial operation above 5%RATED THERMAL POWER after each fuel loading, by consideration of the factors of e below, with the control banks at the maximum insertion limit of Specification 3.1.3.6.See Special Test Exception 3.10.1.COOK NUCLEAR PLANT-UNIT 2 Page 3/4 I-I AMENDMENT 8R, 40S, 434 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.1 REACTIVITY CONTROL SYSTEMS CHARGING PUMP-SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 a.One charging pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency bus.One charging flow path associated with support of Unit 1 shutdown functions shall be available.'PPLICABILITY:

ACTION: Specification 3.1.2.3.a.

-MODES 5 and 6 Specification 3.1.2.3.b.

-At all times when Unit 1 is in MODES 1, 2, 3, or 4.a.With no charging pump OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes." With more than one charging pump OPERABLE or with a safety injection pump(s)OPERABLE when the temperature of any RCS cold leg is less than or equal to 152'F, unless the reactor vessel head is removed, remove the additional charging pump(s)and the safety injection pump(s)motor circuit breakers from the electrical power circuit within one hour.c.The provisions of Specification 3.0.3 are not applicable In addition to the above, when Specification 3.1.2.3.b is applicable and the required flow path is not available, return the required flow path to available status within 7 days, or provide equivalent shutdown capability in Unit 1 and return the required flow path to available status within the next 60 days, or have Unit 1 in HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and HOT SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .The requirements of Specification 3.0.4 are not applicable when Specification 3.1.2.3.b applies.SURVEILLANCE RE UIREMENTS 4.1.2.3.1 The above-required charging pump shall be demonstrated OPERABLE by verifying, that on recirculation flow, the pump develops a differential pressure of greater than or equal to 2290 psid when tested pursuant to Specification 4.0.5.A maximum of one centrifugal charging pump shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 152'F."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 3.1.2.7.b.2.

I COOK NUCLEAR PLANT-UNIT 2 Page 3/4 1-11 AMENDMENT 8$, 407, 446 i'

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.1 REACTIVITY CONTROL SYSTEMS CHARGING PUMPS-OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE.APPLICABILITY:

MODES 1, 2, 3 Gild 4.ACTION: With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1%hk/k at 200'F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ;restore at least two charging pumps to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .SURVEILLANCE RE UIREMENTS 4.1.2.4 At least two charging pumps shall be demonstrated OPERABLE by verifying, that on recirculation flow, each pump develops a differential pressure of R 2290 psid when tested pursuant to Specification 4.0.5.COOK NUCLEAR PLANT-UNIT 2 Page 3/4 1-12 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.1 REACTIVI'IY CONTROL SYSTEMS BORATED WATER SOURCES-SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.7 As a minimum, one of the following borated water sources shall be OPERABLE: a.A boric acid storage system and associated heat tracting with: 1.A minimum usable borated water volume of 4300 gallons, 2.Between 20,000 and 22,500 ppm of boron, and 3.A minimum solution temperature of 145'F.b.The refueling water storage tank with: 1.A minimum usable borated water volume of 90,000 gallons, 2.A minimum boron concentration of 2400 ppm, and 3.A minimum solution temperature of 70'F.APPLICABILITY:

MODES 5 and 6.ACTION: With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes'ntil at least one borated water source is restored to OPERABLE status.SURVEILLANCE RE UIREMENTS 4.1.2.7 The above required borated water source shall be demonstrated OPERABLE: At least once per 7 days by: 1.Verifying the boron concentration of the water, Verifying the contained borated water volume, and Verifying the boric acid storage tank solution temperature when it is the source of borated water.At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when it is the source of borated water.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 3.1.2.7.b.2.

I COOK NUCLEAR PLANT-UNIT 2 Page 3/4 1-15 ill 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.1 REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES-OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.8 Each of the following boratcd water sources shall be OPERABLE: a.A boric acid storage system and associated heat tracing with: 1.A minimum contained borated water volume of 5650 gallons, 2.Between 20,000 and 22,500 ppm of boron, and 3.A minimum solution temperature of 145'F.b.The refueling water storage tank with: APPLICABILITY:

ACTION: 1.A minimum contained borated water volume of 350,000 gallons of water, 2.Between 2400 and 2600 ppm of boron, and 3.A minimum solution temperature of 70'F.MODES 1,2, 3 and 4.With the boric acid storage system inoperable, restore the storage system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWN MARGIN equivalent to at least 1%Delta k/k at 200'F;restore the boric acid storage system to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .b.With the refueling water storage tank inoperable, restore the tank to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .SURVEILLANCE RE UIREMENTS 4.1.2.8 Each borated water source shall be demonstrated OPERABLE: COOK NUCLEAR PLANT-UNIT 2 Page 3/4 1-16 AMENDMENT 04, 434, AS 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)SURVEILLANCE RE UIREMENTS Continued At least once per 18 months by: 1.Verifying automatic isolation and interlock action of the RHR system from the Reactor Coolant System when the Reactor Coolant System pressure is above 600 psig.A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.)show no evidence of structural distress or corrosion.

'.At least once per 18 months, during shutdown, by:t 1.Verifying that each automatic valve in the flow path actuates to its correct position on a Safety Injection test signal.Verifying that each of the following pumps start automatically upon receipt of a safety injection signal: a)Centrifugal charging pump b)Safety injection pump c)Residual heat removal pump By verifying that each of the following pumps develops the indicated differential pressure on recirculation flow when tested pursuant to Specification 4.0.5: 1.Centrifugal charging pump-Greater than or equal to 2290 psid 2.Safety Injection pump Greater than or equal to 1385 psid 3.Residual heat removal pump Greater than or equal to 160 psid By verifying the correct position of each mechanical stop for the following Emergency Core Cooling System throttle valves: 1.Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE.'The provisions of Technical Specification 4.0.8 arc applicable.

COOK NUCLEAR PLANT-UNIT 2 Page 3/4 5-5 AMENDMENT 43k, 434, 488, 480

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.5 The refueling water storage tank (RWST)shall be OPERABLE with: a.A minimum contained volume of 350,000 gallons of borated water, b.Between 2400 and 2600 ppm of boron, and c.A minimum water temperature of 70'F.APPLICABILITY:

MODES 1, 2, 3 and 4.ACTION: With the refueling water storage tank inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .SURVEILLANCE RE UIREMENTS 4.5.5 The RWST shall be demonstrated OPERABLE: a.At least once per 7 days by: 1.Verifying the contained borated water level in the tank, and 2.Verifying the boron concentration of the water.b.At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature.

COOK NUCLEAR PLANT-UNIT 2 Page 3/4 5-11 AMENDMENT BQ, 94

3/4 BASES 3/4.1 REACTIVlTY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL 3/4.1.1.1 Gild 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1)the reactor can be made subcritical from all operating conditions, 2)the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3)the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T,~.The most restrictive condition occurs at EOL, with T,~at no load operating temperature, and is associated with a postulated steam line break accident and resulting uncontrolled RCS cooldown.In the analysis of this accident, a minimum SHUTDOWN MARGIN of 1.3%Delta k/k is initially required to control the reactivity transient and automatic ESF is assumed to be available.

With T,~less than 200'F, the reactivity transients resulting from a postulated steam line break cooldown are minimal and a 1%Delta k/k SHUTDOWN MARGIN provides adequate protection for this event.The SHUTDOWN MARGIN requirements are based upon the limiting conditions described above and are consistent'ith FSAR safety analysis assumptions.

3/4.1.1.3 BORON DILUTION A minimum flow rate of at least 2000 GPM provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual during boron concentration reductions in the Reactor Coolant System.A flow rate of at least 2000 GPM will circulate an equivalent Reactor Coolant System volume of 12,612 cubic feet in approximately 30 minutes.The reactivity change rate associated with boron reductions will therefore be within the capability for operator recognition and control.COOK NUCLEAR PLANT-UNIT 2 Page B 3/4 1-1 AMENDMENT S2, 408, 434 I

3/4 BASES 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.5 REFUELING WATER STORAGE TANK The OPERABILITY of the RWST as part of the ECCS ensures that sufficient negative reactivity is injected into the core to counteract any positive increase in reactivity caused by RCS system cooldown, and ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA.Reactor coolant system cooldown can be caused by inadvertent depressurization, a LOCA or steam line rupture.The limits of RWST minimum volume and boron concentration ensure that 1)sufficient water is available within containment to permit recirculation cooling flow to the core, and 2)the reactor will remain subcritical in the cold condition following a LOCA assuming mixing of the RWST, RCS, ECCS water, and other sources of water that may eventually reside in the sump, with all control rods assumed to be out.These assumptions are consistent with the LOCA analyses.The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 7.6 and 9.5 for the solution recirculated within containment af'ter a LOCA.This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The ECCS analyses to determine F<limits in Specifications 3.2.2 and 3.2.6 assumed a RWST water temperature of 80'F.This temperature value of the RWST water determines that of the spray water initially delivered to the containment following LOCA.It is one of the factors which determines the containment back-pressure in the ECCS analyses, performed in accordance with the provisions of 10 CFR 50.46 and Appendix K to 10 CFR 50.COOK NUCLEAR PLANT-UNIT 2 Page B 3/4 5-3 AMENDMENT 407, 44k

3/4 BASES 3/4.6 CONTAINMENT SYSTEMS 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 I I A9 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 11.49 psig which is 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.This linutation ensures that the structural mtegrity of the contauunent 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 2 Page B 3/4 6-2 AMENDMENT 0

ATTACHMENT 3 TO AEP:NRC:1207 CURRENT PAGES MARKED-UP TO REFLECT PROPOSED CHANGES TO THE DONALD C.COOK NUCLEAR PLANT UNIT NOS.1 AND 2 TECHNICAL SPECIFICATIONS

CURRENT PAGES MARKED-UP TO REFLECT PROPOSED CHANGES TO THE DONALD C.COOK NUCLEAR PLANT UNIT NO.I TECHNICAL SPECIFICATIONS gE!'le}Cc.-

mcus qua 6=-a-6ia c (gO Pgi ySrr~Q'r~~Irgg CP~%7'.QN 6c8-'loo dg(~00'ye 6ts 0>s]<Al:CEPTABt.E QPERAT10f~5~4 5~IS~R tft oCtjgrs Of nOecngl't t.t.t t.z PREHURE (PS(A)SREAKNlHTS (FRAC7LQN RATEO THER'.PQMER, T-AVQ D OEGREES)?8CQ 200 QQ 22SQ 24QQ (0.0, 622.1), (0.0, 633.8), (0.0, 6m.8), (0.0, 650.7), (0.0, 660.1), (1.13, 581.3), (1.08, 601.<).(1.06, 609.8), (1.02, 621.9), (0.98, 633.j), (1.20, D7.5)(1.20, 586.0)(1.20, 591.3)(1.20, 598.9)(1.ZO, 606.2)F16URE 2.1 1 REACTOR CORE mEn t.rzrTS COOKÃLC~W P~VC-4 fX7 L~~~~wo.7N XS2.168

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'ara ra%%ra ,nrarrraaaaa TABLE 2.2-1 REACTOR TRXP SYSTEN INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRXP SETPOINT', ALLOWABLE'VALUES 1.Manual Reactor Tri.p Noc Appli.cable Noc Applicable 2.Power Range, Neutron Flux Low Secpoint-less'chan or equal eo 25%of RATED THERE'LL POWER High Setpoinc-less than or equal, co 109%of RATED THER'iAL POWER Low Setpoint-less than or equal co 26%of RATED THERtfAL POWER High Secpoine-less chan or equal to 110%of BATED THERHAL POi:ER.3.Po~er Range, Neutron Flux, Hi.gh Positive Race Less than or equal co 5%of RATED THERMAL POWER with a ci.Ie constant greater chan or equal to 2 seconds Less than or equal co 5.5%of RATED THEKfAL POWER wi.ch a cime conscanc greater chan or equal co 2 seconds (o Power Range, Neutron Flux, Hi.gh Negative Rate Xncermedf.ate Range, Neutron Flux Less than or equal.co 5S of RATED THER'iAL POWER with a cime constant greacer than or equal co 2 seconds Less than or equa3 to 25%of RATED THER iAL POWER Less than or equal co 5.5%of RATED THER.'fAL POWER wich a cime conscane greaeer chan or equal co 2 seconds Less chan or equal to 30$of RATED THER%6.POWER 6.Source Range, Neutron Flux Less chan or equal co 10 5 councs per second Less chan or equal co 1.3 x 10 counts per second 7.Overtemperacure Delta T See Noce 1 See Note 3 8.Overpower Del.ca T See Note 2 See Note 4 9.Pressurizer Pressure--Low Greater than or equal eo 1875 psig Greacer chan or equal co 1.865.psig LO.Pressurizer Pressure--High 11.Pressurizer Water Level--High Less chan or equal eo 2385 psig Less than or equal co 92%of instrument span Less than or equal.co 2395 psig, Less than or equal to 93%o f instrument span 12.Loss of Flow Gteacer than or equal co 905 of design flow per loop*0 r e ace r chan o r equal co 89.1.%of design flow per COOK NUCLEAR PLANT-UNIT L 2-5 loop*~/~V R.6 4cdat Elo~~z Lc o~7'pl l~gq g AiiENOSENT NO.9i fN 151

REACTOR TltlV SYS'1'DI INI'ItttHI,"N'I'ATIt)H

'I'III V St".I'I'OlH'1'S I I l.ta>'I'A'I'I ttN Hote l: Overtemperature AT 4 AT Q-K.-o 1 2 1 IT.>~Hl>ere h I 0~Indicated h'I'L Ith'I'I tt'I'Itl tetht.tiiHI'.I<U=AVerage temperat.uru I: Indicated'I'L Ilail'Ill)

TliHINAI.VilHI I((~c'Y)=Pressurizer prvu"urv, l+t lS 1+v S 1'-'I'ime cuti La>>L'Lx I x...a:.I i>>Lltv.lead-lag controller for T avg~l 4 st'cs~2~Laplace transform oiterator 22 sectt~IndicaLed Itt'S>>omi>>al opuraLilliJ harv""urc ('235 psig or 2085 psig)~The function getterated 1>y L)w lead-lag co>>troller for T dynattti ynattt c compensation

2~)Lg+~co~~~Tali IMJLJNH II&I a aLH~Operation uilh 4 Loops I 17 l h K.~().()~)o K." UaOUllU the pcta-and f>(AI)is a function of the indicated difference between top and bottoa d t to f e peter-range nuclear ion chaabers;Mith gains to ba selected based on aeaiury4 oa~ec rg of inotnuaant response during plant startup tests such that: (i)For q-q bstvaan-37 Parcant and iRpercent, f (kf)0 (share~and ara p rcan RA77>>Till i..i.(RR in tha top and bhtt oahalves of resp ctively, and qt f qb~s total THEBNLL poMER in percent of amati THERMAL POMER).(ii)For each Percent that the magnitude of (q-q)oxcaoda-37 porc 0,,~lg tr)p aatPcint shall bo outsoar(calif rsdu5ad b$0.33 percent of fte value et I aavm THERMhL POMER.(iii)For each percent that tha aagnituda of (g-g)arcasds+kpercent, tba ag trip aatpoint shall ba autoaatically rsdubad bIw~percent cf lte Value et mrs TfKQNLL POMER.C 2~8 0

o 4 S S S U 0 N~Pg 0 Co ue Overpower AT<AT (K-K o 4 5 S 1+1 T-K (T-T")-f (AI)]6 2~here: AT~I tidic:ale.d AT a t lych')'I I)T~Average temperature, Indicated T.dt Ith'l'L'D dVg TNERHAL POWER.5 63',D TllEENAE E'OWER K5 1.003 0 0.0l77/F for increasing average texperature and 0 for decreasing average temperature K6 00015 for T>T";K 0 for T<T" 6 1+r3S Tha function generated by tha rata lag controller tor T dynaxic compensation kg~~~k I T3 Tixe constant utilized in the rate lag controller tor T~3~10 sacs.4VQ Laplace transforx operator I)i t.e)c l.2(AI)~0 Tha channel's xaxixux xore than~percent The channel's xaxixux xora than 2,1, percent trip point AT span.Qh4 1 1 not gxcood its coxputad tX'ip pQQQ trip point shall not exceed its coxputed trip point.b AT span.I n.y

3 4.1 REAcTzvzTY coNTROL sYsTBts.3 4.L.1 IORATION CONTROL SNJTDCNN NARCZN TAVC CREATER THAN 200 F LDQTINC CONDITION fOR OPERATION e)$3el.lel the SHUTDOWN MARCIA ahall be greater shan or equal to~Delta Q/k.APB.ZnSnXTY:

&DES 1, 2>>, 3, and S,.l.3 Wish she SHtJTDOW NARCZN leea shan~+A%Delta R/k, haaediately initiate and continue boration at'greaser shan or equal so 10 gpa of a solution consaini greaser shan or equaL to 20,000 pea boron or equivalent until she required SHUTDON NARCZM ta reatored.SURVEILLANCE UIREKXNTS (>g~4.1.1.1.1 qha SimTDOMN halOIN shall be dacezaiuad ca be dzeacza, hca'aa zequal so~t Delta~a.within one hour after detecsion of an tnoyerabl~control zod(a)and.'s leaat once per 12 houra thereafter vhile the zod(a)ia inoperable.

Zf the inoperable contzol rod ia iaIIovabLe or unsrippable, the above required NUXDOQW MhIQZN ahall be verified aycepsabLe vish an increaeed allovanca for the vithdravn vorsh of th>>fundable or unsrippable controL zod(a).b.%hen in MDC 1 oz'DC 2 vith Xetf greater than or equaL to 1.0, as least once yer 12 houra by verifying that control baak vishdravaL ia vithin the Ltaita of Specification 3.1.3.5.c.%hen in MDC 2 vith Xeff leaa shan 1.0, vithin 4 houza prior to achieving reactor criticality by verifying that she yred'icsed czittcal contzoL zo4 poaision ia vithin the LiILLta of 5pecificasion 3.1.3.5.d.?rior.te initial operation above 5%RATED THHQtAZ.PtÃEX after each faeL Loading.by conai4eration of the factors of e belov, vish she conszoL banIca at the aaabaua ineertion Linis of Specification 3.1.3.5.+See Special Teat Exception$.10.1., COOK NUCLEAR BLAST UNXT 1 3/4 1-1 mme'O 71tcV>c)48

REACTIVITY CONTROL SYSTEMS CHARCINC PUMP-SHUTDOWN LLGTINC CONDITION FOR OPERATION 3.1.2.3 a.One charging pump in the boron in]ection flo~path required by Specification 3.1.2.1 shall ba OPERABLE and capable of being powered from an OPERABLE emergency bus.b.One charging flovpach associated vith support of Unit 2 shucdoMn functions shall be availabla.*

APPLICABILITY:

Specification 3.1.2.3.a.

-MODES 5 and 6 Specification 3.1.2.3.b.

-Ac aQ times vhen Unit 2 is in MODES 1, 2, 3, or 4.ACT'KON: a.Mi.ch no charging pump OPERABLE, suspend all oparacions involving CORE ALTERATIONS or positive raaccivicy changes.~b.Qi.ch mora chan one charging pump OPERABLE or arith a safety infection pump(s)OPERABLE vhan tha temperature of any RCS cold leg is less chan 0 or equal co 152 F, uxor.ass tha reactor vassal head i.s removed, remove the addi.tional charging pump(s)and che safety infection pump(s)mocor circuic breakers from che elaccrical pover circuit vichin one hour.c.Tha provisions of Specification 3.0.3 ara not applicable.

d.In addi.tion co che abo~e, shen Specificacion 3.1.2.3.b is ipplicable and che required flov path is noc available, recurn che required flo~path co available scacus vichin 7 days, or provide equivalenc shutdown capability in Unit 2 and return the required flov path co available status within the next 60 days, or have Unic 2 in HOT STANDBY within che next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and HOT SHUTDOWN vithin che folloving 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .e.Tha raquiremencs of Specification 3.0.4 are noc applicable

<hen Spacifi.cacion 3.1.2.3.b appli.as.SURVEILLANCE UI~ENTS oir~sC~'~~

4.1.2.3.1.

The above required charging pump shall ba demo crated OPERABL'E by veri.fying, chat on recirculation flov, cha pump develops a 44si4a~pressure of greacar chan or equal co+3&9 ps+Mhan cascad pursuant co Speci.ficacion 4.0.5.~Z,go*A maximum of one centrifugal charging pump shall be OPERABL'E whenever che temperature of one or more of che RCS cold legs i.s lass chan or equal co 152 F.o~for purposas of this specification, addition of vatar from cha RUST does not consti.cuca a positive reaccivicy additi.on providad tha boron concentration in che RUST is.graacar chan the minimum required by Specification 3.1.2.7.b.2.

COOK NUCLEAR PLANT-UNIT 1 3/4 1-11 AMENDMENT No.yg, ZAP.Z7Z, ZN

3.1.2.4 At least tvo charging pumps shall be OPERASLE.MODES 1, 2, 3,and 4.hRDQH: arith only one charging pump OPERhSIZ, restore at least tvo charging pumps to OPERhhLE status vithin 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SThHDbY vithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ;restore at least tvo charging pumps to OPEMLE status vithin the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN vithin the folloving 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .pygmy 5+hlYHf4 4.1.2.4 ht least tvo charging pumps shaQ be de onstrated OPEMLE by verifying, that on recirculation flov, each pump develops apg&aekmge-pressure of greater than or equal to~paid vhen teated pursuant to Specification 4.0.S.x~so J (0 (COOK NUCLEAR PLurr-UHZT l 3/4 l-l2~mm Ho.os, 164

I RcACTIVITY CONTROL SYSTEHS 1 BORATEO WATER SOURCES-SHL'TXWN I<<LIHITING CONDITION FOR O~ciRAT;ON

<I~3.1.2.7 As a minimum, one of the following borated water sources shall be: i OPERABLi: a.A boric acid storage system and associated heat tracing with: 2.3.A minimum usab'.e borated water volume of 4300 gallons,'etween 20,000 and 22,500 ppm of boron, and A minimum solution temperature of 145'F.b.The refueling water storage tank with: l.A minimum usable borated~ater volume of 90,000 gallons, 2.A minimum boron concentration of 2400 ppm.and 3.A minimum solution temperature of~F.0 APPLICABILITY:

HOOKS=.'."." 6.ACitON: With no borated water so r"e OPERABLE, suspend al operations involving CORE i" ALTiRATIONS or positive~eac ivity changes until at least one borated water'.source is restored.o OPERABLE status.~~'SURVEILLANCE RE UIRcvENTS 4.1.2.7 The above required borated water source shall be demonstrated OPiRABLi: I a.At least once per 7 days by: 2.3.Verifying the boron concentration of the water, Verifying the.water level volume of the tank, and Verifying the boric acid storage tank solution temperature when it is tne source of borated water.b.At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when it is tne source.of borpted water.4" For purposes of this speci.ica ion, addition of water from the RMST does 11not constitute a positive reactivity addition provided the boron concentra-

~;tion in the RMST is greater tnan the minimum required by Specification ii~.i.~.7.b.Z.

9.C.COOK-UNIT 1 3/4 1-15 Amendment No.$2.ill

I I<EAC I IV ITY CONTROL SYSTENS" BORATEO MATER SOURCES-OPERATIONS

~~I 1;: LIH:T:NG CONDITION FOR OPEPATION~~~~~~:: 3.1.2.8 Each of the fol/owing berated water sources shall be OPERABLE:~I a.A boric acid storage system and associated heat tracing with: l.A minimum usable borated water volume of 5650 gallons, I I1'~~I~~~,I~1~~Ii~~I~I 2.Between 20.000 and 22.500 ppm of boron, and 3.A minimum solution temperature of 145'F.b.The refueling water storage'tank with: I.A miniru.-.contained volume of 350,000 gallons of water, 2.Between-'".3 and 2600 ppm of boron, and~~~I APPLŽ.'8Il I: Y: I: I'AC ION: 3.A mini-.-solution temperature of i9'F.7d DOGES I,", 3 and 4.~~~~~~Mith the boric acid storage system inoperable.

restore the storage system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOMN HARGIN equivalent to at least 1"~k/k at 200'F;restore the boric acid storage system to OPERABLE status within the next 7 days or be in COLO SHUTOOMN wi hin the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .~0 I I~~~~b.Mith the refueling water storage tank inoperable, restore the tank to OPERABLE status within one hour or be in at least HOT STANOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLO SHUT-00MN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .: SU'1" ILLANCE RE UIRE1".E" 5~4.1.2.5 Each boratec a:=r;=rce shall be demons-rated OPERABLE: CsQK-UNIT 1 3/+1-10 Amendment No.58,

TABLE 3.2-1 DNB PARAMETERS PARAMETER Reactor Coolant System Tavg Pressurizer Pressure LLKZTS 4 Loops in Operation at RATED THE'-hfAL POWER o*~59,7 3oF>2050 psig*Reactor Coolant System Total Plov Rate++*>HEHMOO.gpm*Indicated average of at least three OPERABLE instrument loops.Limit not applicagle during either a THELML POWER ramp increase in exces of 5 percent RATED THERMAL POWER.per minute or a THER.'tAL POWER step increase in excess of 10 percent RATED.'THEKf~

POWER.Indicated value.COOK NUCLEAR PLANT-UNIT 1 3/4 2-14 AMENDMENT NO.~~i f/', 152 I, t'I

'(TABLE 3.3-3 Concinuad EHCKHEERED SAFETY FEATURE ACTUhTION SYSTEM INSTRUMENTATION FUNCTIONAL UNIT DEl.ETZ, f.Sceam FLov in Ter Sceam Lines-High TOTAL NO.OF CHANNELS RININJH CHANNELS CHANNELS APPLICABLE TO TRIP OPERABLE MODES ACTION Four Loops Operacing 2/steam line 1/s am I/steam line y 2 Line steam 1 nes 2, 3~14*ree Loops 0 racing 2/op racIng steam ine L~/any 1/operacing 3~operating steam line steam Line L5 COZNCZDENT ZTH EITHER (.(T~~Lov~Lov avg Four Loops Operating Three Loops Operating OR, COINCID PITH Sceam Line Pressure-Lov 1-T/Loop avg T vt o era8i,ng Loop T any lo'sg L~in any operacin Loop LT in 3 any No operating loops L5 I.T any, 1, 2, 3ERER La*3 Loofs Four Loops Operating Three Loops Operating 1 pressure/I.oop 1 pressure/operating loop 2 pressures any loops L~presssure in any operacing loop 1 pressure any 3 loops 1 pressure in any 2 operating loops 12, 3sEe I.5 ,.I NUCLEAR PLANT-UNIT 1 3/4 3-17 eu2tDHENT NO 91, fgg, 153

r (TABLE 3.3-3 Continued ENCZNEEREO SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION FUNCTIONAL UNIT NININJN TOTAL NO.CHANNELS CHANNELS APPLICABLE OF CNANNELS TO TRIP OPERABLE KODES ACTION COINCIDENT METH%%%BBR-T u~Lovu Lov avg Four Loops Operating Three Loops Operating 1 T/loop avg 1 T av operating loop 2 T'any'Y5opa 1¹¹¹T in any operating loop 1 T any Yg loops 1 T in an/cvo operacing loops 1, 2, 3¹¹'5 Stean Line Presaure-Lo~j 4 Four toopo Operating 1 pressure/loop 2 pressures any loops 1 pressure I., 2, 3¹¹14 any 3 loops Three Loops Operating 5.TURBINE TRIP&FEEDMATER ISOLATION 1 pressure/operating loop 1¹pressure in any operating loop 1 pressure 3¹>>in any 2 operacing'oops 15 a.Steam Generator Wacer Level--High-High 3/loop 2/loop in 2/loop in 1,2.3 any oper-each oper-ating loop acing loop , COOK NUCLEAR PLANT-UNIT 1 3/4 3-21 ANENDNENT NO.9f, f29, 153 0 0 ENGINEERED ShFETY FEATURES INTERLOCKS DES IC NATION F-12 ComrrZON um SETFoae Vith 2 of 3 pressuriser pressure channels greater than.or equal to 1915 psig.Vi,th 2 of 4 T channels less than or eqqal to Setpoint.Setpoint greater than or equal to 541 F FUNCTZON 2 IL prevents or defeats manual block of safety infection actuation on los pressurizer pressure.2-12 allovs the manual block of safety in]ection gcrubriou od lov ateaa line pressure.~~c~c-5 s 74'~AIR/8 i~~j7~N~/A/A'5f~+AM.4-Lgigfects steca dump blocks.Vith 3 of 4 T channels above the reset%44+)aug Point@Pgc,~c~i o(08 F&~i yttE.hafuwWc.~dF ShFBTy/~J8c77on/A~gg~pyp~~<A~Ag8+c sr~PRcssu/Z, 3/4 3-23a ThSLE 3.3-4 ENCQiEERED SAFETY FEhTURE hCTUhTION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONhL UNIT TRIP SETPOIHT l.ShFETY INJECTION, TURSINE TRIP e FEED VhTER ISOLhTIOH e AND MOTOR DRXVEN FEEDVhTER PUMPS a.Manual Iatttation b.huteaattc Actuation Logtc-------------

See Functional Unit 9-------------

Hot hpplfcabla Hoe hpplicabla co ContatIcant Pressure High d.Pressurtser Pressure--

Lmr a.DSMerenttaL Pressure Secvean Steca Ltnes--High eaa Floe tn Steaa s-High Co cidant eh a-Lcw-r Steaa Line Pressura-Lmr Less chan or equal to 1.L psig Greater chan or equal eo LdlS paig Less chan or equal to 100 pst ss than6or qual to 1.x 10 lbs froa 0%d to 20%1 d.ILp frofa 1.42 x LO lbs a)20%1 d to 3.d!x 0 Ibs/hr 00%load T greater t or a eo 541 P Less chan or equal to 1.2 psig Creatar than or equal to Ld05 psig Less than or equaL eo J 112 pst ss than6or qual to L.6 x 10 Lba Croa 0%d to 20%ad.Idea from 1.56 10 lb ag 20%ad to 3.93 10 lbs/hr t 100%load T greater or e to 539 F'+eater then or equal/+eater than or equal eo'to-500 paig staaa Line 4d0 pgg steaa line pressure pressure COOX HUCma PLhHT-UNIT J.3/4 3-24 hMEHDMEHT HO.49, 1ZS 153

TARLK 5.5-4 Continued ENCINEERED ShFETY PEATQRE hCTUATION SYSTEM INSTRUNENTATION TRIP SETPOIBTS FUNCTIONAL UNIT Z.Contatamcnt Radio-acttvtty--High Train A (VRS-LLOL.

ERS-1301, ERS-1305)3.Containment Radio-activity--High Tratn 5 (VRS-1201, ERS-L40L, KRS-N05)TRIP SETPOINT See Table 3.3-6 See Table 3.3-6 Not Applicable Not Applicable 4.Smx LINE ISOLATION a.manual See functional Unit 9 b.Automatic Actuation Logic c.Containment Pressure--

Htgh-High d.Steam llov tn Tvo Steam Lines--High Coincident vith T--Lov-Lov~Not Applicable Less than or equaL to 2.9 patg Less rhan6or equal to L.42 x 10 Lbs/hr froa Ot load to 20%load.Ltpar froa L.42 x 10 lbs/hr ag 20%load to S.bb x 10 lbe/hr at LOOt Load.Not Applicable Less than or equal to 3 psig Less thaa6or equal to 1.56 x 10 Lbs/hr from 0%Load to 20%load.ILpar from 1.56 x LO Lbs/hr6at 20%load to 3.93 10 Lbs/hr at 100%load.5(gag c.(Pc PNM~~" Lou)5~TURBINE TRIP AHD PEEDVATER ISOLATION T greater thea or c@l to 54L P Creater than or equal to 500 patg steaa Ltae pressure T greater than or c~L to 539 F Creater than or cquaL to 48Q'stg steam Line pressure a.Steam Ceacrator Vater Level--Htgh-High Less than or equal to 67%of narrov-range tnstrmacat span each ste4%generator Less than or equal to 6N of aarrov-raagc instrument span each steaa generator COOX NUCLEAR PLANT UNIT 1 3/4 3-26 JQKHDNEHT NO.gg, f/), 153 TABLE 4.3 2 ENCAGE?DAD SAPETT PEATmE ACTUATION SYSTEM INSTRUMENTATION FUNCTIONAL UNIT TRZ?ACTUATING CHAHHEL DEVICE CHQWEL CHAHNEL TUMCTIOHAL OPERATIONAL CEECE ClEEEIU TECH'TEST TEST NODES IN QHZCH SURVEILLANCE L.SAPETT INJECTION, TURSZNE TRIP, FEEDVATER ISOLATION, AHD MOTOR DRIVEH AUXILIARY FEEDVATER PUMPS a.NanuaL Znitiation b.Automatic Actuation H.A.H.A.H(2)H.A.1, 2.3.4-------------------

See tuactionaL Unit 9----~-------------

Logic coContainment Pteaae ux'e-High d.Preaautiaet Pt'eas-use ed%H.A.H.A.,1,2.3 L, 2, 3 e.Dtffet'entiaL Preaa-ut'Secween Steam Linea High H.A.H.A 1, 2~3 1.2~3<Pc'e a accee e LNI'5~/A*<wc, 2.CONTAQQKHT SPRlT a.Kaaual Initiation aeeeeeeaeeeeeeeaeee See functionaL Unit 9 aaeeaeae eaeaaaaeaa b.Automatic Actuation Logic H.A.H.A.H(2)K.AT 1, 2, 3, 4 c.Containment Preaa-uz'e-High-High

', M(5)H.A L,2,3 Caor.NUCLEAR PLurr-mIT 1, 3/4 3a51 SEEECEEEE EC.gg, (P, FUNCTIONAL UNIT TABLE 4.3-2 Continued EHCZNEERED SAPETT TEA'CURED ACTUATION STSTEM IHSTRQlKHTATZON SURVEZLIANCE RE UIRDQBiTS Tr"'ZP ACTUATZNC MODES IN CHANKf"~DEVICE%HIGH RGLHHEL CHANNEL TUN CT ZONAL OPERATZOHAL SURVEILLANCE CZZCZ CALZZlULTZCII TZZT TZST 4.STEP~<<LINE ISOLATION a.Manual eaaaaaeaaeaeaeaaeaea See Tuactioaal Unit 9 b.Automatic Actuation H.h.Logic H.A.M(2)N.A.-I., 2, 3 c.Containmeat'Preaa>>

ure--High-High d.Steam Tlov ia (, Tvo Steam Liaea-High Coiacident vith Tavga LoveLov Svhgm NuC PQGSRtPc/

5.TURBINE TRZP AHD TEEDVATER ISOLATION a.Steam Cene'rator Mater Level--High-High 6.MOTOR DRIVEN MXZLZART FEED WATER PUMPS a~Steam Ceaerator Pater Level--Lov-Lov b.4 kv Eua Loaa of Voltage c.Safety Ia]ectioa d.Loca of Main Teed Pumpa H.A.H.A.k N.A H A.M(3)M(2)H.A.H.A.H.A.H.A.H.A.H.A.H.A.1, 2, 3 1, 2, 3>i 2>Z 1, 2~3 lc2c3 1, 2, 3 1,:2, 3 2 coor.NUCLEAR plurr-UHZT 1 3/4 3-33 Jl!ZHCHZBT IIC~Zg g)f Zf,

3.4.2 h minimum of one,pressurizer code safecy valve shall be OPERQLE arith a lift setting of 2485 PSIC++a.*NODES 4 and 5.LEONE: Pith no pressurf,e'er code safecy valve OPERABLE: a.'Immediately'uspend all operations involving positive reactivity changeable'nd place an OPEMLE RHR loop into operation in the shutdown cooling mode.b.Imiediately render all Safety In)ection pumps and all but one charqui.ng pump inoperable by removing the applicable mocor circuic breakers from the electric pover circuit%thin one hour,.r 4.4.2 The pressurizer code safety valve shall be demonstrated OPEMLE per S'urveillance Requirement 4.4.3.*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 vater from the RUST does noc constitute a posicive reactivity addition provided the boron concentracion in the RUST is greater than the minimum required by Specification 3.1..2.8.b.2 (MODE 4)or 3.1.2.7.b.2 (MODE S).D.C.COOK-.ORXT 1 3/4 4-4~MEHT ttoSB, 1,20

3.4.3 Al.l preaauriaer code, aafety valvea ahall be OZXMZX with a life aettiag ot X4aS ZSZC g~.hQXIQE: With one presauriaez code safety valve inoperable, either reatore the..inoperable valve to OPEMLE status vithin lS rLinutes or be in HOT SHUTDOVN vitharr 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .4.4.3 Ho additional surveillance requirements other than those repaired by Specificatioa 4.0.S.The l.haft setting pressure shall correspond to ambient conditions of the valve at noainel operating teaparatura and pressure.

d.At least once per Lg months by: 1.Verifying aueomaetc isolation and interlock action of ehe RHX system from the Reactor Coolant System vhen the Reactor Coolant System pressure is above C00 psig.2.A visual inspection of the coataiameat sump and verifying ehae the subsystem suction inlets are aot restricted by debris and chat ehe sump components (trash racks screens, etc.)shov no evidence of structural distzess or abnormaL corrosion+

eo ht lease once per'8 months, during shuedovn, by: 1.Verifying chat each automatic valve in the f le paeh actuates eo its correct position on a Safety Injection test signal.'2.Verifying thee each of the foLLovtag pumps seazt automaeically upon receipt of a safety injeceioa signal: a)Centrifugal charging pump b)Safeey injection pump c)Residual heat removal pump g]QPEg~i~Sy verifying ehae e of the folloviag pumps deveLops the indicated 44aahe~>pressuza on recirculation flov vhen tested pursuaat to Specif icatioa 4.0.5.22yO~1.Centrifugal charging pump greater than or equal to%405 psig-I3A,&2.Safety injection pump greater than or equal to+09 psig-IS'O 3.keaiduaL heat removal pump greater than or equal to 404 psig-Sy verifying the correct position of<<ach mechanicaL stop for the foLLovtag Eiaergency Coze Cooling Systea throttle valves: 1.Viehin 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months foLLovtng completion of each valve stroking, operation or maintenance oa ehe valve vhen the ECCS sub-systems are required to be OFKRASLX.COOK NUCLEhR PLANT-UNZT L 3/4$-5

EMERGENCY CORE COOLING SYS""=.'iS REFUEL!HG MATER STOPPAGE Tklsi(LIHITING COHQI;IOH F".R'3P-""=-'".:"H 3.5.5 The refueling wa:er storage tank (RMST)shall be OPERABLE with: 1 a.A minimum con'tained volume of 350,000 gallons of borated water.b.Between 24OO and 26OO ppm of boron, and c.A minimum water=e~pera"ure ofPPF.APPLICABILITY:

HOOES 1, 2.3 and 4.7'0 ACTION: With the refueling wa-.e~s=orage tank inoperable.

restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ".r o in at least HOT STANOBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLO Si{UTNMN within the f" i:.~ng 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ..SURVEILLANCE RE UIRE.".-N".S i4.5.5 The RMST shall be de.znstrated OP=P~BLE a.At least once per 7 days by: Verifying tre contained borated water volume in the tank.and 2.Verifying the boron concentration of the water.b.At least once per 2-'ours by verifying the RMST temperature.

Ir jl II.t II D.C.COOV,-UNIT 1 Amendment No.$3.III

HTS Conc ued 4.7.I..2 Ecch aus:iliaxy feedvaeer pump shaLl be demonstrated OPERABLE vhea tested pursuant co Specification 4.0.5 by: ai b.Verifying that each motor driven pump develops an equivaleac discharge pressure of greater than or equal eo QR-psig ae 60 P in recirculation flovc,-c.-

o+~kRO i@0 Verifying that'he steam carbine driven pump devel s an equivalent discharge pressure of greater thaa or equal eo psig ac 60 P and at a flov of greater than or equal eo 700 gpm vhen che secondary sceam supply pressure is greater than 310 psig.The provisions of Speciiicacioa 4.0.4 are not applicable for encry ineo HOOK 3.c.Verifying that each noa-automatic vaIve in the flov path that is noc locked, sealed, or othervfse secured in position fs in its correct position.d.Verf.fyfng thee each automatic vaLve in the flov path is in ehe fully open position vhenever che auxiLf.azy feedvater system is placed in automatic concroL or vhen above LO\RATED THECAL POQc3..This requirement is noc applicabl.e for those porcioas of the auxiliary feedvaeer system being used intermittently to maintain steam generator vater level.e, Verifying ae lease once per I.8 monehs duxing shuedovn thae each automatic valve in the fIov paeh actuates to ies correct posI.cion upon recefpc of the approprfate engineered safeey features accuaeioa test signal required by Speciffcatioa 3(4.3.2.Verifying at lease once per 18 moachs during shucdova that each aus:i.Liaxy feedvaeer pump scares as desf.gned aucomacically upoa receipt of.the appropriate engineexed safety features actuation cesc signal requixed by Specification 3/4.3.2.Verifying ae least once per 18 moaehs during shutdovn that the unf.t cross-cie valves caa cycle full travel.FoLLovtng cycling, the valves vill be verified to be in theM closed positfons.

N AT ItE In accordance with the code requirements specified in Section 4.1.6 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.C.~VLUME 5.4.2 For a pressure of 2485 psig, and For a temperature of 650'F, except for the pressurizer which is 680'F.approximately 12.466 cubic feet at 0%steam generator tube plugging and 11,551 cubic feet at 30%steam generator tube plugging Th*K"a I Pd*I e eubte-feet at a nominal T,, of 70'F.5.5 MER EN Y ORE LTN Y TEMS 5.5.1 The emergency core cooling systems are designed and shall be maintained in accordance with the original design provisions contained in Section 6.2 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements, with one exception.

This exception is the CVCS boron makeup system and the BIT.5.6 L T RA RITI LTTY-PENT L 5.6.1.1 The spent.fuel storage racks are designed and shall be maintained with: A~equivalent to less.than 0.95 when Qooded with unborated water.A nominal 8.97 inch center-to-center distance between fuel assemblies placed in the storage'racks.C.The fuel assemblies will be classified as acceptable

'for Region 1, Region 2, or Region 3 storage based upon their assembly average burnup versus initial nominal enrichment.

Cells acceptable for Region 1, Region 2, and Region 3 assembly storage are indicated in Figures 5.6-1 and 5.6-2.Assemblies that are acceptable for storage in Region 1, Region 2, and Region 3 must meet the design criteria that define the regions as follows: COOK NUCLEAR PLAiNT-UNIT I AMENDMENT NO.I,, 169 CORRECTED PAGE I

1 2.l SAFETY BASES 4 Loop Operation VescinghoQse POOL (15xL5 OFh)(WRS-1 Correlation)

Typical Cell Thimble Cell Correlation Limit Design Limic DNBR Safecy Analysis Limit DNBR'.L7~3-/,g3 I.VO 1.17~~~J.zx l.Va The curves of Figure 2.1-1 sho<<che loci of points of THEK<AL POt:ER, Reactor CooLanc System pressure and average temperature for<<hich che='ni"um DNBR is no ess='.".an"'.".e-ppL'cable design DNBR limit, or che average e..chalpy ac"he:essel exic is equal co che enchalpy of sacuraced Quid trepresents typical fuel rod represents fuel rods near guide tube ,0 Cook Nuclear Plant Unit 1 B 2-L(a)g~I@4'

~o The parer lac~Negati~mate Trip provides protacti.on for control rod drop accidence" At: hf.gh povec 4 rod drop accident could caus~1ocal f1~peWag Wi~orna calle an uxLco~ar ative local DNhk co exist.Tha Pover'Range Negative Race Trip vf11 prevent this froa occuziing by tripping the reactor.No credit is taken for operation of the Pover Range Negative Rata Trip for those control rod drop accidents for vhich DNfffL j vill be greater than the applicable deaf gn limit NQ value for each fuel type.Intermediate and Source Ran e Nuclear Flux 7he Intermediate and Source Range.Nuclear Flux trips provide reactor-eore protection durf.ng reactor start'p.These trips provide redundant protection to t..e lov setpoin trip of the Pover Range, Neutron Flux channels..he source Range Channels<<LLL init'ate a reactor trip ac~r5 about 10 counts per second.bless manually blocked vhen P-6 becomes active.he Inter=ediatc Range Channels vill,'niciata a reaccor trip at eve , opor anal 0 approxima e y 25 percent of RAZHI~a K JLL 2:;-R..n'ass can a ly b'octad'-..en 2-'0 beco=es active.No credit vas=<en="." operat on o"=he="'ps associated ci"h either the Intermediate I 1 cr.""-=e Range ha".ze s in"he ac"'Cent a..a'ysas; hovever, theft f nt t onal ckoab I (at tne spec if ie'd t ip settings is required by this s=ec=icstion to e."".ance=.".e c;era'e.Lab':y o=--.Reactor S's em (4..".e Cvertemperature delta.t"'p provides core prote ion to prevenc"YS a"""bina"'ons of pressure, pover, coolant temperature, and axfaL"o-cr Cist"'but'on.

provided thac the transient is slov vith respect to piping trinsi.t Celays from tha core to the temperatu e Cececcors (abouc 4 seconds), and pressure f.s vithin the range becvean the High and Lov Pressure reactor trips.This setpoint includes corrections for changes i l dens ity and heat capacity of vacer vich cemperacure and dynamic compensacion for i in dela s from the core to the Loop temperature Catectors.

The refer ce average emperacure an~d re erence crating p ssure (P')re sec eq to the fu pover ind aced 7avg a the nomfn RCS opera ng pressur respecciv y,:o ensu pro ctf.on of t core limf.and to pr erve the a t tiod cim of the Overt perature lta 7 trip or the rang of'full po" average tempera ures assume in the sa t anaL ses t no:=a ax a poMer distribution, this reactor trip fmic f.s aLvays below=he core safety Limi<>nuclear detectors, the reactor trip f.s automatically reduced according to the notations Ln 7abla 2.2-L.DELE (8 COOK~~CLM%PLAÃE UNI7 L 5 2-4 h~EÃM~7 NO~7)L 26 The Overpover delta T reactor ertp provtdei assuraace of fueL tneegriey,~.g.~ao meletng, under aLL possible overpover coadtttoas, Ltitts the required range fot'vertenyeracure delta T protection, aad yrovtdes'a backup to ehe High Heueroa Plus:,erty.

~~The seepotnt tncludaa corrections for changes tn density and heat capacity of vacer vith tewperaeure, and dyaaato conpeasattoa for t ta dele s froa the cora co the looy temperature detectors.~~

The referea average t peraeura ()is see e to ehe 1 over tndtca d Tavg to e sure fuel egrtty dur g overpove conditions or P~T~eh range of L pont av ag>>there es aeeae tn'he aa ey aaal sis.~overpover e ta reactor cr p prov es protect on or c-uy protection for ac pover staaultne break events.Credtt vea taken for operation of this trip tn the st~an Lfne break nasa/energy releases outside coaeatnnene aaalysts.Zn addttton, its functional capability ae the syecified trip seettng is requtred by this specification co enhance ehe overall reliability of ehe reactor proeeceioa syseea,.?resaurtser Pre u e The tressurtaer High and Lov Pressure trips are provtded to Ltnte the pressure range tn vhich reactor opecatioa ts pemftted.The High tressure trip ts backed up by ehe yrassuriaer code safacy valves for PCS overpressure protection.

and is therefore see lover than ehe see pressure for these valves (2485 pstg).The High Pressure trip provides proeection for a Loss of hccernal Load evene.The Lov tressure eriy yrovtdes protection by trfyptng che reactor in che eveae of a loss of reactor coolaat pressure.Pressu txer Vaeer Lave The Pressurizer High Qatar Level crip ensures procectioa agatasc reactor Coolane System overpressurtxaetoa by Ltmttiag the vaeer level eo a volune sufficient to retain a acean bubble end praveat vater relief through the pressurtrer safety valves.The prassurtaer high vater level crip precludes vaeer relief for the Uncontrolled RCCA N.thdraval ae Pover event.cooK NUCLuk?err-UNIT l I 2-5 AHzHuKNT No-XSSM ltd, 158 0 0 3/4~L~a~~CQHTRQL SY51!IAS ES 3 4.L.L EmVTTQH ecmZOL 3 4.L.L.L and 3/4.L.L.2 SH~iVH NARC'sufficient SHUZDQRf MMtCDf<<nsuzes thee I)the reactor can b>>lsade subczitioaL fzoa all operating conditions, 2}the reactivity tzansiencs associated vith postulated accident conditions aze controllable vithin acceptable Iinits, and 3}the reactor vtLL be Iaincained sufQciently subczitioal to preclude inadvertent criticality in the shutdovn condition.

/.3 SjglTDdt'5?QRCXH recpLx'aae vazy dcooghout co e Life as a function of fueL deplecion, RCS boron c entzation, and g5 T.The aoat restrictive condition occurs t tOt., vith'7 at no Hid operating tenperatuze, and is associa a vith a poseZiked stela Line break accident and resulting unconczolled R cooldovn.Za the analysis of M accident, a nin~SHUTÃtV NhRCZH of~~DeLta klk is initially zeqhzed to conc=oL the reac ivity transient and autoaatic'SP is assumed to be available.

Vf.ch Tivg less than 200 F, the reactivity czansients resulting fxoa a postulated stean Line break cooldovn aza lintel and a Li Delta~SRZDON MH'-i provides adecpzata protection for this event.Tha%KEN%QRCIM reqxheaents are based upon the limiting conditions desc ibad above and ara consistent vich gQ safety analysis assunptions

~3 4.L.L.3 hORQH DILIOH~0 h ainu+floe rata of at Least 2000 CCK provides ade~ia~g.pzevencs stzatf.ication and ensures that reactivity changes vill be grsduaL during boron concentration reductions in the Reactor Coolant Systaa.k flov rata of at least 2000 CPI vQL circulate an equivalent Reactor Coolant Systaa volune of I2,02 plus or ainus I00 cubic feet in approximately 30 ainutas.The reaccivity change rata associated vith boron reductions

>end of each fuel cycle ia adequate to confizn the NC value since this coefficient changes slovly due pzincipaLly to th>>reduction in KCS boron PlAPi,-UHZT I h 3/4 I-L NO.1$,723, t48 4.4 R:-~:"R COO&v-SvS.=M%ASKS 3/4.4.1 R~aCTQR COO'>M~~OPS st~~puhxv$ls l.im The plane is desk~ed to operace A all reactor coolant Loops fn operacion, aad maintain ONER above++}-during all normal operations and anc c pated transients. A Loss of flov ia tvo loops vQL cause a reactor cr'p if operating above P-7 (ll percent of RATED THERNhL PQQKR)vhile a Loss of ilaw in one Loop vill causa a reactor trip if operating above P-8 (31 perceac of.Re TED THHQfAL POWER)., Zn NODE 3, a si'agle reactor coolanc Loop provides suffic'eac heac removal capability for removf.ag decay heat;hovever, single failure considerations require that c~o loops be OPERON.Three loops are required-o be OPHASL~an4 to operate if the controL rods are.capable of vithdraval and-".e reactor tr'p breakers a e closed.ne requirameat assures adequate DN R margin'n=".e a~eat of an uncontroLLed rod vi-Crave'".."h'X=ode.Xa NODES 4 and 5, a single reactor coolant loop or RHR Loop provides sufzicianc heat removal capability for removtag decay heat;but single failure considaratf.ons require chat a Least tvo loops bc OPMLE.Thus, i the reac o-coo1xat loops are noc OPML=, this specification requires tvo RHR loops to be OPRVJLL~.The operation of one Reactor Coolant Pump or oae RKL pump provides adequate ilov co ensure mincing, preveac stratiQcatioa aad produce gadual reactivity changes during boron concentration reductions in the Reactor Coolant System.The reaccivity change race associace4 vith boroa reduction vill.therefore, b>>vithin the capability of operacor racogxitioa an4 coacrol.The restriccions on start~my a Reaccor.Coolanc Pump bclov P-7 vith oac or more RCS cold legs less thea or equal co L5Z c are provided to prevent RCS pressure transiaacs, caused by eaerg additions from thc secondary syscem, vhich could exceed chc limits of hppeadix C to 10 CPR Par~50.The RCS vil'>>prateccad against overpressure t=ansieat>> aad vill aoc exceed the Limits of Appends C by either (1)restricting tha vatar volume in the pressuriser aad thereby providing a volume for the primary coolant to expand into or (X)by restricting starting of the RCp's to vhea the secondary vacar temperacure of each steam generator is Lass than 50 F above each of the RCS cold Leg temperatures. C OK hJCL"=<8, P'v7-4.lTT L 3/4 4 1~".ZWO.".ES. WO-PS.f4 BhSES 3 4.5.5 REZt~LIZC WATER STOHACZ maC The OPERhBILITY of the EST as part of the ECCS ensures that sufficient negative reactivity is infected into the core to counteract any'ositive increase in reactivity caused by RCS system cooldovn, and ensures that a sufficient supply of., borated vater is available for injection by the ECCS in the event of a LOCA., Reactor coolant system cooldovn can be caused by inadvertent depressuzization, a loss of coolant accident or a steam line rupture.The limits on RUST minimum volume and boron concentration ensure that 1)sufficient vater is available vt.thin containment to permit recirculation cooling flov to the core, and 2)the reactor vf.ll remain subczitical in the cold condition following assumptions are consistent vith the LOCA analyses.The s e 4xosF.~A The contained~ater volume limit includes an allovance for vater not usable because of tank discharge line location or other physical characteristics; The limits on contained vatez-volume and boron concentration of the RUST also ensure a pH value of beareen 7.6 and 9.5 for the solution recirculated vithin containment after a LOCA.This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corx'osion on mechanical systems and components. The ECCS analyses to determine F limits in Specifications 3.2.2 and 3.2.6 assumed a RUST~ater temperature of 70 F.This temperatux'e value of the RUST eater decermines chat of the spray vater initially delivered to the containment. folloving LOCA.It is one of the factors vhich determines the containment back-pressuze in the ECCS analyses, performed in accordance with.the rovisions of 10 CFR 50.46 and A endQc K to 10 CFR 50.va e o e min R T te era re i Tec ical ec icat n 3..5 h be nse tive'ch ged o 80 to cze e th co seen be een nit an.e lo erR T te era re r ults inl er nta en res e om nta ent pra and afe rds ov sume to it e b k.ve con i t p ssu res ts inc asc floM esi ance of s earn iti the c e reb slo g.r oo d cre in INSERT A a LOCA assuming mixing of the RWST, RCS ECCS water, and other sources of water that may eventually reside in the sump, with all control rods assumed to be out. S~L~ggfj CO 3 A.&.1.<PRX55URIl The lfafcaciona oo concaiasent internal pressure ensura that l)the contaiamenc struccura is pravanca4 from exceeding ita design nagLcivg pressure differential with respecc co the outside atmosphere of 8 paid and 2)the concainmenc peak pressure does noc exceed the design pressure of 12 psfg during LOCA condi.tions.,II.+9 The/maximum peak pressure resulting from a LOCA event is calculace4 to be~.-.89-psf.g, vhich includes 0.3 psig for initial'positive containment pressure.3 4.6.1.5 AIR T~~PERATCRE e Limf.tat: "..s on cont@i"."enc average af,r tempezacure ensure that l.)the con af..Den i'ss is ed o an ini.ial mass su f c encly Lov to""event exceeding the desi'ress re du in'C~~conditions and 2)"he-"bienc air=a="era=.ra='ces."."" excee4=hat=e"peracure allowable for the cont'nuous duty tati-..g specified for equipmenc and instrumentatfon Located'<<i=hf.n cents'r e..t.py.+9.he containment press re" snsient'.s sensitive to--e'n'c'alLv con"a'ned air"ass-"" i"..g a'A..;.e cotta'."ed a'ass'ncreases<<ith ecreasing=a~era=...".e o<<er tempera~are 'a'-=60 F<<iLL Limit"he pea, pressure to++psig"h'ch's less than;-contain<<sent desip pressure of 12 psig.the."per temperature limic 1:.antes the peak accident tempers"'e sli~hcly d'ing a MCA: however:his limf.t is based<<i t p'=ar'upon equ'pmenc pro"ect'on and anticipated operating conditions. Boch the upper and lo<<er temperacure Limits are consistent Mich the parameters sed in the accfdenc anaLyses.3 4.6.L.6 C"NTAI'.iNHlT VESSEL STRUCTURAL I iTECRITY This Lf.=itacion ensures thac che scructural integrity of che concainmenc sceel vessel.viLL be maincained comparable to che otic'naL design standards. for the life of che facility.Structural integrity is required to ensure that (1)che steel liner remains leak tight snd (2)che concrete surrounding che steeL Lf.ner remains capable=providing.externaL missile procection for the sceel liner and r=f.*cion shf,e".-.g in the evenc of a LOCh,.h visual inspection in con]unction vf.th Type A, Leakage tests is sufficient to demonstrate thfs capabi'ty, (0 COOK NUCLEAR PLANT UNIT 1 B 3/L 6-2 Alfie)gENT NO.1"-6 CURRENT PAGES KQQKD-UP TO REFLECT PROPOSED CHANGES TO THE DONALD C.COOK NUCLEAR PLANT UNIT'NO.2 TECHNICAL SPECIFICATIONS 3 4.1 REACTIVITY CONTROL SYSTEMS 3 4.1.1 BORATION CONTROL SHUTDOWN MARGIN-T GREATER THAN 200 F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to~Delta k/k.APPLICABILITY: MODES 1, 2*, 3, and 4.ACTION: J(2%y Pith the SHUTDORf MARGIN less Chan~Delta k/k, immediately initiate and continue boration at greater than or equal to 10 gpm of a solution containing greater than or equal to 20,000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.SURVEILLANCE RE UIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to~a Delta k/k: i zoo a.Vithin one hour after detection of.an inoperable control rod(s)~and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter vhile the rod(s)is inoperable. If the inoperable control rod is immovable or~untrippable, the above required SHUTNMN MARGIN shall be verified acceptable vith an increased allovance for the vithdravn vorth of the immovable or untrippable control rod(s).b.Ken in MODE 1 or MODE 2 vith K ff greater than or equal to 1.0, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that control bank vithdraval is vithin Che limits of Specification 3.1.3.6.c~@hen in MODE 2 vith K f less than 1.0, vithin 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor cri8fcali.ty by verifying that the predicted critical control rod position is vithin the limits of Specification 3.1.3.6.Prior Co initial operation above 5%RATED THERMAL POMER after each fuel loading, by consideration of the factors of e belov, vith the control banks at the maximum insertion limit of Specification 3.1.3.6.*See Special Test Exception 3,10.1 COOK NUCLEAR PLANT-UNIT 2 3/4 1-1 maeMENT NO.82,158,13a REACTIUITY CONTROL SYSTEMS CHARGING PUMP-SHUTDOWN LXMXTXNG CONDXTXON FOR OPERATION 3.1.2.3 a.One charging pump in the boron infection flow path required by Specification 3.1.2.1 shaD be OPERABLE and capable of being powered from an OPERABLE emergency bus.b.One charging flow path associated with support of Unit 1 shutdown functions shall be available..+ APPLICABXLXTY: Specification 3.1.2.3.a. -MODES 5 and 6 Specification 3.1.2.3.b. -At all times when Unit 1 is in MODES 1, 2, 3, or4.ACTION: a.Wi.th no charging pump OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.~b.'i.th more than one charging pump OPERABLE or with a safety infection pump(s)OPERABLE when the temperature of any RCS cold leg is less than or equal to 0 152 F, unless the reactor vessel head is removed, remove the additional charging pump(s)and the safety injection pump(s)motor circuit breakers from the electxical power cixcuit within one hour.c.The provisions of Specification 3.0.3 are not applicable. d.In addition to the above, when Specification 3.1.2.3.b is applicable and the required flow path is not available, return the required flow path to available status within 7 days, or provide equivalent shutdown capability'n Unit 1 and return the required flow path to available status within the next 60 days, or have Unit 1 in HOT STANDBY~ithin the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and HOT SHUTDOQH within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .e.The requirements of Specification 3.0.4 are not appLicable when Specification 3.1.2.3.b applies.SURVEILLANCE RE UXREMENTS 4.1.2.3.1 The above-required charging pump shaD be demonstrated OPERABLE by verifying, that on recirculation f.ow, the pump develops a-44.-sehaege pressure of greater than or equal to 4%0-pekg when tested pursuant togpecificacion 4.0.5.gg,fg p5I8*A maximum of one centrifugal charging pump shall be OpERABLE whenever the temperature !of one or more of the RCS cold legs is less than or equal.to 152 F.~<<For purposes of this specification, addition of water from the RUST does not (~i onstituce a positive reactivity addition provided the boron concentration in the:t4ST is greater than the minimum required by Specification 3.1.2.T.b.2. COOK NUCLEAR PLANT-UNIT 2 3/4 1-11 AMENDMENT NO.B5 10T 116 ICITIITY CalltlaL SffIEflS~a*pe meS-OPERATI%LIMITS CO~OITIO~POR OPERATIO~3.1.2.4 At Ieast~charging pumps sha11 be OPERABt c.APPLICABIUTY: MOQES I, 2, 3 and 4.ACTION: Nth onIy one charging pump OPERABLE, restore at Ieast two charging pumps to OPHABLK situs wfthfn 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be fn at Ieast HOT STAHOBY and borated to a SHUTDOWN MARQIH equfvaIent to at Ieast I~~k at 200 F withfn the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ;restore at Ieast t'~o chargfng pumps to OPBNBl~situs within the nex 7 days or be fn COLO SRUTUOMH within the nex 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .SURYEIL>MHCE REOUIREHEHTi 4.1.2.4 At Ieas--~o cha.~:rg pumps sha11 be demonstra zd OPBABLE"y verifying, that on recfrcuIatfon fIow, each pvxo deveIops a'o.>H&psfg when wsta6.pursuant 4 Speciffcatfon 4.0.5.~haJ gqfO psiq REACTIVITY CONTROL SYSTEMS BORATED MATER SOURCES-SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.7 As a minimum, one of the following borated water sources shall be OPERABLE: a.A boric acid storage system and associated heat tracting with: l.A minimum.usable borated water voIume of 4300 gallons, 2.Between 20,000 and 22,500 ppm of boron, and 3.A minimum solution temperature of 145'F.b.The refueling water storage tank with: l.A minimum usable borated~pter, yo.lume of.90,000 gallons, 2.A minimum boron concentration of 2400 ppm, and 3.A minimum solution temperature of-SPf=.0~9A>F APPLICABILITY: MODES 5 and 6.ACTION: Mith no borated water source OPERABLE, suspend all operations involving CORE ALTFRATIONS or positive reactivity changes" until at least one borated water source is restored to OPERABLE status.SURVEILLANCE RE UIREMENTS 4.1.2.7 The above required borated water source shalT be.demonstrated OPERABLE: a.At least once per 7 days by: 2.3.Verifying the boron concentration of the water, Verifying the contained borated water volume, and Verifying the boric acid storage tank solution temperature when it is the source of borated water.b.At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when"it is the source of borated water.*For purposes of this specification, addition of water from the RMST does not constitute a dilution activity provided the boron concentration in.the RMST is greater than or equal to the minimum required by Specification 3.1.2.7.b.2. D..C, COOK-UNIT 2 3I4 1-15 Amendment No.8~.94 REACTIVITY CONTROL SYSTENS BORATED WATER SOURCES-OPERATING LIHITING CONDITION FOR OPERATION 3.1.2.8 Each of the following borated vater sources shall be OPERABLE: A boric acid storage syseem and associaeed heat eracing vith: 1.A minimum concained boraeed vater volume of 5650 gallons, 2.Between 20,000 and 22,500 ppm of boron, and 3, A mi.ni.mum soluci.on temperature of 145 F.0 b.The refueling water scorage tank vt.th: A minimum contained borated vater volume of water, 350,000 gallons of~2.Between 2400 and 2600 ppm of boron, and 3.APPLICABILITY: NODES 1, 2, 3 and 4.h minimum solucion temperature of 8(A.<~o'F.ACTION: a~Wieh ehe boric acid storage system inooerable, restore ehe storage system co OPERABLE seatus vithi,n 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in ae least HOT STANDBY vithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWN NARGIN equivalent to ae lease 1X Delta k/k at 200 F;restore the boric acid scorage system to OPERABLE status vt.thin the next 7 days or be in COLD SHUTDOWN vithin the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .Wi.th the refueling vater seorage eank inoperable, restore the tank eo OPERABL'E status vt.thin one hour or be in at lease HOT STANDBY vithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT-DOWN vi.chi.n the folloving 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .SURVEILLANCE RE UIREHEVES 4.1.2.8 Each bo'rated vater source shall be demonstrated OPERABLE: COOK NUCLEAR PLANT-UNIT 2 3/4 1-16\AMENDMENT NO.N, ZN, lw EM-RCEN~Y CORE COOL:NC SYS i MS'E SURVETLLANCE RE UTR~~NTS (Cortinuedl d.At least once per 18 months by: 1.Verifying automati.c i.solat'on and interlock ac"'on of the R~system from the Reactor Coolant System when the Reactor Coolant System pressure S.s above 600 psig.A visual inspection of the containment sump and the subsystem suction inlets are not rest"icted that the sump components (t ash racks, sc eens, evidence of structural distress or corrosion.t v.ri.fying that by debris and etc.)show no At least once per 18 months, during shutdown, by:t 1.Verifying that each automatic valve i.n the flow path actuates to its correct position on a Safety Znjection test~signal 2~Verifying that each of the following pumps start auto-matically upon receipt of a safety injection test signal: a)Centrifugal charging pump b)Safety injection pump c)Residual heat removal pump By verifying that each of the following pumps develops the indicated~Kaehmge pressure on recirculation flow when tested pursuant to Specification 4.0.5: p g~su8>~Centrifugal charging pump 2.Safety injection pump 3.Residual heat removal pump f 0 psia Creater than or eaual to~~ah@" f38'S PSid ICc'Ps(g Creater than or ecpaal to~Q-pe4g-gi By verS.fying the correct position of each mechanical stop for the following Emergency Core Cooling System throttle valves: W.thin 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following completion of each valve st oking operation or maintenance on the valve when the ECCS sub-systems are required to be OPERABLE.l'he provisions of Technical Soecification 4.0.8 are applicable. COOK NUCLEONS PLZ~XT-UN77 2 3/4 5-5 a=-NOV:-NT NO."',~,-'o, 159 EMERGENCY CORE COOLING SYSTEMS REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.5 The refueling water storage tank (RWST)shall be OPERABLE with: a.A minimum contained volume of 350,000 gallons of borated water, I b.Between 2400 and 2600 ppm of boron, and c.A minimum water temperature of 49-F-.APPLICABILITY: MQOES 1, 2, 3 and 4.M')0'F ACTION: With the refueling water storage tank inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .SURVEILLANCE RE UIREMENTS 4.5.5 The RWST shall be demonstrated OPERABLE: a.At least once per 7 days by: l.Verifying the contained borated water volume in the tank, and 2.Verifying the boron concentration of the water.b.At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature. 0.C.COOK-UNIT 2 3/4 5-11 Amendment No.3984 3 4.1 REACTIVITY CONTROL SYSTEMS'3 4-1.1 BORATION CONTROL 3 4.1.1.1 and 3 4.1.1.2 SHUTDOWN MARGIN h sufficient SHUTDOWN MARGIN ensures that 1)tha reactor'an be made subcritical from all operating conditions, 2)the x'eactivity transients associated vith postulated accident conditions are controllable vithin acceptable limits, and 3)the reactor vill be maintained sufficiently subcritical to preclude inadvertent criticality. in the shutdovn condition. l.3%SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T.The most xestrictive condition occurs't EOL, vith T at no)oad operkfng temperature, and is associated vith a postulated steal line break accident and resulting av uncontrolled RCS cooldovn.In the analysii of this accident, a minimum SHUTDOWN MARGIN of~Delta k/k is initially required to~control the reactivity transient and automatic ESF is assumed to be available. Qith T less than 200 F, the reactivity transients resulting from a 0 au postulated st5am line break cooldovn az'e minimal and a 1%Delta k/k SHUTDOQN MARGIN provides adequate protection for this event.~r The SHUTDOWN MARGIN requirements are based upon the limiting conditions described above and ara consistent vith FSAR safety analysis assumptions. 3 4.1.1.3 BORON DILUTION h minimum R.ov rate of at least 2000 CPM provides adequate mixing, prevents stratification and ensures that reactivity changes vill be gradual during boron concentration reductions in the Reactor Coolant System.h flov xata of at least 2000 GPM vill circulate an equivalent Reactor Coolant System volume of 12,612 cubic feet in approximately 30 minutes.The reactivity change xate associated vith boron reductions vill therefore be vithin the capability for operator recognition and contxol.0 COOK NUCLEAR PLANT-UNIT 2 B 3/4 1-1 AMENDMENT NO.HZiXSSs$34 ( I l EMERGENCY COOLING SYSTEMS BASES 3 4.5.5 REFUELING VhTER STORAGE TANK The OPERABILITY of che RVST aa pare of the ECCS ensures that sufficient negative reactivity ia in]ected into the cora to counteract any positive increase in feactivity caused by RCS ayacam cooldovn, and ensures that a suffi.cient supply of boraced vater ia availabla for injection by cha ECCS in the event of a LOCh.Reactor coolant system cooldovn can be caused by.inadvertent depreaauriration, a LOCh or steam line rupture.The limits of RVST minimum volume ind boron concentration ensure that 1)sufficient vater is available vithin containment co permit recirculation cooling flov to the core, and 2)che reactor vill remain aubcritical in the cold condition folloving These assumpti.ons <xone,kv A are consistent vith the LOCh analyses.The contained vater volume limit includes an allovance for vater not usable because of tank discharge line location.or other physical characteristi.cs. The limits on contained vater volume and boron concentration of the RVST also ensure a pH valu~of betveen 7.6 and 9.5 for che solution recirculated vithin containment afcer a LOCh, This pH band minimizes the evolution of iodine and mlnlmires che effect.of chloride and caustic stress corrosion on mechanical systems and components. ~~F The ECCS analyses to determine F limi in Specifications 3.2.2 and 3.2.6 assumed a RVST vacer temperature of.This temperature value of the RVST vater determines that of che spray vater ini.cially delivered to che concainmenc folloving LOCh.It is one of che factors vhich determines the containment back'-pressure in che ECCS analyses, performed in accordance vith.-." the provisions of l0 CFR 50.46 and hppendhc K to 10 CFR 50.1NSERT A a LOCA assuming mixing of the RWST, RCS ECCS water, and other sources of water that may eventually reside in the sump, with all control rods assumed to be out./A COOK NUCLEAR PLANT-UNIT 2 5 3/4 5-3 AMENDMENT NO.797, 14Z CONTAINMEHT S YSTEHS 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. rasul6>>ii.qqThe maximum peak pressure is 4-.4 psig from a LOCA event 0.5 ps'~taittwL~aH~covl~A~ pvcssvhc.~3/4.6.1.5 AIR TEHPERATURE (Qg~t~4&5 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'nstrumentation located, within containment. I~'The containment pre sure transient. is sensitive to the initially contained air mass duri g a LOCA.The contai'ned aii mass increases with decreasing temperature. The lower temperature limit of 60'F will limit the peak pressure to.psig which is 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 para-meters used in the accident analyses.3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the con-tainment 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 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. D.C.COOK-UNIT 2 B 3/4 6-2 ATTACHMENT 4 TO AEP:NRC:1207
SUMMARY

DESCRIPTION OF PROPOSED INCREASED STEAM GENERATOR TUBE PLUGGING TECHNICAL SPECIFICATIONS Attachment 4 to AEP:NRC:1207 Page 1 Key for Summary Table Page Section Group Technical Specification Page Technical Specification Related Groups Discussed in Attachment 1, Description of Proposed Changes and 10 CFR 50.92 Significant Hazards Consideration Analysis SGTP Margin Both Group 1, Changes Directly Related to Increased Steam Generator Tube Plugging Group 2, Changes Proposed to Increase unit 1 Operating Margin Group 3, Changes Proposed to Increase the Operating Margin of Both units.Description Remarks Admin Group 4, Administrative Change A Brief Description of Each Proposed Change Brief Comments with a Cross Reference to the Analyses 0
Attachment 4 to AEP:NRC:1207 Page 2 Page 2-2 2-5 2-7 2-8 2<<8 2-9 2-9 Section Figure 2.1-1 Table 2.2-1 Footnote Table 2.2-1 Table 2.2-1 Table 2.2-1 Table 2.2-1 Table 2.2-1 Group Margin SGTP Margin Margin Margin Margin Margin Description Revise Reactor Core Safet Limits Redefine design flow in footnote of Table 2.2-1 to be 1/4 MMF.The upper limit on T'ncreased to reflect analyses.Decrease K1 from 1.32 to 1.17.Change f (n,I)to increase the region of positive al which is without enalt Decrease the upper limit on Tto reflect analyses.Change the allowable values in note 2 and 3.Remarks The new thermal design is discussed in Section 3.3.2.1 of Attachment 6, WCAP 14285.MMF for DNB is discussed in Section 3.3.2.1 of Attachment 6, NCAP 14285.T.S.MMF is 1.025 times thermal design flow (TDF).TDF is specified in Section 3.3.3.1 of NCAP 14285.The MMF employed in the DNB analysis is 1.019 times TDF.This was done to support a range of MMF's from 1.019 to 1.025 times TDF as indicated in Section 2.1 of WCAP 14285.Design flow in current technical specification Table 2.2-1 is MMF/4.The OTDT trip is discussed in Section 3.3.2.1 of NCAP 14285.Details, including T', of the analyzed setpoint are in Table 3.3-3 of WCAP 14285.This change and the next change to f (al)are being requested to optimize operating margin.Some load rejection capability is sacrificed for instrumentation margin, increased allowance for core burndown effects on hot leg streaming, and an increase in the positive dl break point for the f(~I)penalty.The OTDT trip is discussed in Section 3.3.2.1 of WCAP 14285.Details of the analyzed setpoint are in Table 3'-3 of NCAP 14285.See previous discussion of K1 decrease.Cook Nuclear Plant unit 1 is operated in a low temperature, low pressure mode to extend the life of the steam generators.
Therefore, the analysis of the OPDT setpoint was analyzed with a low upper limit on Tto convert unused margin to operating margin.The OPDT trip is discussed in Section 3.3'.1 of NCAP 14285.Details, including T, of the analyzed setpoint are in Table 3.3-3 of WCAP 14285.The values indicated in the markups of Attachment 3 and in the proposed technical specifications of Attachment 2 were calculated by our organization.
Attachment 4 to AEP:NRC:1207 Page 3 Page UNIT 1 3/4 1-1 Section UNZT 1 Section 3.1.1.1 4;1.1.1.1 Group Both Description Reduce required shutdown margin.Remarks The new value is supported by analyses.Unit 1: For core response steam break (CRSB), see Section 3.3.5.6 of WCAP 14285~For steamline mass and energy release (SM&E)inside containment, see Section 3.5.4.2 of WCAP 14285.For SM&E outside containment, see Section 3.3.4.7 of WCAP 14285.UNIT 2 3/4 1-1 UNIT 1 3/4 1-11 UNIT 2 Section 3.1.1.1 4.1.1.1 UNIT 1 Section 4.1.2.3.1 Both Change CCP surveillances to be consistent with 10%degradation.
Change pump surveillance requirements from discharge pressure to differential pressure.Unit 2: For CRSB, see Section B.3.11 of the Vantage 5 Reload Transition Safety Report for Cook Nuclear Plant unit 2 (RTSR).For SM&E inside containment, see Section 3.5.4.2 of WCAP 14285.For SM&E outside containment, see Section 3.3.4.7 of WCAP 14285.A copy of Section B.3.11 has been included in Attachment 7 to this submittal.
The new surveillance criterion is supported by analyses.The surveillance criteria are presented in Section 3.10.1.1 of WCAP 14285.The value given for CCP applies to both-units Unit 1: For LOCA see Sections 3.1.1 and 3.1.2 of WCAP 14285.For core response steam break (CRSB), see Section 3.3.5.6 of WCAP 14285.For steamline mass and energy release (SM&E)inside containment, see Section 3.5.4.2 of WCAP 14285.For SM&E outside containment, see Section 3.3.4.7 of WCAP 14285.UNIT 2 3/4 1-11 UNIT 2 Section 4.1.2.3.1 Unit 2: The technical specification changes supported by the RTSR are delineated in Table 6.1 of the RTSR.For LOCA see Sections C.3.1.2 and C.3.2 of the RTSR.For CRSB, see Section BE 3.11 of the RTSR.For SM&E inside containment, see Section 3'.4.2 of WCAP 14285.For SM&E outside containment, see Section 3.3.4.7 of WCAP 14285.Copies of Table 6.1 and Sections BE 3.11, C.3.1.2, and C.3.2 have been included in Attachment 7 to this submittal.
Attachment 4 to AEP:NRC:1207 Page 4 Page UNIT 1 3/4 1-12 UNIT 2 3/4 1-12 Section UNIT 1 Section 4.1.2.4 UNIT 2 Section 4.1.2.4 Group Both Description Change CCP surveillances to be consistent with 10%degradation.
Change pump surveillance requirements from discharge pressure to differential Remarks See previous discussion of CCP degradation increase for page 3/4 1-11.UNIT 1 3/4 1-15 UNIT 2 3/4 1-15 UNIT 1 3/4 1-16 UNIT 2 3/4 1-16 3/4 2-14 3/4 2-14 UNIT 1 Section 3.1.2.7 UNIT 2 Section 3.1.2.7 UNIT 1 Section 3.1.2~8 UNIT 2 Section 3.1.2.8 Table 3.2-1 Table 3.2-1 Both Both SGTP SGTP Reduce the Minimum RWST temperature to 70 F.Reduce the Minimum RWST temperature to 70 F.Increase DNB temperature limit Reduce MMF limit The mode 5 and 6 minimum RWST temperature is conservatively maintained at the same value as that required for modes 1, 2, 3, and 4.The new value is supported by the core response LBLOCA.Unit 1: See Section 3.1.1, Table 3.1-2, of WCAP 14285'nit 2: See Table C.3.1-2 of the RTSR.A copy of this table is included in Attachment 7.The calculation of the new DNB temperature limit is described in Section 1.2 of WCAP 14285 under the heading"DNB Parameters, RCS Tavg and RCS Flow".The readability error is 2.14F.The resulting DNB temperature limit is 579.30F.See the discussion for page 2-5.
Attachment 4 to AEP:NRC:1207 Page 5 Page 3/4 3-17 3/4 3-21 3/4 3-23a 3/4 3-24 3/4 3-26 3/4 3-31 Section Table 3.3-3 Table 3.3-3 Table 3.3-3 Table 3.3-4 Table 3.3-4 Table 4.3-2 Group Margin Margin Margin Margin Margin Margin Description Change ESF actuation logic to support 12%AFW Pump degradation Change ESF actuation logic to support 12%AFW Pump de radation Change ESF actuation logic to support 12%AFW Pump degradation Change ESF actuation logic to support 12\AFW Pump de radation Change ESF actuation logic to support 12%AFW Pump degradation Change ESF actuation logic to support 12k AFW Pump degradation Remarks The revised part of Table 3.3-3 incorporates the safeguards logic used in Cook Nuclear Plant unit 2.This will allow for the use of 12%auxiliary feedwater head degradation (AFW)in unit l.All analyses, other than an"information only" feedline break analyses, have been performed using the flow from an AFW pump with 12%head degradation.
The safeguards logic itself will be modified via design change prior to implementation of these revised T/S pages (i.e.before unit 1, cycle 16).After this modification, the unit 2 feedline break analysis using 12%degraded flow will bound unit 1.The evaluation which shows that the unit 2 feedwater line break will bound unit 1 is discussed in Section 3.3.4.8 of WCAP 14285.See discussion for page 3/4 3-17 See discussion for page 3/4 3-17 See discussion for page 3/4 3-17 See discussion for page 3/4 3-17 See discussion for page 3/4 3-17
Attachment 4 to AEP:NRC:1207 Page 6 Page Section Group Description Remarks 3/4 3-33 Table 4.3-2 Margin Change ESF actuation logic to support 12k AFN Pump degradation See discussion for page 3/4 3-17 3/4 4-4 3/4 4-5 Unit 1 3/4 5-5 Unit 2 3/4 5-5 UNIT 1 3/4 5-5 UNIT 2 3/4 5-5 Section 3.4.2 Section 3.4.3 Section 4.5.2.f.2 4.5.2.f.3 Section 4.5.2.f.2 4.5.2.f.3 UNIT 1 Section 4.5.2.f.1 UNIT 2 Section 4.5.2.f.l Margin Margin Margin Admin Both Increase Pressurizer Valve Tolerance Increase Pressurizer Valve Tolerance Change RHR/SI pump surveillances to be consistent with 15%degradation.
Change RHR/SZ pump surveillance requirements from discharge pressure to differential pressure.Change RHR/SI pump surveillance requirements from discharge pressure to differential pressure.Change CCP surveillance to be consistent with 10%degradation.
Change pump surveillance requirements from discharge pressure to differential pressure.The Non-LOCA accidents were reanalyzed or reevaluated based on a pressurizer valve setpoint tolerance of 3%.This is noted in section 1.1 and 3.3.2.3 of NCAP 14285.See discussion for page 3/4 4-5.15%degradation of the SZ and RHR pumps is discussed in Sections 1.1 and 1.2 of WCAP 14285.LBLOCA is discussed in Section 3.1.1g SBLOCA is discussed Section 3.1.2;and LOCA mass and energy release (M&E)is discussed in Section 3.5.2.1 of WCAP 14285.The new surveillance criteria are supported by analyses.The surveillance criteria are presented in Section 3.10.1 of NCAP 14285'his an administrative change.The discharge pressure criteria in the current technical specifications correspond to the same pump performance characteristics as the proposed differential pressure criteria.The change ensures that surveillance criteria use similar acceptance criteria.The surveillance acceptance criteria for 10%degraded pumps were provided in the technical specification mark-ups of the RTSR.A copy of the mark-up of page 3/4 5-5 is included in Attachment 7.Refer to the discussion given for page 3/4 1-11 for information concerning the 10%CCP degradation.
Attachment 4 to AEP:NRC:1207 Page 7 Page Section Group Description Remarks UNIT 1 3/4 5-11 UNIT 1 Section 3.5.5 Both Reduce minimum RWST temperature to 70oF See discussion for page 3/4 1-16.UNIT 2 3/4 5-11 3/4 7-6 5-5 B 2-1(a)B 2-4 B 2-5 UNIT 2 Section 3.5.5 Sections 4.7.1.2.a and 4.7.1.2.b Section 5.4.2 Bases Section 2.1.1 Bases Section 2.2.1 Bases Section 2.2.1 Margin SGTP Margin Margin Margin Change AFW Pump surveillance to be consistent with 12%degradation Reduce system volume to account for plugged steam enerator tubes.Change DNB Values for Fuel Remove detail from the discussion of the OTDT rotection tri Remove detail from the discussion of the OPDT protection trip.See discussion for page 3/4 3-17.The proposed surveillance criteria is identical to the criteria in the unit 2 technical specifications.
These criteria correspond to the auxiliary feedwater flows used in all analyses for both units except the"information only" unit 1 feedwater line break.As noted in the discussion for page 3/4 3-17, after the changes to the unit 1 safeguards actuation logic, unit 1 will be bounded b the unit 2 feedwater line break.A volume range corresponding to 0%to 30%plugging is specified.
See section 1.2 of WCAP 14285.The values for DNBR for typical and thimble cells are being revised.The revised values are noted in Sections 1.2 and 3.3.2'of WCAP 14285.This change is related to the new thermal design and the new OTDT and OPDT protection trip setpoints.
Therefore, see also discussions for pages 2-2, 2-7, 2-8, and 2-9.The discussion of the proper normalization of T'nd P's being removed.This information is documented in Section 3.3.2.1 of WCAP 14285 and will be controlled administrativel The discussion of the proper normalization of Tis being removed.This information is documented in Section 3.3.2.1 of WCAP 14285 and will be controlled administratively.
Attachment 4 to AEP:NRC:1207 Page 8 Page UNZT 1 B 3/4 1-1 UNIT 2 B 3/4 1-1 B 3/4 4-1 UNIT 1 B 3/4 5-3 UNIT 2 B 3/4 5-3 UNIT 1 B 3/4 6-2 UNIT 2 B 3/4 6-2 Section UNIT 1 Bases Section 3/4.1.1.1 and 3/4.1.1.2 UNIT 2 Bases Sections 3/4.1.1.1 and 3/4.1.1.2 Bases Section 3/4.4.1 UNIT 1 Bases Section 3/4.5.5 UNIT 2 Bases Section 3/4.5.5 UNIT 1 Bases Sections 3/4.6.1.4 3/4.6.1.5 UNIT 2 Bases Sections 3/4.6.1.4 3/4.6.1.5 Group Both Margin Both Both Description Reduce required shut down margin Change DNB Values for Fuel Reduce the minimum RWST temperature to 70'F.Change peak containment pressure to reflect analysis result Remarks See discussion for page 3/4 1-1.Change"1.69" to"the safety analysis limit".Clarify conditions under which the reactor will remain subcritical.
Specifically, LBLOCA is called out as the initiating condition and the control rods are assumed to be out instead of being inserted except for the most reactive assembly.In addition, the explanation that a conservatively high value of the RWST temperature is included in the technical specifications for unit 1 is being removed because the proposed value of 704F is based on the analyses.See the discussion for a e 3/4 1-16.Discussion of maximum calculated containment pressure is given in Sections 1.2 and 3.5.3.4 of WCAP 14285.
ATTACHMENT 5 TO AEP:NRC:1207 DISCUSSION OP PREVIOUS RELATED SUBMISSIONS 0
Attachment 5 to AEP:NRC:1207 Page 1 Introduction Attachment 6 to this submittal is WCAP 14285.It describes the analyses and evaluations performed by Westinghouse Electric Corporation in order to support a reduced thermal design flow and a reduced minimum measured flow which are expected to result from increased steam generator tube plugging to the level of 30%in the unit 1 steam generators.
Zt also describes analyses and evaluations performed simultaneously to support certain increases in operating margin such as increased setpoint tolerance for the pressurizer safety valves.As discussed in Section 2.0 of WCAP 14285, the new analyses replace analyses performed earlier to support the operation of Cook Nuclear Plant and the evaluations described in WCAP 14285 are based on those earlier analyses.The earlier analyses are described in WCAP 11902 and WCAP 11902 Supplement 1, references 3 and 10.They are referred to as the"Rerating Program" in WCAP 14285.The purposes of this attachment are to: 1.indicate those aspects of earlier analyses which have been submitted for NRC review and approved, 2.indicate those portions of these analyses which have not previously been submitted for review.3.describe the earlier analyses, 4.provide references for previous submittals for the convenience of the reviewer, and This submittal includes some proposed technical specification changes for both units.Therefore, the discussion of this attachment describes submittals for both units.The discussion of this attachment describes the applications made by us to implement the features supported by the earlier analyses and the approvals received.This is significant because it will assist in clarifying available margin and because there are increased operating margins supported by the earlier analyses which we have not previously implemented.
Zn some cases, we have not submitted a request due to the desire to maintain the technical specifications for the two Cook units as nearly alike as possible.The following lists summarizes the status of analysis features of earlier analyses: Princi al Features of the Earlier Anal ses Which Have been Reviewed and A roved 2.3.4~5.6.Reduced temperature and pressure operation for unit 1.Reduced temperature operation for unit 2.10%degradation for the RHR and HHSI pumps for both units.Increased MSZV response time for both units.BIT 0 ppm boric acid concentration for both units.Reduced MMF for unit 1.
Attachment 5 to AEP:NRC:1207 Page 2 Princi al Features of the Earlier Anal ses Which Have Not Been Submitted for Review 2.3.4~5.unit 1 rerate to 3413 MWt.The available power margin is allocated in this submittal to allow for increased steam generator tube plugging.unit 2 rerate to 3588 MWt.Additional analytic work remains to be completed.
10%degradation for the centrifugal charging pumps for both units.Approval to implement this feature is requested in this submittal.
Minimum RWST temperature of 704F.Approval to implement this feature is requested in this submittal.
SDM requirement of 1.3%.Approval to implement this feature is requested in this submittal.
Pu ose of the Earlier Anal ses Reratin Pro ram The earlier analyses were performed to accomplish a number of goals.The most urgent 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 units had to be reviewed or revised, all the analyses were performed to position unit 1 for subsequent uprating to 3413 MWt core power and unit 2 to 3588 MWt core power.As of this time, We have not requested NRC review of uprating either unit.This submission proposes to use the margin between the unit 1 analyzed, uprated power and licensed, rated thermal power to accommodate the increased tube plugging.Finally, the earlier analyses 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), removal of the boron injection tanks (BIT), and slower response times for certain components and systems.This submittal requests approval for the implementation of an allowed 10%degradation for the ECCS centrifugal charging pumps, reduction of required SDM, and a reduction, in the minimum temperature of the RWST's for both units, which is supported in part by the earlier analyses.Descri tion and Review Histo of Prior Submittals 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.Reference 2 requested approval for operation with RHR cross ties closed.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 Attachment 5 to AEP:NRC:1207 Page 3 analyses presented in reference 3 were performed at a core power of 3413 MWt.However, the evaluations described in reference 3 supported operation at a core power of 3250 MWt.Reference 3 also supported 10%degradation of the unit 1 RHR and high head safety injection (HHSZ)pumps and a minimum RWST temperature of 704F.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 with 10%degraded RHR and HHSZ pumps.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 a 2 Licensing Report, describes the balance of the analyses which were performed by Westinghouse Electric Corporation to support the operation of unit 1 at 3413 MWt.Zn particular, an analysis of the steam mass and energy release (SMaE)to containment, the associated containment analysis, and the SM&E outside containment are included in this report.These two analyses were performed to bound both units at the unit 2 uprated core power of 3588 MWt.Together with reference 3, the analyses of reference 10 completed the Westinghouse scope of analyses to support an additional 3 seconds for the response time of the main steam isolation valves (MSIV), 0 ppm boric acid concentration in the BIT, and 10%degradation of the CCP's for unit 1.The two SMaE analyses were performed assuming a SDM of 1.3%.However, the core response steam break analysis reported in reference 3 assumed a SDM of 1.6%.Reference 11, Vantage 5 Reload Transition Safety Report for Cook Nuclear Plant unit 2 (RTSR), together with reference 1, Containment Integrity Analysis, reference 3, WCAP 11902, Reduce'd 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 cross tie 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 10%degradation of the CCP's, HHSI pumps, and RHR pumps, an increase of 3 seconds in MSZV response time for unit 2, 0 ppm boric acid concentration in the BZT 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 SM&E to the containment.
The letters of references 14, 15, and 16 provided supplementary information to the staff xelated to reference 13.Operation of unit 2 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.
Attachment 5 to AEP:NRC:1207 Page 4 The letters of references 18 and 19 proposed technical specifications that implemented an increase of 3 seconds in the MSIV response times.These proposals were supported by reference 3, WCAP-11902, reference 10, WCAP-11902, Supplement 1, reference 11, RTSR, 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 SM&E to the containment.
The proposals to increase the MSIV 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.An evaluation, performed by Westinghouse Electric Corporation, allocated available margin in MMF to the flow reduction.
The evaluation was included in the submittal.
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.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 setpoints for both Cook 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, 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.This proposal was approved by reference 28.References 2.3.4~5.6.WCAP-11908, Containment Integrity Analysis for Cook Nuclear Plant units 1 and 2, M.E.Wills, July 1988.Letter AEP:NRC:1024D, Containment Long Term Pressure Analysis to Support RHR Cross Tie 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.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.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.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.
Attachment 5 to AEP:NRC:1207 Page 5 7.8.9.10.12.13.14.15.16~17.18.19.20.21.22.23.24 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.Letter AEP:NRC:1067D, Modification of Reduced Temperature and Pressure Program technical specification Changes, from M.P.Alexich to T.E.Murley, June 5, 1989.Amendment No.126 to Facility Operating License No.DPR-58.WCAP 11902, Supplement 1, Rerated Power and Revised Temperature and Pressure Operation for Cook Nuclear Plant units 1 6 2 Licensing Report, September 1989.Vantage 5 Reload Transition Safety Report for Cook Nuclear Plant unit 2, B.W.Gergos, Editor, January 1990.No reference 12.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.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, 1990.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.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.Amendment No.148 to Facility Operating License No.DPR-58 and Amendment No.134 to Facility Operating License No.DPR-74.Letter AEP:NRC:1120, Expedited technical specification Change Request Steam Generator Stop Valves, from M.P.Alexich to T.E.Murley, January 31, 1990.Letter AEP:NRC:1123, technical specification Change Request, Steam Generator Stop Valves, from M.P.Alexich to T.E.Murley, May 14, 1990.Amendment No.147 to Facility Operating License No.DPR-58.Amendment No.135 to Facility Operating License No.DPR-74.Letter AEP:NRC:1130, technical specification Change for unit 1 Cycle 11, from M.P.Alexich to T.E.Murley, July 23, 1990.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.
t Attachment 5 to AEP:NRC:1207 Page 6 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 ll, 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.
1 It I ATTACHMENT 6 TO AEP:NRC:1207 DESCRIPTION OF ANALYSES PERFORMED BY WESTINGHOUSE ELECTRIC CORPORATION FOR COOK NUCLEAR PLANT UNIT 1
Attachment 6 to AEP:NRC:1207 Page 1 WCAP 14285
ATTACHMENT 7 TO AEP:NRC:1207 DESCRIPTION OF ANALYSES PERFORMED BY WESTINGHOUSE ELECTRIC CORPORATION FOR COOK NUCLEAR PLANT UNIT 2
ATTACHMENT 6 TO AEP:NRC:1207 DESCRIPTION OF ANALYSES PERFORMED BY WESTINGHOUSE ELECTRIC CORPORATION FOR DONALD C.COOK NUCLEAR PLANT UNIT 1 Attachment 6 to AEP:NRC:1207 Page 1 WCAP 14285 WESTINGHOUSE NONPROPRIETARY CLASS 3 WCAP-14285, Revision 1 DONALD C.COOK NUCLEAR PLANT UNIT 1 STEAM GENERATOR TUBE PLUGGING PROGRAM LICENSING REPORT MAY 1995 WESTINGHOUSE ELECTRIC CORPORATION Energy Systems Business Unit P.O.Box 355 Pittsburgh, Pennsylvania 15230 1995, Westinghouse Electric Corporation, All Rights Reserved mA1 944-1 w.wpf: 1d%51895 TABLE OF CONTENTS Title~Pa e List of Tables List of Figures List of Acronyms and Abbreviations Definitions Vll XXIII Summary and Conclusions XXIV

1.0 INTRODUCTION

-DESCRIPTION OF LICENSE AMENDMENT REQUEST 1.1 Purpose for Change 1.2 Current License Basis and Function of Identified Technical Specification and Description of Proposed Change 1.1-1 1.1-1 1.2-1 2.0 BASIS FOR EVALUATIONS/ANALYSES PERFORMED 2.1 Design Power Capability Parameters 2.2 NSSS Design Transients

2.3 ControVProtection

System Setpoints 2.0-1 2.1-1 2.2-1 2.3-1 3.0 SAFETY EVALUATIONS/ANALYSES PERFORMED 3.1 Loss of Coolant Accident Analyses 3.2 LOCA Hydraulic Forces 3.3 Non-LOCA Analyses 3.4 Post-LOCA Hydrogen Production

3.5 Containment

Analyses 3.6 Steam Generator Tube Rupture Accident Analysis 3.7 Post-LOCA Hot Leg Recirculation Time 3.8 Reactor Cavity Pressure Analysis 3.9 Radiological Analysis 3.10 Fluid and Auxiliary Systems Evaluations 3.11 Primaly Components Evaluations 3.11.1 Steam Generators 3.11.2 Reactor Vessel 3.11.3 Reactor Internals 3.11.4 Control Rod Drive Mechanisms 3.11.5 Reactor Coolant Pumps 3.1-1 3.1-1 3.2-1 3.3-1 3.4-1 3.5-1 3.6-1 3.7-1 3.8-1 3.9-1 3.10-1 3.11-1 3.11-1 3.11-4 3.11-8 3.11-11 3.11-12 m 51944-1w.wpf:1d~1295 TABLE OF CONTENTS (continued)
Section Title Pacae 3.11.6 Pressurizer 3.11.7 Reactor Coolant Loop Piping and Supports 3.11.8 Auxiliary Components 3.12 Fuel Structural Evaluation 3.11-13 3.11-15 3.11-16 3.12-1

4.0 CONCLUSION

S APPENDIX A Proposed Technical Specification Changes 4-1 m 61944-1w.tNpf:1d~1295
LIST OF TABLES Title 1.2-1 2.1-1 3.1-1 3.1-2 3.1-3 3.1-4 3.1-5 3.1-6 Summa'f Technical Specification Changes NSSS Performance Parameters for SGTP Program Large Break LOCA Results Plant Input Parameters Used in Large Break LOCA Analysis Large Break Containment Data (Ice Condenser Containment)
Mass and Energy Release Rates, Minimum Sl Nitrogen Mass and Energy Release Rates Safety Injection Flow Rate: Rerating Program Analysis 3.1-7 Plant Input Parameters Used in Small Break LOCA Analyses: Rerating Program Analysis 3.1-8 3.1-9 3.1-10 3.1-11 Small Break LOCA Calculation:
Rerating Program Analysis Time Sequence of Events for Condition III Events: Rerating Program Analysis Small Break LOCA Calculation:
Rerating Program Analysis Time Sequence of Events for Condition III Events: Rerating Program Analysis 3.1-12 Plant Input Parameters Used in Small Break LOCA Analysis:+/-3%Main Steam Safety Valve Setpoint Tolerance Analysis 3.1-13 Time Sequence of Events for Condition III Events:+/-3%Main Steam Safety Valve Setpoint Tolerance Analysis 3.1-14 Small Break LOCA Calculations:
+/-3%Main Steam Safety Valve Setpoint Tolerance Analysis 3.1-15 Plant Input Parameters Used in Small Break.LOCA Analysis: 30%SGTP Program Analysis with HHSI Cross-Ties Closed m f1944-1w.wpf:1d~1195 f!0 LIST OF TABLES (continued)
Title 3.1-16 Time Sequence of Events for Condition III Events: 30%SGTP Program Analysis with HHSI Cross-Ties Closed 3.1-17 Small Break LOCA Calculations:
30%SGTP Program Analysis with HHSI Cross-Ties Closed 3.3-1 3.3-2 NSSS Performance Parameters Used in Non-LOCA Safety Analyses Trip Points and Time Delays to Trip Assumed in Non-LOCA Accident Analysis 3.3-3 OTbT and OPbT Setpoint Equation and Safety Analysis Limit Coefficient Values 3.3-4 3.3-5 3.3-6 Summary of Initial Conditions and Computer Codes Used Sequence of Events for Loss of Flow and Locked Rotor Accidents Sequence of Events for Loss of External Electrical Load 3.3-7 Limiting Steamline Break Statepoint Double Ended Rupture Inside Containment with Offsite Power Available 3.3-8 Time Sequence of Events-Double Ended Rupture Inside Containment with Offsite Power Available 3.3-9 Parameters Used in the Analysis of the Rod Cluster Control Assembly Ejection Accident 3.5-1 3.5-2 System Parameters, Initial Conditions Safety Injection Flow, Minimum Sl m:51944-1w.wpf:1d441195 IV LIST OF TABLES (continued)
Title 3.5-3 Double-Ended Pump Suction Guillotine Minimum Sl Blowdown Mass and Energy Release 3.5-4 Double-Ended Pump Suction Guillotine Minimum Sl Ref lood Mass and Energy Release 3.5-5 Double-Ended Pump Suction Guillotine Minimum Sl Principal Parameters During Ref lood 3.5-6 Double-Ended Pump Suction Guillotine Minimum Sl Post Ref lood Mass and Energy Release 3.5-7 3.5-8 Double Ended Pump Suction Guillotine Minimum Sl Mass Balance Double Ended Pump Suction Guillotine Minimum SI Energy Balance 3.5-9 3.5-10 3.5-11 3.5-12 Energy Accounting in Millions of BTU Energy Accounting in Millions of BTU Structural Heat Sink Table Material Properties Table 3.5-13-3.11-1 Steamline Break Mass/Energy Releases Inside Containment Performance Characteristics at 3262 MWt 3.11-2 Assumed Operating Parameters for Reactor Vessel Structural Evaluation for Cook Nuclear Plant Unit 1 3.11-3 3.12-1 3.12-2 Pressurizer Components Calculated Fatigue Usages Considering 30%SGTP Maximum LOCA and DBE Grid Load Results Fuel Rod Design Analysis Parameters m%1944-1w.wpf:1d 441195 LIST OF TABLES (continued)
Title 3.12-3 3.12-4 30%SGTP Program Thermal Hydraulic Design Parameters DNBR Limits and Margin Summary m:$1944-1w.wpf:1d~1195 Vl
LIST OF FIGURES~Fi ere Title 3.1-1a-f 3.1-2a-f 3.1-3a-f 3.1-4a-f 3.1-5a-f 3.1-6a-f 3.1-7a-f 3.1-8a-f 3.1-9a-f Reactor Coolant System Pressure.Cases A-F Break Flow During Blowdown, Cases A-F Core Pressure Drop, Cases A-F Core Flowrate, Cases A-F Accumulator Flow During Blowdown, Cases A-F Vessel Liquid Levels During Ref lood, Cases A-F Core Inlet Flow During Ref lood Accumulator and Sl Flow During Ref lood, Cases A-F Integral of Core Inlet Flow, Cases A-F 3.1-10a-f Mass Flux at Peak Temperature Elevation, Cases A-F 3.1-11a-f Rod H.T.C.at Peak Temperature Elevation, Cases A-F 3.1-12a-f Vapor Temperature, Cases A-F 3.1-13a-f Fuel Rod Peak Clad Temperature, Cases A-F 3.1-14 Containment Pressure, CD=0.4, Min.Sl 3.1-1 5 3.1-16 3.1-17 3.1-18 3.1-1 9 Upper Compartment Structural Heat Removal Rate, CD=0.4, Min.Sl Lower Compartment Structural Heat Removal Rate, CD&.4, Min Sl Heat Removal by Sump, CD=0.4, Min.Sl Heat Removal by Lower Compartment Spray, CD=0.4, Min.Sl Containment Temperature, CD=0.4, Min.Sl m%1944-1w.wpf:1d441195 VII LIST OF FIGURES (continued)
~Fi ere 3.1-20 3.1-21 Title Safety Injection Flow Rate Donald C.Cook Unit 1 Hot Rod Power Distribution Donald C.Cook Unit 1 3.1-22 RCS Pressure (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-23 Core Mixture Height (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-24 Hot Spot Clad Temperature (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-25 Core Steam Flowrate (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-26 Hot Spot Heat Transfer Coefficient (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-27 Hot Spot Fluid Temperature (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-28 Total Break Flow (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-29 Intact Loop Pumped Sl Flow (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-30 RCS Pressure (2 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-31 Core Mixture Height (2 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-32 Hot Spot Clad Temperature (2 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-33 Core Steam Flowrate (2 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mal 944-1w.wpl:1d 441195 Vill
LIST OF FIGURES (continued)
Ficiure 3.1-34 Title Hot Spot Heat Transfer Coefficient (2 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-35 Hot Spot FLuid Temperature (2 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-36 Total Break Flow (2 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-37 Intact Loop Pumped Sl Flow (2 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-38 RCS Pressure (r Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-39 Core Mixture Height (4 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-40 Hot Spot Clad Temperature (4 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-41 Core Steam Flowrate (4 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-42 Hot Spot Heat Transfer Coefficient (4 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-43 Hot Spot Fluid Temperature (4 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-44 RCS Pressure (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-45 Core Mixture Height (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-46 Hot Spot Clad Temperature (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 m&1944-1w.wpf:1d~l 195 IX LIST OF FIGURES (continued)
~Fi ure Title 3.1-47 Core Steam Flow Rate (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-48 Hot Spot Heat Transfer Coefficient (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-49 Hot Spot Fluid Temperature (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-50 Total Break Flow (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-51 Intact Loop Pumped Sl Flow (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.1-52 RCS Pressure (3 Inch)High Temperature, High Pressure Donald C.Cook Unit 1 3.1-53 Core Mixture Height (3 Inch)High Temperature, High Pressure Donald C.Cook Unit 1 3.1-54 Hot Spot Clad Temperature (3 Inch)High Temperature, High Pressure Donald C.Cook Unit 1 3.1-55 Core Steam Flowrate (3 Inch)High Temperature, High Pressure Donald C.Cook Unit 1 3.1-56 Hot Spot Heat Transfer Coefficient (3 Inch)High Temperature, High Pressure Donald C.Cook Unit 1 3.1-57 Hot Spot Fluid Temperature (3 Inch)High Temperature, High Pressure Donald C.Cook Unit 1 3.1-58 Total Break Flow (3 Inch)High Temperature, High Pressure Donald C.Cook Unit 1 3.1-59 Intact Loop Pumped Sl Flow (3 Inch)High.Temperature, High Pressure Donald C.Cook Unit 1 m:i'll 944.1w.wpf:1d441195 LIST OF FIGURES (continued)
~Fi ere 3.1-86 Title Hot Rod Power Distribution (3 Inch, 30%SGTP)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 3.3-1 Illustration of Overtemperature and Overpower hT Protection, Nominal Tavg=576.3'F, Nominal Pressure=2100 psia 3.3-2 Illustration of Overtemperature and Overpower AT Protection, Nominal Tavg=576.3'F, Nominal Pressure 2250 psia 3.3-3 Illustration of Overtemperature and Overpower dT Protection, Nominal Tavg=553.0'F, Nominal Pressure=2250 psia 3.3-4 3.3-53.3-6 Illustration of Overtemperature and Overpower bT Protection, Nominal Tavg=553.0'F, Nominal Pressure=2100 psia I Nuclear Power and Hot Channel Heat Flux vs.Time for the Rod Withdrawal From Subcritical Event Fuel Average and Clad Temperature vs.Time for the Rod Withdrawal from Subcritical Event 3.3-7 Nuclear Power vs.Time for the RCCA Withdrawal at Power Event, Full Power, 80 PCM/sec Insertion Rate, Maximum Reactivity Feedback 3.3-8 Pressurizer Pressure and Pressurizer Water Volume vs.Time for the RCCA Withdrawal at Power Event, Full Power, 80 PCM/sec Insertion Rate, Maximum Reactivity Feedback 3.3-9 Core Average Temperature and DNBR vs.Time for the RCCA Withdrawal at Power Event, Full Power, 80 PCM/sec Insertion Rate, Maximum Reactivity Feedback 3.3-10 Nuclear Power vs.Time for the RCCA Withdrawal at Power Event, Full Power, 4 PCM/sec Insertion Rate, Maximum Reactivity Feedback mal 944-1w.wpf:1d~1195 XIII LIST OF FIGURES (continued)
~Fi rrre Title 3.3-11 Pressurizer Pressure and Pressurizer Water Volume vs.Time for the RCCA Withdrawal at Power Event, Full Power, 4 PCM/sec Insertion Rate, Maximum Reactivity Feedback 3.3-1 2 Core Average Temperature and DNBR vs.Time for the RCCA Withdrawal at Power Event, Full Power, 4 PCM/sec Insertion.
Rate, Maximum Reactivity Feedback 3.3-1 3 Minimum DNBR vs Reactivity Insertion Rate for the RCCA Withdrawal at Power Event, 100%Power 3.3-14 Minimum DNBR vs.Reactivity Insertion Rate for the RCCA Withdrawal at Power Event, 60%Power 3.3-1 5 Minimum DNBR vs.Reactivity Insertion Rate for the RCCA Withdrawal at Power Event, 10%Power 3.3-16 3.3-17 Nuclear Power and Core Heat Flux vs.Time for a Typical Response to a Dropped RCCA(s)in Automatic Control Average Coolant Temperature and Pressurizer Pressure vs.Time for a Typical Response to a Dropped RCCA(s)in Automatic Control 3.3-18 Total Core Flow vs.Time for the Complete Loss of Flow Event 3.3-19 Nuclear Power and Pressurizer Pressure vs.Time for the Complete Loss of Flow Event 3.3-20 Average and Hot Channel Heat Fluxes and DNBR vs.Time for the Complete Loss of Flow Event 3.3-21 Total Core Flow and Faulted Loop Flow vs.Time for the Partial Loss of Flow Event m."I1944-1w.wpf:1d~1195 XIV LIST OF FIGURES (continued)
~Fi ere 3.3-22 Title Nuclear Power and Pressurizer Pressure vs.Time for the Partial Loss of Flow Event 3.3-23 Average and Hot Channel Heat Fluxes and DNBR vs.Time for the Partial Loss of Flow Event 3.3-24 3.3-25 3.3-26 Total Core Flow and Faulted Loop Flow vs.Time for the Locked Rotor Event Nuclear Power and RCS Pressure vs.Time for the Locked Rotor Event Average and Hot Channel Heat Fluxes vs.Time and Clad Inner Temperature vs.Time for the Locked Rotor Event 3.3-27 Nuclear Power and DNBR vs.Time for Loss of Load, Minimum Reactivity Feedback with Pressurizer Spray and PORVs 3.3-28 Pressurizer Pressure and Pressurizer Water Volume vs.Time for Loss of Load, Minimum Reactivity Feedback with Pressurizer Spray and PORVs 3.3-29 Core Average and Loop 1 Temperature vs.Time for Loss of Load, Minimum Reactivity Feedback with Pressurizer Spray and PORVs 3.3-30 Total Reactivity and Pressurizer Steam Relief vs.Time for Loss of Load, Minimum Reactivity Feedback with Pressurizer Spray and PORVs 3.3-31 Steam Generator Mass and Safety Valve Relief vs.Time for Loss of Load, Minimum Reactivity Feedback with Pressurizer Spray and PORVs 3.3-32 Nuclear Power and DNBR vs.Time for Loss of Load, Maximum Reactivity Feedback with Pressurizer Spray and PORVs 3.3-33 Pressurizer Pressure and Pressurizer Water Volume vs.Time for Loss of Load, Maximum Reactivity Feedback with Pressurizer Spray and PORVs m:11944-1w.wpt:1d441195 LIST OF FIGURES (continued)
Ficiure 3.3-34 Title Core Average and Loop 1 Temperatures vs.Time for Loss of Load, Maximum Reactivity Feedback with Pressurizer Spray and PORVs 3.3-35 Total Reactivity and Pressurizer Steam Relief vs.Time for Loss of Load, Maximum Reactivity Feedback with Pressurizer Spray and PORVs 3.3-36 Steam Generator Mass and Safety Valve Relief vs.Time for Loss of Load, Maximum Reactivity Feedback with Pressurizer Spray and PORVs 3.3-37 Nuclear Power and DNBR vs.Time for Loss of Load, Minimum Reactivity Feedback without Pressurizer Spray and PORVs 3.3-38 Pressurizer Pressure and Pressurizer Water Volume vs.Time for Loss of Load, Minimum Reactivity Feedback without Pressurizer Spray and PORVs 3.3-3t." Core Average and Loop 1 Temperatures vs.Time for Loss of Load, Minimum Reactivity Feedback without Pressurizer Spray and PORVs 3.3-40 Total Reactivity and Pressurizer Steam Relief vs.Time for Loss of Load, Minimum Reactivity Feedback without Pressurizer Spray and PORVs 3.3-41 Steam Generator Mass and Safety Valve Relief vs.Time for Loss of Load, Minimum Reactivity Feedback without Pressurizer Spray and PORVs 3.3-42 Nuclear Power and DNBR vs.Time for Loss of Load, Maximum Reactivity Feedback without Pressurizer Spray and PORVs 3.3-43 Pressurizer Pressure and Pressurizer Water Volume vs.Time for Loss of Load, Maximum Reactivity Feedback without Pressurizer Spray and PORVs 3.3-44 Core Average and Loop 1 Temperatures vs.Time for Loss of Load, Maximum Reactivity Feedback without Pressurizer Spray and PORVs m 31944-1w.wpf:1d~1195 xvl
LIST OF FIGURES (continued)
~Fi ure 3.3-45 Title Total Reactivity and Pressurizer Steam Relief vs.Time for Loss of Load, Maximum Reactivity Feedback without Pressurizer Spray and PORVs 3.3-46 Steam Generator Mass and Safety Valve Relief vs.Time for Loss of Load, Maximum Reactivity Feedback without Pressurizer Spray and PORVs 3.3-47 Variation of Reactivity with Core Temperature at 1050 psia for the End of Life Rodded Core with One Control Rod Assembly Stuck (Zero Power)for the Steamline Break Double Ended Rupture Event 3.3-48 Doppler Power Feedback for the Steamline Break Double Ended Rupture Event 3.3-49 Safety Injection Flow Supplied by One Charging Pump for the Steamline Break Double Ended Rupture Event 3.3-50 Nuclear Power and Core Heat Flux vs.Time for the Steamline Break Double Ended Rupture Event (Inside Containment with Power)3.3-51 Core Average Temperature and RCS Pressure vs.Time for the Steam!inc Break Double Ended Rupture Event (Inside Containment with Power)3.3-52 Pressurizer Water Volume vs.Time for the Steamline Break Double Ended Rupture Event (Inside Containment with Power)3.3-53 Reactivity and Core Boron Concentration vs.Time for the Steamline Break Double Ended Rupture Event (Inside Containment With Power)3.3-54 Nuclear Power vs.Time for the Rod Ejection Event, Hot Zero Power, End of Life 3.3-55 Fuel Centerline, Fuel Average and Clad Outer Surface Temperature vs.Time for the Rod Ejection Event, Hot Zero Power, End of Life 3.3-56 Nuclear Power vs.Time for the Rod Ejection Event, Hot Full Power, End of Life m&1944-1 w.wpf:1d 441 195 XVII LIST OF FIGURES (continued)
~Fi ure 3.3-57 Title Fuel Centerline, Fuel Average, and Clad Outer Surface Temperature vs.Time for the Rod Ejection Event, Hot Full Power, End of Life ,3.5-1 LOCA Mass and Energy Release Containment Integrity, Containment Pressure Transient 3.5-2 LOCA Mass and Energy Release Containment Integrity, Upper Compartment Temperature Transient 3.5-3 LOCA Mass and Energy Release Containment Integrity, Lower Compartment Temperature Transient 3.5-4 LOCA Mass and Energy Release Containment Integrity, Active and Inactive Sump Temperature Transient 3.5-5 LOCA Mass and Energy Release Containment Integrity, Ice Melt Transient 3.5-6 1.4 ft'ouble-Ended Rupture, 102%Power, MSIV Failure, Upper Compartment Temperature 3.5-7 1.4 ft'ouble-Ended Rupture, 102%Power, MSIV Failure, Lower Compartment Temperature 3.5-8 1.4 ft'ouble-Ended Rupture, 102%Power, MSIV Failure, Upper Compartment Pressure 3.5-9 1.4 ft'ouble-Ended Rupture, 102%Power, MSIV Failure, Lower Compartment Pressure 3.5-10 0.942 ft'plit Break, 30%Power, MSIV Failure, Upper Compartment Temperature m 31944.1w.wpf:1dM1195 XVIII LIST OF FIGURES (continued)
~Fi tjre Title 3.5-11 0.942 ft'plit Break, 30%Power, MSIV Failure, Lower Compartment Temperature 3.5-12 3.5-13 0.942 ft'plit Break, 30%Power, MSIV Failure, Upper Compartment Pressure 0.942 ft'plit Break, 30%Power, MSIV Failure, Lower Compartment Pressure m:$1944.1w.wpf:1d441195 XIX LIST OF ACRONYMS AND ABBREVIATIONS AFWPR ANS ASME APC BIT BOP CCWS CHG/Sl COLR CRDM CS CVCS DBE DECL DEHL DEPS DF DNB DNBR EAB ECC ECCS ECT EDG EFPM EOP ESF ESFAS ESW Fd,H F~FHA FSAR GPM HELB HFP HZP IFBA IFM ITDP LB Auxiliary Feedwater Pump Runout American Nuclear Society American Society of Mechanical Engineers Alternate Plugging Cnteria Boron Injection Tank Balance of Plant Component Cooling Water System Charging/Safety Injection Core Operating Limits Report Control Rod Drive Mechanism Condensate System Chemical and Volume Control System Design Basis Earthquake Double-Ended Cold Leg Double-Ended Hot Leg Double-Ended Pump Suction Decontamination Factor Departure from Nucleate Boiling Departure from Nucleate Boiling Ratio Exclusion Area Boundary Emergency Core Cooling Emergency Core Cooling System Eddy Current Testing Emergency Diesel Generator Effective Full Power Months Emergency Operating Procedure Engineered Safety Features Engineered Safety Feature Actuation System Essential Service Water Hot Channel Enthalpy Rise Factor Total Peaking Factor Fuel Handling Accident Final Safety Analysis Report Gallons per Minute High Energy Line Break Hot Full Power Hot Zero Power Integral Fuel Burnable Absorbers Intermediate Flow Mixing Improved Thermal Design Procedure Large Break m%1944-1w.wpf:1d441295 LIST OF ACRONYMS AND ABBREVIATIONS (continued)
LCO LOCA LO~LOOP LPZ M/E or M8E MMF MSLB Mwt NRC NSSS OPBT OTAT PCT PLOF PORV PTS PSSM PWR RC RCCA RCL RCP RCPB RCS RHR RHRS RPS RSE RSR RTDP RTP RTS RWST RWFS SAL SDM SER SB SFPCS SI Limiting Condition for Operation Loss of Coolant Accident Loss of Load/Turbine Trip Loss of All AC Power to the Station Auxiliaries Low Population Zone Mass and Energy Minimum Measured Flow Main Steam Line Break Megawatt Thermal Nuclear Regulatory Commission Nuclear Steam Supply System Overpower Delta T Overtemperature Delta T Peak Clad Temperature Partial Loss of Reactor Coolant Flow Power Operated Relief Valve Pressurized Thermal Shock Power Shape Sensitivity Model Pressurized Water Reactor Reactor Coolant Rod Cluster Control Assembly Reactor Coolant Loop Reactor Coolant Pump Reactor Coolant Pressure Boundang Reactor Coolant System Residual Heat Removal Residual Heat Removal System Reactor Protection System Reload Safety Evaluation Relative Stability Ratio Revised Thermal Design Procedure Rated Thermal Power Reactor Trip System Refueling Water Storage Tank RCCA Bank Withdrawal from a Subcritical Condition Safety Analysis Limit Shutdown Margin Safety Evaluation Report Small Break Spent Fuel Pool Cooling System Safety Injection m%1944-1w.wpf:
1 d441295 XXI sos SG SGTP SGTR SLB SLB-CR SR TA TAUG THQT TCOLO TDF LIST OF ACRONYMS AND ABBREVIATIONS (continued)
Safety Injection System Steam'enerator Steam Generator Tube Plugging Steam Generator Tube Rupture Steam Line Break Steam Line Break Core Response Surveillance Requirement Total Allowance RCS Average Temperature Vessel Outlet Temperature Vessel Inlet Temperature Thermal Design Flow m.11944-1w.wpf:1d~1 295 XXII DEFINITIONS Rerating Program: WCAP-11902 documented the analyses and evaluations performed to support reduced temperature and pressure operation of Donald C.Cook Nuclear Plant Units 1 and 2.Subsequently, a supplement to WCAP-11902 was issued to summarize the additional efforts performed to support a rerating of Cook Nuclear Plant Unit 1 and to provide part of the support for a Unit 2 rerating.These analyses and evaluations are described in Sections 2.0 and are documented in References 1 and 2 of Section 2.0.Throughout this report, the analyses and evaluations documented in WCAP-11902 and Supplement are referred to as the Rerating Program.Steam Generator Tube Plugging Program: Analyses and evaluations to support operation of Cook Nuclear Plant Unit 1 with up to a level of 30%steam generator tube plugging.In addition to the increased tube plugging level (and corresponding reduced thermal design flow), several increased operating margins were also addressed in the Steam Generator Tube Plugging Program.These operating margins are described in Section 1.0.mA1 944-1w.wpt:1d441295 XXIII

SUMMARY

AND CONCLUSIONS PROGRAM

SUMMARY

The purpose of this document is to provide the safety analysis and evaluation results to support operation of Donald C.Cook Nuclear Plant Unit 1 with up to a level of 30%steam generator tube plugging (SGTP).In addition to the increased level of steam generator tube plugging, the analyses and evaluations support a corresponding reduction in Thermal Design Flow (TDF)and a 5%loop flow asymmetry.
The analyses and evaluations were performed over a range of primary temperatures (553'F and 576.3'F)and for two values of primary pressure (2100 psia and 2250 psia).The evaluations in this report are based on analyses performed for the Rerating Program.This program is discussed in more detail in Section 2.0.In addition to addressing an increased SGTP level of 30%, the following increased operating margins were also addressed:
(1)Reduction of Sl and RHR discharge pressure on recirculation
-The RHR and SI minimum safeguards pump head curves were reduced by 15%, an additional 5%reduction from the current analysis degradation of 10%.The charging pump head curve degradation is maintained at the current value of 10%.(2)The emergency diesel generator (EDG)start time was increased from 10 seconds to 30 seconds (3)To support increased dT drift, the margin between the safety analysis limits (SAL)and the nominal values of the K, and K, gains of the Donald C.Cook Nuclear Plant Unit 1 OTAT and OPbT setpoint equations were adjusted.(4)An increase in the pressurizer code safety valve (PSV)setpoint tolerance from+/-1%to+/-3%(5)Decreased shutdown margin for T, greater than 200'F.The analyses and evaluations in this document support all of these changes.Discussions of specific analyses address issues most relevant to those analyses.The operating parameters for the increase in steam generator tube plugging level and the additional operating margins listed above will be referred to throughout this report as the"Steam Generator Tube Plugging (SGTP)Program".This report provides the necessary documentation to support the Technical Specification changes associated with the Steam Generator Tube Plugging Program.The topics addressed in this report are as follows: m."i1 944-1w.wpf:1d~1295 XXIV Description of License Amendment Summary of Technical Specification Changes Basis for Evaluations/Analyses Performed Loss of Coolant Accident Analyses Post-LOCA Hydrogen Production Post-LOCA Hot Leg Recirculation Time LOCA Hydraulic Forces Non-LOCA Analyses Containment Analyses Steam Generator Tube Rupture Analyses Reactor Cavity Pressure Evaluation Radiological Analysis Primary Components Evaluations Fluid and Auxiliary Systems Evaluations Fuel Structural Evaluation Also provided in the Appendix to this report are the proposed Technical Specification changes.A brief summary of the results of each analysis and evaluation is provided below.ACCIDENT ANALYSIS CONCLUSIONS The results of the accident analyses and evaluations performed for the SGTP Program demonstrate that safe operation of the Donald C.Cook Nuclear Plant Unit 1 is maintained.
The bases for the evaluations and analyses performed are provided in Section 2.1 A summary of the conclusions of each of the accident analyses is provided below.Lar e Break I OCA (Section 3.1.1)The large break LOCA analysis was reanalyzed for the impact of the increased tube plugging level, reduced TDF, loop flow asymmetry, revised ECCS flows, and the increased EDG start time.The large break LOCA analysis was not impacted by the pressurizer code safety valve tolerance increase, the revised K1/K4 values, or the decreased shutdown margin.The large break LOCA analysis was performed with the 1981 version of the Westinghouse ECCS Evaluation Model using the BASH computer code.Analysis assumptions included ECCS flow with the RHR cross-tie valves closed, a total peaking factor of 2.15, a hot channel enthalpy rise peaking factor of 1.55, and an accumulator temperature of 100'F.A full spectrum break analysis was performed at the nominal RCS conditions (initial RCS pressure of 2250 psia and initial hot leg temperature of 609.1'F)from which the limiting break discharge coefficient was determined.
The limiting break was then reanalyzed at the reduced hot leg temperature and nominal RCS pressure of 2250 psia, and also at nominal hot leg temperature and an initial RCS pressure of 2100 psia.The above cases were all analyzed with minimum safety injection flow, which was determined to be limiting.The limiting break was determined to be C,=0.4 at the nominal hot leg temperature (T>>=609.1'F)and a pressure of 2100 psia with m:51944-iw.wpf:1d441295 minimum safety injection flow.The peak cladding temperature was calculated to be 2164'F, which is less than the 2200'F limit in 10CFR50.46.
Small Break LOCA (Section 3.1.2)The small break LOCA analysis was reanalyzed for the impact of the increased tube plugging level, reduced TDF, loop flow asymmetry, revised ECCS flows, and the increased EDG start time.The small break LOCA analysis was not impacted by the pressurizer code safety valve tolerance increase, the revised K1/K4 values, or the decreased shutdown margin.The small.break LOCA analysis was performed with the Westinghouse small break LOCA ECCS Evaluation Model using the NOTRUMP code (including the recent model changes submitted in WCAP-10054-P, Addendum 2 and WCAP-10081-NP, Addendum 2).The key analysis input assumptions included ECCS flows with the HHSI cross-tie discharge valves closed, a total peaking factor of 2.32 and hot channel enthalpy rise peaking factor of 1.55.Other analysis input assumptions incorporated in the small break LOCA analysis are reduced hot assembly average power (P)and a power shape based on a reduced axial offset of+20%.A single break size analysis was performed at the previously-limiting break size of three inches.The calculation used the reduced temperature, reduced pressure operating condition.
An evaluation of the break spectrum and the range of operating conditions concluded that the analyzed case would remain bounding with respect to peak clad temperature.
The calculation was performed with minimum safety injection flow, which was limiting.The peak cladding temperature was calculated to be 1443'F, which is less than the 2200'F limit in 10CFR50.46.
LOCA H draulic Forcin Functions (Section 3.2)LOCA hydraulic forces are relatively insensitive to specific SGTP levels.The Donald C.Cook Nuclear Plant LOCA hydraulic forces were most recently analyzed for the Rerating Program.The RCS parameters used in the existing analysis-of-record conservatively bound the conditions at 30%tube plugging.Therefore, the existing LOCA forces analyses remain conservative relative to the SGTP Program.Non-LOCA Anal ses (Section 3.3)The non-LOCA events were addressed by a combination of evaluations and analyses for the impact of the increased tube plugging level, reduced TDF, loop flow asymmetry, revised ECCS flows, pressurizer code safety valve tolerance increase, increased EDG start time, revised K1/K4 values, and decreased shutdown margin.The computer codes and methods used for the non-LOCA analyses have been previously approved by the NRC.The non-LOCA safety analyses were reviewed on the basis of both DNB and non-DNB acceptance criteria.All DNB event reanalyses were found to yield a minimum DNBR which remains above the limit value.The analyses demonstrate that all licensing basis criteria continue to be met and the conclusions presented in the UFSAR remain valid.mh1944 1 w.wpf: 1 d~1295 xxvl Post-LOCA H dro en Generation (Section 3.4)The post-LOCA hydrogen generation rates that were reviewed as part of the Rerating Program were determined to remain applicable to the SGTP Program.~Ci'S The containment integrity analyses were addressed for the impact of the increased level of tube plugging, reduced thermal design flow, loop flow asymmetry, revised ECCS flows, and the increased EDG start time.The containment analyses were not impacted by the pressurizer code safety valve tolerance increase, the revised K1/K4 values, or the increased shutdown margin.The increase in the containment pressure and temperature following a LOCA was analyzed.The mass and energy release rates calculated as part of the SGTP Program formed the basis to evaluate the structural integrity of the containment following a postulated accident to satisfy the acceptance criteria, General Design Criterion 38.Even though Cook Nuclear Plant is licensed to GDC's in Appendix H of the original FSAR, more conservative acceptance criteria were used.The containment integrity analysis for the most limiting case (i.e., RHR cross tie valve closed)resulted in a maximum calculated containment pressure of 11.49 psig, for the double-ended pump suction minimum safeguards break case.Since the calculated pressure is below the design pressure of 12.0 psig, the results of the LOCA containment integrity analysis are acceptable.
The Mainsteam Line Break (MSLB)mass and energy releases were used as input into the containment integrity analysis to demonstrate that the peak containment temperature resulting from a design basis MSLB will not exceed the equipment qualification criterion for the plant.The containment pressure response determined for the LOCA containment integrity analysis is calculated to be more severe than for the MSLB, and therefore, bounds the MSLB analysis.For the large break case, the limiting case among the double-ended ruptures is the 1.4 ft'ouble-ended rupture, 102%power, MSIV failure case.This case yielded a calculated peak temperature of 322.7'F.For the small break case, the most limiting case in terms of peak calculated temperature is the 0.942 ft'plit break, 30%power with an MSIV failure.This case resulted in a calculated peak temperature of 326'F.Both cases are within the Environmental Acceptance Criteria.Therefore, the analysis demonstrates that the containment heat removal systems function to rapidly reduce the containment pressure and temperature in the event of a MSLB.General Design Criterion 50 and Appendix K are satisfied.
Short Term Containment Anal sis (Section 3.5.1)The short term containment analysis that was performed for the Rerating Program was reviewed and it was determined that the conclusions provided for the Rerating Program remain valid for the SGTP Program.That is, the resulting peak pressures remain below the allowable design peak pressures for the pressurizer enclosure, the fan accumulator room and the steam generator enclosure.
mA1 944-1w.wpf:1d441295 xxvil Steam Generator Tube Ru ture (Section 3.6)The SGTR event was analyzed for the impact of the increased tube plugging level and associated reduced TDF and loop flow asymmetry.
The SGTR analysis was not impacted by any of the SGTP Program increased operating margins.The thyroid and whole body doses estimated for Cook Nuclear Plant Unit 1, based on the 30%SGTP evaluation, remain within a"small fraction" (10%)of the 10CFR100 exposure limit guidelines.
Small fraction is the smallest of the exposure guidelines defined in NUREG-0800.
Therefore, the conclusions of the UFSAR remain valid.Post-LOCA Hot Le Recirculation Time (Section 3.7)The hot leg switchover to preclude boron precipitation and post-LOCA long term cooling are n'ot adversely affected by the 30%SGTP Program.The proposed changes do not significantly affect the normal plant operating parameters, the safeguards systems actuations, the accident mitigation capabilities important to these events, or the assumptions used in the analysis of these events.The proposed changes do not create conditions more limiting than those assumed in the LOCA-related analyses.Reactor Cavi Pressure Anal sis (Section 3.8)The Reactor Cavity Pressure Analysis that was performed for the Rerating Program was reviewed and it was determined that the conclusions provided for the Rerating Program remain valid for the SGTP Program.The SGTP Program parameters affect the Reactor Cavity Pressure Analysis through the mass and energy releases provided as input into the analysis.There is no direct impact of SGTP level on short-term mass and energy release rate calculations and containment subcompartment response analysis.The mass and energy releases used as input for the Reactor Cavity Pressure Analysis reflected limiting conditions and therefore, the NSSS performance parameters for the SGTP Program did not impact the results.Radiolo ical Doses (Section 3.9)A reanalysis of the offsite doses following a large break LOCA was performed for the increase in emergency diesel generator start time to 30 seconds.While there was a slight increase in the offsite thyroid doses, the doses are within the applicable limits.The source terms for LOCA and the fuel handling accident are unaffected by the increase in SGTP level or any of the other SGTP Program adders.mA1944-1 w.wpf:1d441295 XXVIII
FLUID AND AUXILIARY SYSTEMS EVALUATION CONCLUSIONS (Section 3.10)The fluid systems proof of design calculations were reviewed for the SGTP conditions.
This review demonstrated'that the NSSS fluid systems will continue to function adequately as designed for all conditions of the SGTP Program.ECCS flowrates were revised as part of the SGTP Program and were used in the safety analyses and evaluations.
In the NSSS/BOP interface area, the proposed NSSS Performance Parameters for the SGTP Program were compared with those of the Rerating Program.The results of the evaluation show that a SGTP level of 30%will have no adverse effects on the Balance Of Plant (BOP)systems performance (Main Steam System, Condensate and Feedwater System, Auxiliary Feedwater System, Steam Generator and Blowdown System).They will continue to perform acceptably at the conditions associated with 30%SGTP.The evaluations for the fluid and auxiliary systems are described in more detail in Section 3.10.PRIMARY COMPONENTS EVALUATION CONCLUSIONS Steam Generators (Section 3.11.1)In the thermal-hydraulic areas: the modified moisture separator packages on Cook Nuclear Plant Unit 1 will permit operation at steam pressures down to 700 psia and below without exceeding 0.25%moisture carryover.
Steam Generator operating characteristics will be acceptable down to the minimum steam pressure of 589 psia.The evaluation of thermal-hydraulic stability indicates satisfactory results for all SGTP cases.The evaluation performed for the effects of the SGTP program on U-bend tube fatigue for Cook Unit 1 is documented in WCAP-13814.
It was concluded in WCAP-13814 that four tubes were susceptible to high cycle fatigue at the 30%SGTP Program conditions and would require preventative action.Structural analyses and evaluations performed for the Cook Unit 1 steam generators indicate that the steam generator components remain in compliance with the applicable ASME Code requirements under the SGTP conditions.
Reactor Vessel (Section 3.11.2)The results of the structural.
evaluations performed for the reactor vessel demonstrate that operation of Cook Nuclear Plant Unit 1 within the parameters of the SGTP program does not result in stress intensities or fatigue usage factors which exceed the acceptance criteria of the applicable ASME Code versions.The SGTP Program does not result in an increase in the fast neutron fluence values calculated for the Rerating Program.Therefore, the reactor vessel rn 31944-1w.wpf:1d441295 xxlx
integrity analyses performed as part of the Rerating Program will remain applicable after 30%SGTP.Reactor Internals (Section 3.11.3).Results of the thermal-hydraulic analyses performed for the reactor internals indicate that the SGTP Program for Cook Nuclear Plant Unit 1 results in acceptable values of core bypass flow, pressure drops, component lift forces, and momentum flux values.It was also confirmed that the control rod drop time limit of 2.4 seconds remains applicable for the SGTP Program conditions.
From the component stress analysis and the flow induced vibration evaluations, it is concluded that the margins of safety are within acceptable limits per the original design basis.Control Rod Drive Mechanisms (Section 3.11.4)The conclusion of structural evaluations performed for the 30%SGTP conditions for the CRDMs demonstrate that the operability, service life, and structural integrity of the CRDM latch assembly, drive rod, and coil stack will not be adversely affected.Reactor Coolant Pum s (Section 3.11.5)The review performed of the reactor coolant pumps for the 30%SGTP conditions demonstrate that the conditions are acceptable for the 93A RCP, and no additional thermal or structural analyses are required to demonstrate compliance with the applicable codes and standards.
The RCP motor evaluation revealed that the motors are acceptable for operation at the 30%SGTP conditions.
Pressurizer (Section 3.11.6)A fatigue analysis performed for the Cook Unit 1 pressurizer, incorporating the most conservative conditions of the SGTP program, demonstrated that the pressurizer remains in compliance with the applicable ASME Code criteria.Reactor Coolant Pi in and Su rts (Section 3.11.7)An evaluation was performed to determine the effects of the 30%SGTP conditions on the primary loop piping, primary equipment supports, and the primary equipment nozzles.Operation at the SGTP Program conditions was found to be acceptable because these conditions are already enveloped by the Rerating Program.In addition, the rerating transients and plant parameters associated with the Rerating and SGTP Programs for Donald C.Cook Nuclear Plant Unit 1 have been reviewed.The impact on the design basis analysis for the mh1 944.1w.wpf:1d441295 NRC Bulletin 88-08 evaluation of the auxiliary spray piping and the NRC Bulletin 88-11 evaluation of the pressurizer surge line piping is insignificant.
Auxilia Com onents (Section 3.1 f.8)Evaluations were performed for the auxiliary tanks, pumps, valves, and heat exchangers to determine the effects of the revised RCS parameters due to the SGTP Program.The results of these evaluations demonstrated that, due to conservatively specified parameters used in the procurement of the auxiliary equipment, the 30%SGTP parameters are not expected to adversely affect the function or structural integrity of this equipment, Fuel Structural Evaluation (Section 3.12)Evaluations were performed of the fuel for Cook Nuclear Plant Unit 1 for the SGTP Program conditions in the areas of fuel rod and fuel assembly structural integrity, core design, and thermal-hydraulic design.The fuel assembly structural integrity is not affected by the SGTP Program, and the core eoolable geometry is maintained for the 15x15 OFA fuel in the Unit 1 core.The evaluation of the fuel rod structural integrity indicates these conditions will be acceptable, although it is noted that cycle-specific verification during the normal reload will still be performed.
The results of the core design evaluation indicated that the SGTP Program conditions result in no impact to the core design except for the values of the statepoint for the Steamline Break Analysis, and the Dropped Rod Analysis.Thermal-hydraulic analyses were made for the fuel for the limiting 30%SGTP parameters using RTDP methodology.
The analysis showed that the DNBR design basis was met for the limiting DNB events.This analysis caused the available DNB margin to increase.This margin can be used for flexibility of design and to offset unanticipated DNBR penalties.
m:51944-1 w.wpf:1d441295 xxxl

2.1 DESIGN

POWER CAPABILITY PARAMETERS This section describes the parameters which were used as the basis for the evaluations and analyses performed to support the$GTP Program for Cook Nuclear Plant Unit 1.The NSSS performance parameters feature the current licensed NSSS power of 3262 MWt, a T,~temperature range from 553'F to 576.3'F, two primary pressure values of 2250 psia or 2100 psia, a maximum average and peak SGTP level of 30%, reduced TDF, and 5%loop flow asymmetry.
The RCS temperature range is bounded by the Rerating Program.Also incorporated into the SGTP Program was'an RCS flow measurement uncertainty range of 1.9%to 2.5%.The Technical Specifications will be revised to incorporate the Minimum Measured Flow (MMF)corresponding to a flow measurement uncertainty of 2.5%.This increase will provide additional margin for instrumentation.
If additional flow margin is needed in the future.the margin can be reallocated from instrumentation margin to RCS flow margin by revising the MMF Technical Specification because all of the accident analyses evaluated a MMF of 339,100 gpm.A MMF of 339,100 gpm total reflects a 1.9%flow measurement uncertainty.
A MMF of 341,100 gpm total reflects a 2.5%flow measurement uncertainty.
The RCS flow margin has been reviewed and sufficient margin exists to maintain the TDF of 83,200 gpm/loop with the 2.5%flow measurement uncertainty.
A brief description of each set of parameters is provided below: Case 1: These are the original NSSS performance parameters for Unit 1 and are shown for comparison with the revised parameters.
The NSSS power level of 3250 MWt does not account for reactor coolant pump heat;at the time that Unit 1 was designed, it was the custom to indicate only the core power level value.Case 2: These parameters incorporate a core power level of 3250 MWt, an NSSS power level of 3262 MWt (which includes 12 MWt of reactor coolant pump heat), an average steam generator tube plugging level of 30%, primary pressures of either 2250 psia or 2100 psia, and a lower bound vessel average temperature of 553.0'F.Case 3: These parameters incorporate the same features as case 2, except that the primary temperatures and resulting secondary parameters incorporate an upper bound vessel average temperature of 576.3'F.This case was used as the basis for selected analyses, where high pnmaiy temperatures were limiting.Case 4: These parameters incorporate the same features as case 2, except that the TDF was reduced to 79,000 gpm/loop to bound 5%loop flow asymmetry.
Case 5: These parameters incorporate the same features as case 4, except that the primary temperatures and resulting secondary parameters incorporate an upper bound vessel average temperature of 576.3'F (the highest vessel average temperature considered for the SGTP'31944-1 w.wpf: 1 d~1195 2.1-1
,

1.0 INTRODUCTION

-DESCRIPTION OF LICENSE AMENDMENT REQUEST 1.1 PURPOSE FOR CHANGE The Donald C.Cook Nuclear Plant Unit 1 has experienced tube corrosion problems in its steam generators and, as a result, an increasing number of tubes have been plugged during the last several outages.Steam generator tube plugging potentially decreases reactor coolant system flow due to increased flow resistances through the steam generators.
As the number of plugged tubes increases, the RCS flow may be reduced to a value below that which is currently analyzed in the licensing basis.Currently, the licensing basis analyses for the Donald C.Cook Nuclear Plant Unit 1 are documented in the Updated Final Safety Analysis Report{UFSAR).These analyses are bounding for a maximum average steam generator tube plugging (SGTP)level of up to 10%with a peak level of 15%in any one steam generator.
This amendment request reflects the changes to the safety analysis assumptions and results due to the revised operating conditions resulting from an increased level of steam generator tube plugging.While the analyses and evaluations were being performed for the increased level of tube plugging, several operating margins were increased and incorporated into the analyses and evaluations in order to maximize the benefit of the reanalysis.
Therefore, in addition to addressing an increased SGTP level of 30%, the following increased operating margins were also addressed:
(1)Reduction of Sl and RHR discharge pressure on recirculation
-The RHR and Sl minimum safeguards pump head curves were reduced by 15%, an additional 5%reduction from the current analysis degradation of 10%.The charging pump head curve degradation is maintained at the current value of 10%.(2)The emergency diesel generator startup time was increased from 10 seconds to 30 seconds (3)To support increased hT drift, the margin between the safety analysis limits{SAL)and the nominal values of the K, and K, gains of the Donald C.Cook Nuclear Plant Unit 1 OTZT and OPbT setpoint equations were adjusted.{4)An increase in the pressurizer code safety valve (PSV)setpoint tolerance from+/-1%to+/-3%(5)Decreased shutdown margin for T,~greater than 200'F.The revised parameters associated with the, increase in tube plugging level to 30%SGTP, both directly and indirectly, are referred to throughout this report as the"Steam Generator m%1944.1w.wpf:1d~1295 Tube Plugging Program".This program resulted in changes to the Donald C.Cook Nuclear Plant Unit 1 Technical Specifications, including:
Core Safety Limits OPb,T/OTbT Setpoints Shutdown Margin for Modes 1,2,3, and 4 DNB Parameters RTS Response Times ESFAS Instrumentation Logic Pressurizer Code Safety Valve Lift Setting Tolerance ECCS Pump Discharge Pressure on Recirculation Minimum RWST Temperature Containment Internal Pressure RCS Volume Emergency Diesel Generator Start Time m:11944.1w.wpt:
1d441295 1.1-2 been based on the new core safety limits and account for instrument uncertainties.
The reference temperatures are now indicated values and the temperature range that was analyzed is specified.
Shutdown Mar in for MODES 1 2 3 and 4 Shutdown margin (SDM)requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences.
As such, the SDM defines the degree of subcriticality that would be obtained immediately following the insertion or scram of all shutdown and control rods, assuming that the single rod cluster assembly of the highest reactivity worth is fully withdrawn.
The Limiting Condition for Operation (LCO)3.1.1.1 Shutdown Margin has been revised from 1.6'lo dk/k to 1.3'k Ak/k.This change is also included in the associated Bases.The acceptability of the decrease in the SDM is based on re-analysis of the most limiting accident, core response to a steam line break.DNB Parameters RCS Tav and RCS Flow LCO 3.2.5, DNB Parameters specifies RCS parameters assumed as initial conditions in the transient and accident analyses.In Table 3.2-1, DNB Parameters, the RCS T,~has been changed from<570.9'F to<(576.3+5.1)-(readability error)'F.The 576.3'F value has been verified in the re-analyses.
The additional 5.1'F is included to account for temperature uncertainty factors such as cold leg streaming, as documented in WCAP-12568, Rev.1.The readability error will be determined by AEPSC.Additionally, in Table 3.2-1, the RCS Total Flow Rate has been reduced from 2 361,600 to>341,100 gpm.This reduction is based on the increase in the SGTP limit to 30/o and includes a 2.5/o instrument uncertainty.
RTS Res onse Times Reactor Trip System (RTS)Instrumentation response times are assumed in accident analyses for the time interval from when the monitored parameter (level, pressure, temperature, etc.)exceeds its setpoint at the sensor until loss of stationary gripper coil voltage.RTS Instrumentation response times specified in Table 3.3-2 for Functions 9, 10, 14, and 16 (Pressurizer Pressure-Low, Pressurizer Pressure-High,.
Steam Generator Water Level-Low-Low, and Undervoltage-Reactor Coolant Pumps)relaxed from 1.0 second to 2.0 seconds m."i1944-1w.wpf:1d 441295 1.2-2 (Functions 9, 10, and 14)or to 1.5 seconds (Function 16).The acceptability of these relaxations has been verified by accident analyses.ESFAS Instrumentation Lo ic The Engineered Safety Feature Actuation System (ESFAS)instrumentation initiates necessary safety systems, based on the values of selected unit parameters, to protect against violating core design limits, protect the RCS pressure boundary, and to mitigate accidents.
ESFAS Instrumentation logic for Functional Units 1 and 4 and for the P-12 interlock have been revised.These revisions reflect the Unit 1 implementation of the"hybrid" steamline break protection logic that is presently used on Unit 2.Function 1-Safety Injection, Turbine Trip, Feedwater Isolation and Motor-driven Feedwater Pumps: Actuation on Steam Flow in Two Steam Lines-High coincident with either T,-Low-Low or Steam Line Pressure-Low has been replaced by actuation on Steam Line Pressure-Low Function 4-Steam Line Isolation:
Actuation on Steam Flow in Two Steam Lines-High coincident with either T,~-Low-Low or Steam Line Pressure-Low has been replaced by actuation on Steam Flow in Two Steam Lines-High coincident with T,-Low-Low and on Steam Line Pressure-Low Pressurizer Code Safe Valve Lift Settin Tolerance The pressurizer safety valves provide, in conjunction with the Reactor Protection System, overpressure protection for the RCS.Accident and safety analyses which require safety valve actuation assume operation of three pressurizer safety valves to limit increases in RCS pressure.The limit protected by this specification is the reactor coolant pressure boundary Safety Limit of 110%of design pressure.The LCO 3.4.2 (MODES 4 and 5)and LCO 3.4.3 (MODES 1, 2, and 3)pressurizer code safety valve lift setting tolerance has been increased from a 1%to a 3%.The acceptability of the increased safety valve tolerance has been established by evaluation or analysis of applicable events including loss of load, turbine trip, locked rotor, loss of normal feedwater, feedwater line break, and loss of all power to station auxiliaries.
m:51944-1w.wpf:1d 441295 1.2-3 ECCS Pum Flows LCO 3.5.2, ECCS Subsystems
-T,~t 350'F, specifies requirements applicable to pumps.heat exchangers, and flow paths credited with core cooling following an accident.The Surveillance Requirement (SR)4.5.2.f (centrifugal charging pump, safety injection pump and residual heat removal pump tests)requirements have been revised to specify measurement of differential pressure along with revised acceptance criteria pressures.
The revised centrifugal charging pump differential pressure criteria represents a 10%pump head-degradation.
The differential pressure criteria specified for the safety injection pump and the residual heat removal pump reflects relaxations in the associated pump curves which represent 15%pump head degradation.
The relaxation for centrifugal charging pumps is also applicable to LCOs 3.1.2.3 and 3.1.2.4.The acceptability of the relaxed pump curves has been verified for all applicable accidents.
Minimum RWST Tem erature The minimum Refueling Water Storage Tank (RWST)temperature has been changed from 80'F to 70'F in LCOs 3.1.2.7, 3.1.2.8, and 3.5.5 and Bases 3.5.5.The 70'F minimum temperature is acceptable based on the LOCA and non-LOCA analyses performed for the Cook Nuclear Plant Unit 1 licensing basis.I Containment Internal Pressure The maximum calculated post-accident containment pressure must remain below the containment design pressure of 12.0 psig.The results of the containment integrity analyses performed for the SGTP program resulted in a maximum calculated containment pressure of 11.49 psig.Thus, the value in the Bases for LCO 3.6.1.4 (11.89 psig), Internal Pressure, is being revised to reflect the analysis results.Emer en Diesel Generator EDG Start Time LCO 3.8.1, AC Sources-Operating, specifies requirements for off-site and on-site (diesel generator)
AC sources, including EDG testing requirements to demonstrate the capability to achieve the required voltage and frequency within the specified time.The EDG start time, plus the load sequencer loading times, plus the equipment actuation/start times, establishes the total time until the function of ESF equipment is assumed in accident analyses.The SR 4.8.1.1.2.a.4, SR 4.8.1.1.2.e.4.b, SR 4.8.1.1.2.e,6.b, and SR 4.8.1.1.2.f.3 requirements have been revised to specify a relaxed EDG start time of 30 seconds.In the safety analysis, the 30 second time is the time at which the load sequencer is assumed to m:51944-1w.wpt:1d441295 1.2<
start loading.Additionally, the SRs have been revised to specify that voltage and frequency shall be achieved rather than engine RPM, consistent with the safety analysis, Regulatory Guide 1.9, and NUREG-1431.
Relaxation of the EDG start time to 30 seconds has been shown to be acceptable based on re-analysis of limiting accidents.
Only the response times specified in Table 3.3-5, Engineered Safety Features Response Times which include the diesel generator start time are affected by this change.The longer, response times assumed in LOCA, Non-LOCA, and containment analyses have been specified in Table 3.3-5.RCS Volume Design Features Section 5.4.2 specifies the total contained volume of the RCS.With the increase of the SGTP limit to 30%, a corresponding reduction in RCS volume must be specified.
Since the actual level of tube plugging may change each outage, a range of RCS volume corresponding to the range of 0%to 30%tube plugging has been specified:
approximately 12,466 ft'o 11,551 ft'.m:41944.1w.wpf:
1 d441295'.2-5 TABLE 1.2-1

SUMMARY

OF TECHNICAL SPECIFICATION CHANGES PAGE SECTION DESCRIPTION OF CHANGE BASIS 2-2 B 2-1{a)2.1 Bases 2.1 Core Safety Limits Replace Figure 2.1-1, Reactor Core Safety Limits The safety analysis limit DNBR specified in the Bases for Section 2.1 has been revised from 1.45 to 1.42.Figure 2.1-1 has been replaced with a new figure based on the latest analyses, reflecting 30%SGTP, reduced rated thermal power, reduced RCS flow, etc.Section 3.3.2.1 2-7 2-8 2-9 B 2-4 B 2-5 2.2 B 2.2.1 OTbT&OPbT Setpoints Revise OTAT and OPET Trip Setpoint and Allowable Value notes.Proposed settings based on new core safety limits and account for instrument uncertainties.
Section 3.3.2.1 8 Table 3.3-3 3/4 1-1 B 3/4 1-1 3.1.1.1 Shutdown Margin Relaxation based on re-Shutdown Margin limit relaxed analysis of limiting accident from 1.6 to 1.3%hk/k-core response to steam line break.Sections 3.3.4.7 8 3.3.5.6 3/4 2-14 3.2.5 Table 3.2-1 DNB Parameters Revise Table 3.2-1, DNB Parameters, RCS T, from 570.9'F to{581.4-readability error)'F Readability error is responsibility of AEPSC.T, input assumption verified by reanalyses.
Sections 2.1, 3.3.2.1 8 3.12.4 m'81944.1w.wpt:1d441295 1.2-6
TABLE 1.2-1 (continued)

SUMMARY

OF TECHNICAL SPECIFICATION CHANGES PAGE SECTION DESCRIPTION OF CHANGE BASIS 2-5 3/4 2-14 2.2 Table 2.2-1 3.2.5 Table 3.2-1 DNB Parameters Revise Table 3.2-1, DNB Parameters, RCS Total Flow from>361,600 to a 341,100 gpm.Change based on 30%SGTP limit.The value of 341,100 includes a 2.5%instrument uncertainty.
Analysis used 339,100 which includes a 1.9%instrument uncertainty.
Sections 2.1, 3.3.2.1 8 3.12.4 3/4 3-10 3/4 3-11 3.3.1 Table 3.3-2 RTS Response Times The acceptability of these RTS Instrumentation response relaxations verified by times for Functions 9, 10, 14, accident analyses.and 16 (Pressunzer Pressure-Low, Pressurizer Pressure-Sections 3.1 8 3.3 High, Steam Generator Water Level-Low-Low, and Undervoltage-Reactor Coolant Pumps)relaxed from 1.0 second to 2.0 seconds (Functions 9, 10, and 14)or to 1.5 seconds (Function 16)~3/4 3-17 3/4 3-21 3/4 3-23a 3/4 3-24 3/4 3-26 3/4 3-28 3/4 3-31 3/4 3-33 Table 3.3-3 Table 3.3-4 Table 3.3-5 Table 4.3-2 ESFAS LOGIC ESFAS Instrumentation logic for Functional Units 1 and 4 and for the P-12 interlock have been revised.ESFAS logic change has been shown to be acceptable by non-LOCA analyses.Section 3.3 m:11944-1w.wpf:
1 d 441295 1.2-7 TABLE 1.2-1 (continued)

SUMMARY

OF TECHNICAL SPECIFICATION CHANGES PAGE SECTION DESCRIPTION OF CHANGE BASIS B 3/4 4-1 3/4 4.1 DNBR Limit The DNBR limit specified in Bases 3/4 4.1 is no longer applicable.
"1.69" replaced with"the safety analysis limit" No basis required.3/4 4-4 3/4 4-5 3.4.2 3.4.3 Safety Valve Lift Setting Pressurizer code safety valve lift setting pressure tolerance increased to 3%Relaxation based on evaluation or analysis of several limiting events.Sections 1.18 3.3 3/4 1-1 1 3.1.2.3 3/4 1-12 3.1.2,4 3/4 5-5 3.5.2 ECCS Pump Flows Centrifugal charging, safety injection and residual heat removal pumps'est acceptance criteria relaxed.CCP-10%RHR 8 Sl-15%Relaxed pump curves have been verified to be acceptable for all applicable accidents.
Section 3.3.4.7&3.10 3/4 1-15 3.1.2.7 3/4 1-16 3.1.2.8 3/4 5-11 3.5.5 B 3/4 5-3 B 3.5.5 RWST Temperature Minimum RWST temperature reduced from 80'F to 70'F The 70'F minimum temperature is acceptable based on the LOCA and non-LOCA analyses.Section 3.1.1, W-letter 92AE-G-074, dated June 12, 1992 mA1 944-1 w.wpf:1d~1295 1.2-8 TABLE 1.2-1 (continued)

SUMMARY

OF TECHNICAL SPECIFICATION CHANGES PAGE B 3/4 6-2 3.6.1.4 Internal Pressure Thus, the value in the Bases for LCO 3.6.1.4 (11.89 psig), Internal Pressure, is being revised to reflect the analysis results.SECTION DESCRIPTION OF CHANGE BASIS The results of the containment integrity analyses performed for the SGTP program resulted in a maximum calculated containment pressure of 11.49 psig.Section 3.5 3/4 8-3 3/4 8-5 3/4 8-7 3/4 3-27 3/4 3-28 3/4 3-29 3/4 3-30 3.8.1.1 3.3.2 Table 3,3-5 EDG Start Time/ESFAS Response Times EDG start time relaxed to 30 seconds.ESFAS response times affected by EDG start time relaxation revised.The increase of 20 seconds in the EDG start time has been shown to be acceptable for limiting accidents.
Sections 3.3 8 3.5 5.4.2 RCS Volume Change based on 30%Design Feature, RCS volume SGTP limit.reduced.Section 3.10 m:51944-1w.wpf:1d
~1295 1.2-9

1.2 CURRENT

LICENSE BASIS AND FUNCTION OF IDENTIFIED TECHNICAL SPECIFICATIONS AND DESCRIPTION OF PROPOSED CHANGE The proposed changes to the Donald C.Cook Nuclear Plant Unit 1 Technical Specifications are summarized in Table 1.2-1.These changes reflect the impact on the design, analytical methodology, and safety analysis assumptions outlined in this amendment request.The proposed Technical Specification changes are included in Appendix A of this report.A brief overview of the significant Technical Specification changes follows.The changes are based on analyses and evaluations associated with the SGTP Program.Since new analyses and evaluations were required to establish the acceptability of the SGTP level, several related Technical Specification relaxations were veriTied.Core Safe Limits Technical Specification Figure 2.1-1, Reactor Core Safety Limits, shows the loci of points of thermal power, RCS pressure and average temperature below which the calculated DNBR is no less than the design DNBR value and the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid.The figure is based on the enthalpy hot channel factor.Figure 2.1-1 has been replaced with a new figure based on the latest analyses, reflecting 30%tube plugging, reduced rated thermal power, reduced RCS flow, etc.OPbT/OTbT Set pints Technical Specification Table 2.2-1 lists the reactor protection system instrumentation trip setpoints for the various trip functions.
The reactor trip setpoint limits specified in Table 2.2-1 are the nominal values at which the reactor trips are set for each functional unit.The Thermal Overpower bT (OPAT)trip function provides assurance of fuel integrity (e.g., no fuel melting and less than 1%cladding strain)under all possible conditions, limits the required range for Thermal Overtemperature hT (OTBT)protection, and provides a backup to the High Neutron Flux trip.The OTbT trip function provides sufficient core protection to preclude departure from nucleate boiling (DNB)over a range of operating and transient conditions.
The setpoint is automatically varied with temperature.
pressure, and the axial power distribution.
The F(d,l)penalty function adjusts the trip setpoint for axial peaks greater than design.Revisions to the limiting safety system settings for the OTZT and OPbT trip functions (Table 2.2-1, Notes 1, 2, 3, and 4)are proposed to maintaio consistency with the non-LOCA Accident Analysis.These trip functions provide primary protection against DNB and fuel centerline melting (excessive kw/ft)during postulated transients.
The proposed settings have m 31944-1w.wpt:1d441295 1.2-1
2.2 NSSS DESIGN TRANSIENTS The NSSS design transients evaluation for the Donald C.Cook Nuclear Plant Unit 1 SGTP Program was completed and confirmed that the NSSS design transients developed as part of the Donald C.Cook Nuclear Plant Units 1 and 2 Rerating Program continue to apply to Donald C.Cook Nuclear Plant Unit 1 at the increased SGTP conditions.
The evaluation consisted of a comparison of the NSSS performance parameters for the SGTP Program with the parameters for the Rerating Program.The comparison concluded that the SGTP Program parameters that have the potential to impact the NSSS design transients (i.e., temperatures, pressures, and power levels)are bounded by the parameters used in the Rerating Program.The NSSS performance parameters that were not bounded in this manner (i.e., SGTP level and RCS flow)were evaluated and determined to not have a significant impact on the NSSS design transients.
Overall, the evaluation concluded that the NSSS design transients developed as part of the Rerating Program continue to apply to Donald C.Cook Nuclear Plant Unit 1.m:51944.1w.wpf:1d441195 2.2-1 Program).These parameters were used for selected analyses, where high primary temperatures were limiting, and sensitive to RCS loop flow.m&1944-1w.wpf:1d~1195 2.1-2 TABLE 2.1-1 COOK NUCLEAR PLANT UNIT 1 NSSS PERFORMANCE PARAMETERS FOR SGTP PROGRAM Parameter NSSS Power, MWt Core Power, MWt RCS Flow,(gpm/loop)
Minimum Measured Flow, (total gpm)RCS Temperatures,'F Core Outlet Vessel Outlet Core Average Vessel Average Vessel/Core Inlet Steam Generator Outlet Zero Load RCS Pressure, psia Steam Pressure, psia Steam Flow, (10'b/hr.tot.)
Feedwater Temperature,'F%SG Tube Plugging'Unit 1, Original)Case 1 3250 3250 88,500 361,600 602.0 599.3 570.5 567.8 536.3 536.3 547.0 2250 758 14.12 434.8 Flow Definitions:
RCS Flow (Thermal Design Flow)-The conservatively low flow used for thermal/hydraulic design.The design parameters listed above are based on this flow.mal 944.1w.wpf:1d~1195 2.1-3
TABLE 2.1-1 (continued)
COOK NUCLEAR PLANT UNIT 1 NSSS PERFORMANCE PARAMETERS FOR SGTP PROGRAM Parameter (Revised)(Revised)Case 2 Case 3 (Revised)(Revised)Case 4 Case 5'SSS Power, MWt Core Power, MWt RCS Flow, (gpm/loop)
Minimum Measured Flow, (total gpm)-RCS Temperatures,'F Core Outlet Vessel Outlet Core Average Vessel Average Vessel/Core Inlet Steam Generator Outlet Zero Load RCS Pressure, psia Steam Pressure, psia Steam Flow, (10'b/hr.tot.)
Feedwater Temperature,'F%SG Tube Plugging 3262 3250 83,200 339,100 589.7 586.8 555.8 553.0 519.2 518.9 547.0 2250 or 2100 595 14.12 434.8 30 3262 3250 83,200 339,100 611.9 609.1 579.4 576.3 543.5 543.2 547.0 2250 of 2100 749 14.17 434.8 30 3262 3250 79,000 339,100 591.5 588.5 556.0 553.0 517.5 517.2 547.0 2250 of 2100 589 14.12 434.8 30 3262 3250 79,000 339,100 613.6 610.8 579.7 576.3 541.8 541.6 547.0 2250 or 2100 742 14.17 434.8 30 Flow Definitions:
RCS Flow (Thermal Design Flow)-The conservatively low flow used for thermal/hydraulic design.The design parameters listed above are based on this flow.-Minimum Measured Flow-The flow specified in the Tech.Specs.which must be confirmed or exceeded by the flow measurements obtained during plant staftup and is the flow used in reactor core DNB analyses for plants applying the Revised Thermal Design Procedure.
MMF based on a flow measurement uncertainty of 1.9%.Analyses also bound a MMF of 341,100 gpm total which reflects a flow measurement uncertainty of 2.5%.m%1944-1w.wpf:1d441195 2.1-4
2.3 CQNTROl/PROTECTION SYSTEM SETPQINTS Control Systems were evaluated and found to be bounded by the analyses performed as part of the Rerating Program.These analyses reflected the objective of optimizing control parameters, primarily with respect to two aspects of plant behavior: stability of the control systems and operating margins to the various reactor protection system trips.The flexibility identified during the Rerating Program to adjust the full load vessel average temperature and primary pressure as necessary on a cycle-to-cycle basis remains applicable to the SGTP Program.Control systems setpoints are selected for each fuel cycle from those analyzed for the Rerating Program.Therefore, the plant will be adequately stable for all SGTP Program operating conditions and operates with adequate margin to reactor protection system setpoints.
m 81944-1w.wpf:1d~1195 2.3-1 rod in the three phases.The Revised PAD Fuel Thermal Safety Model, described in References 13, generates the initial fuel rod conditions input to LOCBART.SATAN-Vl calculates the RCS pressure, enthalpy, density;and the mass and energy flow rates in the RCS, as well as steam generator energy transfer between the primary and secondary systems as a function of time during the blowdown phase of the LOCA.SATAN-Vl also calculates the accumulator water mass and internal pressure and the pipe break mass and energy flow rates that are assumed to be vented to the containment during blowdown.At the end of the blowdown, information on the state of the system is transferred to the REFILL code which performs the calculation of the refill period to bottom of core recovery time.Once the vessel has refilled to the bottom of the core, the ref lood portion of the transient begins.The BASH code is used to calculate the thermal-hydraulic simulation of the RCS for the ref lood phase.Information concerning the core boundary conditions is taken from all of the above codes and input to the LOCBART code for the purpose of calculating the core fuel rod thermal response for the entire transient.
From the boundary conditions, LOCBART computes the fluid conditions and heat transfer coefficient for the full length of the fuel rod by employing mechanistic models appropriate to the actual flow and heat transfer regimes.Conservative assumptions ensure that the fuel rods modeled in the calculation represent the hottest rods in the entire core.The large break analysis was performed with the December 1981 version of the Evaluation Model modified to incorporate the BASH (Reference 7)computer code.Input Parameters and Initial Conditions:
The analysis presented in this section was performed with a reactor vessel upper head temperature equal to the RCS hot leg temperature and a uniform SGTP level of 30%.The analysis is also based on plant operation with the RHR cross-tie valves closed, and an EDG start time of 30 seconds which results in a safety injection delay time of 47 seconds.A list of plant input parameters used in the large break LOCA analysis is provided in Table 3.1-2.A range of reactor operating temperatures was analyzed in order to justify plant operation at a reactor power level of 3250 MWt between 609.1'F to 586.8'F in the hot legs and 543.5'F and 519.2'F in the cold legs.In addition to the temperature range analyzed, initial.RCS pressure was also varied to justify plant operation at 2250 and 2100 psia.A full spectrum break analysis was performed for nominal RCS conditions (initial RCS pressure of 2250 psia'and hot leg temperature of 609.1'F)from which the limiting break size was determined.
The limiting break was then reanalyzed for the reduced hot leg temperature of 586.8'F and nominal RCS pressure of 2250 psia.The limiting break was also reanalyzed for the nominal hot leg m:4)944-2w.wpt:1d~1195 3.1-6

2.0 BASIS

FOR EVALUATIONS/ANALYSES PERFORMED The purpose of the SGTP Program was to perform the necessary NSSS-related efforts to support an increase'in the level of SGTP to as high as 30%and continue operational flexibility in terms of primary temperature and pressure.In addition to the change in parameters associated with the increased SGTP level, additional changes were incorporated into the analyses, as described in Section 1.0 (e.g., EDG delay time, pressurizer safety valve tolerance, etc.).Previously, AEPSC submitted a report for NRC review in October 1988, which provided the necessary analysis, documentation, and licensing effort to support operation at reduced primary temperatures and pressures.
These analyses were performed in an effort to reduce the propensity for the initiation and propagation of corrosion in the Cook Nuclear Plant Unit 1 Series 51 steam generator tubes.The Westinghouse input for this submittal was provided in WCAP 11902 (Reference 1).The efforts performed for WCAP 11902 supported 100%thermal power operation (3250 MWt core power)in the range of vessel average temperatures between 547'F and 576.3'F, at primary pressure values of 2100 psia and 2250 psia.The pnmary pressures were intended as two discrete values;the program was not structured to support a continuous range of primary pressures.
The intent of the reduced primary pressure value is to minimize the primary to secondary pressure drop across the steam generator tubes at reduced temperature operation.
In addition, the analyses and evaluations performed support a maximum average tube plugging level of 10%, with a peak steam generator tube plugging level of 15%.Subsequently, a supplement to WCAP 11902 was issued as the Westinghouse input for a second submittal to the NRC to summarize the additional efforts performed to support a rerating of Cook Nuclear Plant Unit 1 and to provide part of the support for a Unit 2 rerating (Reference 2).The impact of this document on Cook Nuclear Plant Unit 1 is to support the licensing of a power uprating (in addition to supporting the range of operating conditions described above)to 3425 MWt NSSS.Only the reduced temperature and pressure portion of this program and associated operational improvements have been approved and implemented at this time.AEPSC currently selects the desired operating conditions from within the range addressed in the Rerating Program on a cycle-to-cycle basis.The efforts documented in WCAP-11902 and Supplement are referred to throughout this report as the"Rerating Program".The RCS temperatures of the SGTP Program were chosen to be within the bounds of the Rerating Program.The two primap pressure values of 2100 psia and 2250 psia were evaluated.
The maximum average and peak SGTP level was increased to 30%.A corresponding reduction in thermal design flow (TDF)and also a 5%loop flow asymmetry were also evaluated.
Because the range of NSSS parameters was chosen to be within the bounds of the Rerating Program, many of the analyses performed for the Rerating Program (Reference 1 and 2)remain applicable to the SGTP Program.Upon approval of the analyses m 31944-1w.wpf:1d441295 2.0-1 and evaluations in this report, AEPSC will select the desired operating conditions from within the range addressed in the SGTP Program on a cycle to cycle basis.References 1.WCAP-11902,"Reduced Temperature and Pressure Operation for Donald C.Cook Nuclear Plant Unit 1 Licensing Report", October 1988 2.WCAP-11902, Supplement,"Rerated Power and Revised Temperature and Pressure Operation for Donald C.Cook Nuclear Plant Units 1 and 2 Licensing Report," September 1989 m:11944.1w.wpf:1d441295 2.0-2 TABLE 3.1-14 SMALL-BREAK LOCA CALCULATIONS k3%MAIN STEAM SAFETY VALVE SETPOINT TOLERANCE ANALYSIS RESULTS NOTRUMP Peak Clad Temperature
('F)Peak Clad Temperature Location (ft)Peak Clad Temperature Time (sec)Local Zr/H,O Reaction Maximum (%)Local Zr/H,O Reaction Location (ft)Total Zr/H,O Reaction (%)Rod Burst Burst and Blockage Penalty ('F)Total Peak Clad Temperature
('F)12.0 12.0 1890 4042 5.06 3.75 12.0 12.0<1.0<1.0 None None 117 15 2068 Reduced Pressure, Reduced Temperature 3-Inch 2-inch 1951 1833 Main steam safety valve setpoint tolerance increase case at 3250 MWt core power.m%1 944-2w.wpt:1d441195 3.1-35

3.0 SAFETY

EVALUATIONS/ANALYSES PERFORMED 3.1 LOSS OF COOLANT ACCIDENT (LARGE BREAK AND SMALL BREAK)3.1.1 Large Break LOCA A large break LOCA analysis was performed for Donald C.Cook Nuclear Plant Unit 1 to support an increase in the steam generator tube plugging level to 30%, while maintaining the operational flexibility of the plant by analyzing a range of initial RCS temperature conditions and two discrete RCS pressures.
The large break analysis was performed with the 1981 version of the Westinghouse ECCS Evaluation Model using the BASH computer code.Analysis assumptions included 30%steam generator tube plugging, operation at a reactor power level of 3250 MWt with the RHR cross-tie valves closed, a total peaking factor of 2.15, and a hot channel enthalpy rise peaking factor of 1.55.Safety injection flows were based on pump head degradation of 15%for the high head safety injection pumps and RHR pumps, and 10%for the centrifugal pumps.The EDG start time was also increased to 30 seconds.The analysis assumed a range of operating temperatures in order to justify plant operation between 609.1'F and 586.8'F in the hot legs and 543.5'F and 519.2'F in the cold legs.These temperature ranges represent the Unit 1 power capability parameters for 30%peak uniform steam generator tube plugging displayed in Table 2.1-1.Initial RCS pressure was also varied to justify plant operation at 2100 and 2250 psia.A full spectrum break analysis was performed for nominal RCS conditions (initial RCS pressure of 2250 psia and hot leg temperature of 609.1'F)from which the limiting break discharge coefficient was determined.
The limiting break was then reanalyzed for the reduced hot leg temperature and nominal RCS pressure of 2250 psia, and also for the nominal hot leg temperature and RCS pressure of 2100 psia.The above cases were all analyzed with minimum safety injection flow.The limiting break was also analyzed with maximum safety injection flow.The limiting break size was determined to be C~=0.4 at the nominal hot leg temperature (Type=609.1'F)and a pressure of 2100 psia with minimum safety injection flow.The peak cladding temperature was calculated to be 2164'F which is less than the 2200'limit in 10 CFR 50.46.A detailed description of the large break LOCA analysis is presented below.Identification of Causes and Fre uen Classification A LOCA is the result of a pipe rupture of the RCS pressure boundary.For the analyses reported here, a major pipe break (large break)is defined as a rupture with a total cross-sectional area equal to or greater than 1.0 ft'.This event is considered an ANS Condition IV event, a limiting fault, in that it is not expected to occur during the lifetime of Cook Nuclear Plant Unit 1, but is postulated as a conservative design basis.mal 944-2w.wpf:1d~1195 3.1-1 The Acceptance Criteria for the LOCA are described in 10 CFR 50.46 (10 CFR 50.46 and Appendix K of 10 CFR 50, 1974-Reference 1)as follows: 1.The calculated peak fuel element clad temperature is below the requirement of 2200'F.2.The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of Zircaloy in the reactor.3.The localized cladding oxidation limit of 17 percent is not exceeded during or after quenching.
4.The core remains amenable to cooling during and after the break.5.The core temperature is reduced and decay heat is removed for an extended period of time, as required by the long-lived radioactivity remaining in the core.These criteria were established to provide a significant margin in emergency core cooling system (ECCS)performance following a LOCA.WASH-1400 (USNRC 1975), Reference 2, presents a study in regards to the probability of occurrence of RCS pipe ruptures.II Se uence of Events and S stems 0 erations Should a major break occur, depressurization of the RCS results in a pressure decrease in the pressurizer.
The reactor trip signal subsequently occurs when the pressurizer low pressure trip setpoint is reached.A safety injection signal is generated when the appropriate setpoint is reached.These countermeasures will limit the consequences of the accident in two ways: Reactor trip and borated water injection supplement void formation in causing rapid reduction of power to a residual level corresponding to fission product decay heat.No credit is taken in the LOCA analysis for the boron content of the injection water.However, an average RCS/sump mixed boron concentration is calculated to ensure that the core remains subcritical.
In addition, the insertion of control rods to shut down the reactor is neglected in the large break analysis.2.Injection of borated water provides for heat transfer from the core and prevents excessive clad temperatures.
In the present Westinghouse design, the large break singiedailure is the loss of one RHR (low head)pump.This means that credit could be taken for two charging pumps, two safety injection pumps, and one low head pump.The following is a discussion of the modeling m:51944.2w.wpt:1d441195 3.1-2 procedure for the minimum safeguards and the flow spilling from a break of an ECCS branch injection line{i.e., the spilling line assumptions).
The current procedure for large break analyses assumes that at least one train of ECCS is available for delivery of water to the RCS.Although the single failure is an RHR pump, only one pump in each subsystem is assumed to deliver to the primary loops.However, both EDGs are assumed to start in the modeling of the containment deck fans and sprays.Modeling full containment heat removal systems operation is required by Branch Technical Position CSB 6-1 and is conservative for the large break LOCA.The charging pump starts and delivers flow through the injection lines{one per loop)with one branch injection line spilling to the containment backpressure.
To minimize delivery to the reactor, the branch line chosen to spill is selected as the one with the minimum resistance.
When one safety injection pump and one low head residual heat removal pump start, flow is delivered to the RCS through the accumulator injection lines.Again, one line, with the minimum resistance, is assumed to spill to containment backpressure.
In addition, the safety injection pump and low head RHR pump performance curves were degraded by 15%.For the charging pumps, the performance cuwes were degraded by 10%and a 25 gpm flow imbalance was assumed.Therefore, in the large break ECCS analysis performed by Westinghouse, single failure is conservatively accounted for via the loss of an ECCS train, and the spilling of the minimum resistance injection line despite full containment active heat removal system operation{i.e., two EDGs)~The time sequence of events following a large break LOCA is presented in Table 3.1-1.Before the break occurs, the unit is in an equilibrium condition; that is, the heat generated in the core is being removed via the secondary system.During blowdown, heat from fission product decay, hot internals and the vessel, continues to be transferred to the reactor coolant.At the beginning of the blowdown phase, the entire RCS contains subcooled liquid which transfers heat from the core by forced convection with some fully developed nucleate boiling.After the break develops, the time to departure from nucleate boiling is calculated, consistent with Appendix K of 10 CFR 50 (Reference 1).Thereafter, the core heat transfer is unstable, with both nucleate boiling and film boiling occurring.
As the core becomes uncovered, both turbulent and laminar forced convection and radiation are considered as core heat transfer mechanisms.
The heat transfer between the RCS and.the secondary system may be in either direction, depending on the relative temperatures.
In the case of continued heat addition to the secondary system, the secondary system pressure increases and the main steam safety valves may actuate to limit the pressure.Makeup water.to.the secondary side is automatically provided by the auxiliary feedwater system.The safety injection signal actuates a feedwater isolation signal which isolates normal feedwater flow by closing the main feedwater isolation mA1944-2w.wpf:1 d441195 3.1-3 valves, and also initiates e'mergency feedwater flow by starting the auxiliary feedwater pumps.The secondary flow aids in the reduction of RCS pressure.When the RCS depressurizes to 600 psia, the accumulators begin to inject borated water into the reactor coolant loops.The conservative assumption is made that accumulator water injected bypasses the core and goes out through the break until the termination of bypass.This conservatism is again consistent with Appendix K of 10 CFR 50.Since loss of offsite power (LOOP)is assumed, the RCPs are assumed to trip at the inception of the accident.The effects of pump coastdown are included in the blowdown analysis.The blowdown phase of the transient ends when the RCS pressure{initial values with uncertainty assumed to be 2317 psia or 2033 psia)falls to a value approaching that of the containment atmosphere.
Prior to or at the end of the blowdown, the mechanisms that are responsible for the emergency core cooling water bypassing the core are calculated not to be effective.
At this time{called end-of-bypass) refill of the reactor vessel lower plenum begins.Refill is completed when emergency core cooling water has filled the lower plenum of the reactor vessel, which is bounded by the bottom of the fuel rods (called bottom of core recovery time).The ref lood phase of the transient is defined as the time period lasting from the end-of-refill until the reactor vessel has been filled with water to the extent that the core temperature rise has been terminated.
From the latter stage of blowdown and then the beginning-of-ref lood, the safety injection accumulator tanks rapidly discharge borated cooling water into the RCS, contributing to the filling of the reactor vessel downcomer.
The downcomer water elevation head provides the driving force required for the ref looding of the reactor core.The low head and high head safety injection pumps aid in the filling of the downcomer and subsequently supply water to maintain a full downcomer and complete the ref looding process.Continued operation of the ECCS pumps supplies water during long-term cooling.Core temperatures have been reduced to long-term steady state levels associated with the dissipation of residual heat generation.
After the water level of the refueling water storage tank (RWST)reaches a minimum allowable value, coolant for long-term cooling of the core is obtained by switching to the cold leg recirculation phase of operation in which spilled borated water is drawn from the engineered safety features (ESF)containment sumps by the low head safety injection (residual heat removal)pumps and returned to the RCS cold legs.The containment spray system continues to operate to further reduce containment pressure.Approximately 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the initiation of the LOCA, the ECCS is realigned to supply water to the RCS hot legs in order to control the boric acid concentration in the reactor vessel.Long-term cooling includes long-term criticality control.Griticality control is achieved by determining the RWST and accumulator concentration necessary to maintain subcriticality without credit for RCCA insertion.
The necessary RWST and accumulator concentration is a m:$1944.2w.wpf:1d~1195 3.1-4 function of each core design and is checked each cycle.The current Technical Specification value is 2400 ppm to 2600 ppm boron (Reference 3).Core and S stem Performance Mathematical Model: The requirements of an acceptable ECCS evaluation model are presented in Appendix K of 10 CFR 50 (Federal Register 1974), Reference 1.Large Break LOCA Evaluation Model The analysis of a large break LOCA transient is divided into three phases: (1)blowdown, (2)refill, and (3)ref lood.There are three distinct transients analyzed in each phase, including the thermal-hydraulic transient in the RCS, the pressure and temperature transient within the containment, and the fuel and clad temperature transient of the hottest fuel rod in the core.Based on these considerations, a system of interrelated computer codes has been developed for the analysis of the LOCA.A description of the various aspects of the LOCA analysis methodology is given by Bordelon, Massie, and Zordan (Reference 4).This document describes the major phenomena modeled, the interfaces among the computer codes, and the features of the codes which ensure compliance with the Acceptance Criteria.The SATAN-VI, WREFLOOD, BASH and LOCBART codes, which are used in the LOCA analysis, are described in detail by Bordelon et al.(1974)"';Kelly et al.(Reference 6);Young et al.(Reference 7);and Bordelon et al.(Reference 4).Code modifications are specified in References 8, 9, 10 and 11.It is noted that the WREFLOOD code, which was previously used to calculate the RCS behavior during vessel lower plenum refill, has been replaced by the REFILL code as reported in Reference 18.The REFILL code is identical to the section of the WREFLOOD code that modeled the refill phase.These codes assess the core heat transfer geometry and determine if the core remains amenable to cooling throughout and subsequent to the blowdown, refill, and ref lood phases of the LOCA.The SATAN-Vl computer code analyzes the thermal-hydraulic transient in the RCS during blowdown and the REFILL computer code calculates this transient during the refill phase of the accident.The BASH code is used to determine the system response during the ref lood phase of the transient.
The LOTIC computer code, described by Hsieh and Raymund in WCAP-8355 and WCAP-8345 (Reference 12), calculates the containment pressure'ransient.
The containment pressure transient is input to BASH for the purpose of calculating the ref lood transient.
The LOCBART computer code calculates the thermal transient of the hottest fuel m A1944.2w.wpl:1d~1 195 3.1-5
temperature of 609.1'F and RCS pressure of 2100 psia.The cases analyzed are identified in Table 3.1-1.The bases used to select the numerical values that are input parameters to the analysis have been conservatively determined from extensive sensitivity studies (Westinghouse 1974 (Reference 14);Salvatori 1974 (Reference 15);Johnson, Massie, and Thompson 1975 (Reference 16).In addition, the requirements of Appendix K regarding specific model features were met by selecting models which provide a significant overall conservatism in the analysis.The assumptions which were made pertain to the conditions of the reactor and associated safety system equipment at the time that the LOCA occurs, and include such items as the core peaking factors, the containment pressure, and the performance of the ECCS.Decay heat generated throughout the transient is also conservatively calculated.
Another input parameter that affects LOCA analysis results is the assumed axial power shape at the beginning of the accident.Large break LOCA analyses have been traditionally performed using a symmetric, chopped cosine axial power shape.Recent calculations have shown that there was a potential for top-skewed power distributions to result in peak cladding temperatures (PCT)greater than those calculated with a chopped cosine axial power distribution.
Westinghouse has developed a process, which was applied to the Cycle 13 and 14 reloads for Cook Nuclear Plant Unit 1, that reasonably ensures that the cosine remains the limiting power distribution, by defining appropriate power distribution surveillance data.This process, called the power shape sensitivity model (PSSM), is described in a topical report (WCAP-12909-P), Reference 19, and further clarified in ET-NRC-91-3633, Reference 20, which are currently under NRC review.With implementation of the PSSM in the reload design process, top skewed axial power distributions that are potentially more limiting than the power distribution used in the ECCS analysis are reasonably precluded from occurring by the design and surveillance data provided to monitor the power distribution.
A meeting was held at the Westinghouse Licensing Office in Bethesda on December 17, 1981, between members of the U.S.Nuclear Regulatory Commission and members of the Westinghouse Nuclear Safety Department to discuss the impact of maximum safety injection on the large break ECCS analysis on a generic basis.Further discussion of this issue is provided in a letter from E.P.Rahe, Manager of Westinghouse Nuclear Safety Department, to Robert L.Tedesco of the U.S.Nuclear Regulatory Commission (Reference 17).A brief description of this issue is given below.Westinghouse ECCS analyses currently assume minimum safeguards for the safety injection flow, which minimizes the amount of flow to the RCS by assuming maximum injection line resistances, degraded ECCS pump performance, and the loss of one RHR pump as the most limiting single failure.This is conservatively modeled as a loss of one train of safety injection, including RHR pump, safety injection pump and centrifugal charging pump.Both containment spray pumps are assumed operable.This is the limiting single failure assumption when offsite mA1 944-2w.wpf:1d441195 3.1-7 power is unavailable for most Westinghouse plants.However, for some Westinghouse plants, the current nature of the Appendix K ECCS evaluation models is such that it may be more limiting to assume the maximum possible ECCS flow delivery.In that case, maximum safeguards which assume minimum injection line resistances, enhanced ECCS pump performance, and no single failure, result in the highest amount of flow delivered to the RCS.Therefore, the worst break for Cook Nuclear Plant Unit 1 (C~=0.4 for nominal hot leg temperature of 609.1'F and RCS pressure of 2100 psia)was reanalyzed assuming maximum safeguards.
Results: Based on the results of the LOCA sensitivity studies (Westinghouse 1974 (Reference 14);Salvatori 1974 (Reference 15);Johnson, Massie, and Thompson 1975 (Reference 16)the limiting large break was found to be the double-ended cold leg (DECL)guillotine.
Therefore, only the DECL guillotine break is considered in the large break ECCS performance analysis.Calculations were performed for a range of Moody break discharge coefficients.
The results of these calculations are summarized in Table 3.1-1.The containment data used to generate the LOTIC backpressure transient are shown in Table 3.1-3.The mass and energy release data used for the limiting minimum safeguards case are shown in Table 3.1A.Nitrogen release rates to the containment are given in Table 3.1-5.Figures 3.1-1 a through 3.1-19 present the results of the cases analyzed for the large break LOCA.The alpha designation in the figure number corresponds to the cases as described in Table 3.1-1.Fi ures 3.1-1a to 1f The system pressure shown is the calculated core pressure.Fi ures 3.1-2a to 2f The flow rate from the break is plotted as the sum of both ends of the guillotine break.Fi ures 3.1-3a to 3f The core pressure drop shown is from the lower plenum, near the core, to the upper plenum at the core outlet.Fi ures 3.1-4a to 4f The core flow is shown during the blowdown phase of the transient.
Fi ures 3.1-5a to 5f The accumulator flow during blowdown is plotted as the sum of that injected into the intact cold legs.mal 944-2w.wpf:1d~1195 3.1-8 Fi ures 3.1-6a to 6f The core and downcomer collapsed liquid water level, and the core quench front are plotted during the ref lood phase of the transient.
Fi ures 3.1-7a to 7f The core inlet flow is shown as it is calculated during the ref lood phase.Fi ures 3.1-8a to 8f The total accumulator and pumped ECCS flow injected into the intact cold legs during ref lood is shown.Fi ures 3.1-9a to 9f The integral of the core inlet flow during ref lood as calculated with BASH is plotted.Fi ures 3.1-10a to 10f The mass flux is plotted at the hot spot (the node which produced the peak clad temperature) on the hot rod.Fi ures 3.1-11a to 11f The heat transfer coefficient is plotted at the hot spot on the hot rod.Fi ures 3.1-12a to 12f Fi ures 3.1-13a to 13f The vapor temperature at the hot spot on the hot rod is plotted.4 The clad temperature at the hot spot is shown for the hot rod.Fi ure 3.1-14 The containment pressure transient used in the analysis is provided for the minimum Sl case.Fi ures 3.1-15 to 18 These figures show the heat removal rates of the heat sinks found in the lower and upper compartment and the heat removal by the sump and lower compartment spray.Fi ure 3.1-19 This figure shows the temperature transients in both the lower and upper compartments of containment.
As shown in Table 3.1-1, the limiting case for Cook Nuclear Plant Unit 1 is Case E (C,=0.4 for nominal hot leg temperature of 609.1'F and RCS pressure of 2100 psia).The maximum clad temperature calculated for a large break is 2164'F, which is less than the Acceptance Criteria limit of 2200'F.The maximum local metal-water reaction is 14.30 percent, which is well below the embrittlement limit of 17 percent as required by 10 CFR 50.46.The total core metal-water reaction for all breaks is less than the 1 percent criterion of 10 CFR 50.46.The clad temperature transient is terminated at a time when the-core geometry is still amenable to cooling.As a result, the core temperature will continue to drop and the ability to remove decay heat generated in the fuel for an extended period of time will be provided.mA1 944-2w.wpt:1dM1195 3.1-9 References 1."Acceptance Criteria for Emergency Core Cooling System for Light Water Cooled Nuclear Power Reactors," t0 CFR 50.46 and Appendix K of t0 CFR 50, Federal VI 39M 1 3.2.U.S.Nuclear Regulatory Commission 1975,"Reactor Safety Study-An Assessment of Accident Risks in U.A.Commercial Nuclear Power Plants," WASH-1400, NUREG-75/014.3.Attachment 13 to letter, M.P.Alexich, IMECo, to H.R.Denton, NRC, March 26, 1987, AEP: NRC:0916W.
4.Bordelon, F.M.;Massie, H.W.;and Zordan, T.A."Westinghouse ECCS Evaluation Model-Summary," WCAP-8339, 1974.5.Bordelon, F.M.et al~,"SATAN-Vl Program;Comprehensive Space, Time Dependent Analysis of Loss-of-Coolant," WCAP-8302 (Proprietary) and WCAP-8306 (Non-Proprietary), 1974.~~e 0 6.Kelly, R.D.et al.,"Calculation Model for Core Reflooding After a Loss-of-Coolant Accident (WREFLOOD Code)," WCAP-8170 (Proprietafy) and WCAP-8171 (Non-proprietary), 1974.7.Young, M.Y.et al,"The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code," WCAP-10266-P-A Revision 2 (Proprietary), 1987.8.Rahe, E.P.(Westinghouse), letter to J.R.Miller (USNRC), Letter No.NS-EPRS-2679, November 1982.9.Rahe, E.P.,"Westinghouse ECCS Evaluation Model, 1981 Version," WCAP-9920-P-A (Proprietafy Version), WCAP-9221-P-A (Non-Proprietafy version), Revision 1, 1981.10.Bordelon, F.M., et al.,"Westinghouse ECCS Evaluation Model-Supplementary Information," WCAP-8471 (Proprietary) and WCAP-8472 (Non-proprietary), 1975.11.Thomas, C.O., (NRC),"Acceptance for Referencing of Licensing Topical Report WCAP-10484(P)/1 0485(NP),'Spacer Grid Heat Transfer Effects During Ref lood,'" Letter to E.P.Rahe (Westinghouse), June 21, 1984.m:11944.2w.wpf:1d~1195 3.1-10
12.Hsieh, T., and Raymund, M.,"Long-Term Ice Condenser Containment LOTIC Code Supplement 1," WCAP-8355, Supplement 1, May 1975, WCAP-8345 (Proprietary), July 1974.13."Westinghouse Revised PAD Code Thermal Safety Model," WCAP-8720, Addendum 2 (Proprietary) and WCAP-8785 (Non-propnetary).
14."Westinghouse ECCS-Evaluation Model Sensitivity Studies," WCAP-8341 (Proprietary) and WCAP-8342 (Non-proprietary), 1974.15.Salvatori, R.,"Westinghouse ECCS-Plant Sensitivity Studies," WCAP-8340 (Proprietary) and WCAP-8356 (Non-pro prietary), 1974.16.Johnson, W.J.;Massie, H.W.;and Thompson, C.M."Westinghouse ECCS-Four Loop Plant (17x17)Sensitivity Studies," WCAP-8565-P-A (Proprietary) and WCAP-8566-A
{Non-proprietary), 1975.17.Rahe, E.P.(Westinghouse).
Letter to Robert L.Tedesco (USNRC), Letter No.NS-EPR-2538, December 1981.18.Liparulo, N.J.(Westinghouse), letter to W.T.Russel (USNRC), Letter No.~~~NTD-NRC-94-4143, May 23, 1994.19.Stucker, D.L.et al.,"Westinghouse ECCS Evaluation Model: Revised Large Break LOCA Power Distribution Methodology," WCAP-12909-P (Proprietary) and WCAP-12935-NP (Non-Proprietary), May 1991.20.Tritch, S.R.(Westinghouse), letter to R.C.Jones (USNRC), Letter No.ET-NRC-91-3633, October 25, 1991.m&1944-2w.wpf:1d441195 3.1-11

3.1.2 Small

Break LOCA A small break LOCA analysis has been performed for the Donald C.Cook Unit 1 Nuclear Plant to support an increase in the steam generator tube plugging level to 30%, while maintaining the operational flexibility of the plant by demonstrating that the 10 CFR 50.46 Acceptance Criteria can be met for a range of initial RCS pressure and temperature conditions.
The small break LOCA analysis was performed with the Westinghouse small break LOCA ECCS Evaluation Model using the NOTRUMP code""", including the recent changes in Addendum 2"'o incorporate modeling of safety injection into the broken loop and the COSI condensation model.The key analysis input assumptions included 30%peak uniform steam generator tube plugging, operation at a reactor power level of 3250 MWt with the HHSI cross-tie discharge valves closed, a total peaking factor of 2.32, and a hot channel enthalpy rise peaking factor of 1.55.Also incorporated in the analysis are a reduced hot assembly average power and a power shape based on a reduced axial offset of+20%.Safety injection flows are based on pump head degradation of 15%for the high head safety injection pumps and 10%for the centrifugal charging pumps, and the emergency diesel generator start time was increased to 30 seconds.The analysis was performed in order to bound plant operation within the range of RCS temperatures specified in the Unit 1 power capability parameters for 30%uniform steam generator tube plugging in Table 2.1-1, and at RCS pressures of 21 00 and 2250 psia.A single break size analysis was performed at the previously-limiting break size of three inches.The calculation used the reduced temperature, reduced pressure operating condition which was previously demonstrated to be the limiting operating condition for the small break analysis.Based on an evaluation of the break spectrum and the range of operating conditions, it was concluded that the analyzed case would remain bounding with respect to peak clad temperature.
The peak cladding temperature was calculated to be 1443'F which is less than the 2200'F limit in 10 CFR 50.46.A detailed description of the analysis is presented in the following sections.Since the analysis to support 30%steam generator tube plugging is an extension of previous small break LOCA analyses performed for Cook Nuclear Plant Unit 1, the description also includes a discussion of the previous analyses.These include the Rerating Program analyses currently in the FSAR which were performed for a reactor power level of 3588 MWt, and the analysis performed for a reactor power level of 3250 MWt to support an increase in the main steam safety valve (MSSV)setpoint tolerance to k3%.m:11944.2w.wpt:1d~1195 3.1-12 3.1.2.1 Rerating Program Analysis Analysis of Effects and Consequences Method of Analysis For loss-of-coolant accidents due to small breaks less than one square foot, the NOTRUMP computer code"~'s used to calculate the transient depressurization of the RCS as well as to describe the mass and enthalpy of flow through the break.The NOTRUMP computer code is a state-of-the-art one-dimensional general network code incorporating a number of advanced features.Among these are calculation of thermal non-equilibrium in all fluid volumes, flow regime-dependent drift flux calculations with counter-current flooding limitations, mixture level tracking logic in multiple-stacked fluid nodes and regime-dependent heat transfer correlations.
The NOTRUMP small-break LOCA emergency core cooling system (ECCS)evaluation model was developed to determine the RCS response to design basis small break LOCAs, and to address NRC concerns expressed in NUREG-0611,"Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse-Designed Operating Plants." The reactor coolant system model is nodalized into volumes interconnected by flow paths.The broken loop is modelled explicitly, while the three intact loops are lumped into a second loop.Transient behavior of the system is determined from the governing conservation equations of mass, energy, and momentum.The multinode capability of the program enables explicit, detailed spatial representation of various system components which, among other capabilities, enables a proper calculation of the behavior of the loop seal during a loss-of-coolant accident.The reactor core is represented as heated control volumes with associated phase separation models to permit.transient mixture height calculations.
Detailed descriptions of the NOTRUMP code and the evaluation model are provided in References 1 and 2.Safety injection systems consist of gas pressurized accumulator tanks and pumped injection systems.Minimum emergency core cooling system availability is assumed for the analysis.Assumed pumped safety injection characteristics as a function of RCS pressure used as boundary conditions in the Rerating Program analysis are shown in Figure 3.1-20 and in Table 3.1-6.The safety injection flow rates presented are based on pump performance curves degraded 10 percent from the design head and are consistent with closure of the high head safety injection system cross-tie valve.The effect of flow from the RHR pumps is not considered in the small break analyses since their shutoff head is lower than the RCS pressure during the time portion of the transient considered here.Safety injection is delayed 27 seconds after the occurrence of the injection signal to account for diesel generator startup and emergency power bus loading in case of a loss of offsite power coincident with an accident.mA1 944-2w.wpf:1d~1195 3.1-13 Peak clad temperature calculations are performed with the LOCTA-IV"'ode using the NOTRUMP calculated core pressure, fuel rod power history, uncovered core steam flow and mixture height as boundary conditions.
Figure 3.1-21 depicts the hot rod axial power shape used to perform the small break analysis for the Rerating Program.This shape was chosen because it represents a distribution with power concentrated in the upper regions of the core.Such a distribution is limiting for small-break LOCAs because it minimizes coolant level swell, while maximizing vapor superheating and fuel rod heat generation in the uncovered elevations.
The small break LOCA analysis assumes the core continues to operate at full rated power until the control rods are completely inserted.Results This section presents results of the limiting break analysis (as determined by the highest calculated peak fuel rod clad temperature) for a range of break sizes and RCS pressures and temperatures for a reactor power level of 3588 MWt.The limiting break was found to be a 3-inch diameter cold leg break initiated at reduced RCS pressure and temperature conditions.
The maximum temperature attained during the transient was 2122'F.A list of input assumptions used in the Rerating Program analysis for reduced pressure and temperature conditions is provided in Table 3.1-7.The results of a three break spectrum analysis performed at reduced RCS pressure and temperature conditions are summarized in Table 3.1-8, while the key transient event times are listed in Table 3.1-9.Figures 3.1-22 through 3:1-29 show the limiting three-inch break transient, respectively:
-RCS pressure,-Core mixture level,-Peak clad temperature,-Core outlet steam flow,-Hot spot rod surface heat transfer coefficient,-Hot spot fluid temperature,-Cold leg break mass flow rate, and-Safety injection mass flow rate.During the initial period of the small-break transient, the effect of the break flow is not strong enough to overcome the flow maintained by the reactor recirculation cooling pumps as they coast down.Normal upward flow is maintained through the core and core heat is adequately removed.At the low heat generation rates following shutdown, the fuel rods continue to be well cooled as long as the core is covered by a two-phase mixture level~From the clad temperature transient for the 3-inch break calculation shown in Figure 3.1-24, it is seen that the peak clad temperature occurs near the time at which the core is most deeply uncovered when the top of the core is steam cooled.This time is also accompanied by the highest vapor superheating above the mixture level.A comparison of the total break flow to containment shown in Figure 3.1-28 to the safety injection flow rate shown in Figure 3.1-29 shows that at m 31944-2w.wpf:1d~
1 195 3.1-14 the time the transient was terminated, the safety injection flow being delivered to the RCS exceeded the mass flow out the break.Although the inner vessel core mixture level has not yet covered the entire core, there is no longer a concern of exceeding the 10 CFR 50.46 criteria since the pressure is gradually decaying and there is a net mass inventory gain.As the RCS inventory continues to gradually increase, the core mixture level will continue to increase and the fuel clad temperatures will continue to decline.Reratin Pro ram Break S ectrum Cases Studies documented in Reference 4 determined that the limiting small-break size occurred for breaks less than 10 inches in diameter.To ensure that the 3-inch diameter break was limiting for the reduced temperature and pressure RCS conditions, calculations were also run with breaks of 2 inches and 4 inches.The results of these calculations are shown in the Results Table 3.1-8 and Sequence of Events Table 3.1-9.Plots of the following parameters are shown in Figures 3.1-30 through 37 for the 2-inch break, and Figures 3.1-38 through 43 for the 4-inch break.-RCS pressure,-Core mixture level,-Peak clad temperature,-Core outlet steam flow,-Hot spot rod surface heat transfer coefficient,-Hot spot fluid temperature,-Cold leg break mass flow rate, (for the 2-inch case only), and-Safety injection mass flow rate (for the 2-inch case only).As seen in Table 3.1-8, the maximum clad temperatures were calculated to be less than that for the 3-inch break.Additional Reratin Pro ram Anal ses Calculations were also performed for Cook Nuclear Plant Unit 1 with the NOTRUMP"~'nd LOCTA-IV"'odes to examine the influence of initial loop fluid operating temperatures and operating pressures on small break LOCA peak clad temperature.
These additional analyses confirmed that the most limiting PCT result was that from the reduced temperature and pressure limiting 3-inch diameter break described previously.
To support operation of the Gook Nuclear Plant Unit 1 at RCS pressures of 2100 psia and 2250 psia for a range of loop operating temperatures, two additional analyses were performed.
Calculations were performed for a 3-inch diameter break for an initial RCS pressure of 2250 psia at initial loop fluid operating temperatures corresponding to T,~program setpoints of 547'F and 578'F.The results of these calculations are shown in the Results Table 3.1-10 m 31944-2w.wpf:1d441195 3.1-15 and the Sequence of Events Table 3.1-11.Plots of the following parameters are shown in Figures 3.1~through 51 for the reduced temperature and high pressure case, and Figures 3.1-52 through 59 foi the high temperature and high pressure case.-RCS pressure,-Core mixture level,-Peak clad temperature,-Core outlet steam flow,-Hot spot rod surface heat transfer coefficient,-Hot spot fluid temperature,-Cold leg break mass flow rate, and-Safety injection mass flow rate.As seen in Table 3.1-10, the maximum clad temperatures were calculated to be less than that for the 3-inch break initiated at reduced temperature and pressure conditions.
NUREG-0737<",Section II.K.3.31, required plant-specific small break LOCA analysis using an Evaluation Model revised per Section II.K.3.30.
In accordance with NRC Generic Letter 83-65"', generic analyses using NOTRUMP"~'ere performed and are presented in WCAP-11145"'.Those results demonstrate that in a comparison of cold leg, hot leg and pump suction leg break locations, the cold leg break location is limiting.3.1.2.2 Main Steam Safety Valve Setpoint Tolerance Relaxation Analysis Additional small break LOCA analyses were performed at a reactor power level of 3250 MWt to support an increase in the MSSV lift setpoint tolerance from~1%to~3%.Prior to the analysis performed to support MSSV tolerance relaxation, this mode of operation was supported by an evaluation limiting core power to 3250 MWt.The MSSV analyses were.performed for operation with the HHSI cross-tie valves closed and assuming a 25 gpm charging pump flow imbalance.
This resulted in a reduction in the charging pump flow, and thus a reduction in the total safety injection flow rate relative to the Rerating Program analysis.The limiting 3-inch break for reduced pressure and reduced temperature operating conditions was analyzed, since the Rerating Program analysis demonstrates that this case results in the most limiting clad temperature.
Since the basis for the limiting case determination remains valid, it was not necessary to analyze the full spectrum of cases.However, an analysis was also performed for a 2-inch break since a reduction in safety injection flow rate can potentially shift the limiting break to a smaller break size.The analysis for the 2-inch break confirmed that the limiting break did not shift to a smaller break size.-m:51944-2w.wpf:1d441195 3.1-16 A list of the plant input parameters for the t3%MSSV setpoint tolerance analysis is provided in Table 3.1-12.The results of the limiting 3-inch break analysis are presented in the Sequence of Events Table 3.1-13 and the Results Table 3.1-14.Results of the non-limiting 2-inch case are also provided in Tables 3.1-13 and 3.1-14.Plots of the following parameters for the 3-inch break analysis are shown in Figures 3.1-60 through 3.1-67, and for the 2-inch break in Figures 3.1-69 through 3.1.2-76:-RCS pressure-Core mixture level-Peak clad temperature
-Core outlet steam flow-Hot spot rod surface heat transfer coefficient
-Hot spot fluid temperature
-Cold leg break mass flow rate, and-Safety injection mass flow rate Figure 3.1-68 contains the power shape for a core power level of 3250 MWt which is applicable to both cases.The 3-inch break with HHSI cross-ties closed, initiated at reduced pressure and temperature operating conditions and a core power level of 3250 MWt, represents the licensing basis small break analysis for an increased MSSV setpoint tolerance of k3%.Application of a burst and blockage penalty resulted in a peak clad temperature of 2068'F, which was less than the 2200'F limit.3.1.2.3 30%Steam Generator Tube Plugging Analysis An additional small break LOCA analysis has been performed to support an increase in steam generator tube plugging level from 15%to a maximum of 30%in each steam generator.
The analysis was performed in order to bound plant operation between 609.1'F and 586.8'F in the hot legs and 543.5'F and 519.2'F in the cold legs.These temperature ranges are defined in the Unit 1 power capability parameters for 30%peak uniform steam generator tube plugging displayed in Table 2.1-1.The analysis also supports plant operation at RCS pressures of 2100 and 2250 psia.The analysis was performed for the limiting 3-inch break with HHSI cross-tie valves closed at reduced pressure and reduced temperature operating conditions and a core power level of 3250 MWt, which was previously demonstrated to result in the most limiting clad temperature.
An evaluation of the basis for the limiting case determination was performed and it was concluded that it was not necessary to perform a full break spectrum for this case.mh1 944.2w.wpf:1d441195 3.1-17
The analysis incorporates a 20 second increase in emergency diesel generator starting delay to 30 seconds, which results in a total safety injection delay of 47 seconds after the occurrence of the injection signal.The safety injectioh flow rates used in the analysis include a 5%increase in high head safety injection pump degradation, for a total of 15%degradation.
For the high head charging pumps, the performance curve degradation remained at 10%and a 25 gpm flow imbalance was assumed.The analysis also includes a reduction in the maximum axial offset from+30%to+20%and a reduction in the maximum hot assembly peaking factor from 1.433 to 1.38, with a corresponding change in the axial power shape used in the analysis.The use of the revised axial offset and hot assembly power factor in the small break LOCA analysis are consistent with the current core design and operation limits.An evaluation of up to 5%RCS loop flow asymmetry was also performed to support the analysis.A list of the plant input parameters for the 30%SGTP analysis is provided in Table 3.1-15.Previously, safety injection into the broken loop was not modeled in the Westinghouse small break LOCA analyses since it was assumed that the additional safety injection would be a benefit.Because recent studies have shown that the response to broken loop safety injection can result in an increase in the calculated PCT, modeling of safety injection into the broken loop has now been incorporated into the NOTRUMP small break evaluation model.A more realistic model for condensation of steam by pumped safety injection based on data from the COSI test facility has also been incorporated, which provides a benefit larger than the penalty for safety injection in the broken loop.The methodology for modeling safety injection to the broken loop in small break LOCA analyses and application of the COSI condensation model are presented in the NOTRUMP Small Break ECCS Evaluation Model, Addendum 2"'.The analysis for 30%steam generator tube plugging modeled the pumped safety injection and an accumulator in the broken loop, and used the more realistic COSI condensation model in Reference 8.The results of the 3-inch break analysis are presented in the Sequence of Events Table 3.1-16 and the Results Table 3.1-17.Plots of the following parameters for the 3-inch break analysis are shown in Figures 3.1-77 through 3.1-85:-RCS pressure-Core mixture level-Hot Spot Clad Temperature
-Core Outlet Steam Flow-Hot spot rod surface heat transfer coefficient
-Hot spot fluid temperature
-Cold leg break mass flow rate-Broken loop safety injection mass flow rate, and-Lumped intact loop safety injection mass flow rate Figure 3.1-86 contains the power shape used in the analysis., mal 944-2w.wpf:1d441195 3.1-18
Due to the modeling of safety injection in the broken loop with the COSI condensation model change, in conjunction with the reduced peaking factors, the PCT for the 30%steam generator tube plugging small break LOCA analysis is lower than for previous small break analyses.Because no rod burst was calculated to occur and the beginning of life calculated peak clad temperature is low enough to preclude a Zr/H,O reaction temperature excursion following burst, no burst and blockage penalty is applied.The resulting total peak clad temperature of 1443'F is less than the 2200'F limit.Small Break LOCA Analysis Conclusions Analyses presented in this section show that the high head portion of the emergency core cooling system, together with the accumulators, provide sufficient core flooding to keep the calculated peak clad temperatures below required limits of 10 CFR 50.46.Hence, adequate protection is afforded by the emergency core cooling system in the event of a small break loss-of-coolant accident, References 1.Meyer, P.E.,"NOTRUMP-A Nodal Transient Small Break and General Network Code," WCAP-1 0079-P-A, August 1985.2.Lee, N.et.al.,"Westinghouse Small Break ECCS Evaluation Model Using The NOTRUMP Code," WCAP-1 0054-P-A, August 1985.3.Bordelon, F.M., et al~,"LOCTA-IV Program: Loss-of-Coolant Transient Analysis," WCAP-8305, June 1974, WCAP-8301, (Proprietary), June 1974.4."Report on Small Break Accidents for Westinghouse NSSS System,"Vols.I to III, WCAP-9600, June 1979.5."Clarification of TMI Action Plan Requirements," NUREG-0737, November 1980.6.NRC Generic Letter 83-35 from D.G, Eisenhut,"Clarification of TMI Action Plan Item II.K.3.31," November 2, 1983.7.Rupprecht, S.D., et.al.,"Westinghouse Small Break LOCA ECCS Evaluation Model Generic Study With the NOTRUMP Code;" WCAP-11145-P-A, October 1986.8.Thompson, C.M., et.al.,"Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection IntO the Broken Loop and COSI Condensation Model", WCAP-10054-P, Addendum 2 (Proprietary) and WCAP-10081-NP, Addendum 2 (Non-Proprietary)), August 1994.mal 944-2w.wpf:1d441195 3.1-19 TABLE 3.1-1 LARGE BREAK LOCA RESULTS Peak Clad Temperature
('F)Peak Clad Location{ft)Local Zr/H,O Reaction (Max%)Local Zr/H,O Location (ft)Total Zr/H,O Reaction (%)Case A C0=0.4 THOT 609.1'F P=2250 psia Min.Sl 2069 5.75 7.59 5.75 Case B Co=0.6 THOT 609.1'F P=2250 psia Min.Sl 1993 6.25 8.19 6.00<1.0 Case C CO=0.8 THQT 609.1'F P=2250 psia Min.Sl 1965 6.25 6,62 6.00<1.0 Case D C0=0.4 THOT 586.8'F P=2250 psia Min.Sl 2036'.00 8.45 6.00 (1.0 Case E C0=0.4 THOT 609.1'F P=2100 psia Min.Sl 2164 6.25 14.30 6.25<1.0 Case F C,=0.4 THOT 609.1'F P=2100 psia Max.Sl 2149 6.25 12.01 6.25<1.0 Hot Rod Burst Time (s)43.6 41.8 45.7 46.5 42.0 42.0 Hot Rod Burst Loc.(ft)5.75 6.00 6.00 6.00 6.25 6.25 m 31944.2w.wpf:
I d441195 3.1-20 A
TABLE 3.1-1 (continued)
LARGE BREAK LOCA RESULTS Start Reactor Trip Signal Safety Injection Signal Accumulator Injection Case A C,=0.4 THQT 609.1'F P=2250 psia Min.Sl 0.0 0.64 4.80 18.70 Case B Co=0.6 THOT-609.1'F P=2250 psia Min.Sl 0.0 0.64 4.60 13.90 Case C Co=0.8 THOT 609.1'F P=2250 psia Min.Sl 0.0 0.63 4.50 11.60 Case D CO=0.4 THOT-586.8'F P=2250 psia Min.Sl , 0.0 0.55 4.40 17.80 Case E Co=0.4 THQT-609.1'F P=2100 psia Min.SI 0.0~0.49 4.10 18.70 Case F Co=0.4 HOT 609.1'F P=2100 psia Max.SI 0.0 0.49 4.10 18.70 End of Blowdown 40.75 31.77 28.05 40.61 39.96 39.96 Pump Injection Bottom of Core Recovery 51.80 54.30 51.60 44.60 51.50 41.80 51.50 55.30 51.10 54.20 51.10 54.00 Accumulator Empty 69.09 62.30 48.75 70.09 68.96 69.78 m%1944.2w.wpf:1d441195 3.1-21 TABLE 3.1-2 PLANT INPUT PARAMETERS USED IN LARGE BREAK LOCA ANALYSIS Core Power (MWt)Peak Linear Power (kW/ft)Total Core Peaking Factor, F~Hot Channel Enthalpy Rise Factor, F,Maximum Assembly Average Power, P-Fuel Assembly Array Steam Generator Tube Plugging Level (%)Accumulator Water Volume (ft'/tank)
Accumulator Tank Volume (ft'/tank)
Minimum Accumulator Gas Pressure (psia)Accumulator Water Temperature
('F)Refueling Water Storage Tank Temperature
('F)Thermal Design Flowrate (gpm/loop)
RCS Loop Average Temperature
('F)Nominal Initial RCS Pressure (psia)Nominal Steam Pressure (psia)Safety Injection Delay Time (sec)RHR Pump Head Degradation
(%)HHSI Pump Head Degradation
(%)Charging Pump Head Degradation
(%)Charging Pump Flow Imbalance (gpm)RHR Cross-Tie Valve Position 102%of 3250 102%of 14.434 2.15 1.55 1.38 15 X 15 OFA 30 946 1350 600 100 70-105 83,200 553.0 and 576.3 2100 and 2250 595 and 749 47 15 15 10 25 Closed m%1944.2w.wpf:
1 d~1295 3.1-22 NET FREE VOLUME TABLE 3.1-3 LARGE BREAK CONTAINMENT DATA (ICE CONDENSER CONTAINMENT)(Includes Distribution Between Upper, Lower, and Dead-Ended Compartments)
Initial Conditions Pressure Maximum Temperature for the Upper, Lower, and Dead-Ended Compartments Minimum Temperature for the Upper, Lower, and Dead-Ended Compartments UC 746,829 ft'C 249 446 ft'E 116,168 ft IC 163,713 ft'4.7 psia UC 100'F LC 120'F DE 120'F UC 60'F LC 60'F DE 60'F RWST Temperature Temperature Outside Containment Initial Spray Temperature Spray System Runout Flow for a Spray Pump Number of Spray Pumps Operating Post-Accident Initiation of Spray System Distribution of Spray Flow to the Upper and Lower Compartments Deck Fan Post-Accident Initiation of Deck Fans Flow Rate per Fan Assumed Spray Efficiency of Water from Ice Condenser Drains 70oF 22oF 70oF 3600 gpm 36 sec LC 2700 gpm UC 4500 gpm 480 sec 43,890 cfm per fan 100%m:11944-2w.wpf:1d441195 3.1-23 TABLE 3.1-3 (continued)
STRUCTURAL HEAT SINKS 1 LC 2 LC 3 LC 4 LC LC LC LC 8 LC 9 LC 10 LC 11 LC 12 LC 13 UC 14 UC~Area tr'2,105 11,701 65,979 5,462 5,273 290'4,896 4,515 5,775 57,317 9,404 2,623 378 34,895 Thickness ft 0.0469/2.0 2.0 4.0 0.0833 0.0103 0.25 0.0078 0.1042 0.009 0.00833 0.0313 0.0313 0.0365/0.1667 0.0078 Material Steel/concrete Concrete Concrete Steel Steel Lead Steel Steel Steel Steel Steel Steel Stee Vconcrete Steel 15 16 17 18 19 UC UC UC UC UC 8,060 420 29,332 34,125 420 0.0208 0.0052 2.0 0.0469/2.0 0.0052 Steel Steel Concrete Steel/concrete Steel UC: Upper Compartment LC: Lower Compartment DE: Dead-Ended Compartment IC: Ice Compartment m:51944 2w.wpf:1d441195 3.1-24 TABLE 3 1-4 MASS AND ENERGY RELEASE RATES, MINIMUM SI Time sec 10 12 12.4 14 16 18 20 24 28 32 36 40 52 65 75 86 95 124 294 Mass Ibm/sec 57910 48870 33500 25260 22660 19580 16980 16000 14530 12140 10410 9170 7010 6750 5640 3580 4390 230 280 390 810 420 430 330 Ene BTU/sec 3.081(10)2.542(10')
1.762(10')
1.357(10')
1.223(10')
1.096(10)9.838(10)9.346(10)8.608(10')
7.313(10')
6,254(10')
5.472{10')
3.871(10')
2.839(10')
1.757(10')
7.951(10')
9.057(10')
1.267(10')
6.321(10')
2.073(10')
2.884{10')
2.464(10')
1.666(10')
1.452(10')
1.314(10')
m'A1944-2w.wpt:
1 d~1 195 3.1-25 TABLE 3.1-5 NITROGEN MASS AND ENERGY RELEASE RATES~Time eec 69.2 73.2 77.2 81.2 85.2 89.2 93.2 97.2 101.2 105.2 109.2 113.2 117.2 121.2 125.2 129.2 137.2 141.2 145.2 153.2 161.2 169.2 177.2 Flow Rate Ibm/sec 231.8 166.4 120.8 87.3 62.1 42.9 28.8 19.5 14.1 11.1 9.0 7.3 5.9 4.8 3.9 3.2 2.1 1.8 1.4 1.0 0.7 0.5 0.3 mal 944-2'.wpt:1d441195 3.1-26 TABLE 3.1-6 SAFETY INJECTION FLOW RATE RERATING PROGRAM ANALYSIS RCS PRESSURE (psia)415 515 615 715 815 915 1015 1115 1215 1315 1415 1515 HHSI FLOW (Ib/sec)20.48 19.38 18.18 16.89 15.54 14.09 12.50 10.46 7.99 4.48 0.00 0.00 CHARGING FLOW (Ib/sec)37.32 35.29 33.29 31.26 29.06 26.80 24.48 22.16 19.74 17.27 14.70 12.02 TOTAL FLOW (Ib/sec)57.80 54.67 51.47 48.15 44.60 40.89 36.98 32.62 27.73 22.05 1¹.70 12.02 m:51944-2w.wpf:
1 d441 195 3.1-27 TABLE 3.1-7 PLANT INPUT PARAMETERS USED IN SMALL BREAK LOCA ANALYSIS RERATING PROGRAM ANALYSIS Core Power Total Core Peaking Factor Steam Generator Tube Plugging Level 102%of 3588 MWt 2.32 15%(peak uniform)Accumulator Conditions:
Cover Gas Pressure Water Volume Total Volume 600 psia 946.0 ft 1350 ft RCS Initial Conditions:
Reduced Temperature, Reduced Pressure Case Loop Temperatures Consistent With T, Program Setpoint of, Pressure Vessel Flowrate 547OF 2100 psia 354000 gpm m%1944-2w.wpf:1d441295 3.1-28 TABLE 3.1-7 PLANT INPUT PARAMETERS USED IN SMALL BREAK LOCA ANALYSIS RERATING PROGRAM ANALYSIS Core Power Total Core Peaking Factor Steam Generator Tube Plugging Level 102%of 3588 MWt 2.32 15%(peak uniform)Accumulator Conditions:
Cover Gas Pressure Water Volume Total Volume 600 psia 946.0 ft'350 ft'CS Initial Conditions:
Reduced Temperature, Reduced Pressure Case Loop Temperatures Consistent With T, Program Setpoint of, Pressure Vessel Flowrate 547'F 2100 psia 354000 gpm m&1944-2w.wpl:1d~1195 3.1-28
TABLE 3.1-8 SMALL-BREAK LOCA CALCULATION RERATING PROGRAM ANALYSIS RESULTS PARAMETER VALUE Reduced Tem erature Reduced Pressure Break SIze: 2-Inch 3-Inch 4-Inch Peak clad temperature
('F)Elevation (ft)1899 2122 12.00 12.00 1414 11.25 Zr/H,O cumulative reaction Maximum local (%)Elevation (ft)Total core (%)7.16 7.70 0.25 12.00 12.00 11.50<0.3<0.3<0.3 Rod Burst None None None CALCULATION:
NSSS Power MWt 102%of Peak Linear Power kW/ft 102%of Hot Rod Power Distribution (kW/ft)Accumulator Water Volume, cu.ft.3588 16.426 See Figure 3.1-21 946 Does not include pump heat.mA1944-2w.wpf:1d441195 3.1-29 TABLE 3.1-8 SMALL-BREAK LOCA CALCULATION RERATING PROGRAM ANALYSIS PARAMETER RESULTS VALUE Reduced Tem erature Reduced Pressure Break Size: 2-Inch 3-Inch 4-Inch Peak clad temperature
('F)Elevation (ft)Zr/H,O cumulative reaction Maximum local (%)Elevation (ft)1899 12.00 7.16 12.00 2122 1414 12.00 11.25 7.70 0.25 12.00 11.50 Total core (%)Rod Burst (0.3 None (0.3 None (0.3 None CALCULATION:
NSSS Power MWt 102%of Peak Linear Power kW/ft 102%of Hot Rod Power Distribution (kW/ft)Accumulator Water Volume, cu.ft.3588 16.426 See Figure 3.1-21 946 Does not include pump heat.m'31944-2w.wpf:1d441295 3.1-29 TABLE 3.1-9 TIME SEQUENCE OF EVENTS FOR CONDITION III EVENTS RERATING PROGRAM ANALYSIS Small-Break Loss of Coolant Accident Event Time (s)Reduced Tem erature Reduced Pressure Break SIze: 2-Inch 3-Inch 4-Inch Break occurs Reactor trip signal Safety injection signal Start of safety injection delivery Loop seal venting 25.37 36.54 63.54 1634.4 17.10 44.10 10.74 37.74 652.1 420A 11.24 6.85 Loop seal core uncovery Loop seal core recovery Boil-off core uncovery Accumulator injection begins Peak clad temperature occurs Top of core covered Sl flow exceeds break flow N/A N/A 2216.7 N/A 4143.8 N/A 4587.5 645.8 680.3 1045.7 1711.5 1958.7 N/A 2197.1 424.6 439.2 696.5 901.0 969.5 1982.7 N/A m:II1 944-2NI'.WPf:1d441295 3.1-30 TABLE 3.1-9 TIME SEQUENCE OF EVENTS FOR CONDITION III EVENTS RERATING PROGRAM ANALYSIS Small-Break Loss of Coolant Accident Event Time (s)Reduced Tem erature Reduced Pressure Break Size: 2-Inch 3-Inch 4-Inch Break occurs Reactor trip signal Safety injection signal Start of safety injection delivery Loop seal venting 25.37 36.54 63.54 1634.4 11,24 17.10 44.10 652.1 420.4 6.85 10.74 37.74 Loop seal core uncovery Loop seal core recovery N/A 645.8 424.6 N/A 680.3 439.2 Boil-off core uncovery Accumulator injection begins Peak clad temperature occurs Top of core covered Sl flow exceeds break flow 2216.7 N/A 1711.5 1045.7 1958.7 4587.5 2197.1 4143.8 N/A N/A 696.5 901.0 969.5 1982.7 N/A mal 944.2w.wpf:
1 d441195 3.1-30 TABLE 3.1-10 SMALL-BREAK LOCA CALCULATION RERATING PROGRAM ANALYSIS RESULTS PARAMETER VALUE High Temp.Reduced Temp.Hi h Pressure Hi h ressure 3-Inch 3-Inch Peak clad temperature
('F)Elevation (ft)Zr/H 0 cumulative reaction aximum local (%)Elevation (ft)Total core (%)Rod Burst 1756 11.75 1.99 11.75<0.3 None 1887 11.75 4.64 11.75<0.3 None CALCULATION'SSS Power MWt 102%of Peak Linear Power kW/ft 102%of Hot Rod Power Distribution (kW/ft)Accumulator Water Volume, cu.ft.3588 16.426 See Figure 3.1-21 946 Does not include pump heat.m (1944-2w.wpf:1d441295 3.1-31 TABLE 3.1-10 SMALL-BREAK LOCA CALCULATION RERATING PROGRAM ANALYSIS RESULTS PARAMETER VALUE High Temp.Reduced Temp.Hi h Pressure Hi h ressure 3-Inch 3-Inch Peak clad temperature
('F)Elevation (ft)Zr/H 0 cumulative reaction aximum local (%)Elevation (lt)Total core (%)Rod Burst 1756 11.75 1.99 11.75<0.3 None 1887 11.75 4.64 11.75<0.3 None CALCULATION:
NSSS Power MWt 102%of Peak Linear Power kW/ft 102%of Hot Rod Power Distribution (kW/ft)Accumulator Water Volume, cu.ft.3588 16.426 See Figure 3.1-21 946 Does not include pump heat.m%1 944 2)N.wpf:1d~1195 3.1-31 TABLE 3.1-11 TIME, SEQUENCE OF EVENTS FOR CONDITION III EVENTS RERATING PROGRAM ANALYSIS Small-Break Loss of Coolant Accident Event Time (s)High Temp.Reduced Temp.Hi h Pressure Hi h Pressure Break occurs Reactor trip signaI Safety injection signal Start of safety injection delivery Loop seal venting Loop seal core uncover Loop seal core recovery Boil-off core uncovery Accumulator injection begins Peak clad temperature occurs Top of core covered Sl flow exceeds break flow 3-Inch 19.03 29.74 51.74 666.96.N/A N/A 1070.4 1672.0 1793.7 N/A 2022.0 3-Inch 15.97 20.95 47.95 698.78 691.54 726.85 1166.8 1855.2 1986.2 N/A 2282.7 mh1 944-2IN.WPf:1d~1195 3.1-32 TABLE 3.1-11 TIME SEQUENCE OF EVENTS FOR CONDITION III EVENTS RERATING PROGRAM ANALYSIS Small-Break Loss of Coolant Accident Event Time (s)'igh Temp.Reduced Temp.Hi h Pressure Hi h Pressure Break occurs Reactor trip signal Safety injection signal Start of safety injection delivery Loop seal venting Loop seal core uncovery Loop seal core recovery Boil-off core uncovery Accumulator injection begins Peak clad temperature occurs Top of core covered Sl flow exceeds break flow 3-Inch 19.03 29.74 51.74 666.96 N/A N/A 1070.4 1672.0 1793.7 N/A 2022.0 3-Inch 15.97 20.95 47.95 698.78 691.54 726.85 1166,8 1855.2 1986.2 N/A 2282.7 mh1 944-2w.wpf:1dM1295 3.1-32
TABLE 3.1-12 PLANT INPUT PARAMETERS USED IN SMALL BREAK LOCA ANALYSIS k3%MAIN STEAM SAFETY VALVE SETPOINT TOLERANCE ANALYSIS Core Power (MWt)Peak Linear Power (kW/ft)Total Core Peaking Factor, F~Hot Channel Enthalpy Rise Factor, F,Maximum Assembly Average Power, PAxial Offset (%)Fuel Assembly Array Steam Generator Tube Plugging Level (%)Accumulator Water Volume (ft'/tank)
Accumulator Tank Volume (ft'/tank)
Minimum Accumulator Gas Pressure (psia)Accumulator Water Temperature
('F)Refueling Water Storage Tank Temperature
('F)Thermal Design Flowrate (gpm/loop)
RCS Loop Average Temperature
('F)Nominal Initial RCS Pressure (psia)Nominal Steam Pressure (psia)Safety Injection Delay Time (sec)HHSI Pump Head Degradation
(%)Charging Pump Head Degradation
(%)Charging Pump Flow Imbalance (gpm)HHSI Cross-Tie Valve Position Auxiliary Feedwater Total Flowrate (gpm)102%of 3250 102%of 14.921 2.32 1.55 1.433+30 15 X 15 OFA 15 946 1350 600 130 120 88,500 547.0 2100 607 27 10 10 25 Closed 750 m 61944-2w.wpf:1d~1295 3.1-33 TABLE 3.1-12 PLANT INPUT PARAMETERS USED IN SMALL BREAK LOCA ANALYSIS k3%MAIN STEAM SAFETY VALVE SETPOINT TOLERANCE ANALYSIS Core Power (MWt)Peak Linear Power (kW/ft)Total Core Peaking Factor, Fa Hot Channel Enthalpy Rise Factor, F,Maximum Assembly Average Power, PAxial Offset (%)Fuel Assembly Array Steam Generator Tube Plugging Level (%)Accumulator Water Volume (ft'/tank)
Accumulator Tank Volume (ft'/tank)
Minimum Accumulator Gas Pressure (psia)Accumulator Water Temperature
('F)Refueling Water Storage Tank Temperature
('F)Thermal Design Flowrate (gpm/loop)
RCS Loop Average Temperature
('F)Nominal Initial RCS Pressure (psia)Nominal Steam Pressure (psia)Safety Injection Delay Time (sec)HHSI Pump Head Degradatlon
(%)Charging Pump Head Degradation
(%)Charging Pump Flow Imbalance (gpm)HHSI Cross-Tie Valve Position Auxiliary Feedwater Total Flowrate (gpm)2100 10 130 120 88,500 547.0 607 27 10 750 Closed 102%of 3250 102%of 14.921 2.32 1.55 1.433+30 15 X 15 OFA 15 946 1350 mal 944.2w.wpf:1d441195 3.1-33 TABLE 3.1-13 TIME SEQUENCE OF EVENTS FOR CONDITION III EVENTS%MAIN STEAM SAFETY VALVE SETPOINT TOLERANCE ANALYSI Small-break Loss of Coolant Accident Time (s)Event Reduced Pressure, Reduced Temperature 3-Inch 2-Inch Break Occurs Reactor trip signal Safety injection signal Start of safety injection Start of auxiliafy feedwater delivery Loop seal venting Loop seal core uncovefy Loop seal core recovefy Boil-off core uncovery Accumulator injection begins Peak clad temperature occurs Top of core covered Sl flow rate exceeds break flow rate 0.0 8.64 17.13 44.13 68.6 592 N/A N/A 1680 1890 N/A 1890 0.0 19.03 37.11 64.11 79.1 1390 N/A N/A 2312 N/A 4042 N/A 4091 Main steam safety valve setpoint tolerance increase case at 3250 MWt core power.m%1944-2IN.WPf:1d441195 3.1-34 TABLE 3.1-13 TIME SEQUENCE OF EVENTS FOR CONDITION III EVENTS%3%MAIN STEAM SAFETY VALVE SETPOINT TOLERANCE ANALYSIS Small-break Loss of Coolant Accident Time (s)'vent Reduced Pressure, Reduced Temperature 3-Inch 2-Inch Break Occurs Reactor trip signal Safety injection signal Start of safety injection Start of auxiliary feedwater delivery Loop seal venting Loop seal core uncovery Loop seal core recove1y Boil-off core uncovery Accumulator injection begins Peak clad temperature occurs Top of core covered Sl flow rate exceeds break flow rate 0.0 8.64 17.13 68.6 592 N/A'680 1890 N/A 1890 0.0 19.03 37.11 64.11 79.1 1390 N/A N/A 2312 N/A 4042 N/A 4091'ain steam safety valve setpoint tolerance increase case at 3250 MWt core power.m&1944.2W.Wpf:
1 d~1195 3.1-34 TABLE 3.1-14 SMALL-BREAK LOCA CALCULATIONS
%MAIN STEAM SAFETY VALVE SETPOINT TOLERANCE ANALYSIS RESULTS NOTRUMP Peak Clad Temperature
('F)Peak Clad Temperature Location (ft)Peak Clad Temperature Time (sec)Local Zr/H,O Reaction Maximum (%)12.0 12,0 1890 4042 5.06 3.75 Reduced Pressure, Reduced Temperature 3-Inch 2-Inch 1951 1833 Local Zr/H,O Reaction Location (ft)Total Zr/H,O Reaction (%)Rod Burst Burst and Blockage Penalty ('F)Total Peak Clad Temperature
{F)12.0<1.0 None 117 2068 12.0<1.0 None 15 1848 Main steam safety valve setpoint tolerance increase case at 3250 MWt core power.m%1944-2W.wpf:1d44'f f 95 3.1-35
TABLE 3.1-15 PLANT INPUT PARAMETERS USED IN SMALL BREAK LOCA ANALYSIS 30%SGTP PROGRAM ANALYSIS WITH HHSI CROSS TIES CLOSED Core Power (MWt)Peak Linear Power (kW/ft)Total Core Peaking Factor, F~Hot Channel Enthalpy Rise Factor, F,Maximum Assembly Average Power, PAxial Offset (%)Fuel Assembly Array Steam Generator Tube Plugging Level (%)Accumulator Water Volume (ft'/tank)
Accumulator Tank Volume (ft'/tank)
Minimum Accumulator Gas Pressure (psia)Accumulator Water Temperature
('F)Refueling Water Storage Tank Temperature
('F)Thermal Design Flowrate (gpm/loop)
RCS Loop Average Temperature
('F)Nominal Initial RCS Pressure (psia)Nominal Steam Pressure (psia)Safety Injection Delay Time (sec)HHSI Pump Head Degradation
(%)Charging Pump Head Degradation
(%)Charging Pump Flow Imbalance (gpm)HHSI Cross-Tie Valve Position Auxiliary Feedwater Total Flowrate (gpm)102%of 3250 102%of 14.12 2.32 1.55 1.38+20 15 X 15 OFA 30 946 1350 600 130 120 83,200 553.0 2100 595 47 15 10 25 Closed 750 m:$1944-2w.wpf:1dM1 195 3.1-36
TABLE 3.1-15 PLANT INPUT PARAMETERS USED IN SMALL BREAK LOCA ANALYSIS 30%SGTP PROGRAM ANALYSIS WITH HHSI CROSS-TIES CLOSED Core Power (MWt)Peak Linear Power (kW/ft)Total Core Peaking Factor, Fa Hot Channel Enthalpy Rise Factor, F,Maximum Assembly Average Power, PAxial Offset (%)Fuel Assembly Array Steam Generator Tube Plugging Level (%)Accumulator Water Volume (ft'/tank)
Accumulator Tank Volume (ft'/tank)
Minimum Accumulator Gas Pressure (psia)Accumulator Water Temperature
('F)Refueling Water Storage Tank Temperature
('F)Thermal Design Flowrate (gpm/loop)
RCS Loop Average Temperature
('F)Nominal Initial RCS Pressure (psia)Nominal Steam Pressure (psia)Safety Injection Delay Time (sec)HHSI Pump Head Degradation
(%)Charging Pump Head Degradation
(%)Charging Pump Flow Imbalance (gpm)HHSI Cross-Tie Valve Position Auxiliary Feedwater Total Flowrate (gpm)102%of 3250 102%of 14.12 2.32 1.55 1.38+20 15 X 15 OFA 30 946 1350 600 130 120 83,200 553.0 2100 595 47 15 10 25 Closed 750 m 31944-2w.wpf:1d441295 3.1-36 TABLE 3.1-16 TIME SEQUENCE OF EVENTS FOR CONDITION III EVENTS 30%SGTP PROGRAM ANALYSIS WITH HHSI CROSS-TIES CLOSED Small-break Loss of Coolant Accident Event Break occurs Reactor trip signal Safety injection signal Start of safety injection Start of auxiliary feedwater delivery Loop seal venting Loop seal core uncovefy Loop seal core recovery Boil-off core uncovery Accumulator injection begins Peak clad temperature occurs Top of core recovered Combined pumped Sl flow rate exceeds break flow rate Time (s)Reduced Pressure, Reduced Temperature 3-Inch 0.0 8.8 17.4 64.4 88.8 528 N/A N/A 1054 1648 1748 2995 1856 30%steam generator tube plugging case at 3250 MWt core power.m&1944-2w.wpf:1d~1295 3.1-37 TABLE 3.1-16 TIME.SEQUENCE OF EVENTS FOR CONDITION III EVENTS 30%SGTP PROGRAM ANALYSIS WITH HHSI CROSS-TIES CLOSED Small-break Loss of Coolant Accident Event Time (s)Reduced Pressure, Reduced Temperature 3-Inch Break occurs Reactor trip signal Safety injection signal Start of safety injection Start of auxiliary feedwater delivery Loop seal venting Loop seal core uncovery Loop seal core recovery Boil-off core uncovery Accumulator injection begins Peak clad temperature occurs Top of core recovered Combined pumped Sl flow rate exceeds break flow rate 88.8 0.0 8.8 17.4 64.4 528 N/A N/A 1054 1648 1748 2995 1856 30%steam generator tube plugging case at 3250 MWt core power.m:11944-2w.wpf:1d441195 3.1-37 TABLE 3.1-17 SMALL-BREAK LOSS OF COOLANT ACCIDENT CALCULATIONS 30%SGTP PROGRAM ANALYSIS WITH HHSI CROSS-TIES CLOSED RESULTS Reduced Pressure, Reduced Temperature 3-Inch NOTRUMP Peak Clad Temperature
('F)Peak Clad Temperature Location (ft)Peak Clad Temperature Time (sec)Local Zr/H,O Reaction Maximum (%)Local Zr/H,O Reaction Location (ft)Total Zr/H,O Reaction (%)Rod Burst Burst and Blockage Penalty Total Peak Clad Temperature
('F)1443'F 11.5 1748 (1.0 11.5 (1.0 None None 1443'F 30%steam generator tube plugging case at 3250 MWt core power.mh1 944-2w.wpf:1d441295 3.1-38 TABLE 3.1-17 SMALL-BREAK LOSS OF COOLANT ACCIDENT CALCULATIONS 30%SGTP PROGRAM ANALYSIS WITH HHSI CROSS-TIES CLOSED RESULTS Reduced Pressure, Reduced Temperature 3-Inch NOTRUMP Peak Clad Temperature
('F)Peak Clad Temperature Location (ft)Peak Clad Temperature Time (sec)Local Zr/H,O Reaction Maximum (%)Local Zr/H,O Reaction Location (ft)Total Zr/H,O Reaction (%)Rod Burst Burst and Blockage Penalty Total Peak Clad Temperature
('F)1 1.5 1443'F 11.5 1748 (1.0 (1.0 None None 1443'F 30%steam generator tube plugging case at 3250 MWt core power.m:51944-2w.wpf:1d441195 3.1-38
2500 2000 1500 1000 500 0 0 20 30 TIME (S)40 50'gure 3.1-1a Reactor Coolant System Pressure Case A, CD=0.4, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 I%1944-2w.wpf:1d~1195 3.1-39 2500 2000 1500 LLI 1000 LIJ 500 10 15 20 TIME (S)25 30 35 igure 3.1-1b Reactor Coolant System Pressure Case B, CD=0.6, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 m&1944-2w.wpf:1
&)41 195 3.140
250D 2000 150D 1000 10 15 20 TIME (S)25 30 igure 3.1-1 c Reactor Coolant System Pressure Case C, CD=0.8, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 mh1 944-2w.wpf:1d~1195 3.1<1 250D 2000 1500 1000 500 10 2D 30 TIME (S)40 50 igure 3.1-1 d Reactor Coolant System Pressure Case D, CD=0,4, Thot=586.8'F, P=2250 psia Donald C.Cook Unit 1 mA1 944-2w.wpf:1d~1195 3.1-42
2500 2000 1500 1000 500 10 20 TIME (S)30 igure 3.1-1 e Reactor Coolant System Pressure Case E, CD=0.4, Thot=609.1'F, P=2100 psia Donald C.Cook Unit 1 m&1944-2w.Wpf:1d441195 3.1' 2500 2000 1500 LU 1000 500 20 TlhlE (S)igure 3.1-1 f Reactor Coolant System Pressure Case A, CD=0.4, Thot=609.1'F, P=2100 psia, max Sl Donald C.Cook Unit 1 m&1944-2w.wpf:1d441195 E 60000 50000<0000 30000 LIJ 20000 4 10000 10 20 30 TIME (S)40 50 igure 3.1-2a Break Flow During Blowdown Case A, CD=0.4, Thot=609.1'F, P=2250 psia Donald C, Cook Unit 1 m&1944-2w.wpf:1d441195 3.145 70000 60000 50000 40000 I-30000 C7 20000 10000)0 15 20 25 TIME (S)50 35 Figure 3.1-2b Break Flow During Blowdown Case B, CD=0.6, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 mA1944.2w.wpf:1d441195 3.1<6 SOOOO CD 4J M CQ 60000 40000 20000 10 15 20 TIME (S)25 30 Figure 3.1-2c Break Flow During Blowdown Case C, CD=0.8, That=609.1'F, P=2250 psia Donald C.Cook Unit 1 m:51944-2w.wpf:1dM1195 3.1-47 70000 60000 50000 LJ 40000 30000 C)20000$0000 10 20 3D TIME (S)Figure 3.1-2d Break Flow During Blowdown Case D, CD=0.4, Thot&86.8'F, P=2250 psia Donald C.Cook Unit 1 mal 944-2W.wpf:1d441195 3.1<8 60000 50000 40000 30000 20000 10000 10 20 T I ME (S)Figure 3.1-2e Break Flow During Blowdown~~Case E, CD=0.4, Thot=609.1'F, P=2100 psia Donald C.Cook Unit 1 mh1 944-2w.wpt:1dC41195 3.149
60000 50000<0000 30000 20000 10000 10 20 TIME (S)30 Figure 3.1-2f Break Flow During Blowdown Case F, CD=0.4, Thot=609.1'F, P=2250 psia, max Sl Donald C.Cook Unit 1 m%1944.2w.wpf:1d~1195 3.1-50
20 10 C/)cn]0 CL-20 I-4J-30 4 4 C5-40-50 10 20 30 TIME (S)40 50 Figure 3.1-3a Core Pressure Drop Case A, CD=0.4, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 m 61944-2w.wpf:1d441195 3.1-51 10 0-50 10 15 20 25 TIME (S)30 Figure 3.1-3b Core Pressure Drop Case B, CD=0.6, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 mA1 944-2w.wpf:1d441195 3.1-52' 10 CL-10 LIJ CL C/l-20 LLl LX CL-30 I-LQ-40 L4-50-60 10 15 20 TIME (5)25 Core Pressure Drop Case C, CD=0.8, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 m%1944-2w.wpf:1d441195 3.1-53 40 20 CA CL-60 10 20 30 T I ME (S)Figure 3.1-3d Core Pressure Drop~~Case D, CD=0.4, Thot=586.8'F, P=2250 psia Donald C.Cook Unit 1 m (1944-2w.wpf:1dM1195 3.1-54 20 10 I-QJ 20 4J 4-30 10 20 TIME (S)30 Figure 3.1-3e Core Pressure Drop Case E, CD=0.4, Thot=609.1'F, P=2100 psia Donald C.Cook Unit 1 mA1944.2w.wpf:1 d~1195 3.1-55
20 10 0 C/)n--10 I-QJ 20 4J 4-30-40 10 20 TIME (S)30 Figure 3.1-3f Core Pressure Drop Case F, CD=0.4, That=609.1'F, P=21 00 psia, max Sl Donald C.Cook Unit 1 m&1944-2w.wpf:1d441195 3.1-56
-CORE INLET~CORE OUTLET 40000 30000 20000 10000-10000 10 20 30 TIME (S)40 50 Figure 3.1<a Core Flowrate Case A, CD=0.4, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 m&1944-2w.wpt:1d~1195 3.1-57
-CORE INLET~CORE OUTLET 40000 30000 20000 10000-10000-20000 10 15 20 TIME (S)25 30 35 Figure 3.1Mb Core Flowrate Case B, CD=0.6, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 m%1944-2w.wpf:1d~1195 3.1-58
-CORE INLET a a CORE OUTLET 40000 30000 20000 10000 CO-10000-20000-30000 10 15 20 TIME (S)-25 30 Figure 3.1<c Core Flowrate Case C, CD=0.8, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 mal 944-2w.wpf:1d441195 3.1-59
-CORE INLET~CORE OUTLET 40000 30000 20000 10000-10000 10 20 30 TIME (S)40 50 Figure 3.1<d Core Flowrate Case D, CD=0.4, Thot&86.8'F, P~O psia Donald C.Cook Unit 1 m&1944-2w.wpf:1d~1195 3.1-60 I'Ilail IIII 4%t I I I I t4f~~~I~~~I'tl~~~.~~~~~~
-CORE INLET~CORE OUTLET 40000 30000 20000 10000-10000 10 20 TIME (S)30 Figure 3.1<f Core Flowrate Case F, CD=OA, Thot=609.1'F, P=2100 psia, max Sl Donald C.Cook Unit 1 mA1 944-2w.wpf:1d~1195 3.1-62 6000 5000~4000 3000 2000 1000 10 20 30 TIME (S)40 50 e Figure 3.1-5a Accumulator Flow During Blowdown Case A, CD=0.4, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 m&1944-2w.wpf:1d~1195 3.1-63 6000 5000~<000 3000 I-2000 1000 10 15 20 25 TIME (S)30 Figure 3.1-5b Accumulator Flow During Blowdown Case B, CD=0.6, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 m:11944-2w.wpf:1d~1195 3.1-64 6000 5000 4000 3000 2000 1000 10 15 20 TIME (S)25 30 Figure 3.1-5c Accumulator Flow During Blowdown Case C, CD=0.8, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 mA1944.2w.wpf:1dM1195 3.1-65 5000 4000 3000 I-2000 1000 10 20 30 TIME (S)40 50 Figure 3.1-5d Accumulator Flow During Blowdown Case D, CD=0.4, Thot=586.8'F, P=2250 psia Donald C.Cook Unit 1 m&1944-2w.wpf:1d~1 195 3.1-66 6000 5000~<000 3000 2000 1000 20 TIME (S)30 Figure 3.1-5e Accumulator Flow During Blowdown Case E, CD=0.4, Thot=609.1'F, P=2100 psia Donald C.Cook Unit 1 m 81944-2w.wpt:1d~1195 3.1-67 6000 5000~<000 m 3000 2000 1000 10 20 TIME (5)30 Figure 3.1-5f Accumulator Flow During Blowdown Case F, CD=0.4, Thot=609.1'F, P=2100 psia, max Sl Donald C.Cook Unit 1 mA1 944-2w.wpf:1d~1195 3.1-68 CORE IllXTURE LEVEL OUENCH FRONT LOCATION~DOWNCOIIER LEVEL 20 u 15 10 50 1 0 150 2 0 250 TIME AFTER REFLOOD (S)3 0 Figure 3.1-6a Vessel Liquid Levels During Ref lood Case A, CD=0.4, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 m&1944-2w.wpf:1d441195 3.1-69
-CORE MIXTURE LEVEL~QUENCH FRONT LOCATION~OOWNCOMER LEVEL 20 15 10 0 0 50 1 0 150 200 250 TtME AFTER REFLOOD (S)300 Figure 3,1-6b Vessel Liquid Levels During Ref lood Case B, CD=0.6, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 mh1 944-2w.wpf:1d441195 3.1-70
-CORE MIXTURE LEVEL a-a OUENCH FRONT LOCATION~OOWNCOMER LEVEL 25 20 I-u 15 10 50 1 0 150 2 0 250 TIME AFTER REFLOOD (S)3 0 Figure 3.1-6c Vessel Liquid Levels During Ref lood Case C, CD=0.8, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 mA1 944-2w.wpf:1d441195 3.1-71 CORE MIXTURE LEVEL~QUENCH FRONT LOCAT'ION~OOWNCOIIER LEVEL 20 I-10 50 1 0 1 0 2 0 250 TIME AFTER REFLOOD (S)3 0 Figure 3.1-6d Vessel Liquid Levels During Ref lood Case D, CD=0.4, Thot=586.8'F, P=2250 psia Donald C.Cook Unit 1 mA1 944-2w.wpf:1dM1195 3.1-72 CORE MIXTURE LEVEL~QUENCH FRONT LOCATION~OOWNCOMER LEVEL 25 20 15 10 50 1 0 150 20 250 TIME AFTER REFLOOO ($)300 Figure 3.1-6e Yessel Liquid Levels During Ref lood Case E, CD=0.4, Thot=609.1'F, P=2100 psia Donald C.Cook Unit 1 mh1 944-2W.Wpt:1 d441195 3.1-73 CORE MIXTURE LEVEL e-a QUENCH FRONT LOCATION~DOWNCOIIER LEVEL 25 20 50 1 0 150 2 0 250 TIME AFTER REFLDDD (S)Figure 3.1-6f Vessel Liquid Levels During Ref load Case F, CD=0.4, That=609.1'F, P=2100 psia, max Sl Donald C.Cook Unit 1 m&1944-2w.Wpf:1d~1195 3.1-74 1~1 F 9 LLJ I 0'C9 CD o 0~7 o 4 0'0'0'50 100 150 2 TIME AFTER REFLOOD (S)250 Figure 3.1-7a Core Inlet Flow During Reflood Case A, CD=0.4, Thot=609,1'F, P=2250 psia Donald C.Cook Unit 1 m%1944-2w.wpf:1d441195 3.1-75 Cf)0'I-0'C9 CD o 0'CO 4 0'C)CO 0'0'50 0 150 2 0 TIME AFTER REFI.OOD (S)250 Figure 3.1-7b Core Inlet Flow During Ref lood Case B, CD=0.6, Thot=609.1'F, P=2250 psia Donald C.Cook Unit i m%1944-2w.wpf:1d~1195 3.1-76 0'CQ C)o 0'C)4 F 6 C)tD 0'0'50 0 150 20 TIME AFTER REFLOOD (S)Figure 3.1-7c Core Inlet Flow During Reflood Case C, CD=0.8, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 m%1944-2w.wpt:1d~1195 3.1-77 1~2 F 9 C9 z,'~8 C)C)0'o F 6 C)0'A@50 1 0 150 200 TIME AFTER REFLOOD (S)250 Figure 3.1-7d Core Inlet Flow During Ref load Case D, CD=0.4, Thot=586.8'F, P=2250 psia Donald C.Cook Unit 1 m 31944-2W.wpf:1d~1195 3,1-78 0'OI C9 0 AS C5 C)0'I-o 0.6 C)0~5 0~4 50 1 0 150 200 TIME AFTER REFLOOD (S)250 Figure 3.1-7e Core Inlet Flow During Reflood Case E, CD=0.4, Thot=609.1'F, P=2100 psia Donald C.Cook Unit 1 mA1 944-2w.wpf:1d~1195 3.1-79 1~2 CQ 0~8 CD CD CD 4 0.6 CD 0~4 50, 1 0 150 2 0 ()TIME AFTER REFLOOD S 250 Figure 3.1-7f Core Inlet Flow During Ref lood Case F, CD=0.4, That=609.1'F, P=2100 psia, max Sl Donald C.Cook Unit 1 m:(1944-2w.wpt:1dM1195 3.1-80 5000 4000 LIJ 3000 2000 1000 0 50 1 0 150 2 0 2 0 3 0 TIME AFTER REFLOOD S Figure 3.1-8a Accumulator and Sl Flow During Ref lood Case A, CD=0.4, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 mA1 944-2w.wpf:1d441195 3.1-81 5000 4000 LLI 3000 LIJ I-2000 1000 50 0 1 0 20 0 250 300 TIME AFTER REFLOOD S Figure 3.1-8b Accumulator and Sl Flow During Ref lood Case B, CD=0.6, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 mA1 944-2w.wpt:1d441195 3.1-82 5000 4000 3000 2000 1000 50 1 0 150 2 0 250 TIME AFTER REFLOOD (S)300 Figure 3.1-8c Accumulator and Sl Flow During Ref lood Case C, CD=0.8, Thot=609.1 F, P=2250 psia Donald C.Cook Unit 1 m%1944-2w.wpf:1d441195 3.1-83 5000 4000 3000 2000 1000 50 1 0 150 2 0 250 300 T I ME AFTER REF LOOD S Figure 3.1-8d Accumulator and Sl Flow During Ref lood Case D, CD=0.4, Thot=586,8'F, P=2250 psia Donald C.Cook Unit 1 mA1 944-2w.wpf:1dM1195 3.1-84 5000 4000 3000 2000 1000 50 1 0 150 2 0 250 TIME AFTER REFLOOD (S)Figure 3.1-8e Accumulator and Sl Flow During Ref lood Case E, CD=0.4, Thot=609.1'F, P=2100 psia Donald C.Cook Unit 1 m31 944.2w.wpf:1d441195 3.1-85
'50,00 4000 3000 2000 1000 0 0 50 100 150 200 250 TIME AFTER REFLOOD (S)300 Figure 3.1-8f Accumulator and Sl Flow During Ref lood o Case F, CD=0.4, Thot=609.1'F, P=2100 psia, max Sl Donald C.Cook Unit 1 m%1944-2w.wpf:1d441195 3.1-86 CD CV 15 I I 10 50 10 150 20 20 TIME AFTER REFLOOD (S)Figure 3.1-9a Integral of Core Inlet Flow~~Case A, CD=0.4, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 m 31 944-2w.wpf:1 d~1195 3.1-87 20 10 LK 0 I 50 1 0 150 200 250 TIME AFTER REFLOOD (S)Figure 3.1-9b Integral of Core Inlet Flow Case B, CD=0.6, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 m&1944-2w.wpf:1d441195 3.1-88' 20 C)C4 15 I I-10 CL 0 I-50 1 0 150 200 250 TIME AFTER REFLOOD (S)300 Figure 3.1-9c Integral of Core Inlet Flow Case C, CD=0.8, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 m%1944-2w.wpf:1d~1195 3.1-89 20 tD z 15 I 10 CQ LD C)o 5 CL CP 0 I-50 10 10 20 20 TIME AFTER REFLOOD (S)3 0 Figure 3.1-9d Integral of Core Inlet Flow Case D, CD=0.4, Thot=586.8'F, P=2250 psia Donald C.Cook Unit 1 m:41 944-2W.wpf:1d~1 195 3.1-90 20 C)CV x 15 I 50 1 0 150 2 0 250 TIME AFTER REFLOOD (S)300 e Figure 3.1-9e Integral of Core Inlet Flow Case E, CD=0.4, Thot=609.1'F, P=2100 psia Donald C.Cook Unit 1 m51 944.2w.wpf:1dM1195 3,1-91 20 I 10 C9 CD CD o 5 CL 0 I-50 1 0 150 2 0 250 TIME AFTER REFLOOD (S)300 e Figure 3.1-9f Integral of Core Inlet Flow Case F, CD&.4, Thot=609.1'F, P=2100 psia, max Sl Donald C.Cook Unit 1 mh1 944-2w.wpt:1dM1199 3.1-92 800 600 400 X 200-200 50 1 0 150 2 0 250 T.l ME (5)0 350 Figure 3.1-10a Mass Flux at Peak Temperature Elevation Case A, CD=0.4, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 m&1944-2w.wpf:1d441195 3.1-93 SOO 600 400 OC 200-200 50 00 150 TIME (S}250 3 D Figure 3.1-10b Mass Flux at Peak Temperature Elevation Case B, CD=0.6, Thot=609.1'F, P=2250 psia Donald C, Cook Unit 1 mh1 944-2w.wpt:1d441195 3.1-94 800 600 400 200-200 50 10 150 20 TIME (S)250 3 0 Figure 3.1-10c Mass Flux at Peak Temperature Elevation Case C, CD=0.8, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 mA1 944-2w.wpf:1d441195 3.1-95 800 600 I-400 X 200-200 50 10 150 2 0 250 TIME (S)30 30 Figure 3.1-10d Mass Flux at Peak Temperature Elevation Case D, CD=0.4, Thot=586.8'F, P=2250 psia Donald C.Cook Unit 1 m 51944-2w.wpf:1d~1 195 3.1-96 800 600 400 X 200-200 50 1 0$50 2 0 250 3 0 350 TIME (S)Figure 3.1-10e Mass Flux at Peak Temperature Elevation Case E, CD=0.4, Thot=609.1'F, P=2100 psia Donald C.Cook Unit 1 mh1 944-2w.wpf:1d441195 3.1-97
800 600 I-400 X 200-200 50 100 150 2 0 250 300 350 T IME S Figure 3.1-1 Of Mass Flux at Peak Temperature Elevation Case F, CD=0.4, Thot=609.1'F, P=2100 psia, max Sl Donald C.Cook Unit 1 m%1944-2w.wpf:1d441195 3.1-98 i l f 4J Cf)~10 I 4J$0 50 100 150 200 250 TIME (S)300 350 Figure 3.1-11a Rod H.T.C.at Peak Temperature Elevation Case A, CD=0.4, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 mh1944<w.wpf:1d~1195 3.1-99 10 LaJ CO~10 I--'I$0 0 50 100 150 200 TIME (S)250 300 Figure 3.1-1 1b Rod H.T.C.at Peak Temperature Elevation Case B, CD=0.6, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 m:L19444w.wpf:1dM1195 3.1-100 10~)0 I--1 10 0 50 100 150 200 TIME (S)250 300 Figure 3.1-1 1c Rod H.T.C.at Peak Temperature Elevation Case C, CD=0.8, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 mh19444w.wpf:1d~1195 3.1-1 01 10 CQ~10 10 0 50 100 150 2DD 250 TIME (S)300 350 Figure 3.1-11d Rod H.T.C.at Peak Temperature Elevation Case D, CD=0.4, Thot&S6,8'F, P=2250 psia Donald C.Cook Unit 1 m&1 9444w.wpt:1dM1195 3.1-1 02 10~10~10 LJ C7 1 o 10 4J Cf)~10-1 10 0 50 100 150 200 250 TIME (S)300 350 Figure 3.1-11e Rod H.T.C.at Peak Temperature Elevation Case E, CD=0.4, Thot=609.1'F, P=2100 psia Donald C.Cook Unit 1 mh1944<w.wpf:1d441195 3.1-103 4 10~10 cn~10 4~~10 C/7~10 I-10 50 100 150 200 250'00 TIME (S)350 Figure 3.1-1 1f Rod H.T.C.at Peak Temperature Elevation Case F, CD&.4, Thot=609.1'F, P=2100 psia, max Sl Donald C.Cook Unit 1 mA1944<w.wpt:1d441195 3.1-104 1800 1600 1400 1200 LLJ 1000 800 600 400 200 5010102 0 2 0 300 350 ()TIME S Figure 3.1-1 2a Vapor Temperature Case A, CD=0.4, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 m:519444w,wpf:1d~1195 3.1-105 1800 1600 1400 1200 1000 800 IJJ 600 400 200 50 0 150 TIME (S)2 0 250 300 Figure 3.1-1 2b Vapor Temperature Case B, CD=0.6, That=609.1'F, P=2250 psia Donald C.Cook Unit 1 mh19444w.wpf:1d~1 195 3.1-106
1800 1600 1400 1200 1000 800 600 400 200 50 1 0 150 2 0 TIME (S)250 300 Figure 3.1-1 2c Vapor Temperature Case C, CD=0.8, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 mA19444w.wpf:1d441195 3.1-107 1800 1600 1400 1200 10DO 800 600 400 200 50 1 0 150 200 250 3 0 350 TIME (S)Figure 3.1-1 2d Vapor Temperature Case D, CD=0.4, Thot=586.8'F, P=2250 psia Donald C.Cook Unit 1 mh19444w.wpf:1d441195 3.1-108 2000 1500 LJ 1000 LJ 500 50 1 0 150 200 250 3 0 350 TtME (5)Figure 3.1-1 2e Vapor Temperature Case E, CD=0.4, Thot=609.1'F, PM100 psia Donald C.Cook Unit 1 m%1 944<w.wpf:1d~1 195 3.1-109 f i/i 2000 1500 1000 500 50 1 0 150 2 0 250 300 350 TIME S Figure 3.1-1 2f Vapor Temperature Case F, CD=0.4, Thot=609.1'F, P=2100 psia, max Sl Donald C.Cook Unit 1 mh1944<w.wpt:1dM1195 3.1-110 2200 2000 1800 1600 1 1400 1200 1000 800 600 50 1 0 150 2 0 250 TIME (S)30 0 350 Figure 3.1-1 3a Fuel Rod Peak Clad Temperature Case A, CD=0.4, That=609.1'F, P=2250 psia Donald C.Cook Unit 1 mh1944<w.wpf:1d441195 3.1-1 11 2000 1800 1600 1400 1200 I-1000 800 600 50 100 150 200 TIME (S}250 300 Figure 3.1-1 3b Fuel Rod Peak Clad Temperature Case B, CD=0.6, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 m:L19444w.wpf:1d441195 3 1-112 2000 1800 1600 1400 1200 I-1000 800 600 50 150 TIME (S)200 250 Figure 3.1-1 3c Fuel Rod Peak Clad Temperature Case C, CD=0.8, Thot=609.1'F, P=2250 psia Donald C.Cook Unit 1 mh1944<w.wpf:1d441195 3,1-113 2200 2000 1800 1600 1400 1200 I-1000 800 600 50 1 0 150 2 0 2 0 TIME (S)3 0 350 Figure 3.1-1 3d Fuel Rod Peak Clad Temperature Case D, CD=0.4, Thot&86.8'F, P=2250 psia Donald C.Cook Unit 1 mh1944<w.wpf:1d441195 3.1-114 5~4 l5 II 2200 2000 1800 1600 1400 LJ 1200 I-1000 800 600 50 1 0 1 0 2 0 250 3 0 350 TIME (5)Figure 3.1-13e Fuel Rod Peak Clad Temperature Case E, CD=0.4, Thot=609.1'F, P&1OO psia Donald C.Cook Unit 1 mh1944<w.wpf:1d441195 3.1-<<S
2200 200D 1B00 1600 1400 120D LLI 1000 800 600 50 0 150 2 0 250 3 0 TIME (S)350 Figure 3.1-1 3f Fuel Rod Peak Clad Temperature Case F, CD&.4, Thot=609.1'F, P=2100 psia, max Sl Donald C.Cook Unit 1 m:L19444w.wpf:1 d441195 3.1-1 16
-UPPER COIIPARTMEHT
~LOWER COIIPARTMEHT 50 10 20 TIME (SEC)250 300 Figure 3.1-14 Containment Pressure CD=0.4, Min Sl Donald C.Cook Unit 1 m 51944<w.wpf:1d~1 195 3.1-117
14000 12000>0000 I SOOO O.C)6000<000 2000 50 0 1 0 2 TlhlE (SEC)250 Figure 3.1-15 Upper Compartment Structural Heat Removal Rate CD=0.4, Min Sl Donald C.Cook Unit 1 m319444w.wpf:1dM1195 3.1-118 10 CD CA~10 C)S~10 4$0 50 100$50 200 TIME (SEC)250 300 Figure 3.1-16 Lower Compartment Structural Heat Removal Rate CD=0.4, Min Sl Donald C.Cook Unit 1 mh19444w.wpf:1d~1195 3.1-119
$60000 140000 120000 100000 80000 60000 O.tD 40000 2000D-2DOOD 50 100 150 2 0 TIME (SEC)2 0 3 0 Figure 3.1-17 Heat Removal by Sump CD=0.4, Min Sl Donald C.Cook Unit 1 mal 944<w.wpf:1d~1185 3.1-120
100000 80000 m 60000 I 40000 O LIJ 20000 50 101020 T I ME (SEC)250 3 0 Figure 3.1-18 Heat REmoval by Lower Compartment Spray CD=0.4, Min Sl Donald C.Cook Unit 1 I&1 9444w.wpf:1dM1195 3.1-121 UPPER COMPARTMENT a a LOWER COMPARTMENT 250 200 150 LIJ I-100'0 50 1 0 150 2 0 TIME (SEtl)250 300 Figure 3.1-1 9 Containment Temperature CD=0.4, Min SI Donald C.Cook Unit 1 mh1944<w.wpf:1d~
1 195 3.1-122 60 50 io 5 30 20 10 400 600 800 1000 PRESSURE (PRA)1200 1ioo 1600 Figure 3.1-20 Safety Injection Flow Rate Donald C.Cook Unit 1 mA1944<w.wpf:
1 d441195 3.1-123 18.0 16.0 14.0 12~0 10.0 F 0 l l I 6~0 4~0 2~0 0'2'4'6.0 8-0 ZLVTLTION (PT)Figure 3.1-21 Hot Rod Power Distribution Donald C.Cook Unit 1 m L1944<w.wpt:1d~1195 3.1-124 2288.2888.1888.K 7c)1688.4J g 1488.K~1288.~1888.888.688.488@1888, 1588.TINE (SEC l 2888.2688.Figure 3,1-22 RCS Pressure (3 Inch}Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mA1944<w.wpf:1d441195 3.1-125 f
36.34.32.I 4 w 28.LJ)Ld u 26.OC I z 24~4J OC C)22.TOP QF CORE 28.18.16@588.1888.'1588.TINE (SEC)2888.Figure 3.1-23 Core Mixture Height (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mA1944<w.wpf:1d441195 3.1-126
~2588.2888.Cl 1588.o~)888.L5 K 5.588.~~688.$88.1888.1288~$488.l688.l888.2888.2288.2488.26BB.TlHE tSEC)Figure 3.1-24 Hot Spot Clad Temperature (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 m%1 944<w.wpf:1d441195 3.1-127
'258.Lal z 288.E CD~158.CI w l 88~CD l888.1588.TINE(SEC)2888.2688.Figure 3.1-25 Core Steam Flowrate (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mA1 944<w.wpf:1d~1195 3.1-12B 18>cD 182 4 cn 181 188 688.888.1888.1288.1488, 1688.1888.2888.2288.2488.2688.TlfC (SECf Figure 3.1-26 Hot Spot Heat Transfer Coefficient (3 inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mA1944<w.wpf:1d441 195 3.1-129 4~2588.CI-2888.1588.y 1888.o 588.S.688.SBS.1888.1288.1488.1688.ISSS.2888.2288.2488.2688.TltK 1 SEC)Figure 3.1-27 Hot Spot Fluid Temperature (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mh19444w.wpf:1d441195 3.1-130 1B88.1688.1488.~1288.c 1888.B88.688.488.288.258.588.758.1888.1258.1588.1758.2888.2258.2588.TINE (SEC)Figure 3.1-28 Total Break Flow (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mal 9444w.wpf:1d441195 3.1-131 45,~~48.X:~55.o~38.D~25.x: 0 a.28.C)C)iS.18.2SB.588.7SB.1888.1258.1588.1758.2888.2258.2588.TINE (SEC)Figure 3.1-29 Intact Loop Pumped Sl Flow (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mA1944<w.wpf:1d441195 3.1-132 2288.2888.-1888.4J P)1688.K f)1488.+1288.188gl.888@1888.2888.5888.4888.5888.6888.TINE (SEC)Figure 3.1-30 RCS Pressure (2 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 m:51944<w.wpf:1d~1195 3.1-133 56.34, 52.38.I 4 a 28.4J LJ~26.OC 2 24 4l C)47 22.TOP OF CORE 28.18.1888.2888.5888.4888.TINE (SEC)5888.6888.Figure 3.1-31 Core Mixture Height (2 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mh19444w.wpf:1dM1195 3.1-134 4~2588.2888.Cl Cl 1588.~1888.588.R S.2888.2588.5888.5588.48BB.4588.5888.5588.6888..T1NE t SEC)Figure 3.1-32 Hot Spot Clad Temperature (2 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mA1944<w.wpf:1dM1195 3.1-135
168.148.CJ 4J z 128.Z g 188 4 C)88.I X 4J 68.C)48.28.1888.2888.5888.4888.5888.TINE t SEC)Figure 3.1-33 Core Steam Flowrate (2 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mA1944<w.wpf:1dM1195 3.1-1 36 0
CV cD]82 cn]8]2888.2588.5888.5588.4888.4588.5888.5588.6888.7 lHE l SEC)Figure 3.1-34 Hot Spot Heat Transfer Coefficient (2 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 m:11944<w.wpf:1dM1195 3.1-137
~2588.2888.l588.5 tZ o.1BSS.CI B.2888.2588.5888.5588..4888.l588.SISS.5588.6888.'Tlirt (SECl Figure 3.1-35 Hot Spot FLuid Temperature (2 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mA1S44<w.wpf:1d441195 3.1-138 888.788, 688.K w 588.288.188.1888.2888.3888.4888.5888.TIME (SEC)Figure 3.1-36 Total Break Flow (2 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mA19444w.wpf:1d441195 3.1-139 45.48.CD 4J CO~38.C7 L 25.C7 a 28.8 15.18.2888.3888.4888.TINE t SEC)5888.6888.Figure 3.1-37 Intact Loop Pumped Sl Flow (2 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 m%1944-3w.wpf:1d441195 3.1-140 2288.2888.'888.~1688.ac 1488.a.1288, K~~1888.688.688.488.288~588.1888.1588.TINE f SEC)2588.Figure 3.1-38 RCS Pressure (r Inch)Reduced Temperature, Reduced Pressure Donafd C.Cook Unit 1 mh1944<w.wpf:1d~1195 3.1-141 56.54.52.4 a 28.4l LLJ u 26.GC I x: 24.LLl OC C)CJ 22.TOP OF CORE 28.18'6@588.1888'.1588.TINE f SEC)2588.Figure 3.1-39 Core Mixture Height (4 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mh1944<w.wpf:1d~1195 3.1-142
~2588.-2888.Cl Cl 1588.CI~1888, L5 I: O 5 588.e.i88.688.888.1888.1288.fi88.1688.1888.2888.TlNE lSEC1 Figure 3.1-40 Hot Spot Clad Temperature (4 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mA1944<w.wpf:1dM1195 3.1-143 275.258.225.-288.LJ g 175.~~168.cr 125.~l88.CI 75.25.588.i888.1588.TiME (SEC)2588.F igure 3.1<1 Core Steam Flowrate (4 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mA1944<w.wpt:1d~1 195 3.1-144 Cg 4.cD l82 Ch C/l tX I o co l8l CI I I Cf.l88 488.$88.888.l888.l 288.1488.T 1NC 1 SEC I l688, l888, 2888.Figure 3.1-42 Hot Spot Heat Transfer Coefficient (4 inch)e Reduced Temperature, Reduced Pressure Oonald C.Cook Unit 1 mA1944<N.Wpl:1d441195 3.1-145 4~2588.D 2888.1588.5 a.1888.K O 588.8.488.688.888.1888.1288.1488.168B.1888.2888.71NE t SEC)Figure 3.1-43 Hot Spot Fluid Temperature (4 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 m&1 944<w.wpt:1dM1195 3.t-<46 2488.2288.2888'1888.ar.1688~a-1488.X~~1288.~1888.688.488@1888.1688.2888.2688.5888.TINE (SEC)Figure 3.1-44 RCS Pressure (3 Inch)Reduced Temperature, Reduced Pressure Donaid C.Cook Unit 1 m:51 944<w.wpf:1d~f195 3.1-147 36.34, 32, I 4 a 28.4J Ll~26.OC I OC E 24'J C)47 22.TOP 0 F CORE 28.js.i888.1588.2888.2588.TINE (SEC)Figure 3.1-45 Core Mixture Height (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 m%1 944<w.wpf:1d441195 3.1-148 cI, 2SSB.2888.Cl CD 1SSS.R~1888.5 SSB.688.888." 1888.1288.1488.1688.1888.2888.2288.2488.2888.Tlat%1SKC)Figure 3.1<6 Hot Spot Clad Temperature (3 Inch)Reduced Temperature, Reduced Pressure Donald C, Cook Unit 1 m&1944<w.wpf:1d 441195'3.1-149 258.225.288'17S.X-158.o 4~125.CL~188.LJ 5 CJ 25.588.1888.1588.2888.TINE (SEC)Figure 3.1<7 Core Steam Flow Rate (3 Inch)~~Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mA19444w.wpf:1d441195 3.1-150 58~Ck I OJ 4 582 tC W CI cA 585 I 688.888.l888.5288.5488.l688.l688.2888.2288.2488.2688.TlNE lSECl Figure 3.1-48 Hot Spot Heat Transfer Coefficient (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mA19444w.wpf:1d441195 3.1-161 I l I I t
~2588.2888.1588.gg~I CI a 1888.K I O SSS.QSS.SSS.1888..1288.1488.1688.1888.2888.2288 2488.2688.71NE tSEC)Figure 3.1<9 Hot Spot Fluid Temperature (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mh19444w.wpt:td441195 3.1-152 1888.1688.1488.~1288.~1888.hC 888.688.488.288.S88.1888.1588.2888.2S88.5888.TINE (SEC)Figure 3.1-50 Total Break Flow (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 m:51944<w.wpf:1d~1195 3.t-153 55.58.45.48, CD K w 35.C)~S8.CD Cl~25.X: CL a 28.C)C)~~~5.i8.588.1888.1588.2888.TIME (SEC)Figure 3.1-51 Intact Loop Pumped Sl Flow (3 Inch)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 m:11944<w.wpf:1d441195 3.1-154 0
2488, 2288.2888.~1B88.ac 1688.~1488.K~~1288.1888.888.688.488~1888.1S88.2888.2S88.5888.TINE t SEC)Figure 3.1-52 RCS Pressure (3 Inch)High Temperature, High Pressure Donald C.Cook Unit 1 m:51944<w.wpf:1d~1195 3.1-155
36.34, 32.38.I 4 a 28.4J 4J 4 26.I~24.4J C)22.TOP OF CORE 28.18.16@1888.1588.2888.TINE (SEC)Figure 3.1-53 Core Mixture Height (3 Inch)High Temperature, High Pressure Donald C.Cook Unit 1 m519444w.wpf:1d441 195 3.1-156
5888.~2888.4J 4J 4l Cl 2888.o 1688.F888.5.688.l888.f288.1488.3688~ISSS.2888.2288.2488.2688.2888.SBSS.TlNE (SEC)Figure 3.1-54 Hot Spot Clad Temperature (3 Inch)High Temperature, High Pressure Donald C.Cook Unit 1 m51944<w.wpf:1d441195 3.1-157 h r ll H f 228.288.188.-168.CJ~148.~~128.cr 188.I 88.4J C7 88.48.28.588.1888.1588.2888.2588.TINE (SEC)Figure 3.1-55 Core Steam Flowrate (3 Inch)High Temperature, High Pressure Donald C.Cook Unit 1 m:L1944<w.wpt:1d441195 3.1-158
le>OC AJ I cD 182 CL CI cn 1 el CZ lee 1888.1288.1488.l688.1888.2888.2288.2488.2688.2888.5888.TllK tSEC)Figure 3.1-56 Hot Spot Heat Transfer Coefficient (3 Inch)High Temperature, High Pressure Donald C.Cook Unit 1 mal 944<w.wpt:1 d~1 195 3.1-159 4.~2588.Ch 2888.1588.y 1888.la)P 588.1888..1288.1488.1688.1QSS.2888.2288.2488.2888.2888.5888.Tll1C lSECl Figure 3.1-57 Hot Spot Fluid Temperature (3 Inch)High Temperature, High Pressure Oonald G.Gook Unit 1 mh1 944-3w.wpf:1d441195 3.1-160 1688.1488.1288.K g 1888.o 888.g 688.I 488.288.588.1888.1588.2888.TINE (SEC)2S88.5888, Figure 3.1-58 Total Break Flow (3 Inch)High Temperature, High Pressure Donald C.Cook Unit 1 mA1 9444w.wpf:1dC41195 3.1-161 I,
58.45.~48.K z KS.o~S8.o~~2S.X: lL a 28.C)CI 15.28.1888.1588.2888.2588.TINE (SEC)Figure 3.1-59 Intact Loop Pumped Sl Flow (3 Inch)High Temperature, High Pressure Donald C.Cook Unit 1 mh1944<w.wpf:1d~1195 3.1-162 2DC3 15CO LJ lJl lOCO 500 0 1000 2000 3000 TIME (SEC)Figure 3.1-60 RCS Pressure (3 Inch, 3%MSSV Tolerance)
Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 4000 50(mA1944<w.wpf:1d441195 3.1-163 4~30 oc 25 Top of Cnre 20 15 0 f000 2000 3000 TIME (SEC)~000 5CC Figure 3.1-61 Core Mixture Level (3 Inch, 3%MSSV Tolerance)
Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 m%1 944<w.wpf:1dC41195 3.1-164
'8" G.16"0.1400.Ll 12".0.I OC 4J K)000.I 800.600.400 500.1000.1500.2000.(s)2500.3000.350 Figure 3.1-62 Peak Clad Temperature (3 Inch, 3%MSSV Tolerance)
Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 m%1 944<w.wpf:1d~1195 3.1-165
(
53 100 50 0 0 1000 2000 3000 TIME (SEC)4000 5COO Figure 3.1-63 Core Outlet Steam FLow Rate (3 Inch, 3%MSSV Tolerance)
Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 m L1944<w.wpf:1d441195 3.1-166
oc 103)01 500.1000.i500.2000.2500.T:ME: (S)3000.35GG Figure 3.1-64 Hot Spot Rod Surface Heat Transfer Coefficient (3 Inch, 3%MSSV Tolerance)
Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mh1944<w.wpf:1d441195 3.'I-167 2ODO.1800.1600.1400.1200.K 1000.SOO.600.400 500.1000.1500.2000.2500.3000.TIME (S)Figure 3.1-65 Hot Spot FLuid Temperature (3 Inch, 3%MSSV Tolerance)
Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mh1944<w.wpf:1d 441195 3.1-168
>5CO E CD~tOCO 500 0 0 1000 2000 3000-IMK (SEC)4000 Figure 3.1-66 Cold Leg Break Mass Flow Rate (3 Inch, 3%MSSY Tolerance)
Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 m%1 944<w.wpf:1d~1195 3.1-169 6: Vi>20 0 0 1000 2000 3000 TIME (SEC)4000 5CGC e Figure 3.1-67 Safety Injection Mass Flow Rate (3 Inch, 3%MSSV Tolerance)
Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mA1944<w.wpf:1d441195 3.1-170 14 12 10 o'8 CL 4 6 8 Elevation (ft)Figure 3.1-68 Hot Rod Power Distribution Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 m%1 944<w.wp1:1d441195 3.1-171 2200 2000 1800 1600 1400 1200 1000 800 600 1000 2000 3000 4000 5000 6000 7000 TIME (SEC)Figure 3.1-69 RCS Pressure (2 Inch, 3%MSSV Tolerance)
Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mh1944<w.wpf:1d441195 3.1-172 40 LLJ 30 25 I Top of Co 20 15 1000 2000 3000 4000 5000 6000 7000 T I ME (SEC)Figure 3.1-70 Core Mixture Level (2 Inch, 3%MSSV Tolerance)
Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 m%1 944<w.wpf:1d441195 3.1-173
2000 1800 1600 1400 1200 1000 800 600 400 2000 3000 4000 5000 TIME (S)6000 7000 1 Figure 3.1-71 Peak Clad Temperature (2 Inch, 3%MSSV Tolerance)
Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mA1944<w.wpt:1d441195 3.1-174
200 CA cn 150 100 o 50-50 1000 2000 3000 4000 5000 6000 7000 TIME (SEC)Figure 3.1-72 Core Outlet Steam Flow Rate (2 Inch, 3%MSSY Tolerance)
Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mht9444w.wpf:td44t t95 3.1-175
5 10 I 4 L 10 I i LZI~10 UJ CO V 2 10 W I-10 0 10 2000 3000 4000 5000 TIME (S)6000 7000 Figure 3.1-73 Hot Spot Rod Surface Heat Transfer Coefficient (2 Inch, 3%MSSV Tolerance)
Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 m%1 944<w.wpf:1d441195 3.1-176 1800 1600 1400 1200 I-1000 LU I 800 600 400 2000 3000 4000 5000 TIME (S)6000 7000 Figure 3.1-74 Hot Spot Fluid Temperature (2 Inch, 3%MSSV Tolerance)
Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mA19444w.wpf:1dM1195 301 1 77 700 600 500 m 400 300 C)200 100 1000 2000 3000 4000 5000 6000 7000 TIME (SEC)Figure 3.1-75 Cold Leg Break Mass Flow Rate (2 Inch, 3%MSSV Tolerance)
Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 m:119444w.wpf:1d441195 3.1-178 50 40 30 o 20 10 1000 2000 3000 4000 5000 6000 7000 TIME (SEC)Figure 3.1-76 Safety Injection Mass Flow Rate (2 Inch, 3%MSSV Tolerance)
Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 m319444w.wpf:1d441195 3.1-1 79
2200 2000 1800 1600 1400 4J 1200 LaJ 1000 800 600 400 1000 2000 TIME (S}3000 4000 Figure 3.1-77 RCS Pressure (3 Inch, 30 Io SGTP)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mA19444W.wpf:1d441195 3.1-180 40 35 LaJ 4J 30 LJ 25 I 20 15 0 1000 2000 TIME (S)3000 4000 Figure 3.1-78 Core Mixture Level (3 Inch, 30%SGTP)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mh19444w.wpf:1d441195 3.1-181 1600 1400 l200 1000 LJ 800 600 400 1000 1500 2000 2500 TIME (Sj 3000 3500 t Figure 3.1-79 Hot Spot Clad Temperature (3 Inch, 30%SGTP)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 m&1 944<w.wpf:1d
~1195 3.1-182 250 200 150 o 100 0 0 l000 2000 TIME (Sj 3000 4000 Figure 3.1-80 Core Outlet Steam Flow (3 Inch, 30%SFTP)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mA1944<wwpf:1d441 195 3.1-183 5 10 I 4 1 10 I CQ 3~10 C)V 2 10 1 i-10 0 10 1000 150Q 2000 2500 TIME (Sj 3000 3500 Figure 3.1-81 Hot Spot Rod Surface Heat Transfer Coefficient (3 Inch, 30%SGTP)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mA19444w.wpf:1dM1195 3.1-184 1100 1QQQ 900 800 I-700 CL I-600 500 400 0 1000 2000 TIME (S)3000 4000 I Figure 3.1-82 Hot Spot Fluid Temperature (3 Inch, 3%SGTP)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mA19444w.wpf:1d441195 3,1-185 1600 1400~cn 1200 1000 4J I 800 CO 600 400 200 1000 2000 TIME (Sj 3000 4000 Figure 3.1-83 Cold Leg Break Mass Flow Rate (3 inch, 30%SGTP)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 m%1 944<w.wpf:1d~1 195 3.1-186 35 30 25 4J I CL 20 15 CD 4 cn I 0 1000 2000 TIME (S)3000 4000 Figure 3.1-84 Broken Loop Safety Injection Mass Flow Rate (3 Inch, 30%SGTP)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 mh1944<w.wpf:1d~1195 3.1-1 87 60 50 40 20 10 1000 2000 TIME (S)3000 4000 Figure 3.1-85 Lumped Intact Loop Sl Mass Flow Rate (3 Inch, 30%SGTP)Reduced Temperature, Reduced Pressure Donald C.Cook'Unit 1 m%1 9444w.wpf:1d441195 3.1-188 3.2 LOCA HYDRAULIC FORCES LOCA hydraulic forces are relatively insensitive to specific steam generator tube plugging levels and the associated changes in thermal design flow, provided the RCS temperatures remain unchanged.
The LOCA hydraulic forces analyzed for the Rerating Program are documented in Section 3.2 of WCAP-11902 and used conservative values of 582.3'F and 511.7'F for Tand T~, respectively.
The LOCA hydraulic forcing functions in Section 3.2 of WCAP 11902 conservatively bound the RCS parameters in Table 2.1-1 of this report, even with consideration of 5%asymmetric flow.WCAP-11902, Section 3.2, remains valid and no changes and/or additions are required.m%1 944<w.wpf:1d441195 3.2-1 3.3 NON-LOCA ANALYSES 3.3.1 Introduction This section evaluates the effects of reduced temperature and pressure operation with a maximum average SGTP level of 30%for Donald C.Cook Nuclear Plant Unit 1 with respect to the non-LOCA safety analyses.The effort performed is to support Unit 1 operation with a core power of 3250 MWt in the range of full-power reactor vessel average temperatures between 553'F and 576.3'F at primary pressure values of 2100 psia or 2250 psia (Cases 1 and 2 of Table 3.3-1, which are identical to cases 2 and 3 of Table 2.1-1).The current non-LOCA analyses of record for Unit 1 support a rerated core thermal power of 3411 MWt (3425 MWt NSSS)with a full power vessel average temperature between 547'F and 578.7'F at a primary system pressure of 2100 psia or 2250 psia.Cases 3 and 4 of Table 3.3-1 present the range of conditions supported by the current non-LOCA analyses of record.It.is important to note that the current non-LOCA safety analyses of record support the rerating of Donald C.Cook Nuclear Plant Unit 1.However, the unit has never been licensed to operate in accordance with the parameters defined as Cases 3 and 4 of Table 3.3-1.The Donald C.Cook Nuclear Plant Unit 1 licensing basis, as reported in the UFSAR (Reference 14)includes analyses and evaluations of sixteen non-LOCA events, which are delineated on the next two pages.This licensing-basis has been reviewed to assess the impact associated with the SGTP Program.The following events were~re-anal zed as part oi the SGTP Program: Unit 1 UFSAR Section Accident 14.1.1 14.1.2 Uncontrolled RCCA Bank withdrawal from a Subcritical Condition Uncontrolled RCCA Bank withdrawal At Power 14.1.3 14.1.4 14.1.6 14.1.8 14.2.5 Rod Cluster Control Assembly Misalignment RCCA Drop Loss of Reactor Coolant Flow (Including Locked Rotor)Loss of External Electrical Load and/or Turbine Trip Rupture of a Steam Pipe (core response analysis)m:11944<w.wpt:1d441195 3.3-1
~,
14.2.6 Rupture of a CRDM Housing (Rod Ejection)14.3.4.4 Mass and Energy Release Analysis for Postulated Secondafy System Pipe Ruptures Inside Containment.
The following events/analyses have been evaluated to support the operating conditions associated with the SGTP program: Unit 1 UFSAR Section Accident 14.1.5 14.1.7 14.1.9 Chemical and Volume Control System Malfunction Start-up of an Inactive Loop Loss of Normal Feedwater 14.1.10 14.1.11 Excessive Heat Removal Due to Feedwater System Malfunctions Excessive Load Increase Incident 14.1.12 Loss of All AC Power to the Plant Auxiliaries 14.2.8 (Unit 2)Rupture of a Main Feedwater Pipe 14.4.11.3 Steamline Break Mass/Energy Release Outside Containment 3.3.2 Non-LOCA Safety Analysis Assumptions Requiring Technical Specification Changes To enhance operating flexibility for RCS reduced temperature and pressure operation with a maximum average steam generator tube plugging of 30%, certain reactor protection system setpoints were revised.The following requirements were relaxed to enhance operating flexibility as well: EDG start time from ambient conditions; pressurizer code safety valve setpoint tolerance; and shutdown margin for T,~greater than 200'F.The revised RPS setpoints include the overtemperature bT (OTET)and the overpower AT (OPAT)reactor trips.The general equations for the OTAT and OPBT reactor trip setpoints and the safety analysis limit coefficient values are presented in Table 3.3-3: A detailed The 30%SGTP program also included an evaluation of the Major Rupture of a Feedwater Pipe event (UFSAR Section 14.2.8), which is not part of the Unit 1 licensing basis and is provided for informational purposes only.mA1944<w.wpf:1d441195 3.3-2 discussion of the revised setpoint equations for these reactor trip functions is provided in Section 3.3.2.1.Discussions of the EDG start time requirement, the pressurizer code safety valve setpoint tolerance adjustment, and shutdown margin relaxation are also presented in the sections that follow.The applicable Technical Specification updates for these revisions/relaxations are provided in Appendix A.3.3.2.1 Reactor Protection System Trip Setpoints Revised OTET and OPAT setpoints are based upon new core thermal safety limits, which account for the effects of the RCS parameter changes associated with the increased level of steam generator tube plugging, using the methodology described in Reference 1.These setpoints were revised to increase the available margin between the safety analysis setpoint values and the nominal, or Technical Specification values, such that more hT-drift could be accommodated between instrumentation calibrations during the fuel cycle.Presently, the power margin associated with the Rerating Program is being utilized to offset the bT-drift that is being experienced during core bumup (i.e., the core power of 3411 MWt is supported by the analyses, but the plant is actually operated with a core full-power value of 3250 MWt).However, since the 30%SGTP parameters do not have this power margin available, there was a need to revise the OTBT and OPET setpoints as part of the SGTP Program.Figures 3.3-1 through 3.3-4 present the allowable reactor coolant loop average temperature and bT conditions as a function of primary coolant pressure, based upon a minimum measured flow (MMF)of 339,100 gpm and a 1.55 chopped cosine axial power distribution.
Figure 3.3-1 represents the most limiting 30%SGTP operating configuration (nominal full-power T,~=576.3'F, nominal pressure=2100 psia)of the range of conditions described in Table 3.3-1 (Cases 1 and 2)for the calculation of the OTLT and OPET setpoints.
The boundaries of operation defined by the OTBT and OPIT trips are represented as"protection lines" on this diagram.The protection lines are drawn to include all adverse instrumentation and setpoint errors so that under nominal conditions, a trip would occur well within the area bounded by these lines.The utility of this diagram is in the fact that the limit imposed by any given DNBR can be represented as a line.The DNB lines represent the locus of conditions for which the DNBR equals the limit value (1.40 and 1.42 for typical and thimble cells, respectively; see Table 3.12-3).All points below and to the left of a DNB line for a given pressure have a DNBR greater than the Safety Analysis Limit DNBR value.The diagram shows that DNB is prevented for all cases if the area enclosed with the maximum protection lines is not traversed by the applicable Safety Analysis Limit DNBR at any point.The area of permissible operation (power, pressure, and temperature) is bounded by the combination of reactor trips: high neutron flux (fixed setpoint);
high and low RCS pressure (fixed setpoints);
overpower and overtemperature hT (variable setpoints), and the opening of m:51944<w.wpf:Id~1195 313 3 t ll the steam generator safety valves, which limit the maximum RCS average temperature.
The Safety Analysis Limit DNBR value (1.40 typical and 1.42 thimble), which was used as the DNBR limit for all accidents analyzed with the Revised Thermal Design Procedure (RTDP;Reference 2), is conservative compared to the actual Design Limit DNBR value (1.23 and 1.22 for typical and thimble cells, respectively), required to meet the DNB design basis.Table 3.3-2 presents the limiting trip setpoints assumed in the accident analyses and the time delay values assumed for each trip function.The difference between the limiting trip point assumed for the analysis and the normal trip point represents an allowance for instrumentation channel error and setpoint error.Nominal trip setpoints are specified in the plant Technical Specifications.
During plant start-up tests, it is demonstrated that actual instrument time delays are equal to or less than the assumed values.Additionally, protection system channels are calibrated and instrument response times are determined periodically in accordance with the Technical Specifications.
The 30%SGTP effort assumed that the reference average temperatures (T'nd T")used in the OTAT and OPAT setpoint equations are scaled to the full-power average RCS temperature each time the cycle average temperature is changed.It is also assumed that the reference pressure (P')in the OTZT equation is set equal to the appropriate nominal primary system pressure for a particular cycle (either 2100 psia or 2250 psia).These assumptions are key to ensure that the actual plant conditions required to result in an OTET and/or OPAT trip signal to be generated are conservative with respect to assumptions made in the safety analyses.Figures 3.3-1 through 3.3-4 illustrate the OTET and OPAT protection setpoints for the endpoints of the range of full-power vessel average temperatures for the SGTP Program at either 2100 psia or 2250 psia.The calibration of the NIS excore detectors, to compensate for the changes in coolant density each time the cycle operating conditions are changed, is also assumed in the analyses.The OTAT and OPbT reactor trip functions provide primary protection against fuel centerline melting, among other concerns (i.e., DNB and hot-leg boiling).The criterion for no fuel melt is, the uranium dioxide melting temperature shall not be exceeded for at least 95 percent of the limiting fuel rods at a 95 percent confidence level (Reference 1).This criterion is met by limiting the calculated fuel centerline temperature to 4700'F (valid for 60000 MWD/MTU bumup per Reference 16).In many cases, fuel centerline melting can be prevented by limiting gross core thermal power to a prescribed limit (historically 118%of nominal power)independent of axial power distribution.
As part of the reload process (via the Reload Safety Analysis Checklist, or RSAC), the peak linear heat generation rate of the core (i.e., peak kw/ft)is determined specifically for fuel centerline melting concerns.Even though the revised OTZT and OPET reactor trip setpoint equations allow the typical gross core average thermal power to slightly exceed the historical value of 118%(Cook Unit 1-analyses indicate that a peak overpower of 119.03%can be achieved with the revised setpoints), fuel centerline melt m%1 944<w.wpf:1d441195 3.3A l7 L f concerns are specifically evaluated on a cycle-by-cycle basis as part of the formal reload process to ensure fuel centerline melting does not occur.Since the revised OTBT and OPAT reactor trip setpoint equations allow the typical gross core average thermal power to slightly exceed this 118%value, as noted above, this fact was addressed with respect to the steamline break-core response (SLB-CR)methodology.
The full-power steamline break analysis for core response considerations is not in the Donald C.Cook Nuclear Plant licensing basis.Nevertheless, it has been determined that the revised OTBT and OPAT setpoint equations, with the OPZT reference average temperature (T")restricted to values no greater than 563.0'F, provides sufficient assurance that minimum DNBR will be protected during a HFP SLB.3.3.2.2'mergency Diesel Generator Start-up Time Relaxation Those events that must consider a loss of offsite power (i.e., Loss of All AC Power to the Station Auxiliaries, and Steam line Break for core response)have been evaluated with respect to an increase in the EDG start time from 10 seconds to 30 seconds, This start time relaxation of the EDG has been found to be acceptable.
This feature also affects the Main Steamline Break Mass/Energy Releases Inside Containment analysis.However, the impact is limited to the containment response portion of the analysis, as the steam mass and energy release calculations are based on the conservative assumption that offsite power is available for the duration of the blowdown.3.3.2.3 Pressurizer Code Safety Valve Setpoint Tolerance Increase The following events, which are potentially impacted by an increase in the pressurizer code safety valve setpoint tolerance, have been shown to support an increase from+1%to+3%setpoint tolerance:
Loss of External Electrical Load;Loss of Normal Feedwater; Loss of All AC Power to the Station Auxiliaries; and Locked Rotor/Shaft Break events.Thus, a setpoint tolerance of+3%for the pressurizer code safety valves is acceptable.
3.3.2.4 Shutdown Margin Relaxation All of the current Donald C.Cook Nuclear Plant Unit 1 licensing-basis analyses (i.e., the analyses supporting the Rerating Program)that model shutdown margin (SDM)assume 1.3%6k/k, except for the Steamline Break for Core Response (SLB-CR)event.The re-analysis of the SLB-CR event for the SGTP Program was performed with a SDM assumption of 1.3%6k/k.As such, all of the non-LOCA safety analyses that model SDM support the reduced SDM value of 1.3%6k/k.mA1944<w.wpt:1d441195 3.3-5
'H t]I II f 3.3.2:5 Steamline Break Protection System Modification The coincidence logic.currently required for safety injection initiation and steamline isolation on high steam flow and low steam pressure or low-low T,~for Unit 1 is to be modified to match that installed at Unit 2.This logic is part of the steamline break protection system.A detailed description of both of the steamline break protection systems currently installed in each of the units is presented in Section 3.5.4.The proposed Unit 1 modification, which will result in the two units having identical steamline break protection systems, consists of replacing Sl actuation on high steam flow coincident with low steam pressure, or high steam flow coincident with low-low T,~, with Sl actuation on low steam pressure only.The proposed Unit 1 modification also replaces steamline isolation on high steam flow coincident with low steam pressure with SLI on low steam pressure only.The coincidence requirement for high steam flow with low steam pressure of the current Unit 1 design increases the likelihood that safeguards initiation might be delayed compared to the proposed Unit 1 modified design, where only a low steam pressure signal is required.In the case where the coincidence logic prohibits safety injection and steamline isolation on high steam flow with low steam pressure, one of the other signals must be received before the safeguards are initiated.
As such, the currently installed Unit 1 steamline break protection system design assumed in the Steamline Break Mass/Energy Releases Outside Containment calculations (Section 3.3.4.7), the Rupture of a Steam Pipe analysis (Section 3.3.5.6), and the Steamline Break Mass/Energy Releases Inside Containment analysis (Section 3.5.4)bounds the proposed modifications to the Unit 1 steamline break protection system, as the requirement to satisfy the coincidence, discussed above, can result in Sl and/or steamline isolation later in a transient than had Sl and/or SLI actuated by low steam pressure alone.A delay in the initiation of safeguards is conservative for all three of the previously listed events.Also, an evaluation regarding the Major Rupture of a Feedwater Pipe event for Unit 1 has been performed (Section 3.3.4.8), which specifically addresses the proposed Unit 1 steamline break protection system modifications.
It should be noted that the deleted Sl actuation function, i.e., high steam flow coincident with low-low T,~, is not modeled in any of the non-LOCA safety analyses.3.3.3 Methodology The Unit 1 non-LOCA safety evaluation for the SGTP Program was performed using current Westinghouse methodology and computer codes.The following four sub-sections discuss: the Initial Conditions assumed, which reflects the change from the Improved Thermal Design Procedure (ITDP)(utilized for the Rerating Program of Unit 1)to the Revised Thermal Design Procedure (RTDP)for most of the events that are DNB limited;the Computer Codes Utilized;and the 5%RCS Flow Asymmetry.
m51944<w.wpt:
i d~1195 3.3-6 3.3.3.1 Initial Conditions For most accidents which are DNB limited, nominal values of initial conditions and minimum measured flow (339,100 gpm)are assumed.The allowances on reactor power, RCS temperature and pressure are determined on a statistical basis and are included in the limit DNBR as described in WCAP-11397 (Reference 2).This procedure is known as the"Revised Thermal Design Procedure".
For accidents that are not DNB limited or in which RTDP is not employed the initial conditions are obtained by adding the maximum steady-state errors to rated values.The following steady-state errors are considered:
a.Core Power+2%calorimetric error allowance b.Average RCS Temperature
+4.1'F controller deadband and measurement error allowance; also a+1.0'F bias for cold-leg streaming c.Pressurizer Pressure+67 psi steady-state fluctuations and measurement error allowance (see paragraph below)d.Reactor Flow Thermal Design Flow (332,800 gpm)It should be noted that the pressurizer pressure uncertainty includes an allowance of 23 psi for"readability," which is only applicable for DNB considerations.
However, the 67 psi u'ncertainty was conservatively applied to all non-LOCA analyses for simplicity.
Thus, there is an additional 23 psi of pressure margin that can be realized, if necessafy, for non-DNB events.Table 3.3-4 summarizes initial conditions and computer codes used in the accident analysis, and shows which accidents employed a DNB analysis using the RTDP.3.3.3.2 Computer Codes Utilized Summaries of the principal computer codes used in the transient analyses are given below.The codes used in the analysis of each transient have been listed in Table 3.3-4.FACTRAN FACTRAN calculates the transient temperature distribution in a cross-section of a metal clad UO, fuel rod and the transient heat flux at the surface of the clad using as input the nuclear m%1 944<w.wpf:1d441195 3.3-7 power and the time-dependent coolant parameters (pressure, flow, temperature, and density).The code uses a fuel model which simultaneously exhibits the following features: A.A sufficiently large number of radial space increments to handle fast transients such as rod ejection accidents.
B.Material properties which are functions of temperature and a sophisticated fuel-to-clad gap heat transfer calculation.
C.The necessafy calculations to handle post-departure from nucleate boiling transients:
film boiling heat transfer correlations, Zircaloy-water reaction, and partial melting of the materials.
.FACTRAN is further discussed in Reference 3.LOFTRAN The LOFTRAN program is used for transient response studies of a pressurized water reactor (PWR)system to specified perturbations in process parameters.
LOFTRAN simulates a multiloop system by a model containing the reactor vessel, hot and cold leg piping, steam generators (tube and shell sides), and the pressurizer.
The pressurizer heaters, spray, relief, and safety valves are also considered in the program.Point model neutron kinetics, and reactivity effects of the moderator, fuel, boron, and rods are included.The secondary side of the steam generator utilizes a homogenous, saturated mixture for the thermal transients and a water level correlation for indication and control.The reactor protection system is simulated to include reactor trips on high neutron flux, overpower hT, overtemperature dT, high and low pressure, low flow, and high pressurizer level.Control systems are also simulated including rod control, steam dump, feedwater control, and pressurizer pressure control.The ECCS, including the accumulators, is also modeled.LOFTRAN also has the capability of calculating the transient value of DNBR based on the input from the core limits.The core limits represent the minimum value of DNBR as calculated for typical or thimble cell.LOFTRAN is further discussed in Reference 4.TWINKLE The TWINKLE program is a multi-dimensional spatial neutron kinetics code, which was patterned after steady-state codes presently used for reactor core design.The code uses an implicit finite-difference method to solve the two-group transient neutron diffusion equations in one, two, and three dimensions.
The code uses six delayed neutron groups and contains a mA1944<w.wpf:1d 441195 3.3-8
detailed multi-region fuel-clad-coolant heat transfer model for calculating pointwise Doppler and moderator feedback effects.The code handles up to 2000 spatial points and performs its own steady-state initialization.
Aside from basic cross-section data and thermal-hydraulic parameters, the code accepts as input basic driving functions such as inlet temperatures, pressure, flow, boron concentration, control rod motion, and others.Various edits are provided;e.g., channel wise power, axial offset, enthalpy, volumetric surge, pointwise power, and fuel temperatures.
The TWINKLE code is used to predict the kinetic behavior of a reactor for transients which cause a major perturbation in the spatial neutron flux distribution.
TWINKLE is further described in Reference 5.THING IV The THING IV computer program, as approved by the NRC, is used to determine coolant density, mass velocity, enthalpy, vapor void, static pressure, and DNBR distributions along parallel flow channels within a reactor core under all expected operating conditions.
The THING IV code is described in detail in Reference 6.3.3.3.3 5%RCS Flow Asymmetry A 5%RCS flow asymmetry is supported by the non-LOCA safety analyses.Specifically, a reduction of RCS flow in one loop up to 5%below the nominal average per loop flow rate is acceptable, as long as the total minimum measured RCS flow is equal to or greater than 339,100 gpm.Should more than one loop be below the 84,775 gpm/loop flow rate, the sum of the loop flow shortfalls can be no greater than 5%of one loop.The non-LOCA events that potentially are sensitive to asymmetric RCS flow include: Uncontrolled RCCA Bank Withdrawal from a Subcritical Condition (RWFS), Partial Loss of Forced Reactor Coolant Flow (PLOF), Reactor Coolant Pump Locked Rotor/Shaft Break (Locked Rotor), Loss of Normal Feedwater, Excessive Heat Removal Due to Feedwater System Malfunctions, Loss of All AC Power to the Station Auxiliaries, Steamline Break for Core Response, Rod Ejection at zero power conditions (HZP Rod Ejection), and Rupture of a Main Feedwater Pipe.The balance of the non-LOCA events are not sensitive to RCS flow asymmetry.
The following events explicitly accounted for the effects of asymmetric RCS flow as part of the SGTP Program analyses: RWFS, PLOF, Locked Rotor, and HZP Rod Ejection.Specifically, the PLOF and Locked Rotor analyses model the fault to oocur in the loop with the highest flow (i.e., 5%above the nominal per loop minimum measured flow).Thus, the low flow reactor trip (conservatively assumed as a percentage of nominal flow as opposed to a percentage of mh1944<w.wpf:
1 d 441195 3.3-9 normalized flow)is delayed as much as possible, also the largest overall flow reduction is obtained with this model.For the RWFS and Rod Ejection analyses, a RCS flow corresponding to two out of four reactor coolant pumps in service (Mode 3 flow)is assumed.A conservative flow fraction is used, which bounds a worst case 5%flow asymmetry scenario in Mode 3 where the loops that would be providing the most flow are out of service.The safety analysis criteria for all of the aforementioned reanalyses continued to be met after explicitly accounting for the asymmetric RCS flow.The remainder of the events that are potentially sensitive to asymmetric RCS flow, but did not explicitly account for the effects in the specific analysis, have been evaluated and were found to be able to accommodate a RCS flow asymmetry of 5%.Thus, it can be concluded that 5%RCS flow asymmetry is supported (either directly or indirectly) by the Cook Nuclear Plant Unit 1 non-LOCA safety analyses and evaluations.
3.3.4 Non-LOCA Safety Evaluation:
Transients Evaluated The sections that follow contain the detailed descriptions of the impact of the SGTP Program on the applicable non-LOCA transients.
This first grouping of transients are those which could be evaluated and the second grouping (Section 3.3.5)are transients which required re-analysis.In all cases the appropriate UFSAR acceptance criteria are satisfied.
3.3.4.1 Chemical and Volume Control System Malfunction Reactivity can be added to the core by feeding pnmary grade water into the RCS via the reactor makeup portion of the CVCS.Boron dilution is a manual operation.
A boric acid blend system is provided to permit the operator to match the boron concentration of reactor coolant makeup water during normal charging to that in the RCS.The CVCS is designed to limit, even under various postulated failure modes, the potential rate of dilution to a value which provides the operator sufficient time to correct the situation in a safe and orderly manner.The opening of the Primary Water Makeup Control Valve supplies water to the RCS which can dilute the reactor coolant.Inadvertent dilution can be readily terminated by closing this valve.In order for makeup water to be added to the RCS, at least one charging pump must also be running in addition to the primary water pumps.The rate of addition of unborated water makeup to the RCS is limited by the capacity of the primary water pumps.The maximum addition rate in this case is 225 gpm with both primary water pumps running.The 225 gpm reactor makeup water delivery rate is based on a pressure drop calculation comparing the pump curves with.the system resistance curve.This is the maximum delivery based on the unit piping layout.Normally, only one primary water supply pump is operating white the other is on standby.m:41944<w.wpf:1d 441 195 3.3-10 Jt The boric acid from the boric acid tank is blended with primary grade water in the blender and the composition is determined by the preset flow rates of boric acid and primary grade water on the control board., In order to dilute, two separate operations are required.First, the operator must switch from the automatic makeup mode to the dilute mode;second, the start button must be depressed.
Omitting either step would prevent dilution.This makes the possibility of inadvertent dilution very remote.Information on the status reactor coolant makeup is continuously available to the operator.Lights are provided on the control board to indicate the operating condition of pumps in the CVCS.Alarms are actuated to warn the operator if boric acid or demineralized water flow rates deviate from preset values as a result of system malfunction.
To cover the phases of the plant operation and to account for the reduction in the volume of the RCS due to the increase in the level of SGTP up to 30%, boron dilution during startup and power operation were examined.Included in the evaluation was the effect of the difference in the density of unborated makeup water and the density of the reactor coolant.The evaluation is to show that, from initiation of the event, sufficient time is available to allow the operator to determine the cause of the addition and take corrective action before excessive shutdown margin is lost.Results and Conclusions Because of the steps involved in the dilution process, an erroneous dilution is considered highly unlikely.Nevertheless, if it does occur, numerous alarms and indications are available to alert the operator to the condition.
The maximum reactivity addition due to the dilution is slow enough to allow the operator to determine the cause of the addition and take corrective action before excessive shutdown margin is lost for the phases of the plant operation (start-up and at-power)affected by the SGTP Program.3.3.4.2 Startup of an Inactive Loop In accordance with Technical Specification 3/4.4.1, Cook Nuclear Plant Unit 1 operation during Modes 1 and 2 with less than four reactor coolant loops is not permitted.
Since three loop operation during Modes 1 and 2 is prohibited, the Startup of an Inactive Loop event does not have to be considered as part of the 30%SGTP Program.3.3.4.3 Loss of Normal Feedwater A loss of normal feedwater (from pump failures, valve malfunctions, or loss of offsite AC power)results in a reduction in capability of the secondary system to remove the heat mA1944<w.wpf:
1 d~1195 3.3-11
generated in the reactor core.If an alternative supply of feedwater were not supplied to the plant, core residual heat following reactor trip would heat the primary system water to the point where water relief from the pressurizer would occur, resulting in a substantial loss of water from the RCS.Since the plant is tripped well before the steam generator heat transfer capability is reduced, the primary system variables never approach a DNB condition.
The reactor trip on low-low water level in any steam generator provides the necessary protection against a loss of normal feedwater.
The auxiliary feedwater system is started automatically.
The turbine driven auxiliary feedwater pump utilizes steam from the secondary system and exhausts to the atmosphere.
The motor driven auxiliary feedwater pumps are supplied by power from the diesel generators if a loss of offsite power occurs.The pumps take suction directly from the condensate storage tank for delivery to the steam generators.
An evaluation of the system transient has been performed to show that following a loss of normal feedwater, with iriitial plant conditions consistent with those defined in the SGTP Program, the auxiliary feedwater system is capable of returning the plant to a safe condition by removing the stored and residual heat, thus preventing either overpressurization of the RCS or uncovery of the core.The results of the evaluation demonstrate that the Loss of Normal Feedwater event can support the 30%SGTP conditions.
The limiting peak pressurizer level occurs under low T,~(553'F)conditions.
This is consistent with the current Loss of Normal Feedwater analysis of record performed as part of the Rerating Program.The conclusions presented in the Donald C.Cook Nuclear Plant Unit 1 UFSAR (Reference 14)remain applicable for 30%SGTP conditions, since the Loss of Normal Feedwater analysis under rerated conditions (i.e., Cases 3 and 4 of Table 3.3-1)yield more severe results than those obtained from the sensitivity cases investigated for the SGTP Program.This is due to the benefits from the power level reduction (3411 MWt~3250 MWt)and the increase in the lower bound T,~(547'F~553'F)more than offsets the heat removal penalties caused by the increase in SGTP level (15%~30%)and the thermal design flow reduction (354,000 gpm-+332,800 gpm).3.3.4.4 Excessive Heat Removal due to Feedwater System Malfunctions Reductions in feedwater temperature or additions of excessive feedwater are means of increasing core power above full power.Such transients are attenuated by the thermal capacity of the secondary plant and of the RCS.The Overpower-Overtemperature Protection (high neutron flux, overpower dT, and overtemperature hTAnps)prevents any power increase which could lead to DNBR less than minimum allowable value in the event that the steam generator High-High Level Protection has not been actuated.m:11944<w.wpf:
1 d~1195 3.3-12 Excessive feedwater flow may be caused by full opening of a feedwater control valve due to a Feedwater Control System malfunction or an operator error.At power conditions, this excess flow causes a greater.load demand on the RCS due to increased subcooling in the steam generator.
With the plant at no load conditions, the addition of cold feedwater may cause a decrease in RCS temperature and thus a reactivity insertion due to the effects of the negative moderator coefficient of reactivity.
The excessive heat removal due to Feedwater System Malfunction events are examined primarily to demonstrate core protection.
For.the SGTP Program, an evaluation of the system transient has been performed to show that acceptable consequences will occur in the event of an excessive feedwater addition, due to control system malfunction or operator error which allows one or more feedwater control valve(s)to open fully.This evaluation considered both at power and zero power scenarios with the reactor being operated under both automatic and manual rod control conditions.
A feedwater malfunction event as described above results in an increase in the rate at which heat is removed from the reactor coolant.An increase in the level of tube plugging in the steam generators results in a reduction in the heat transfer characteristics between the primary coolant and the steam system.Thus, a less severe cooldown would be experienced for this event under the 30%SGTP conditions.
However, the RCS flow reduction due to the larger number of tubes being plugged is a DNB penalty for the at power events.Furthermore, the reduction in core power from the rerated value of 3411 MWt to 3250 MWt provides a DNB benefit.The evaluation performed for the SGTP Program conservatively ignored the benefit associated with the reduced ability of the excessive feedwater flow to cool the primary coolant.The evaluation conservatively minimized the benefit associated with the rated thermal power reduction and conservatively maximized the penalty due to the RCS flow reduction with respect to the power and flow values assumed in the analyses of record.The parameters assumed in the analyses of record for the feedwater malfunction events are consistent with those presented as Cases 3 and 4 of Table 3.3-1.The evaluation concluded that both the at power and zero power feedwater malfunction transients can support 30%SGTP conditions.
The reactivity insertion rate assumed in the current UFSAR analysis (120 pcm/sec)for the zero power event continues to be conservative for the 30%SGTP conditions.
It should be noted that the revised OTbT and OPbT setpoint equations do not impact the feedwater malfunction events, due to the fact that the analyses do not take credit for the protection offered by these trip functions.
The conclusions presented in the Donald C.Cook Nuclear Plant Unit 1 UFSAR for the Excessive Heat Removal Due to Feedwater System Malfunctions (UFSAR Section 14.1.10)remain applicable.
m:11 944<w.wpf:1d~1195 3.3-13 3.3.4.5 Excessive Increase in Secondary Steam Flow An excessive load increase incident is defined as a rapid increase in steam flow that causes a power mismatch between the reactor core power and the steam generator load demand.The reactor control system is designed to accommodate a 10%step load increase and a 5%per minute ramp load increase in the range of 15 to 100%of full power.Any loading rate in excess of these values may cause a reactor trip actuated by the reactor protection system.This accident could result from either an administrative violation such as excessive loading by the operator or an equipment malfunction in the steam dump control or turbine speed control.During power operation, steam dump to the condenser is controlled by reactor coolant condition signals, i.e., high reactor coolant temperature indicates a need for steam dump.A single controller malfunction does not cause steam dump;an interlock is provided which blocks the opening of the valves unless a large turbine load decrease or turbine trip has occurred.Protection against an excessive load increase accident is provided by the following reactor protection system signals: Overpower AT Overtemperature d, T Power range high neutron flux Low pressurizer pressure An excessive increase in steam load results in an increase in the rate at which heat is removed from the reactor coolant.An increase in the level of tube plugging in the steam generators results in a reduction in the heat transfer characteristics between the primary coolant and the steam system.Thus, a less severe cooldown would be experienced for this event under the 30%SGTP conditions.
However, the RCS flow reduction due to the large number of tubes being plugged is a DNB penalty for this event.Conversely, the reduction in the core power from the rerated value of 3411 MWt to 3250 MWt provides a DNB benefit.The evaluation performed for the SGTP Program conservatively ignored the benefit associated with the reduced ability of the excessive steam flow to cool the primary coolant.The evaluation conservatively minimized the benefit associated with the rated thermal power reduction and conservatively maximized the penalty due to the RCS flow reduction with respect to the power and flow values assumed in the analyses of record.The parameters assumed in the analyses of record for the Excessive Load increase Incident are consistent with those presented as Cases 3 and 4 of Table 3.3-1....m%1 944<w.wpf:1d~1195 3.3-14 The evaluation concluded that the Excessive Load Increase Incident can support the 30%SGTP conditions.
The revised OTET and OPBT setpoint equations do not impact this event, as the current analysis of record resulted in the plant reaching a stabilized condition at the higher power level, i.e., no reactor trip occurred for this event.The conclusions presented in the Donald C.Cook Nuclear Plant Unit 1 UFSAR for the Excessive Load Increase Incident (UFSAR Section 14.1.11)remain applicable.
3.3.4.6 Loss of All AC Power to the Station Auxiliaries The loss of all AC power to the station auxiliaries event, as with the loss of normal feedwater incident, is a limiting transient with respect to pressurizer overfill~The decrease in primary to secondary heat transfer ability, due to the increase in SGTP, aggravates the heatup portion of the transient, and increases the potential for filling the pressurizer.
As such, the loss of all AC power to the station auxiliaries is evaluated for the SGTP Program.A complete loss of all (non-emergency)
AC power (e.g.offsite power)may result in the loss of all power to the plant auxiliaries, i.e., the RCPs, condensate pumps, etc.The loss of power may be caused by a complete loss of the offsite grid accomplished by a turbine generator trip at the station, or by a loss of the onsite AC distribution system.This transient is analyzed to show the adequacy of the heat removal capability of the auxiliary feedwater system.The transient is more severe than the loss of load event analyzed because in this case the decrease in heat removal by the secondary system is accomplished by a flow coastdown which further reduces the capacity of the primary coolant to remove heat from the core.The reactor will trip due to: (1)turbine trip;(2)upon reaching one of the trip setpoints in the primary and secondary systems as a result of the flow coastdown and decrease in secondary heat removal;or (3)due to loss of power to the control rod drive mechanisms as a result of the loss of power to the plant.Following a loss of power with turbine and reactor trips, the sequence described below will occur: A.Plant vital instruments are supplied from emergency DC power sources.B.As the steam system pressure rises following the trip, the steam generator power-operated relief valves may be automatically opened to the atmosphere.
The condenser is assumed not to be available for steam dump.If the steam flow rate through the power relief valves is not available, the steam generator safety valves may lift to dissipate the sensible heat of the fuel and coolant plus the residual decay heat produced in the reactor.mA1944<w.wpf:1d~1195 3.3-15
As the no load temperature is approached, the steam generator power-operated relief valves (or safety valves, if the power operated relief valves are not available) are used to dissipate the residual decay heat and to maintain the plant at the hot standby condition.
The standby diesel generators, started on loss of voltage on the plant emergency busses, begin to supply plant vital loads.The motor driven auxiliary feedwater pumps are supplied power by the diesels and the turbine-driven pump utilizes steam from the main steam system.Both type pumps are designed to supply rated flow within 80 seconds of the initiating signal even if a loss of all non-emergency AC power occurs simultaneously with loss of normal feedwater.
The turbine exhausts the used steam to the atmosphere.
The auxiliary feedwater pumps take suction from the condensate storage tank for deliver to the steam generators.
Following the RCP coastdown caused by the loss of AC power, the natural circulation capability of the RCS will remove decay heat from the core, aided by auxiliang feedwater in the secondang system.The results of an evaluation is presented here to show that the natural circulation flow in the RCS, following a loss of all AC power to the station auxiliaries with initial plant conditions consistent with those defined in the SGTP Program is sufficient to remove decay heat from the core.The results of the evaluation demonstrate that the Loss of AII AC Power to the Station Auxiliaries (LOOP)event can support the 30%SGTP conditions.
The limiting peak pressurizer level occurs under low T, (553'F)conditions.
This is consistent with the current LOOP analysis of record, which was performed for the Rerating Program.The conclusions presented in the Donald C.Cook Nuclear Plant Unit 1 UFSAR (Reference 14)remain applicable for the 30%SGTP conditions, since the LOOP analysis under rerated conditions (i.e, Cases 3 and 4 of Table 3.3-1)yield more severe results than those obtained from the sensitivity cases investigated for the SGTP Program.This is due to the benefits from the'ower level reduction (3411 MWT-+3250 MWt)and the increase in the lower bound T,~(547'F~553'F)more than offsets the heat removal penalties caused by the increase in SGTP level (15%~30%), the additional delay in AFW delivery (60 seconds-+80 seconds)due to the relaxed EDG start time delay (10 seconds~30 seconds), and the thermal design flow reduction (354,000 gpm~332,800 gpm).3.3.4.7 Steamline Break Mass/Energy Releases Outside Containment The existing mass and energy (M/E)releases following a steamline break (SLB)outside containment were performed to support the range of conditions possible for the Rerating Program of Unit 1 (Cases 3 and 4 of Table 3.3-1), as well as to position Unit 2 for a potential future uprating (i.e., 3600 MWt NSSS).Thus, the M/E releases are based upon a rated m 51944<w.wpf:1d441195 3.3-16 thermal power of 3600 MWt, The core reactivity parameters were chosen to conservatively maximize the'reactivity feedback effects of the cooldown resulting from a blowdown from either Donald C.Cook Nuclear Plant unit.The changes associated with the SGTP Program for Unit 1, i.e., RCS flow reduction, reduced primary-to-secondary heat transfer capability, and reduction in the rated thermal power, are less limiting parameters relative to the assumptions currently made for the M/E release calculations following a SLB outside containment.
Furthermore, the adjustment in the K, safety analysis value of the OPbT setpoint equation (discussed in Section 3.3.2.1)does not impact the SLB M/E Release Outside Containment analysis, which is the only non-LOCA safety analysis that relies on this trip function for primary protection.
This is because a conservatively larger K, value of 1.18 was assumed in the SLB M/E Release Outside Containment analysis.The revised safety analysis K, value for Unit 1 is 1.172.The increase in the EDG start time delay from 10 seconds to 30 seconds has no effect on this analysis as well, since it is conservative to maintain offsite power such that reactor coolant pump operation is maintained (which aids in maximizing the steam releases).
Therefore, it can be concluded that the current licensing basis outside containment SLB M/E releases (UFSAR Section 14.4.11.3) continue to bound Unit 1 operation as defined by the SGTP Program.It is key to notice that the existing outside containment SLB M/E release analysis-of-record became part of the Cook Nuclear Plant Unit 1 licensing basis (and Unit 2 for that matter)following the approval of the Boron Injection Tank (BIT)Removal submittal (Reference 18).The existing SLB M/E Release Outside Containment analysis assumed: a.End-of-life shutdown margin of 1.3 IDk/k at no-load, equilibrium xenon conditions, and the most reactive RCCA stuck in its fully withdrawn position.Minimum capability for the injection of boric acid solution corresponding to the most restrictive single failure in the safety injection system.The ECCS consists of the following systems: 1)the passive accumulators, 2)the low head safety injection (residual heat removal)system, 3)the high head safety injection system, and 4)the charging system.Only the charging system and the passive accumulators are modeled for the steamline break accident analysis for M/E releases outside containment.
Centrifugal Charging pump head degradation of 10%was assumed.Coincidence logic required for Sl and SLI consistent with the current Unit 1 steamline break protection system.A proposed modification will change the logic associated with this system.However, as discussed in Section 3.3.2.5, the current analysis, which assumes the cqrrant Unit 1 steamline break protection system, bounds the proposed modifications to the Unit 1 steamline break protection system.mh1 944<w.wpf:1d441195 3.3-17 3.3.4.8 Major Rupture of a Feedwater Pipe The feedline break event is currently presented in the Unit 1 UFSAR (Section 14.2.8)for"informational purposes only," as this event is not part of the Unit 1 licensing-basis.
However, Cook Unit 2 does have the feedline break event in its licensing scope.The SGTP Program for Unit 1 includes an evaluation to demonstrate that the response of Unit 1 to a feedline break is bounded by the existing Unit 2 feedline break analysis.A key stipulation for this evaluation is that the Unit 1 steamline break protection logic, which is currently classified as the"Old" system, must be modified to match the"Hybrid" steamline break protection logic that is in place at Cook Nuclear Plant Unit 2.A detailed evaluation was performed, which included sensitivity cases using the LOFTRAN code.The evaluation specifically assessed the plant parameter changes associated with the SGTP Program (Cases 1 and 2 of Table 3.3-1)relative to the Unit 2 parameters.
corresponding to rerated conditions (i.e., 3600 MWt NSSS).Sensitivity cases investigated the effects of increasing the pressurizer safety valve setpoint tolerance from a1%to a3%and increasing the EDG start time delay from 10 seconds to 30 seconds.The sensitivity cases also included the effects associated with the setpoint tolerance increasing from t1%to t3%, which has been previously evaluated (Reference 19).The evaluation concluded that the results presented in the Donald C.Cook Nuclear Plant Unit 2 UFSAR for the Major Rupture of Main Feedwater Pipe event (UFSAR Section 14.2.8)are applicable to Unit 1, provided that the steamline break protection logic installed at Unit 1 is modified to match that installed in Unit 2.Furthermore, the evaluation concluded that an increase in the EDG start time from 10 seconds to 30 seconds;an increase in the pressurizer safety valve setpoint tolerance from a1%to a3%, as well as the inclusion of the 1.0'F bias to account for the cold-leg streaming phenomenon can be accommodated.
mA1944<w.wpf:1d441195 3.3-18
3.3.5 Non-LOCA Safety Evaluation:
Transients Analyzed The subsections that follow contain the details of the accidents re-analyzed to support 30%SGTP operation of Unit 1.In all cases, the applicable UFSAR acceptance criteria are satisfied.
3.3.5.1 Uncontrolled RCCA Withdrawal From A Subcritical Condition The uncontrolled RCCA withdrawal from a subcritical condition event is analyzed to determine the impact of the reduced RCS flow as a result of the increased steam generator tube plugging level of 30%.This event is analyzed to demonstrate core protection.
Although the no-load temperature does not change for the SGTP Program, the reduction in nominal RCS flow is non-conservative with respect to the DNB transient.
An RCCA withdrawal incident is defined as an uncontrolled addition of reactivity to the reactor core by withdrawal of RCCA banks resulting in a power excursion.
While the occurrence of a transient of this type is highly unlikely, such a transient could be caused by a malfunction of the Reactor Control or Control Rod Drive Systems.This could occur with the reactor either subcritical or at power.The"at power" case is discussed in Section 3.3.5.2.Reactivity is added at a prescribed and controlled rate in bringing the reactor from a shutdown condition to a low power level during startup by RCCA bank withdrawal.
Although the initial startup procedure uses the method of boron dilution, the normal startup is with RCCA bank withdrawal
~RCCA bank motion can cause much faster changes in reactivity than can be made by changing boron concentration.
The control rod drive mechanisms are wired into preselected banks, and these bank configurations are not altered during the core life.The RCCA's are therefore physically prevented from being withdrawn in other than their respective banks.Power supplied to the rod banks is controlled such that no more than two banks can be withdrawn at any time.The RCCA drive mechanism is of the magnetic latch type and the coil actuation is sequenced to provide variable speed rod travel.The maximum reactivity insertion rate is analyzed by assuming the simultaneous withdrawal of the combination of the two banks of the maximum combined worth at maximum speed.Should a continuous control rod assembly withdrawal be initiated, the transient will be terminated by the following reactor trip functions.
Source range neutron flux level trip-actuated when either of two source range channels indicates a flux level above a preselected, manually adjustable value.This trip function may be manually bypassed when either intermediate range flux channel indicates a flux level above the source range cutoff level.It is mh1944<w.wpf:1d~1195 3.3-19
automatically reinstated when both intermediate range channels indicate a flux level below the source range cutoff level.Intermediate range neutron flux level trip-actuated when either of two intermediate range channels indicates a flux level above a preselected, manually adjustable value.This trip function may be manually bypassed when two of the four power range channel are reading above approximately 10 percent of full power flux and is automatically reinstated when three of the four power range channels indicate a flux level below this value.Power range neutron flux level trip (low setting)-actuated when two out of the four power channels indicate a flux level above approximately 25 percent of full power flux.This trip function may be manually bypassed when two of the four power range channels indicate a flux level above approximately 10 percent of full power flux and is automatically reinstated when three of the four channels indicate a flux level below this value.Power range neutron flux level trip (high setting)-actuated when two out of the four power range channels indicate a flux level above a preset setpoint.This trip function is always active.In addition, control rod stops on high intermediate range flux level and high power range flux level serve to discontinue rod withdrawal and prevent the need to actuate the intermediate range flux level trip and the power range flux level trip, respectively.
The neutron flux response to a continuous reactivity insertion is characterized by a ver fast power rise terminated by the reactivity feedback effect of the negative fuel temperature coefficient.
This self-limitation of the initial power burst results from a fast negative fuel temperature feedback (Doppler effect)and is of prime importance during a startup incident since it limits the power to a tolerable level prior to protective action.After the initial power burst, the neutron flux is momentarily reduced and then, if the incident is not terminated by a reactor trip, the neutron flux increases again, but at a much slower rate.Termination of the startup incident by the previously discussed protection channels prevents core damage.In addition, the reactor trip from pressurizer high pressure serves as a backup to terminate the incident before an overpressure condition could occur.Method of Anal sis The analysis of the uncontrolled RCCA bank withdrawal from subcritical accident is performed in three stages: first an average core nuclear power transient calculation, then an average core heat transfer calculation, and finally the departure from nucleate boiling ratio (DNBR)m:<1944<w.wpf:
td~1 195 3.3-20 calculation.
The average core nuclear calculation is performed using spatial neutron kinetics methods (TWINKLE)to determine the average power generation with time including the various total core feedback effects, i.e., Doppler reactivity and moderator reactivity.
The average heat flux and temperature transients are determined by performing a fuel rod transient heat transfer calculation in FACTRAN.The average heat flux is next used in THING IV for transient DNBR calculations.
Analysis of this transient incorporates the neutron kinetics, including six delayed neutron groups and the core thermal and hydraulic equations.
In addition to the neutron flux response, the average fuel, clad and water temperature, and also the heat flux response, are computed.In order to give conservative results for a startup incident, the following additional assumptions are made concerning the initial reactor conditions:
Since the magnitude of the neutron flux peak reached during the initial part of the transient, for any given rate of reactivity insertion, is strongly dependent on the Doppler power reactivity coefficient, a conservatively low value (i.e., small in absolute value)is used for the startup incident (-0.9 x 10 Ak/%power).The contribution of the moderator reactivity coefficient is negligible during the initial part of the transient because the heat transfer time constant between the fuel and the moderator is much longer than the neutron flux response time constant.However, after the initial neutron flux peak, the succeeding rate of power increase is affected by the moderator temperature reactivity coefficient.
Although during normal operation (100%rated power)the moderator coefficient will not be positive at any time in core life, a highly conservative value has been used in the analysis to yield the maximum peak core heat flux.The analysis is based on a moderator coefficient which was at least+5 pcm/'F at the zero power nominal average temperature, and which became less positive for higher temperatures.
This was necessary since the TWINKLE computer code used in the analysis is a diffusion theory code rather than a point kinetics approximation and the moderator temperature feedback cannot be artificially held constant with temperature.
3.The reactor is assumed to be at hot zero power (547'F).This assumption is more conservative than that of a lower initial system temperature.
The higher initial system temperature yields a larger fuel to water heat transfer, a larger fuel thermal capacity, and a less negative (smaller absolute magnitude)
Doppler coefficient.
The less negative Doppler coefficient reduces the Doppler feedback effect thereby increasing the neutron flux peak.The high neutron flux peak combined with a high fuel thermal capacity and larger thermal conductivity m:51944<w.wpi:1d~1195 3.3-21
yields a larger peak heat flux.Initial multiplication factor (k,)is assumed to be closely approaching

1.0 since

this results in the maximum neutron flux peak.Two reactor coolant pumps are assumed to be in operation.
This is conservative with respect to the DNB transient.
The most adverse combination of instrumentation and setpoint errors, as well as delays for trip signal actuation and control rod assembly release, are taken into account.A 10%increase has been assumed for the power range flux trip setpoint raising it from the nominal value of 25%to a value of 35%in addition to taking no credit for the source and intermediate range protection.
Reference to Figure 3.3-5, however, shows that the rise in nuclear flux is so rapid that the effect of errors in the trip setpoint on the actual time at which the rods are released is negligible.
In addition to the above, the rate of negative reactivity insertion corresponding to the trip action is based on the assumption that the highest worth control rod assembly is stuck in its fully withdrawn position.The accident is analyzed using the Standard Thermal Design Procedure with the initial conditions listed in Table 3.3-4.The analysis was performed for a reactivity insertion rate of 75 pcm/sec.This reactivity insertion rate assumed is greater than that for the simultaneous withdrawal of the combination of the two sequential control banks having the greatest combined worth at maximum speed (45 inches/minute).
1 pcm=10'hk/k Results and Conclusions The nuclear power, heat flux, fuel average temperature, and clad temperature versus time for a 75 pcm/sec insertion rate are shown in Figures 3.3-5 and 3.3-6.This insertion rate, coupled with the 30%SGTP conditions, yields a minimum DNBR which remains above the limit value.For the Rod Withdrawal from subcritical event, the core axial power distribution is severely peaked to the bottom of the core.The W-3 DNB correlation is used to evaluate DNBR in the span between the lower non-mixing vane grid and the first mixing vane grid.The WRB-1 correlation remains applicable for the rest of the fuel assembly.For all regions of the core the DNB design bases are met.mA1944<w.wpf:1d441195 3.3-22 3.3.5.2 Uncontrolled Control Rod Assembly Bank Withdrawal At Power An uncontrolled Rod Cluster Control Assembly (RCCA)withdrawal at power results in an increase in core heat flux.Since the heat extraction from the steam generator lags behind the power generation until the steam generator pressure reaches the relief or safety valve setpoint, there is a net increase in reactor coolant temperature.
Unless terminated by manual or automatic action, the power mismatch and resultant coolant temperature rise would eventually result in DNB.Therefore, to minimize the possibility of breaching the cladding, the Reactor Protection System is designed to terminate any such transient before the DNBR falls below the limit value.The automatic features of the Reactor Protection System which minimize adverse effects to the core in an RCCA Bank Withdrawal incident at power include the following:
Nuclear power range instrumentation actuates a reactor trip on high neutron flux if two out of four channels exceed an overpower setpoint.Reactor trip is actuated if any two out of four hT channels exceed an overtemperature bT setpoint.This setpoint is automatically varied with axial power distribution, coolant average temperature and pressure to protect against DNB.3.Reactor trip is actuated if any two out of four hT channels exceed an overpower bT setpoint.This setpoint is automatically varied with coolant average temperature so that the allowable fuel power rating is not exceeded.4.A high pressure reactor trip, actuated from any two out of four pressure channels, is set at a fixed point.This set pressure is less than the set pressure for the pressurizer safety valves.5.A high pressurizer water level reactor trip, actuated from any two out of three level channels, is set at a fixed point.In addition to the above listed reactor trips, there are the following RCCA Withdrawal blocks.a.High neutron flux (one out of four)b.Overpower AT (two out of four)c.Overtemperature dT (two out of four)The manner in which the combination of overpower and.overtemperature bT trips provide protection over the full range of Reactor Coolant System conditions is illustrated in Figures 3.3-1 through 3.3-4.These figures represent the allowable conditions of reactor mal 944<w.wpf:1d441195 3.3-23 coolant loop average temperature and power with the design power distribution in a two-dimensional plot.The purpose of this analysis is to demonstrate the manner in which the above protective systems function for various reactivity insertion rates from different initial conditions.
Reactivity insertion rates and initial conditions govern which protective function occurs first.Method of Anal sis This transient is analyzed by the LOFTRAN code.The core limits as illustrated in Figure 3.3-1 through 3.3-4 are used as input to LOFTRAN to determine the minimum DNBR during the transient.
The analysis is performed to bound the conditions of high and low average temperature with high and low RCS pressures for Unit 1.This accident is analyzed with the RTDP described in Reference 2.Plant characteristics and initial conditions are listed in Table 3.3-4.For an uncontrolled rod withdrawal at power accident, the following conservative assumptions are made: A.Nominal values are assumed for the initial reactor power, pressure, and RCS temperatures.
Uncertainties in initial conditions are included in the limit DNBR as described in Reference 2.B.Reactivity coefficients
-two cases are analyzed: Minimum Reactivity Feedback.A+5 pcmPF moderator temperature coefficient of reactivity and a least negative Doppler only power coefficient (see Table 3.3A)are assumed.2.Maximum Reactivity Feedback.A conservatively large negative moderator temperature coefficient and a most negative Doppler only power coefficient (See Table 3.3-4)are assumed.The reactor trip on high neutron flux is assumed to be actuated at a conservative value of 118 percent of nominal full power.The bT trips include all adverse instrumentation and setpoint errors, while the delays for the trip signal actuation are assumed at their maximum values.D.'The RCCA trip insertion characteristic is based on the assumption that the highest worth assembly is stuck in its fully withdrawn position.mA1944<w.wpf:1d441195 3.3-24 E.The maximum positive reactivity insertion rate is greater than that for the simultaneous withdrawal of the combinations of the two control banks having the maximum combined worth at maximum speed.Results Figures 3.3-7 through 3.3-9 show the transient response for a rapid RCCA bank withdrawal incident starting from full power.Reactor trip on high neutron flux occurs shortly after the start of the accident.Since this is rapid with respect to the thermal time constants of the plant, small changes in T,~and pressure result and margin to DNB is maintained.
The transient response for a slow RCCA bank withdrawal from full power is shown in Figures 3.3-10 through 3.3-12.Reactor trip on overtemperature dT occurs after a longer period and the rise in temperature and pressure is consequently larger than for rapid RCCA bank withdrawal
~Again, the minimum DNBR is greater than the limit value.Figure 3.3-13 shows the minimum DNBR as a function of reactivity insertion rate from initial full power operation for minimum and maximum reactivity feedback.It can be seen that two reactor trip functions provide protection over the whole range of reactivity insertion rates.These are the high neutron flux and overtemperature bT functions.
The minimum DNBR is always greater than the limit value.Figures 3.3-14 and 3.3-15 show the minimum DNBR as a function of reactivity insertion rate for RCCA bank withdrawal incidents starting at 60 and 10 percent power respectively.
The results are similar to the 100 percent power case, except as the initial power is decreased, the range over which the overtemperature hT trip is effective is increased.
In neither case does the DNBR fall below the limit value.Conclusions The high neutron flux and overtemperature bT trip channels provide adequate protection over the entire range of possible reactivity insertion rates, i.e., the minimum value of DNBR is always larger than the limit value for all fuel types.Also, the pressurizer does not fill.mal 944<w.wpf:1 d~1195 3.3-25 3.3.5.3 Rod Cluster Control Assembly Misalignment The rod cluster control assembly misalignment events are primarily examined to demonstrate core protection.
Although the reduction in rated thermal power is a benefit for the DNB evaluation, the reduction in RCS flow is non-conservative with respect to the DNB transient.
As such, the rod cluster control assembly misalignment events are analyzed to determine the impact of the SGTP Program.Rod cluster control assembly misalignment accidents include: A.A dropped RCCA B.A dropped RCCA bank C.Statically misaligned RCCA Each RCCA has a position indicator channel which displays position of the assembly.The displays of assembly positions are grouped for the operator's convenience.
Fully inserted assemblies are further indicated by rod bottom light.Group demand position is also indicated.
RCCAs are always moved in preselected banks, and the banks are always moved in the same preselected sequence.The rods comprising a group operate in parallel through multiplexing thynstors.
The two groups in a bank move sequentially such that the first group is always within one step of the second group in the bank.A definite schedule of actuation (or deactuation of the secondary gripper, movable gripper, and lift coils of a mechanism) is required to withdraw the RCCA attached to the mechanism.
Since the stationaiy gripper, movable gripper, and lift coils associated with the RCCAs of a rod group are driven in parallel, any single malfunction which would cause rod withdrawal would affect a minimum of one group.Mechanical malfunctions are in the direction of insertion, or immobility.
A dropped RCCA or RCCA bank is detected by: a.Sudden drop in the core power level as seen by the nuclear instrumentation system;b.Asymmetric power distribution as seen on out-of-core neutron detectors or core exit thermocouples; c.Rod at bottom signal;d.Rod position deviation monitor, e.Rod position indication.
m%1 944<w.wpf:
1 d441195 3.3-26 Misaligned RCCA are detected by: a.Asymmetric power distribution as seen on out-of-core neutron detectors or core I'xit thermocouples; b.Rod position deviation monitor;c.Rod position indicators.
The resolution of the rod position indicator channel is+5 percent (+12 steps).Deviation of any assembly from its group by twice this distance will not cause power distributions worse than the design limits.The rod position deviation monitor alerts the operator to rod deviation before it can exceed ten percent of span (+24 steps).If the rod position deviation monitor is not operable, the operator is required to take action as required by the Technical Specifications.
Method of Anal sis A.One or more dropped RCCAs from the same group.For evaluation of the dropped RCCA event, the transient system response is calculated using the LOFTRAN code.The code simulates the neutron kinetics, RCS, pressurizer, pressurizer relief and safety valves, pressurizer spray, steam generator, and steam generator safety valves.The code computes pertinent plant variables including temperatures, pressures, and power level.Nominal values for initial reactor power, temperature, and RCS pressure are assumed to bound the operation of Unit 1 with 30%SGTP.The initial conditions are presented in Table 3.3Q.Uncertainties for initial conditions are included in the limit DNBR.Statepoints are calculated and nuclear models are used to obtain a hot channel factor consistent with the primary system conditions and reactor power.By incorporating the primary conditions from the transient and the hot channel factor from the nuclear analysis, the DNB design basis is shown to be met using the THING IV code.The transient response, nuclear peaking factor analysis, and DNB design basis confirmation are performed in accordance with the methodology described in Reference 9.Note that operation with automatic rod control is assumed for the analysis.Also note that the analysis does not take credit for the negative flux rate reactor trip.mh1944<w.wpf:
1 d~1195 3.3-27 Statically Misaligned RCCA Steady state power distributions are analyzed using the methodology described in Reference 9.The peaking factors are then used as input to the THING IV code to calculate the DNBR.Results One or more Dropped RCCAs Single or multiple dropped RCCAs within the same group result in a negative reactivity insertion.
The core is not adversely affected during this period, since power is decreasing rapidly.Following plant stabilization, normal rod retrieval or shutdown procedures are followed.The operator may manually retrieve the RCCA by following approved operating procedures.
Power may be reestablished either by reactivity feedback or control bank withdrawal.
Following a dropped rod event in manual rod control, the plant will establish a new equilibrium condition.
The equilibrium process without control-system interaction is monotonic, thus removing power overshoot as a concern, and establishing the automatic rod control mode of operation as the limiting case.For a dropped RCCA event in the automatic rod control mode, the Rod Control System detects the drop in power and initiates control bank withdrawal.
Power overshoot may occur due to this action by the automatic rod controller after which the control system will insert the control bank to restore nominal power.Figures 3.3-16 and 3.3-17 show a typical transient response to a dropped RCCA (or RCCAs)in automatic control.Uncertainties in the initial condition are included in the DNB evaluation as described in Reference 9.In all cases, the minimum DNBR remains above the limit value.Dropped RCCA bank A dropped RCCA bank typically results in a reactivity insertion greater than 500 pcm.The core is not adversely affected during the insertion period, since power is decreasing rapidly.The transient will proceed as described in"A" above;however, the return to power will be less due to the greater worth of an entire bank.Following plant stabilization, normal rod retrieval or shutdown procedures are followed to further cool down the plant.m:41 944<w.wpt:1d441195 3.3-28 C.Statically Misaligned RCCA The most severe misalignment situations with respect to DNBR at significant power levels arise from cases in which one RCCA is fully inserted, or where bank D is fully inserted with one RCCA fully withdrawn.
Multiple independent alarms, including a bank insertion limit alarm, alert the operator well before the postulated conditions are approached.
The bank can be inserted to its insertion limit with any one assembly fully withdrawn without the DNBR falling below the limit value.The insertion limits in the Technical Specifications may vary from time to time depending on a number of limiting criteria, It is preferable, therefore, to analyze the misaligned RCCA case at full power for a position of the control bank as deeply inserted as the criteria on minimum DNBR and power peaking factor will allow.The full power insertion limits on control bank D must then be chosen to be above that position and will usually be dictated by other criteria.Detailed results will vane from cycle to cycle depending on fuel arrangements.
With bank D inserted to its full insertion limit and one RCCA fully withdrawn, DNBR does not fall below the limit value.This case is analyzed assuming the initial reactor power, pressure, and RCS temperatures are at their nominal values (as given in Table 3.3A)but with the increased radial peaking factor associated with the misaligned RCCA.DNB calculations have not been performed specifically for RCCAs missing from other banks;however, power shape calculations have been done as required for RCCA ejection analysis.Inspection of the power shapes shows that the DNB and peak kw/ft situation is less severe than the bank D case discussed above assuming insertion limits on the other banks equivalent to a bank D insertion limit.For RCCA misalignments with one RCCA fully inserted, the DNBR does not fall below the limit value.This case is analyzed assuming the initial reactor power, pressure, and RCS temperatures are at their nominal values, (as given in Table 3.34)but with the increased radial peaking factor associated with the misaligned RCCA.DNB does not occur for the RCCA misalignment incident and thus the ability of the primary coolant to remove heat from the fuel rod is not reduced.The peak fuel temperature corresponds to a linear heat generation rate based on the radial peaking factor penalty associated with the misaligned RCCA and the design axial power distribution.
The resulting linear heat generation is well below that which would cause fuel melting.mh1944<w.wpt:1d441195 3:3-29 Following the identification of a RCCA group misalignment condition by the operator, the operator is required to take action as required by the plant Technical Specifications and operating instructions.
m:51 944<w.wpt:1d 441195 3,3-30 3.3.5.4 Loss of Reactor Coolant Flow (Including Locked Rotor Analysis)A loss of forced reactor coolant flow may result from a simultaneous loss of electrical supplies to all reactor coolant pumps.If the reactor is at power at the time of the accident, the immediate effect of loss of coolant flow is a rapid increase in the coolant temperature which is magnified by a positive MTC.This increase could result in DNB with subsequent adverse effects to the fuel if the reactor were not tripped promptly.The trip systems available to mitigate the consequence of this accident are discussed in the UFSAR.Simultaneous loss of electrical power to all reactor coolant pumps at full power is the most severe credible loss of flow condition.
For this condition reactor trip together with flow sustained by the inertia of the coolant and rotating pump parts will be sufficient to prevent RCS overpressurization and the DNB ratio from exceeding the limit values.The decrease in reactor coolant system flow rate events are primarily examined to demonstrate core protection.
The reduction in RCS flow, as a result of the increase in steam generator tube plugging to 30%, is non-conservative with respect to the DNB transient.
As such, analyses are presented to discuss the impact of this change.Method of Anal sis The following loss of flow cases are analyzed: 1.Loss of four pumps from nominal full power conditions with four loops operating.
2.Loss of one pump from nominal full power conditions with four loops operating.
The normal power supplies for the pumps are four buses connected to the generator.
Each bus supplies power to one pump.When a generator trip occurs, the pumps are automatically transferred to a bus supplied from external power lines, and the pumps will continue to supply coolant flow to the core.The simultaneous loss of power to all reactor coolant pumps is a highly unlikely event.Since each pump is on a separate bus, a single bus fault would not result in the loss of more than one pump.A full plant simulation is used in the analysis to compute the core average and hot spot heat flux transient responses, including flow coastdown, temperature, reactivity and control rod insertion effects.These data are then used in a detailed thermal-hydraulic.
computation to compute the margin to DNB using the RTDP.This computation solves the continuity, momentum and energy equations of fluid flow together with the WRB-1 DNB correlation.
mA1944<w.wpf:1dO41195 3.3-31
The analyses are performed to bound the conditions of the SGTP Program.Uncertainties in initial conditions are included in the limit DNBR as described in Reference 2.Nominal values are assumed for the initial reactor power, pressure, and RCS temperatures.
The initial conditions used are listed in Table 3.3-4.This transient is analyzed by three digital computer codes.First the LOFTRAN code is used to calculate the loop and core flow during the transient, the time of reactor trip based on the calculated flows, the nuclear power transient, and the primary system pressure and temperature transients.
The FACTRAN code is then used to calculate the heat flux transient based on the nuclear power and flow from LOFTRAN.Finally, the THING IV code is used to calculate the DNBR during the transient based on the heat flux from FACTRAN and flow from LOFTRAN.The DNBR transients presented represent the minimum of the typical or thimble cell for each type of fuel.Results Figures 3.3-18 through 3.3-20 show the transient response for the loss of power to all RCPs with four loops in operation.
The reactor is assumed to be tripped on undervoltage signal.Figure 3.3-20 shows the DNBR to be always greater than the limit value for the most limiting fuel assembly cell.Figures 3.3-21 through 3.3-23 show the transient response for the loss of one RCP with four loop operation.
The reactor is assumed to be tripped on low flow signal.Figure 3.3.-23 shows the DNB to be always greater than the limit value for the most limiting fuel assembly cell.The sequence of events following each of these transients is included in Table 3.3-5..Since DNB does not occur, the ability of the primary coolant to remove heat from the fuel rod is not significantly reduced.Thus, the average fuel and clad temperature do not increase significantly above their respective initial values.Conclusions The analysis shows that the DNBR will not decrease below the limit value at any time during the transient.
Thus, no fuel adverse effects or clad rupture is predicted, and all applicable acceptance criteria are met.m:11944<w.wpf:1d441195 3.3-32 Locked Rotor Accident A transient analysis has been performed for the instantaneous seizure of a reactor coolant pump rotor.Flow through the affected reactor coolant loop is rapidly reduced, leading to a reactor trip on a low flow signal.Following the trip, heat stored in the fuel rods continues to pass into the core coolant, causing the coolant to expand.At the same time, heat transfer to the shell side of the steam generator is reduced, first because the reduced flow results in a decreased tube side film coefficient and then because the reactor coolant in the tubes cools down while the shell side temperature increases (turbine steam flow is reduced to zero upon plant trip).The rapid expansion of the coolant in the reactor core, combined with the reduced heat transfer in the steam generator causes an insurge into the pressurizer and a pressure increase throughout the RCS.The insurge into the pressurizer causes a pressure increase which in turn actuates the automatic spray system, opens the power-operated relief valves, and opens the pressurizer safety valves, in a sequence dependent on the rate of insurge and pressure increase.The power-operated relief valves are designed for reliable operation and would be expected to function property during the accident.However, for conservatism, their pressure-reducing effect as well as the pressure-reducing effect of the spray are not included in this analysis.The locked rotor event is examined to determine the DNB transient and to demonstrate that the peak RCS pressure and peak clad temperature remain below the limit values.The reduction in RCS flow, due to the increase in the SGTP level, is non-conservative with respect to the DNB evaluation.
As such, the locked rotor event was re-analyzed.
Method of Anal sis Two digital-computer codes are used to analyze this transient.
The LOFTRAN code is used to calculate the resulting loop and core flow transients following the pump seizure, the time of reactor trip based on the loop flow transients, the nuclear power following reactor trip, and to determine the peak pressure.The thermal behavior of the fuel located at the core hot spot is investigated using the FACTRAN code, using the core flow and the nuclear power calculated by LOFTRAN.The FACTRAN code includes the use of a film boiling heat transfer coefficient.
The analysis is performed to bound the conditions associated with the SGTP Program.As in the previous UFSAR analysis, the analysis assumes offsite power is available following the reactor trip and turbine trip.Evaluation of the Pressure Transient After pump seizure, the neutron flux is rapidly reduced by Gontrol rod insertion.
Rod motion begins 1 second after the flow in the affected loop reaches 87 percent of nominal flow.No m:51944<w.wpf:1d441195
~3.3-33 credit is taken for the pressure reducing effect of the pressurizer relief valves, pressurizer spray, steam dump or controlled feedwater flow after plant trip.Although these operations are expected to occur and would result in a lower peak RCS pressure, an additional degree of conservatism is provided by ignoring their effect.The pressurizer safety valves are assumed to initially open at 2575 psia and achieve rated flow at 2580 psia.This analysis assumed an initial pressurizer pressure of 2317 psia.Table 3.3-4 presents the initial conditions assumed for the peak pressure transient.
Evaluation of the Peak Clad Temperature For this accident, DNB is assumed to occur in the core;therefore an evaluation of the consequences with respect to fuel rod thermal transients is performed.
The assumption of rods going into DNB as a conservative initial condition is made in order to determine the clad temperature and zirconium water reaction.This analysis assumed an initial pressurizer pressure of 2100 psia.Results obtained from analysis of this hot spot condition represent the upper limit with respect to clad temperature and zirconium water reaction.In the evaluation, the rod power at the hot spot is assumed to be 2.5 times the average rod power (i.e., FQ=2.5)at the initial core power level.Table 3.3-4 presents the initial conditions assumed for the peak clad temperature transient.
Film Boi%ng Coefficient The film boiling coefficient is calculated in the FACTRAN code using the Bishop-Sandberg-Tong film boiling correlation (Reference 13).The fluid properties are evaluated at film temperatures (average between wall and bulk temperatures).
The program calculates the film coefficient at every time step based upon the actual heat transfer conditions at the time.The neutron flux, system pressure, bulk density, and mass flow rate as a function of time are used as program input.For the peak clad temperature analysis, the initial values of the pressure and the bulk density are used throughout the transient since they are the most conservative with respect to clad temperature response.For conservatism, DNB was assumed to start at the beginning of the accident.Fuel Clad Gap Coefficient The magnitude and time dependence of the heat transfer coefficient between fuel and clad (gap coefficient) has a pronounced influence on the thermal results.The larger the value of the gap coefficient, the more heat is transferred between pellet and clad.Based on investigations on the effect of the gap coefficient upon the maximum clad temperature during m:51944<w.wpf:1d~1195 3.3-34 the transient, the gap coefficient was assumed to increase from a steady state value consistent with initial fuel temperature to 10,000 BTU/hr-ft'-'F at the initiation of the transient.
Thus the large amount of energy stored in the fuel because of the small initial value is released to the clad at the initiation of the transient.
Zirconium-Steam Reaction The zirconium-steam reaction can become significant above 1800'F (clad temperature).
In order to take this phenomenon into account, the following correlation, which defines the rate of the zirconium-steam reaction, was introduced into the models (Reference 10).=33.3 x ttP exp (-'1.366 where: w=amount reacted, mg/cm'=time, seconds T=temperature, K The reaction heat is 1510 cal/g Evaluation of Rods-in-ONB An evaluation is made to determine what percentage, if any, of rods are expected to be in DNB during the transient.
For this evaluation, the predicted core conditions are used as input to a THING IV calculation of the minimum DNBR during the transient.
Results of the THINC IV evaluation are then used to determine the percentage of fuel rods which experience DNB.Table 3.3-4 presents the initial conditions assumed for the rods-in-DNB evaluation.
Reeeite The transient results for the locked rotor accident are shown in Figures 3.3-24 through 3.3-26.The peak RCS pressure (2641 psia)reached during the transient is less than that which would cause stresses to exceed the faulted condition stress limits (this peak pressure is also below 110%of the design pressure).
The pressure response shown in Figure 3.3-25 is the response at the point in the Reactor Coolant System having the maximum pressure.Also, the peak clad surface temperature (1934'F, shown in Figure 3.3-26)is considerably less than 2700'F.The sequence of events is included in Table 3.3-5.For the most limiting fuel assembly, less than 7%of the.rods reach a DNBR value less than the limit value for the 30%SGTP conditions.
mA1944<w.wpf:1d441295 3.3-35 Conclusions A.Since the peak RCS pressure reached during any of the transients is less than that which would cause stresses to exceed the faulted conditions stress limits, the integrity of the primary coolant system is not endangered.
Since the peak clad surface temperature calculated for the hot spot during the worst transient remains considerably less than 2700'F (the temperature at which clad embrittlement may be expected), the core will remain in place and intact with no loss of core cooling capability.
mA1944<w.wpt:
1 d441195 3.3-36 3.3.5.5 Loss of External Electrical Load The complete loss of steam load from full power is examined primarily to show the adequacy of the pressure relieving devices and also to demonstrate core protection.
The reduction in RCS flow, as result of increasing the level of SGTP, is non-conservative with respect to the DNB behavior.Primary protection for this event is provided by the high pressurizer pressure, OTLT, high pressurizer water level, and low-low steam generator water level reactor trips.The loss of external electrical load may result from an abnormal variation in network frequency or other adverse network operating conditions.
It may also result from a trip of the turbine generator or in an unlikely opening of the main breaker from the generator which fails to cause a turbine trip but causes a rapid large NSSS load reduction by the action of the turbine control.Method of Anal sis The loss of load transients are analyzed by employing the detailed digital computer program LOFTRAN.The program simulates the neutron kinetics, RCS, pressurizer, pressurizer relief and safety valves, pressurizer spray, steam generator, and steam generator safety valves.The program computes pertinent plant variables including temperatures, pressures, and power level.An analysis is performed to bound the conditions of the SGTP Program.Nominal values are assumed for the initial reactor power, temperature, and pressure.This accident is analyzed with the RTDP.Plant characteristics and initial conditions are listed in Table 3.3-4.Major assumptions are summarized below: A.Initial Operating Conditions
-nominal conditions for reactor power, pressure, and RCS temperatures are assumed for statistical DNB analyses.B.Moderator and Doppler Coefficients of Reactivity
-the loss of load is analyzed with both maximum and minimum reactivity feedback.The maximum feedback cases assume a large negative moderator temperature coefficient and the most negative Doppler power coefficient.
The minimum feedback cases assume a positive moderator temperature coefficient and the least negative Doppler coefficients.
C.Reactor Control-from the standpoint of the maximum pressures attained it is conservative to assume that the reactor is i'anual control.If the reactor were in automatic control, the control rod banks would move prior to trip and reduce the severity of the transient.
mal 944<w.wpf:1d~1195 3.3-37 D.Pressurizer Spray and Power-Operated Relief Valves-two cases for both the minimum and maximum moderator feedback cases are analyzed: Full credit is taken for the effect of pressurizer spray and power-operated relief valves in reducing or limiting the coolant pressure.Safety valves are also available.
2.No credit is taken for the effect of pressurizer spray and power-operated relief valves in reducing or limiting the coolant pressure.Safety valves are operable.E.Steam Release-no credit is taken for the operation of the steam dump system or steam generator power-operated relief valves.The steam generator pressure rises to the safety valve setpoint where steam release through the safety valves limits the secondary steam pressure.F.Feedwater Flow-main feedwater flow to the steam generators is assumed to be lost at the time of turbine trip.No credit is taken for auxiliary feedwater flow since a stabilized plant condition will be reached before auxiliary feedwater initiation is normally assumed to occur;however, the auxiliary feedwater pumps would be expected to start on a trip of the main feedwater pumps.The auxiliary feedwater flow would remove core decay heat following plant stabilization.
G.Reactor trip is actuated by the first Reactor Protection System trip setpoint reached.Trip signals are expected due to high pressurizer pressure, overtemperature bT, high pressurizer water level, and low-low steam generator water level.Results The transient responses for a loss of load from full power operation are shown for four cases: minimum and maximum reactivity feedback, with and without pressure control (Figures 3.3-27 through 3.3-46).Figures 3.3-27 through 3.3-31 show the transient responses for the loss of load with minimum reactivity feedback assuming full credit for the pressurizer spray and pressurizer power-operated relief valves.No credit is taken for the steam dump.The reactor is tripped by the overtemperature AT trip signal.The minimum DNBR remains well above the limit value.The pressurizer relief and safety valves prevent overpressurization of the pnmary system.The steam generator safety valves mA1944<w.wpf:1d441195 3.3-38
prevent overpressurization of the secondary system, maintaining pressure below 110 percent of design value.Figures 3.3-32 through 3.3-36 show the responses for the total loss of steam load with maximum reactivity feedback.All other plant parameters are the same as the above.The DNBR increases throughout the transient and never drops below its initial value.Pressurizer relief valves and steam generator safety valves prevent overpressurization in primary and secondary systems, respectively.
The reactor is tripped by the low-low steam generator water level signal.The pressurizer safety valves are not actuated for this case.In the event that feedwater flow is not terminated at the time of turbine trip for this case.flow would continue under automatic control with the reactor at a reduced power.The operator would take action to terminate the transient and bring the plant to a stabilized condition.
If no action were taken by the operator the reduced power operation would continue until the condenser hotwell was emptied.A low-low steam generator water level reactor trip would be generated along with auxiliary feedwater initiation signals.Auxiliary feedwater would then be used to remove decay heat with the results less severe than those presented in Section 3.3.4.3, Loss of Normal Feedwater Flow.The loss of load accident was also studied assuming the plant to be initially operating at full power with no credit taken for the pressurizer spray, pressurizer power-operated relief valves, or steam dump.The reactor is tripped on the high pressurizer pressure signal.Figures 3.3-37 through 3.3-41 show the transient responses with minimum reactivity feedback.The neutron flux remains essentially constant at full power until the reactor is tripped.The DNBR never goes below its initial value throughout the transient.
In this case the pressurizer safety valves are actuated, and maintain system pressure below 110 percent of the design value.Figures 3.3-42 through 3.3-46 show the transient responses with maximum reactivity feedback with the other assumptions being the same as in the preceding case.Again, the DNBR increases throughout the transient and the pressurizer safety valves are actuated to limit primary pressure.The sequence of events following each of these transients is included in Table 3.3-6.Conclusions Results of the analyses show that the plant design is such that a loss of load without a direct or immediate reactor trip presents no hazard to the integrity of the RCS or the main steam system.Pressure relieving devices incorporated in the two systems are adequate to limit the maximum pressures to within the design limits.The integrity of the core is maintained by mhl944<w.wpf:1d441195 3.3-39 operation of the reactor protection system, i.e., the DNBR will be maintained above the limit value.Thus the conclusions presented is the UFSAR remain valid for 30%SGTP.3.3.5.6 Rupture of a Steam Pipe Although the no-load temperature does not change for the SGTP Program, and a reduction in the heat transfer capability, due to the increased number of plugged steam generator tubes, would result in a less severe cooldown, the impact of the RCS flow reduction needs to be addressed for the steamline break accident.The reanalysis also assumed a reduction in the available shutdown margin from 1.60 to 1.30%d,k/k at no-load conditions.
An evaluation has been performed for those cases that model a coincident loss of offsite power in order to address the increase in the EDG start time from 10 to 30 seconds.This analysis was performed assuming the coincidence logic required for SI and SLI consistent with the current Unit 1 steamline break protection system.A proposed modification to the Unit 1 steamline break protection system will change this logic.However, this analysis bounds the proposed modifications to the Unit 1 steamline break protection system, as discussed in Section 3.3.2.5.A rupture of a steam pipe results in an uncontrolled steam release from a steam generator.
The steam release results in an initial increase in steam flow which decreases during the accident as the steam pressure falls.The energy removal from the causes a reduction in coolant temperature and pressure.In the presence of a negative moderator temperature coefficient (MTC), the cooldown results in a reduction of core shutdown margin.If the most reactive RCCA is assumed stuck in its fully withdrawn position, there is an increased possibility that the core will become critical and return to power.A return to power following a steam pipe rupture is a potential concern mainly because of the high hot channel factors which exist when the most reactive assembly is assumed stuck in its fully withdrawn position.The core is ultimately shut down by boric acid delivered by the ECCS.The analysis of a steam pipe rupture is performed to demonstrate that: Assuming a stuck assembly, with or without offsite power, and assuming a single failure in the engineered safety features, there is no consequential damage to the primary system and the core remains in place and intact.Although DNB and possible clad perforation following a steam pipe rupture are not necessarily unacceptable, the following analysis, in fact, shows that no DNB occurs for any rupture assuming the most reactive assembly stuck in its fully withdrawn position.m:51 944<w.wpf:1d~1195 3.3<0
Method of Anal sis The analysis of the steam pipe rupture has been performed to determine:
A.The core heat flux and RCS temperature and pressure resulting from the cooldown following the steam line break.The LOFTRAN code has been used.The thermal and hydraulic behavior of the core following a steam line break.A detailed thermal and hydraulic digital-computer code, THING IV, has been used to determine if DNB occurs for the core conditions computed in item A above.The following conditions were assumed to exist at the time of a main steam line break accident: End-of-life shutdown margin (1.30%hk/k)at no load, equilibrium xenon conditions, and the most reactive RCCA stuck in its fully withdrawn position: Operation of the control rod banks during core bumup is restricted in such a way (to not violate the rod insertion limits presented in the Technical Specifications) that addition of positive reactivity in a steam line break accident will not lead to a more adverse condition than the case analyzed.A negative moderator coefficient corresponding to the end-of-life rodded core with the most reactive RCCA in the fully withdrawn position: The variation of the coefficient with temperature and pressure has been included.The k,versus temperature at 1050 psia corresponding to the negative moderator temperature coefficient used is shown in Figure 3.3-47.The Doppler power feedback assumed for this analysis is presented in Figure 3.3-48.The core properties associated with the sector nearest the affected steam generator and those associated with the remaining sector were conservatively combined to obtain average core properties for reactivity feedback calculation.
Further, it was conservatively assumed that the core power distribution was uniform.These two conditions cause underprediction of the reactivity feedback in the high power region near the stuck rod.To verify the conservatism of this method, the reactivity as well as the power distribution was checked for the limiting conditions for the cases analyzed.This core analysis considered the Doppler reactivity from the high fuel temperature near the stuck RCCA, moderator feedback from the high water enthalpy near the stuck RCCA, power redistribution and non-uniform core inlet temperature effects.For cases in which steam generation occurs in the high, flux regions of the core, the effect of void formation was also included.It was determined that the reactivity employed in the kinetics analysis was always larger than the reactivity mh1944<w.wpf:1d441195 3.341
calculated including the above local effects for the statepoints.
These results verify conservatism; i.e., underprediction of negative reactivity feedback from power generation.
Minimum capability for injection of boric acid (2400 ppm)solution corresponding to the most restrictive single failure in the safety injection system.The ECCS consists of the following systems: 1)the passive accumulators, 2)the low head safety injection (residual heat removal)system, 3)the high head safety injection system, and 4)the charging system.Only the charging system and the passive accumulators are modeled for the steam line break accident analysis.Centrifugal Charging pump head degradation of 10%was assumed.The modeling of the safety injection system in LOFTRAN is described in Reference 4.Figure 3.3-49 presents the safety injection flow rates as a function of RCS pressure assumed in the analysis.The flow corresponds to that delivered by one charging pump delivering its full flow to the cold leg header.No credit has been taken for the low concentration borated water, which must be swept from the lines downstream of the RWST prior to the delivery of boric acid to the reactor coolant loops.For this analysis, a boron concentration of 0 ppm for the boron injection tank is assumed.It should be noted that this analysis also considers the operation of the Centrifugal Charging Pump Minimum Flow Isolation Valves.These valves are assumed to close following the receipt of a SI signal and reopen when RCS pressure rises above 2000 psig.The Sl flow rates assumed in the steamline break analysis, graphically shown in Figure 3.349, correspond to Centrifugal Charging Pump Minimum Flow Isolation Valves being in the closed position.For the cases where offsite power is assumed, the sequence of events in the safety injection system is the following.
After the generation of the safety injection signal (appropnate delays for instrumentation, logic, and signal transport included), the appropriate valves begin to operate and the charging pump starts.In 27 seconds, the valves are assumed to be in their final position and the pump is assumed to be at full speed and to draw suction from the RWST.The volume containing the low concentration borated water is swept into core before the 2400 ppm borated water reaches the core.This delay, described above, is inherently included in the modeling.In cases where offsite power is not available, a 30 second delay is assumed to start the EDGs and to commence loading the necessary safety injection equipment onto them.mal 944<w.wpf:1d~1195 3.3<2 D.Design value of the steam generator heat transfer coefficient including allowance for fouling factor.E.Four combinations of break sizes and initial plant conditions have been considered in determining the core power transient which can result from large area pipe breaks.a.Complete severance of a pipe downstream of the steam flow restrictor with the plant initially at no load conditions and all reactor coolant pumps running.b.Complete severance of a pipe inside the containment at the outlet of the steam generator with the same plant conditions as above.c.Case (a)above with loss of off-site power simultaneous with the generation of the Safety Injection Signal (loss of AC power results in coolant pump coastdown).
d.Case (b)above with the loss of off-site power simultaneous with the Safety Injection Signal.A fifth case, in which the spurious opening of a steam dump, relief, or safety valve occurs, was considered.
An evaluation concluded that the DNBR remains above the limit value for this case.e.A break equivalent to a steam flow of 247 Ibs per second at 1100 psi from one steam generator with off-site power available.
F.Power peaking factors corresponding to one stuck RCCA and nonuniform core inlet coolant temperatures are determined at end of core life.The coldest core inlet temperatures are assumed to occur in the sector with the stuck rod.The power peaking factors account for the effect of the local void in the region of the stuck control assembly during the return to power phase following the steam line break.This void in conjunction with the large negative moderator coefficient partially offsets the effect of the stuck assembly.The power peaking factors depend upon the core power, temperature, pressure, and flow, and, thus, are different for each case studied.The analyses assumed initial hot shutdown conditions at time zero since this represents the most pessimistic initial condition.
Should the reactor be just critical or operating at power at the time of a steam line break, the reactor will be tripped by the normal overpower protection system when power level mal 944<w.wpf:1d441195 3.3-43
reaches a trip point.Following a trip at power the RCS contains more stored energy than at no-load, the average coolant temperature is higher than at no-load and there is appreciable energy stored in the fuel.Thus, the additional stored energy is removed via the cooldown caused by the steam line break before the no-load conditions of RCS temperature and shutdown margin assumed in the analyses are reached.After the additional stored energy has been removed, the cooldown and reactivity insertions proceed in the same manner as in the analysis which assumes no-load condition at time zero.In addition, since the initial steam generator water inventory is greatest at no-load, the magnitude and duration of the RCS cooldown are more severe than steam line breaks occurring at power.G.In computing the steam flow during a steam line break, the Moody Curve (Reference 11)for fl/D=0 is used.The fast acting steamline isolation valves are assumed to close in less than 11 seconds from receipt of actuation signal.The 11 second closure time of the isolation valves is based upon the actuating signal being generated by the steam flow in two steam lines-high coincident with steam line pressure-low functions.
For breaks downstream of the isolation valves, closure of all valves would completely terminate the blowdown.For any break, in any location, no more than one steam generator would experience an uncontrolled blowdown even if one of the isolation valves fails to close.Results The limiting case for Cases a through e was shown to be the double-ended rupture located upstream of the flow restrictor with offsite power available.
Table 3.3-7 lists the limiting statepoint for this worst case.The results presented are conservative indication of the events which would occur assuming a steam line rupture since it is postulated that all of the conditions described above occur simultaneously.
Figures 3.3-50 through 3.3-53 show the RCS transients and core heat flux following a main steam line rupture (complete severance of a pipe)upstream of the flow restrictor at initial no-load condition.
The sequence of events for this transient is presented in Table 3.3-8.Offsite power is assumed available so that full reactor coolant flow exists.The transient shown assumes an uncontrolled steam release from only one steam generator.
Should the core be critical at near zero power when the rupture occurs the initiation of safety injection by high differential pressure between any steamline and the remaining steamlines or by high steam flow signals in coincidence with either low-low RCS temperature or low steam line mA1944<w.wpt:1d441195 pressure will trip the reactor.Steam release from more than one steam generator will be prevented by automatic trip of the fast acting isolation valves in the steam lines by high containment pressure signals or high.steam flow coincident with low steamline pressure or low-low T,.Even with the failure of one valve, release is limited to approximately 13 seconds for the other steam generators while the one generator blows down.The steam line stop valves are assumed to be fully closed in less than 11 seconds from receipt of a closure signal (steam flow in two steam lines-high coincident with steam line pressure-low).
As shown in Figure 3.3-53, the core attains criticality with the RCCAs inserted (with the design shutdown assuming one stuck RCCA)before boron solution at 2400 ppm enters the RCS.A peak core power less than the nominal full power value is attained.The calculation assumes the boric acid is mixed with, and diluted by the water flowing in the RCS prior to entering the reactor core.The concentration after mixing depends upon the relative flow rates in the RCS and in the safety injection system.The variation of mass flow rate in the RCS due to water density changes is included in the calculation as is the variation of flow rate in the safety injection system due to changes in the RCS pressure.The safety injection system flow calculation includes the line losses in the system as well as the pump head curve.Note that since the RCS pressure (Figure 3.3-51)drops below 2015 psia and never repressurizes above that value, the automatic operation to open the Centrifugal Charging Pump Minimum Flow Isolation Valves would not occur during this event.Therefore, there would be not reduction in Sl flow below that assumed in the safety analysis.The assumed steam release for an accidental depressurization of the main steam system (Case e)is the maximum capacity of any single steam dump, relief, or safety valve.Safety injection is initiated automatically by low pressurizer pressure.Operation of one centrifugal charging pump is assumed.Boron solution at 2400 ppm enters the RCS providing sufficient negative reactivity to prevent core damage.The transient is quite conservative with respect to cooldown, since no credit is taken for the energy stored in the system metal other than that of the fuel elements or the energy stored in the other steam generators.
Since the transient occurs over a period of about 5 minutes, the neglected stored energy is likely for this event to have a significant effect in slowing the cooldown.The DNB transient is bounded by the limiting case for a steamline rupture.The DNB analysis for the limiting case (double-ended rupture located upstream of the flow restrictor) showed that the minimum DNBR remained above the limit value.Conclusions The analysis has shown that the criteria stated earlier are satisfied.
mA1944<w.wpf:1d~1195 3.3-45 Although DNB and possible clad perforation following a steam pipe rupture are not necessarily unacceptable and not precluded by the criteria, the above analysis, in fact, shows that no DNB occurs for the rupture (includiog.an accidental depressurization of the main steam system)assuming the most reactive RCCA stuck in its fully withdrawn position.3.3.5.7 Rupture of Control Rod Drive Mechanism Housing (RCCA Ejection)This accident is defined as the mechanical failure of a control rod mechanism pressure housing resulting in the ejection of a RCCA and drive shaft.The consequence of this mechanical failure, in addition to being a small break loss-of-coolant accident, is a rapid positive reactivity insertion together with an adverse core power distribution, possibly leading to localized fuel rod damage.This event has been analyzed as part of the 30%SGTP Program to address the reduction in RCS flow due to the increase in the SGTP level.If an RCCA ejection accident were to occur, a fuel rod thermal transient which could cause DNB may occur together with limited fuel damage.The amount of fuel damage that can result from such an accident will be governed mainly by the worth of the ejected RCCA and the power distribution attained with the remaining control rod pattern.The transient is limited by the Doppler reactivity effects of the increase in the fuel temperature and is terminated by reactor trip actuated by neutron flux signals, before conditions are reached that can result in damage to the reactor coolant pressure boundafy, or significant disturbances in the core, its support structures or other reactor pressure vessel internals which would impair the capability to cool the core.The neutron flux response to a continuous reactivity insertion is characterized by a very fast flux increase terminated by the reactivity feedback effect of the Doppler coefficient, This self limitation of the power burst is of primary importance since it limits the power to a tolerable level during the delay time for protective action.Should a RCCA Ejection accident occur, the following automatic features of the RPS are available to terminate the transient, The source-range high neutron flux reactor trip is actuated when either of the independent source-range channels indicates a neutron flux level above a preselected manually adjustable setpoint.This trip function may be manually bypassed when either intermediate-range flux channel indicates a flux level above a specified level.It is automatically reinstated when both intermediate-range channels indicate a flux level below a specified level.The intermediate-range high neutron flux reactor trip is actuated when either of two independent intermediate-range channels indicates a flux level above a preselected manually adjustable setpoint."7his trip function may be manually bypassed when two of the four power-range channels give readings above mA1944<w.wpf:1d~1195 3.3<6
approximately 10%of full power and is automatically reinstated when three of the four channels indicate a power below this value.The power-range high neutron flux reactor trip (low setting)is actuated when two-out-of-four power-range channels indicate a power level above approximately 25%of full power.This trip function may be manually bypassed when two of the four power-range channels indicate a power level above approximately 10%of full power and is automatically reinstated when three of the four channels indicate a power level below this value.d.The power-range high neutron flux reactor trip (high setting)is actuated when two-out-of-four power-range channels indicate a power level above a preset setpoint (typically 109%of full power).This trip function is always active.e.The high nuclear flux rate reactor trip is calculated when the positive rate of change of neutron flux on two-out-of-four nuclear power-range channels indicates a rate above the preset setpoint.This trip function is always active.Due to the extremely low probability of a RCCA Ejection accident, this event is classified as an ANS Condition IV event (Limiting Fault).The ultimate acceptance criteria for this event is that any consequential damage to either the core or the RCS must not prevent long-term cooling, and that any offsite dose consequences must be within the guidelines of 10 CFR 100.To demonstrate compliance with these requirements, it is sufficient to show that the RCS pressure boundary remains intact, and that no fuel dispersal into the coolant, gross lattice distortions, or severe shock waves will occur in the core.Therefore, the limiting criteria is described in Reference 12,and summarized below: A.Average fuel pellet enthalpy at hot spot below 225 caVg for unirradiated fuel and 200 caVg for irradiated fuel.B.Average clad temperature at the hot spot below the temperature at which clad embrittlement may be expected (3000'F).C.Peak reactor coolant pressure less than that which could cause stresses to exceed the faulted condition stress limits.D.Fuel melting will be limited to less than ten percent 10%of the fuel volume at the hot spot even if the average fuel pellet enthalpy is below'the limits of criterion A above.The analysis performed is to bound the parameters associated with the increased SGTP level of 30%.mA1944<w.wpf:1d~1 185 3.3<7 Method of Anal sis The calculation of the RCCA ejection transient is performed in two stages, first an average core channel calculation and then a hot region calculation.
The average core calculation is performed using spatial neutron kinetics methods to determine the average power generation with time including the various total core feedback effects, i.e., Doppler reactivity and moderator reactivity.
Enthalpy and temperature transients in the hot spot are then determined by multiplying the average core energy generation by the hot channel factor and performing a fuel rod transient heat transfer calculation.
The power distribution calculated without feedback is pessimistically assumed to persist throughout the transient.
A detailed discussion of the method of analysis can be found in Reference 12.Average Core Analysis The spatial kinetics computer code, TWINKLE, is used for the average core transient analysis.This code solves the two group neutron diffusion theory kinetic equation in one, two or three spatial dimensions (rectangular coordinates) for six delayed neutron groups and up to 2000 spatial points.The computer code includes a detailed multi-region, transient fuel-clad-coolant heat transfer model for calculation of pointwise Doppler and moderator feedback effects.In this analysis, the code is used as a one dimensional axial kinetics code since it allows a more realistic representation of the spatial effects of axial moderator feedback and RCCA movement.However, since the radial dimension is missing, it is still necessary to employ very conservative methods (described below)of calculating the ejected rod worth and hot channel factor.Further description of TWINKLE appears in Section 3.3.3.2.Hot Spot Analysis In the hot spot analysis, the initial heat flux is equal to the nominal times the design hot channel factor.During the transient, the heat flux hot channel factor is linearly increased to the transient value in 0.1 second, the time for full ejection of the rod.Therefore, the assumption is made that the hot spot before and after ejection are coincident.
This is very conservative since the peak after ejection will occur in or adjacent to the assembly with the ejected rod, and prior to ejection the power in this region will necessarily be depressed.
The hot spot analysis is performed using the detailed fuel and clad transient heat transfer computer code, FACTRAN.This computer code calculates the transient temperature distribution in a cross section of a metal clad UO, fuel rod, and the heat flux at the surface of the rod, using as input the nuclear power versus time and the local coolant'conditions.
The zirconium-water reaction is explicitly represented, and all material properties are represented as functions of temperature.
A conservative radial power.distribution is used.within the fuel rod.m:51944<w.wpf:1d441195 3.3<8
FACTRAN uses the Jens-Lottes or Dittus-Boelter correlation (References 8 and 15, respectively) to determine the film heat transfer before DNB, and the Bishop-Sandberg-Tong correlation (see Reference 13)to determine the film boiling coefficient after DNB.The Bishop-Sandberg-Tong correlation is'consewatively used assuming zero bulk fluid quality.The DNB ratio is not calculated, instead the code is forced into DNB by specifying a conservative DNB heat flux.The gap heat transfer coefficient can be calculated by the code;however, it is adjusted in order to force the full power steady-state temperature distribution to agree with the fuel heat transfer design codes.Further description of FACTRAN appears in Section 3.3.3.2.A detailed three-dimensional calculation of a worst case scenario (Reference 12)demonstrates an upper limit to the number of rods-in-DNB for the RCCA Ejection accident as 10%.Since the severity of the Cook Nuclear Plant Unit 1 analysis does not exceed this worst case analysis, the maximum number of rods in DNB following a RCCA Ejection will be less than 10%, although neither the number of rods in DNB nor the minimum DNBR value is explicitly calculated in the Cook Nuclear Plant Unit 1 analysis.The most limiting break size resulting from a RCCA Ejection will not be sufficient to uncover the core or cause DNB at any later time.Since the maximum number of fuel rods experiencing DNB is limited to 10%, the fission product release will not exceed that associated with the guidelines of 10 CFR 100.System Overpressure Analysis Because safety limits for fuel damage specifled earlier are not exceeded, there is little likelihood of fuel dispersal into the coolant.The pressure surge may therefore by calculated on the basis of conventional heat transfer from the fuel and prompt heat generation in the coolant.The pressure surge is calculated by first performing the fuel heat transfer calculation to determine the average and hot spot heat flux versus time.Using this heat flux data, a THING IV calculation is conducted to determine the volume surge.Finally, the volume surge is simulated in the LOFTRAN computer code.This code calculates the pressure transient taking into account fluid transport in the RCS and heat transfer to the steam generators.
No credit is taken for the possible pressure reduction caused by the assumed failure of the control rod pressure housing.Input parameters for the analysis are conservatively selected on the basis of values calculated for this type of core.The more important parameters are discussed below.Table 3.3-9 presents the parameters used in this analysis.mA1944<w.wpf:1d441195 3.3-49
Ejected Rod Worths and Hot Channel Factors The values for ejected rod worths and hot channel factors are calculated using either three dimensional static methods or by a synthesis method employing one dimensional and two dimensional calculations.
Standard nuclear design codes are used in the analysis.No credit is taken for the flux flattening effects of reactivity feedback.The calculation is performed for the maximum allowed bank insertion at a given power level, as determined by the rod insertion limits.Adverse xenon distributions are considered in the calculation to provide worst case results.Appropriate margins are added to the ejected rod worth and hot channel factors to account for any calculational uncertainties, including an allowance for nuclear power peaking due to densification.
Power distribution before and after ejection for a worst case can be found in Reference 12.During plant startup physics testing, ejected rod worths and power distributions are measured in the zero and full power configurations and compared to values used in the analysis.Experience has shown that the ejected rod worth and power peaking factors are consistently overpredicted in the analysis.Reactivity Feedback Weighting Factors The largest temperature rises, and hence the largest reactivity feedbacks occur in channels where the power is higher than average.Since the weight of a region is dependent on flux, these regions have high weights.This means that the reactivity feedback is larger than that indicated by a simple channel analysis.Physics calculations have been carried out for temperature changes with a flat temperature distribution, and with a large number of axial and radial temperature distributions.
Reactivity changes were compared and effective weighting factors determined.
These weighting factors take the form of multipliers which, when applied to single channel feedbacks, correct them to effective whole core feedbacks for the appropriate flux shape.In this analysis, since a one dimensional (axial)spatial kinetics method is employed, axial weighting is not necessary if the initial condition is made to match the ejected rod configuration.
In addition, no weighting is applied to the moderator feedback.A conservative radial weighting factor is applied to the transient fuel temperature to obtain an effective fuel temperature as a function of time accounting for the missing spatial dimension.
These weighting factors have also been shown to be conservative compared to three dimensional analysis (Reference 12).Moderator and Doppler Coefficient The critical boron concentrations at the beginning of life and end of life are adjusted in the nuclear code in order to obtain moderator density coefficient curves which are conservative m%1 944<w.wpl:
1 d441195 3.3-50 compared to actual design conditions for the plant.As discussed above, no weighting factor is applied to these results.The resulting moderator temperature coefficient is at least+5 pcm/'F at the appropriate zero or, full power nominal average temperature, and becomes less positive for higher temperatures.
This is necessary since the TWINKLE computer code utilized in the analyses is a diffusion-theory code rather than a point-kinetics approximation and the moderator temperature feedback cannot be artificially held constant with temperature.
The Doppler reactivity defect is determined as a function of power level using a one dimensional steady-state computer code with a Doppler weighting factor of 1.0.The Doppler weighting factor will increase under accident conditions, as discussed above.Delayed Neutron Fraction, P,Calculations of the effective delayed neutron fraction (P,)typically yield values no less than 0.70%at beginning of life and 0.50%at end of life.The accident is sensitive to P,if the ejected rod worth is equal to or greater than p,as in zero power transients.
In order to allow for future cycles, pessimistic estimates of p,of 0.50%at beginning of a cycle and 0.40%at end of a cycle were used in the analysis.Trip Reactivity Insertion The trip reactivity insertion assumed is given in Table 3.3-9 and includes the effect of one stuck RCCA adjacent to the ejected rod.These values are reduced by the ejected rod reactivity.
The shutdown reactivity was simulated by dropping a rod of the required worth into the core.The start of rod motion occurred 0.5 seconds after the high neutron flux trip points is reached before significant shutdown reactivity is inserted into the core.This is particularly important conservatism for hot full power accidents.
The minimum design shutdown margin available for this plant at HZP may be reached only at end of life in the equilibrium cycle.This value includes an allowance for the worst stuck rod, an adverse xenon distribution, conservative Doppler and moderator defects, and an allowance for calculational uncertainties.
Physics calculations have shown that the effect of two stuck RCCAs (one of which is the worst ejected rod)is to reduce the shutdown by about an additional 1%hk/k.Therefore, following a reactor trip resulting from an RCCA ejection accident, the reactor will be subcrltical when the core returns to HZP..&Depressurization calculations have been performed assuming the maximum possible size break (2.75 inch diameter)located in the reactor pressure vessel head.The results show a rapid pressure drop and a decrease in system water mass due to the break.The ECCS is actuated on low pressurizer pressure within one minute after the break.The RCS pressure continues to drop and reaches saturation (1100 to 1300 psi depending on the system temperature) in about two to three minutes.Due to the large thermal inertia of primary and rn A1 944<w.wpf:1d441 195 3.3-51
secondary system, there has been no significant decrease in the RCS temperature below no-load by this time, and the depressurization itself has caused an increase in shutdown margin by about 0.2%d,k/k due to the, pressure coefficient.
The cooldown transient could not absorb the available shutdown margin until more than 10 minutes after the break.The addition of borated safety injection flow (supplied from the RWST)starting one minute after the break is much more than sufficient to ensure that the core remains subcritical during the cooldown.Reactor Protection Reactor protection for a rod ejection is provided by high neutron flux trip (high and low setting)and high rate of neutron flux increase trip although the analysis modeled the high neutron flux trip (high and low setting)only.These protection functions are part of the reactor trip system.No single failure of the reactor trip system will negate the protection functions required for the rod ejection accident, or adversely affect the consequences of the accident.Results Table 3.3-9 summarizes the results.Cases are presented for both beginning and end of life at zero and full power.A.Beginning of Cycle, Full Power Control Bank D was assumed to be inserted to its insertion limit.The worst ejected rod worth and hot channel factor were conservatively calculated to be 0.15%b,k/k and 6.8 respectively.
The peak clad average temperature was 2299'F.The peak spot fuel center temperature reached melting, conservatively assumed at 4900'F.However, melting was restricted to less than 10%of the pellet.B.Beginning of Cycle, Zero Power For this condition, Control Bank D was assumed to be fully inserted and banks B and C were at their insertion limits.The worst ejected rod is located in Control Bank D and has a worth of 0.65%d,k/k and a hot channel factor of 12.0.The peak clad average temperature reached 2130'F, the fuel center temperature was 3120'F.C.End of Cycle, Full Power Control Bank D was assumed to be inserted to its insertion limit.The ejected rod worth and hot channel factors were conservatively calculated to be 0.19%m:0944<w.wpf:1d~1195 3.3-52 d,k/k and 7.1 respectively.
This resulted in a peak clad average temperature of 2245'F.The peak hot spot fuel center temperature reached melting at 4800'F.However, melting was.restricted to less than 10%of the pellet.D.End of Cycle, Zero Power The ejected rod worth and hot channel factor for this case were obtained assuming Control Bank D to be fully inserted and banks B and C at their insertion limits.The results were 0.75%d,k/k and 19.0 respectively.
The peak clad average and fuel center temperatures were 2322'F and 3258'F.The , Doppler weighting factor for this case is significantly higher than for the other cases due to the vefy large transient hot channel factor.For all the cases analyzed, average fuel pellet enthalpy at the hot spot remains below 200 cal/g.The nuclear power and hot spot fuel and clad temperature transients for two cases (end of life zero power and end of life full power)are presented in Figures 3.3-54 through 3.3-57.The ejection of an RCCA constitutes a break in the RCS, located in the reactor pressure vessel head.Following the RCCA ejection, the operator would follow the same emergency instructions as for any other LOCA to recover from the event.Pressure Surge A detailed calculation of the pressure surge for an ejection worth of one dollar at beginning of life, hot full power, indicates that the peak pressure does not exceed that which would cause stress to exceed the faulted condition stress limits (Reference 12).Since the severity of the present analysis does not exceed the"worst case" analysis, the accident for Cook Nuclear Plant Unit 1 will not result in an excessive pressure rise or further adverse effects to the RCS.Lattice Deformatf'ons A large temperature gradient will exist in the region of the hot spot.Since the fuel rods are free to move in the vertical direction, differential expansion between separate rods cannot produce distortion.
However, the temperature gradients across individual rods may produce a differential expansion tending to bow the midpoint of the rods toward the hotter side of the rod.Calculations have indicated that this bowing would result in a negative reactivity effect at the hot spot since Westinghouse cores are under-moderated, and bowing will tend to increase the under-moderation at the hot spot.In practice, no significant bowing is anticipated, since the structural rigidity of the core is more than sufficient to withstand the forces produced.Boiling in the hot spot region would produce a net flow away from that region.However, the heat mA1944<w.wpf:1d~1195 3.3-53
from the fuel is released to the water relatively slowly, and it is considered inconceivable that cross flow will be sufficient to produce significant lattice forces.Even if massive and rapid boiling, sufficient to distort the lattice, is hypothetically postulated, the large void fraction in the hot spot region would produce a reduction in the total core moderator to fuel ratio, and a large reduction in this ratio at the hot spot.The net effect would therefore be a negative feedback.lt can be concluded that no conceivable mechanism exists for a net positive feedback resulting from lattice deformation.
In fact, a small negative feedback may result.The effect is conservatively ignored in the analysis.Conclusions Even on a pessimistic basis, the analyses indicate that the described fuel and clad limits are not exceeded.It is concluded that there is no likelihood of sudden fuel dispersal into the coolant.Since the peak pressure does not exceed that which would cause stresses to exceed the faulted condition stress limits, it is concluded that there is no likelihood of further consequence to the RCS.The analyses have demonstrated the fission product release as a result of fuel rods entering DNB is limited to less than 10%of the fuel rods in the core.3.3.5.8 Steamline Break Mass/Energy Releases Inside Containment The non-LOCA discussion regarding the reanalysis of the steamline break mass and energy releases inside containment can be found in Sections 3.5.4 and 3.5.5.It should be noted that the changes associated with the SGTP Program for Unit 1, i.e., RCS flow reduction, reduced pnmary-to-seconda1y heat transfer capability, and reduction in the rated thermal power, are less limiting parameters relative to the assumptions currently made for the M/E release calculations following a SLB inside cont'ainment.
The parameter changes associated with the SGTP program do not warrant reanalysis of this event.However, evaluations are currently in place (References 7 and 17)to address several non-conservative assumptions in the analysis.A reanalysis effort was undertaken for the steamline break mass and energy releases inside containment as part of the SGTP Program, such that the Reference 7 and 17 evaluations will no longer be required.3.3.6 Conclusions of the Non-LOCA Safety Evaluation The non-LOCA safety analyses and evaluations presented in this section support the operation of Donald C.Cook Nuclear Plant Unit 1 with SGTP, as described in Table 3.3-1 (Cases 1 and 2).References 1.Ellenberger S.L.et al.,"Design Bases for the Thermal Overpower hT and Thermal Overtemperature dT Trip Functions," WCAP-8746, March, 1977.m:11944<w.wpf:1d~1195 3.3-54 2.Friedland, A.J., Ray, S.,"Revised Thermal Design Procedure," WCAP-11397-A, April, 1989.3.Hargrove, H.G.,"FACTRAN-A FORTRAN-IV Code for Thermal Transients in a UO, Fuel Rod," WCAP-7908, June, 1972.4.Burnett, T, W.T., et al.,"LOFTRAN Code Description," WCAP-7907-A, April 1, 1984.5.Risher, D.H., Jr., and Barry, R.F.,"TWINKLE-a Multi-Dimensional Neutron Kinetics Computer Code," WCAP-8028-A, January, 1975.6.Friedland, A.J.and Ray, S.,"Improved THINC-IV Modeling for PWR Core Design," WCAP-12330-P, August 1989.7."American Electric Power Service Corporation, Donald C.Cook Nuclear Power Plant Units 1 and 2, Increased U r 8 Lower Com artment S ra Delive Times," W Letter AEP-94-712, June 13, 1994.8.W.H.Jens, P.A.Lottes,"Analysis of Heat Transfer, Burnout.Pressure Drop, and Density Data for High-Pressure Water," U.S.AEC Report ANL<627 (1951).9.Haessler, R.L., et.al.,"Methodology for the Analysis of the Dropped Rod Event," WCAP-11394-P-A and WCAP-11395-NP-A, January 1990.10.Baker, L., and Just, L.,"Studies of Metal Water Reactions of High Temperatures, III Experimental and Theoretical Studies of the Zirconium-Water Reaction," ANL-6548, Argonne National Laboratory, May, 1962.11.Moody, F.S.,"Transitions of the ASME, Journal of Heat Transfer," Figure 3, Page 134, February 1965.12.Risher, D.H., Jr.,"An Evaluation of the Rod Ejection Accident of Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," WCAP-7588, Revision 1A.13.Bishop, A.A., Sandberg, R.O., and Tong, L.S.,"Forced Convection Heat Transfer at High Pressure After the Critical Heat Flux," ASME 65-HT-31, August, 1965.14.Donald C.Cook Nuclear Plant Unit 1 Updated Final Safety Analysis Report, USNRC Docket Number 50-31 5, updated through 1993.mA1944<w.wpf:
1 d441195 3.3-55 15.F.W.Dittus, C.N.Boelter,"Heat Transfer in Automobile Radiators of the Tubular Type," Calif.Univ.Publication in Eng., 2 of 13, 4433-461 (1930).16.Christensen, J.A., et.al.,"Melting Point of Irradiated Uranium Dioxide," Transactions of the American Nuclear Socie 7, 1964.17."American Electric Power Service Corporation, Donald C.Cook Nuclear Plant Units 1 and 2, Feedwater Isolation Valve Evaluation Su ort" W Letter AEP-93-528, April 8, 1993.18.Letter from William O.Long, Sr.(USNRC)to Eugene E.Fitzpatrick (AEPSC),

Subject:

Amendment Nos.158 and 142 to Facility Operating License Nos.DPR-58 and DPR-74 (TAC Nos.80262 and 80263), dated November 20, 1991.19.SECL-91-429,"Donald C.Cook Units 1 and 2 Main Steam Safety Valve Lift Tolerance Relaxation," March 1992.m:51944<w.wpt:1d441195 3.3-56 TABLE 3.3-1 DONALD C.COOK NUCLEAR PLANT UNIT 1 NSSS PERFORMANCE PARAMETERS USED IN NON-LOCA SAFETY ANALYSES Parameter NSSS Power, MWt (30%SGTP Program)Case 1 Case 2 3262 3262 Case 3 3425 Case 4 3425 (Rerating Program)"'ore Power, MWt RCS Flow, gpm/loop"'inimum Measured Flow, total gpm@RCS Tem erature'F Core Outlet 3250 83200 339,100 589.7 3250 83200 339,100 611.9 341 3 366,400 t')583.6 3413 88500 366,400 t'~614.0 Vessel Outlet 586.8 Core Average 555.8 Vessel Average 553.0 Vessel/Core Inlet 519.2 Steam Generator Outlet 518.9 Zero Load 547.0 609.1 579.4 576.3 543.5 543.2 547.0 580.7 549.7 547.0 513.3 513.1 547.0 611.2 581.8 578.7 546.2 546.0 547.0 RCS Pressure, psia 2250 or 2100 2250 or 2100 2250 or 2100 2250 or 2100 Steam Pressure, psia Steam Flow (10'b/hr total)Feedwater Temp.,'F SG Tube Plugging,%595 14.12 434.8 30 749 14.17 434.8 30 603 14.98 10 820 15.07 442.~10 Cook Unit 1 is not licensed to operate at the rerated conditions specified by Cases 3 and 4 with 30%steam generator tube plugging (SGTP)levels.However, several events that were previously performed using these conditions were subsequently evaluated to support the 30%SGTP program.Hence, the rerated conditions are aho specified in this table for completeness.
RCS Flow (Thermal Design Fhw)-The conservatively low fhw used for thermal/hydraulic design.The design parameters listed above are based upon this flow.Minimum Measured Flow-The flow specified in the Technical Specifications which must be confirmed or exceeded by the flow measurements obtained during phnt startup and is the flow used in reactor core DNB analyses for phnts applying the Revised Thermal Design Procedure.
MMF based upon a 1.9%flow measurement uncertainty..Analyses also bound a MMF of 341,100 gpm which reflects a 2.5%flow measurement uncertainty.
A MMF of 366,400 gpm was assumed in the Rerating Program analyses.A safety evaluation was performed to support a reduction of MMF to 361,600 gpm (SECL-90-280).
m:51 944<w.wpf:11441195 3.3-57 f i 1 II II TABLE 3.3-2 TRIP POINTS AND TIME DELAYS TO TRIP ASSUMED IN NON-LOCA ACCIDENT, ANALYSIS'ri Function Limiting Trip Point Assumed Time Delay~Seconds Power range high neutron flux, high setting Power range high neutron flux, low setting Overtemperature dT Overpower hT High pressurizer pressure Low pressurizer pressure High pressurizer water level Low reactor coolant flow (From loop flow detectors)
Undervoltage trip Low-low steam generator level High steam generator level Turbine Trip Feedwater Isolation 118 percent 35 percent Variable, see Figure 3.3-1 through 3.3-4 and Table 3.3-3 Variable, see Figure 3.3-1 through 3.3-4 and Table 3.3-3 2420 psig 1825 psig 100%NRS 87 percent loop flow 0.0 percent of narrow range level span 82 percent of narrow range level span 0.5 0.5 8.0'.0 2.0 2.0 2.0 1.0 1.5 2.0 11.0 Toute time delay (including RTD bypass loop fluid transport delay effect, bypass loop piping thermal capacity, RTD time response, and trip circurt, channel electronics dehy)from the time the temperature difference in the coolant hops exceeds the trip setpoint until the rods are free to fall.The time delay assumed in the analysis supports the 6 second response time of the RTD time response, trip circuit delays, and channel electronics dehy presented in the Technical Specifications.
No explicit value assumed in the analysis.Undervoltage trip setpoint assumed reached at initiation of analysis.The control rod scram time to dashpot is 2.4 seconds Overpower hT reactor trip was assumed in the steamline break masslenergy release outside containment calculations.
mA1944<w.wpt:1d 441195 3.3-58 t,l , It TABLE 3.3-3 OTLT AND OPbT SETPOINT EQUATION AND SAFETY ANALYSIS LIMIT COEFFICIENT VALUES Overtemperature hT equation: OT~T 5 4TO[K)-K2[](T-T)+(P-P)fi(>l)1+t~s where, K,=1.35 K=0.023 22 seconds 4 seconds s=Laplace transform operator T'553.0 to 576.3'F 0.0011 P=2100 or 2250 psia f,(d I): Dead-band:
from-37 to+3%6,l Positive Wing: 2.34/J'/od,l for each percent Al>+3%6,l Negative Wing: 0.33'/d/&I for each percent bl<-37%5,l Overpower dT equation: OPBT-~To[K4 K,['T-K,(T-T")-f,(~l)1+c s where, K, 73 s Tll PS f (bl): 1.172 0.0177;this gain is not modeled in the non-LOCA safety analyses 10 seconds Laplace transform operator 553.0 to 563.0'F 0.0015 2100 or 2250 psia 0 mal 944<w.wpf:1d~1195 3.3-59 TABLE 3.3.4

SUMMARY

OF INITIAL CONDITIONS AND COMPUTER CODES USED Faults Uncontrolled Rod Cluster Assembly Bank Withdrawal from a Subcritical Condbon Computer Codes Utilized TWINKLE FACTRAN THING IV Moderator Temperature J~P~F Moderator Density~hK/ml~cc~Do er Refer to Section 3.3.5.1 Min (I)Reactivi Coefficients Assumed DNB Correlation WQIWRB-I See Section 3.3.4.3 Revised Thormal Design Procedure No Reactor Initial NSSS Thermal Power~Oo ol MWl Vessel Vessel Coolant~Flow GPM 146,432 Average Temperature
~F Pressurizer Pressure~PGIA r 2033 Uncontrolled Rod Cluster Assembly Bank Withdrawal At Power (2)Rod Cluster Control Assembly Misalignment LOFTRAN LOF TRAN THING IV+5 NA'in and Max (3)NA WRB-1 WRB-1 Yes Yes 3270 1962 327 3270 339,100 339,100 576.3 564.58 549.93 576.3 2100 2100 Uncontrolled Boron Dilution NA NA NA NA NA 3425 0 NA NA NA Loss of Forced Reactor Coolant Flow LOFTRAN FACTRAN THING IV t5 NA WRB-1 Yes 3270 339.100 576.3 2100 Locked Rotor (Peak Pressure)Locked Rotor (Peak Clad Temp)Locked Rotor (Rods-in-DN8)
LOF TRAN LOFTRAN FACTRAN LOFTRAN FACTRAN THING IV+5+5+5 NA NA NA NA NA WRB-1 NA NA Yes 3335 3335 3270 332,800 332,800 339,100 581.4 581.4 576.3 2317 2100'A-Not Applicable (1)Minimum Doppler power defect (porn//.power)
=-9.55+0.0350 where 0 is in%power.(2)Multiple power levels, Tavg, and reactivity feedback cases were examined.(3)Maximum Doppler power defect (pcmP/G power)=-19.4+0.0650.(4)Minimum and Maximum reacbvity feedback cases were examined.m tt 944.4w.wpf:1d-041195 3.3-60 TABLE 3.3.4 (continued)

SUMMARY

OF INITIAL CONDITIONS AND COMPUTER CODES USED Faults Computer Codes Utilized Reactivi Coefficients Assumed Moderator Moderator Temperature Density Qc~ml'F~alt'Jr~Ice DoODler DNB Correlation Revised Thermal Design Plooedule Initial NSSS Thermal Power~Ou ul MWt Reactor Vessel Coolant~Flow GPM Vossel Average Temperature
~F Pressurizer Pressure~PSIA Loss of Electrical Load andfor Turbine Trip (4)LOFTRAN+5.54 Max and Min WRB-1 Yes 3262 339,100 576.3 2100 Excessive Heat Removal Due to Feedwater System Malfunction (5)LOF TRAN Loss of Normal Feedwator (5)LOFTRAN+5.54 Min NA WRB-1 Yes 3494 3425 0 354,000 366,400 551.5 578.7 547 2285 2100 Excess Load Increase Incident (5)LOFTRAN NA 0 and.54 Max and WRB-1 Min Yes 3425 366,400 578.7 2100 Loss of Olfsite Power lo the Station Auxiliaries (5)LOFTRAN+5 NA NA 354,000 542.5 2285 Rupture ol a Steam Pipe Rupture of a Control Rod Drive Mechanism Housing LOF TRAN THING IV TWINKLE FACTRAN See Figure NA 3.3w47 See Section NA 3.3.5.7 See Figure W.3 3.3-48 NA 3335 0 332,800 332,800 146,432 547 581.4 547 2100'A-Not Applicable (1)Minimum Doppler power defecl (pcs%%dpower)
=-9.55+0.035Q where Q is in%power.(2)Multiple power levels, Tavg, and reactivity feedback cases were examined.(3)Maximum Doppler power defect (pcml%%d power)=~19.4+0.065Q.(4)Minimum and Maximum reactivity feedback cases were examined.(5)Values presented conespond to the respective rerating analysis.Subsequent evaluabons support the 30%%d SGTP parameters given as Cases 1 and 2 of Table 3.3-1.m'.11944.4 w.wpf:1d.041195 3.3-61 TABLE 3.3-5 SEQUENCE OF EVENTS FOR LOSS OF FLOW AND LOCKED ROTOR ACCIDENTS Accident Complete Loss of Flow Partial Loss of Flow Locked Rotor Event All pumps lose power and begin coasting down, undervoltage trip signal generated Rods begin to drop Minimum DNBR occurs One operating pump loses power and begins coasting down Low reactor coolant flow trip setpoint reached in faulted loop Rods begin to drop Minimum DNBR occurs One pump rotor seizes Low reactor coolant flow trip setpoint reached in faulted loop Rods begin to drop Maximum percentage of rods in DNB predicted Maximum RCS pressure occurs Maximum clad temperature occurs Time sec.0.0 1.50 3.40 0.0 1.74 2.74 3.90 0.0 0.04 1.04 2.6 3.20 3.49 m:51944<w.wpf:1d441195 3.3-62 TABLE 3.3-6 SEQUENCE OF EVENTS FOR LOSS OF EXTERNAL ELECTRICAL LOAD Case Minimum Feedback with Pressure Control Maximum Feedback with Pressure Control Minimum Feedback without Pressure Control Maximum Feedback without Pressure Control Event Loss of external electrical load OTBT trip setpoint reached Peak RCS pressure occurs Rods begin to drop Minimum DNBR occurs Loss of external electrical load Minimum DNBR occurs Peak RCS pressure occurs Low-low steam generator level trip setpoint reached Rods begin to drop Loss of external electrical load Minimum DNBR occurs High pressurizer pressure trip setpoint reached Rods begin to drop Peak RCS pressure occurs Loss of external electrical load Minimum DNBR occurs High.pressurizer pressure trip setpoint reached Rods begin to drop Peak RCS pressure occurs Time sec.0.0 14.2 15.5 16.2 18.0 0.0 0.0 10.0 68.1 70.1 0.0 0.0 8.4 10.4 12.0 0.0 0.0 8.9 10.9 12.5 mA1944<w.wpf:1d~1195 3.3-63
TABLE 3.3-7 LIMITING STEAMLINE BREAK STATEPOINT DOUBLE ENDED RUPTURE INSIDE CONTAINMENT WITH OFFSITE POWER AVAILABLE Heat Inlet Time Pressure Flux Cold Temp.Flow Boron Reactivity sec psia Fraction'F Hot'F Fraction PPM Percent Density gm/cc 180.2 601.93.228 336.6 463.3 1.0 7.13.001.849 mA1944<w.wpf:1d~1 195 3.3-64 TABLE 3.3-8 TIME SEQUENCE OF EVENTS DOUBLE ENDED RUPTURE INSIDE CONTAINMENT WITH OFFSITE POWER AVAILABLE Event Steam line rupture occurs Low steam line pressure coincident with high steam flow in two steam lines reached Feedwater Isolation (All loops)Criticality attained Steamline Isolation (Loops 2, 3 and 4)Pressurizer empties Sl flow starts Boron from Sl reaches the core Peak heat flux attained Core becomes subcritical Time sec 0.00 2.06 10.06 12.40 13.06 13.20 29.06 39.80 179.2 180.0 m:11 944<w.wpf:1d~1195 3.3-65
TABLE 3.3-9 PARAMETERS USED IN THE ANALYSIS OF THE ROD CLUSTER CONTROL ASSEMBLY EJECTION ACCIDENT Time in Life Power Level (%)Ejected Rod Worth (%b,k)Delayed Neutron Fraction Feedback Reactivity Weighting Trip Reactivity
('iMk)F, Before Rod Ejection F, After Rod Ejection Number of Operational Pumps Maximum Fuel Pellet Average Temperature
('F)Maximum Fuel Center Temperature
('F)Maximum Clad Average Temperature
('F)Maximum Fuel Stored Energy (cal/gm)Fuel Melt in Hot Pellet,%HZP~Be innin 0.65 0.0050 2.071 2.2.50 12.2764 3120 2130 112.7 HFP Be<einning 102 0.15 0.0050 1.30 2.50 6.8 4056 4968 2299<10 HZP End 0.75 0.0040 2.755 2.50 19.2963 3258 2322 122.2 HFP End 102 0.19 0.0040 1.30 2.50 7.1 3969 4872 2245 172.7<10 m%1 944<w.wpf:1d441195 3.3-66 1 I I t 85 80 75 70 65 60~55 CI 50 45 i 1840 psia~oo oooo~oooo~ooooo 2400 psia x 2100 psia~,"'toooo~~~ooo\OPET Trip~ooooo~oooo~oooo~oooo oooo'oooo ooooo SG Safety Open 35 Care Limits OTBT Trip%~30 560 580 600 T('F)620 DONALD C.COOK NUCLEAR PLANT UNIT 1 FIGURE 3.3-1 Illustration of Overtcntperature and Overpower dT Protection Nominal Tavg=576.3'F Nominal Pressure=2100 psia m:51944-5w.wpt:1d441195 3.3%7 85 80 75 70 65~55 CI 50 45 35 l840 psia~k T'ooooo OPhT Trip~o~oo'o o o oooo~o o~ooo~oo'oo Core Limits 2400 psia%50 psia~ooo oooo~oooo//>r SG Safety rx OTBT Trip~Valves Open 30 560 580 600 T,('F)620 DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3-2 Blustration of Overtemperature and Overpower AT protection Nominal Tave=576.3'F Nominal Pressure~2250 psia m:11944.5w.wpt:1dM1195 3.348
85 80 75 70 65~60+55 CI 50 1840 psia%oooooooooooooo 2250 psia%oooo oooo~oooo\2250 psia 1840 psia S\2400 psia oooo~".....OphT Trip~ooooo~ooo~ooo oooo~oo 45 35 Core Limits OTET Trip 30 560 580 600 T,s ('8 620 DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3-3 Illustration of Overtemperature and Overpower 4T Proteeuon Nominal Tavg=553.0'F Nominal Pressure=2250 psia m:11944.5w.wpf:1dO41195 3.3<9 85 80 75 70 65~60~55 50 45 35~sos basso cross SG Safety Valves Open Core Limits OTET Trip 2100 psia 2400 s~sss~oYosooso't i840 psra~oooo\2400psia t i 2100psia i i840 psut\'\\\'\OPET Trip~oss cross~sos~~~~oos 30 580 600 T,g ('8 620 DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3<Illustration of Overtemperature and Overpower hT Protection Nominal Tavg=553.0'F Nominal Pressure=2 l00 psia m:$1944-5w.wpt:1d441195 3.3-70 1.0E+1 CU C E C 0 C 0 C$O n C5c3 z 0.0 1.0E-I 1.0E-2 0 10 Time (s)15 20 25 0.5 I E g 0.4 0 0.3 u m 0.2 Z C C~0.1 O O x 0.0 0 10 Time (s]15 20 25 DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3-5 Nuclear Power and Hot Channel Heat Flux vs.Tiara For , The Rod Withdrawal From Subcritical Event m:119444w.wpf:1d441195 3.3-71
2,400 o 2,QQQ c1,600 E~ID e g 1,200 P Cl 800 10 Time[s]15 20 25 750 u~700 K~650~I~600 0 550 500 0 10 Time[s]15 20 25 DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3-6 Fuel Average and Clad Temperature vs.Time For The Rod Withdrawal From Subcritical Event m:11944.5w.wpt:1d~1195 3.3-72 1.4+1.2~1.0 Q I-0.8~0.6 Q 0.4 z 0.2 0.0 0 Time[s)10 DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3-7 Nuclear Power vs.Time For The RCCA Withdrawal At Power Event, Full Power, 80 PCM/Sec.Insertion Rate Maxiamn Reactivity Feedback m:$1944-5w.wpf:
1 d441195 303 73 2,300 2,250 o 2300 n 2,150 N g 2,100 n.2,050 2,000 0 Time[s]10 1,200 o 1,150 E I m 1,100 l I 1,050 Q.1,000 0 Time[s)10 DONALD C.COOK NUCLEAR PLANT UNIT 1 FIGURE 3.3-8 Pressurizer Pressure and Pressurizer Water Volume vs.Time For The RCCA Withdrawal At Power Event, Full Power, 80 PCM/Sec.Insertion Rate Maximum Reactivity Feedback m&1944-5w.wpf:1d441195 3.3-74
590 u o 580 f-570 I 560 O O 550 0 Time[s]10 4.0 3.5 3.0 2.5 2.0 1.5 0 Time fs]10 DONALD C.COOK NUCLEAR PLANT UNIT 1 FIGURE 3.3-9 Core Average Temperature and DNBR vs.Tiara For Tbe RCCA Wahdrawal At Power Event, Full Power, 80 PCM/Sec.Insertion Rate Maximum Reactivity Feedback m:11944-5w.wpf:1d
~1195 3.3-75 1.41.2 I~1.0 0.I-0.8~0.6 O~0.4 z 0.2 0.0 0 50 150 200 250 300 350 Time[s]DONALD C.COOK NUCLEAR PLANT UNIT 1 FIGURE 3.3-10 Nuclear Power vs.Time For The RCCA Withdrawal At Power Event, Full Power, 4 PCM/Sec.Insertion Rate Maxittmn Reactivity Feedback m&1944-5w.wpf:1dM1195 3.3-76
2,200~2,160 g 2,120 n..a 2,080 5 n-2,040 2,000 0 50 100 150 200 250 300 350 Time fs]1,500 1,400 N>1,300'5 e 1,200 N'C 8 n 1,100 1,000 0 50 100 150 200 250 300 350 Time (s]DONALD C.COOK NUCLEAR I'LAM'NIT 1 FIGURE 3.3-11 Pressurizer Pressure and Pressurizer Water Volume vs.Time For The RCCA Withdrawal At Power Event, Full Power, 4 PCM/Sec.Insertion Rate Maximum Reactivity Feedback m:11944.5w.wpf:1d~1195 303 77 o 595 cL.590 E~$m 585 o 580 575 0 50 100 150 200 250 300 350 Time[s]4.0 3.5 3.0 2.5 2.0 1.5 0 50 150 200 250 300 350 Time[s]DONALD C.COOK NUCLEAR PLANT UNIT 1 FIGURE 3.3-12 Core Average Temperature and DNBR vs.Tune For The RCCA Withdrawal At Power Event, Full Power, 4 PCM/Sec.Insertion Rate Maximum Reactivity Feedback m:$1944-5w.wpl:1dO41195 3.3-78
2.0 I I//m z 1.8 0 E'a 1.7 0venemperature 4T Trip High Neutron Flux Trip 1.6 Min.Feedback Max.Feedback 1.5 0.3 3 10 30 Reactivity Insertion Rate[PCM/Sec)100 DONALD C.COOK NUCLEAR PLANT UNIT 1 FIGURE 3.3-13 Miniinim DNBR vs.Reactivity Insertion Rate For The RCCA Withdrawal At Power Event, 1009o Power m i1944-5w.wpt:1d441195 3.3-79
2.4 2.2 tL z 2.0 0 E a 1.8 0vertemperature hT Trip Kgh Neutron Flux Trip 1.6 Min.Feecback Max.Feedback 1.4 0.3 1 3 10 30 Reactivity Insertion Rate[PCM/Sec)DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3-14 Minitmun DNBR vs.Reactivity Insertion Rate For The RCCA Withdrawal At Power Event, 60%Power m:11944-5w.wpt:1d441195 3.3-80 2.8 Min.Feetataek Max.Feedback 2.6 K z 2.4 0 E 2 2 High Neutron RuxTrip~2.0 Overtemperarure ET Trip 1.8 0.3 3 10 30 Reactivity Insertion Rate (PCM/Sec)100 DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3-15 Minitmtm DNBR vs.Reactivity Insertion Rate For The RCCA Withdrawal At Power Event, 10%Power m 119444w.wph1d4141195 3.3-81
l.t000 l.1000 l.0000 Q C O.SOOOO.00000 O.70000.COOOO R.50000 o 8.S le t000 l 1000 C E c 1.0000 Q I, SOOOO.00000 u.70000 Z~C0000.50000 8 8 Time[Ij DONALD C;COOK NUCLEAR PLANT UNIT 1 FIGURE 3.3-16 Nuclear Po~and Core Heat Flux vs.Tine for a Typical Response to a Dtopped RCCA(s)in Automatic Control m:(1944.5w.wpf:1d~1195 3.3-82 550.0$10.0 550.0 I-530.0 e o$10.0 O 490.0 8 8 8 tV f200.0 2100,0 g 2000.0 n e 1NN.O g e 1000+0 n 1100o0 8 Time fs]DONALD C.COOK NUCLEAR PLANT UNIT 1 FIGURE 3.3-17 Average Coohnt Tempertaure and Pressurizer Pressure vs.Tine for a 7ypical Response to a Dropped RCCA(s)in Auromanc Control mal 944-5w.wpf:1d441195 3.3-83 1.4 1.2 c~1.0 c 0 c 0.8 0.6 O i" 0.4 0 0.2 0.0 0 4 Time[s]10 DONALD C.COOK NUCLEAR PLANT UNIT 1 FIGURE 3.3-18 Total Core Flow vs.Time for The Complete Loss Of Flow Event m:i1 944-5w.wpf:1d441195 3.3-84 fi 0 1.4~1.2 S~1.0 0 C'=0.8~0.6.Q~0.4 O~0.2 0.0 0 Time[s)10 2,600 EU'v)2,400 CL C7 g 2.200 lN 2,000 CL 1,800 0 Time[s]10 DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3-19 Nuclear Power and Pressurizer Pressure vs.Time for The Complete Loss Of Flow Event m:11944.5w.wpt:
id~1 195 3.3-85 1.6 C5~1.2 I.O C=0.8 X G:~0.4 I T Average Channel Hot Channel 0.0 0 Time[s]10 4.0'L z O 3.0 2.0 1.0 0 4 Time[s]'eat fluxes are shown as a fraction of the nominal average channel heat flux DONALD C.COOK NUCLEAR PLANT UNIT 1 FIGURE 3.3-20 Average and Hot Channel Heat Fluxes and DNBR vs.Time for the Complete Loss Of Flow Event m:$1944-5w.wpf:1d441 195 3.3-86 f'l 1.4 1.2 C E 1.O 0 c pg 0.6 O K p4 0.2 0.0 0 Time[s]10 1.4~1.2 C o 1.0 C 0~o.s e'.6~0.4 3 D~0.2 0.0 0 Time[s]10 DONALD C.COOK NUCLEAR PLANT UNIT 1 FIGURE 3.3-21 Total Core Flow and Faulted Loop Flow vs.Time for The Partial Loss Of Flow Event m:51944-5w.wpi:1d~1195 3.3-87 1.4~1.2~1.0 0 C 0'=0.8~0.6 O~0.4~0.2 0.0 0 Time (s]10 2,600 C5~2,400 CL I g 2,200 I N'C~2,000 L, 1,800 0 Time[s]10 DONALD C.COOK NUCLEAR PLANT UNIT 1 FIGURE 3.3-22 Nuclear Power aml Pressurizer Pressure vs.Tiara for The Partial Loss Of Flow Event m:$1944.5w.wpt:1d441195 3.3-88 1.6~1.2 E r 0=0.8 8 td I)C LL~0.4 z Average Channel 0.0 0 Time[s)10 4.0 3.0 2.0 1.0 0 4 Time[s]'eat fluxes are shown as a fraction of the nominal average channel heat flux DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3-23 Average and Hot Channel Heat Fluxes and DNBR vs.Time for the Partial Loss Of Flow Event mA1 944-5w.wpf:1d~1195 3.3-89 1.4 1.2 C 5 1.O 0 C 08 0.6 u-O.4 I 0 0.2 0.0 0 Time[s]10 1.4 1.2 I 1.O 0.8 C O 0.6 0.4 u.0.2 8 o.o'D e-0.2 cl~4-0.6 0 Time (s]10 DONALD C.COOK NUCLEAR PLANT UNIT 1 FIGURE 3.3-24 Total Core Flow and Faulted Loop Flow vs.Tiara For The Locked Rotor Event m:(1944.5w.wpt:1d441195 3.3-90
1.4 a$~1.0 0 L O'=0.8~0.6 O~0.4 z 0.2 0.0 0 Time[s]10 2,800 2,600.5 I 2,400 o-2,200 Co O K 2,000 1,800 0 Time ts]10 DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3-25 Nuclear Power and RCS Pressure vs.Time For The Locked Rotor Event m f1944-5w.wpf:id%41195 3.3-91 1.6.I 1.2 Ha Clannel r 0 r.=0.8 X LL~0.4 rD Average Channel 0.0 0 Time (s]10 3,000.0 2,500.0 0~~2,000.0~rn r-1,500.0 O 1,000.0 500.0 0 Time[s]10'eat fluxes are shown as a fraction of the nominal average channel heat flux DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3-26 Average and Hot Channel Heat Fluxes vs.Tine and Clad Inner Temperature vs.Tine For The Locked Rotor Event mA1944.5w.wpf:1d441195 3.3-92 1.4 Cl I+1.,0 Q C O'=0.8~0.6 O~0.4 z 0.2 0.0 0 20 Time fs)60 80 5.0 4.0 P 3.0 A 2.0 1.0 0 20 40 60 Time[s]80 DONALD C.COOK NUCLEAR PLANT UNIT1 FIGURE 3.3-27 Nuciear Power and DNBR vs.Tiara For Loss of Load, Minimum Reactivity Feedback With Pressurizer Spray and PORVs m:51944.5w.wpt:1d 441195 3.3-93 2.800 2,600 Ct o 2;400 g.2,200 N g 2,000 G.1,800 1,600 0 20 Time[s]80 2,000 1,800 I~~1,600)I m 1,400'~~1,200 e-1,000 800 0 20 Time[s]80 DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3-28 Pressurizer Ptessute and Pressurizer Water Volund vs.Tun: For Loss of Load, Minimum Reactivity Feedback With Pressurizer Spray and PORVs m:11944.5w.wpt:1d041195 3.3-94 u el 650 CO CL E i-600 I 550 O O 500 0 20 Time[s]80 100 G.: 650 n.600 E~Q~550 8~ThOf Toold 500 0 20 Time fs]80 100 DONALD C.COOK NUCLEAR PLANT UNIT 1 FIGURE 3.3-29 Core Average and Loop 1 Temperatures vs.Time.For Loss of Load, Minimum Reactivity Feedback With Pressurizer Spray and PORVs m:11944.5w.wpt:1d~1195 3.3-95 1,000 5 ci-1,000 0 u-2.000 CUK 9-3,000~O-5,000 0 20 40 60 Time[s]80 50~~30 E~20 N I 10 CL 20 Time[s]80 DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3-30 Total Reactivity and Pressurizer Steam Relief vs.Tine For Loss of Load, Minimum Reactivity Feedback With Pressurizer Spray and PORVs m."tl 944-5w.wpf:1d441 195 3.3-96
1 i0,000 E 100,000 tO, 90.000 C)m 80,000 U E to 70,000 60.000 0 20 Time[s]80~~300 K g 200@100 (9 V)0 0 20 lime (s)80 DONALD C.COOK NUCLEAR PLANT UNIT 1 FIGURE 3.3-31 Steam Generator Mass and Safety Valve Relief vs.Tirn: For Loss of Load, Minimum Reactivity Feedback With Pressurizer Spray and PORVs m:11944-5w.wpf:1d~1195 3.3-97 1.4<1.2+1.0.O C'=0.8~0.6 O~0.4 z 0.2 0.0 20 Time (s]80 5.0 4.0 2.0 1.0 0 20 Time[s]80 DONALD C.COOK NUCLEAR PLANT UNIT 1 FIGURE 3.3-32 Nuclear Power and DNBR vs.Time For Loss of Load, Maximum Reactivity Feedback With Pressurizer Spray and PORVs m:11 944-5w.wpt:1d441 195 3.3-98 2,800 2,600 m 2;400 2.200 I N g 2,000 I tt.1,800 1,600 0 20 Time (s]80 2,000 1,800 I~~1,600 I m 1,400'~~1,200 P n-1,000 800 0 20 Time[s]80 DONALD C.COOK NUCIXm PLANT UNIT I FIGURE 3.3-33 Pressurizer Pressure and Pressurizer Water Volume vs.Time For Loss of Load, Maxitrtum Reactivity Feedback With Pressurizer Spray and PORVs m:t1 944-5w.wpt:
1d441195 3.3-99 u 0 o 650 f-600 r o 550 O O 500 0 20 Time[s]80<650 P a.600 E~8~550 y Thot Tcold 500 0 20 Time[s]80 DONALD C.COOK NUCLEAR PLANT UNIT 1 FIGURE 3.3-34 Core Average and Loop I Temperatures vs.Time For Loss of Load, Maximum Reactivity Feedback With Pressurizer Spray and PORVs mal 944.5w.wpf:1 d441195 3.3-100 j
1.000 5~-1;000)v 2)000 C$Cl 9-3,000~O-5,000 0 20 40 60 Time[s]80 50$40(t-30 E~20 n 10 O.0 0 20 Time[s)80 DONALD C.COOK'UCLEAR PLANT UNIT I FIGURE 3.3-35 Total Reactivity and Pressurizer Steam Relief vs.Tine For Loss of Load, Maximum Reactivity Feedback With Pressurizer Spray and PORVs m:$1944.5w.wpf:ld441195 3.3-101
110,000 E 100,000~90,000 m 80,000 Q E to 70,000 60,000 0 20 Time[s]80 400.0~300.0 K o 200.0~100.0 (9 to 0.0 0 20 Time[s]80 100 DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3-36 Steam Generator Mass and Safety Valve Relief vs.Tirade For Loss of Load, Maximum Reactivity Feedback With Pressurizer Spray and PORVs m:51944-5w.wpf:1d~1195 3.3-102 t I I 1.4 co]2 I~1.0-0.8 8~0.6 0~0.4~0.2 0.0 0 20 Time[s]80 5.0 4.0 g 3.0 O 2.0 1.0 0 20 Time[s]80 DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3-37 Nuclear Power and DNBR vs.Time For Loss of Load, Mioimum Reactivity Feedback Without Ptessurizer Spray and PORVs m:51944.5w.wpf:1d~1195 3.3-103
2,800 2,600 ca o 2;4M g.2,200 I N g 2,000 Q.1,800 1,600 0 20 Time[s]80 2,000-1,800~~1,600 I~1,400 I'~1,200 P L 1,000 800 0 20 Time[s]80 100 DONALD C.COOK NUCLEAR PLANT UNIT 1 FIGURE 3.3-38 Pressurizer Pressure and Pressurizer Water Volume vs.Tine For Loss of Load, Miainnun Reactivity Feedback Without Pressurizer Spray and PORVs m:51944-5w.wpktd441 195 3.3-104
u I 650 t-600 o 550 0 0 500 0 20 Time (s]60 80 100 G.: 650 a.600 E~ID~550 Tcold 500 0 20 Time[sj 80 DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3-39 Core Average and Loop I Temperatures vs.Tine For Loss of Load, Minimum Reactivity Feedback Without Pressurizer Spray and PORVs mA1944 5w.wpt:1d441195 3.3-105
1,000 E c-1,000 u-2.000 6$tr.9-3,000~O-5,000 0 20 Time (s]80 100 50 rr-30 E~20N'C I 10 n 0 0 20 Time fs]80 DONALD C.COOK NUCLEAR PLANT UNIT 1 FIGURE 3.3~Total Reactivity and Pressurizer Steam Relief vs.Time For Loss of Load, Minimum Reactivity Feedback Without Pressurizer Spray and PORVs m:11 944.5w.wpf:
1 dC41195 3.3-106 0
110.000 E 100,000 Cl 90.000 O S 80,000 Q E to 70,000 60,000 0 20 Time[s]60 80~300 K>>200 to 100 (9 to 20 Time[s]80 100 DONALD C.COOK NUCLEAR PLANT UNIT 1 FIGURE 3.3<1 Steam Generator Mass and Safety Valve Relief vs.Tine ,'or Loss of Load, Minimum Reactivity Feedback Without Pressurizer Spray and PORVs m:$1944.5w.wpk 1 d441195 3.3-107
1.4 g 1.2 10 0 C O'=0.8~0.6 O~0.4 O z 0.2 0.0 0 20 Time[s]80 5.0 4.0~3.0 O 2.0 1.0 0 20 Time[s]80 DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3Q2 Nuclear Power and DNBR vs.Time For Loss of Load, Maximum Reactivity Feedback Without Pressurizer Spray and PORVs m:11944.5w.wpf:1d~1195 3.3-108
2,800 2,600 (0 e 2;400 2,200 I N g 2,000 0 1,800 1,600 0 20 Time[s]60 80 2,000 n-1,800 I~~1,600 L m 1,400 I'~1,200 P L 1.000 800 0 20 Time[s]80 100 DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3Q3 Pressurizer Ptessute and Pressurizer Water Volume vs.Ttm.For Loss of Load, Maximum Reactivity Feedback Without Pressurizer Spray and PORVs m:51944.5w.wpf:
1 d441195 3.3-109
u 0 m 650 tO t-600 I 550 O O 500 0 20 Time (s]80 t.650 n.600 E~Q~550 3 Tcold 500 0 20 Time[s]80 DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3~Core Average and Loop I Temperatures vs.Time For Loss of Load, Maximum Reactivity Feedback Without Pressurizer Spray and PORVs m."t1 944-5w.wpt:1d441195 3.3-110
1,000 E g-1,000 o-2.000 C5 ID tr.m-3.000~O-5,000 0 20 Time[s]80 50 I~30 E~20 Q hl N 10 Q.20 Time[s]80 DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3A5 Total Reactivity and Pressurizer Steam Relief vs.Tire.For Loss of Load, Maximum Reactivity Feedback Without Pressurizer Spray and PORVs mht 944.5w.wpf:
1d~t 195 3.3-111
110,000 E 100,000~90,000 5 80,000 (9 E ro 70,000 60,000 0 20 Time[s]60 80~300 K g 200@100 (0 0 0 20 Time[s]80 DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3M Steam Generator Mass and Safery Valve Relief vs.Tine For Loss of Load, Maximum Reactivity Feedback Without Pressurizer Spray and PORVs mal 944.5w.wpf:1d~1195 3.3-112
1.030 1.020 1.010 QQQ o~~~~~~r 0.990 I I 0.980 I I I 240 280 320 360 400, 440 480 520 560 Core Average Temperature
['F]DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3%7 Variation of Reactivity With Core Temperature At 1050 psia For The End Of Life Rodded Core With One Control Rod Assembly Stuck (Zero Povver)For The Steamline Break Double Ended Rupture Event mht944-5w.wpf:id~1195 3.3-113 3.60 E u 3.20 2.80 X 2.40+~o 2.00 o 3"-1.60~o 1 20 0 0.80 0.40 0.00 0 10 15 20 25 30 35 40 45 50 Core Power[Precent of Nominal]DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3A8 Doppler Power Feedback For The Steamline Break Double Ended Rupture Event m:51944-5w.wpf:
1 d~1195 3.3-114
2,400 2,000~1,600 g 1,200 Q.o 800 K////0 0 5 10 15 20 25 30 35 40 45 50 Cold Leg Safety Injection tlbm/sec]DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3<9 Safety In~ecaon Flow Supplted By One Chargtng Pump For The Steamline Break Double Ended Rupture Event m$1944.5w.wpf:1d44]
195 3.3-115 0.4 E o 03 0 C 0 0.2 l 0 o I 0.1 z 0.0 0 50 150 Time[s]250 0.4 I 0.3 O C O 0.2 LL K 0.1 P O 0 0.0 0 50 150 Time[s]250 DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3-50 Nuclear Power and Core Heat Flux vs.Tine For The Steamline Break Double Ended Rupture Event fInside Containment With Power]m:51944-5w.wpt:1d~1195 3.3-116
u o 520 CL E t-480 I 440 O O 400 0 50 100.150 Time[s]250 2,500 2,000 I u)1.5000 M O tt: 1,000 500 0 50 150'Time[s]250 DONALD C.COOK NUCLEAR PLAN r UNIT 1 FIGURE 3.3-51 Core Average Temperature and RCS Pressure vs.Time For The Steamline Break Double Ended Rupture Event[inside Containment With Power]m:11944-5w.wpt:1d~1195 3.3-117
500~~400 L e 300 I'~200 n-100 50 150 Time (s]250 DONALD C.COOK NUCLEAR E'LAN'I'NIT I FIGURE 3.3-52 Pressurizer Water Volume vs.Tim For The Steamline Break Double Ended Rupture Event Pnside Containment With Power]mA1944.5w.wpf:1d~1195 3.3-118
5 CL-500 tt:-1,000-1,500 0 50 100 150 Time[s]250 40 I.30 C O C a 20 O O 0~10 P O O 50 150 Time[s]250 DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.53 Reactivity and Core Boron Concentration vs.Time For The Steamline Break Double Ended Rupture Event Pnside Containment With Power]mh1944.5w.wpf:1d
~1195 3.3-119 1.0E+2 C E o 1.0E+1 0.8 0.0 O 0 g 1.0E-1 O z 1.0E-2 0 2 Time (s]DONALD C.COOK NUCLEAR PLANT UNIT 1 FIGURE 3.3-54 Nuclear Power vs.Time For The Rod Ejection Event, Hot Zero Power, End Of Life m%1944-5w.wpf:1d~1195 3.3-120 u3000 t5 i-2000 D 0 D-1000 LL Fuel Centerline Fuel Average Clad Outer Surface Time[s]10 DONALD C.COOK NUCLEAR PLANT UNIT I FIGURE 3.3-55 Fuel Centerline, Fuel Average, and Clad Outer Surface Tetnperature vs.Tiara For Tkie Rod Ejection Event, Hot Zero Power, End Of Life m:51944-5w.wpi:
1 d441195 3.3-121 3.0 25 I o 2.0 I G"->.5~+1.0~0.5 z 0.0 0 2 Time[s]DONALD C.COOK NUCLEAR PLANT UNIT 1 FIGURE 3.3-56 Nuclear Power vs.Time For The Rod Ejection Event, Hot Full Power, End Of Life m%1944-5w.wpf:1d 441195 3.3-122 u 5000 ca 4000 O.E i-3000 D O~2000 u 1000 Fuel Centeriinc Fuel Average Clad Outer Surface Time[s)10 DONALD C.COOK NUCLEAR PLANT UNIT 1 FIGURE 3.3-57 Fuel Centerline, Fuel Average, and Clad Outer Surface Temperature vs.Tice For The Rod Ejection Event, Hot Full Power, End Of Life mA1944<w.wpt:Ld441195 3.3-123 r 3.4 POST LOCA HYDROGEN PRODUCTION As part of the Rerating Program, Westinghouse provided hydrogen generation rates and inventories inside containment from the various sources;including core radiolysis, sump radiolysis, corrosion-generated hydrogen and the zirconium/water reaction.A comparison of the key parameters and assumptions employed in these analyses were compared to the data provided for the SGTP Program.This comparison indicates that the maximum normal operation containment temperature of 120'(maximum)remains unchanged.
Since a containment temperature of 120'prior to the accident is the bases for the current analysis of record, the Rerating Program results remain applicable.
However, the post-accident time-temperature profile inside containment and the fraction of the core that undergoes a zirc-water reaction resulting from this analysis have been reviewed in order to ensure that the values employed in the Rerating Program analysis remain bounding for the SGTP Program.This review has been completed with the following conclusions:
The analysis for the Rerating Program considered zirconium-water reactions of 1.5%, 3.0%, and 5.0%.The limiting PCT calculations for the small break LOCA occurring within the pressurizer doghouse shows that only 0.128%of the core clad is oxidized.Applying the 10CFR50.44(d)(1) factor of 5 increase results in a zirc-water reaction percentage of 0.06%.The analyses remain highly conservative, since the value is less than the minimum value considered in the calculations (i.e., 1.5%).Also, the updated values do not contradict the UFSAR statement that"..zirconium-water reaction is calculated to be a maximum of 0.1%by weight of the total quantity of zirconium in the core." (UFSAR, page 14.3.6-2 dated July, 1982)For the large break LOCA, the zirc-water reaction is 4.93%.Since this value is less than the value that was considered in the Rerating Program (i.e., 5%), the results from the analysis remain applicable.
2.The impact of, the revised post-LOCA temperatures on the post-accident hydrogen generation has been reviewed and found to be negligible.
3.Another consideration is the current inventory of corrodible materials inside containment.
AEPSC provided the changes to the inventory of corrodible materials inside containment and confirmed that UFSAR Tables 14.3.6-3 and 14.3.6-7 remain valid.Therefore, the Rerating Program analysis remains bounding for the SGTP Program.I'A1944<w.wpf:1d441 195 3.4-1 16 14 12 10 w 2 0 4 6 8 ELEVATION (FT)Figure 3.1-86 Hot Rod Power Distribution (3 Inch, 30%SGTP)Reduced Temperature, Reduced Pressure Donald C.Cook Unit 1 m%1 944<w.wpf:1d~1195 3.1-189

3.5 CONTAINMENT

ANALYSES 3.5.1 Short-Term Containment Analysis The short term containment integrity analysis is used to verify the adequacy of interior structures and walls by demonstrating that calculated differential pressures are less than design limits.The functionality of the ice condenser is demonstrated and containment integrity is also verified.The efforts performed for the short term containment analysis, applicable to the Pressurizer Enclosure, as described in Section 3.4.1 of WCAP-11902, support operation of Cook Nuclear Plant Units 1 8 2 over the full range of rerated parameters.
The major impacts on LOCA short term mass and energy release rate calculations and containment subcompartment response analysis, are the effects due to RCS temperature changes.For the steam generator enclosure, mass and energy releases and the subsequent containment response are performed at zero power, which maximizes effects because steam pressure is maximum.All relevant analyses and evaluations performed for the Rerating Program assumed values which would bound both Units 1 8 2 at the rerated power levels and revised temperatures and pressures described in Table 2.1-1 of WCAP-11902, Supplement 1.The results of the short term containment analyses and evaluations for the SGTP Program demonstrate that, for the pressurizer enclosure, the fan accumulator room and the steam generator enclosure, the resulting peak pressures remain below the allowable design peak pressures.
Since the calculated pressures in WCAP-11902, Supplement 1 for the loop compartments exceeded the design pressure, demonstration of structural adequacy was required.This issue was addressed by AEPSC and is documented in UFSAR Section 14.3.4.2.3.4.
3.5.2 Loss-of-Coolant Mass and Energy Release 3.5.2.1 Purpose The purpose of this analysis was to calculate the long term LOCA mass and energy releases with the proposed revised plant conditions and increased operating margins.The increased operating margins include increased EDG start time to 30 seconds and the revised RHR and HHSI pump flow rates.This section provides the analytical basis with respect to the LOCA containment mass and energy release for the operation of the Donald C.Cook Nuclear Plant Units 1 and 2 at the SGTP Program conditions.
This containment integrity analyses bounds both units.Rupture of any of the piping carrying pressurized high temperature reactor coolant, termed a LOCA, will result in release of steam and water into the containment.
This, in turn, will result in an increase in the containment pressure and temperature.
The mass and energy release rates described in this document form the basis of further computations to evaluate the structural integrity of the containment following a postulated accident to satisfy the Nuclear m Al 9444w.wpf:1d441195 3,5-1 Regulatory acceptance criteria, General Design Criterion 38, which is more restrictive than the GDC criteria in Appendix H of the original FSAR, to which the Donald C.Cook Nuclear Plants are licensed.Section 3.5.3 presents the long term containment integrity analysis for containment pressurization evaluations.
3.5.2.2 System Characteristics and Modeling Assumptions The mass and energy release analysis is sensitive to the assumed characteristics of various plant systems, in addition to other key modeling assumptions.
Some of the most critical items are the RCS initial conditions, core decay heat, safety injection flow, and metal and steam generator heat release modeling.Specific assumptions concerning each of these items are discussed next.Tables 3.5-1 and 3.5-2 present key data assumed in the analysis.For the long term mass and energy release calculations, operating temperatures to bound the highest average coolant temperature range were used as bounding analysis conditions.
The modeled core rated power of 3413 MWt adjusted for calorimetric error (+2 percent of power)was the basis in the analysis.The use of higher temperatures is conservative because the initial fluid energy is based on coolant temperatures which are at the maximum levels attained in steady state operation.
Additionally, an allowance of+5.1'F is reflected in the temperatures in order to account for instrument error and deadband.The initial RCS pressure in this analysis is based on a nominal value of 2250 psia.Also included is an allowance of+67 psi, which accounts for the measurement uncertainty on pressurizer pressure.The selection of 2250 psia as the limiting pressure is considered to affect the blowdown phase results only, since this represents the initial pressure of the RCS.The RCS rapidly depressurizes from this value until the point at which it equilibrates with containment pressure.The rate at which the RCS blows down is initially more severe at the higher RCS pressure.Additionally the RCS has a higher fluid density at the higher pressure (assuming a constant temperature) and subsequently has a higher RCS mass available for releases.Thus, 2317 psia initial pressure was selected as the limiting case for the long term mass and energy release calculations.
These assumptions conservatively maximize the mass and energy in the RCS.The selection of the fuel design features for the long term mass and energy calculation is based on the need to conservatively maximize the core stored energy.The margin in core stored energy was chosen to be+15 percent.Thus, the analysis very conservatively accounts for the stored energy in the core.The fuel conditions were adjusted to provide a bounding analysis for current Cook Nuclear Plant Units 1 and 2 fuel design features..The following items serve as the basis to ensure conservatism in the core stored energy calculation:
time of maximum fuel densification and highest BOL temperatures.
Margin in RCS volume of 3%(which is composed of 1.6%allowance for thermal expansion and 1.4%for uncertainty) is modeled.m319444w.wpt:1d~1195 3.5-2 Regarding safety injection flow, the mass and energy calculation considered the limiting scenario of minimum safety injection flow, with the RHR crosstie valve is assumed to be closed, in conjunction with a 15%pump head degradation for the RHR and Sl pumps and 10%pump head degradation for the charging pumps.This configuration conservatively bounds other respective alignments.
Closure of the RHR crosstie was considered over the HHSI crosstie because this would have a more severe impact on the analysis (i.e., closure of the RHR crosstie would bound closure of the HHSI crosstie).
This results in the conservative minimum safety injection flowrate used.The following assumptions were employed to ensure that the mass and energy releases are conservatively calculated, thereby maximizing energy release to containment:
Maximum expected operating temperature of the RCS (100%full power conditions) 2.An allowance in temperature for instrument error and dead band (+5.1'F)1 3.Margin in volume of 3%(which is composed of 1.6%allowance for thermal expansion, and 1.4%for uncertainty) 4.Core rated power of 3413 MWt 5.Allowance for calorimetric error (+2 percent of power)6.Conservative coefficient of heat transfer (i.e., steam generator primary/secondary heat transfer and RCS metal heat transfer)7.Allowance in core store energy for effect of fuel densification 8.A margin in core stored energy (+15 percent included to account for manufacturing tolerances) 9.An allowance for RCS initial pressure uncertainty
(+67 psi)10.Steam generator tube plugging leveling (0%uniform)~Maximizes reactor coolant volume and fluid release~Maximizes heat transfer area across the SG tubes~Reduces coolant loop resistance, which reduces delta-p upstream of break and increases break flow mA1944+w.wpf:1d441195 3.5-3 Thus, based on the above conditions and assumptions, a bounding analysis of Cook Nuclear Plant Units 1 and 2 is made for the release of mass and energy from the RCS in the event of a LOCA to support the SGTP Program.3.5.2.3 Long Term Mass and Energy Release Analysis 3.5.2.3.1 Introduction The evaluation model used for the long term LOCA mass and energy release calculations was the March 1979 model described in Reference 1.This evaluation model has been reviewed and approved by the NRC, and has been used in the analysis of other ice condenser plants.This report section presents the long term LOCA mass and energy releases that were generated in support of the SGTP Program.These mass and energy releases are then subsequently used in the LOTIC-1 containment integrity analysis peak pressure calculation.
3.5.2.3.2 LOCA Mass and Energy Release Phases The containment system receives mass and energy releases following a postulated rupture in the RCS.These releases continue over a time period, which, for the LOCA mass and energy analysis, is typically divided into four phases: Blowdown-the period of time from accident initiation (when the reactor is at steady state operation) to the time that the RCS and containment reach an equilibrium state at containment design pressure.2.Refill-the period of time when the lower plenum is being filled by accumulator and ECCS water.At the end of blowdown, a large amount of water remains in the cold legs, downcomer, and lower plenum.To conservatively consider the refill period for the purpose of containment mass and energy releases, it is assumed that this water is instantaneously transferred to the lower plenum along with sufficient accumulator water to completely fill the lower plenum.This allows an uninterrupted release of mass and energy to containment.
Thus, the refill period is conservatively neglected in the mass and energy release calculation.
3.Ref lood-begins when the water from the lower plenum enters the core and ends when the core is completely quenched.4.Post-ref lood (Froth)-describes the period.following the ref lood transient.
For the pump suction break, a two-phase mixture exits the core, passes through the hot legs, and is superheated in the steam generators.
After the broken loop steam generator cools, the break flow becomes two phase.m:41944+w.wpf:1d441195 3.5-4 3.5.2.3.3 Computer Codes The Reference 1 mass and energy release evaluation model is comprised of mass and energy release versions of the following codes: SATAN Vl, WREFLOOD, and FROTH.These codes were used to calculate the long term LOCA mass and energy releases for Cook Nuclear Plant Units 1 and 2.SATAN calculates blowdown, the first portion of the thermal-hydraulic transient following break initiation, including pressure, enthalpy, density, mass and energy flowrates, and energy transfer between primary and secondary systems as a function of time.The WREFLOOD code addresses the portion of the LOCA transient where the core ref looding phase occurs after the pnmary coolant system has depressurized (blowdown) due to the loss of water through the break and when water supplied by the Emergency Core Cooling refills the reactor vessel and provides cooling to the core.The most important feature is the steam/water mixing model (See Section 3.5.2.6.2).
FROTH models the post-ref lood portion of the transient.
The FROTH code is used for the steam generator heat addition calculation from the broken and intact loop steam generators.
3.5.2.4 Break Size and Location Generic studies have been performed with respect to the effect of postulated break size on the LOCA mass and energy releases.The double ended guillotine break has been found to be limiting due to larger mass flow rates during the blowdown phase of the transient.
During the ref lood and froth phases, the break size has little effect on the releases.Three distinct locations in the RCS loop can be postulated for pipe rupture: 1.Hot leg (between vessel and steam generator) 2.Cold leg (between pump and vessel)3.Pump suction (between steam generator and pump)The break location analyzed for the SGTP Program is the pump suction double ended rupture, DEPS (10.48 ft').Break mass and energy releases have been calculated for the blowdown, ref lood, and post-reflood phases of the LOCA for each case analyzed.The following information provides a discussion on each break location.The hot leg double ended rupture has been shown in previous studies to result in the highest blowdown mass and energy release rates.Although the.ce'e flooding rate would be the highest for this break location, the amount of energy released from the steam generator secondary is minimal because the majority of the fluid which exits the core bypasses the steam generators venting directly to containment.
As a result, the reflood mass and energy m:51944+w.wpf:ad~1195 3.5-5 releases are reduced significantly as compared to either the pump suction or cold leg break locations where the core exit mixture must pass through the steam generators before venting through the break.For the hot leg break, generic studies have confirmed that there is no ref lood peak (i.e., from the end of the blowdown period the containment pressure would continually decrease).
The mass and energy releases for the hot leg break have not been included in the scope of this containment integrity analysis because for the hot leg break only the blowdown phase of the transient is of any significance.
Since there are no ref lood and post-ref lood phases to consider, the limiting peak pressure calculated would be the compression peak pressure and not the peak pressure following ice bed meltout.The cold leg break location has also been found in previous studies to be much less limiting in terms of the overall containment energy releases.The cold leg blowdown is faster than that of the pump suction break, and more mass is released into the containment.
However, the core heat transfer is greatly reduced, and this results in a considerably lower energy release into containment.
Studies have determined that the blowdown transient for the cold leg is, in general, less limiting than that for the pump suction break.During reflood, the flooding rate is greatly reduced and the energy release rate into the containment is reduced.Therefore, the cold leg break is not included in the scope of the SGTP Program.The pump suction break combines the effects of the relatively high core flooding rate, as in the hot leg break, and the addition of the stored energy in the steam generator.
As a result, the pump suction break yields the highest energy flow rates during the post-blowdown period by including all of the available energy of the RCS in calculating the releases to containment.
This break location has been determined to be the limiting break for all ice condenser plants.In summary, the analysis of the limiting break location for an ice condenser containment has been performed and is shown in this report.The double-ended pump suction (DEPS)guillotine break has historically been considered to be the limiting break location, by virtue of its consideration of all energy sources present in the RCS.This break location provides a mechanism for the release of the available energy in the RCS, including both the broken and intact loop steam generators.
3.5.2.5 Application of Single Failure Criteria An analysis of the effects of the single failure criteria has been performed on the mass and energy release rates for the DEPS break.An inherent assumption in the generation of the mass and energy release is that offsite power is lost.This results in the actuation of the emergency diesel generator, required to power the safety injection system.This is not an issue for the blowdown period which is limited by the compression peak pressure.The limiting minimum safety injection case has been analyzed for the effects of a single failure.In the case of minimum safeguards, the single failure postulated to occur is the loss of an emergency diesel generator.
This results in the loss of one pumped safety injection train, mA19444w.wpf:1d441195 3.5-6 thereby minimizing the safety injection flow.An additional conservatism has been included in this analysis in that the closure of the RHR crosstie valve has been considered because it results in a further reduction in safety.injection flow.The analysis further considers the RHR and Sl pump head curves to be degraded by 15%and the charging pump head curve to be degraded by 10%.This results in the greatest Sl flow reduction for the minimum safeguards case.3.5.2.6 Mass and Energy Release Data 3.5.2.6.1 Blowdown Mass and Energy Release Data A version of the SATAN-Vl code is used for computing the blowdown transient, which is the code used for the ECCS calculation in Reference 2.The code utilizes the control volume (element)approach with the capability for modeling a large variety of thermal fluid system configurations.
The fluid properties are considered uniform and thermodynamic equilibrium is assumed in each element.A point kinetics model is used with weighted feedback effects.The major feedback effects include moderator density, moderator temperature, and Doppler broadening.
A critical flow calculation for subcooled (modified 2aloudek), two-phase (Moody), or superheated break flow is incorporated into the analysis.The methodology for the use of this model is described in Reference 1.Table 3.5-3 presents the calculated mass and energy releases for the blowdown phase of the DEPS break.For the pump suction breaks, break path 1 in the mass and energy release tables refers to the mass and energy exiting from the steam generator side of the break;and, break path 2 refers to the mass and energy exiting from the pump side of the break.3.5.2.6.2 Ref lood Mass and Energy Release Data The WREFLOOD code used for computing the ref lood transient, is a modified version of that used in the 1981 ECCS evaluation model (Reference 2).The WREFLOOD code consists of two basic hydraulic models-one for the contents of the reactor vessel, and one for the coolant loops.The two models are coupled through the interchange of the boundary conditions applied at the vessel outlet nozzles and at the top of the downcomer.
Additional transient phenomena such as pumped safety injection and accumulators, reactor coolant pump performance, and steam generator release are included as auxiliary equations which interact with the basic models as required.The WREFLOOD code permits the capability to calculate variations during the core ref looding transient of basic parameters such as core flooding rate, core and downcomer water levels, fluid thermodynamic conditions (pressure, enthalpy, density)throughout the primary system, and mass flow rates through the primary system.The code permits hydraulic modeling of the two flow paths mal 944+w.wpf:1d441195 3 5-7 available for discharging steam and entrained water from the core to the break;i.e.the path through the broken loop and the path through the unbroken loops.A complete thermal equilibrium mixing condition for the steam and ECCS injection water during the ref lood phase has been assumed for each loop receiving ECCS water.This is consistent with the usage and application of the Reference 1 mass and energy release evaluation model in the Rerating Program analyses.Even though the Reference 1 model credits steam/mixing only in the intact loop and not in the broken loop, justification, applicability, and NRC approval for using the mixing model in the broken loop has been documented (Reference 3).This assumption is justified and supported by test data, and is summarized as follows: The model assumes a complete mixing condition (i.e., thermal equilibrium) for the steam/water interaction.
The complete mixing process, however, is made up of two distinct physical processes.
The first is a two phase interaction with condensation of steam by cold ECCS water.The second is a single phase mixing of condensate and ECCS water.Since the steam release is the most important influence to the containment pressure transient, the steam condensation part of the mixing process is the only part that need be considered.(Any spillage directly heats only the sump.)The most applicable steam/water mixing test data has been reviewed for validation of the containment integrity reflood steam/water mixing model.This data is that generated in 1/3 scale tests (Reference 4), which are the largest scale data available and thus most cleady simulates the flow regimes and gravitational effects that would occur in a PWR.These tests were designed specifically to study the steam/water interaction for PWR ref lood conditions.
From the entire series of 1/3 scale tests, a group corresponds almost directly to containment integrity ref lood conditions.
The injection flowrates for this group cover all phases and mixing conditions calculated during the ref lood transient.
The data from'these tests were reviewed and discussed in detail in Reference 1.For all of these tests, the data clearly indicate the occurrence of very effective mixing with rapid steam condensation.
The mixing model used in the containment integrity reflood calculation is therefore wholly supported by the 1/3 scale steam/water mixing data.Additionally, the following justification is also noted.The post-blowdown limiting break for the containment integrity peak pressure analysis is the pump suction double ended rupture break.For this break, there are two flowpaths available in the RCS by which mass and energy may be released to containment.
One is through the outlet of the steam generator, the other via reverse flow through the reactor coolant pump.Steam which is not condensed by ECCS injection in the intact RCS loops passes around the downcomer and through the broken loop cold leg and pump in venting to containment.
This steam also encounters ECCS injection water as it passes through the broken loop cold leg, complete mixing occurs and a portion of it is condensed.
It is this portion of steam which is condensed that is taken credit for in this mA19444w.wpt:1d 441195 3.5-8 analysis.This assumption is justified based upon the postulated break location, and the actual physical presence of the ECCS injection nozzle.A description of the test and test results is contained in References 1 and 4.Table 3.5-4 presents the calculated mass and energy release for the ref lood phase of the pump suction double ended rupture with minimum safety injection.
The transients of the principal parameters during ref lood are provided in Table 3.5-5.3.5.2.6.3 Post-Ref lood Mass and Energy Release Data The FROTH code (Reference 5)is used for computing the post-ref lood transient.
The FROTH code calculates the heat release rates resulting from a two-phase mixture level present in the steam generator tubes.The mass and energy releases that occur during this phase are typically superheated due to the depressurization and equilibration of the broken loop and intact loop steam generators.
During this phase of the transient, the RCS has equilibrated with the containment pressure, but the steam generators contain a secondary inventory at an enthalpy that is much higher than the primary side.Therefore, there is a significant amount of reverse heat transfer that occurs.Steam is produced in the core due to core decay heat.For a pump suction break, a two phase fluid exits the core, flows through the hot legs and becomes superheated as it passes through the steam generator.
Once the broken loop cools, the break flow becomes two phase.The methodology for the use of this model is described in Reference 1.After containment depressurization, the mass and energy release available to containment is generated directly from core boiloff/decay heat.Table 3.5-6 presents the two phase post-ref lood (froth)mass and energy release data for the pump suction double ended case.3.5.2.7 Decay Heat On November 2, 1978 the Nuclear Power Plant Standards Committee (NUPPSCO)of the American Nuclear Society (ANS)approved ANS standard 5.1 for the determination of decay heat.This standard was used in the mass and energy release model with the following input: Significant assumptions in the generation of the decay heat cuwe: Decay heat sources considered are fission product decay and heavy element decay of U-239 and N,-239.mal 9444w.wpl:1d
~1195 3.5-9 2.Decay heat power from fissioning isotopes other than U-235 is assumed to be identical to that of U-235.3.Fission rate is constant over the operating history of maximum power level.4.The factor accounting for neutron capture in fission products has been taken from Table 10 of Reference 6.5.Operation time before shutdown is 3 years.6.The total recoverable energy associated with one fission has been assumed to be 200 MeV/fission.
7.Two sigma uncertainty (two times the standard deviation) has been applied to the fission product decay.3.5.2.8 Steam Generator Equilibration and Depressurization Steam generator equilibration and depressurization is the process by which secondary side energy is removed from the steam generators in stages.The FROTH computer code calculates the heat removal from the secondary mass until the secondary temperature is the saturation temperature (T,)at the containment design pressure.After the FROTH calculations, steam generator secondary energy is removed based on first and second stage rates.The first stage rate is applied until the steam generator reaches T, at the user specified intermediate equilibration pressure, when the secondary pressure is assumed to reach the actual containment pressure.Then, the second stage rate is used until the final depressurization.
The heat removal of the broken loop and intact loop steam generators are calculated separately.
During the FROTH calculations, steam generator heat removal rates are calculated using the secondary side temperature, primary side temperature and a secondary side heat transfer coefficient determined using a modified McAdam's correlation (Reference 7).Steam generator energy is removed during the FROTH transient until the secondary side temperature reaches saturation temperature at the containment design pressure.The constant heat removal rate used during the first heat removal stage is based on the final heat removal rate calculated by FROTH.The SG energy available to be released during the first stage interval is determined by calculating the difference in secondary energy available at the containment design pressure and that at the (lower)user specified intermediate equilibration pressure, assuming saturated conditions.
This energy is then divided by the first stage energy removal rate, resulting in an intermediate equilibration time.At this time, the rate of energy release drops substantially to the second stage rate.The second stage rate is determined as the fraction of the difference in secondary energy available between the intermediate equilibration and final depressunzation.
mA19444w.wpi:1 d~1195 3.5-10 3.5.2.9 Sources of Mass and Energy The sources of mass considered in the LOCA mass and energy release analysis are given in Table 3.5-7.These sources are the RCS, accumulators, and pumped safety injection.
The energy inventories considered in the LOCA mass and energy release analysis are given in Table 3.5-8.The energy sources include: 1.RCS Water 2.Accumulator Water Pumped Injection Water Decay Heat Core Stored Energy RCS Metal-Primary Metal (includes SG tubes)7.Steam Generator Metal (includes transition cone, shell, wrapper, and other internals) 8.Steam Generator Secondary Energy (includes fluid mass and steam mass)Secondary Transfer of Energy (feedwater into and steam out of the steam generator secondary)
Energy Reference Points: Available Energy: 212'F;14.7 psia Total Energy Content: 32'F;14.7 psia It should be noted that the inconsistency in the energy balance tables from the end of Ref lood to 3600 seconds, i.e.,"Total Available" data versus"Total Accountable", resulted from the omission of the reactor upper head in the analysis following blowdown.It has been concluded that the results are more conservative when the upper head is neglected.
This does not affect the instantaneous mass and energy releases, or the integrated values, but causes an increase in the total accountable energy within the energy balance table.mA$944<w.wpt:1d441195 3.5-11
The mass and energy inventories are presented at the following times, as appropriate:
1.2.3.4.5.6.Time zero (initial conditions)
End of blowdown time End of refill time End of ref lood time Time of broken loop steam generator equilibration to pressure setpoint Time of intact loop steam generator equilibration to pressure setpoint In the mass and energy release data presented, no Zirc-water reaction heat was considered because the clad temperature did not rise high enough for the rate of the Zirc-water reaction heat to be of any significance.
The consideration of the various energy sources in the mass and energy release analysis provides assurance that all available sources of energy have been included in this analysis.Although Cook Nuclear Plant Unit 1 is not a Standard Review Plan Plant, the review guidelines presented in Standard Review Plan Section 6.2.1.3 have been satisfied.
3.5.2.10 References"Westinghouse LOCA Mass and Energy Release Model for Containment Design-March 1979 Version", WCAP-10325-P-A, May 1983 (Proprietary), WCAP-10326-A (Non-Propnetary).
2."Westinghouse ECCS Evaluation Model-1981 Version", WCAP-9220-P-A, Rev.1, February 1982 (Proprietary), WCAP-9221-A, Rev.1 (Non-Proprietary) 3.Docket No.50-315,"Amendment No.126, Facility Operating License No.DPR-58 (TAC No.71 06), for D.C.Cook Nuclear Plant Unit 1", June 9, 1989.4.EPRI 294-2, Mixing of Emergency Core Cooling Water with Steam;1/3 Scale Test and Summary, (WCAP-8423), Final Report June 1975.5."Westinghouse Mass and Energy Release Data For Containment Design", WCAP-8264-P-A, Rev.1, August 1975 (Proprietary), WCAP-8312-A (Non-Proprietary).
6.ANSI/ANS-5.1 1979, American National Standard for Decay Heat Power in Light Water Reactors", August 1979.7.W.H.McAdam, Heat Transmission, McGraw-Hill 3rd edition, 1954, p.172.mal 944+w.wpf:1d441195 3.5-12 3.5.3 LOCA Containment Integrity Analysis 3.5.3.1 Description of LOTIC-1 Model Early in the ice condenser development program, it was recognized that there was a need for modeling of long term ice condenser performance.
It was realized that the model would have to have capabilities comparable to those of the dry containment (COCO)model.These capabilities would permit the model to be used to solve problems of containment design and optimize the containment and safeguards systems.This has been accomplished in the development of the LOTIC code, described in Reference 1.The model of the containment consists of five distinct control volumes, the upper compartment, the lower compartment, the portion of the ice bed from which the ice has melted, the portion of the ice bed containing unmelted ice, and the dead ended compartment.
The ice condenser control volume with unmelted and melted ice is further subdivided into six subcompartments to allow for maldistribution of break flow to the ice bed.The conditions in these compartments are obtained as a function of time by the use of fundamental equations solved through numerical techniques.
These equations are solved for three phases in time.Each phase corresponds to a distinct physical characteristic of the problem.Each of these phases has a unique set of simplifying assumptions based on test results from the ice condenser test facility.These phases are the blowdown period, the depressurization period, and the long term.The most significant simplification of the problem is the assumption that the total pressure in the containment is uniform.This assumption is justified by the fact that after the initial blowdown of the RCS, the remaining mass and energy released from this system into the containment are small and very slowly changing.The resulting flow rates between the control volumes will also be relatively small.These flow rates then are unable to maintain significant pressure differences between the compartments.
In the control volumes, which are always assumed to be saturated, steam and air are assumed to be uniformly mixed and at the control volume temperature.
The air is considered a perfect gas, and the thermodynamic properties of steam are taken from the ASME steam table.The condensation of steam is assumed to take place in a condensing node located, for the purpose of calculation, between the two control volumes in the ice storage compartment.
The exit temperature of the air leaving this node is set equal to a specific value which is equal to the temperature of the ice filled control volume of the ice storage compartment.
Lower compartment exit temperature is used if the ice bed section is melted.m:119444w.wpf:1d441195 3.5-13 l
3.5.3.2 Containment Pressure Calculation The following are the major input assumptions used in the LOTIC analysis for the pump suction pipe rupture case with the steam generators considered as an active heat source for the Cook Nuclear Plant Containment:
Minimum safeguards are employed in all calculations, e.g., one of two spray pumps and one of two spray heat exchangers; one of two RHR pumps and one of two RHR heat exchangers providing flow to the core;one of two safety injection pumps and one of two centrifugal charging pumps;and one of two air return fans.2.2.11 x 10'bs.of ice initially in the ice condenser.
3.The blowdown, ref lood, and post reflood mass and energy releases described in Section 3.5.2 are used.4.Blowdown and post-blowdown ice condenser drain temperatures of 190'F and 130'F are used, respectively.
5.Nitrogen from the accumulators in the amount of 4510 lbs.is included in the calculations.
6.Essential setvice water temperature of 87.5'F is used on the spray heat exchanger and the component cooling heat exchanger, 7.The air return fan is effective 10 minutes after the transient is initiated.
8.No maldistribution of steam flow to the ice bed is assumed.(This assumption is conservative since it contributes to early ice bed melt out time.)9.No ice condenser bypass is assumed.(This assumption depletes the ice in the shortest time and is thus conservative.)
10.The initial conditions in the containment are a temperature of 57'F in the upper compartment volume, and 60'F in the lower and dead-ended compartment volumes.All volumes are at a pressure of 0.3 psig.11.Containment structural heat sinks are assumed with conservatively low heat transfer rates.(See Tables 3.5-11 and 3.5-12)mA1944+w.wpf:1d%51895 3.5-14 12.The operation of one containment spray heat exchanger (UA=3.107 x 10'tu/hr-'F), for containment cooling and the operation of one RHR heat exchanger.(UA=2.22 x 10'tu/hr-'F) for core cooling.The component cooling heat exchanger was modeled at 3.58 x 10'tu/hr-'F.
13.The air return fan returns air at a rate of 39,000 cfm from the upper to the lower compartment.
14.An active sump volume of 40,600 ft's used.15.102%of 3413 MWt power is used in the calculations.
16.Subcooling of ECCS water from the RHR heat exchanger is assumed.17.Essential service water flow to the containment spray heat exchanger was modeled as 2000 gpm.Also the nuclear service water flow to the component cooling heat exchanger was modeled as 5000 gpm.18.RHR Spray initiation is assumed after switchover from injection to recirculation has been completed and containment pressure is greater than or equal to 8 sl 3.5.3.3 Structural Heat Removal Provision is made in the containment pressure analysis for heat storage in interior and exterior walls.Each wall is divided into a number of nodes.For each node, a conservation of energy equation expressed in finite difference forms accounts for transient conduction into and out of the containment structural heat sinks used in the analysis.The material property data used is found in Tables 3.5-11 and 3.5-12.The heat transfer coefficient to the containment structure is based primarily on the work of Tagami (Reference 2).When applying the Tagami correlations, a conservative limit was'laced on the lower compartment stagnant heat transfer coefficients.
They were limited to a steam-air ratio of 1.4 according to the Tagami correlation.
The imposition of this limitation is to restrict the use of the Tagami correlation within the test range of steam-air ratios where the correlation was derived.With these assumptions, the heat removal capability of the containment is sufficient to absorb the energy releases and still keep the maximum calculated pressure below the design pressure.'al 944<w.wpf:1d441195 3.5-15 3.5.3.4 Analysis Results The results of the analysis shows that the maximum calculated containment pressure is 11.49 psig for the DEPS minimum safeguards break case.This pressure peak occurs at approximately 7752 seconds, with ice bed meltout at approximately 5423 seconds.The following plots show the containment integrity transient, as calculated by the LOTIC-1 code.Figure 3.5-1: Containment Pressure Transient Figure 3.5-2: Upper Compartment Temperature Transient Figure 3.5-3: Lower Compartment Temperature Transient Figure 3.5-4: Active and Inactive Sump Temperature Transient Figure 3.5-5: Ice Melt Transient Tables 3.5-9 and 3.5-10 give energy accountings at various points in the transient.
3.5.3.5 Relevant Acceptance Criteria The LOCA mass and energy analysis has been performed in accordance with the criteria shown in the Standard Review Plan (SRP)Section 6,2.1.3.In this analysis, the relevant requirements of General Design Criteria (GDC)50 and 10 CFR Part 50 Appendix K have been included by confirmation that the calculated pressure is less than the design pressure, and because all available sources of energy have been included, which is more restrictive than the GDC criteria in Appendix H of the original FSAR, to which the Donald C.Cook Nuclear Plants are licensed.These sources include reactor power, decay heat, core stored energy, energy stored in the reactor vessel and internals, metal-water reaction energy, and stored energy in the secondary system.The containment integrity peak pressure analysis has been performed in accordance with the criteria shown in the SRP Section 6.2.1.1.b for ice condenser containments.
Conformance to GDC's 16, 38, and 50 is demonstrated by showing that the containment design pressure is not exceeded at any time in the transient.
This analysis also demonstrates that the containment heat removal systems function to rapidly reduce the containment pressure and temperature in the event of a LOCA.3.5.3.6 Conclusions
/Based upon the information presented, it may be concluded that operation with the revised plant conditions and increased operating margins for Donald C.Cook Nuclear Plant is acceptable.
Operation with the RHR crosstie valve closed was also shown to be more limiting than operation with the valve open since there is less safety injection water available for steam condensation.
Operation with the revised plant conditions, increased operating margins mA1944+w.wpf:1d~1195 3.5-16 and the RHR crosstie valve closed results in a calculated peak containment pressure of 11.49 psig, as compared to the design pressure of 12.0 psig.Thus, the most limiting case has been considered, and has been demonstrated to yield acceptable results.3.5.3.7 References 1."Long Term Ice Condenser Containment Code-LOTIC Code", WCAP-8354-P-A, April 1976 (Proprietary), WCAP-8355-A (Non-Proprietary).
2.Tagami, Takasi, Interim Report on Safety Assessments and Facilities Establishment Project in Japan for Period Ending June, 1965 (No.1).3.5.4 Steamline Break Mass/Energy Releases Inside Containment The mass/energy releases for the inside containment analysis of record at the time the 30%SGTP Program was underway is based on the Rerating Program, which was performed to bound both units.The calculation of the mass/energy releases following a steamline break is described in the Cook Nuclear Plant Unit 1 UFSAR Section 14.3.4.4.Steamline ruptures occurring inside a reactor containment structure may result in significant releases of high energy fluid to the containment environment, possibly resulting in high containment temperatures and pressures.
The quantitative nature of the releases following a steamline rupture is dependent upon the many possible configurations of the plant steam system and containment designs as well as the plant operating conditions and the size of the rupture.These variations make it difficult to reasonably determine the single"worst case" for both containment pressure and temperature evaluations following a steambreak.
The analysis performed as part of the Rerating Program determined that the limiting scenario of the steambreak cases analyzed for the containment response evaluation were a break size of 0.86 ft'ccurring at full power for the split rupture scenario and a break size of 4.6 ft'ccurring at full power for the double-ended rupture scenario.As part of the 30%SGTP Program, several of the limiting cases of the steamline break mass and energy release calculations inside containment have been reanalyzed to assess a longer feedwater motor operated (FMO)valve stroke time, larger unisolatable feedline and steamline volumes, and revised maximum AFW flow rates.A relaxation in the EDG start-up time from 10 to 30 seconds, and an increase in the upper containment and lower compartment spray delay time are also addressed.
However, these latter analysis assumption changes only affect the containment response analysis.It should be noted that the changes associated only with the SGTP Program for Unit 1, i.e., RCS flow reduction, reduced primary-to-secondary heat transfer capability, and reduction in the rated thermal power, are less limiting parameters relative to the assumptions currently made for the mass/energy release calculations following a steamline break inside mal 9444w.wpf:1d441195 3.5-17 containment.
The parameter changes associated with the SGTP Program do not warrant reanalysis of this event.However, evaluations are currently in place (References 1 and 2)to address several non-conservative assumptions in the analysis.A reanalysis effort was undertaken for the steamline break mass and energy releases inside containment as part of the SGTP Program, so that the Reference 1 and 2 evaluations will no longer be required and to support the increased operating margins included in the discussion of the previous paragraph.
This section discusses a series of steamline breaks, consistent with the cases presented in the UFSAR, which were analyzed to determine the mass and energy releases from a variety of postulated pipe breaks encompassing wide variations in plant operation, safety system performance, and break sizes.The mass and energy release data is subsequently used as input to the containment integrity analysis discussed in Section 3.5.5.3.5.4.1 Method of Analysis The LOFTRAN computer code (Reference 3)was used to calculate the break flows and enthalpy of the releases through the steamline break as a function of time.Blowdown masslenergy releases determined using LOFTRAN include the effects of core power generation, main and auxilianJ feedwater additions, engineered safeguards systems, reactor coolant thick metal heat storage, and reverse steam generator heat transfer.A bounding analysis was performed to address the range of conditions possible for Unit 1 operation and the potential Unit 2 uprating.The assumptions on the initial conditions are taken to maximize the mass and total energy released.The higher primary temperatures along with the higher uprated power level associated with the Unit 2 uprating parameters are conservative for the mass/energy release calculations.
The upper bound temperature of Table S-2.1-1, Case 8 of WCAP-11902, Supplement (Reference 4), was used.Since the mass blowdown rate is dependent on steam pressure, and the steam pressure is less for the lower bound temperature case, the steam pressure of the upper bound temperature case is limiting for the range of operating conditions possible for the uprating of Unit 2.The functions which actuate safety injection and steamline isolation during a steamline rupture event are commonly referred to as the Steamline Break Protection System.A plant's steamline break protection system design can have a large effect on steamline break results.The current steamline break protection system designs for Unit 1 and Unit 2 are different.
The current Unit 1 design is referred to as an"OLD" steamline break protection system design.The Unit 2 design (and the proposed modified Unit 1 design;see Section 3;3.2.5)is referred to as a"HYBRID" steamline break protection system design.The two systems have the following characteristics:
mal 9444w.wpf:
1 d~1195 3.5-18 Current Unit 1-"OLD" Steamline Break Protection Safety Injection Signals 1.High steam flow coincident with low steam(inc pressure (two out of four lines)2.High steam flow coincident with low-low T,~(two out of four lines)3.Two out of three differential pressure signals between a steam line and the remaining steam lines 4.Two out of three low pressurizer pressure signals 5.Two out of three high containment pressure signals Steamline Isolation Signals 1.High steam flow coincident with low steamline pressure (two out of four lines)2.High steam flow coincident with low-low T,~(two out of four lines)3.Two out of four high-high containment pressure signals Unit 2-"HYBRID" Steamline Break Protection Safety Injection Signals Low steamline pressure (two out of four lines)2.Two out of three differential pressure signals between a steam line and the remaining steam lines 3.Two out of three low pressurizer pressure signals 4.Two out of three high containment pressure signals Steamline Isolation Signals 1.Low steamline pressure (two out of four lines)2.High steam flow coincident with low-tow T, (two out of four lines)3.Two out of four high-high containment pressure signals The only differences between the current Unit 1 and Unit 2 steamline break protection logic designs are the actuations from a high steam flow and low-low T,~signal and the logic associated with the low steamline pressure signal required to actuate safety injection and mA1944+w.wpf:1d~1195 3.5-19 steamline isolation.
Currently, for Unit 1, a high steam flow coincident with low-low T,~signal actuates both safety injection and steamline isolation.
For Unit 2, a hf'gh steam flow coincident with low-low T, signal actuates only steamline isolation.
However, the difference is not significant for the calculation of the mass/energy releases since the analysis does not take credit for any ESF actuations on a high steam flow coincident with low-low T, signal.The current Unit 1 design requires a coincidence between the low steamline pressure and high steam flow for protection actuation.
The Unit 2 design only requires the low steamline pressure signal for protection actuation; no coincidence with steam flow is required.The coincidence logic required for safety injection initiation and steamline isolation on high steam flow and low steam pressure for the current Unit 1 design is more limiting for the calculation of mass/energy releases inside containment than Unit 2's design.Actuation of safety injection and steamline isolation will limit the mass/energy released to the containment.
Delaying the safeguards initiation will result in a conservative calculation of the mass/energy releases for the containment pressure and temperature evaluation.
The coincidence requirement for high steam flow with low steam pressure of the current Unit 1 design increases the likelihood that safeguards initiation might be delayed compared to Unit 2's design where only a low steam pressure signal is required.In the case where the coincidence logic prohibits safety injection and steamline isolation on high steam flow with low steam pressure, one of the other signals must be received before the safeguards are initiated.
As such, the current Unit 1 steamline break protection system design was assumed in this bounding analysis for the calculation of the mass/energy releases inside containment.
3.5.4.2 Assumptions Several steamline breaks were analyzed to determine a limiting break condition for the containment temperature and pressure response.The following assumptions were used in the analysis: Double-ended pipe breaks were assumed to occur at the nozzle of one steam generator and also downstream of the flow restrictor.(Since neither Unit 1 nor Unit 2 has integral flow restnctors.)
Split ruptures were assumed to occur at the nozzle of one steam generator.
b.The blowdown is assumed to be dfy saturated steam.c.As discussed previously, the Unit 1 steamline break protection system design is assumed.However, credit was not taken for safeguards actuation on high steam line differential pressure or high steam flow coincident with low-low T,.d.Steamline isolation is assumed complete 11 seconds after the setpoint is reached for either high steam flow coincident with low steam pressure or m:L1944+w.wpf:1d~1195 3.5-20 high-high containment pressure.The isolation time allows 8 seconds for valve closure plus 3 seconds for electronic delays and signal processing.
As part of the Rerating Program, 4.6 ft'nd 1.4 ft'ouble-ended pipe breaks were analyzed at 102, 70, 30, and zero percent power levels.For the SGTP Program, a sub-set of these double-ended pipe breaks have been re-analyzed.
This sub-set, listed below, corresponds to the most limiting double-ended pipe breaks, as determined during the Rerating Program effort.4.6 ft', 102%power, with an MSIV failure;4.6 ft', 70%power, with an MSIV failure;1.4 ft', 102%power, with an MSIV failure;1.4 ft', 70%power, with an MSIV failure.As part of the Rerating Program, split pipe ruptures were analyzed at 0.86 ft', 102%power, 0.908 ft', 70%power, 0.942 ft', 30%power;and 0.4 ft', hot shutdown.These split break sizes for each power level were modeled because they reflect the largest breaks for which ESF actuations (i.e., steamline isolation, feedwater isolation, and safety injection) must be generated by high containment pressure trips.The high steam flow coincident with low steam pressure is not reached for these break sizes or smaller break sizes at the respective power levels (Reference 5).For the SGTP Program, a sub-set of these split breaks have been re-analyzed.
This sub-set, listed below, corresponds to the most limiting split break sizes, as determined during the Rerating Program effort.0.86 ft', 102%power, with an auxiliary feedwater runout protection (AFWRP)failure;0.942 ft', 30%power, with an AFWRP failure;0.942 ft', at 30%power, with an MSIV failure.Failure of an MSIV, failure of a feedwater isolation valve or main feed pump trip, and failure of auxiliary feedwater runout control were considered.
As part of the Rerating Program two cases for each>reak size and power level scenario were analyzed with one case modeling the MSIV failure and the other case modeling the AFW runout control failure.Each case assumed conservative main feedwater addition to bound the feedwater isolation valve or mA1944+w.wpt:1d441195 3.5-21 main feed pump trip failure.For the SGTP Program, the sub-set of case re-analyzed assumed a failure as previously noted (items e.and f.).The case re-analyzed for the SGTP Program also assumed.Feedline isolation via FMO valves complete 44 seconds after the setpoint is reached for either high steam flow coincident with low steam pressure or high containment pressure.The isolation time allows 41 seconds for valve closure plus 3 seconds for electronic delays and signal processing.
h.The auxiliary feedwater system is manually re-aligned by the operator after 10 minutes.A shutdown margin of 1.3%Ak/k is assumed.A moderator density coefficient of 0.54 Ak/gm/cc is assumed.Minimum capability for injection of boric acid (2400 ppm)solution corresponding to the most restrictive single failure in the safety injection system.The ECCS consists of the following systems: 1)the passive accumulators, 2)the low head safety injection (residual heat removal)system, 3)the high head (intermediate head)safety injection system, and 4)the charging safety injection system.Only the charging safety injection system and the passive accumulators are modeled for the steam line break accident analysis.The modeling of the safety injection system in LOFTRAN is described in Reference 3.Figure 3.3-49 of this report presents the safety injection flow rates as a function of RCS pressure assumed in the analysis.The flow corresponds to that delivered by one charging pump delivering its full flow to the cold legs.The safety injection flows assumed in this analysis take into account the 10%degradation of the charging pump performance.
No credit has been taken for any borated water that might exist in the injection lines, which must be swept from the lines downstream of the boron injection tank isolation valves prior to the delivery of boric acid to the RCS loops.For this analysis, a boron concentration of 0 ppm for the boron injection tank is assumed.After the generation of the safety injection signal (appropriate delays for instrumentation, logic, and signal transport included), the appropriate valves begin to operate and the safety injection charging pump starts.In 27 seconds, the valves are assumed to be in their final position (VCT charging pump suction valve has closed following opening of RWST charging pump suction valve)and the pump is assumed to be at full speed and to draw suction from the RWST.The volume containing the low concentration borated water is swept into the core before the 2400 ppm borated water reaches the core.This delay, described above, is inherently included in the modeling.Note that the relaxed m%1 944<w.wpf:1d441195 3.5-22 EDG start-up time is not reflected in the steamline mass/energy releases, as conservative releases are obtained if offsite power is maintained (see item m.below):.For the at-power cases, reactor trip is available by safety injection signal, overpower protection signal (high neutron flux reactor trip or OPET reactor trip), and low pressurizer pressure reactor trip signal.m.For RCP operation, offsite power is assumed available.
Continued operation of the RCPs maximizes the energy transferred from the RCS to the steam generator.
n.No steam generator tube plugging is assumed to maximize the heat transfer characteristics.
3.5.4.3 Single Failure Effects Failure of an MSIV increases the volume of steam piping which is not isolated from the break.When all valves operate, the piping volume capable of blowing down is located between the steam generator and the first isolation valve.If this valve fails, the volume between the break and upstream of the isolation valves in the other steamlines, including safety and relief valve headers and other connecting lines, will feed the break.For the cases which modeled a failure of an MSIV, the steamline volumes associated with Unit 1 were assumed since the volume available for blowdown for this scenario is greater than that for Unit 2.For the cases which did not model a failure of an MSIV, the steamline volumes associated with Unit 1 were assumed since the volume available for blowdown for this scenario is greater than that for Unit 2.b.Failure of a diesel generator would result in the loss of one containment safeguards train resulting in minimum heat removal capability.
Failure of a feedwater isolation valve would result in additional inventory in the feedwater line which would not be isolated from the faulted steam generator.
The mass in this volume can flash into steam and exit through the break.For consistency with the UFSAR steamline break masslenergy release analysis, all cases conservatively assumed failure of the feedwater isolation valve, which resulted in the additional inventory available for release through the steambreak and in higher than normal main feedwater flows.d.Failure of the auxiliary feedwater runout control equipment would result in a higher auxiliary feedwater flow entering the faulted steam generator prior to re-alignment of the AFW system.For cases where the runout control operates properly, a bounding m:51944+w.wpf:Id~
1 195 3.5-23 constant AFW flow of 775 gpm to the faulted steam generator was assumed.This value was increased to 1375 gpm to simulate a failure of the runout control.3.5.4.4 Results The steamline break mass/energy releases inside containment were calculated to account for the range of conditions possible for the potential uprating of Unit 2.One set of mass/energy releases were calculated to bound both Units incorporating the limiting steamline break protection design currently installed in Unit 1.Section 3.5.5 presents the containment integrity evaluation for a main steamline break using the mass/energy releases calculated here.As discussed in Section 3.5.5, the limiting scenarios of the steambreak cases analyzed for the containment response evaluation were a break size of 1.4 ft'ccurring at 102%power with an MSIV failure for the double-ended rupture scenario and a break size of 0.942 ft'ccurring at 30%power with an MSIV failure for the split rupture scenario.Table 3.5.13 presents the mass/energy releases for these limiting steambreak cases of the containment response evaluation.
3.5.4.5 References 1."American Electric Power Service Corporation, Donald C.Cook Nuclear Power Plant Units 1 and 2, Increased Upper 8 Lower Compartment Spray Delivery Times," W Letter AEP-94-712, June 13, 1994.2."American Electric Power Service Corporation, Donald C.Cook Nuclear Power Plant Units 1 and 2, Feedwater Isolation Valve Evaluation Support," W Letter AEP-93-528, April 8, 1993.3.Burnett, T, W.T., et al.,"LOFTRAN Code Description," WCAP-7907-A, April 1, 1984.4."Rerated Power and Revised Temperature and Pressure Operation for Donald C.Cook Nuclear Plant Units 1&2 Licensing Report," WCAP-11902, Supplement, September 1989.5.Land, R.E.,"Mass and Energy Releases Following a Steam Line Rupture," WCAP-8860, September 1976.mal 944+w.wpf:1d441195 3.5-24 3.5.5 Main Steam Line Break Containment Integrity 3.5.5.1 Introduction and Background A series of main steam line split and double-ended breaks were analyzed as part of the Rerating Program for Cook Nuclear Plant Units 1 and 2 to determine the most severe break condition for containment temperature and pressure response for this design basis event.The analysis and evaluation conducted are discussed in Reference 1.The results from the Rerating Program, which are documented in the FSAR, show that the worst case of the double-ended breaks was a 4.6 square foot break, occurring at 102%power with a main steam isolation valve failure.The worst case for the split breaks was the 0.86 square foot break, occurring at 102%power, with the failure of auxiliary feedwater runout protection.
The calculated peak containment temperature was 324.9'F and 324.4'F, respectively.
3.5.5.2 Purpose of Analysis An analysis was performed as a part of the SGTP Program, to demonstrate that the peak containment temperature resulting from a design basis main steam line break will not exceed the equipment qualification criterion for the plant.The analysis was performed to bound Cook Nuclear Plant Units 1 and 2 operation under uprated conditions (3600 MWt NSSS).The containment pressure response generated for the LOCA double-ended pump suction break (Section 3.5.3)is calculated to be more severe, and therefore is not a concern here.3.5.5.3 Major Analytical Assumptions An analysis consistent with the Reference 1 analysis was performed.
The analytical effort provides bounding calculations for both Units 1 8 2 at a power level of 3600 MWt.A spectrum of the limiting split breaks from Reference 1 were analyzed: 0.86 ft', 102%power, with an AFWRP failure;0.942 ft', 30%power, with an AFWRP failure;0.942 ft', at 30%power, with an MSIV failure.Also, the following double-ended breaks from Reference 1 were analyzed: 4.6 ft', 102%power, with an MSIV failure;4.6 ft', 70%power, with an MSIV failure;1.4 ft', 102%power, with an MSIV failure;1.4 ft', 70%power, with an MSIV failure.a The mass and energy release to containment as a result of the postulated steam line break were calculated using the LOFTRAN computer code (Reference 2).Consistent with the m:$19444w.wpf:
1 d 441195 3.5-25 Reference 1 analysis, no credit was taken for entrainment.
Section 3.5.4 presents additional details regarding the calculation of the inside containment steam line break mass and energy releases.The consequences of these releases;in particular the peak containment temperature, was calculated using the LOTIC-3 computer code (Reference 3).The following are the major input assumptions used in LOTIC-3: The containment integrity calculations were performed with an additional failure of one of the containment safeguards trains, e.g., one of two spray pumps, which results in the minimum spray flow and one of two air return fans.Where applicable, plant data consistent with the LOCA containment integrity analysis (Section 3.5.3)was used.2.The total initial ice mass used is 2.11 x 10'bs.3.The initial conditions in the containment are a temperature of 120'F in the lower and dead-ended compartments, a temperature of 27'F in the ice condenser, and a temperature of 57'F in the upper compartment.
All volumes are at a pressure of 0.3 psig and a relative humidity of 15%.4.The RWST temperature was assumed to be 105'F.5.A containment spray pump flow of 2075 gpm to the upper compartment and 1006 gpm to the lower compartment was assumed.6.Containment spray response time following high-high containment pressure setpoint is 115 seconds.7.The mass and energy releases are given in Section 3.5.4.3.5.5.4 Results Lar e Break The limiting case among the double-ended ruptures, which yielded a calculated peak temperature of 322.7'F, is.the 1.4 ft'ouble-ended rupture, 102%power, MSIV failure case.Figures 3.5-6 and 3.5-7 provide the upper and lower compartment temperature profiles.Figures 3.5-8 and 3.5-9 illustrate the upper and lower compartment pressure transients.
mh19444w.wpf:1d441195 3.5-26
Small Break The most limiting case in terms of peak calculated temperature is the 0.942 ft'plit break, 30%power, with an MSIV failure.This case resulted in a calculated peak temperature of 326'F.Figures 3.5-10 and 3.5-11 provide the upper and lower compartment temperature profiles.Figures 3.5-12 and 3.5-13 present the upper and lower compartment pressure transients.
3.5.5.5 Conclusion The main steam line break containment integrity analysis has been performed consistent with the current licensing basis analysis and SGTP Program, considering the present operating plant conditions.
The results of this analysis show that the Environmental Acceptance Criteria (Reference 4)applicable for Cook Nuclear Plant Units 1 and 2 are met.This analysis therefore demonstrates that the containment heat removal systems function to rapidly reduce the containment pressure and temperature in the event of a main steam line break accident.GDC 50 and 10CFR Part 50 Appendix K are satisfied, which is more restrictive than the GDC criteria in Appendix H of the original FSAR, to which the Donald C.Cook Nuclear Plants are licensed.3.5.5.6 References WCAP-11902, Supplement 1, September 1989,"Rerated Power and Revised Temperature and Pressure Operation for Donald C.Cook Nuclear Plant Units 1&2 Licensing Report." 2.WCAP-7907-P-A (Proprietary),"LOFTRAN Code Description", April 1984.3.WCAP-8354-P-A (Proprietary), Supplement 2,"Long Term Ice Condenser Containment Code-LOTIC-3 Code", February 1979.4.AEP/W SGTP-19,"Donald C.Cook Nuclear Plant Steam Generator Tube Plugging Analysis Technical Documentation Transmittal", August 10, 1994.mA1944+w.wpt:1d~1195 3.5-27
TABLE 3.5-1 DONALD C.COOK NUCLEAR PLANT UNITS 1 AND 2 SYSTEM PARAMETERS"INITIAL CONDITIONS PARAMETERS Core Thermal Power (MWt)Reactor Coolant System Flowrate, per Loop (gpm)Vessel Outlet Temperature
('F)Core Inlet Temperature
('F)Vessel Average Temperature
('F)Initial Steam Generator Steam Pressure (psia)Steam Generator Design Steam Generator Tube Plugging (%)Initial Steam Generator Secondary Side Mass (Ibm)Accumulator Water Volume (ft')N, Cover Gas Pressure (psia)Temperature
('F)Safety Injection Delay (sec)(includes time to reach pressure setpoint)VALUE 3413 79000 615.2 547.4 581.3 836.3 Model 51 114075 946 600 120 48.0 (analysis value includes an additional
+5.1'F allowance for instrument error and deadband)mh1944+w.wpf:1d441195 3.5-28 TABLE 3.5-2 DONALD C.COOK NUCLEAR PLANT UNITS 1 AND 2 SAFETY INJECTION FLOW Minimum Sl RCS Pressure (psig)0 20 40 60 80 100 120 140 160 180 200 INJECTION MODE Total Flow (gpm)3635.5 3447.2 3235.3 3003.7 2738.0 2425.6 2041.3 1493.3 889.5 883.0 876.4 RECIRCULATION MODE (w/o RHR Spray)Total Flow (gpm)3011 RECIRCULATION MODE (w/RHR Spray)Total Flow (gpm)mA1944<w.wpf:1d451895 3.5-29 TABLE 3.54 DONALD C.COOK NUCLEAR PLANT UNITS 1 AND 2 DOUBLE-ENDED PUMP SUCTION GUILLOTINE MINIMUM SI SLOWDOWN MASS AND ENERGY RELEASE TIME BREAK PATH NO.l FLOW BREAK PATH NO.2 FLOW SECONDS.000.101.201.301.401.601.800 1.10 1.40 2.30 2.80 3.00 3.40 3.90 4.60 5.20 6.20 6.60 8.00 8.40 8.80 9.40 1 1.0 13.8 16.4 18.4 18.8 19.0 19.2 19.4 19.6 19.8 20.0 20.2 20.6 20.8 21.0 21.4 21.8 22.0 22.2 22.4 23.6 23.8 24.4 24.6 25.8 27.0 28.0 LBM/SEC.0 40932.2 41809.9 46791.4 47304.8 44920.5 45062.1 41715.9 38993.8 31603.5 26248.9 21790.0 19301.6 18078.7 15511.6 13955.8 12621.1 12409.8 12926.3 12289.7 10124.1 9779.3 9523.6 7137.6 5492.8 4381.2 4172.2 4012.7 3961.1 3820.3 3734.8 3554.0 3322.7 3066.4 2642.1 2411.6 2237.3 1909.0 1705.5 1562.8 1445.6 1337.0 1176.0 914.5 833.4 475.4 413.0 140.8 66.5 19.0 THOUSAND BTU/SEC.0 22363.5 22984.5 25944.9 26505.7 25777.3 26395.2 25037.9 23888.1 20634.8 17695.2 14847.8 13347.2 12537.0 10751.8 9663.8 8669.3 8475.8 8608.2 8355.6 7526.0 7313.0 6842.3 5509.2 4593.4 3819.6 3729.5 3668.6 3685.6 3688.4 3730.0 3671.6 3582.1 3451.8 3188.3 2951.7 2755.6 2365.5 2122.8 1947.9 1805.1 1672.2 1475.1 1151.5 1050.2 600.7 523.1 179.4 85.2 24.5 , LBM/SEC.0 21842.5 23698.7 23586.2 23083.2 21588.6 19910.9 18871.6 18466.4 17983.4 17008.4 16632.6 15908.2 15034.4 13993.7 13348.7 12570.9 13180.5 12562.2 12442.3 12205.8 11841.3 10832.1 9219.1 7813.7 6705.5 9594.3 5136.1 8561.9 8618.3 5453.2 8394.0 4907.9 6807.8 4697.0 5832.6 3815.9 4740.7 2922.4 6501.6 4567.9 5030.5 33M.6 1329.9 2964.0 2402.3 2843.3 2309.9 299.7 126.0 THOUSAND BTU/SEC.0 11892.7 12911.5 12864.8 12605.8 11803.0 10889.6 10327.0 10106.3 9842.5 9311.5 9107.5 8714.5 8239.8 7674.4 7323.9 6900.4 7237.4 6907.8 6843.7 6713.4 6514.6 5959.4 5074.5 4311.0 3474.5 4972.0 2661.2 4245.4 4403.8 2703.5 4148.0 2449.3 3152.7 2086.8 2567.9 1683.9 1974.2 1187.0 2457.3 1744.3 1926.8 1231.8 459.4 796.5 633.4 754.2 648.0 93.9 45.3 mA1944<w.wpI:1d~1195 3.5-30 TABLE 3.&4 DONALD C.COOK NUCLEAR PLANT UNITS 1 AND 2 DOUBLE ENDED PUMP SUCTION GUILLOTINE MINIMUM Sl REFLGOD MASS AND ENERGY RELEASE TIME SECONDS 28.0 28.3 28.5 28.7 29.7 30.0 30.7 34.1 36.2 38.1 39.1 40.1 41.2 42.2 43.2 45.2 47.2 48.2 49.2 51.2 53.2 55.2 57.2 59.2 61.2 62.3 63.3 64.3 68.3 70.3 74.3 78.3 81.3 82.3 86.3 93.3 97.3 101.3 105.3 107.3 117.3 127.3 135.3 143.3 153.3 163.3 175.3 193.3 223.3 249.7 LBM/SEC.0 143.2 122.8 110.9 108.0 119.6 125.7 148.1 160.6 170.9 213.4 351.3 371.3 368.0 362.3 350.6 339.7 355.6 346.0 337.3 329.2 321.6 314.5 307.8 313.7 456.3 448.9 414.7 399.6 372.0 347.1 329.9 324.5 304.0 272.8 257.3 243.3 230.9 225.1 201.1 183.5 173.0 165.1 158.1 153.5 150.'I 148.0 148.4 150.3 THOUSAND BTU/SEC.0 168.1 144.3 130.3 126.8 140.4 147.6 174.1 188.8 200.9 251.1 414.5 438.4 434.6 427.7 413.8 400.8 419.7 419.3 408.2 397.9 388.3 379.3 370.8 362.9 369.5 539.8 531.0 490.1 472.1 439.2 409.5 389.1 382.6 358.3 321.1 302.8 286.3 271.5 264.7 236.3 215.4 203.0 193.7 185.5 180.0 176.0 173.5 173.8 175.9 BREAK PATH NO.1 FLOW BREAK PATH LBM/SEC.0 1818.3 1791.4 1778.9 1717.9 1696.2 1658.1 1508.0 1431.7 1371.5 2017.3 3809.9 4023.2 3989.7 3929.7 3807.4 3691.6 3902.6 3885.2 3787.0 3694.5 3607.3 3524.8 3446.8 3372.8 238.7 301.8 298.4 282.9 276.1 263.7 252.5 244.9 242.6 233.6 220.1 213.5 207.6 202.4 200.0 190.1 183.0 178.8 175.8 173.1 171.4 170.1 169.4 169.6 170.5 NO.2 FLOW THOUSAND BTU/SEC.0 162.7 160.3 159.2 153.8 151.8 148.4 135.0 128.1 122.8 223.3 531.6 581.6 581.3 574.9 560.8 547.3 563.8 561.6 549.9 538.9 528.5 518.7 509.4 500.5 160.2 244.5 240.1 219.4 210.4 194.1 179.5 169.7 166.5 154.9 137.4 128.9 121.3 114.6 111.6 98.9 89.8 84.4 80.4 76.9 74.5 72.7 71.4 70.9 71.4 mA19444w.wpf:1d441195 3.5-31 TABLE 3.5-5 DONALD C.COOK NUCLEAR PLANT UNITS I AND 2 DOUBLE-ENDED PUMP SUCTION GUILLOTINE MINIMUM SI PRINCIPAL PARAMETERS DURING REFLOOD TIME SECONDS 28.0 28.5 28.9 29.2 29.4 30.5 33.'I 36.2 41.2 43.0 44.2 47.2 48.2 50.1 58.2 61.2 62.3 63.3 66.g 74.3 95.3 106.3 121.3 136.3 155.3 171.8 191.3 210.0 229.3 231.3 249.7 DEGREE F 219.9 2 I7.1 215.1 214.8 214.8 215.1 216.7 218.7 221.6 222.6 223.4 225.3 226.0 227.2 232.9 235.1 235.8 236.5 238.4 242.8 244.3 242.3 243.4 244.0 243.2 244.3 243.5 244.3 243.8 244.3 244.2 244.1 IN/SEC.000 19.962 8.452 3.412 4.303 2.570 2.141 2.220 3.530 3.401 3.319 3.149 3.256 3.177 2.899 2.817 3.092 3.715 3.515 3.035 2.628 2.224 1.949 1.694 1.537 1.423 1.375 1.347 1.338 1.334 1.334 1.337 FLOODING TEMP RATE CARRYOVER FRACTION.000.000.018.041.314.534.635.710.724.731.742.745.750.761.764.765.763.766.770.771.769.768.766.764.765.765.768.770.774.774.776 CORE HEIGHT FT.00.56 1.05 1.18 1.24 1.50 1.77 2.00 2.36 2.50 2.60 2.81 2.87 3.00 3.50 3.67 3.73 3.80 4.01 4.52 5.00 5.56 6.00 6.53 7.00 7.55 8.00 8.52 9.00 9.49 9.54 10.00 DOWNCOMER HEIGHT FT.00.35.55.97 1.27 3.00 7.00 11.55 15.99 16.00 16.00 16.00 16.00 16.00 16.00 16.00 15.98 15.79 15.23 13.95 12.92 12.02 11.55 11.31 11.36 11.68 12.07 12.61 13.17 13.75 13.81 14.37 FLOW FRACTION.250 1.000 1.000 1.000 1.000.598.485.445.578.575.573.566.576.575.562.558.612.628.626.62'I.614.604.594.581.571.563.560.559.581.563.563.565 (POUNDS MASS PER.0.0 7168.8 7168.8 7070.4 7070.4 6999.3 6999.3 6953.1 6953.1 6672.9 6672.9 6194.7 6194.7 5728.8 5728.8 4713.3 4713.3 4549.3 4549.3 4455.0 4455.0 4240.3 4240.3 4472.3 4049.6 4397.4 3971.2 3981.5 3548.8 3851.3 3416.7 437.7.0 403.2.0 408.8.0 421.8.0 431.8.0 441.3.0 447.3.0 452.4.0 455.8.0 458.8.0 460.6.0 462.4.0 463.9.0 465.3.0 465.5.0 466.7.0 SECOND).0.0.0.0.0.0.0.0.0.0.0.0.0.0.0.0.0.0.0.0.0.0.0.0.0.0.0.0.0.0.0.0 INJECTION TOTAL ACCUMULATOR SPILL ENTHALPY BTU/LB M..00 89.50 89.50 89.50 89.50 89.50 89.50 I 89.50: 89.50 89.50 89.50 89.50 87.94 87.90 87.71 87.64 72.99 72.99 72.99 72.99 72.99 72.99 72.99 72.99 72.99 72.99 72.99 72.99 72.99 72.99 72.99 72.99 mAI9444w.wpI:1d 041195 3.5-32 TABLE 3.5+DONALD C.COOK NUCLEAR PLANT UNITS 1 AND 2 DOUBLE.ENDED PUMP SUCTION GUILLOTINE MINIMUM SI POST REFL'OOD MASS AND ENERGY RELEASE TIME SECONDS 249.7 254.7 259.7 264.7 269.7 274.7 279.7 284.7 289.7 299.7 304.7 309.7 314.7 324.7 329.7 339.7 344.7 359.7 364.7 399.7 404.7 424.7 449.7 454.7 469.7 474.7 489.7 499.7 519.7 529.7 534.7 544.7 549.7 554.7 574.7 579.7 609.7 614.7 870.8 870.9 874.7 1979.7 1982.3 2222.2 2316.8 LBM/SEC 200.8 201.2 200.4 200.8 200.0 200.3 199.6 199.9 199.0 199.6 198.7 199.0 198.1 198.5 197.6 197.9 197.0 197.2 196.2 195.8 194.8 194.4 193.3 193.8 192.5 192.8 191.7 191.7 190.7 190.9 190.1 190.1 189.5 189.9 188.6 188.6 187.3 82.0 82.0 79.9 79.8 64.8 64.7 64.7 63.4 63.4 THOUSAND BTU/SEC 251.0 251.5 250.6 251.0 250.0 250.5 249.5 249.8 248.8 249.5 248.5 248.7 247.7 248.1 247.0 247.4 246.2 246.5 245.3 244.7 243.5 243.1 241.6 242.3 240.6 241.0 239.7 239.7 238.4 238.7 237.7 237.7 236.9 237.4 235.8 235.8 234.1 102.5 102.5 94.4 99.4 75.2 80.5 80.5 78.8 78.8 BREAK PATH NO.1 FLOW BREAK LBM/SEC 285.3 284.9 285.7 285.3 286.1 285.8 286.6 286.3 287.1 286.6 287.4 287.2 288.0 287.6 288.5 288.2 289.2 289.0 289.9 290.4 291.4 291.7 292.9 292.3 293.7 293.3 294.4 294.4 295.5 295.2 296.0 296.0 296.7 296.2 297.5 297.5 298.8 404.1 404.1 406.3 406.3 421.4 103.3 103.3 332.2 332.2 103.3 103,2 103.2 103.0 103.1 102.9 103.0 102.8 102.9 102.5 102.6 102.5 102.5 102.2 102.3 102.0 102.1 101.7 101.8 101.2 101.3 100.9 100.7 100.4 100.4 100.2 100.2 99.9 99.7 99.5 99.5 99.3 99.4 99.2 99.0 98.9 98.5 120.8 120.8 88.7 155.7 84.8 96.5 96.5 158.0 158.0 PATH NO 2 FLOW THOUSAND BTU/SEC mal 944+w.wpf:1d441195 3.5-33 TABLE 3.5-7 DONALD C.COOK NUCLEAR PLANT UNITS 1 AND 2 DOUBLE-ENDED PUMP SUCTION GUILLOTINE MINIMUM SI'ASS BALANCE TIME (SECONDS).00 28.00 28.00 249.66 870.94 2316.82 INITIAL IN RCS AND ACC ADDED MASS PUMPED INJECTION TOTAL ADDED-TOTAL AVAILABLE"'ASS (THOUSAND LBM)771.32 771.32 771.32 771.32 771.32 771.32.00.00.00 91.01 393.02 1087.36.00.00.00 91.01 393.02 1087.36 771.32 771.32 771.32 862.33 1164.34 1858.68 DISTRIBUTION REACTOR COOLANT 537.32 57.74 67.87 135.93 135.93 135.93 ACCUMULATOR 234.00 171.20 161.07.00.00.00 TOTAl CONTENTS 771.32 228.94 228.94 135.93 135.93 135.93 EFFLUENT BREAK FLOW ECCS SPILL TOTAL EFFLUENT'-TOTAL ACCOUNTABLE-
.00 542.36 542.36 726.39 1028.39 1722.73.00.00.00.00.00.00.00 542.36 542.36 726.39 1028.39 1722.73 771.32 771.30 771.30 862.31 1164.32 1858.66 m%1944-6w.wpt:td.04t 195 3.5-34 TABLE 3.5-8 DONALD C COOK NUCLEAR PLANT UNITS 1 AND 2 DOUBLE-ENDED PUMP SUCTION GUILLOTINE MINIMUM Sl ENERGY BALANCE TIME (SECONDS).00 28.00 28.00 249.66 870.94 2316.82 INITIAL ENERGY IN RCS,ACC,S GEN ADDED ENERGY PUMPED INJECTION DECAY HEAT ENERGY (MILLION BTU)901.43 901.43 901.43.00.00.00 8.96.00 8.96-5.10 TOTAL ADDED'"" TOTAL AVAILABLE-'ISTRIBUTION REACTOR COOLANT ACCUMULATOR CORE STORED PRIMARY METAL SECONDARY METAL EFFLUENT STEAM GENERATOR TOTAL CONTENTS BREAK FLOW ECCS SPILL TOTAL EFFLUENT'"TOTAL ACCOUNTABLE-'00 3.87 901.43 905.30 318.00 12.74 20.94 15.32 28.06 13.71 178.97 168.74 84.19 84.08 271.26 275.82 901.43 570.41.00 334.41.00.00.00 334.41 901.43 904.82 3.87 905.30 13.64 14.42 13.71 168.74 84.08 275.82 570.41 334.41.00 334.41 904.82 HEAT FROM SECONDAR.00-5.10 901.43 6.64 34.20-5.10 35.75 937.18 30.54.00 3.19 143.60 77.83 252.03 507.19 421.73.00 421.73 928.92 901.43 28.69 87.00-2.19 113.49 1014.92 30.54.00 3.16 92.94 59.46 189.66 375.76 630.90.00 630.90 1006.66 901.43 84.57 181.86 4.03.270.46 1171.89 30.54.00 2.92 63.41 35.57 116.14 248.57 916.99.00 916.99 1165.56 rnA19444w.wpf:1d441195 3.5-35 TABLE 3.5-9 ENERGY ACCOUNTING IN MILLIONS OF BTU Ice Heat Removal Structural Heat Sinks RHR Heat Exchanger Heat Removal Spray Heat Exchanger Heat Removal Energy Content of Sump Ice Melted (Pounds)(10')Approx.End of Blowdown (t=10.0 sec.)207.7 17.37 188.94 0.67 Approx.End of Ref lood (t=249.7 sec.)250.3 44.73 250.0 0.84 Integrated Energies m%1 9446w.wpf:
1 d641195 3.5-36 TABLE 3.5-10 ENERGY ACCOUNTING IN MILLIONS OF BTU Approx.Time of Approx.Time of Ice Melt Out Peak Pressure (t=5423 sec.)(t=7752 sec.)Ice Heat Removal'tructural Heat Sinks RHR Heat Exchanger Heat Removal'pray Heat Exchanger Heat Removal'nergy Content of Sump Ice Melted (Pounds)(10')
567.21 82.52 49.0 58.31 583.6 2.11 567.21 112.68 77.31 92.3 599.3 2.11 Integrated Energies m:519444w.wpf:
1 d 441 195 3.5-37 TABLE 3.5-11 STRUCTURAL HEAT SINK TABLE SURFACES Upper Compartment Material 1.Paint Carbon Steel Concrete 2.Paint Concrete 3.Paint Concrete 4.Paint Concrete Lower Compartment Material 5.Paint Concrete 6.Paint Concrete 7.Paint Concrete 8.Paint Concrete AREA (Ft)32500.32500.32500.10086.10086.5880.5880.11970.11970.5069.5069.13660.13660.16730.16730.8665.8665.THICKNESS (Ft)0.001083 0.0469 2.0 0.001083 2.0 0.00125 1.5 0.00125 1.0 0.00125 2.0 0.00125 1.5 0.00125 1.0 0.00125 2.0 Ice Condenser 9.Steel 10.Steel 11.Steel 12.Paint Concrete 180600.76650.28670.3336.3336.0.00663 0.0217, 0.0267 0.000833 0.333 m:519444w.wpf:1d441195 3.5-38 TABLE 3.5-11(continued)
STRUCTURAL HEAT SINK TABLE SURFACES Ice Condenser 13.Steel and Insulation Steel 14.Steel and Insulation Concrete AREA (Ft')19100.19100.13055.13055.THICKNESS (Ft)1.0 0.0625 1.0 1.0 mA1944+w.wpt:1d441195 3.5-39 TABLE 3.5-12 MATERIAL PROPERTIES TABLE Material Concrete Conductivity Volumetric Heat Capacity Btu/hr-'ft-'F Btu/ftF 0.8 Steel 26.0 56.4 m:$19444w.wpf:1d441195 3.5-40 TABLE 3.5-13 UNIT 1/UNIT 2 STEAMLINE BREAK MASS/ENERGY RELEASES INSIDE CONTAINMENT 30%Power, 0.942 ft'plit Break Failure-MSIV TIME~sec.0000.2000 5.600 7.000 10.00 13.00 13.60 14.80 15.60 16.00 18.00 20.00 26.00 35.00 40.00 45.00 50.00 60.00 70.00 80.00 90.00 100.0 110.0 120.0 150.0 200.0 270.0 290.0 292.5 295.0 297.5 320.0 337.5 352.5 367.5 395.0 410.0 432.5 495.0 605.0 MASS~tbm/eec.0000 1873.1744.1734.1718.1703.1698.1688.1681.1677.1629.1522.1284.1061.974.2 905.9 853.1 782.0 741.1 719.1 707.4 701.3 698.3 696.7 695.1 694.1 692.9 691.3 667.1 607.8 554.0 476.8 403.4 344.5 296.7 183.5 136.6 114.3 106.6 109.5 ENERGY~MBtu/sec.0000 2.234 2.085 2.073 2.054 2.036 2.031 2.020 2.011 2.007 1.950 1.825 1.544 1.277 1.173 1.091 1.028.9418.8925.8659.8517.8444.8408.8388.8369.8357.8342.8323.8028.7312.6658.5711.4820.4106.3531.2174.1609.1345.1252.1282 mA19444w.wpf:1d~1195 3.5-41 TABLE 3.5-13 (continued)
UNIT 1/UNIT 2 STEAMLINE BREAK MASS/ENERGY RELEASES INSIDE CONTAINMENT 102%Power, 1.4 ft'ouble Ended Rupture Failure-MSIV TIME~sec.0000.2000 1.400 3.800 6.000 8.000 10.00 11.60 12.00 12.20 13.00 14.20 15.20 16.00 16.40 17.00 22.00 24.00 26.00 28.00 30.00 32.00 34.00 36.00 45.00 75.00 100.0 200.0 280.0 282.5 285.0 287.5 292.5 300.0 320.0 350.0 610.0 MASS~Ibm/sec.0000 9753.8708.7436.7228.7069.6882.6658.6441.6224.5353.4047.2959.2090.1657.1482.1306.1253.1208.1169.1137.1110.1087.1068.1006.878.0 851.6 831.4 825.3 789.4 695.6 619.9 455.2 241.3 112,0 106.5 3.231 ENERGY~MBt u/sec.0000 11.68 10.45 8.940 8.693 8.504 8.281 8.014 7.752 7.490 6.443 4.871 3.562 2.516 1.995 1.784 1.572 1.509 1.455 1.408 1.369 1.337 1.309 1.285 1.211 1.056 1.024.9998.9924.9485.8349.7431.5429.2850.1.308.1244 m'31944 4w.wp1.'1d~
1 195 3.5-42 10 0 10 1 10 2 10 Time (e)3 10 Figure 3.5-1 LOCA Mass Energy.Release Containment Integrity Containment Pressure mh19444w.wpt:1d~1 195 3.5-43 SO 0 10 1 10 2 10 Time{I)Figure 3.5-2 LOCA Mass Energy Release Containment Integrity Upper Containment Temperature m:L1944+W.WPf:1d441195 3.5-44 220 210 O 200 I~~~~~~~~~~~~~~~~~~0~~I~~~~~~~~~~~~~~~~~~~~~~~~~~~~~170~~~~~~~~~~~~~~~~~0~~~~~~~~~0 10 t 10 2 10 Time (e)10 10 Figure 3.5-3 LOCA Mass Energy Release Containment Integrity Lower Containment Temperature mA1944+w.wpf:1d~)
195 3.5A5 180 1?0 LL Q P 160 CL 8 I~~0$~\~0 10 1 10 2 10 Time (sj 3 10 10 Aottve Sump Twnp.Inaotfva Sump Temp.Figure 3.&4 LOCA Mass Energy Reiease Containment integrity Active Sump and Inactive Sump Temperature Transient
,25E+07.2E+07.15f+Ol~~~~.1E+07~~~~~~~t~~~~~~~~~~~~~~0~~~~~500000~~~t~~~~0 10 10 10 Time (e)Figure 3.5-5 LOCA Mass Energy Release Containment Integrity Ice Melt Transient m:51944<w.wpt:1dM1195 3.5-47 u 110 I CO I CL 8 Q+100~~~y~~~o o 80 0 Time (s)~PIRATURt (Fj Figure 3.5-6 102%Power, 1.4 sq.ft.Double Ended Rupture-MSIV Failure~~~Upper Compartment Temperature mal 944<w.wpf:1d~1195 3.5-48
~I I~I'.~~~~~~~~I~~~~
CO 40 CL>e fh thCL Time (s)PRCSSURE (PSIO)F igure 3.5-8 102%Power, 1.4 sq.ft.Double Ended Rupture-MSIV Failure Upper Compartment Pressure m%1 944+w.wpf:1d441195 3.5-50 0 PRSSSURR (PSIO)Figure 3.5-9 102%Power, 1.4 sq.ft.Double Ended Rupture-MSIV Failure~~Lower Compartment Pressure m:519444w.wpf:1d441195 3.5-51 I~~~~-I'~I~~~~~~'
8508 2so K E I~~0~~~t 0~Time (s)TEMPIRATURE P)Figure 3.5-1'i 30%Power, 0.942 sq.ft.Split Break-MSIV Failure Lower Compartment Temperature m:519444w.wpf:1dM1195 3.5-53
CR Yl 54~~~~~~0 Tlma (I)PRRSSURR (tOIO)Figure 3.5-12 30%Power, 0.942 sq.ft.Split Break-MSIV Failure~~~~~~Upper Compartment Pressure m31944+w.wpf:1d~1195 3.5-54
~~~
'

3.6 STEAM

GENERATOR TUBE RUPTURE ACCIDENT ANALYSIS The UFSAR analysis of a steam generator tube rupture (SGTR)transient is performed to conservatively predict the radiological consequences of such an event.An evaluation of this transient, supporting an increase to 30%SGTP for Donald C.Cook Nuclear Plant Unit 1, has been completed to determine the impact on the dose releases.The primary thermal hydraulic parameters which affect the calculated offsite radiation doses for a SGTR event are the assumed radioactivity level in the reactor coolant, the reactor coolant released through the ruptured tube to the secondary steam generator volume, and the steam released from the ruptured steam generator to the atmosphere.
With respect to the UFSAR analysis, a change in the SGTP level does not impact the reactivity level of the reactor coolant.However, both the potential'primary coolant release to the secondary and the secondary steam release to the atmosphere are impacted by the assumed tube plugging level~To evaluate the effect of a 30%SGTP level, the mass releases from the RCS and from the secondary volume to the atmosphere were calculated.
Four cases were considered assuming a nominal RCS temperature of 533.0'F and 576.3'F with both symmetric and asymmetric RCS flow conditions.
A nominal full power level of 3262 MWt (NSSS)was also assumed.The thyroid and whole body doses estimated for Cook Nuclear Plant Unit 1, based on the 30%SGTP evaluation, remain within a"small fraction" (10%)of the 10CFR100 exposure limit guidelines.
Small fraction is the smallest of the exposure guide lines defined in NUREG-0800 (Standard Review Plan).Therefore, the conclusions of the UFSAR remain valid.m:519444w.wpf:1d441195 3.6-1 3.7 POST-LOCA HOT LEG RECIRCULATION TIME The hot leg switchover calculation performed to preclude boron precipitation during cold leg recirculation following a LOCA is not adversely affected by the changes associated with the SGTP Program.The analysis is not affected since the proposed changes do not affect the normal plant operating parameters, the safeguards systems actuations, or the accident mitigation capabilities important to the calculation.
The SGTP Program does not significantly change the assumptions used in the analysis.Any variation in the initial RCS fluid inventory due to tube plugging is judged to be insignificant with respect to the calculated time for hot leg switchover to preclude boron precipitation.
Other fluid inventories and boron concentrations remain unchanged.
Furthermore, the proposed changes do not create conditions more limiting than those assumed in the hot leg switchover calculation.
m:51944+w.wpf:1d441195 3.7 1
!!I

3.8 REACTOR

CAVITY PRESSURE ANALYSIS The Reactor Cavity Pressure Analysis is performed to calculate the initial pressure response in the reactor cavity to a loss of coolant accident.The Reactor Cavity Pressure Analysis that was performed for the Rerating Program was reviewed and it was determined that the conclusions provided for the Rerating Program (WCAP-11902) remain valid for the SGTP Program.The SGTP Program parameters affect the Reactor Cavity Pressure Analysis through the mass and energy releases provided as input into the analysis.There is no direct impact of SGTP level on short-term mass and energy release rate calculations and containment subcompartment response analysis (See Section 3.5.1).The major impact results from changes to RCS temperature.
For short-term effects, higher release rates typically result from cooler RCS conditions.
The mass and energy releases used as input for the Reactor Cavity Pressure Analysis reflected limiting conditions and therefore, the NSSS performance parameters for the SGTP Program did not impact the results.m%1 9444w.wpf:1d441195 3.8-1

3.9 RADIOLOGICAL

ANALYSIS A review was performed of the radielogical analysis in the UFSAR for Cook Nuclear Plant Unit 1 to determine the effects of the SGTP Program.The source terms for LOCA and the fuel handling accident are unaffected by the increase in SGTP level.However, a reanalysis of the offsite doses following a large break LOCA was performed for the increase in EDG start time to 30 seconds.The EDG start time delay resulted in a delay in spray injection flow to containment of 115 seconds, whereas the previous analysis assumed no delay in spray injection flow to containment.
While there was a slight increase in the offsite thyroid doses, the doses are within the applicable limits.Source terms for the SGTR event were recalculated at the SGTP Program conditions.
The radiological consequences of the SGTR are summarized in Section 3.6 of this report.For the steamline break radiological analysis, the offsite thyroid dose and the corresponding steam generator primary-to-secondary leak rate determined for the alternate steam generator tube plugging criteria program (APC)is, by design of the methodology, relatively insensitive to the amount of steam released to the environment.
For the steam generator in the ruptured loop, all of the initial iodine activity along with all of the primary-to-secondary leakage activity is released to the environment.
Any additional steam release from this steam generator would be due to the introduction of clean aux feedwater, which would not increase the activity released to the environment (i.e., can't release more than 100%of the activity from this steam generator).
Steam released form the unaffected steam generators due to boiling of the secondary coolant accounts for approximately 1%of the total activity release.Additional steam released from these steam generators will have no significant impact on either the calculated offsite dose or the allowable primary-to-secondary leak rate.Therefore, a steamline break radiological analysis was not required for the SGTP Program.mal 944+w.wpf'.1d~
1195 3.9-1 3.10 FLUID AND AUXILIARY SYSTEMS EVALUATIONS 3.10.1 Fluid Systems Evaluation 3.10.1.1 Introduction This section addresses the impact of the SGTP Program on the ability of the Reactor Coolant System and auxiliary fluid systems to perform their required functions.
The parameters considered are listed in Table 2.1-1.In order to support the operation of Cook Nuclear Plant Unit 1 at the SGTP Program conditions, the following systems were evaluated at the new conditions:
1)Reactor Coolant System{RCS), 2)Chemical and Volume Control System (CVCS), 3)Emergency Core Cooling System (ECCS), 4)Residual Heat Removal System (RHR), and 5)Spent Fuel Pool Cooling System (SFPCS).A brief description of each system is provided below.The Emergency Core Cooling System flowrates were revised as part of the SGTP Program.These ECCS flowrates reflected a charging pump head degradation of 10%{differential pressure of 2290 psid on recirculation), a safety injection and RHR pump head degradation of 15%(differentia pressures of 1326 psid and 150 psid, respectively, on recirculation).
The ECCS flowrates were used in the safety analyses and evaluations for the SGTP Program.3.10.1.2 Description of Fluid Systems Reactor Coolant S stem The RCS consists of four identical heat transfer loops connected in parallel to the reactor vessel.Each loop contains a reactor coolant pump and a steam generator.
In addition, the system includes a pressurizer, a pressurizer relief tank, interconnecting piping, and instrumentation necessary for operational control.During operation, the RCPs circulate pressurized water through the reactor vessel and the four coolant loops.The water, which serves both as a coolant, moderator, and solvent for boric acid (chemical shim control), is heated as it passes through the core.It then flows to the steam generators, where the heat is transferred to the steam system, and returns to the RCPs to repeat the cycle.RCS pressure is controlled by the use of the pressurizer where water and steam are maintained in equilibrium by electrical heaters and water sprays.Steam can be formed (by the heaters)to increase RCS pressure or condensed (by.the pressurizer spray)to reduce the pressure.Three spring loaded safety valves and three power operated relief valves are connected to the pressurizer and discharge to the pressurizer relief tank, where the steam is condensed and cooled by mixing with water.m:51 944<w.wpt:
1 d441195 3.10-1 Chemical and Volume Control S stem The CVCS provides for boric acid addition, chemical additions for corrosion control, reactor coolant clean-up and degasification, reactor coolant make-up, reprocessing of water letdown from the RCS, and RCP seal water injection.
During plant operation, reactor coolant flows through the shell side of the regenerative heat exchanger and then through a letdown orifice.The regenerative heat exchanger reduces the temperature of the reactor coolant and the letdown orifice reduces the pressure.The cooled, low pressure water leaves the reactor containment and enters the auxiliary building.A second temperature reduction occurs in the tube side of the letdown heat exchanger followed by a second pressure reduction due to the low pressure letdown valve.After passing through one of the mixed bed demineralizers, where ionic Impurities are removed, coolant flows through the reactor coolant filter and enters the Volume Control Tank (VCT).Emer en Core Coolin S stem The ECCS injects borated water into the reactor following a break in either the reactor or steam systems in order to cool the core and prevent an uncontrolled return to criticality.
Two each safety injection pumps and residual heat removal pumps take suction from the RWST and deliver borated water to four cold leg connections via the accumulator discharge lines.In addition, two centrifugal charging pumps take suction from the RWST on Sl actuation and provide flow to the RCS via separate Sl connections on each cold leg.This arrangement of Sl pumps can provide safety injection flow at any RCS pressure up to the set pressure of the pressurizer safety valves.Residual Heat Removal S stem The RHRS is designed to remove sensible and decay heat from the core and reduces the temperature of the RCS during the second phase of plant coo!down.As a secondary function, the RHRS is used to transfer refueling water between the RWST and the refueling cavity at the beginning and end of refueling operations.
The RHRS consists of two residual heat exchangers, two RHR pumps and associated piping, valves and instrumentation.
During system operation, coolant flows from one hot leg of the RCS to the RHR pumps, through the tube side of the residual heat exchangers and back to two RCS cold legs.The residual heat exchangers are of the shell and U-tube type.Reactor coolant circulates through the tubes, while component cooling water circulates through the shell.S ent Fuel Pool Coolin S stem The SFPCS removes the decay heat generated by spent fuel elements stored in the spent fuel pool.A secondary function is to maintain the clarity and purity of the spent fuel pool m:119444w.wpf:1d 441195 3.1 0-2 water.The SFPCS serves the spent fuel pool'which is shared between the two Cook Nuclear Plant units.The system design incorporates two cooling trains.Each of the two cooling trains in the SFPCS consists of a pump, heat exchanger, strainer, associated piping, valves and instrumentation, and a shared discharge.
The pump draws water from the pool, circulates it through the heat exchanger and returns it to the pool.The heat exchangers are of the shell and U-tube type;component cooling water circulates through the shell, and spent fuel pool water circulates through the tubes.3.10.1.3 Fluid Systems Evaluation The impact of the SGTP Program on the ability of the following fluid systems to perform their required functions has been evaluated for the RCS, CVCS, ECCS, RHRS, and SFPCS.Reactor Coolant S stem The capability of the RCS to operate was evaluated at the SGTP Program conditions listed in Table 2.1-1.The capacities of the pressurizer spray and power operated relief valves, pressurizer surge line, relief line, RTD bypass delay times and pressurizer relief tank setpoints were evaluated.
It was concluded that the design pressurizer spray flow rate of 750 gpm can not be achieved at the 30%SGTP Program conditions.
The pressurizer spray flow rate was calculated to be 596 gpm.This was determined to be adequate such that the PORVs are not actuated following a 10%step load decrease from full power.Results of this analysis indicate that the reduced spray valve capacity is adequate to prevent PORV actuation for the nominal operating pressures of 2100 and 2250 psia as well as over the range of full load average RCS temperatures between 553'F and 576.3'F.The pressurizer surge line pressure drop was evaluated during a design basis surge.The design basis surge results from three safety valves relieving at the design capacity.It was determined that the RCS maximum pressure at the discharge of the RCP is 2745 psia which is below the ASME maximum allowable pressure for the RCS.The pressurizer relief line pressure drop calculation was unaffected by the revised NSSS parameters for the SGTP Program.The results of the evaluations showed that the installed PORV capacity of 630,000 Ibm/hr is adequate for the design basis load swings for operation at all SGTP operating conditions.
The RTD Bypass delay time calculations indicate that the fluid transport delay times for the existing piping network remain below 1.0 second in all loops and are, therefore, acceptable.
m:11 9444w.wpt:1d~1195 3.10-3
The pressurizer relief tank setpoints were found to be acceptable.
The PRT pressure will be maintained below the rupture disc set pressure following a design basis discharge with the current level setpoints.
Chemical and Volume Control S stem The regenerative and letdown heat exchangers are designed to cool letdown flow from Tto 115'F.This reduction in temperature is required to ensure that the normal RCP seal injection temperature requirement of 130'F will be maintained, including an allowance for a 15'F temperature rise across the centrifugal charging pump.The variations in T~considered for the SGTP Program are bounded by the design inlet temperature of 547'F for the regenerative heat exchanger.
Therefore, the cooling requirements of the letdown function are met with the revised operating parametets.
The letdown function is designed to reduce the static pressure of the reactor letdown stream from the RCP suction pressure to VCT operating pressure, such that the design pressure of intervening piping and components is not exceeded and fluid is maintained in a subcooled condition throughout the system.The majority of the pressure reduction is taken across the letdown orifices.The pressure control valve, QRV-301, ensures that adequate back pressure is maintained on the letdown onfices to ensure subcooled fluid conditions.
The pressurizer pressures considered (2100 or 2250 psia)are bounded by the design pressurizer operating pressure.In addition, it has been verified that QRV-301 is capable of maintaining sufficient backpressure on the letdown orifices to ensure subcooled fluid conditions when the pressurizer pressure is reduced to 2100 psia.Therefore, the pressure reduction requirements of the letdown function are met with the revised operating parameters.
Emer enc Core Coolin S stem The primary system pressures considered for this program are less than or equal to the primary system pressure against which the original system was designed to deliver.The required core cooling flow rate is proportional to reactor power level which has not changed as a result of this program.Therefore, the revised primary system parameters do not require an increase in either the motive pressure or core cooling capacity of the ECCS.Residual Heat Removal S stem The RHRS is normally placed in operation approximately four hours after reactor shutdown when the pressure and temperature of the RCS are approximately 400 psig and 350'F, respectively.
Under normal operating conditions, the RHRS is designed to reduce the temperature of the reactor coolant to 140'F within 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months following reactor shutdown, with both trains operating.
In the event of a train failure, the RHRS is designed to reduce the reactor coolant temperature to 200'F within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months after reactor shutdown.Since the initiation temperature and decay heat generation rates (power level)have not changed from m:51 9444w.wpf:1d441195 3.10-4 those previously evaluated for the Rerating Program, the demands on the RHRS are not affected.Therefore, the RHRS is still capable of reducing the reactor coolant temperature to 140'F within the 20 hour2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months limit for normal operating conditions, when both trains are operating.
In the event of a train failure, the RHRS is still capable of reducing the reactor coolant temperature to 200'F within the 36 hour4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months limit.S ent Fuel Pool Coolin S stem The primary function of the SFPCS is to remove decay heat which is generated by the spent fuel pool elements stored in the pool.Decay heat generation is proportional to plant power level.Since the plant power level of 3262 MWt remains unchanged from that previously evaluated for the Rerating Program, the demands on the SFPCS are not increased.
The purification function is controlled by SFPCS demineralization and filtration rates, which are not affected by the SGTP Program.3.10.2 NSSS/Balance of Plant Interface Systems Evaluation The proposed NSSS Performance Parameters for the SGTP Program were compared with those of the Rerating Program.The results of the evaluation show that a SGTP level of 30%will have no adverse effects on the Balance Of Plant (BOP)systems performance.
The Donald C.Cook Nuclear Plant Unit 1 BOP fluid systems and components have been evaluated to assess the effects of increasing the SGTP level up to 30%.The evaluation compared the bounding NSSS performance parameters with the current bounding Unit 1 Rerating Program parameters (Cases 1,3,4,5, 8 6 of WCAP-11902, Supplement 1)to determine the impact on the following BOP systems: Main Steam System Condensate and Feedwater System Auxiliary Feedwater System Steam Generator and Blowdown System The proposed performance parameters which affect the BOP systems and components, compared to the Unit 1 Rerating Program parameters, either do not change, or change in a favorable direction with increased SGTP levels of up to 30%.For example, the SGTP Program power level of 3262 MWt corresponds to the minimum power level evaluated for the Unit 1 Rerating Program.The final feedwater temperature remains unchanged as well as no load Tavg and secondary steam pressure;Also, the steam mass flowrates are bounded by the Unit 1 Rerating Program parameter.
One significant change in parameters is the change in the full power steam pressure where the lower boundirig full power steam pressure (589 psia-Case 4 of Table 2.1-1)is below a lower bounding rerating full power steam pressure (603 psia-Case 4, Table 2.1-1 of WCAP 11902, Supplement 1).This would result in an increase in volumetoc steam flow (cubic feet/sec)for the same power rating.However, m&1 9444w.wpf:
1 d441195 3.1 0-5 since the Rerating Program incorporated a power uprating (3425 MWt)and corresponding mass flow increase, the volumetric flow increase at the reduced power level (3262 MWt)also falls within the bounds of the Unit 1-Rerating Program parameters.
Consequently.
the changes in steam'flow rates and the design considerations associated with steam flow rates are not significant.
Therefore, it was concluded that an increase in SGTP levels of up to 30%will have either no impact or an insignificant impact on the NSSS/BOP fluid systems.They will continue to perform acceptably at the conditions associated with the SGTP Program.m:$1944<w.wpf:1d441195 3.10-6 3.11 PRIMARY COMPONENT EVALUATIONS Evaluations were performed for all NSSS primary and auxiliary components to support the SGTP Program for Cook Nuclear Plant Unit 1.In some cases, structural reanalysis was performed.
In general, the evaluations and analyses were performed assuming the associated NSSS performance parameters case(s)(from Table 2.1-1)most limiting for the particular component.
The NSSS components reviewed for the SGTP Program are as follows:.Section Component 3.11.1 3.11.2 3.11.3 3.11.4 3.11.5 3.11.6 3.11.7 3.11.8 Steam Generators Reactor Vessel Reactor Internals Control Rod Drive Mechanisms Reactor Coolant Pumps Pressurizer Reactor Coolant Loop Piping and Supports Auxiliary Components A summary of the evaluations and analyses is provided below.3.11.1 Steam Generators The following sections describe the analyses and evaluations performed under the Cook Nuclear Plant Unit 1 SGTP Program for the Unit 1 Steam Generators.
The Steam Generators evaluated are the original Model 51-series.
Three separate areas of evaluation are addressed for the SGTP Program: Thermal-hydraulic performance characteristics (including moisture separator performance)
U-bend tube fatigue Structural integrity Thermal-H draulic Performance Evaluation The factors governing the thermal-hydraulic performance of steam generators can be reduced to the thermal power and steam pressure.Other factors such as primary temperature, primary flow and plugging level are important only insofar as they affect the steam pressure.Primary pressure, in the range under consideration, does not affect thermal hydraulic performance.
m:519444w.wpf:1d441195 3.11-1 I I' As part of the Rerating Program, thermal hydraulic performance parameters were evaluated for a range of thermal powers and steam pressures.
The pressure range is bounded by the Rerating Program wheh all powers are considered.
At the 3262 MWt power rating, applicable to the 30%plugging evaluation, the lowest steam pressure is slightly below the value analyzed during the Rerating Program.This will be shown to be of no consequence.
The conclusion of the Rerating Program was that the performance characteristics of the steam generators, including moisture carryover, continue to be acceptable at all the Rerating Program conditions.
This conclusion continues to apply for the 30%SGTP conditions.
Moisture Separator Limits Modifications to the moisture separators at Unit 1 were completed in the Spring of 1989.These modifications include the following elements:~primary separator"top hats" which diffuse the jet issuing from the primary separators,"steam chimneys" which vent steam from below the mid deck plate without entraining liquid drops, and additional upper tier dryer drains.With these modifications, moisture carryover values measured in the field have been near or below 0.1%over a wide range of power levels and steam pressures.
In earlier separator systems, including the unmodified Model 51, moisture strongly increased with power.The top hats eliminate a primary cause of this dependence which is the jets issuing from the primary separators.
The trend of increasing moisture with decreasing steam pressure remains, but its effect is small and the moisture level remains low to the lowest limit of the data, 700 psia.Based on the available field data, moisture carryover is expected to remain comfortably below 0.25%for steam pressures down to 700 psi and below.Other Thermal Hydraulic Characteristics In addition to moisture carryover, the Rerating Program evaluated circulation ratio, hydrodynamic stability, and steam generator mass as additional indicators of acceptable performance.
The change in these parameters from the design value to the values at each of the Rerating Program conditions were calculated.
Steam pressures analyzed for the rating of 3262 MWt had a range of 610 to 820 psia.Variation of the three parameters over this range is presented in Table 3.11-1 along with the design value at<he rated power.It is evident from Table 3.11-1 that the parameters listed are minimally affected by steam pressure at constant power.Circulation ratio is essentially unaffected.
Damping factor is the m:hl 944+w.wpf:1dM1195 3.11-2 measure of hydrodynamic stability, a large negative value indicating a stable unit.This parameter, too, is essentially unaffected by the steam pressure.Steam generator mass is slightly affected by reduced steam pressure.As steam pressure decreases, the voids in the bundle increase reducing the mass inventory.
The effect is small and does not affect operability.
Table 3.11-1 displays the parameter variation down to a steam pressure of 610 psia.For the 30%plugging conditions, the minimum steam pressure is 589 psia.The 21 psi pressure change represented in the table was shown to have minimal affect on the parameters reviewed.The additional 21 psi pressure change to 589 will also be small.Steam generator operating characteristics will be acceptable down to the minimum steam pressure of 589 psia.U-bend Fati ue Evaluation A complete U-bend fatigue evaluation is documented in WCAP-13814, December 1993,"D.C.Cook Unit 1-Evaluation for Tube Vibration Induced Fatigue" (Reference 1).The evaluation was performed to determine the susceptibility to fatigue-induced cracking, consistent with NRC Bulletin 88-02.The evaluations were performed for the current operating conditions as well as for a level of 30%tube plugging.The analysis identified preventative actions for tubes identified as potentially susceptible to U-bend vibration induced fatigue.Structural Inte ri Evaluation Structural integrity evaluation of steam generator components performed for the Donald C.Cook Nuclear Plant Unit 1 Rerating Program included NSSS performance parameter cases that bounded steam generator tube plugging (SGTP)levels of up to 15%.The NSSS design transients developed for the Rerating Program continue to apply to Donald C.Cook Nuclear Plant Unit 1 at the 30%SGTP conditions.
Since the performance parameters and the design transients still apply, the evaluations performed for 15%SGTP would be applicable for all components of the steam generator except the divider plate.A new evaluation of the divider plate, therefore, was performed for a higher pressure differential across the divider plate caused by a higher (30%)tube-plugging level.This analysis demonstrated the structural acceptability of the divider plate.It was therefore concluded that the Cook Nuclear Plant Unit 1 steam generator structural integrity would be maintained for operations with a tube-plugging level of up to 30%.Reference WCAP-13814,"Donald C.Cook Unit 1 Evaluatiorilor Tube Vibration Induced Fatigue," December 1993 mal 944+w.wpf:1d 441 195 3.11-3
3.11.2 Reactor Vessel 3.11.2.1 Introduction" The Addendum Report prepared for the Unit 1 reactor vessel (reference 1)to evaluate the stress and fatigue effects of the operating parameters and RCS transients associated with the Rerating Program remains applicable.
Therefore, no new reactor vessel stress calculations were performed for the SGTP Program.This report evaluates the maximum primary plus secondaiy stress intensity ranges and maximum cumulative fatigue usage factors resulting from the Rerating Program conditions which bound the 30%SGTP conditions.
The calculated stress intensity range and usage factor values are compared with the applicable limits of Section III of the ASME Boiler and Pressure Vessel Code.The operating parameters shown in Table 3.11-2 were used as a basis for the evaluation.
These parameters bound the 30%SGTP parameters contained in Table 2.1-1 of this report.3.11.2.2 Summary of Results The results of the reactor vessel analyses and evaluations are summarized below.Based on these results, all of the stress intensity and fatigue usage limits (with the exception of the 3S maximum range of primary plus secondary stress intensity limit for the control rod drive mechanism housings and outlet nozzle safe end)of the applicable ASME Code version for Unit 1 (reference 2)are met.Exceeding the 3S limit for the CRDM housings and outlet nozzle safe end is reconciled by simplified elastic-plastic analyses in accordance with reference 3.Therefore, the reactor vessel for Cook Nuclear Plant Unit 1 continues to remain in compliance with the applicable Code for the conditions associated with the Rerating and SGTP Programs.Control Rod Drive Mechanism Housin Ada ter The maximum range of primary plus secondary stress intensity is calculated to be 77.76 ksi which exceeds the 3S limit of 69.9 ksi.However, a simplified elastic-plastic analysis was performed in accordance with paragraph NB-3228.3 of the 1971 Edition of Section III of the ASME Boiler and Pressure Vessel Code, and the higher range of stress intensity is reconciled.
The maximum cumulative fatigue usage factor is 0.1687 which is below the ASME Code limit of 1.0.Main Closure Re ion The main closure region of the reactor vessel consists of.the vessel flange, the closure head flange and the closure stud assemblies which couple the head to the vessel.The maximum ranges of stress intensity in the closure head flange and the vessel flange are 65.26 ksi and mA1944<w.wpf:1d~1195 3.11-4 61.04 ksi, respectively, compared to an ASME Code 3S limit of 80.1 ksi.The maximum service in the closure studs is 91.8 ksi which compares favorably to the 3S limit 107.7 ksi.The maximum cumulative fatigue usage factor for the closure head flange, vessel flange and closure studs are 0.018, 0.029 and 0.99, respectively.
The usage factors are all less than the 1.0 ASME Code limit.However, it should be noted that the closure stud usage factor of 0.99 was calculated under the assumption that the first 25 percent of the 11,680 occurrences of plarlt loading and plant unloading at 5 percent of full power per minute (2,920 occurrences of each)occurred during the first 10 years of operation when the vessel outlet temperature (T)was 599.3'F.If the 0.99 usage factor is unacceptably high or if cycle counting indicates that 1.00 may be exceeded, the closure studs are readily replaceable.
Outlet Nozzle The maximum range of primary plus secondary stress intensity in the outlet nozzle safe end is calculated to be 59.58 ksi compared to the 3S limit for the austenitic stainless steel material of 50.1 ksi.Since the maximum range exceeds 3S, a simplified elastic-plastic analysis per paragraph NB-3228.3 of the 1971 Edition of Section III of the ASME Boiler and Pressure Vessel Code was performed which justified the higher maximum range of stress intensity.
The maximum usage factor at the safe end is 0.021 which is less than 1.0.The maximum range of stress intensity in the outlet nozzle and nozzle-to-shell juncture is 57.09 ksi compared to the 3S allowable of 80.1 ksi.The maximum cumulative usage factor in the nozzle and nozzle-to-shell juncture is 0.0631 which is also less than 1.0.Inlet Nozzle The maximum range of stress intensity in the inlet nozzle safe end is 49.65 ksi which is less than 3S=50.1 ksi.The maximum range of stress intensity in the inlet nozzle and nozzle-to-shell juncture is 49.86 ksi which compares favorably with a 3S limit of 80.1 ksi.The maximum cumulative usage factors in the nozzle safe end and nozzle-to-shell juncture are 0.01 74 and 0.0977, respectively, which are both less than 1.0.Vessel Wall Transition The maximum range of stress intensity and cumulative fatigue usage factor for the vessel wall transitionbetween the nozzle shell and the vessel beltline, are 33.57 ksi and 0.0066.These values are less than the ASME Code limits of 80.1 ksi and 1.0, respectively.
Bottom Head-to-Shell Juncture The maximum range of primary plus secondary stress intensity at the juncture between the vessel bottom hemispherical head and the vessel beltline shell is 34.53 ksi compared to a 3S m:i1944+w.wpi:
id~i195 3.11-5 allowable of 80.1 ksi.The maximum cumulative fatigue usage factor at the juncture was calculated to be 0.0182 which is less than 1.0.Bottom Head Instrumentation Penetrations The bottom head instrumentation penetrations are acceptable for the SGTP Program based upon a maximum range of primary plus secondary stress intensity of 51.49 ksi and a maximum cumulative fatigue usage factor of 0.1220.These values compare favorably with ASME Code allowables of 69.9 ksi (3S)and 1.0, respectively.
Core Su ort Pads The core support pads were evaluated to have a maximum range of stress intensity of 69.70 ksi compared to a 3S limit of 69.9 ksi.The maximum cumulative fatigue usage factor was calculated to be 0.693 which is less than the 1.0 ASME Code limit.3.11.2.3 Conclusions The results of the evaluations demonstrate that operation of the reactor vessel in accordance with the conditions associated with the 30%SGTP Program does not result in stress intensities or fatigue usage factors which exceed the acceptance criteria of the applicable ASME Code version for Cook Nuclear Plant Unit 1 (reference 2).Some of the stress intensity ranges are higher than the original stress report.However, all of the stress intensity limits specified in the applicable ASME Code version are still satisfied with the incorporation of 30%SGTP conditions, with the exception of the 3S maximum stress intensity range limit for the CRDM housings and outlet nozzle safe ends.Exceeding 3S in the CRDM housings and outlet nozzle safe ends is reconciled by simplified elastic-plastic analyses in accordance with the requirements of paragraph NB-3228.3 of the 1971 Edition of Section III of the ASME Code (reference 3).3.11.2.4 Reactor Vessel Integrity Evaluation The 30%SGTP conditions for Donald C.Cook Nuclear Plant Unit 1 will not result in an increase in the fast neutron fluence values calculated for the Rerating Program.Based on this information, the reactor vessel integrity analyses performed per the methodology of Regulatory Guide 1.99, Revision 2, as part of the Rerating Program will remain applicable after 30%SGTP.These analyses are applicable for reactor vessel inlet temperatures (Tcold)above 525'F.Operation below 525'F down to 510'F has been further evaluated and found to be acceptable (Reference 4).The increase in SGTP to 30%does not impact the technical support provided in Reference 4 for operation below 525 F-.m:El 944<w.wpf:1d441195 3.11-6 3.11.2.5 References 1.WCAP-11967,"TReduction/Rerating Reactor Vessel Evaluation Addendum to Analytical Report for Indiana and Michigan Electric Company.Donald Cook Nuclear Power Plant Unit No.1 Reactor Vessel", August, 1988, by S.L.Abbott.2.ASME Boiler and Pressure Vessel Code Section III, American Society of Mechanical Engineers, New York."Nuclear Vessels", 1965 Edition with Addenda through the Winter of 1966.3.ASME Boiler and Pressure Vessel Code,Section III, American Society of Mechanical Engineers, New York."Nuclear Power Plant Components", 1971 Edition.4.AEP-93-582,"American Electric Power Service Corporation, Donald C.Cook Nuclear Plant Unit 1, Operation Below 525 degree F", Keith F.Matthews to John Jensen, October 18, 1993.m:41944+w.wpf:
1 d441 195 3.11-7 3.11.3 Reactor Internals\3.11.3.1 Introduction This section documents the results and conclusions of the evaluations performed to investigate the impact of the SGTP Program on the Cook Nuclear Plant Unit 1 reactor vessel internals.
In order to assess the impact of the SGTP Program, the following evaluations were performed.
~Review and Evaluation of Thermal Transients
~Review and Evaluation of Power Level Thermal-Hydraulic Analysis-The analyses included: Evaluation of the effects on core bypass flow Pressure drop distribution in the reactor vessel Component hydraulic lift forces Mechanical System Evaluations which include: Asymmetric flow evaluation Flow induced vibrational
~Components Thermal Stress and Fatigue Evaluations 3.11.3.2 Thermal Transients and Power Level Review Thermal Transients Per Section 2,2, the thermal transients and number of occurrences used in the Rerating Program remain unchanged; therefore the thermal transients evaluation for the Rerating Program remain applicable.
Power Level The power level for the SGTP Program per Table 2.1-1 is 3250 MWt reactor power which is the original design basis for Cook Nuclear Plant Unit 1.The Rerating Program used a power of 3588 MWt reactor power.A change in power level will affect the thermal loads on various reactor internals components such as: Lower Core Support Structure Baffle-Barrel Region Thermal Shield m:51 9444w.wpf:
1 d441195 3.11-8 The decrease in power level from 3588 MWt to 3250 MWt will not have an adverse effect on the structural evaluation performed for the above components.
3.11.3.3 Thermal-Hydraulic Analyses Thermal-hydraulic analyses, as part of the reactor internals qualification for the SGTP Program were performed.
The thermal-hydraulic analyses input parameters were taken from Table 2.1-1.Four different conditions were evaluated using the input parameters for Cases 2 and 3 from Table 2.1-1 (Case 2 O 2250 psi, Case 2 O 2100 psi, Case 3 O 2250 psi, and Case 3 O 2100 psi).The most conservative results for pressure drops, lift forces, and core bypass flow was obtained using Case 2 input parameters O 2250 psi from Table 2.1-1.Core B ass Flow Core bypass flow is defined as the total amount of reactor coolant flow that bypasses the core region, and is not considered effective in the core heat transfer process.Consequently, the effect of increasing bypass flow is a reduction in core power capability.
Evaluations show that the input parameters from Case 2 of Table 2.1-1, provide results with the highest total bypass flow of 4.4%.The resulting total core bypass flow is still within the allowable limit of 4.5%specified for the SGTP Program.S stem Pressure Losses and H draulic Lift Forces The Rerating Program used a conservative evaluation for the system pressure losses, and hydraulic lift forces.The system pressure losses and hydraulic lift forces for the SGTP Program are considered bounded by the Rerating Program.The input parameters, which influence the system pressure losses and hydraulic lift forces for the SGTP Program are the same or lower than those used for the Rerating Program.The SGTP parameters will yield system pressure losses and hydraulic lift forces which are bounded by those used in the Rerating Program.The only parameter which would increase the system pressure losses and hydraulic lift forces is the change in power level, and the change in power level is considered to have an insignificant effect for these parameters.
Therefore, the SGTP Program does not have an adverse effect on the system pressure losses and hydraulic lift forces.3.11.3.4 Mechanical System Evaluation Flow Induced Vibration The parameters which can influence the flow induced vibration characteristics of the reactor vessel internals is the flow and temperature.
The mechanical design flow for the SGTP Program are not changing, only the thermal design flow is changing and it is decreasing.
The mechanical design flow is unchanged and the temperature range is enveloped by the temperature range in the Rerating Program.Therefore, flow induced vibration will not be m:519444w.wpf:1d441195 3.11-9 adversely impacted by the SGTP Program since the flow induced vibration loadings are enveloped by the work performed for the Rerating Program.As mmetric Flow Evaluation The effect of asymmetric flow on the reactor vessel internals was evaluated.
The asymmetric flow loads from Table 2.1-1 were used to evaluate the effect of the new flow condition.
Test data from another plant was also used since the internals for Cook Nuclear Plant Unit 1 are similar to the internals of that plant.The evaluation concluded that the maximum displacements for the core barrel beam mode n=1, and shell modes n=2 and n=3 were enveloped by the test data.Therefore, the assymmetric flow condition is considered acceptable.
3.11.3.5 Component Evaluation Reactor Internals Thermal/Stress and Fati ue Evaluation The reactor internals thermal/stress and fatigue evaluation was performed by using the Rerating Program evaluation as the last qualified operating conditions and evaluating the change in loadings due to the SGTP Program on reactor internals.
Loadings which can impact the evaluation performed are: Thermal Transients and the Number of Occurrences Power Level Gamma Heating Rates Mechanical Loadings Flow Rates Seismic Loads (OBE)Operating Temperature The new loadings for the reactor vessel internals are evaluated in the following section.Load Evaluation The NSSS design transients, Section 2.2, for the SGTP Program remain the same as those previously analyzed for the Rerating Program.Seismic loads, mechanical loads, and gamma heating rates are not affected by steam generator tube plugging.The operating temperatures from Table 2.1-1 were chosen such that they would be enveloped by the operating temperatures used in the Rerating Program.Therefore, for this evaluation, power level and flow rates are the only parameters which are changing thaf can affect the reactor vessel internals evaluation.
The reduction in power level, from 3588 MWt to 3250 MWt, will decrease the gamma heating levels for various reactor vessel components.
A lower gamma heating value will cause the metal temperature due to gamma heating to decrease, which will bring m:$1944+w.wpt.'1d~1 195 3.11-10
the metal temperature closer to the fluid temperature.
This will cause a smaller thermal gradient on the various portions of the reactor vessel internals which are affected by gamma heating.The smaller thermal gradient will cause a lower stress in the affected components.
Since the reduced power level will result in a lower stress state, the previous evaluation for the Rerating Program is considered bounding for thermal stress and fatigue.The reduction in thermal design flow is approximately 6%for the normal operating conditions.
The temperature range for the 30%SGTP are within the range of temperatures evaluated and are considered bounded by the Rerating Program.Since there is only a small change in flow rate, this will cause the pressure drop in the reactor vessel to decrease.The decreased pressure drop will result in a smaller calculated stress level for various components, (core barrel and baffle-former plates).The reduction in Thermal Design Flow translates a 5%or less reduction in the forced convection heat transfer coefficients.
A 5%reduction in the film coefficients is not expected to significantly affect the heat convection between the reactor vessel internals and the reactor coolant.Therefore, the thermal evaluation performed for the Rerating Program is considered applicable.
Conclusion The reactor vessel internals stress and fatigue evaluation is considered bounded by the Rerating Program evaluation.
The SGTP Program does not have an adverse effect on the reactor vessel internals since the loads which are changing are actually improving the margins for the reactor vessel internals when compared to the Rerating Program results.Rod Dro Time An assessment was made to confirm the present RCCA drop time limit of 2.4 seconds remains applicable with the SGTP Program conditions.
Based on the analysis performed, it is concluded that the 2.4 second RCCA drop time remains applicable.
3.11.4 Control Rod Drive Mechanisms An evaluation was performed to evaluate the effects of 30%SGTP for Donald C.Cook Nuclear Plant Unit 1.A review of the NSSS Performance Parameters, given in Table 2.1-1, shows that these conditions are bounded and have been evaluated.
Since the NSSS Performance Parameters and the NSSS Design Transients for the SGTP Program are bounded by those of the Rerating Program, the conclusion of the generic analysis performed for the Rerating Program remains valid for the SGTP Program.Varying the hot leg reactor coolant temperature will have no impact on the structural and thermal analysis of the CRDM.Varying the hot leg temperature will affect certain CRDM material properties.
However, the effect will be insignificant.
Since the hot leg temperature range for the SGTP Program is within the bounds of the hot leg temperature range for the m:'11944+w.wpf:1d441 195 3.11-11 Rerating Program, the Rerating Program analysis continues to apply and the design requirements for the CROM pressure boundary are still met.3.11.5 Reactor Coolant Pumps and Motors The Model 93A reactor coolant pumps (RCPs)used in the Cook Nuclear Plant Unit 1 NSSS were reviewed to determine the impact of the NSSS parameters for the SGTP Program provided in Table 2.1-1 of this report.Because the NSSS parameters for the SGTP Program are bounded by those of the Rerating Program and the NSSS design transients are also bounded by the Rerating Program, no additional thermal or structural analysis were required to demonstrate compliance with the codes and standards in effect at the time of the original contract.Varying the cold leg temperature from 536.3'F up to 543.2'F or down to 517.2'F will have no impact on the structural and thermal analysis of the reactor coolant pump.The only difference is related to the material properties used in the analysis.The difference in material properties at the subject temperature is considered negligible.
Therefore, the design requirements for the RCP pressure boundary are still met.The RCP motors were evaluated to determine the worst case loading.The performance of t the motors at these loads has been evaluated and the results are as follows: 1.Continuous operation at the new hot loop rating at 6420 HP.This represents a 7.0%increase over the nameplate rating of the motor.The change in stator winding temperature resulting from the increase will be less than 5'C.Original test data indicates that with this temperature increase included, the NEMA design limits for a Class B winding will not be exceeded.Therefore, continuous operation of the motors under hot loop conditions with 30%SGTP is acceptable.
2.Operation at the new cold loop rating of 8020 HP.The revised load represents a 6.9%increase over the nameplate rating of the motor.Analysis indicates this load increase will cause the stator winding temperature to increase about 7'C.The resulting winding temperature will be less than the Class F NEMA design limits.Therefore, operation of the motors under cold loop conditions with 30%SGTP is acceptable.
mA1944<w.wpf:1d441195 3.11-12 3.Starting with revised load torque under the worst case conditions (maximum reverse flow, cold loop, 80%voltage).The increase in rotor cage winding temperature due to the increased load is small and the total winding temperature is well below the design limit.Therefore, starting under the worst case scenario is acceptable.
The review for the SGTP Program of the Cook Nuclear Plant Unit 1 reactor coolant pumps demonstrates that the SGTP conditions are acceptable for the Model 93A RCP.The design requirements of the RCP pressure boundary are still met.The RCP Motor evaluation determined that the Cook Nuclear Plant Unit 1 motors are acceptable for operation with the 30%SGTP conditions.
3.11.6 Pressurizer 3.11.6.1 Introduction The functions of the pressurizer are to absorb any expansion or contraction of the primary reactor coolant due to changes in temperature and pressure and to keep the RCS at the desired pressure.The first function is accomplished by keeping the pressurizer approximately half full of water and half full of steam at normal conditions, connecting the pressurizer to the RCS at the hot leg of one of the reactor coolant loops and allowing inflow or outflow to or from the pressurizer as required.The second function is accomplished by keeping the temperature in the pressurizer at the water saturation temperature (T,,)corresponding to the desired pressure.The temperature of the water and steam'n the pressurizer can be raised by operating electric heaters at the bottom of the pressurizer and can be lowered by introducing relatively cool water spray into the steam space at the top of the pressurizer.
The limiting locations from a structural standpoint on the pressurizer are the surge nozzle, the spray nozzle, and the upper shell at the point of spray impingement.
The limiting operating condition (relative to the SGTP conditions) of the pressurizer occurs when the RCS pressure is high and the RCS hot leg temperature (T>>)and cold leg temperature (T~,)are low.This is explained as follows: Due to inflow and outflow to and from the pressurizer during various transients the surge nozzle alternately sees water at the pressurizer temperature (T,,)and water from the RCS hot leg at T>>.If the RCS pressure is high (which means that T,, is high)and T>>is low, then the surge nozzle will see maximum thermal gradients and thus experience the maximum thermal stress.Likewise the spray nozzle and upper shell temperatures alternate between steam at T~, and spray which for many transients is at T~p.Thus, if RCS pressure is high (T,, is high)and T~, is low, then the spray nozzle and upper shell will also experience the maximum thermal gradients and thermal stresses.m%1 944<w.wpf:1d441195 3.11-13 3.11.6.2 Description of Analysis and Results The updated analysis performed for.the Donald C.Cook Nuclear Plant Unit 1 SGTP Program for the pressurizer is based on the NSSS design transients provided for the Rerating Program.The design transients are also applicable for the SGTP Program (see Section 2.2).The analysis was performed by modifying the original Cook Pressurizer analysis (Reference 1), which was performed to the requirements of the ASME Code, 1968 Edition (Reference 2).The original analysis was performed using finite element techniques.
Finite element models were constructed for the various parts of the pressurizer.
These were then subjected to the pressure loads, external loads (such as piping loads on the nozzles)and thermal transients.
The models then calculate the primary, secondary and peak stresses for the various conditions.
The, pressurizer maximum pressure and maximum external loads did not increase due to the SGTP Program.Thus, the primary stresses from the original analysis are still valid.Also, the conditions that cause maximum primary plus secondary stress (inadvertent auxiliary spray for spray nozzle and upper shell, and DBE for the surge nozzle)have not changed.Therefore, the only ASME Code requirement affected by the transient modifications was fatigue.The fatigue usage factors are shown in Table 3.11-3 for the critical components.
3.11.6.3 Conclusions A fatigue analysis was performed for the Cook Nuclear Plant Unit 1 pressurizer, incorporating the most conservative conditions of the SGTP Program.The results of this analysis demonstrate that the pressurizer remains in compliance with the applicable ASME Code criteria.3.11.6.4 References 1.Model 51 Series Pressurizer Report, Westinghouse Electric Corporation, October 1974.2.ASME Boiler and Pressure Vessel Code, 1968 Edition,Section III, Article 4.mA19444w.wpt:1d441195 3.11-14 3.11.7 Reactor Coolant Loop Piping and Supports As part of the Rerating Program, an evaluation of the reactor coolant loop piping, primary equipment nozzles, and the primary equipment supports was performed for a set of thermal parameter cases that included 15%SG tube plugging.The program reported results for a rerating and also addressed a number of additional cases to cover various temperatures in the loop piping.In that report the analysis for the loop piping, the primary equipment nozzles, and the primary equipment supports were reconciled to the Rerating Program as well as the SGTP Program conditions.
The 30%SGTP loop piping temperatures are enveloped on both the lower and upper ends by the loop piping temperatures already considered in the Rerating Program.The LOCA hydraulic forcing functions generated for the Rerating Program bound the proposed 30%SGTP conditions.
The NSSS thermal design transients are applicable for both the Rerating and SGTP Program conditions.
The Rerating Program transients and the plant parameters associated with both the Rerating and the SGTP Programs were reviewed for impact on the WCAP-14070 (Reference 1)evaluation for NRC Bulletin 88-08,"Thermal Stresses in Piping Connected to the Reactor Coolant System".The WCAP specifically addressed the auxiliary spray piping.The defined transients are primary loop piping transients and are far enough removed from the auxiliary spray piping to have a negligible impact.The operating parameters for both the Rerating and the SGTP Program conditions have normal operating cold leg temperatures that deviate from the existing design basis values.The range of normal operating cold leg temperatures have been reviewed for impact on the NRC Bulletin 88-08 evaluation and were found to have no impact on the conclusions stated in WCAP-14070.
The Rerating Program transients for Cook Nuclear Plant Unit 1 were reviewed for potential impact on the existing evaluation for the pressurizer surge line thermal stratification analysis.The report that was prepared to demonstrate compliance with NRC Bulletin 88-11,"Pressurizer Surge Line Thermal Stratification" is WCAP-12850 (Reference 2).The reconciliation of the referenced transients applies to both the Rerating Program and the SGTP Program because the transients cover both programs.The results of the evaluation indicate that the fatigue usage factor increases by a small amount (from 0.275 to 0.277).Since the maximum usage factor reported in the WCAP was rounded to a value of 0.30, the result does not change.Because the allowable fatigue usage factor is 1.0, the results are acceptable.
As part of the surge line stratiTication analysis, a set of pressurizer nozzle loadings due to stratification was used as input to the pressurizer evaluation.
Our evaluation shows that the changes to the pressurizer nozzle loadings are not significant, and need not be evaluated further (there are no increases greater that 2%and load decreases were ignored).In conclusion, the reactor coolant loop piping, the primary equipment nozzles, and the primary equipment support loads are acceptable for the SGTP Program conditions because these conditions are already enveloped by the evaluation performed for the Rerating Program.All design basis analysis performed for these components applies to the 30%SGTP condition.
m%1 9444w.wpf:1d441195 3.11-15 I I The Rerating Program transients and plant parameters associated with the Rerating and the SGTP Programs for Donald C.Cook Nuclear Plant have been reviewed, and the impact on the design basis analysis for the NRC Bulletin 88-08 evaluation of the auxiliary spray piping and NRC Bulletin 88-11 evaluation of the pressurizer surge line piping is insignificant.
References WCAP-14070,"Evaluation of Donald C.Cook Units 1 and 2 Auxiliary Spray Piping per NRC Bulletin 88-08," July 1994.2.WCAP-12850,"Structural Evaluation of Donald C.Cook Nuclear Plant Units 1 and 2 Pressurizer Surge Lines, Considering the Effects of Thermal Stratification," January 1991.3.11.8 Auxiliary Components The auxiliary components (pumps, valves, tanks and heat exchangers) were reviewed to determine the impact of the NSSS parameters for the SGTP Program, provided in Table 2.1-1 of this report.Because the NSSS parameters of the SGTP Program are bounded by those of the Rerating Program and the Auxiliary Equipment Transients are either unchanged or still bounded, there is no effect on the auxiliary components of Cook Nuclear Plant Unit 1.m:11944<w.wpt:1d441 195 3.11-16 TABLE 3.11-1 PERFORMANCE CHARACTERISTICS AT 3262 MWT Parameter Design Value Maximum Steam Pressure Minimum Steam Pressure Steam Pressure (psia)Circulation Ratio Damping Factor (hr')Secondary Mass (Ibm x 10')812 5.49-453 113 820 5.45-434 113 610 5.42-482 108 mh1944<w.wpf:1d~1195 3.11-17 I I' TABLE 3.11-2 ASSUMED OPERATING PARAMETERS FOR REACTOR VESSEL STRUCTURAL EVALUATION FOR COOK NUCLEAR PLANT UNIT 1 Design Pressure (psig)Normal Operating Pressure (psig)2485 Upper Bound Lower Bound 2235 1985 Design Temperature
('F)Normal Operating Vessel Inlet Temperature
('F)Normal Operating Vessel Outlet Temperature
('F)650 511.7 615.2 Zero Load Temperature, ('F)547-Design Life: The design life of the reactor vessel is 40 years.The design life is the period of anticipated plant service which is used as a basis for defining the number of occurrences of design transients and external loads to be used in the design fatigue analysis.The design life is not to be considered as a warranty but is used strictly for determining fatigue usage factors for the reactor vessel components.
The reactor vessel is analyzed to operate with normal operating vessel inlet temperatures (T~)from 511.7'F to 547'F and normal operating vessel outlet temperatures (T)from 582.3'F to 615.2'F.The reactor vessel closure studs were analyzed for fatigue usage assuming a normal vessel outlet temperature of 599.3'F for the first 10 years of operation and the maximum normal vessel outlet temperature of 615.2'F for the remaining 30 years.mA19444w.wpf:1d441195 3.11-18 TABLE 3.11-3 DONALD C.COOK 1 PRESSURIZER COMPONENTS, CALCULATED FATIGUE USAGES CONSIDERING 30%SGTP COMPONENT FATIGUE USAGE Surge Nozzle Spray Nozzle Safety and Relief Nozzle Seismic Analysis Lower Head-Heater Well Lower Head Perforations Upper Head and Shell Support Skirt/Flange Heater Vibrations Baffle Vibrations Support Lug Manway Instrument Nozzle Immersion Heater Valve Support Bracket 0.3323 0.99 0.148 0.07 0.0165 0.97 0.011 0.048 0.0 0.1084 0.004 0.01 mal 944+w.wpf:1d441195 3.11-19 3.12 FUEL STRUCTURAL EVALUATION Evaluations were performed of the fuel for Cook Nuclear Plant Unit 1 under the Rerating Program in the areas fuel rod and fuel assembly structural integrity, core design and thermal-hydraulic design.These evaluations assumed a maximum core power level of 3250 MWt and the associated range of operating conditions from Table 2.1-1.3.12.1 Fuel Assembly Structural Evaluation Fuel assemblies are designed to perform as described in the Technical Specifications.
The combined effects of design basis loads are considered in the verification of the fuel assembly and its components to maintain the fuel assembly structural integrity.
This is necessary so that the fuel assembly functional requirements are met, the core eoolable geometry is maintained, and the reactor core can be shut down safely.A structural evaluation of the fuel assembly was performed for the SGTP Program for Cook Nuclear Plant Unit 1, considering the range of operating parameters described in Table 2.1-1.This evaluation assumed 15 x 15 optimized fuel for Unit 1: The summary of the maximum LOCA and DBE grid load results are presented in Table 3.12-1 with consideration of the requirement of grid load combination, the SRSS of the DBE and LOCA maximum loads is less than 2040 lbs.This maximum load is 33.6%of the grid strength for the 15 x 15 OFA fuel assembly design.Thus, the 15 x 15 OFA design has ample margin for resisting faulted conditional loading.The fuel assembly design is structurally acceptable for Donald C.Cook Nuclear Plant Unit 1~In conclusion, the SGTP Program for Cook Nuclear Plant Unit 1 does not significantly increase the operating and postulated transient loads such that they will adversely affect the fuel assembly functional requirements.
The fuel assembly structural integrity is not affected and the core eoolable geometry is maintained for the assumed fuel type for Cook Nuclear Plant Unit 1.3.12.2 Fuel Rod Structural Evaluation An evaluation was performed under the SGTP Program of the impact of NSSS performance parameters in Table 2.1-1 on the ability of fuel to satisfy fuel rod design criteria for Cook Nuclear Plant Unit 1.While fuel rod design analyses are not directly impacted by steam generator tube plugging levels, they are sensitive to core inlet temperature, mass flow rates, and other related parameters.
Table 3.13-2 provides a comparison of the parameters assumed for the Rerating Program against those of the SGTP Program.A review of the thermal models indicates that the-5%reduction in power, or heat flux will generally offset the-6%reduction in mass flow rate, especially when combined with the 2.7'F mA1944+w.wpf:1d441195 3.12-1 reduction in the maximum inlet temperature.
Other fuel performance models, e.g., fission gas release, thermal creep, etc., dependent upon the core power and fuel temperatures.
will also be offset by these effects.As a result, fuel rod design analyses performed for the 30%SGTP parameters would not be anticipated to be more limiting than the Rerating Program analyses for any of the impacted fuel rod design criteria, and the conclusions of Rerating Program will remain valid for SGTP Program for Donald C.Cook Nuclear Plant Unit 1.Finally, as in the past, cycle-specific fuel performance analyses will continue to be performed for each fuel region to confirm that this assessment, and all fuel rod design criteria, are satisfied for the operating conditions specific to each cycle of operation.
These evaluations support the Reload Safety Evaluation (RSE), which is transmitted to AEPSC prior to each cycle of operation.
3.12.3 Core Design The results of the core design evaluation indicated that the increased steam generator tube plugging level and reduced Thermal Design Flow result in no impacts to the core design except for the values of the statepoint for the Steamline Break Analysis, and the Dropped Rod Analysis.See Section 3.3 for the new statepoint values for the Steamline Break Analysis.See Section 3.12.4 for the new limits concerning Dropped Rod Limit Lines.3.12.4.1 Purpose of Analysis The purpose of this section is to describe the thermal-hydraulic analysis necessary to support the decrease in flow associated with an increase in SGTP level to 30%over a range of RCS temperatures.
3.12.4.2 Assumptions Table 3.12-3 summarizes the thermal-hydraulic design parameters used in this analysis.The core inlet temperature is consistent with the high temperature 30%SGTP case.Use of high inlet temperature bounds the range of RCS Tavg with regard to the Departure from Nucleate Boiling (DNB)analysis.Included in Table 3.12-3, for comparison, are the thermal-hydraulic parameters currently in the Donald C.Cook Nuclear Plant Unit 1 Safety Analysis.mA1944+w.wpf:1d441195 3.12-2 3.12.4.3 Discussion of Evaluation 3.12.4.3.1 Calculation Methods The thermal hydraulic design criteria and methods remain the same as those presently in the Donald C.Cook Nuclear Plant Unit 1 UFSAR with the exceptions described in the following paragraphs.
DNB Methodolo The existing thermal-hydraulic analyses use the Improved Thermal Design Procedure (ITDP)(Reference 1).For this methodology, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically, such that there is at least a 95 percent probability at a 95 percent confidence level that the minimum DNBR of the limiting fuel rod is greater than or equal to the DNBR limit of the DNB correlation being used.Plant parameter uncertainties are used to determine the plant DNBR uncertainty.
This DNBR uncertainty, combined with the DNBR limit, establishes Design Limit DNBR values which must be met in plant safety analyses.Since the parameter uncertainties are considered in determining the Design Limit DNBR values, the plant safety analyses are performed using values of input parameters without uncertainties.
In addition, the Design Limit DNBR values are increased to values designated as the Safety Analysis Limit DNBRs.The DNBR margin available between the Safety Analysis Limit DNBR values and the Design Limit DNBR values is used to offset DNBR penalties.
The analysis of the 30%SGTP conditions uses the Revised Thermal Design Procedure (RTDP)(Reference 2).This methodology gives improved DNBR performance over ITDP by statistically combining the DNB correlation uncertainties with the ITDP uncertainties listed above, i.e., uncertainties in plant operating parameters (vessel coolant flow, core power, coolant temperature, system pressure and effective core flow fraction), nuclear and thermal parameters (F~), fuel fabrication parameters (F~,), THING IV, and transient codes.The uncertainty factor obtained is used to define the Design Limit DNBR which satisfies the DNB design criterion.
The DNB design criterion is that the probability that DNB will not occur on the most limiting fuel rod is at least 95 percent at a 95 percent confidence level during normal operation and operational transients (Condition I events)and during transient conditions arising from faults of moderate frequency (Condition II events).Condition I and II events are defined in ANSI 18.2.As was done with ITDP, the design limit DNBR values are increased to values designated as the Safety Analysis Limit DNBR and the DNBR margin available between these limits is used for flexibility of design and operation of the plant and to offset DNBR penalties such as rod bow.The DNBR limits, current penalties, and margin associated with RTDP analysis are listed in Table 3.12-4.mh1944<w.wp1:1d441195 3.12-3
~tll C Md i An improved THING IV model was used in the DNB analysis of this core.This model is described in Reference 3 and has been approved for use by the NRC.3.12.4.3.2 Design Evaluation DNB Performance The change in design parameters in going from the current analysis to the 30%SGTP conditions included decreasing the power, flow and inlet temperatures as shown in Table 3.12-3.This affects the DNB performance of the core.The DNB methodology was changed from ITDP to RTDP to generate DNBR margin.The DNBR Safety Analysis Limits (Table 3.12-4)were set to keep the DNBR limiting portion of the core limits unchanged.
The associated axial offset limits were recalculated.
The DNB events not protected by core limits that were analyzed were Loss of Flow, Locked Rotor, Static Rod Misalignment, Dynamic Dropped Rod (RCCA), and RCCA Bank Withdrawal from subcritical (Rod Withdrawal from Subcritical).
The results of the analyses showed that the thermal hydraulic design criteria were met for each event.Fuel Tem eratures The limiting values of the fuel average and centerline temperatures will not change due to the 30%SGTP conditions.
3.12.4.4 Conclusions Thermal-hydraulic analyses were made for the fuel for the limiting 30%SGTP parameters using RTDP methodology.
The analysis showed that the DNBR design basis was met for the limiting DNB events.This analysis caused the available DNBR margin to increase.This margin can be used for flexibility of design and to offset unanticipated DNBR penalties.
3.12.4.5 References WCAP-8567-P-A,"Improved Thermal Design Procedure," H.Chelemer, L.H.Boman, D.L.Sharp, February, 1989.2.WCAP-8567-P-A,"Revised Thermal Design Procedure," A.J.Friedland and S.Ray, April 1989.3.WCAP-12330-P,"Improved THING IV Modelling for PWR Core Design," A.J.Friedland and S.Ray, August, 1989.mA1944+w.wpt:1d441195 3.12-4 TABLE 3.12-1 MAXIMUM LOCA AND DBE GRID LOAD RESULTS Case Accumulator DBE SRSS (DBE 8 LOCA)394 75<2000<2040 Grid Load (Ibs)X Z Grid Strength)6080 lbs.<33.6%of grid strength m51944%w.wpf:1d441195 3.12-5 TABLE 3.12-2 FUEL ROD DESIGN ANALYSIS PARAMETERS Parameter Core Power, MWt Minimum System Pressure, psia Maximum Inlet Temperature,'F Thermal Design Flow, gpm Bypass Flow,%FDH Reratin Pro ram 3413 2100 546.2 354,000 4.5 1.55 SGTP Pro ram 3250 2100 543.5 332,800 4.5 1.55 m:11944<w.wpf:1'195 3.12-6 TABLE 3.12-3 DONALD C.COOK NUCLEAR UNIT 1 30%SGTP PROGRAM THERMAL AND HYDRAULIC DESIGN PARAMETERS Reactor Core Heat Output, MWt Reactor Core Heat Output, 10'tu/hr Heat Generator in Fuel,%Pressurizer Pressure, Nominal, psia Radial Power Distribution Design Parameters Current Analysis 30%SGTP Program"'41 3f'1 3250 11,646 11,090 97.4 97.4 2100 2100 1.55[1+0.3(1-P)]
1.55[1+0.3(1-P)]
Limit DNBR for Design Transients Flow Channel DNB Correlation HFP Nominal Conditions Vessel Thermal Design Flow, 10'bm/hr Core Flow Rate, 10'bm/hr Bypass Flow,%Normal Vessel/Core Inlet Temp,'F Vessel Average Temp,'F Core Average Temp, F Vessel Outlet Temp,'F Average Temp Rise in Vessel,'F Average Temp Rise in Core, F Heat Transfer" Average Heat Transfer Area, ft'verage Heat Flux, Btu/hr-ftverage Linear Power, kw/ft"'eak Linear Power for Normal Operation, kw/ft" Temperature at Peak Linear Power for Prevention of Centerline Melt,'F Typical 1.45 WRB-1 133.4 127.4 4.5 546.4 578.7 581.8 611.0 64.6 67.3 52,200 217,400 7.04 16.5'700 Thimble Typical Thimble 1.45 1.40 1.42 WRB-1 125.9 120.3 4.5 543.5 576.3 579.4 609.1 65.6 68.4 52,200 207,000 6.70 15.7 4700 (a)(b)(c)(d)Cook Nuclear Plant Unit 1 is currently licensed to operate at 3250 MWt High inlet temperature bounds the proposed temperature range with respect to DNB Based on nominal 144 inch active fuel length Based on 2.35 F Peaking Factor mA19444W.wpf:1d441195 3.12-7
TABLE 3.12M DONALD C.COOK NUCLEAR PLANT UNIT 1 SGTP PROGRAM RTDP DNBR LIMITS AND MARGIN

SUMMARY

DNB Correlation Cell Type Design Limit Safety Analysis Limit Total DNBR Margin DNBR Penalty-Rod Bow 1'F Temperature Bias Net Remaining DNBR Margin Typical 1.23 1.40 12.1 2.6 1.5 8.0 WRB-1 Thimble 1.42 14.1 2.6 1.5 10.0 m%1 944<w.wpf:1d~1195 3.12-8

4.0 CONCLUSION

S Provided in this document are the results and conclusions of the safety analyses and evaluations to support the implementation of the SGTP Program and the revised Technical Specification changes for Cook Nuclear Plant Unit 1.The safety analyses, evaluations, and supporting documentation provided in this submittal demonstrate acceptable results in each case, incorporating the revised operating conditions associated with the SGTP Program.A brief summary of the results of each analysis and evaluation is provided in the'Summary and~Conclusions" section of this report.mhl9444w.wpf:1d 441195 4-1 (t
APPENDIX A Proposed Technical Specification Changes
~I~~I~~I I~~I~I I~I I~~~I I~I~, I~~~
660 650 2400 gala UNACCQPT OPEAATlOH 640 630 2100paIa P-620 610 600 590 580 570 0 0.2 0.4 0.6 0.8 1 Power (fraction of rated thermal power)1.2 PRESSURE LEihhl 1840 2000 21 00 2250 2400 (0.02, 62086), (0.02, 632.79), (0.02, 63985))(0,02, 64986), (0.02, 65952), BREAKPOINTS (1.136, 586.17), (1.094, 60021), (1.068, 608.72), (1.031, 620.83), (0.996, 632.42), (1 2, 577.94)(1 2, 58652)(12, 591.77)(12, 599.40)(12, 606.63)FIGURE 2.1-1 REACTOR CORE SAFETY LIMITS TABLE 2.2-1 REACTOR TRIP SYSTEÃINSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWASLE'VLUES l.Manual Reactor Trip Noc Applicable Nor, Applicable 2.Pover Range, Neutron Flux 3.Pover Range, Neuczon Flux, High Positive Rate Lov Serpoint-less than or equal to 25\of RATED THERMAL POWER High Setpoinr-less chan or equal to 109\of RATED THERMIC POWER Less than or equal to 5%of RATED THER.QL POWER vich a time conscanc greater than or equal co 2 seconds Lov Setpoint-less than or equal to 26%of RATED THER%0.POWER High Setpoint-less chan or equal to 110%o" RATED THERMAL POWER Less chan or equal to 5.5%of RATED THER'.9J." POWER vith a time constanc greater chan or equal ro 2 seconds 4.Pover Range, Neutron Flux, High Negative Race 5.Intermediace Range, Neutron Flux 6.Source Range, Neutron Flux 7.Overtemperature Delta T Less than or equal to 5%of RATFD THERMO.POWER vith a time constant greacez chan or equal ro 2 seconds Less than or equal to 25%of RATED THERMAL POWER Less than or equal to 10 5 counts per second See Nore 1 Less than or equal to 5.5%of RATED THEB.fAL POWER vith a rime conscant greater chan oz equal to 2 seconds Less than or equal ro 30~of RATED THERMAL POWER'ess chan or equal to 1.3 x 10 councs per second See Note 3 8.Overpover Delta T See Note 2 See Noce 4 9.Pressurizer P essure--Lov l.O.Pressurizer Pressure--High 11.Pressuzirer Water Level--High Greater rhan or equal to 1875 psig Less than or equal to 2385 psig Less than or equal co 92%of instrument span Greater chan or equal ro 1865 psig Less than or equal ro 2395 psig Less than or equal to 93'f f.nstrument span AMENDMENT NO.9i 125 158 12.Loss of Flov Greater than or equal to 90%Greater than or equal co of~ekgn-flov per loop+89.1%of-4askgs.flov pe" loop*85)%75*Se..~n.flov is~468.gpm pez loop.~~~iwuw NERSuCZD COOK NUCLEAR PLANT-UNIT 1 2-5
REACKN TIIII'Y~'I'IJI Itl~"I'l<IIHI N'I'h'I'll>N
'I'III I'l I'I'UIN'I~
N>>'I'A'I'l>N Note 1s Ovurteaperature AT l AT g-K.o i 2$>T.'I I l I l>-'I I I'-T'>K ll~-I~'-l{hl)]I'I wl>uru l A'I'-" lndiCated A I 4L Ith'II.I>llll.lOIAI
~I'L>WI.I.u~Average te>II>orat>>l>>, I'7(.9 u, Indicated'I'L IIA'I'I:I>
'I'III I>I%I.I'>>wl;II (~~I I')4V>J A Preeeur Lxl.'I I>ruu~ut a:~I~:>le p>+~,s>tv S~l~II>dicated II>'5 nu>>>i>>4I ul>U>4ti>>g l>c>: ule ('235 l>sig or 2085 psig)~The function>J>.>>er4I>:>I l>y I.l>U lua>l l4>J col>lrol ler for T dynaaic compensation avg'I>>><<~>>>>I4>>t~>>>>l.:..>>>>thu lead-la>J cuntloller for T t~22 scca vg SUCS e l2=Laplace tra>>stor>>>
op<<rator
n'fhL~I.I." 2,2-j+c~ot nnngi 1:Ct:-IMILJIUI go~A ggo~~r.o~~~
Ol)er~t iun Mich i loupe l h,~4 I K-i)u~lu K.=-u.oui s v 3 I q)exceeds-37 percent, tbe if bII 0.33 percent nf 1tn rains at ,I q)exceeds+k>percent, tbe it bg~percent nf ltn rains at C 2.N'ii)For each percent that the aagnituds oC (q trip sstpoint shall be autoaaklcally reduced aavae THERHhL POMER.(iii)For each percent that the aagnituds og (q trip sstpoint shall be autosatically rsdu5aS ga'rm TllERlQ L POMER.~Inl f (sf)is a function of the indicated dif ference batueen top any botton detectors nf ths er-range nuclear ion chasbers;uith gains to be selected based on naasuryg instrunsnt response during plant startup tests such that:~3 (i)For qw%, between-37 percent and+R.percent, C (hI)p (vhsro q and g are phrCent HA'i'I:.l~
Tll)I, i.lVR in the tOp and bhttOa halVea Ol the COrl rdfspactively, a>>d q i q i" total'tHENlhL pOMER in percent oC a@~TBERCaL POMER).TQ%I a XMI.L 2..~Loco jt judge~S S'Q~SII~~~p g~o~~nca~g~~ii Mote 2: Overpower nT<nT iK-K o i 5<3S S T-K (T-T")-f~nIi i 6 2 where: n T o litic-'ala:8
.>'I't lu'<'>:>>'I'IIVIIHAL PQWEII Average teapc rature,}'ndicated T at iwra>>avg 1.003'I'l VIU4hl I'OWER I i'-~>>l 0~0.0171/1'or increasing average teaperature hand 0 for decreasing average tesperature f c l+S o.uai5 for T>T";K6 0 for T<T" The function generated by the rate lag controller for T dynaaic coapensation LVQ Mot~3$Mote 4!Tfae constant utilized in the rate lag controller for t~l0 secs.3 4VQ Laplace transfora operator f>(nIl~0 The channel's aaxfaua trip point shall not exceed its coaputed trfp pofnC Qy~pre than~percent nT span-~(PEP~RrsPowI8~rv)
The channel'aaxiaua trip point shall not exceed its deputed trip pong gore than~percent nT span.
~++h 2.1 S BASE 4 Loop Operation'estinghouse Fuel (15xl5 OFh)(LR3-1 Correlation)
Typical Cell Thimble Cel].**Correlation Limit Design Limit DNBR Safety Analysis Limit D.':BR 1.17/~23 I.WO~A 1.17 Jo22 A/r+2.A~Reac"or oo lant System""ess'e s~N I A~o~volBR s 0 ess~~~o a":era"e enthalpv a" the'se'.e:(it p loci of points of i.-:"=K~AL PO~ER, x:etage te=perature for which the b e desi n"'.lBR limit, or th is eq al to the en=halpy of saturated o o~represents typical fuel rod*represents fuel rods near g ide tube Cook Nuclear Plant Unit 1 B 2-1(a)Amendment No.7$$J?ZS~
P~g Nsg~Ie jetfve~te Tz'fp pravf des protectfon for control rad drop~id'+: At hf jb paver, a rod drop accident could cause local f1~>+~ay v}6<cauld cause an uncansezvatfve local DNR to exist.Paver Rsn je Ne jatfve Rate Trip vill prevent this tram ace~in j by trfppinj the reactor..No credft fs taken for operation of the Po r Negative Rate Tz f p for'hose contro 1 rod drop ace f dents far~ich the DNERis vfll be greater than the applicable design 1fait DNR value far each fuel type.Intermediate and Source Ran~Nuclear Flux The Intermedi,ate and Source Range, Nuclear Flux trips provide reactor core pro ection during.eactor start"p.These trips provide redundant protection to the 1ov setpoint tr'p of the Pover Range, Neutron Flux channc1s.Thc source Range"hanna's vill in't'ate a zcactor trip at abou" 1C coun"s pcr second, nless manually blocked vhen P-6 becomes=tive.he ntc-.cd'a"c Range Channe's vill initiate a reactor trip at"=cnt.ere'."c"or='ai
=c approximately 25 percent of RATED K"=KQL?-':-R n'css=an ay b'ac~cd-hcn?-'.0 bcco=cs active.4o credit vas=kcn="=operation of=hc trips associated vi=h ci her the Intermediate c"="==Range hannc s n=hc ac=cn ana.yses;hovever, their i notional capabi=r at=hc speci'cd tz'p settings is required bv this s=cci=ication to c-.'".ance
=hc c<<rcra: zci ab'v o=--Reactor~y<<<<p<<p<<~C rcrtc-.pcraturc delta.:"'p prov'Ccs core prate:'on to prevent"NS a'=""bina='ons o: pressure, pover, coolant temps"aturc, and axial-over Cis=".'bu='on.
provided that the transient is s ov vith respect to pi"'ng""a..sit Cciays fzom the core to the temperature detectors (about 4 seconds), and pressure is vithin the range betveen thc High and Lov Prcssure reactor tzips.This setpoint includes correc=iona for changes in dcnsi y and heat capacity of vatez vith temperature and dynamic compensation for i in dela s from the core to the loop temperature detectors.
The refer ce average emperatuze an".e re erence crating p ssure (P')re set eq to the fu povcr ind ated Tavg a the nomi RCS apera ng pressur zespectiv y,:o ensu pro ction of t core limi and to pr erve the a t.=ion tim of the Over perature lta T trip or the rang of full po'cz average tempera ures assume in t'h e sa t anal ses t nor=a ax a povcz 4'stzibution, this reactor tzip imit, is alvays belov=he core safety limit as shovn in Figure 2.1-1.If axial peaks are greater than design, as indicated by the difference betveen top and bottom-over range nuclear detectors, the reactor trip is automatically reduced according to the notations in Table 2.2-1.DnE-rZ COOK NUC~~PLANT UNIT 1 B 2-4~MZNT No, 74, 126
~r ta The overpover delta T reactor trip provides assurance of fuel integrity,~.g..no melting, under all possible overpover conditions, liaits the required range for Overtemperature delta T protection, and provides a backup to the High Heutron flux trip.The setpoint includea corrections for changes in density and heat capacity of vater vith temperature, and dynastic compensation for i in dela s froa the core to the loop temperature detectors~The referen average t perature ()is se'c e to e 1 over indica d Tavg to e sure fuel egrity dur g overpova conditions or PRi range of 1 pover ave age tempera es assume in'he sa ty ana]sis, e overpover e te reactor tr p p'rov as protect on or c~up protection for at pover steamline break events.Credit vaa taken for operation of this trip N the steam line break mass/energy releases outside containment ana]ysis.1n addition, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the reactor protection system.Pressuriser tre u e The tressuriser High and Los Pressure trips are provided to limit the pressure range in vhich reactor operation is permitted.
The High Pressure trip is backed up by the pressuriser code safety valves for LCS overpressure protection, and is therefore set lover than the set pressure for these valves{2485 psig), The High Pressure trip provides protection for a Loss of Kxternal Load event.The Lov Pressure trip provides protection by tripping the reactor in the event of a loss of reactor coolant pressure.Pzessuriser Uater Leva The Pressuriser High Pater Level trip ensures protection against Reactor Coolant System overpressurization by limiting the vater level to a volume sufficient to retain a steam bubble end prevent vater relief through the pressuriser safety valves.The pressuriser high vater level trip precludes vater relief for the Uncontrolled RCCA Withdrsval at Pover event, COOK NUCLEAR P EST UNZT 1~2-5 AMZgu~HO.XW.LN, 15
'll 3 4.1 REACTIVTTY CONTROL SYR~3 4.l.1 ADORATION CONTROL SHUTDOQH MARCIN-TAUC CREATER THAN 200 F LIMITINC CONDITION POR OPERATION/.3 3.1.1.1 The SHUTDOMN MARCIN shall be greater chan or equal to~Delta k/k, APPLICASILITY:
NODES l.2+, 3, and 4.ACTION: 1.3 Uich the SHUTDOMH MARCIH less chan~~lDelta k/k,%mediately tnitiace and concinue boration at greater than or equal to 10 gpss ot a solution containing greater than or equal to 20.000 ppa boroa or equivalent uncil the required SMUTDORt MARCIH is reacored.SURVEILIANCE UIREMENTS 4.1.1.1.1 Ths SIICTOOIOI IIASCIN shall be datartltsd ta be dresser that sit steal ss+H\Delta h/h: ae b.C~Vithin one hour after detectioa of an inoperable control rod(a)and ac least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter vhile the rod($)ia inoperable.
If the inoperable concrol rod ts hearable or untrippablet the above required SttUTDOQN MARCIH shall be verified acceptable vith an increased allovance for the vichdzavn vorth ot che iaaovable or untrippable control rod(a).Shen ia NODE 1 or MODE 2 vith jeff greater than or equal to 1.0, ac j least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that control bank vithdraval is vithin the liaits of Specification 3.1.3.5.%hen in MODE 2 vich Xeff less than 1.0, vithia 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criticality by verifying that the predicted critical control rod position ia vichin the liatta of Specification 3.1.3.5.d.Prior to initial operation above 50 lATED SHOAL NMEk after each fuel loading, by consideracion of the factors of e belov, vith the coacrol banks at the aaxitua insertion liait of Specificacioa 3.1.3.5.+See Special Toss iizceptioa 3 e 104 1 e COOK HUCLLC PLAHT~UNIT 1 3/4 1 1 AMEHDmrr NO.7k.128.148
REACTIVITY CONTROL SYS~CHARCINC PUMP-SHUTDO'4N LLNTINC CONDITION FOP.OPTATION 3.1.2.3 a.One charging pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency bus.b.One chazging flowpach associated vith support of Unit 2 shutdown functions shall be available.
>>APPLICABILITY:
Specification 3.1.2.3.a.
-MODES 5 and 6 Specification 3.1.2.3.b.
-At all times vhen Unit 2 is in MODES 1, 2, 3, or 4.ACTION;a.Mi.th no charging pump OPERABLE, suspend all operations involving CORE ALTERATIONS or posi.tive reactivity changes.>>>>
b.With more than one charging pump OPERABLE or vith a safecy injection pump(s)OPERABLE vhen the temperature of any RCS cold leg is less that or equal co 152 F, unless che reaccor vessel head is removed, remove 0 the additional charging pump(s)and the safety injection pump(s)motor circuit breakers izom the electrical povez circuit within one houz.c.The provisions of Specification 3.0.3 are not applicable.
d.In addition to the above, vhen Specif'cation 3.1.2.3.b is applicable and che required flov path is not available, return the required flow path to available status within 7 days, or provide equivalent shutdown capability in Unit 2 and return the required flov path to available status vithin the nexc 60 days, or have Unit 2 in HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and HOT SHUTDOWN vithin the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .e.The requiremencs of Specification 3.0.4 are not applicable when Specification 3.1.2.3.b applies.SURVEILLANCE RE UIMHENTS p>rrsg~'~<
4.1.2.3.1 The above zequized charging pump shall be demo trated OPERABLE by verifying, thac on recirculation flov, che pump develops a~hasp+pressure of greater than or equal to~ps+when tested pursuant to Specification 4.0.5.gZ,Vo+A maximum of one centrifugal charging pump shall be OPERABLE vhcnever the temperature of one or more of che RCS cold legs is less than or equal to 152 F.>>>>Foz puzposes of this specification, addition of vatez from the RUST does noc consticuce a positive reactivity addition provided the boron concentration in the RUST is.greater than the minimum required by Specification 3.1.2.7.b.2.
COOK NUCLFML PLANT-UNIT 1 3/4 1-11 AMENDMENT NO.9'f, f/', 167, 3.1.2.4 At least tvo charging pumps shall be OPERABLE.MODES 1, 2, 3 and 4.hEZQH: With only one charging pump OPERABLE, restore at least tvo charging pumps to OPERABLE status vithin 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBT vithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ;restore at least tvo charging pumps to OPERABLE status vithin the next 4S hours or be in COLD SHUTDOWS vithin the follovtng 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .piggy QEA/YNK 4.1.2.4 At least tvo charging pumps shall be de onstrated OPERABLE by verifying, that on recirculation flov, each pump develops agBaeherge-pressure of greater than or equal to~paid vhen tested pursuant to Specification 4.0.5.~~No COOK NUCLEAR PIhHT-UNIT 1 3/4 1-12 AMENDMENT NO.04, 164
[Rr>CTIVITY CONTROL SYST=MS l!BORATEO WATER SOURCES S~UT"OWN;>LIMIT!NG CONDITION FOR".o=~A ION-;'I'.3.1.2.7 As a minimum, one of the following borated water sources shall be.-'>OPER."BLE:
a.A boric acid s:orage system and associated heat tracing with: l.A minimum usab'.e borated water volume of 4300 gallons, 2.Between 20,0:0 and 22,500 ppm of boron, and 3.A minimum solution temperature of 145'F.b.The refueling water storage tank with: l.A minimum usable borated water volume of 90,000 gallons, 2.A minimum boron concentration of 2400 ppm, and 3.A minimu.-., solution temperature of~F.O APPLICABILITY MOO'S=an" 6.ACTION:>With no borated water so r"e OPERABLE, suspend a>', operations involving CORE i" ALTERATIONS or positive~eactivity changes until at least one borated water resource is restored o GP=.=.'BLE status.i>>'URVEILLANCE RE UIRE'".ENTS (s j>4.1.2.7 The above required borated water source shall be demonstrated
>OPERABLE: a.At least once per 7 days by: 2.3.Verifying the boron concentration of the~ater, Verifying the water level volume of the tank, and Verifying the boric acid storage tank solution temperature when it is tne source of borated water.b.At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when it is tne source, of borgted water.I>II Ii For purposes o.tni s speci.ica.ion, addi tion of water from the RWST does!lnot cons itute a positive reactivity addition provided the boron concentra-
>;tion in the RMST is greater tnan the minimum required by Specification ii>I>Tb>.Q.C.COOK-UNIT 1 3/4 1-15 Amendment No.$2, Ill
", RE'C IVITY CONTROL SYSTEMS!I" BORATED WATER SOURCES-OPERA.fONS i: LI.'1.'T:NG CONDITION FOR OPEPAT!ON I:: 3.l.2.8~I I'~~i~I I~~I I I~~I~Each of the following borated water sources shall be OPERABLE: a.A boric acid storage system and associated heat tracing with: l.A minimum usable borated water volume of 5650 gallons, Between 20I000 and 22,500 ppm of boron, and~\3.A minimum solution temperature of 145'F.b.The refueling water storage tank with:~g~~I I~~l.A miniŽu.".contained volume of 350,000 gallons of water, 2.3.'I APPL: ".<<9 IL ITY:<l f~AC~f QaI BetweenŽ3 and 2600 ppm of boron, and A min;--s"1Žion temperature of g&F.70 MODES I,", 3 and 4.With the d'or'.c acid storage system inoperable, restore the storage system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTOOWN MARGIN equivalent to at least 1'k/k at 200'F;restore the boric acid storage system to OPERABLE s atus within the nex.7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .'I~~I~~With the refue'iing water storage tank inoperable, restore the tank to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT-DOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .~.'UR'I'5 i LLANCE"4.1.2.8 Each boratec~a:er s=urc shall be demons rated OPERABLE:~~3.C.COOK-UNIT 1 3/4 1-16 Amendment No.gg,-I TABLE 3.2-1 DNB PARAMETERS LENITS 4 Loops in Operation at RATED THEBKWL POWER Reactor Coolant System Tavg Pressurirer Pressure 4 (+7/'S+5'./)(g8~j)QL]/if Qggog~pf epsc REsp~asuiTvg
>2050 psig**Reactor Coolant System Total Flov Rate/CD/)/dg>~~gpm*Indicated average of at least three OPERABLE instrument loops.Limit not applicable during either a THERMAL POWER ramp increase in exce of 5 percent RATED THERMAL POWER per minute or a THELRQ.POWER step increase in excess of 10 percent RATED THFKM.POWER.Indicated value.COOK NUCLEAR PLANT-UNIT 1 3/4 2-14
TASKS 3.3iI RZACTOR TRIP SYST5t INSTNlMBtTATIOÃ RESPONSt TIMES l.Manual Reactor Trip 2.Pover Range,'eutron Flux (High and Lov Secpoinc)NOT AP?LICASLE Less chan or equal co 0.$seconds+3.Pover Range, Neutron Flux, High Positive Rate 4.Pover Range, Neutron Flux, High Negacive Rate 5.Incermediace Range, Neutron Flux 6.Source Range, Neutron Flux 7.Overtemperature delta T 8.Overpover delta T s than or equal to 0.5 seconds'OT APPLICASLK NOT AP?LICASLE Less than or equal to 6.0 seconds*Less chan or equal to 6.0 seconds+9.Pressurirer Pressure--Lov Less than or equal to seconds IO.Pressurizer Pressure--High Less chan or equal to 2,0~seconds II.Pressurizer Mater Level-High Less than or equal to 2.0 seconds Neutron dececcors are exempc from response time testing.Response rime of che neutron flux signal portion of the channel shall be measured from detector output, or input of firsc electronic component in channel.COOK NUCDhR PlhFT NIT l 3/4 3 10 AMEHDMENT NO.98.L1;sa TABLE 3.3-2 Continued REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES P>CTIONAL UNIT 12.Loss of Flov-Single Loop (Above P-8)RESPONSE TIME Less than or equal to 1.0 seconds 13.Loss of Flov~Tvo Loops (Above P-7 and belov P~8)Less than or equal to 1.0 seconds 14.Steam Generator Vater Level~-Lov-Lov 15.Steam/Feedvater Flo~Mismatch and Los Steam Generator Mater LeveL 16.Undervoltage-Reactor Coolant Pumps 17.~Žderfrequency-Reactor Coolant Pumps Less than or equal to 2..0++seconds NOT APPLICABLE Less than or equal to I<~seconds Less than or equal to 0.6 seconds 13.Turbine Trip A.Lou Fluid Oil, Pressure B.Turbine Stop Valve'.9.5afe=y Injection Input from ESF 20.Reactor Coolant Pump Breaker Position Trip NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE COOK NUCLEAR PLANT-UNIT 1 3/4 3-11 hMKHDMENT NO.ZlS, 15S TABLE 3.3-3 Continued ENCINEERED SAFETY FEATURE ACTUATION SYSTEN INSTRPAENTATION FUNCTIONAL UNIT Dc AF7c.f.Steam Flow in Tv Steam Lines-Hi.gh KININJM TOTAL NO.CHANNELS CHANNELS APPLICABLE OF CHANNELS Four Loops Operacing'2/steam l.ine 1/s am line y 2 steam 1 ne5 1/steam line ree Loops 0 rating COINCIDENT ITH EITHER 2/op rating steam ine l~/any 1/operating 3'peracing steam line steam line T~~Lov~Lov avg Four Loops Operating 1 T/loop avg T any avg 1T any 1, 2, 3~14*3 loomis Three Loops Operati.ng OR, COINCID VITH T avd o crating loop in any operatin loop 1T in.3 any c5o operacing loops 1.5 Steam Line Pressure-Lov Four Loops Operacing Three Loops Operating 1 pressure/loop 1 pressure/operating loop 2 pre55ures any loops 1~pre555ure in any operating loop 1 pressure any 3 loops 1 pressure in any 2 operati.ng loops 1, 2, 3'4".OOK NUCLEAR PLhNT-UNIT 1 3/4 3-17 AHEIHENT NO.Pf, f28 153
TABLE 3.3-3 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEH INSTRUMEÃI'ATION FUNCTIONAL UNIT COINCIDE'C MITH~HBR-TOTAL NO.OF CHANNELS HINIMUH CHANNELS CHANNELS APPLICABLE TD TRIP OPERABLE MODES ACTION e e Loge Log avg Four Loops Operating Three Loops Operating 1 T/loop avg opera8fng loop 2 T any 75ops 1~T in any operating loop 1 T any'P loops 1 T in an)c o operating loops 1, 2, 3'5 Steam Line Pressure-Lov Four Loops Operating Three Loops Operacing 5.TURBINE TRIP&FEEDWATER ISOLATION 1 pressure/loop 1 pressure/operating loop 2 pressures any loops lee>>pressure in any operating loop 1 pressure 1, 2, 3e 14 any 3 loops 1 pressure 3~in any 2 operacing'oops a.Steam Generator Mater Level--High-High 3/loop 2/loop in 2/loop in 1,2,3 any oper-each oper-ating loop ating loop COOK NUCLEAR PLANT-UNIT 1 3/4 3-21 AHENDMENT NO.gf, f26, 153
ENCZNEERED SAPETT FEATURES IN?ERLOCXS DES IC NATION F-11 P-12 CONDITION AND SETPOINT Vith 2 of 3 pressuriser pressure channels greater than or equal to 1915 psig.Pith 2 of 4 T channels ave less than or equal to Setpoint.Setpoint greater than or equal to 541 F FUNCTION P-11 prevents or defeats manual block of safety in)ection actuation on loT pressurizer ptessure.P-12 allovs the manual block of safety in]ection gau87iou owl lov steam line pressure.~~CAccfc:S ST~g AzMZ ISdMD4 oN HI4H Stcam.FC~,~~facts steam dump blocks.Pith 3 of 4 T channels avg above the reset WQev-Polw, P/ZPFPV/pg pgA~I yh'g rNfAJHhc 8LdCC PF Srtr=BTy j~J6G770~Ac 7&t97ppw o~w~~EA~r~Z PR~ss~g~COOK NUCLEAL HAH'UHIT 1 3/4 3-23a ammmrr NO.153
TASLE 3.3-4 EHCINEERED SAFETY FEATURE ACTUATION SYSTEH INSTRUHENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOVASLE VALUES 1.SAFETY INJECTION, TURBINE TRIP, FEEDVATER ISOLATION, AHD HOTOR DRIVEN FEEDVATER PIPS a.Hanual Zniti.ation See Functional Qnfc 9 b.Automatic Actuation Logic c.Contaizuaent Preaaure--
High d.Preaauriser Pressure--
Lcnr Hot Applicable Less than or equal to 1.1 paig Creater than or equal to ldl5 psig Hot Applicable Less than or equal to 1.2 paig Creater than or equal to Id05 paig~.Differential Presaur~hetveen Steam Lines-High f.eaa Plov tn Steam a--Hi.gh Co ci.dent so+a r Steam Line Preasure--
Loe Less than or equal to 100 psi aa than6or qual to 1.x 10 lbs from Ot d to 20%1 d.Linea from 1.42 x 10 lba a)20%1 d to 3.1$z 0 lbs/hr 00%load Less than or equaL to 112 psi aa than6or qual to 1.6 x 10 lbs from 0%d to 20%, ad.Lipa from 1.56 10 Ib a)20'd to 3.93 10 lbs/hr t 100%load T greacer c oc'o 541 P T greacer oc'to 539 F Q g'eater chan or equal (+eater than or equal to to 500 yaig steam line 480 p+g steam line pressure pressure COOK NUCLEAR KLFZ UHZT 1 3/4 3-24 AHEIHEHT HO.49, 128 153 TABLE 3.3-4 Continued ENCINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT 2.Containment Radio-activityy--High Train A (VRS-1101, ERS 1301, ERS-1305)3.Contatnmeat Radio-activity--High Train B (VRS-1201, ERS-1401, ERS-1405)4.STEAM LINE ISOIhTION TRIP SETPOINT See Table 3.3-6 See Table 3.3-6 ALMVABLE VALUES Hot Applicable Not Applicable a.Manual b.Automatic Actuattoa Logic--------See ruaeaaoaaX UnSe 9------...Not Applicable Hot Appltcable c.Containment Pressure--
High High d.Steam Flov tn Tvo Steam Linea--Htgh Cotncident vtth T--Lov-Lcnr
~Less than or equal to 2.9 patg Less than6or equal to 1.42 z 10 lba/hr from 0%load to 20%load.Linear from 1.42 x 10 lbs/hr a(20%load to 3.4B s 10 lba/hr at 100%load.Less than or equal to 3 patg Leaa than6or equal to 1.56 x 10 lbs/hr from 0%load to 20%load.LiIIear from 1.56 x 10 lbs/hr6at 20%load to 3.93 10 lba/hr at 100%load.Q, 5ig44 4.inc, MN54&~-Lusa)5.TURBZ?R TRIP ASD FEEDVATER ZSOLATZOX T greater thea or a@i to 54loF Creater than or equal to 500 pstg steam line pressure T greater than or~qQ1 to 539 F Creater than or equal to 4BO paig steam ltae pressure a.Staaa Cenerator Qatar Level,--High-High Less than or equal to 67%of narra'-range instrument span each steam generator Less than or equal to 6$%of nazrcw-range instrument spaa each steam generator COOK NUCLEAR PALS?-UNIT 1 3/4 3-26 AHKHDMEST EO.94, Qg, 153 ThhLE 3.3 5 BfC NEELED SAFETY FEATURES kESPONSE TIMES INITIATINC SXCNAL AND FUNCTION RESPONSE TIME IN SECONDS 1.Manual ao Safety In)ection (ECCS)Feedvater Isolation Reactor Trip (SI)Containment Isolation-Phase
'A Containment Purge and Exhaust Isolation Auxiliary Feedvater Puips Essential Service Water System Not, Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable b.Containment Spray Containment Isolation-Phase 5 Containmenc Purge and Exhaust Isolation Containment hir Recirculation Fan c.Containment Isolation-Phase
'h'ontainment Purge and Exhaust Isolation d.Steam Line Isolation 2.Containment Pressure-Hi h a.Safety In]ection (ECCS)b.Reactor Trip (froa SI)c.Feedvater Isolation d.Containment Isolation-Phase h'.Containment Purge and Exhaust Isolation f.Auxiliary Fee@Water Pumps g.Essential Service Water System Not Applicable Not Applicable Not Applicable Not Applicable Not Ap'placable Not Applicable Not Applicable
+7,0 Less than or equal 27.OQQ/~M Less than or equal Less than or equal Less than or equal 18.04/28.0H Not Applicable Not Applicable Less than or equal 13.0f/48.0H to to 3,0 to 8.0 to'to COOK NUCLEAR PLANT-UNIT 1 3/4 3-27 AMnmENT gag, 158 TABLE 3.3 5 Continu EÃCZNEERED SAFETY FEATURES RE5PONSE TZMZ5 IHITZATINO 5ZCNAL AHD FUNCTION RE5POHSE TINE IN SECONDS 3.pressurizer Pressure-t,ov a.Safety In)ection{ECCS)b.C~d.~o Reactor Trip (from SI)Feedvater Zsolation Containment Isolation Phase A~Containment Purge and Exhaust Zsolation Auxiliary Feedwater Pumps Essential Service Mater System y7.0 Less than or equal to 21~000/9+vO++
Less than or equal to 3.0 Z,ess than or equal to 8.0 I.ess than or equal to 18.0P Not Applicable Not Applicable I,ess ghan or equal to 48.0/13.00 4.D!.fferential Pressure Between Steam Lines-Hi h a.Safety Zn)ection{ECCS)b.Reactor Trip (from SI)c.Feedvater Isolation d.Containment Isolation-Phase"A" e.Containment Purge and Exhaust Isolation f.Auxiliary Feedvater Pumps g.Essential Service Mater System S7,~I.ess than or equal 27.088/~~Less than or equal Less than or equal Less than or equal 18.08/28.Of I Not Applicable Not Applicable Less than or equal 13.0f/48.0tt to to 3.0 to 8.0 to to 5.Steam Flow in Two Steam Lines-Hi h Coincident wit.".Tav Low Lov gd7 R?Pl cd 8/-r a.Safety In)ection (ECCS)b.C~d.Reactor Trip (from SI)Feedwater Isolation Containment Isolation-Phase A" e~f.go Containment Purge ar'xhaust Isolation Auxiliary Feedvater Pumps Essential SeFfice Mater Sys'em Steam I,ine Isolation Not Applica'".'~
Not Applicable HWP Less than or equal to 13." COOK NUCLEAR PLANT UNIT 1 3/4 3 28 AMENDMXNT NO N e XI' TASL~3.3'Continued EHCINE~SAFETY FM~iS RESPONSE TLKS INITIATINC SICNAL AND FUNCTIQN 6.Steam Line Pressuz'Lov RESPONSE TO%IN SECQNDS a.Safety Ia)ection (ECCS)b.Reactor Trip (fzom SI)c.Feedvater Isolatioa d.Containmeac Isolation-Phase
'A e.Containment Pur$e and Exhaust Isolation f.Auxiliary Feedvacer Pumps$.Essencial Service Qatar System h.Steam Line Isolation 7.Con aizaent P essu e--High-High Lagg chan or equal 27.~>.%~Legs than or equal Lagg than or equal Legs thaa or equal 1$,0e/2$.0>>Not Applicable Noc Applicable Lass thaa or equal 14.0e/48.0>>
Legs thaa or equal to to 3.0 to$.0 Co Co to 11.0 a~b.C.Concainmenc Spray Containment Isolation-Phase b Steam Line Isolation Containment Air Recirculation Fan Less chan or equal Not Applicable Lass chan or equal Less than oz equal g5.o to~to 10.0 o 600~0 S.S earn Cenerator"ater Level--Hi h-Hi h a.Turbine Tr'p b.Feedvater Isolatioa 9.Steam Cenerator Pacer Level-Lov-Lov a.Motor Driven Auxiliary Feedvater Pu"ps b.Turbine Driven Auxiliary Feedvacer P"~s 10.4160 vole Emer ercv$us Loss of Vol age Lass than or equal to 2.5 Lass than or equal to 11.0 a/go.o~~Less than oz equal co 60.0Lass than or equal to 50.0~a.Aotor Driven Auxiliary Feedvatar Mps 11.Loss of~gin Feedvacer MŽos Less than or equal co 60.0a.Motor Driven Auxiliary Feedvater Pumps Less than or equal to 60.0>12.Reactor Coolant Pm~Sus Undcrvolta
~a.Turbine Driven Auxiliary Feedvacar Pumps Lass than or equal to 60.0>CQQK NUC~~PLANT UNIT 1'3/4 3 29 AMZNDKBtT NQ.49 t 229.168 TASLE 3~3-5 Continued TA31Z NOTlTXON e Diesel generator starting and sequence loading delays nos inclu4ed.Otfsite paver available.
Response time limit includea opening of valves to establish SZ path and attainment of discharge pseaaure for centrifugal charging pumps.~Diesel generator starting and sequence loading delays included.Response time limit includes opening of valves to establish SI path and~astainmenc of discharge pressure for centrifugal charging pumps.++Diesel generator starting and sequence loading delays included.Response time limit includea opening of valves to establish SZ path and attainment
'f discharge pressure for centrifugal charging, SZ.and MR pumps.Sequential transfer of charging pump auction from the VCT to the RVST (RVST valves open.then VCT valves close)is NOT included.Diesel generator starting and sequence loading delays included.Response sine li&t includea opening of valves to establish SZ path and attainment of discharge pressure for censrifugal charging pumps.Sequential transfer of charging pump sucsion from the VCT to the RUST (EST valves open, then VCT valves close)ia include4.98 Diesel generator starting and sequence loading delays NOT included.Offsite pover available.
Response time limit includea opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps.Sequensial transfer of charging pump auction from the VCT so the RVST (RVST valves open, then VCT valves close)is included.REsP((<r TINFs u$cO F'o4 ofFEiTE Scowl(s'4afgcF A'%vs(5.~~(P(=~~/~C'~~&V.g(Qg(=('(=~(=+~~+5'7~+((mg+pLIO S(=+~su~c.uR (Cy~+~~/+5+((+0M (/9 T+@Sparse 7(4(=(ea vp(=S eP'En(ec oF v'+<egg (a c+((yi~pggrv(o(=Q<acpp)QZ PR~~4c oW/~~M.CCOK HUC~~PLhBT-UHZT 1 3/4 3-30 hHENDMEPZ NO~7'li8 TAbLE 4.3-2 ENCINEHtED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEIILANCE RE UZKDKNTS FUNCTIONAL UNIT 1.SAFEST INJECTZOH, TURSZNE TRIP s PEEDVATER ZSOIATIOH, ASD MOTOR DRZVEH AUXZIZhRY FEEDVATER PUMPS TRIP ACTUATING CHANNEL DEVICE C~lHEL CHANNEL TUHCTIONAL OPERATIONAL CHECK CBETBEATTOE
'TEST TEST MODES IN MHZCH SURVEILLANCE RE UIRED a.Menuel Znitfstfon e e o o s e e e e o e e o e e s o s o See Punct f one 1 Unf t 9 e e e o e s e e e o e e e s e e e o b.iutomstfc Actuetfoa Logic S.A.S.A.M(2)H.A.12,3,4 c.Containment Press-ure-High d.Pressurfser Press-ureeiAK e.DDferentfs1 Press ure betveea Steam Lines--High M(3)H.A.H.A.H.A.H'A.l.2, 3 1,2,3 1, 2.3 1, 2.3 PressureeeLolf (57rg&LBa/2 2.COHTAZNMEHT SPRAT a.Mam+1 Initiation s o o s o o o o e os s so o o os o See Punct f oae1 Unit 9 e s o e o o o e o e e o o o o s e e o b.hutomstfc Actuation Logic H.A.H.A.M(2)H.A.1,2.3B4~c.Containment Press-ureeHf gheHfgh M(3)H.A.1, 2, 3 CQOX NUCLEAR PIAST UHZT 1 3/4 3e31 Sl!EEOI!EET EO.fII,/It, TABLE 4.3-2 Continued ENCINEERED SAFETY FEA'TURED ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UZREMENTS FUNCTIONAL UNIT T ZP ACTUATINtH QRNNEL DEVICE CHANNEL QGQiNEL FUNCTIONAL OPHbLTIONAL CHECK CBIIBEBTICH TEST TEST MODES ZN RiICH SURVEILLANCE 4~SThB LINE ISOLATION a.Manual---See Punctional Unit 9--------b.Automatic Actuation Logic c.Containment Preaa-ur~--High-High d.Steam Flov in Tvo Steam Linea--High Coincident vith Tavg~~Lov~Lov N.A.N.A.M(2)M(3)N.A.H.A.H.A.1,2,3 1, 2, 3 1,2.3 Serge uuB=Wg E":SSqee-~S~~5.TURBINE TRIP AND FEEDVATER ISOLATION a.Steam Cenerator Vater Level--High-High 6.MOTOR DRIVEN AUXZLIART FEEDVATER PUMPS a.Steam Cenerator Qatar Level-Lov Lov H.A.1.A.IH2>Z l.2, 3 1, 2.3 b.4 kr Eua Loaa of Voltage c.Safety Injection d.Loaa of Main Peed Pumpa H.A.1.A.1.A.H.A.M(2)N.A.H.A.H.A.1,2.3 1,2,3 1, 2 COOK NUCLEAR PLANT-UN'3/4 3-33 JHEHBHEHT HO.TI, g], (ET 3.4.2 h minimum of one pressurirer code safety valve shall be OPERABLE with a lift setting of 2485 PSIC<<+t.*h.3 MODES 4 and 5, MEZQE: Pith no pressuri.ter 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 Infection pumps and all but one charging pump inoperable by removing the applicable motor circuit breakers from the electric power circuit within one hour.4.4.2 The pressurizer code safety valve shall be demonstrated OPERABLE per Surveillance Requirement 4.4.3.~e 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 RUST 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.S.b.2 (MODE 4)or 3.1.2.7.b.2 (MODE 5).D.C.COOK-3hGT'1 3/4 4 4 AMENDMENT NOSE~1; 3.4.3 All pressuriser code safety valves shall be OPERASLX rith a lift settinI, of 24IS PSZC gm.TChEIJXt: hQXZQI: With one pressuriaer code safety valve inoperable, either restor~the inoperable valve to OPERASLE status vithin 15 minutes or be in HOT SHUTDOWN vithin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .4.4.3 No additional surveillance requirements other than those required by Specification 4.0.S.'The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.COOI IUCLZhk PLhHT UHIT 1 3/4 4ej AMEHDHEIT 50.440,
it least once per 11 months by: 1.Verifying automatic isolation and interlock action of the RHR system from the Reactor Coolant System shen the Reactor Coolant System pressure is above i00 paig.2.L vfaual inspection of the containment sump and verifying that the subsystem auction inlets are not restricted by debrfa and that the sump components (trash racks, screens, etc.)ahov no evidence of structural distress or abnormal corroa f on.e.ht least once per 1$months, during shutdown, by: 1.Verifying that each automatic valve in the flov path actuates to ita correct position on~Safety Injection teat signal.2.Verifying that each of the folloving pumps atatt automatically upon receipt of a safety injection signal: a)Centrifugal charging pump b)Safety injection pump c)Residual heat removal pump g>FFggwphc.f.by verifying that e of the follcnrfng pumps develops the indicated 4kool~zpresaure on zecizculation flov vhen tested puzsuant to Specification 4.0.5.g2 PO 1.Centrifugal charging pump greater than or equal to4405 paf~13k&2.Safety injection pump greater than or equal to+40%paig-l5O 3.4aidual heat removal pump greater than or equal to 404 pafg-g.by verffyfng the correct position of each mechanfcal atop for the follovfng Emergency Coze Cooling System throttle valves: 1.within 4 houzs following completion of each valve stroking operation or maintenance on the valve vhen the ECCS sub-systems are required to be OPSNLIl.COOK NUCLEAR HAUNT UNIT 1 3/4 5-5 AHESDMEHT 80.44%, 404, 444, 444~g{
EMERGENCY CORE COOLING SYST.".'AS REFUELING MATER STOPPAGE."A'K LIMITING CONOI".ION;
~R OPE-"..-"":3'<
3.5.5 The refueling wa:e.storage tank (RMST)shall be OPERABLE with: A minimum contained volume of 350,.000 gallons of borated water.b.Between 2400 and 2600"pm of boron, and c.A minimum water temperature of~F.7'0 APPLICABILI; Y: NODES 1, 2, 3 anC 4.ACTION: With the refueling w=e~storage tank inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ".r"e in a least HOT STANOBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLO SHUTDOWN within the;":.Nin" 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ., SURVEILLANCE RE UIRE.".E".'l.5 The RWST shall be de:-.onstrated OPE?wBL!!4.5.5 II 2.Verifying tne boron concentration of the water.b.At least once per 2'ours by verifying the RMST temperature.
a.At least once per 7 days by: l.Verifying-.ne contained borated water volume in the tank, and 0.C.COOK-UNIT 1;,'-" 5-11 Amendment No.Q,lll ELECTRICAL POWER SYSTEMS SURVEILLANCE RE UIREMENTS Continued 4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE: in accordance with the frequency specified in Table 4.8-1 on a STAGGERED TEST BASIS by:~lEVM I H i 30 M~OSi udge.P4t I9%'6D~%26 (l>LyZ 4+0 RfEQC&cy Rr 44<<r.2da.b.C.1.2.3.Verifying the fuel level in the day tank, 57W~OQ y Verifying the fuel level in the fuel storage tank Verifying that the fuel transfer pump can be st ted and that it transfers fuel from the storage system to t e day tank, Verifying that the diesel starts from~~4condition and 5.Verifying that the generator is loaded to greater than or equal to 1750 kw and that it operates for greater than or equal to 60 minutes and verifying that the generator output breaker to the emergency bus is OPERABLE, and 6..Verifying that the diesel generator is aligned to provide standby power to the associated emergency busses.By removing accumulated water"~: 1)From the day tank at least once per 31.days and after each occasion when the diesel is operated for greater than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and 2)From the storage tanks at least once per 31 days.By sampling new fuel oil"~in accordance with the applicable guidelines of ASTM 04057-81 prior to adding new fuel to the storage tanks and 1)By verifying, in accordance with the tests specified in ASTM 0975-81 and prior to adding the new fuel to the storage tanks, that the sample has: he d>es genera r start 1 seco pe ormed a least o e per 1 days engin starts r the rpose o this action be at educed celera n that m h i 1 t s and ar on t s rom mb ent nd t>o sha these s veillan tests.All ot r su eillanc testing nd comp satory rate s reco ended b the man acture o diese n ine a minim ed.*~The actions to be taken should any of the properties be found outside of specified limits are defined in the Bases.D.C.COOK-UNIT 1 3/4 8-3 AMENDMENT NO.125 ELECTPICAL POWER SYSTEMS SURVEILLANCE RE UIREMEMTS Continued 2.Verifying the generator capability to reject a load greater" than or equal to 600 kw while maintaining voltage at 4160+420 volts and frequency at 60+1.2 Hz, 3.Verifying the generator capability to reject a load of 3500 kw without exceeding 75X of the difference between nominal speed and the overspeed trip setpoint, 4.Simulating a loss of offsite power by itself, and: a)Verifying de-energization of the emergency busses and load shedding from the emergency busses, b)Verifying that the.diesel starts on the auto-start signal, energizes the emergency busses witli permanently connected loads within seconds, energizes the auto-connected shutdown loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads.After load sequencing is completed, the steady state voltage and frequency of the emergency busses shall be maintained at 4160.+420 volts and 60+1.2 Hz during.the test.5.Ve~ifying that, on a Safety In)ection actuation test signal (without loss of offsite power), the diesel generator starts on the auto-start signal and operates on standby for greater than or equal to 5 minutes, 6.Simulating a loss of offsite power in con)unction with a Safety Injection actuation test signal, and by: a)Verifying de-ener gization of the emergency busses and load shedding from the emergency busses, Verifying the diesel starts on th>>auto-start signal, ene izes the emergency busses with permanently connected oa s withinseconds, energizes the auto-connected emergency (accident) loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads.After load sequencing is completed, the steady state voltage and frequency of the emergency busses shall be 4160+420 volts and 60+1.2 Hz.The voltage and frequency shalT be maintained within these limits for the remainder of this test, and D.C.COOK-UNIT 1 3/4 8-5 NENOHENT NO.125 ELECTRICAL POWER SYSTEMS SURVEILLANCE RE UIREMENTS Continued b)Agitate the fuel oil in the storage tank awhile pumping the oil from the bottom of the tank through a 5-micron filter, and back to the opposite end of the tank.Three successive samples shall be taken and analyzed according to ASTM 02276-83.If the contaminant level in any of the samples is greater than 10 mg per liter, the agitation, filtration, and sampling processes shall be repeated.If the contaminant level remains above 10 mg per liter after 3 iterations, the draining and cleaning method described in surveillance requirement 4.8.1.1.2.f.l.a shall be employed.2)Performing a precision leak detection test to verify that the leakage rate from the fuel oil system is less than or, equal to.05 gallons per hour.sZ,~s-.~~~ea~~~.~3)...Starting both diesel generators simultaneouslyduring shutdown, and verifying that both diesel generators It'lE rw~py gp<wOS>vvcrhgE-yr wCn i+to V spaz.g~8ocy pr fog/.2 fFg.t a e per orme a ter any mo>scatsons w c cou a ect geese genera or interdependence.
O.C.COOK-UNIT 1 3/4 8-7 AMENDMENT NO.125
3 4.1 REACTIVITY CONTROL SYSTEMS hhSES 3 4.1.1 ADORATION epamOL 3 4.1.1.1 and 3 4.1.1.2 SHUTDOWNÃMARCIN A sufficienC SHUTD(RN MARCIN ensures ChaC 1)the reaccor can be made subcritical from all operating cond.tions, 2)the reaccivicy transieacs associated vith postulated accident conditioas aze controllable vithin acceptable limits, and 3)the reactor vill be maintained sufficiently subcricical to preclude inadvertent criticality ia the shutdovn condition.
/.3 SjiUTDOQN MARCIN requireme vary throughout co e life as a fuaction of fuel depletion, RCS boron c ncentration, and RCS T~The most restrictive conditioa occurs c EOL, vith T at ao j~ofd operating I temperature, and is associate vith a postdated steam line break accident and resulting uncontrolled R cooldova.Ia the analysis of this accident,.a minimum SHUTDON MARCIN of~+HH Delta k/k is iaitially required to control the reactivity transient and aucomatic'F is assumed to be available.
Vith Tavg less than 200 F, Che reactivicy transients resulting from a postulated steam line break cooldovn are minimal and a 1%Delta k/k SHUTDOWN MARCIN t provides adequate proceccion for this event~The SHUTDOWNÃMARCIN requirements are based upon che limiciag conditions described above and are consistent vith FSAR safety analysis assumptions.
3 4.1.1.3 EORON DIT ION A ainhaaa flov rate of at least 2000 CPM provides adequate miming, prevents stratification and ensures Chat reactivicy changes vill be gradual during boron concentratioa reductions ia the Reactor Coolant System.A flov rate of at least 2000 CPM vill circulate aa equivalent Reactor Coolant Syscem volume of 12,612 plus or minus 100 cubic feet ia approximately 30 minutes.The f reactivicy change x'atc usociated vith boron reductions vill therefore be vithin the capability for operator recognition and coatrol.3 4.1.1.4 MODERATOR TEMPGULTURE COEFFICIBIT MTC The limitations on MTC are provided to ensure that the assumptions used in the accident and transient analyses remaia valid through each fuel cycle.The suzveQlancc requirement for measurement of the ETC at the beginning, and near the cnd of each fuel cycl>>is adequate to coafizm thc MTC value since this coefficient changes slovly duc principally to the reductioa in RCS boron COOK NUCLM PEST UNIT 1 5 3/4 1 1 ANQiDMVT NO.74 i 08~>->
3Ae 3/4 4>qt'(C OR C~~')'p mr gS 5+g g+$&lfc V5/S I.I The plant is designed=o operate'2 all reactor coolant loops in operation, and maintain 5NR above~during all no~1 operations and ant'c pated transients.
A loss of f lou in tvo loops vill cause a reactor t"ip Lf operating above P-7 (11 percent of RATED THER'OWER) vhile a loss of flew in one loop vill cause a reactor tz'p if operating above P-8 (31 percent of.RATED THEq.'LAI.
POWER).In!LODE 3, a s'ngla reactor coolant loop provides suffic'ant heat removal capability for removing decay heat;hovavez, single failure considerations requize that tvo loops be OPMLE.Three loops aza raqu'"ed to be OP~LE and to operate'f the control rods aza capable of vithdzaval and t~e reactor-z'p breakers are c'osed..ne requirement assures adequate Dh3R mazgf.n'n che event of an uncontrolled rod v'=.".dza-a'.".
-".'s=ode.In LODES 4 and 5, a single reactor coolant loop oz RHR loop provides sufficient heac removal capabiLity for removing decay heat;but single ailuza consideracions require that at least tvo loops be OPBABLE.Thus, if the reacto coolant loops are not OPMI=, this speci ication zequi es cvo RHR Loops to be OPHABIZ.Tha operation of ona Reactor Coolant Pump or one RBR pump provides adequate flov to ensuza mixing, prevent stratification and produce gradual reactivicy changes during boron concanczacion reductions in the Reactor Coolant System.Tha reaccivity change rata associated vith boron reduction vi11, cherefoze, be vithin the capabilicy of operator recognition and contro L The rescrictions on starting a Reactor, Coolanc Pump belov P-7 vith one or more RCS cold legs Lass than or equal to 152 7 aza provided to pravenc RCS pressure cransiencs, caused by energy additions fzom cha secondary system, vhich could exceed the limits of Appendix C to LO CFR Paz-50.The RCS vil'e protected against overprassure transients and vill noc exceed the limits of, Appendix C by either (1)restzicting tha vataz volume in the pressurizer and thereby providing a volume for the primary cooLant to expand into oz (2)by restricting stazting of tha RCP's to vhen,the secondazy vatar temperature of beach steam generatoz's less than 50 F above each of the RCS cold leg temperacuzas
~COOK h"QCI"-Qt, P~~Z-4.lTT 1 3/4 4 A~~'fDl".~'iT NO.N, fa3, 167 ardEf 3 4.5.5 RE~AC VOTER STOay:t mm gEPCACZ gh Ttt 1>ski h The OPERAbILITY of the RVST as part of the ECCS ensures that sufficient negative reactivity is in)ected into the core to counteract any positive increase in reactivity caused by RCS systea cooldcnm, and ensures that a sufficient supply of.borate4 vater is available for in)ection by the ECCS in the event of~LOCA.Reactor coolant systeN cooldova can be caused by inadvertent depressurisation, a loss ot coolant accident or a steam line rupture.The limits on RQST minhtua volume and boron concentration ensur~t}1)sufficient vater is available vithin containment to ermit recirculation coolin flov to the core.and 2)~rea or vil rema s c tical n the old co ditio ol o ng a x ng of e RQS and e RCS a'ter lumes ith a: c trol ds i erted ce t or the st r tive ontro esse 1.These assumptions are consistent vith the IANNA analyses.The contained vater volume limit includes an allovance for vater not usable because of tank discharge line location or other physical characteristics.
The limits on contained vater volume an4 boron concentration of the EST als ensure a pH value of betveen 7.6 and 9.5 for the solution recirculated vithi containment after a LOCA.This pH band minimises the evolution ot iodine an mi.nimires the effect of chloride an4 caustic stress corrosion on mechanical systems and components.
The ECCS analyses to determine F~limip in Specifications 3.2.2 and 3.2.6 assumed a RVST vater temperature of 70 F.This temperature value of the RMS'ater determines that of the spray vater initially delivere4 to the containment following LOCA.It is one of the factors which determines the containment back-pressure in the ECCS analyses, performed in accordance vith the rovisions of 10 CFR 50.46 and h endix K to 10 CFR 50.~a ue o th mini um RV temp ature n Pc cal ecifi tion 5.5 h been conse ative cha e4 to 0 F t increa~the onsis ncy be een its 1~gZTE nd 2.The 1 er R T temp rature result in lo er con ainmen press re fro, c ntai nt sp y an4 afe 4s fl ass d to it th break.Love co tainme t pres ure re ults incre sed fl v res tance f ste exit g the ore reb evin<<floe and i reasi PCT.INSERT A the reactor will remain subcritical in the cold condition following a LOCA assuming mixing of the RWST, RCS, ECCS water, and other sources of water that may eventually reside in the sump, with all control rods assumed to be out.COON.'asm PmN-VÃIT 1 I 3/4 5 3 AHRf5lcÃT M.fg, li CONTAI~~3 a.g.1.4-PRES SVRE The limitations oa contaiaaant internal pressure ensure that 1)th+containment structure is prevented froa exceeding its design negative re!sure differential with respect, to the outside atmosphere of d psig and 2)the containment peak pressure does not exceed the design pressure of 12 psig during LOCA conditions.
<<.+9 The maximum peak pressure resulting from a LOCA event is calculated to be~~psig, which includes 0.3 ps'g for initial positive containment pressure.3 4 6 1 5 AIR TeuaERA L%E.he 1'=isa"."..s on conta'"."cn average air temperature ensure that 1)the containment
."=ass's'=.=ed to an i..i"ial mass sufficiently low to"".event exceeding t..e design"ress re d ring L"CA condi"'ons and 2)=he a.".biens a'r=e para='e='.es.-.==ex"eed:hat temperature allowable or the cont'mucus duty rating specified fsr cqu'pment and instrumentation located within ccn aŽen".~e ccntainment press re"rsnsient'.s sensit've:o
-'.~'ni='allv con=a'ned air-.ass"" itg a'"A...".e santa'ned a'ss'"..c"eases with"ecreasi.".z
=emperatu...-.e
'ower=e"perature
'-.=6O c will li pell pressure 0++s.,'h'less han.~co'n'tainmen't design pressure of 12"sig..he=per temperature limit I;.ences the peak ac"ident temperat re s'igh ly d r'g a'CA;however=his limit is based pr=ar'.v upon equ'pment"ro"ec='on and ant'cipa=ed
==era"ing conditions.
Both the upper and lover temperature limits are cons's"ent with the parameters.sed in=he accident analyses.3n.6.1.6 C"NTAI>~ENT VESSEL S.RUCT';RAL INTEGRITY This li='cation ensures that the structural integrity of the containment steel vessel will be maintained comparable to the ori-nal design standards for the life of the facility.Structural i-..=egrity is requi=ed o ensure thar, (1)the steel liner remains leak tight x-..d (2)the concrete surrounding the steel liner remains capable.=providing external missile protection for the steel liner and r-"iation shielc;;.g in the event of a LOCA.A visual inspection in con]u"..stion with Type A leakage tests is sufficient to demonstrate this capabi ity.COOK NUCLEAR PLANT UNIT 1 B 3/4 6-2 AMENDMENT NO, 1" 6
In accordance with the code requirements specified in Section 4,1.6 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.C.M>LUhK 5.4.2 5.5 For a pressure of 2485 psig, and For a temperature of 650'F, except for the pressurizer which is 680'F.HPF'Ro)OehVKLy IZ,,+&4 C~Slc.FeE7 AP O%Sl&PAP QE~FRdmg MM PCu~z~Auld Il>55/Cg8IC F'EZ7 AT 3OL'ran~GEAIEA!lb@
P~~iud.Th a'M I fh I y~hie-fees at a nominal T~of 70'F.5.5.1 The emergency core cooling systems are designed and shall be maintained in accordance with the original design provisions contained in Section 6.2 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements, with one exception.
This exception is the CVCS boron makeup system and the BIT.5.6 5.6.1.1 The spent f'uel storage racks are designed and shall be maintained with: a.A~equivalent to less than 0.95 when flooded with unborated water.b.A nominal 8.97 inch center-~nter distance between fuel assemblies placed in the storage racks.C.The fuel assemblies will'be classified as acceptable for Region 1, Region 2, or Region 3 storage based upon their assembiy average burnup versus initial nominal enrichment.
Cells acceptable for Region 1, Region 2, and Region 3 assembly storage are indicated in Figures 5.6-1 and 5.6-2.Assemblies that are acceptable for storage in Region 1, Region 2, and Region 3 must meet the design criteria that define the regions as follows: COOK NUCLEAR PLANT-UNIT 1 5-5 AMENDMENT NO.l,, 169 CORRECTED PAGE ATTACHMENT 7 TO AEP:NRC:1207 DESCRIPTION OF ANALYSES PERFORMED BY WESTINGHOUSE ELECTRIC CORPORATION FOR DONALD C.COOK NUCLEAR PLANT UNIT 2 Vantage 5 Reload Transition Safety Report foI Donald C.Cook Nuclear Plant Unit 2 6.0 SUhB~Y OP TECHNICAL SPECIFICATIONS CHANGES Table 6.1 presents a list of the Technical SpeciGcations changes.The changes noted in Table 6.1 are given in the proposed Technical Specifications page changes in Appendix A.43 TABLE 6.1

SUMMARY

OF TECHNICAL SPECIFICATIONS CHANGES SECIION PAGE CHANGE REASON FOR CHANGE 1.0, Add COLR COLR implementation pgI to index 1.12a, pg 1-3 Add COLR COLR implementation Figure 2.1-1, pg 2-2 Revised safety limits Reanalysis supports VANTAGE 5 reload 2.2.1, pg 2-5 Design Qow change&trip setpoint Change in design Qow due to VANTAGE 5 fuel reload, RTDP implementation Table 2.2-1, pg 2-7&24 Revise Overtemperature Reanalysis supports VANTAGE 5 reload hT limits Table 2,2-1, pg 2-9 Revise Overpower b.T limits Reanalysis supports VANTAGE 5 reload 2.1.1 Bases, pgB2-1&B 2-2 Update to bases VANTAGE 5 fuel reload and COLR implementation (relocation of FNgH)2.1.1 Bases, pg B 2-4 Update to bases VANTAGE 5 fuel reload and delete Cycle 6 speciQc information 2.1.1 Bases, pg B 2-5 Revise bases Reanalysis supports VANTAGE 5 reload 2.1.1 Bases, pg B 2-7 Revise bases circuit breaker time Reanalysis supports VANTAGE 5 reload 3/4.1.1.1, pg 3/4 1-1&1-2 Decrease shutdown margin Reanalysis with reduced SDM 3/4.1.1.2, pg 3/4 1-3&1-3b Decrease shutdown margin Reanalysis with reduced SDM.Change to Vfestinghouse dilution accident methodology TABLE 6.1

SUMMARY

OF TECHNICAL SPECIFICATIONS CHANGES (continued)
SECIION PAGE CHANGE REASON FOR CHANGE 3/4.1.1.4, pg 3/4 1-5&3/4 1-6 MTC relocated to COLR&revised EOL limit VANTAGE 5 fuel reload and COLR implementation (relocation of MTC)3/4.1.1.5, pg 3/4 1-7 Minimum temperature for surveillance req.Reanalysis with reduced temp 3/4.1.2.3, pg 3/4 1-11 Change ch.pump discharge head Make consistent with the analysis 3/4.1.2.4, pg 3/4 1-12 Change ch.pump discharge head Make consistent with the analysis 3/4.1.2.7, (pg 3/4 1-15 3/4.1.2.8, pg 3/4 1-16 Change 80 OF to 70 GF Change volume from 5650 to 7715 gallons&change 80 F to'70 F Make spec consistent with the analysis limit Make spec consistent with the VANTAGE 5 reload analysis limit to accommodate reduced rod worth and management flexibility 3/4.1.3.1, pg 3/4 1-19 Delete reference to Fig.3.1-1 COLR implementation 3/4.1.3.4, pg 3/4 1-23 Change rod drop time from 2.2 to 2.7 sec Relocate steps withdrawn to COLR Make spec consistent with the analysis limit&COLR implementation 3/4.1.3.5, pg 3/4 1-&Relocate shutdown rod insertion limits to COLR COLR implementation (relocation of shutdown rod insertion limits)45
TABLE 6.1

SUMMARY

OF TECHNICAL SPECIFICATIONS CHANGES (continued)
SECTION PAGE CHANGE REASON FOR CHANGE 3/4.19.6, pg 3/4 1-25 Relocate control rod insertion limits to COLR COLR implementation (relocation of control rod insertion limits)3/4.19.6, pg 3/4 1-26 Delete Qgure 3.1-1 COLR implementation 3/49.2.1, pg 3/4 2-1&2-3 Relocate axial Qux difference limits to COLR COLR implementation (relocation of AFD limits)3/4.3.2.1, pg 3/4 2-4 Relocate axial Qux difference allowable deviation Fig.to COLR COLR implementation (relocation of AFD allowable deviation) 3/43.2.2, pg 3/4 2-5 Relocate Fq limits to COLR COLR implementation (relocation of Fq limit)3/4.3.2.2, pg 3/4 24, 2-8a&24b Relocate K(Z)&V(Z)Qgures to COLR COLR implementation (relocation of Fq limit)3/4.3.2.3, pg 3/4 2-9 Relocate FN~limits to COLR'OLR implementation (relocation of FNgH limit)3/4.25.1, pg 3/4 2-15 Reformat DNB spec Change DNB parameter values and add low Tavg window Adopt planned Cook Nuclear Plant Unit 1 spec format consistent with VANTAGE 5 reload 3/4.2.5.1, pg 3/4 2-16&2-17&2-18 Delete tables 3.2-1 and 3.2-2 Delete 3.2.5.2 Adopt planned Cook Nuclear Plant Unit 1 spec format Not required TABLE 6.1

SUMMARY

OF TECHNICAL SPECIFICATIONS CHANGES (continued)
SECTION PAGE CHANGE REASON FOR CHANGE 3/4.3.2.6, pg 3/4 2-19 Relocate Fg limits COLR implementation (relocation to COLR-'f Fg limit)Changed deGnition of Fg Westnghouse CAOC methodology Table 33-2, pg 3/43-9&3-10 Changed and added RPS Make consistent with the analysis response times limits Table 3.4-4, pg 3/4 3-25 Change ESFAS setpoint Make consistent with analysis Table 3.3-5, pg 3/4 3-26&3/4 3-27&3/4 3-28 Changed ESF response time Make consistent with the analysis times limits 3/4.4.1.2, pg 3/4 4-2&4-3 Reduce number of RCPs Make consistent with the analysis required operable in limits mode 3 3/4.4.4, pg 3.4 4-6 Change water volume from 62%to 92%Make consistent with the analysis limit 3/4.4.6.2, pg 3/4 4-15&3/4 4-16 Controlled leakage in terms of resistance Consistent with analysis 3/45.1b, pg 3/4 5-1 Revise minimum contained borated water volume&min/max cover-pressure Make consistent with analysis limits (0 3/4.5.2.f, pg 3/4 5-5 Revised SI pump performance Reanalysis with degraded SI performance 47 TABLE 6.1 SUMMMRY OF TECHNICAL SPECIFICATIONS CHANGES (continued)
SECTION PAGE CHANGE REASON FOR CHANGE 3/45.2.h, pg 3.4 5-6 Revised SI pump Qow balance limits Adopt limits similar to Cook Nuclear Plant Unit 1 3/4$.5, pg 3/4 5-11 Reduce RWST min temp Make spec consistent with to 70'F analysis limit 3/4.1.1.1, pg B 3/4'1-1 Decrease shutdown margin Reanalysis with reduced shutdown margin B 3/4.1, pg B 3/4 1-3 Revise concentrations and volumes Make spec consistent with analysis limits B 3/4.2.1, pg B 3/4 2-1&2-2&2-3 Revise to reQect COLR implementation Changed to WCAP-8385 COLR implementation
'relocation of AFD limits)'estinghouse methodology B 3/4.2.2&3, pg B 3/4 2-4 thru 2-4b Revised to reQect COLR implementation
&VANTAGE 5 reload'ANTAGE 5 reload T-H analysis and COLR implementation (relocation of Fg and F gH limits)B 3/4.M, pg B 3/4 2-5 B 3/4.2.6, pg B 3/4 2-5 Revise to reQect reduced temp DNB limit Revise to reQect CAOC control Reanalysis with reduced temp\Make spec consistent with analysis B 3/4.55, pg B 3/4 5-3 Reduce RWST temp to 70 OF Make spec consistent with the analysis limit B 3/4.7.1, pg B 3/4 7-1 Reformat valve lift criteria Make consistent with the analysis limit TABLE 6.1

SUMMARY

OF TECHNICAL SPECIFICATIONS CHANGES (continued)
SECfION PAGE CHANGE REASON FOR CHANGE 3.4.9.1, pg B 3/4 9-1 Delete reference to refueling reactivity calcs at 2000 ppm Reanalysis of refueling reactivity at 2400 ppm boron 6.9.1.11, pg 6-18 Add COLR to section 6 COLR implementation 49 EHERCENCY CORE COOLENC SYSTE.'fS SURVEILLANCE RE UZREHENTS Continued d.At least once per 18 months by: 1.Verifying automat'c isolation and interlock action of the RHR system from che Reactor Coolant System Mhen the Reactor Coolant System pressure is above 600 psig,*2.h visual inspeccion of the containment sump and verifying thac the subsystem suction inlets are noc restricted by debris and tha-the sump components (trash racks, screens, etc.)shoe no ev'dence of sttuctural distress or corros ion.e.At least once per 18 months, during shucdovn, by: 1.Vexifying that each automatic valve in the floe path actuates co ics correct position qp a Safecy Infection test signal.2, Verifying that each of the following pumps"start automaticallv upon receipt of a safety XnJection test signal: a)Centrifugal charging pump b)Safety injection pump c)sidual heat r oval pump J jg~fi4 By verif inggthat each o the following pumps develops the indicate 4+ee~h ssure on recircu'ation fl.ou shen tested pursuant t icat on 4.0.5: 1.Centrifugal charging pump 2.Safecy In)eccion pump 3.Residual heat removaL pump~Za rO pseud$8'$foal~40 Psi g.By verifying the correct posicion of eac mechanical stop fo" the folloving Emergency Core Cooling System throctle valves: 1.within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months folloving complecion of each valve stroking operation or maintenance on the valve Mhen the ECCS subsyste'ms are required to be OPERABLE,*The provisions of Specification 4.0.7 are applicable.
D.C.COOK-UNTT 2 3/4 5-5 Amendment 8.~89 APPENDIX B NON-LOCA ANALYSES FOR THE DONALD C.COOK NUCLEAR PLANT UNIT 2 TRANSITION TO 17X17 VANTAGE 5 FUEL B9.11 Rupture of a Steamline (Steamline Break)B3.11.1 Introduction Although the no load temperature does not,change due to the plant rerating and VANTAGE 5 fuel, the impact of the various fuel parameter changes as well as various temperature and pressure operation was addressed.
Also, the nominal low steam pressure setpoint for steamline isolation and safety injection actuation is revised (lowered from 600 psig to 500 psig)to provide operating margin.As such, the rupture of a steam pipe event was analyzed.Included in the analysis are the design changes associated with the VANTAGE 5 transition and other modifled safety analysis assumptions as discussed in Section B.1.A rupture of a steam pipe results in an uncontrolled steam release from a steam generator.
The steam release results in an initial increase in steam flow which decreases during the accident as the steam pressure falls.The energy removal from the Reactor Coolant System causes a reduction of coolant temperature and pressure.In the presence of a negative coolant temperature coefficient, the cooldown results in a reduction of core'shutdown margin.If the most reactive RCCA is assumed stuck in its fully withdrawn position, there is an increased possibility that the core will become critical and return to power.A return to power following a steam pipe rupture is a potential concern mainly because of the high hot channel factors which exist when the most reactive RCCA is assumed stuck in its fully withdrawn position.The core is ultimately shut down by boric acid delivered by the Emergency Core Cooling System.The analysis of a steam pipe rupture is performed to demonstrate that: A.Assuming a stuck RCCA, with or without offsite power, and assuming a single failure in the engineered safety features, there is no consequential damage to the core and the core remains in place and intact.B.Although DNB and possible clad perforation following a steam pipe rupture are not necessarily unacceptable, the following analysis,'n fact, shows that no DNB occurs for any rupture assuming the most reactive RCCA stuck in its fully withdrawn position.~~~B.3.11.2 Method of Analysis The analysis of the steam pipe rupture has been performed to determine:
A.The core heat Qux and RCS temperature.;and pressure resulting from the cooldown following the steam line break.The LOFIRAN Code (Reference 5)has been used.B.The thermal and hydraulic behavior of the core following a steam line break.A detailed thermal and hydraulic digital computer code, THINC, has been used to determine if DNB occurs for the limiting core conditions computed in item A above.The following conditions were assumed to exist at the time of a main steam line break accident: A.End-of-life shutdown margin (19%6k/k)at no load, equilibrium xenon conditions, and the most reactive RCCA stuck in its fully withdrawn position.8-94 B.A negative moderator temperature coefficient corresponding to the endwf-life rodded core with the most reactive RCCA in the fully withdrawn position: The variation of the coefflcient with temperature and pressure has been included.The keff versus temperature at 1050 psia corresponding to the negative moderator temperature coefficien used plus the Doppler temperature effect, is shown in Figure B3-55.The Doppler power feedback assumed for this analysis is presented in Figure B3-56.The core properties associated with the sector nearest the affected steam generator'nd those associated with the remaining sector were conservatively combined to obtain average core properties for reactivity feedback calculation.
Further, it was conservatively assumed that the core power distribution was uniform.These two conditions cause underprediction of the reactivity feedback in the'high power region near the stuck rod.To verify the conservatism of this method, the reactivity as well as the power distribution was checked for the limiting conditions for the cases analyzed.This core analysis considered the Doppler reactivity from the high fuel temperature near the stuck RCCA, moderator feedback from the high enthalpy water near the stuck RCCA, power redistribution and non-uniform core inlet temperature effects.For cases'n which steam generation occurs in the high flux regions of the core, the effect of void formation was also included.It was determined that the reactivity employed in the kinetics analysis was always larger than the reactivity calculated including the above local effects for the statepoints.
These results verify conservatism; i.e., underprediction of k negative reactivity feedback from power generation.
G Minimum capability for injection of boric acid (2400 ppm)solution from the RWST corresponding to the most restrictive single failure in the safety injection system.The Emergency Core Cooling System (ECCS)consists of the following systems: 1)the passive accumulators, 2)the low head.safety injection (residual heat removal)system, 3)the intermediate head safety injection system, and 4)the high head safety injection (charging) system.Only the high head safety injection (charging) system and the passive accumulators are modeled for the steam line break accident analysis.Centrifugal Charging pump flow degradation of 10%was assumed.The modeling of the safety injection system in LOFTRAN is described in Reference 5.Figure B3-57 presents the safety injection flow rates as a function of RCS pressure assumed in the 8-95 analysis.The Qow corresponds to that delivered by one charging pump delivering its full Qow to the cold leg header.No credit has,.been taken for the low concentration borated water which must be swept from the lines downstream of the RWST isolation valves prior to the delivery of boric acid to the reactor coolant loops.For this analysis, a boron concentration of 0 ppm for the boron injection tank is assumed.For the cases where ofBite power is assumed, the sequence of events in the safety injection system is the following.
After the generation of the safety injection signal (appropriate delays for instrumentation, logic, and signal transport included), the appropriate valves begin to operate and the high head safety injection pump starts.In 27 seconds, the valves are assumed to be in their Gnal position and the pump is assumed to be at full speed and to draw suction from the RWST.The volume containing the low concentration borated water is swept into core before the 2400 ppm borated water reaches the core.This delay, described above, is inherently included in the modeling.In cases where offsite power is not available, an additional 10 second delay is assumed to start the diesel generators and to commence loading the necessary safety injection equipment onto them.D.Design value of the steam generator heat transfer coefQcient.
K Four combinations of break sizes and initial plant conditions have been considered in determining the core power transient which can result from large area pipe breaks.a Complete severance of a pipe downstream of the steam"Qow restrictor with the plant initially at no load conditions and all reactor coolant pumps running.b.Complete severance of a pipe inside the containment at the outlet of the steam generator (upstream of the steam Qow restrictor) with the same plant conditions as above.~c.Case (a)above with loss of off-site power simultaneous with the generation of the Safety Injection Signal (loss of AC power results in reactor coolant pump coastdown).
d.Case (b)above with the loss of offsite power simultaneous'with the Safety Injection Signal.A fifth case was analyzed to show that the DNBR remains above.the limit value in the event of the spurious opening of a steam dump or relief, valve.e.A break equivalent to a steam Qow of 265 lbs per second at 1100 psia from one steam generator with offsite power available.
F.Power peaking factors corresponding to one stuck RCCA are determined at end of core life assuming non-uniform core inlet coolant temperatures.
The coldest core inlet temperatures are assumed to occur in the sector with the stuck rod.The power peaking factors account for the effect of the local void in the region of the stuck control assembly during the return to power phase following the steam line break.This void in conjunction with the large negative moderator coeKcient partially offsets the effect of the stuck assembly.The power peaking factors depend upon the core power, temperature, pressure, and flow, and are thus different for each case studied.The analyses assumed initial hot shutdown conditions at time zero since this represents the most pessimistic initial condition.
Should the reactor be just critical or operating at power at the time of a steam line break,-.'the reactor will be tripped by the normal overpower protection system when power level reaches a trip point.Following a trip at power the reactor coolant system contains more stored energy than at no-load, the average coolant temperature is higher than at no-load and there is appreciable energy stored in the fuel.Thus, the additional stored energy is removed via the cooldown caused by the steam line break before.the no-load conditions of RCS temperature and shutdown margin assumed in the analyses are reached.After the additional stored energy has been removed, the cooldown and reactivity insertions proceed in the same manner as in the analysis which assumes no-load conditions at time zero.(In addition, since the initial steam generator water inventory is greatest at no-load, the magnitude and duration of RCS cooldown are more severe than steam line breaks occurring at power.8-97 G.In computing the steam flow during a steam line break, the Moody Curve (Reference 11)for~~fL/D=0 is used.H.The total delay time assumed for the steamline isolation is 11 seconds from receipt of actuation signaL The 11 second steamline isolation time includes valve closure time, and electronics and sensor delay.The Technical Speciflcations require a maximum 8 second valve closure time.For breaks downstream of the isolation valves, closure of all valves would completely terminate the blowdown.For any break, in any location following steamline isolation, no more than one steam generator would experience an uncontrolled blowdown even if one of the isolation valves fails to close.Plant characteristics and initial conditions are shown in Table B.2.4.B.3.11.3 Results The limiting case for Cases a through e was shown to be the double-ended rupture located upstream of the flow restrictor with oKsite power available (case b).Table B.3-10 lists the limiting statepoint for this worst case.The results presented are a conservative indication of the events which would occur assuming a steam line rupture.Figures B.3-58 through B3-60 show the RCS transient and core heat flux following a main steam line rupture (complete severance of a pipe),.upstream of the flow restrictor at initial no-load conditions.
Offsite power is assumed available so that full reactor coolant flow exists.The transient shown assumes an uncontrolled steam release f'rom only one steam generator.
Should the core be'critical at near zero power when the rupture occurs the initiation of safety injection by high differential pressure between any steamline and the remaining steamlines or by low steam line pressure in two steamlines will trip the reactor.Steam release from more than one steam generator will be prevented by automatic trip of the fast acting isolation valves in the steam lines by high-high'ontainment pressure signals or low steamline pressure or high steam flow coincident with low-low T-avg.Even with the failure of one valve, release from the other steam generators is terminated by steamline isolation while the one generator blows down.The steam line stop valves are assumed t to be fully closed in less than 11 seconds from receipt of a closure signal.B-98 As shown in Figure B3M, the core attains criticality with the RCCAs inserted (with the design shutdown margin assuming one stuck RCCA)before boron solution (2400 ppm from RWST)enters the RCS.A peak core power less than the nominal full power value is attained.The calculation assumes the boric acid is mixed with, and diluted by, the water flowing in the RCS prior.to entering the reactor core.The concentration after mixing depends upon the relative flow rates in the RCS and in the safety injection system.The variation of mass flow rate in the RCS due to water density changes is included in the calculation as is the variation of flow rate in the safety injection system due to changes in the RCS pressure.The safety injection system flow calculation includes the line losses in the system as well as the pump head curve.The assumed steam release for an accidental depressurization of the main steam system (case e)is the maximum capacity of any single steam dump, relief, or safety valve.Safety injection is initiated automatically by low pressurizer pressure.Operation of one centrifugal charging pump is assumed.Boron solution at 2400 ppm enters the RCS providing sufficien negative reactivity to prevent core damage.The transient is quite conservative with respect to cooldown, since no credit is taken for the energy stored in the system metal other than that of the fuel elements or the energy stored in the other steam generators.
Since the transient occurs over a period of about 5 minutes, the neglected stored energy is likely for this event to have a signiflcant effect in slowing the cooldown.The DNB transient is bounded by the limiting case for a steamline rupture.The DNB analysis for the limiting case (double-ended rupture located upstream of the flow restrictor) showed that the minimum DNBR remained above the limit value.The DNBR design basis limit for the hypothetical steamline break event is 1.45.The pressures for this event fall in the low pressure range (500-1000 psia)where the W-3 based DNB correlation is used with a 1.45 limit DNBR.This design limit for low pressure applications of the W-3 correlation has been approved by the NRC in Reference 15.Although the low pressure limit was approved in conjunction with WCAP-9227-NP, which is not referenced in the Cook Nuclear Plant Unit 2 UFSAR, the SER is an applicable reference for reload designs.The calculated sequence of events for the limiting case (doublewnded rupture located upstream of the flow restrictor) are shown in Table B.3-11.8.99
B3.11.4 Conclusions
~~~The analysis has shown that the criteria stated earlier are satisGed.Although DNB and possible clad perforation following a steam pipe rupture can be acceptable and is not precluded by the criteria, the above analysis, in fact, shows that no DNB occurs for the rupture (including an accidental depressurization of the main steam system)assuming the most reactive RCCA stuck in its fully withdrawn position.
TABLE B.3-10 LIMI'.HNG S'ZEAMLINE BREAK STATEPOINT DOUBLE ENDED RUPTURE INSIDE CONTAINMENT WITH OFFSITE POWER AVAILABLE Sec Psia 100.2 598.7 Fraction~F OF Frac PPM Percent 0.107 330.2 441.8 1.0 1.51 0.044 Inlet Temp Time Pressure Heat Flux Cold Hot Flow Boron Reactivity Density~GM CC 0.863 8-132
(g TABLE B.3-11 TIME SEQUENCE OF EVENTS Accidents Rupture of a Steamline Events Y~ime sec 1.Inside Containment With Offsite Power available Steam line ruptures 0.0 Low steamline pressure setpoint reached 0.26 Feedwater Isolation (All loops)8.26 Steamline Isolation (Loops 2, 3 and 4)11.26 Pressurizer empties 13.8 SI fiow starts 27.26 Criticality attained 29.4 Boron from SI reaches cores 38.2 Peak heat flux attained 100.2 Core becomes subcritical 116.2 8-133 TABLE B3-11'continued)
TIME SEQUENCE OF EVENTS Accidents Rupture of a Steamline Events T~ime sec 2.Inside Containment Without Offsite Power available Steam line ruptures 0.0 Low steamline pressure setpoint reached 0.26 Feedwater Isolation (All loops)8.26 Steamline Isolation (Loops 2, 3 and 4)s 11.26 Pressurizer empties 15.4 Criticality attained 37.0 SI flow starts 37.26 Boron from SI reaches cores 51.4 Peak heat flux attained 236.0 Core becomes subcritical 291.7 8-134 1,05 102 1.015 1.01 0,995 240 280 320%0 400 440 480 520 560 CORE AVERAGE TENPERATURE
('F)Figure B.3-55 Variation of Reactivity with Core Temperature at 1050 PSIA for the End of Life Rodded Core with One Control Rod Assembly Stuck (Assumes Zero Power)
g I-UJ 1.4 i2 2 C/l O o ED (~08 0.6 CORE POWER (FRACTION OF NOMINAL)Figure B.3-56 Dopp1er Power Feedback for Steamline Break 8.227 1.6 0,6 COLD LEG SAFETY INJECTION (LB/SEC)Figure B.3-57 Safety Injection Flow Supplied by One Charging Pump B-228
.275.25.225.2.175 U.15.'125 CD~1 cC~.875.65.825 58.188.158.288.258.588.(y..275-~25.225 VJ~2 C/l.175.15.125~cD.875.825 8.8, C:n vO~108.158.208.256.588.TINE (SEC)Figure B.3-58 Nuclear Power'and Core Heat Flux Versus Time Steamline Break OER Inside Containment with Power
550.CLI P-5 LU CD LP 568 L 458.488.358.560.250.0, t 25CG.2250.2000.1750.150C.1259.1908.7c6 500.25C.-'L~56.IEG.156.260.250.300.IL...15~cCO.c56.300.r I)'5)5-.C))608.1460.LLJ CD CCl)-cc CL'200.1600.680~698.400.0.50~180.158.208~250'CO.TIME (SEc)Figure B.3-59 Core Average Temperature, RCS Pressure, and Pressurizer Water Volume Versus Time Steamline Break DER Inside Containment with Power 8-230 2588.2888.1=08.1808.O 568.8~I--588.-1668.CX 1588-2888.-2568.50.168 158 288 258 388 258.288.O O 158.m 188.ED 58.c8 8.50.168.158.288.258.388.TINE (SEC)Figure B.3-60 Reactivity and Core Boron Concentration Versus Time Steamline Break OER Inside Containment with Power 8-231 APPENDIX C LOCA ANALYSES FOR THE DONALD C.COOK NUCLEAR PLANT UNIT 2'IRANSITION TO 17x17 VANTAGE 5 FUEL C.3.1.2 MAJOR LOCA ANALYSES APPLICABLE TO WFDTINGHOUSE FUEL Identification of Causes and Fre uen Classi6cation A losswf-coolant accident (LOCA)is the result of a pipe rupture of the RCS pressure boundary.For the analyses reported here, a major pipe break (large break)is deGned as a rupture with a total cross-sectional area equal to or greater than 1.0 ft2.This event is considered an ANS Condition IV event, a limiting fault, in that it is not expected to occur during the lifetime of the Donald C.Cook Nuclear Plant Unit 2, but is postulated as a conservative design basis.The Acceptance Criteria for the LOCA are described in 10 CFR 50.46 (10 CFR 50.46 and Appendix K of 10 CFR 50 1974)(1)as follows: The calculated peak fuel element clad temperature is below the requirement of 2200oF.<2.The amount of fuel element cladding that reacts chemically with water or steam to generate hydrogen, does not exceed 1 percent of the total amount of Zircaloy in the fuel rod cladding.3.The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling.The localized cladding oxidation limit of 17 percent is not exceeded during or after quenching.
4.The core remains amenable to cooling during and after the break.~5.The core temperature is reduced and decay heat is removed for an extended period of time, as required by the long-lived radioactivity remaining in the core.These criteria were established to provide a significant margin in emergency core cooling system (ECCS)performance following a LOCA.WASH-1400 (USNRC 1975)(10)presents a study in regards to the probability of occurrence of RCS pipe ruptures.
li Se uence of Events and S tems 0 erations Should a major break occur, depressurization of the RCS results in a pressure decrease in the pressurizer.
Loss-Of-OKsite Power (LOOP)is assumed coincident with the occurence of thc break.The reactor trip signal subsequently occurs when the pressurizer low pressure trip sctpoint is reached.A safety injection signal is generated when the appropriate setpoint is reached.These countermeasures will limit the consequence of the accident in two ways: Reactor trip and borated water injection supplement void formation in causing rapid reduction of power to the residual level corresponding to fission product decay heat.No credit is taken in the LOCA analysis for the boron content of the injection water, however an average RCS/sump mixed boron concentration is calculated to ensure that thc core remains subcritical.
In addition, the insertion of control rods to shut down the reactor is neglected in the large break analysis.2.Injection of borated water provides for heat transfer from the core and prcvcnts cxccssivc clad temperatures.
In the present Westinghouse design, the large break single failure is the loss of onc RHR (low head)pump.This means that credit could be taken for two high head charging pumps, two safety injection pumps, and one low head pump.The following is a discussion of the modelling proccdurc for the minimum safeguards and the flow splitting from a break of an ECCS branch injection linc (i.e., the spilling line assumptions).
The current procedure for large break analyses assumes that at least one train of ECCS is availahlc for delivery of water to the RCS.Although the single failure is an RHR pump, only onc pump in each subsystem is assumed to deliver to the primary loops.However, both Emcrgcncy Dicscl Generators (EDGs)are assumed to start in the modelling of the containment deck t'ans and sprays.Modelling full containment heat removal systems operation is required by Branch Tcchnical Position CSB 6-1 and is conservative for the large break LOCA.The high head charging pump starts>>nd delivers Oow through the injection lines (one per loop)with one branch injection linc spilling to the containment backpressure.
To minimize delivery to the reactor, the branch linc chosen tn spill C is selected as the one with the minimum resistance.
When one safety injection pump and onc low head residual heat removal pump start, Qow is delivered to the reactor coolant system through thc accumulator injection lines.Again, one line, with the minimum resistance, is assumed to spill.to r containment backpressure.
In addition, all safety injection pump performance curves were degraded s by 10%and a 25 gpm Qow imbalance was assumed for the high head charging pumps.Therefore, in the large break ECCS analysis performed by Westinghouse, single lailure is conservatively accounted for via the loss of an ECCS train, and the spilling of the minimum resistance injection line despite full containment active heat removal system operation (i.c., two EDGs).The time sequence of events following a large break LOCA is presented in Table C.3.1-5.Thc safety injection performance, as modelled for the large break LOCA, is presented in Figures C.3.1.1 and C.3.1.2.'(I Before the break occurs, the unit is in an equilibrium condition; that is, the heat generated in thc core is being removed via the secondary system, During blowdown, heat from emission product decay, hot internals and the vessel, continues to be transferred to the reactor coolant.At thc beginning of the blowdown phase, the entire RCS contains subcooled liquid which translcrs heat l'rom thc core by forced convection with some fully developed nucleate boiling.After thc brcak dcvclops, the time to departure from nucleate boiling is calculated, consistent with Appendix K of l0 CFR 50(1).Thereafter, the core heat transfer is unstable, with both nucleate boiling and l>lm boiling occurring.
As the core becomes uncovered, both turbulent and laminar forced convection
<<nd radiation are considered as core heat transfer mechanisms.
The heat transfer between the RCS and the secondary system may be in either direction, dcpcnding on the relative temperatures.
In the case of continued heat addition to the secondary system, thc secondary system pressure increases and the main steam safety valves may actuate to limit thc pressure.Makeup water to the secondary side is automatically provided by thc auxiliary fccdwatcr system.The safety injection signal actuates a feedwater isolation signal which isolates main feedwater flow by closing the main feedwater isolation valves, and also initiates auxiliary fccdwatcr flow by starting the auxiliary feedwater pumps.The secondary flow aids in the reduction of RCS ipressure.C-3 When the RCS depressurizes to 600 psia, the accumulators begin to inject borated water into the reactor coolant loops.The conservative assumption is made that accumulator water injected bypasses the core and goes out through the break until the termination of bypass.This conservatism is again consistent with Appendix K of 10 CFR 50.Since loss of offsite power (LOOP)is assumed, the RCPs are assumed to trip at the inception of the accident.The effects of pump coastdown are included in the blowdown analysis.The blowdown phase of the transient ends when the RCS pressure (initial values with uncertainty assumed to range from 2037 to 2313 psia)falls to a value approaching that of the containment atmosphere.
Prior to or at the end of the blowdown, the mechanisms that are responsible for the emergency core cooling water bypassing the core, are calculated not to be effective.
At this time (called end-of-bypass) refill of the reactor vessel lower plenum begins.Refill is completed when emergency core cooling water has filled the lower plenum of the reactor vessel, which is bounded by the bottom of the fuel rods (called bottom of core recovery time).The reflood phase of the transient is defined as the time period lasting from the bottom of core f recovery until the reactor vessel has been filled uiith water to the extengthat the core temperature rise has been terminated.
From the latter stage of blowdown and then the beginning-of-reflood, the safety injection accumulator tanks rapidly discharge borated cooling water into the RCS, contributing to the filling of the reactor vessel downcomer.
The downcomer water elevation heaps provides the driving force required for the reflooding of the reactor core.The RHR (low head), safety injection and high head charging pumps aid in the filling of the downcomer and, subsequently, supply water to maintain a full downcomer and complete the reflooding process.Continued operation of the ECCS pumps supplies water during long-term cooling.Core temperatures have been reduced to long-term steady state levels associated with the dissipation of residual heat generation.
After the water level of the easMuab water storage tank (RW~reaches a minimum allowable value, coolant for long-term cooling of the core is obtained by switching to the cold leg recirculation phase of operation in which spilled borated water is drawn from the engineered safety features (ESF)containment sumps by the residual heat removal (low head)safety injection pumps and returned to the RCS cold legs.The containment spray system continues to operate to further reduce containment pressure.
Approximately 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the initiation of the LOCA, the ECCS is realigned to inject water to the RCS hot legs in order to control the boric acid concentration in the reactor vesseL Long-term cooling includes long-term criticality control: Criticality control is achieved by determining the RWST and accumulator concentrations necessary to maintain subcriticality without credit for RCCA insertion.
The necessary RWST and accumulator concentrations are a function of each core design and are checked each cycle.The current Technical Specifications value are 2400 to 2600 ppm boron for the RWST and 2400 to 2600 ppm for the accumulators.
The accumulators are conservatively modelled at 2300 ppm for the post-LOCA subcriticality requirement.
Core and S tern Performance Mathematical Model: The requirements of an acceptable ECCS evaluation model are presented in Appendix K of 10 CFR So(1).Large Break LOCA Evaluation Model The analysis of a large break LOCA transient is divided into three phases: (1)blowdown, (2)refill, and (3)reQood.There are three distinct transients analyzed in each phase, including the thermal-hydraulic transient in the RCS, the pressure and temperature transient within the containment, and the fuel and clad temperature transient of the hottest fuel rod in the core.Based on these considerations, a system of interrelated computer codes has been developed for the analysis of the LOCA.A description of the various aspects of the LOCA analysis methodology is given by Bordelon,-
Massie, and Zordan (1974)().This document describes the major phenomena modeled, the interfaces among the computer codes, and the features of the codes which ensure compliance with the Acceptance Criteria The SATAN-VI, WREFLOOD, BASH and LOCBART codes, which are used in the LOCA analysis, are described in detail by Bordelon et al.(1974)(5);
Kelly et al.(1974)();Young et al.(1987)(");
and Bordelon et al.(1974)(6).
Code modifications are specified in References 2, 7, 13, and 17.These codes assess the core heat transfer geometry and determine if the core remains amenable to cooling through and subsequent to the blowdown, refill, and refiood phases of the LOCA.The SATAN-VI computer code analyzes the thermal-hydraulic C-5 transient in the RCS during blowdown and the WREFLOOD computer code calculates this~~transient during the refill phase of the accident.The BASH code is used to determine the RCS response during the reflood phase of the transient.
The LOTIC computer code, described by Hsieh and Raymund in WCAP-8355 (1975)and WCAP-8345 (1974)(), calculates the containment backpressure transient.
The containment backpressure transient is input to BASH for the purpose of calculating the reflood transient.
The LOCBART computer code calculates the thermal transient of the hottest fuel rod in the three phases.The improved fuel performance model, described in Reference 15, generates the initial fuel rod conditions input to LOCBART.SATAN-VI calculates the RCS pressure, enthalpy, density, and the mass and energy flow rates in the RCS, as well as steam generator energy transfer between the primary and secondaiy systems as a function of time during the blowdown phase of the LOCA.SATAN-VI also calculates the accumulator water mass and internal pressure and the break mass and energy flow.rates that are assumed to be vented to the containment during blowdown.At the end of the blowdown, information on the state of the system is transferred to the WREFLOOD code which performs the calculation of the refill period to bottom of core (BOC)recovery time.Once the vessel has refilled to the bottom of the core, the reflood portion of the transient begins.The BASH code is used to calculate the thermal-hydraulic simulation of the RCS for the reflood phase.Information concerning the core boundary conditions is taken from all of the above codes and input to the LOCBART code for the purpose of calculating the core fuel rod thermal response for the entire transient.
From the boundary conditions, LOCBART computes the fluid conditions and heat transfer coefficient for the full length of the fuel rod by employing mechanistic models appropriate to the actual flow and heat transfer regimes.Conservative assumptions ensure that the fuel rods modeled in the calculation represent the hottest rods in the entire core.The large break analysis was performed with the December 1981 version of the Evaluation Model modified to incorporate the BASH(d)computer code.
Input Parameters and Initial Conditions:
The analysis presented in this section was performed with a reactor vessel upper head temperature equal to the RCS hot leg temperature.
A range of reactor operating temperatures was analyzed in order to justify plant operation at a reactor power level of 3588 Mwt between 582.2 F to 615;2 F in the hot legs and 511.7'F to 547.6 F in the cold legs.In addition to the temperature range analyzed, initial RCS prcssurizcr pressure was also varied to justify plant operation between 2037 and 2313 psia.A full spectrum break analysis was done at the high pressure/high temperature RCS conditions (initial RCS pressurizer pressure, with uncertainty, of 2313 psia and initial hot leg temperature of 615.2 F)from which the limiting break size was determined.
The limiting break was then reanalyzed lor low temperature and high RCS pressure.The limiting break was also reanalyzed for thc high temperature and low initial RCS pressure of 2037 psia.The analysis also considered plant operation at reduced power level with the RHR cross tie valve closed.The reduction in power level was considered necessary to offset the reduction in safety injection flow due to thc closure.of the RHR cross tie valve.This case assumed a reduced power level of 3413.MWt and minimum safeguards with the RHR cross tie valve closed at the limiting RCS conditions.
All cases conservatively assumed 15%steam generator tube plugging in all four steam generators.
Table.C.3.1-1 describes the cases analyzed.Tables C.3.1-2 and C3.1-3 summarize thc kcy input I parameters and setpoints modelled in the Cook Nuclear Plant Unit 2 large brcak LOCA analysis.The bases used to select the numerical values that are input parameters to the analysis have bccn conservatively determined from extensive sensitivity studies (Westinghouse 1974(12);Salvatori 1974();Johnson, Massie, and Thompson 1975(8)).In addition, the requirements of Appendix (11).K to 10 CFR 50(1)regarding speciGc model features were met by selecting models which provide a significant overall conservatism in the analysis.The assumptions which were made pertain to thc conditions of the reactor and associated safety system equipment at the time that the LOCA occurs, and include such items as the core peaking factors, the containment pressure, and thc perl'ofillancc of the ECCS.Decay heat generated throughout the transient is also conservatively calculated as per the requirements of Appendix K to 10 CFR 50(1).C-7 Another input parameter that affects LOCA analysis results is the assumed axial power shape at~~~the beginning of the accident.Power shape sensitivity studies performed with Westinghouse ECCS evaluation models have always demonstrated the chopped cosine shape with the peak at the core midplane to be limiting.Westinghouse has performed"spot check" analyses using the BASH reflood evaluation model for power shapes skewed to the top of the core.Results of these analyses have demonstrated the chopped cosine peaked at the core midplane remains the limiting power shape(18)A meeting was held at the Westinghouse Licensing Office in Bethesda on December 17, 1981, between members of the U.S.Nuclear Regulatory Commission and members of the Westinghouse Nuclear Safety Department to discuss the impact of maximum safety injection on the large break ECCS analysis on a generic basis.Further discussion of this issue is provided in a letter from E.P.Rahe, Manager of Westinghouse Nuclear Safety Department, to Robert I Tedesco of the U.S.Nuclear Regulatory Commission(14).
A brief description of this issue is given below.Westinghouse ECCS analyses currently assume minimum safeguards for the safety injection flow,~~which minimizes the amount of flow to the RCS by assuming maximum injection line resistances, degraded ECCS pump performance, and the loss of one residual heat removal (RHR)pump as the most limiting single failure.This is the'limiting single failure assumption when offsite power is unavailable for most Westinghouse plants.However, for some Westinghouse plants, including Cook Nuclear Plant Unit 2, the current nature of the Appendix K ECCS evaluation models is such that it may be more limiting to assume the maximum possible ECCS flow delivery.In that case, maximum safeguards which assume minimum injection line resistances, enhanced ECCS pump performance, and no single failure, result in the highest amount of flow delivered to the RCS.Therefore, the worst break for Cook Nuclear Plant Unit 2 (CD=Q.6)was reanalyzed, assuming maximum safeguards (Case A vs.Case F of Table C.3.1-1).Examination of the LOCA analysis results in Table C.3.1-6 demonstrates that minimum safeguards assumptions result in the highest peak clad temperature for Cook Nuclear Plant Unit 2.
When assessing the effect of transition cores on the large break LOCA analysis, it must be determined whether the transition core can have a greater calculated peak cladding temperature (PCT)than either a complete core of the 17x17 ANF assembly design or a complete core of the Westinghouse 17x17 VANTAGE 5 design.For a given peaking factor, the only mechanism available to cause a transition core to have a greater calculated PCT than a full core of either fuel is the possibility of flow redistribution due to fuel assembly hydraulic resistance mismatch.Hydraulic resistance mismatch will exist only for a transition core and is the only unique difference between a complete core of either fuel type and the transition core.The 17x17 ANF fuel assembly is nearly identical to the Westinghouse 17x17 OFA assembly in terms of hydraulic and geometric characteristics.
Therefore, the analyses reported in Reference 19 which demonstrate that the 17x17 VANTAGE 5 fuel features result in a fuel assembly that is more limiting than a Westinghouse.17x17 OFA fuel assembly, with respect to large break LOCA ECCS performance, remain valid as applied at Cook Nuclear Plant Vnit 2.The same large break LOCA transition core penalty reported.in Section 5.23 of Reference 19 will be applied to the transition from 17x17 ANF fuel assemblies to Westinghouse 17x17 VANTAGE 5 fuel assemblies.
Westinghouse transition core designs, including specific 17X17 OFA to 17x17 VANTAGE 5 transition core cases, were analyzed.The increase in hydraulic resistance for the VANTAGE 5 assembly was shown to produce a reduction in reflood steam flow rate for the VANTAGE 5 fuel at mixing vane grid elevations for transition core configurations.
The various fuel assembly specific transition core analyses performed resulted in peak cladding temperature increases of up to 50 F for core axial elevations that bound the location of the PCT.Therefore, the maximum PCI'enalty possible for VANTAGE 5 fuel residing in a transition core is 50 F, Reference 19.As stated earlier, this transition core penalty continues to apply to the transition from 17x17 ANF fuel assemblies to Westinghouse 17x17 VANTAGE 5 fuel assemblies due to the near identical design of 17x17 ANF and Westinghouse 17x17 OFA fuel assemblies.
Once a full core of VANTAGE 5 fuel is achieved the large break LOCA analysis will apply without the transition core penalty.C-9
Results: Based on the results of the LOCA sensitivity studies (Westinghouse 1974(1);Satvatori 1974(11);Johnson, Massie, and Thompson 1975(8)), the limiting large break was found to be the doublewnded cold leg guillotine (DECLG).Therefore, only the DECLG break is considered in the large break ECCS performance analysis.Calculations were performed for a range of Moody break discharge coefficient.
The results of these calculations are summarized in Tables C3.1-5 and C3.1-6.The containment data used to generate the LOTIC backpressure transient are shown in Table.C3.1-4.The mass and energy release data for the minimum and maximum safeguards cases (Case A and F)are shown in Tables C.3.1-7 and C.3.1-8 respectively.
In addition, mass and energy release data for Case G.(3413 Mwt, RHR cross tie valve closed)are shown in Table C3.1-9.The mass releases for the remaining cases are not presented, since they do not vary signiflcantly from the data shown in Table C.3.1-7.Nitrogen release rates to the containment are given in Table C.3.1-10.Figures C.3.1-3a through C.3.1-30 present the results of the cases analyzed for the large break LOCA, The alpha designation in the figure number corresponds to the cases as described in Table C.3.1-1.Figures C3.1-3a-g The system pressure shown is the calculated core pressure.Figures C.3.1-4a-g The flow rate from the break is plotted as the sum of both ends of the guillotine break.Figures C3.1-5a-g The core pressure drop shown is from the lower plenum, near the core, to the upper plenum at the core outlet.Figures C.3.1-6a-g The core flow rate is shown during the blowdown phase of the transient.
Figures C.3.1-7a-g The accumulator flow rate during blowdown is plotted as the sum of that injected into the intact cold legs.
P Figures C3.14a-g The core and downcomer collapsed liquid water levels are plotted during the reflood phase of the transient.
Figures C.3.1-9a-g The core inlet flow rate is shown as it is calculated during the reflood phase.Figures C3.1-10a-g The total pumped ECCS fiow rate injecting into the intact cold legs is shown.Figures C3.1-11a-g The integral of the core inlet flow rate as calculated with BASH is plotted.Figures C3.1-12a-g The mass flux is plotted at the hot spot (the node which produced the peak clad temperature) on the hot rod.Figures C.3.1-13a-g The heat transfer coefficient is plotted at the hot spot on the hot rod.Figures C.3.1-14a-g The fluid temperature at the hot spot on the hot rod is plotted.Figures C3.1-16-18 The containment backpressure transient used in the analysis is provided for Cases A, F and G (the minimum and maximum SI flow cases, and the 3413 Mwt cross tie valve closed case).Figures C.3.1-19-27 These flgures show the heat removal rates of the heat sinks found in the lower and upper compartment and the heat removal by the sump and lower compartment spray for Cases A, F and G.Figures C.3.1-28-30 These figures show the temperature transients in both the lower and upper compartments of containment and flow from the upper to lower compartments for Cases A, F and G.The peak clad temperature calculated for a large break is 2140'F, which is less than the acceptance criterion limit of 2200~F.The maximum local metal-water reaction is 6.80 percent, which is well below the embrittlement limit of 17 percent as required by 10 CFR 50.46.The total
metal-water reaction is less than 0.3 percent for all breaks, corresponding to less than 0.3 percent hydrogen generation, as compared with the 1 percent criterion of 10 CFR 50A6.The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling.As a result, the core temperature will continue to drop and the ability to remove decay heat generated in the fuel for an extended period of time will be provided.
DONALD C.COOK NUCLEAR PLANT UNIT 2 TABLE C3.1-2 INPUT PARAMETERS USED IN THE LARGE BREAK LOCA ECCS ANALYSIS Cross T~ies 0 e RHR Cross Ties Closed License Core Power(a), (MWt)Peak Linear Power(), (kw/ft)Total Peaking Factor, FgT Axial Peaking Factor, FZ Hot Channel Enthalpy Rise Factor, F gH Power Shape: Fuel Assembly Array Accumulator Water Volume, Nominal (ft3/accumulator)
Allowance Accumulator Tank Volume, Nominal (ft/accumulator)
Accumulator-Gas Pressure, Minimum (psia)Safety Injection Pumped Flow Rate (All pumps degraded 10'Yo, Charging pump flow rate imbalance=25 Containment Parameters Initial Loop How (GPM)Vessel Inlet Temperature
(~F)Vessel Outlet Temperature (F)Average Reactor Coolant Pressure (psia)Steam Pressure (psia)Steam Generator Tube Plugging Level (%)Refueling Water Storage Tank Temperature
('F)3588 12.714 2.220 1.370 1.620 3413.12.721 2.335 1.420 1.644 Chopped Cosine 17 X 17 VANTAGE 5 946 946+25+25 1350 1350 600'600 See Figures C.3.1.1 gpm)and C3.1.2 See Table C.3.1-4 88,500 88,500 511.7 to 513.3 to 547.6 546.4 582.2 to 580.6 to 615.2 611.2 2037.4 to 2037.4 to 2312.6 2312.6 587 to 603 to 820'20 15 15 70(b)70(b)(a)Two percent is added to this power to account for calorimetric error.(b)The BASH computer code models average RWST temperature during core reflooding (85~F).Other computer codes in the model use 70~F.
C.3.2 LOSS OF REACTOR COOLANT FROM SMALL RUPTURED PIPES OR FROM CRACKS IN LARGE PIPES WHICH ACTUATES THE EMERGENCY CORE COOLING SYSTEM Anal is of Effects and Cons uences Method of Anal is For small breaks (less than 1.0 ft2)the NOTRUMP()()digital computer code is employed to calculate the transient depressurization of the Reactor Coolant System as well as to describe the mass and enthalpy of the Quid flow through the break.Small Break LOCA Anal is Usin NOTRUMP For loss-of-coolant accidents due to small breaks less than 1 square foot, the NOTRUMP computer code()()is used to calculate the transient depressurization of the RCS as well as to describe (10)(11)~the mass and enthalpy of the Quid flow through the break.The NOTRUMP computer code is a state-of-the-art one-dimensional general network code incorporating a number of advanced features.Among these are calculation of thermal nonwquilibrium in all Quid volumes, flow regime-dependent drift flux calculations with counter-current Qooding limitations, mixture level tracking logic in multiple-stacked Quid nodes and regime4ependent heat transfer correlations.
The NOTRUMP small-break LOCA emergency core cooling system (ECCS)evaluation model was developed to determine the RCS response to design basis small break LOCAs, and to address NRC concerns expressed in NUREG-0611,"Generic Evaluation of Feedwater Transients and Small Break Losswf-Coolant Accidents in Westinghouse-Designed Operating Plants".The reactor coolant system model is nodalized into volumes interconnected by Qowpaths.The broken loop is modelled explicitly, while the three intact loops are lumped into a second loop.Transient behavior of the system is determined from the governing conservation equations of mass, energy, and momentum.The multinode capability of the program enables explicit, detailed spatial representation of various system components which, among other capabilities, enables a proper calculation of the behavior of the loop seal during a losswf-coolant accident.The reactor core is represented as heated control volumes with associated phase separation models to permit transient mixture height calculations.
Detailed descriptions of the NOTRUMP code and the evaluation model are provided in References 10 and 11.
<0, After the small break LOCA is initiated, reactor trip occurs due to a low pressurizer pressure signal (1860 psia).Soon after the reactor trip signal is generated, the safety injection signal is actuated due to low pressurizer pressure (1715 psia).Safety injection systems consist of gas pressurized accumulator tanks and pumped injection systems.The small break LOCA analysis assumed an accumulator water volume cflual to the average of that allowed in the technical specification with a cover gas pressure of 600 psia.This is the minimum of the cover gas pressure.allowed in the Technical Specifications.
Mni1num emergency core cooling system availability is assumed for the analysis at the maximum RWST temperature.
Assumed pumped safety injection characteristics as a function of-RCS pressure used as boundary conditions in the analysis are shown in Figure C.3.2-1 and in Table C9.2-6.The safety injection flow rates presented are based on pump performance curves degraded 10 percent from the design head and an assumed charging system branch linc imbalance of 25 gpm.The effect of flow from the RHR pumps is not considered in the snmll break LOCA analyses since their shutoff head is lower than the RCS pressure during tbc time portion of the transient considered here.Safety injection (SI)is delayed 27 seconds after the occurrence of the injection signal to account for diesel generator startup and emergency power bus loading in case of a loss of offsite power coincident with a LOCA.The small break LOCA analysis also assumed that the auxiliary feedwater pumps were degraded by 15 percent and that thc rod drop time was 2.7 seconds.Peak clad temperature calculations are performed with the LOCTA-IV(2) code using the NOTRUMP calculated core pressure, fuel rod power history, uncovered core steam flow and mixture heights as boundary conditions.
Figure C3.2-10 depicts the hot rod axial power shape used to perform the small break LOCA analysis.This shape was chosen because it represents a distribution with power concentrated in the upper regions of the core.Such a distribution is limiting for small-break LOCAs because it minimizes coolant level swell, while maximizing vapor superheating and fuel'rodheat generation at the uncovered elevations.
The small break LOCA analysis assumes the core continues to operate at full power until the control rods are completely inserted.Results'his section presents results of the limiting smaH break LOCA analysis (as determined by the highest calculated peak clad temperature) for a range of break sizes and RCS pressures and temperatures.
The limiting break was found to be a 4-inch diameter cold leg brcak initiated at
reduced RCS pressurizer pressure (2100 psia)and high temperature (core Tavg=581.3 OF)~~conditions.
The peak clad temperature attained during the transient was 1357 F.A list of input assumptions used in the low pressure and high temperature analysis is provided in Table C.3.2-1.The results of a three break spectrum analysis performed at the reduced RCS pressure and high temperature conditions are summarized in Table C3.24, while the key transient event times are listed in Table C3.2-2.Figures C.32-2 through 9 show for the limiting four-inch break transient, respectively:
RCS pressure Core mixture level Peak clad temperature Core outlet steam Qow rate Hot spot rod surface heat transfer coefficient Hot spot Quid temperature Cold leg break mass Qow rate Safety'injection mass Qow rate During the initial period of the small-break transient the effect of the break flow rate is not strong enough to overcome the Qow rate maintained by the reactor coolant pumps as they coast down.Normal upward Qow is maintained through the core and core heat is adequately removed.At the low heat generation rates following reactor trip the fuel rods continue to be well cooled as long as the core is covered by a two-phase mixture level.From the clad temperature transient for the 4-inch break calculation shown in Figure C3.2-4,'it is seen'hat the peak clad temperature occurs near the time when the core is most deeply uncovered and the top of the core is steam cooled.This time is also accompanied by the highest vapor superheating above the mixture level.A comparison of the total break Qow rate to cont'ainment shown in Figure C.3.2-8 to the safety injection flow rate shown in Figure C9.2/shows that at the time the transient w'as terminated, either when the safety injection Qow rate that was delivered to the RCS exceeded the mass flow rate out the break or when the core was covered as in Figure C3.2-20.Although the inner vessel core mixture level has not yet covered the entire core, there is no longer a concern of exceeding the 10 CFR 50A6 criteria since the pressure is gradually decaying and there is a net mass inventory gain.As the RCS inventory continues to gradually increase, the core mixture level will continue to increase and the fuel clad temperatures will continue to decline.
Conclusions Analyses presented in this section show that the high head charging and safety injection subsystems of the Emergency Core Cooling System, together with the accumulators, provide sufficient core Qooding to keep the calculated peak clad temperatures below the required limit of 10 CFR 50.46.Hence adequate protection is afforded by the Emergency Core Cooling System in the event of a smail break loss-of~lant accident.Additional Break Cases Studies documented in Reference 3 determined that the limiting small-break size occurred for breaks less than 10 inches in diameter.To insure that the 4-inch diameter break was limiting, calculations were run with breaks of 3 inches and 6 inches.The results of these calculations are shown in the Sequence of Events Table C.3.2-2, and the Results Table C.3.2-4.Plots of the following parameters are shown in Figures C3.2-11 through 18 for the 3-inch break, and Figures C.3.2-19 through 26 for the 6-inch break.RCS pressure.Core mixture level Peak clad temperature Core outlet steam Qow rate Hot spot rod surface heat transfer coefficient Cold leg break mass Qow rate Safety injection mass Qow rate As seen in Table C3.2Q the peak clad temperatures were calculated to be less than that for the 4-inch break.Additional Anal is Calculations were also performed for Cook Nuclear Plant Unit 2 with the NOTRUMP()(l)and LOCTA-IV()codes to examine the inQuence of initial RCS coolant operating temperatures and operating pressures on small break LOCA peak clad temperature.
The analyses performed demonstrated that the reduced pressure and high temperature conditions analyzed resulted in the limiting PCI'or the 4-inch diameter break.
To support operation of Cook Nuclear Plant Unit 2 at RCS pressures of 2100 psia and 2250 psia for a range of loop operating temperatures, two additional analyses were performed.
Calculations were performed for a four-inch diameter break for an initial RCS pressurizer pressure of 2250 psia at initial RCS coolant operating temperatures corresponding to core Tavg program setpoints of 5813~F and at a Tavg of 547.0 F.The results of these calculations are shown in the Sequence of Events Table C.3.2-3, and the Results Table C.3.2-5.Plots of the following parameters are shown in Figures C3.2-27 through 34 for the high pressure, high temperature case, and Figures C3.2-35 through 42 for the high pressure, low temperature case.RCS pressure Core mixture level, Peak clad temperature, Core outlet steam flow rate, Hot spot rod surface heat transfer coefficient,-Hot spot Quid temperature, Cold leg break mass Qow rate, and Safety injection mass Qow rate.As seen in Table C3.2-5, the peak clad temperatures were calculated to be less than that.for the 4-inch break initiated at reduced pressure and high temperature conditions.
Additional calculations were made to support closure of the high head safety injection cross tie valves.Since the amount of pumped injection Qow would be reduced with the high head cross tie valves closed, it was necessary to lower core power in order to maintain the peak clad temperature within the 10 CFR SOA6 limit.Thus the calculation which supports plant operation with the high head cross tie valves closed assumed an initial RCS pressurizer pressure at 2100 psia and core Tavg at 581.3 F at a core power level of 3413 Mwt.This calculation also assumed a charging system Qow rate imbalance of 25 gpm.The assumed pumped ECCS Qow performance for the high head cross tie valve closed case is listed in Table C.3.2-9.
Past analyses have shown that a reduction in pumped safety injection flow rate increases the peak clad temperature for smaller breaks (3 inches)more than larger small breaks (4 and 6 inches).An important parameter in determining what will be the limiting break size is the reactor power to safety injection flow rate ratio.For the high head cross tie closure case the reactor power to safety injection flow rate ratio was reduced which shifted the limiting break size to the 3-inch diameter cold leg break.Evidence of this effect is the Cook Nuclear Plant Unit 1 small break LOCA analysis which was performed with the high head cross tie valves closed assuming a charging system flow rate imbalance of 10 gpm with a reactor power of 3588 Mwt.The Cook Nuclear Plant Unit 1 small break LOCA analysis had a reactor power to safety injection flow rate ratio approximately equal to Cook Nuclear Plant Unit 2 with the high head cross tie valves closed assuming 25 gpm charging system fiow rate imbalance at a reactor power level of 3413 Mwt.It was concluded that with the high head cross tie valves closed and with reduced reactor power the limiting break would be shifted from the 4-inch diameter cold leg break to the 3-inch diameter break size.To verify this conclusion, two calculations were performed which assumed break sizes of 3-and 4-inch diameters at the reduced pressure, high temperature initial conditions.
Table C.3.24 lists the results of the cross tie closed cases which show that with the reduced snl'ety injection flow the 3-inch diameter break is limiting.The sequence'f events for these calculations is listed in Table C.3.2-7.Past small break LOCA analyses that were performed for plants which are similar to Cook Nuclear Plant Unit 2 but have power to safety injection flow rate ratios less than that of Cook Nuclear Plant Unit 2, have shown that an assumed break size of 2 inches did not result in the limiting peak clad temperature.
Thus, based on the comparision of power to safety injection low rate ratio, it was concluded that a 2-inch diameter break would not yield a peak clad temperature more limiting that that of the 3-inch diameter break size.Plots for the 3-and 4-inch break with the HHSI cross tie valves closed are shown in flgures C.3.2-43 through C.3.2-50 and C.3.2-51 through C.3.2-58 respectively.
NUREG4737(13),Section II.K3.31, required plant-speciflc small break LOCA analysis using an Evaluation Model revised per Section ILK.3.30.In accordance with NRC Generic Let tcr 83-65(4), generic analyses using NOTRUMP()()were performed and are presented in WCAP-11145().Those results demonstrate that in a comparison of cold leg, hot leg and pump suction Icg break locations, the cold leg break location is limiting.C-139
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ML17332A790: Difference between revisions

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Contents

Text

2.0 SMARTY

5.0 DESIGN

SUMMARY

1.0 INTRODUCTION

2.3 ControVProtection

3.5 Containment

4.0 CONCLUSION

SUMMARY

SUMMARY

2.1 DESIGN

1.0 INTRODUCTION

SUMMARY

SUMMARY

SUMMARY

SUMMARY

1.2 CURRENT

2.0 BASIS

3.0 SAFETY

3.1.2 Small

1.0 since

Subject:

SUMMARY

SUMMARY

3.5 CONTAINMENT

3.6 STEAM

3.8 REACTOR

3.9 RADIOLOGICAL

SUMMARY

4.0 CONCLUSION

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

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ML17332A790: Difference between revisions

Revision as of 06:36, 26 April 2019

Text

2.0 SMARTY

5.0 DESIGN

SUMMARY

1.0 INTRODUCTION

2.3 ControVProtection

3.5 Containment

4.0 CONCLUSION

SUMMARY

SUMMARY

2.1 DESIGN

1.0 INTRODUCTION

SUMMARY

SUMMARY

SUMMARY

SUMMARY

1.2 CURRENT

2.0 BASIS

3.0 SAFETY

3.1.2 Small

1.0 since

Subject:

SUMMARY

SUMMARY

3.5 CONTAINMENT

3.6 STEAM

3.8 REACTOR

3.9 RADIOLOGICAL

SUMMARY

4.0 CONCLUSION

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

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