Text

Proposed Tech Specs Supporting Cycle 10 Reload.Proprietary Safety Evaluation Withheld
	ML17324B291
Person / Time
Site:	Cook
Issue date:	03/26/1987
From:	; INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	;
Shared Package
ML17324B290	List: ML17324B289; ML17324B291;
References
	NUDOCS 8704010071
	Download: ML17324B291 (514)
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DEFINITIONS MEMBER S OF THE PUBLIC 1.35 MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors or its vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the plant.

SITE BOUNDARY 1.36 The SITE BOUNDARY shall be that line beyond which the land is not owned, leased or otherwise controlled by the licensee.

UNRESTRICTED AREA 1.37 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY to which access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or any area within the site boundary used for residential quarters or industrial, commercial, institutional and/or recreational purposes.

DFSIGN THERMAL POWER 1.38 DESIGN THERMAL POWER shall be a design total reactor core heat transfer rate to the reactor coolant of 3411 MWt. See Table 1.3.

ALLOWABLE POWER LEVEL APL 1.39 APL means "allowable power level" which is that power level, less than or equal to 100% RATED THERMAL POWER, at which the plant may be operated to ensure that power distribution limits are satisfied.

D. C. COOK - UNIT 1 1-7 AMENDMENT NO.

,$ 704010071 870326 PDR,'DOCK. ' 05000315 P, . '.PDR

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (T ) shall not exceed the limits shown in Figure 2.1-1 for 4 loop operate.5n.

APPLICABILITY: MODES 1 and 2.

ACTION:

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

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

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

ACTION:

MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour.

MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce -the Reactor Coolant System pressure to within its limit within 5 minutes.

D..C. COOK - UNIT 1 2-1 AMENDMENT NO.

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS This page intentionally left blank.

D. C. COOK - UNIT 1 2-3 AMENDMENT NO.

A O

O TABLE 2.2-1 Continued I

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATIONS Continued Operation with 4 Loops K = 1.135 1

K = 0.0130 2

K = 0.000659 I 3 and f (hI) is a function of the indicated difference between top and bottom detectors of 00 the phwer-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that,:

For q - q between -37 percent and +2 percent, f (a'Z)=0 (where q and q are percent DESIGN THERMAL POWER in the top and bhttom halves of She cor8 respectively, and q + q is total THERMAL POWER in percent of DESIGN THERMAL POWER).

(ii) For each percent that the magnitude of (q q ) exceeds -37 percent, the hT trip setpoint shall be automatically reduced b) 2.3 percent ot its value at DESIGN THERMAL POWER.

(iii) For each percent that the magnitude of (q q ) exceeds +2 percent, the aT O trip setpoint shall be automatically reduced bf 1.8 percent of its value at DESIGN THERMAL POWER.

TABLE 2.2-1 Continued O

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS O

I NOTATION Continued T3S Note 2: Overpower bT < ~T 0 [K -K T K6 (T-T' AI) ]

2 "3'here:

AT 0 Extrapolated aT at DESIGN THERMAL POWER Average temperature, 0 F

Indicated T avg at DESIGN THERMAL POWER 577.1 F K4 1.089 K5 0.0177/ F for increasing average temperature and 0 for decreasing average temperature K6 0.0011 for T > T"; K 6

= 0 for T < T" 1+ <3S The function generated by the rate lag controller for T dynamic compensation T3 Time constant utilized in the rate lag controller for T

= 10 secs.

3 Laplace transform operator o f2(b,I) = fl (DI) as defined in Note 1 above.

Note 3: The channel's maximum trip point shall not exceed its computed trip point by more than 2.5 percent hT span.

Note 4: The channel's maximum trip point shall not exceed its computed trip point by more than 3.4 percent ~T span.

3 4 1 REACTIVITY CONTROL SYSTEMS 3 4 1 1 BORATION CONTROL SHUTDOWN MARGIN - STANDBY STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.60% Ak/k.

APPLICABILITY: MODES 1, 2*, and 3.

ACTION:

With the SHUTDOWN MARGIN less than 1.60% Ak/k, immediately initiate and continue boration at greater than or equal to 10 gpm of 20,000 ppm boric acid solution 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 1.60% Ak/k:

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 increased by an amount at least equal to the withdrawn worth of the immovable or untrippable control rod(s).

b. When in MODES 1 or 2

¹ , 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 , at least once during control rod withdrawal and at least once per hour thereafter until the reactor is critical.

d. Prior to initial operation above 5S 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 '

'See Special Test Exception 3.10.1 With K ff greater than or eff equal to 1.0

¹¹With K ff less than 1.0 eff D. C. COOK - UNIT 1 AMENDMENT NO.

REACTIVITY CONTROL SYSTEMS SURVEILLANCE RE UIREMENTS Continued

e. When in MODE 3, at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

1. Reactor coolant system boron concentration,

2. Control rod position,

3. Reactor coolant system average temperature, 4 . Fuel burnup based on gross thermal energy generation,

5. Xenon concentration, and

6. Samarium concentration, 4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within + 1% gk/k at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least those factors stated in Specification 4.1 1.1;l.e, above. The predicted

~

reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days after each fuel loading.

D. C. COOK - UNIT 1 3/4 1-2 AMENDMENT NO.

REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - SHUTDOWN LIMITING CONDITION FOR OPERATION 3,1.1.2 The SHUTDOWN MARGIN shall be:

a. In MODE 4:

1. Greater than or equal to 1.6% A k/k when operating with one or more Reactor Coolant, Loops in accordance with Specification 3.4.1.3.

2. Greater than the value shown in Figure 3 '-3 when operating with no Reactor Coolant Loops but one or more Residual Heat Removal Loops in accordance with Specification 3.4.1.3.

b. In MODE 5:

1. Greater than or equal to 1.08 Ak/k when operating with one or more Reactor Coolant Loops in accordance with Specification 3.4.1.3.

2. Greater than the value shown in Figure 3.1-3 when operating with no Reactor Coolant Loops but one or more Residual Heat Removal Loops in accordance with Specification 3.4.1.3.

APPLICABILITY: MODES 4 and 5 ACTION:

With SHUTDOWN MARGIN less than the above limits, immediately initiate and continue boration at greater than or equal to 10 gpm of 20,000 ppm boric acid solution or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE RE UIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to the above limits:

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 SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable control rod(s).

DE C. COOK - UNIT 1 3/4 1-3 AMENDMENT NO.

REACTIVITY CONTROL SYSTEMS P

SURVEILLANCE RE UIREMENTS Continued

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

1. Reactor coolant system boron concentration,.

2. Control rod position,

3. Reactor coolant system average temperature,

4. Fuel burnup based on gross thermal energy generation, 5, Xenon concentration, and

6. Samarium concentration.

D. C. COOK - UNIT 1 3/4 1-3a AMENDMENT NO.

Cl 0O 5.00 O

I Q 0PE T ON~>

4.00 A I

$ ~f O, D,E Bc5 t-I I

I 3.00 g <-'-g"I-g t

L j. l.

r-p 3

2. 00 QQOO(EE A 1 +t C4 l ACCFPABLE %/~RAT CY L

1.00 L 466 O

0 .00 C +Z t 0 200 400 600 800 1,000 1,200 1,400 1,600 1,800 2,000 BORON CONCENTRATION (PPM)

Figure 3.1-3 REQUIRED SHUTDOWN MARGIN

REACTIVITY CONTROL SYSTEMS BORON DILUTION LIMITING CONDITION FOR OPERATION 3.1.1.3 The flow rate of reactor coolant through the reactor coolant system shall be greater than or equal to 2000 gpm whenever a reduction in Reactor Coolant System boron concentration is being made.*

APPLICABILITY: ALL MODES .

ACTION:

With the flow rate of reactor coolant through the reactor coolant system less than 2000 gpm, immediately suspend all operations involving a reduction in boron concentration of the Reactor Coolant System.

SURVEILLANCE RE UIREMENTS 4.1.1.3 The flow rate of reactor coolant through the reactor coolant system shall be determined to be greater than or equal to 2000 gpm within one hour prior to the start of and at least once per hour during a reduction in the Reactor Coolant System boron concentration by either:

a. Verifying at least one reactor coolant pump is in operation, or

b. Verifying that at least one RHR pump is in operation and supplying greater than or equal to 2000 gpm through the reactor coolant system. t

For purposes of this specification, addition of water from the RWST does not constitute a dilution activity provided the boron concentration in the RWST is greater than or equal to the minimum required by specification 3.1.2.8.b.2 (MODES 1, 2, 3, and 4) or 3.1.2.7.b.2 (MODES 5 and 6).

D. C. COOK - UNIT 1 3/4 1-4 AMENDMENT NO.

REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIEN LIMITING CONDITION FOR OPERATION 3.1.1.4 The moderator temperature coefficient (MTC) shall be:

a. Within the region of acceptable operation in Figure 3.1-2, and

b. Less negative than -3.5 x 10 Ak/k/ F at RATED THERMAL POWER.

APPLICABILITY: MODES 1 and 2*¹ ACTION With the moderator temperature coefficient outside any one of the above limits, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE RE UIREMENTS 4.1.1.4.1 The MTC shall be determined to be within its limits by confirmatory measurements. MTC measured values shall be extrapolated and/or compensated to permit direct comparison with the above limits.

4.1.1.4.2 The MTC shall be determined at the following frequencies and THERMAL POWER conditions during each fuel cycle:

a. Prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading.

b. At any THERMAL POWER, within 7 EFPD after reaching a RATED THERMAL POWER equilibrium boron concentration of 300 ppm.

With Keff greater than or equal to 1.0 See Special Test Exception 3.10.4 D. C. COOK - UNIT 1 3/4 1-5 AMENDMENT NO ~

FIGURE 3.'I 2 Moderator Temperature Coefficient (MTC)

MTC x 10+ Ak/kjdeg.F 1.00 Unac ceptabl Oper tion 0.50 0.00

-0.50

-1.00 Acc ptable Opera on

-1.50

-2.00

-2.50

-3.00 10 20 30 40 50 60 70 80 90 100 PERCENT RATED THERMAL POWER

REACTIVITY CONTROL SYSTEMS 3 4 1 2 BORATION SYSTEMS FLOW PATHS - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.1 As a minimum, one of the following boron injection flow paths shall be OPERABLE:

a. A flow path from the boric acid tanks via a boric acid transfer pump and charging pump to the'Reactor Coolant System if acid storage tank in Specification 3.1.2.7a is OPERABLE, or only the boric

b. The flow path from the refueling water storage tank via a charging pump to the Reactor Coolant System if only the refueling water storage tank in Specification 3.1.2.7b is OPERABLE.

APPLICABILITY: MODES 5 and 6.

ACTION:

With none of the above flow paths OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes* until at least one injection path is restored to OPERABLE status.

SURVEILLANCE RE UIREMENTS 4.1.2.1 At least one of the above required flow paths shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1. Cycling each testable power operated or automatic valve in the flow path through at least one complete cycle of full travel.

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.

D. C. COOK - UNIT 1 3/4 1-7 AMENDMENT NO.

I REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.,1.2.3 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.

APPLICABILITY: MODES 5 and 6.

ACTION

a. With no charging pump OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.*

b. With more than one charging pump OPERABLE or with a safety injection pump(s) OPERABLE0 when the temperature of any RCS cold leg is less than or equal to 170 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.

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 discharge pressure of greater than or equal to 2390 psig when tested pursuant to Specification 4.0.5 at least once per 31 days.

4.1.2.3.2 All charging pumps and safety injection pumps, excluding the above required OPERABLE charging pump, shall be demonstrated inoperable by verifying that the motor circuit breakers have been removed from their electrical power supply circuits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , except when:

a. The reactor vessel head is removed, or

b. The temperature of all RCS cold legs is greater than 0 170 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.

D. C. COOK - UNIT 1 3/4 1-11 AMENDMENT NO.

REACTIVITY CONTROL SYSTEMS BORIC ACID TRANSFER PUMPS - SHUTDOWN LIM T NG C NDI ION FOR OPERATION 3,1.2.5 At least one boric acid transfer pump shall be OPERABLE and capable of being powered from an OPERABLE emergency bus if only the flow path through the boric acid transfer pump of Specification 3.1.2.1a is OPERABLE.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no boric acid transfer pump OPERABLE as required to complete the flow path of Specification 3.1.2.la, suspend all operations involving CORE ALTERATIONS or positive reactivity changes* until at least one boric acid transfer pump is restored to OPERABLE status.

SURVEILLANCE RE UIREMENTS 4.1.2.5 At least the above required boric acid transfer pump shall be demonstrated OPERABLE at least once per 7 days by:

a. Starting (unless already operating) the pump from the control room,

b. Verifying, that on recirculation flow, the pump develops a discharge pressure of greater than or equal to 110 psig,

c. Verifying pump operation for at least 15 minutes, and

d. Verifying that the pump is aligned to receive electrical power from an OPERABLE emergency bus.

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.

D. C. COOK - UNIT 1 3/4 1-13 AMENDMENT NO.

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. 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 0 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 80 F.

APPLICABILITY: MODES 5 and 6.

ACTION'ith no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS 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 shall be demonstrated OPERABLE:

At least once per 7 days by:

1. Verifying the boron concentration of the water,

2. Verifying the water level volume of the tank, and

3. 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 is the source of borated water.

RWST temperature when it I

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.

D. C. COOK - UNIT 1 3/4 1-15 AMENDMENT NO.

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.8 Each 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 5650 gallons,

2. Between 20,000 and 22,500 ppm of boron, and 0

3. A minimum solution temperature of 145 F.

b. 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 80 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% Ak/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 SHUT-DOWN 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 ~ater source shall be demonstrated OPERABLE:

D. C. COOK - UNIT 1 3/4 1-16 AMENDMENT NO.

REACTIVITY CONTROL SYSTEMS SURVEILLANCE RE UIREMENTS Continued a ~ At least once per 7 days by:

1. Verifying the boron concentration in each water source,

2. Verifying the water level of each water source, and

3. Verifying the boric acid storage system solution temperature.

b. At least once per 24 hours by verifying the RUST temperature.

D. C. COOK - UNIT 1 3/4 1-17 AMENDMENT NO.

REACTIVITY CONTROL SYSTEMS 3 4. 1 3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3 '.3.1 All full length (shutdown and control) rods shall be OPERABLE and positioned within + 12 steps (indicated position) of their group step counter demand position.

APPLICABILITY: MODES 1* and 2*

ACTION

a. With one or more full length rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With more than one full length rod inoperable or misaligned from the group step counter demand position by more than + 12 steps (indicated position), be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With one full length rod inoperable due to causes other than addressed by ACTION a, above, or misaligned from its group step counter demand height by more than + 12 steps (indicated position),

POWER OPERATION may continue provided that within one hour either:

1. The affected rod is restored to OPERABLE status within the above alignment requirements, or

2. The affected rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that:

a) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions, and b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and

See Special Test Exceptions 3.10.2 and 3.10.4 D. C. COOK - UNIT 1 3/4 1-18 AMENDMENT NO ~

REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION Continued c) A power distribution map $ s obtained from the movable incore detectors and F (Z) and F" are verified to be within their limits within 79 hours9.143519e-4 days 0.0219 hours 1.306217e-4 weeks 3.00595e-5 months , an5 d) Either the THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within one hour and within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the high neutron flux trip stepoint is re-duced to less than or equal to 85% of RATED THERMAL POWER, or e) The remainder of the rods in the group with the inoperable I rod are aligned to within + 12 steps of the inoperable rod within one hour while maintaining the rod sequence and inser-tion limits of Figure 3 '-1; the THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.5 during sub-sequent operation.

SURVEILLANCE RE UIREMENTS 4.1.3.1.1 The position of each full length rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Rod Position Deviation Monitor is inoperable, then verify the group positions at least bourse'.1.3.1.2 once per 4 Each full length rod not fully inserted shall be determined to be OPERABLE by movement of at least 8 steps in any one direction at least once per 31 days.

D. C. COOK - UNIT 1 3/4 1-19 AMENDMENT NO.

REACTIVITY CONTROL SYSTEMS TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL LENGTH ROD Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Misalignment Loss of Reactor Coolant From Small Ruptured Pipes Or From Cracks In Large Pipes Which Actuates The Emergency Core Cooling System Single Rod Cluster Control Assembly Withdrawal At Full Power Major Reactor Coolant System Pipe Ruptures (Loss of Coolant Accident)

Major Secondary System Pipe Rupture Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection)

D. C. COOK - UNIT 1 3/4 1-19a AMENDMENT NO.

REACTIVITY CONTROL SYSTEMS ROD DROP TIME LIMITING CONDITION FOR OPERATIO 3.1.3.3 The individual full length (shutdown and control) rod drop time from the fully withdrawn position (228 steps) shall be less than or equal to 2.4 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:

a. T greater than or equal to 541 0 F, and avg

b. All reactor coolant pumps operating.

APPLICABILITY: MODES 1 and 2.

ACTION:

With the drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

SURVEILLANCE RE UIREMENTS 4.1.3.3 The rod drop time of full length rods shall be demonstrated through measurement prior to entering MODE 2:

a. For all rods following each removal of the reactor vessel head,

b. For specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and

c. At least once per 18 months.

D. C. COOK - UNIT 1 3/4 1-21 AMENDMENT NO ~

REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.4 All shutdown rods shall be fully withdrawn (228 steps).

APPLICABILITY:, MODES 1* and 2*¹ ACTION:

With a maximum of one shutdown rod not fully withdrawn, except for surveil-lance testing pursuant to Specification 4.1.3.1.2, within one hour either:

a. Fully withdraw the rod, or

b. Declare the rod to be inoperable and apply Specification 3.1.3.1 ~

SURVEILLANCE RE UIREMENTS 4.1.3.4 Each shutdown rod shall be determined to be fully withdrawn:

a. Within 15 minutes prior to withdrawal of any rods in control banks A, B, C or D during an approach to reactor criticality, and

b. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

See Special Test Exceptions 3.10.2 and 3.10.4

¹With Kefff greater than or equal to 1.0 DE C. COOK - UNIT 1 3/4 1-22 AMENDMENT NO.

REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.5 The control banks shall be limited in physical insertion as shown in Figure 3.1-1.

APPLICABILITY: MODES 1

and *¹ 2

ACTION:

With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, either:

a. Restore the control banks to within the limits within two hours, or

b. Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the group position using the above figure, or

c. Be in HOT STANDBY within 6 hours.

SURVEILLANCE RE UIREMENTS 4.1.3.5 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

See Special Test Exceptions 3.10.2 and 3.10.4

¹

~

With K eff greater than or equal to 1.0.

D. C. COOK - UNIT 1 3/4 1-23 AMENDMENT NO.

(FULI.Y WITHDRAWN)

(0.66,228) 228 ~ ~

200 BANK C 82):"

X 0

150 I:

I ~

I-(0, 1 28)

CO

a. 100

~ o O

'BANK II D

0 50 I ~

~-

0 0.0 0.2 0.4 0.6 0.8 1.0 (FULLY INSERTED) FRACTION OF RATED THERMAL POWER (3250 Nlt)

FIGURE 3.1-1 ROD GROUP INSERTION LIMITS YERSUS THERMAL POWER 4 LOOP OPERATION D.'. COOK UNIT 1 0/4 1-24 ANENDNEVaZ NO.

REMOVE THIS PAGE D. C. COOK - UNIT 1 3/4 1-25 AMENDMENT NO.

REMOVE THIS PAGE D. C. COOK - UNIT 1 3/4 1-26 AMENDMENT NO.

3 4.2 POWER DISTRIBUTION LIMITS AXIAL FLUX DIFFERENCE AFD LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the target band (+5% or +3% flux difference units) about a target flux difference.

APPLICABILITY: MODE 1 above 50% RATED THERMAL POWER>>

ACTION:

With the indicated AXIAL FLUX DIFFERENCE outside of the target band about the target flux difference and with THERMAL POWER:

1. Above 90% or 0.9 x APL (whichever is less) of RATED THERMAL POWER, within 15 minutes:

a) Either restore the indicated AFD to within the target band limits, or b) Reduce THERMAL POWER to less than 90% or 0.9 x APL (whichever is less) of RATED THERMAL POWER.

2. Between 50% and 90% or 0' x APL (whichever is less) of RATED THERMAL POWER:

a) POWER OPERATION may continue provided:

1) The indicated AFD has not been outside of the target band for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months penalty deviation cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and

2) The i.ndicated AFD is within the limits shown on Figure 3.2-1. Otherwise, reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b) Surveillance testing of the Power Range Neutron Flux Channels may be performed pursuant to Specification 4.3.1.1.1 provided the indicated AFD is maintained within the limits of Figure 3.2-1. A total of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months operation may be accumulated with the AFD outside of the target band during this testing without penalty deviation.

+See Special Test Exception 3.10.2 D. C. COOK-UNIT 1 3/4 2-1 AMENDMENT NO.

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION Continued THERMAL POWER shall not be increased above 90% or 0.9 x APL (whichever is less) of RATED THERMAL POWER unless the indicated AFD is within the target band and ACTION 2.a) 1), above has been satisfied.

THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD has not been outside of the target band for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months penalty deviation cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE RE UIREHENTS 4.2.1.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:

Monitoring the indicated AFD for each OPERABLE excore channel:

1. At least once per 7 days when the AFD Honitor Alarm is OPERABLE, and

2. At least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring the AFD Honitor Alarm to OPERABLE status.

Monitoring and logging the indicated AXIAL FLUX DIFFERENCE for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereafter, when the AXIAL FLUX DIFFERENCE Honitor Alarm is inoperable.

The logged values of the indicated AXIAL FLUX DIFFERENCE shall be assumed to exist during the interval preceding each logging.

D. C. COOK - UNIT 1 3/4 2-2. AMENDMENT NO.

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS Continued 4.2.1.2 The indicated AFD shall be considered outside of its target band when at least 2 of 4 or 2 of 3 OPERABLE excore channels are indicating the AFD to be outside the target band. Penalty deviation outside of the target band shall be accumulated on a time, basis of:

a. A penalty deviation of one minute for each one minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and

b. A penalty deviation of one-half minute for each one minute of POWER OPERATION outside of the target band at THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER.

4.2.1.3 The target axial flux difference of each OPERABLE excore channel shall be determined in conjunction with the measurement of APL as defined in Specification 4.2.6.2. The provisions of Specification 4 '.4 are not applicable.

4.2.1.4 The axial flux difference target band about the target axial flux difference shall be determined in conjunction with the measurement of APL as defined in Specification 4.2.6.2. The allowable values of the target band are + 5$ or + 3S. Redefinition of the target band from + 3% to

+ 5S between determinations of the target axial flux difference is allowed when appropriate redefinitions of APL are made. Redefinition of the target band from + 5% to + 3% is allowed only in conjunction with the determination of a new target axial flux difference. The provisions of Specification 4.0.4 are not applicable.

D. C. COOK - UNIT 1 3/4 2-3 AMENDMENT NO.

FIGURE 3.2 1 ALLOV/AEILE DEVIATION FROM TARGET FLUX DIFFERENCE O

O O

O I

m1 00 A

Unacceptable (-1o,eo) (+" 0 90) Unacceptable

" Operation;, Operation x 80

+3% Target

( 8.9o},

I

(+ago} .,> +5% Target Band T 4 Acceptable Operation

~o 60 4

(-z5,5o) 'Q(+25.50)

+(-23,50) (+Z3,50) 4 hl 40 -=-

8 20 "- -- ==

I 8

R pl t 4

4 l

a I I

II W 0 i l l i I 0

aZ -30 20 10 0 10 20 30 O

Deviation from Target Flux Difference

POWER DISTRIBUTION LIMITS HEAT FLUX HOT CHANNEL FACTOR-F Z LIMITING CONDITION FOR OPERATION 3.2.2 F (Z) shall be limited by the following relationships:

~

Westin house Fuel Exxon Nuclear Co Fuel Fq(Z> < ~p [K(Z>[ F Q

(Z) < P

[K(Z)[ F ) 0.5 F (Z) < [4.20] [K(Z)] F (Z) < [4.08] [K(Z)] P < 0.5 THERMAL POWER RATED THERMAL POWER

~ F (Z) is the measured hot channel factor including a 3%

m9nufacturing tolerance uncertainty and a 5% measurement uncertainty.

< K(Z) is the function obtained from Figure 3.2-3 for Westinghouse fuel and Figure 3.2-2 for Exxon Nuclear Co. fuel.

APPLICABILITY: MODE 1 ACTION:

With F (Z) exceeding its limit:

Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds the limit within 15 minutes and similarly reduce t9e Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1% for each 1% F (Z) exceeds the limit.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a, above; THERMAL POWER may then be increased provided F (Z) is demonstrated through incore mapping to be within its limit.

D. C. COOK - UNIT 1 3/4 2-5 AMENDMENT NO.

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 F (Z) shall be determined to be within its limit above 5% of RATED THERi4L POWER according to the following schedule:

a. Whenever F (Z) is measured for reasons other than meeting the requiremen9 of 4.2.6.2, or

b. At least once per 31 effective full power days, whichever occurs first.

D. C. COOK - UNIT 1 3/4 2-6 AMENDMENT NO.

FIGURE 3.2 2 EXXON FUEL oo K(Z) NORMALIZED VS. CORE HEIGHT K(Z) NORMALIZED F Q (Z)

I 1.20 t-Q (o.o,1.o) (6.0,1.0) 1.00 (11.01, .936) 0.80-0.60-(12.0, .4902)

-040 .

0.20 I 0.00 L I I .I 0 4 6 8 10 12 o CORE HEIGHT (FT.)

FIGURE 3.2 3 MlESTINGHOUSE FUEL A K(Z) NORMALIZED VS. CORE HEIGHT O

Q K(Z) NORMALIZEO FQ {Z)

I 'l.20 r-(o.o 1.o) (6.0,1.0) 1.00 (1 1.1 83,0.935)

II 0.80 ~ '4 (12.0,0.714) 0.60 "=

I 00 040 --

II 0.20 ~

I 1'

t 0.00 L I

-1 I.

0 6 8 10 o CORE HEIGHT (FT.)

POWER DISTRIBUTION LIMITS NUCLEAR ENTHALPY HOT CHANNEL FACTOR - F LIMITING CONDITION FOR OPERATION 3.2.3 F H

shall be limited by the following relationships:

P~H < 1.49 [1 + 0.3 (1-P)] (for Westinghouse fuel) and F H

< 1.45 [1 + 0.2 (1-P)) (for Exxon Nuclear Co. fuel) where P is the fraction of RATED THERMAL POWER APPLICABILITY: MODE 1 ACTION' With F<H exceeding limit:

a. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,

b. Demonstrate through in-core mapping that F is within its limit within 24 hours after exceeding the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL gOWER; subsequent POWER OPERATION may proceed, provided that F" is demonstrated through in-core mapping to be within its limR at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL POWER and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after attaining 95% or greater RATED THERMAL POWER.

D. C. COOK - UNIT 1 3/4 2-9 AMENDMENT NO.

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS 4.2.3 F shall be determined to be within its limit by using the movable incore detectors to obtain a power distribution map:

a. Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and \

b. At least once per 31 Effective Full Power Days.

c. The provisions of Specification 4.0.4 are not applicable.

D. C. COOK - UNIT 1 3/4 2-10 AMENDMENT NO.

POWER DISTRIBUTION LIMITS UADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION 3.2.4 THE QUADRANT POWER TILT RATIO shall not exceed 1,02 APPLICABILITY: MODE 1 ABOVE 50% OF RATED THERMAL POWER+

ACTION:

a. With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but less than or equal to 1.09:

,1. Within 2 hours:

a) Either reduce the QUADRANT POWER TILT RATIO to within its limit, or b) Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.0 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

2. Verify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours and reduce the Power Range Neutron Flux-High Trip set-points to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour until verified acceptable at 95$ or greater RATED THERMAL POWER.

b. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod:

1. Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.0, within 30 minutes.

2. Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after exceeding the limit or

See Special Test Exception 3.10.2 D. C. COOK - UNIT 1 3/4 2-11 AMENDMENT NO.

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION Continued reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,

3. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour until verified acceptable at 95% or greater RATED THERMAL POWER.

c. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either a shutdown or control rod:

1. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 558 of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

2. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour until verified at 958 or greater RATED THERMAL POWER.

SURVEILLANCE RE UIREMENTS 4.2.4 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:

a ~ Calculating the ratio at least once per 7 days when the alarm is OPERABLE.

b. Calculating the ratio at least once per 12 hours during steady state operation when the alarm is inoperable.

c, Using the movable incore detectors to determine the QUADRANT POWER TILT RATIO at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when one Power Range Channel is inoperable and THERMAL POWER is greater than 75 percent of RATED THERMAL POWER.

D. C. COOK - UNIT 1 3/4 2-12 AMENDMENT NO.

POWER DISTRIBUTION LIMITS DNB PARAMETERS LIMITING COND TION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-1:

a. Reactor Coolant System T avg

b. Pressurizer Pressure

c. Reactor Coolant System Total Flow Rate ACTION:

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

SURVEILLANCE RE UIREMENTS 4.2.5.1 Each of the parameters of Table 3.2-1 shall be verified to be within their limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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

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

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

D. C. COOK - UNIT 1 3/4 2-13 AMENDMENT NO.

TABLE 3 2-1 DNB PARAMETERS LIMITS 4 Loops In Operation PARAMETER at RATED THERMAL POWER Reactor Coolant System T 570.4 F avg Pressurizer Pressure 2205 psig**

6 Reactor Coolant System 138.6 x 10 lbs/hr***

Total Flow Rate

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.

~

D. C. COOK - UNIT 1

~ 3/4 2-14 AMENDMENT NO.

POWER DISTRIBUTION LIMITS ALLOWABLE POWER LEVEL - APL LIMITING CONDITION FOR OPERATION 3.2.6 THERMAL POWER shall be less than or equal to ALLOWABLE POWER LEVEL (APL), given by the following relationships:

Westin house Fuel APL - min over Z of F 2.10 K Z x 100%, or 100%, whichever is less.

(Z)xV(Z)xF P

Exxon Nuclear Co Fuel APL min over Z of 2.04 K Z x or whichever is less.

100%, 100%,

F (Z)xV(Z)xF "

P

~ F (Z) is the measured hot channel factor including a 3%

m9nufactur'ing tolerance uncertainty and a 5% measurement uncertainty.

~ V(Z) is the function defined in the Peaking Factor Limit Report.

~ F - 1.00 except when successive steady-state power F (Z) dkstribution maps indicate an increase in max over Z of with exposure. Then either of the penalties, F , shall K(Z) be taken: P F - 1.02, or P

F 1.00 provided that Surveillance Requirement 4.2.6.2 is sRtisfied once per 7 Effective Full Power Days until 2 successive maps indicate that the max over Z of F (Z)

K(Z) is not increasind.

~ The above limit is not applicable in the following core regions.

1) Lower core region 0% to 10% inclusive.

2) Upper core region 90$ to 100% inclusive.

APPLICABILITY: MODE 1 DE C. COOK - UNIT 1 3/4 2-15 AMENDMENT NO.

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION Continued ACTION:

With THERMAL POWER exceeding APL Reduce THER1AL POWER to APL or less of RATED THERMAL POWER within 15 minutes. Then reduce the Power Range Neutron Flux-High Trip Setpoints by the same percentage which APL is below RATED THEM'fAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower .5T Trip Setpoints have been reduced the same percentage which APL is below RATED THERMAL POWER.

b. THERMAL POKER may be increased to a new APL calculated at the reduced power by either redefining the target axial flux difference or by correcting the cause of the high F (Z) condition.

SURVEILLANCE RE UIREMENTS 4.2.6.1 The provisions of Specification 4.0.4 are not applicable.

4.2.6.2 APL shall be determined by measurement in conjunction with the target flux difference and target band determination~ above 15% of RATED THERMAL POWER, according to the following schedule:

Upon achieving equilibrium conditions after exceeding by 10% or more of RATED THERiAL POWER, the THERfAL POWER at which APL was last determined~'<~, or li At least once per 31 effective full power days, whichever occurs first.

APL can be redefined by remeasuring the target axial flux difference in accordance with ACTION statement b of Specification 3.2.6.

~'~~'During power escalation at the beginning of each cycle, the design target may be used until a power level for extended operation has been achieved.

D. C. COOK - UNIT 1 3/4 2-16 AMENDMENT NO.

I 1

II A y

REMOVE THIS PAGE D. C. COOK - UNIT 1 3/4 2-17

REMOVE THIS PAGE D. C. COOK - UNIT 1 3/4 2-18

REMOVE THIS PAGE D. C. COOK - UNIT 1 3/4 2-19

REMOVE THIS PAGE D. C. COOK - UNIT 1 3/4 2-20

REMOVE THIS PAGE D. C. COOK - UNIT 1 3/4 2-21

I REMOVE THIS PAGE D. C. COOK - UNIT 1 3/4 2-22

REMOVE THIS PAGE D. C. COOK - UNIT 1 ,3/4 2-23

"I REMOVE THIS PAGE D. C. COOK - UNIT 1 3/4 2-24

INSTRUMENTATION This page intentionally left blank.

D. C. COOK - UNIT 1 3/4 3-2 AMENDMENT NO.

TABLE 3.3-1 O REACTOR TRIP SYSTEM INSTRUMENTATION O

O MINIMUM I

TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

1. Manual Reactor Trip 1, 2 and

12

2. Power Range, Neutron Flux 1, 2 and

2~

3. Power Range, Neutron Flux, 1, 2 High Positive Rate

4. Power Range, Neutron Flux, 1, 2 High Negative Rate

5. Intermediate Range, 1, 2and*

Neutron Flux

6. Source Range, Neutron Flux A. Startup 2~~ and

B. Shutdown 3, 4 and 5

7. Overtemperature hT Four Loop Operation 1, 2 W

eg ap 8. Overpower hT Four Loop Operation

~

Pf P4 1, 2 O

1 ntinued REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION O

O 9. Pressurizer Pressure-Low 1, 2 I

10. Pressurizer Pressure High 1, 2 g

ll. Pressurizer Water Level High 1, 2

12. Loss of Flow Single Loop 3/loop 2/loop in 2/loop in 7¹ (Above P-8) any opera- each opera-ting loop ting loop

13. Loss of Flow Two Loops 3/loop 2/loop in 2/loop in 7¹ (Above P-7 and below P-8) two opera- each opera-ting loops ting loop

14. Steam Generator Water 3/loop 2/loop in 2/loop in 1, Level 2

Low-Low any opera- each opera-ting loop ting loop

15. Steam/Feedwater Flow 2/loop-level 1/loop-level 1/loop-level 1, 2 Mismatch and Low Steam and coincident and Generator Water Level 2/loop-flow with 2/loop-flow mismatch in 1/loop-flow mismatch or same loop mismatch in 2/loop-level same loop and 1/loop-flow mismatch

TABLE 3.3-1 Continued REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO- CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

16. Undervoltage-Reactor Coolant 4-1/bus 1 Pumps

17. Underfrequency-Reactor 4-1/bus Coolant Pumps

18. Turbine Trip A. L'ow Fluid Oil Pressure 70 B. Turbine Stop Valve Closure

19. Safety Injection Input 1, 2 from ESF

20. Reactor Coolant Pump Breaker Position Trip A. Above P-8 1/breaker 1/breaker 1 10 B. Above P-7 and below P-8 1/breaker 1/breaker 1 11 per oper-ating loop

21. Reactor Trip Breakers 1I 2 1, 13 3* 4* 5* 14

22. Automatic Trip Logic 1I 2 1 3* 4* 5* 14

TABLE 3 3-1 Continued TABLE NOTATION

With the reactor trip system breakers in the closed position and the control rod drive system capable of rod withdrawal.

¹ The provisions of Specification 3.0.4 are not applicable.

¹¹ High voltage to detector may be de-energized above P-6.

ACTION STATEMENTS ACTION 1 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1.1.

ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied.

a. The inoperable channel is placed in tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing of the other channels per Specification 4.3.1.1.1. ib

c. Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER and the Power Range, Neutron Flux trip setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months per Specification 4.2.4.c.

ACTION 3 - With the number of channels OPERABLE one less than requi.red by the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

DE C. COOK - UNIT 1 3/4 3-6 AMENDMENT NO.

0 TABLE 3 3-1 Continued ACTION 10 - With one channel inoperable, restore the inoperable channel to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to below P-8 within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Operation below P-8 may continue pursuant to ACTION 11.

ACTION ll - With less than the Minimum Number of Channels OPERABLE, operation may continue provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 12 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-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 and/or open the reactor trip breakers.

ACTION 13 - With one of the diverse trip features (Undervoltage or shunt trip attachment) inoperable, restore it to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the breaker inoperable and apply ACTION 1. The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time required for performing maintenance to restore the breaker to OPERABLE status.

ACTION 14 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the reactor trip breakers within the next hour.

REACTOR TRIP SYSTEM INTERLOCKS DESIGNATION CONDITION AND SETPOINT FUNCTION p-6 With 2 of 2 Intermediate Range P-6 prevents or defeats Neutryy Flux Channels less than the manual block of 6xlO amps. source range reactor trip.

D. C. COOK - UNIT 1 3/4 3-8 AMENDMENT NO.

1 TABLE 3.3-1 Continued DESIGNATION CONDITION AND SETPOINT FUNCTION P-7 With 2 of 4 Power Range Neutron P-7 prevents or defeats Flux Channels greater than or the automatic block of equal to 11% of RATED THERMAL reactor trip on: Low POWER or 1 of 2 Turbine First flow in more than one Stage Pressure channels greater primary coolant loop, than or equal to 37 psig. reactor coolant pump under-voltage and under-frequency, turbine trip, pressurizer low pressure, and pressurizer high levels P-8 With 2 of 4 Power Range Neutron P-8 prevents or defeats Flux channels greater than or the automatic block of equal to 31% of RATED THERMAL reactor trip on low POWER. coolant flow in a single loop.

P-10 With 3 of 4 Power Range Neutron P-10 prevents or defeats Flux channels less than 9S of the manual block of:

RATED THERMAL POWER. Power range low setpoint reactor trip, Intermediate range reactor trip, and intermediate range rod stops.

Provides input to P-7.

D. CD COOK - UNIT 1 3/4 3-9 AMENDMENT NO.

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL . SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST RE UIRED

l. Manual Reactor Trip A. Shunt Trip Function N.A. N.A. S/U(1) (10) 3* 4* 5*

B. Undervoltage Trip Function N.A. N.A. S/U(1) (10) 3* 4* 5k

2. Power Range, Neutron Flux D(2,8),M(3,8) M and S/U(1) 1, 2 and

and Q(6,8)

3. Power Range, Neutron Flux, N.A. R (6) 1, 2 High Positive Rate

4. Power Range, Neutron Flux, N.A. R (6) 1, 2 High Negative Rate

5. Intermediate Range, R(6,8) S/U (1) 1, 2 and*

Neutron Flux

6. Source Range, Neutron Flux R(6,8) M(8) and 2(7) I 3(7) I S/U(1) 4 and 5

7. Overtemperature hT R(9) 1, 2

8. Overpower hT R(9) 1, 2

9. Pressurizer Pressure Low R li 2

10. Pressurizer Pressure High li 2 ll. Pressurizer Water Level High 1/ 2

12. Loss of Flow-Single Loop R(8)

The provisions of Specification 4.0.6 are applicable.

TABLE 4.3-1 Continued REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL MODES IN WHICH A CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE O

O FUNCTIONAL UNIT CHECK CALIBRATION TEST RE UIRED I

13. Loss of Flow-Two Loops R(8) N.A.

14. Steam Generator Water Level- R 1, 2 Low-Low

15. Steam/Feedwater Flow Mismatch and S 1, 2 Low Steam Generator Water Level

16. Undervoltage-Reactor Coolant N.A.

Pumps

17. Underfrequency-Reactor Coolant N.A.

Pumps

18. Turbine Trip A. Low Fluid Oil Pressure N.A. N.A. S/U (1) 1, 2 B. Turbine Stop Valve Closure N.A. N.A. S/U (1) 1, 2

19. Safety Injection Input from ESF N.A. N.A. M(4) 1, 2

20. Reactor Coolant Pump Breaker N.A. N.A. N.A.

Position Trip

21. Reactor Trip Breaker A. Shunt Trip Function N.A. N.A. M(5) (ll) and S/U(l) (ll) 2 3* 4* 5*'

B. Undervoltage Trip Function N.A. N.A. M(5) (ll) and S/U(1) (ll) 3* 4* 5*.

O 22. Automatic Trip Logic N.A. N.A. M(5) 2 3* 4* 5*

23. Reactor Trip Bypass Breaker N.A. N.A. M(12) and S/U (1) (13) 3* 4* 5*

The provisions of Specification 4.0.6 are applicable.

TABLE' 3-1 Continued NOTATION With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.

(1) If not performed in previous 7 days.

(2) Heat balance only, above 15% of RATED THERMAL POWER. Adjust channel if absoluted difference greater than 2 percent.

(3) Compare incore to excore axial imbalance above 15% of RATED THERMAL POWER. Recalibrate if absolute difference greater than or equal to 3 percent.

(4) - Manual ESF functional input check every 18 months.

(5) - Each train tested every other month.

(6) - Neutron detectors may be excluded from CHANNEL CALIBRATION.

(7) - Below P-6 (BLOCK OF SOURCE RANGE REACTOR TRIP) setpoint.

(8) - The provisions of Specification 4.0.4 are not applicable.

The provisions of Specification 4.0.4 are not applicable for fl(AI) and f2(AI) penalties. (See also note 1 of Table 2.2-1)

(10) The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip circuits for the Manual Reactor Trip Function. The test shall also verify the OPERABILITY of the Bypass Breaker trip circuit(s).

The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers.

(12) - Local manual shunt trip prior to placing breaker in service.

(13) - Automatic Undervoltage Trip.

D. C. COOK - UNIT 1 3/4 3-14 AMENDMENT NO.

TABLE 3 . 3-3 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION A

O O

MINIMUM I TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 1 SAFETY INJECTION( TURBINE TRIP(

FEEDWATER ISOLATION, AND MOTOR DRIVEN FEEDWATER PUMPS

a. Manual Initiation 1, 2, 3, 4 18

b. Automatic Actuation 1, 2( 3, 4 13 Logic

c. Containment 1, 2, 3 14*

Pressure-High

d. Pressurizer 1, 2, 3g 14*

Pressure - Low

e. Differential Pressure Between Steam Lines High Four Loops 3/steam line 2/steam line 2/steam line 1, 2, 3gg 14*

Operating any steam line Three Loops 3/operating 1 /steam 2/operating 3gg 15 P~

Operating steam line line, any steam line operating steam line R

O

TABLE 3.3-3 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION O

O O

MINIMUM I TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

f. Steam Flow in Two Steam Lines-High Four Loops 2/steam line 1/steam line 1/steam line 14*

Operating any 2 steam lines Three Loops 2/operating 1 5 5 5/any 1/operating 3 15 Operating steam line operating steam line steam line COINCIDENT WITH EITHER T Low-Low 14*

avg /loop Four Loops- 1 T 2 T any Operating loopaXg 3 18Zgs Three Loops Operating 1 T / 148$ T 1 T in any op8riting two Zgerating 3 15 opening any loop loop loops

TABLE 3.3-3 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION oOA MINIMUM I TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION OR, COINCIDENT WITH Steam Line Pressure-Low Four Loops 1 pressure/ 2 pressures 1 pressure 14*

Operating loop any loops any 3 loops Three Loops 1 pressure/ 1 pressure 1 pressure 3 15 Operating operating in any oper- in any 2 loop ating loop operating loops 4l 2. CONTAINMENT SPRAY LA I

a. Manual 1) 2, 3) 4 18 00

b. Automatic Actuation 1, 2, 3, 4 13 Logic

c. Containment Pressure 4 1, 2, 3 16 High-High

TABLE 3.3-3 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION A

O MINIMUM O

TOTAL NO. CHANNELS CHANNELS APPLICABLE I FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

4. STEAM LINE ISOLATION

a. Manual 1/steam line 1/steam line 1/operating 1, 2, 3 18 steam line

b. Automatic 1, 2, 3 13 Actuation Logic

c. Containment Pressure 4 1, 2, 3 16 High-High

d. Steam Flow in Two Steam Lines High Four Loops 2/steam line 1/steam line 1/steam line 1) 2) 3 14*

Operating any 2 steam lines Three Loops 2/operating 1 /any 1/operating 15 Operating steam line operating steam line steam line t7 R

O

TABLE 3.3-3 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION O

o MINIMUM I TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION g

COINCIDENT WITH EITHER T

avg Low-Low Four Loops 1 T /loop 2 T any 14*

Operating looping 3 18@s Three Loops 1 T /oper- 1 T in 1 T in any 15 Operating atiggloop any opNating two%0erating loop loops ORg COINCIDENT WITH Steam Line Pressure-Low Four Loops 1 pressure/ 2 pressures 1 pressure 300 14*

Operating loop any loops any 3 loops Three Loops 1 pressure/ 1 pressure 1 pressure in 3 15 Operating operating in any oper- any 2 oper-loop ating loop ating loops 5 ~ TURBINE TRIP &

FEEDWATER ISOLATION

a. Steam Generator 3/loop 2/loop in 2/loop in 1I 2I 3 ]4*

Water Level any oper- each oper-O High-High ating loop ating loop

TABLE 3 3-3 Continued TABLE NOTATION Trip function may be bypassed in this MODE below P-ll.

Trip function may be bypassed in this MODE below P-12.

¹¹¹ The channel(s) associated with the protective functions derived from the out of service Reactor Coolant Loop shall be placed in the tripped mode.

¹¹¹¹ Manually trip all bistables which would be automatically tripped in the event pressure in the associated active loop were less than the pressure in the inactive loop. For example, if loop 1 is the inactive loop then the bistables which indicate low pressure in loops 2, 3, and 4 relative to loop 1 should be tripped.

The provisions of Specification 3.0.4 are not applicable.

ACTION STATEMENTS ACTION 13 - With the number of OPERABLE Channels one less than the Total Number of Channels, be in 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 ; however, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1.1.

ACTION 14 - With the number of OPERABLE Channels one less than the Total Number of Channels, operations may proceed until performance of the next required CHANNEL FUNCTIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 15 - With a channel associated with an operating loop inoperable, restore the inoperable channel to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in HOT SHUTDOWN within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; however, one channel associated with an operating loop may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1.1.

ACTION 16 - With the number of OPERABLE Channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition and the Minimum Channels OPERABLE requirement is met; one additional channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1.1.

D. C. COOK - UNIT 1 3/4 3-22 AMENDMENT NO.

TABLE 3 3-3 Continued ACTION 17 - With less than the Minimum Channels OPERABLE, operation may continue provided the containment purge and exhaust valves are maintained closed. 'I ACTION 18 With the number of OPERABLE Channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-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 in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 19 With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

The inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

b. The Minimum Channels OPERABLE requirements is met; however, one additional channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1.

ENGINEERED SAFETY FEATURES INTERLOCKS DESIGNATION CONDITION AND SETPOINT FUNCTION With 2 of 3 pressurizer P-11 prevents or defeats pressure channels greater manual block of safety tha'n or equal to 1915 psig. injection actuation on low pressurizer pressure.

With 2 of 4 T channels P-12 allows the manual less than or ave eqttal to block of safety injection Setpoint. from high steam flow coincident with either Setpoint greater than or low steam line pressure equal to 541 F. or low-low T . P-12 in coincidence Nfh high steam flow will result in a steam line isolation.

P-12 affects steam dump blocks.

With 3 of 4 T ave channels above the resetgvalue, the manual block of safe-ty injection from high steam flow coincident with either low steam line pressure or low-low T is prevented or defeate6.

D. C. COOK - UNIT 1 3/4 3-23 AMENDMENT NO.

TABLE 3 3-5 Continued ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

6. Steam Flow in Two Steam Lines-H h Coincident with Steam Line Pressure-Low

a. Safety Injection (ECCS) < 13.0¹/23.0¹¹

b. Reactor Trip (from SI) < 3.0

c. Feedwater Isolation < 8.0

d. Containment Isolation-Phase "A" < 18.0¹/28.0¹¹

e. Containment Purge and Exhaust Isolation Not Applicable

f. Auxiliary Feedwater Pumps Not Applicable

g. Essential Service Water System < 14 0¹/48 0¹¹

h. Steam Line Isolation < 8.0

7. Containment Pressure--Hi h-Hi h

a. Containment Spray < 45.0

b. Containment Isolation-Phase "B" Not Applicable

c. Steam Line Isolation < 7.0

d. Containment Air Recirculation Fan < 660.0

8. Steam Generator Water Level--Hi h-Hi h

a. Turbine Trip < 2.5

b. Feedwater Isolation < 11.0

9. Steam Generator Water Level--Low-Low

a. Motor Driven Auxiliary Feedwater Pumps < 60.0

b. Turbine Driven Auxiliary Feedwater Pumps < 60.0

10. 4160 volt Emer enc Bus Loss of Volta e

a. Motor Driven Auxiliary Feedwater Pumps < 60.0 ll. Loss of Main Feedwater Pum s

a. Motor Driven Auxiliary Feedwater Pumps < 60.0

12. Reactor Coolant Pum Bus Undervolta e

a. Turbine Driven Auxiliary Feedwater Pumps < 60.0 D. C. COOK - UNIT 1 3/4 3-29 AMENDMENT NO.

LE 4. 3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE A FUNCTIONAL UNIT CHECK CALIBRATION TEST RE UIRED O

O lq I

1 SAFETY INJECTIONS TURBINE TRI P g FEEDWATER ISOLATIONg AND MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS

a. Manual Initiation N.A. N.A. M(1) 1, 2, 3, 4

b. Automatic Actuation Logic N.A. N.A. M(2) lg 2g 3g 4

c. Containment Pressure-High M(3) 1, 2, 3

d. Pressurizer Pressure Low 1, 2( 3

e. Differential Pressure 1, 2, Between Steam Lines High 3

f. Steam Flow in Two Steam 1, 2, 3 Lines--High Coincident with T Low or Steam Line PNRsure Low

2. CONTAINMENT SPRAY

a. Manual Initiation N.A. N.A. M(1) 1, 2, 3, 4

b. Automatic Actuation Logic N.A. N.A. M(2) 1, 2, 3, 4

c. Containment Pressure High- M(3) 1, 2, 3 High O

The provisions of Specification 4.0.6 are applicable.

TABLE 4.3-2 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST RE UIRED

4. STEAM LINE ISOLATION

a. Manual N.A. N.A. M(1) 1, 2, 3

b. Automatic Actuation Logic N.A. N.A. M(2) 1, 2, 3

c. Containment, Pressure M(3) 1/ 2/ 3 High-High

d. Steam Flow in Two Steam Lines 1, 2, 3 High Coincident with T Low-Low PNRsure Low

5. TURBINE TRIP AND FEEDWATER ISOLATION

a. Steam Generator Water Level 1/ 2/ 3 High-High

6. MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS

a. Steam Generator Water Level Low-Low 1, 2, 3

b. 4 kv Bus 1, 2, 3 Loss of Voltage

c. Safety Injection N.A. N.A. M(2) 1, 2, 3

d. Loss of Main Feed Pumps N.A. N.A. 1/ 2 The provisions of Specification 4.0.6 are applicable.

A O

O O

I TABLE 4.3-2 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST

7. TURBINE DRIVEN AUXILIARY FEEDWATER PUMPS

a. Steam Generator Water Level Low-Low 1, 2, 3

b. Reactor Coolant Pump N.A. 1, 2, 3 Bus Undervoltage

8. LOSS OF POWER

a. 4 kv Bus 1, 2, 3, 4 Loss of Voltage

b. 4 kv Bus 1, 2, 3, 4 Degraded Voltage a

Pf O

INSTRUMENTATION'his page intentionally left blank.

D. C. COOK - UNIT 1 3/4 3-49 AMENDMENT NO.

INSTRUMENTATION This page intentionally left blank.

D. C. COOK - UNIT 1 3/4 3-50 AMENDMENT NO.

REACTOR COOLANT SYSTEM HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 a. The reactor coolant loops listed below shall be OPERABLE and in operation as required by items b, c, and d:

1. Reactor Coolant Loop 1 and its associated steam generator and reactor coolant pump,

2. Reactor Coolant Loop 2 and its associated steam generator and reactor coolant pump,

3. Reactor Coolant Loop 3 and its associated steam generator and reactor coolant pump,

4. Reactor Coolant Loop 4 and its associated steam generator and reactor coolant pump.

b. At least two of the above coolant loops shall be OPERABLE and at least one loop in operation if the reactor trip breakers are in the open position, or the control rod drive system is not capable of rod withdrawal.*

At least three of the above coolant loops shall be OPERABLE and in operation when the reactor trip system breakers are in the closed position and the control rod drive system is capable of rod withdrawal.

At least three of the above coolant loops shall be OPERABLE and in operation above P-12. (Refer to Technical Specification 3.3.2.1, Table 3.3-3 for instrumentation requirements.)

APPLICABILITY: MODE 3 ACTION:

a. With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With less than the number of operating coolant loops required by item c above, restore the required number of coolant loops within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or open the reactor trip breakers.

With less than the number of operating coolant loops required by item d above, restore the required number of coolant loops within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or lower the reactor coolant system temperature belo~ P-12.

D. C. COOK - UNIT 1 3/4 4-2 AMENDMENT NO.

REACTOR COOLANT SYSTEM HOT STANDBY LIMITING CONDITION FOR OPERATION Continued

d. With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System** and immediately initiate corrective action to return the required coolant loop to operation.

SURVEILLANCE RE UIREMENTS 4.4.1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

4.4.1.2.2 At least one cooling loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

All reactor coolant pumps may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided (1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration+*, and (2) core outlet temperature is maintained at least 10 0 F below saturation temperature.

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

D. C. COOK - UNIT 1 3/4 4-2a AMENDMENT NO.

REACTOR COOLANT SYSTEM SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 a. The coolant loops listed below shall be OPERABLE and in operation as required by items b and c:

1. Reactor Coolant Loop 1 and its associated steam generator and reactor coolant pump,*

2. Reactor Coolant Loop 2 and its associated steam generator and reactor coolant pump,*

3. Reactor Coolant Loop 3 and its associated steam generator and reactor coolant pump,*

4. Reactor Coolant Loop 4 and its associated steam generator and reactor coolant pump,*

5. Residual Heat Removal - East,**

6. Residual Heat Removal - Vest,**

b. At least two of the above coolant loops shall be OPERABLE and at least one loop in operation if the reactor trip breakers are in the open position, or the control rod drive system is not capable of rod withdrawal.***

At least three of the above reactor coolant loops shall be OPERABLE and in operation when the reactor trip system breakers are in the closed position and the control rod drive system is capable of rod withdrawal.

APPLICABILITY: MODES 4 and 5 ACTION'ith less than the above required coolant loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; be in COLD SHUTDOWN within 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b. With less than the number of operating coolant loops required by item c above, restore the required number of coolant loops within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or open the reactor trip breakers.

With no coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System+*** and immediately initiate corrective action to return the required coolant loop to operation.

D. C. COOK - UNIT 1 3/4 4-3 AMENDMENT NO.

REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS 4.4.1.3.1 The required residual heat removal loop(s) shall be determined OPERABLE per Specification 4.0.5.

4.4.1.3.2 The required reactor coolant pump(s),

be determined to be OPERABLE once per 7 days by if not in operation, shall verifying correct breaker alignments and indicated power availability.

4.4.1.3.3 The required steam generator(s) shall be determined OPERABLE by verifying secondary side level to be greater than or equal to 25% of wide range instrument span at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.1.3.4 At least one coolant loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .,

A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures less than or equal to 170 0 F unless 1) the pressurizer. water volume is less than 62% of span or 2) the secondary water temperature of each steam generator is less than 50 0 F above each of the RCS cold leg temperatures. Operability of a reactor coolant loop(s) does not require an OPERABLE auxiliary feedwater system.

. ** The normal .or emergency power source may be inoperable in MODE 5.

- - All reactor coolant pumps and residual heat removal pumps may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided 1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration+***, and 2) core outlet temperature is maintained at least 10 F below saturation temperature.

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

D. C. COOK - UNIT 1 3/4 4-3a AMENDMENT NO.

REACTOR COOLANT SYSTEM This page intentionally left blank.

D. C. COOK - UNIT 1 3/4 4-3b AMENDMENT NO.

PLEASE REMOVE THIS PAGE D. C. COOK - UNIT 1 3/4 4-3c AMENDMENT NO ~

PLEASE REMOVE THIS PAGE D. C. COOK - UNIT 1 3/4 4-38 AMENDMENT NO.

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 + 1%.*

APPLICABILITY: MODES 4 and 5.

ACTION:

With no pressurizer code safety valve OPERABLE:

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

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

D. C. COOK - UNIT 1 3/4 4-4 AMENDMENT NO.

REACTOR COOLANT SYSTE SAFETY VALVES - OPERATING LIMITING CONDITION FOR OPERATION 3.4.3 All pressurizer code safety valves shall be OPERABLE with a lift setting of 2485 PSIG + 1%.*

APPLICABILITY: MODES 1, 2 and 3.

ACTION'ith 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 Each pressurizer code safety valve shall be demonstrated OPERABLE with a lift setting of 2485 psig Boiler and Pressure

+1% in accordance with Section XI of the ASME Vessel Code, 1974 Edition.

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

D. C. COOK - UNIT 1 3/4 4-5 AMENDMENT NO.

REACTOR COOLANT SYSTEM RELIEF VALVES - OPERATING LIMITING CONDITION FOR OPERATION 3.4.11 Three power operated relief valves (PORVs) and their associated block valves shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. PORVs inoperable:*

1. With one PORV inoperable, within 1 hour either restore the inoperable PORV to OPERABLE status or close the associated block valve and remove power from the block valve; otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

2. With two PORVs inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore at least one of the inoperable PORVs to OPERABLE status or close the associated block valves and remove power from the block valves; restore at least one of the inoperable PORVs to OPERABLE status within the following 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 and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3. With three PORVs inoperable, within 1 hour either restore at least one of the PORVs to OPERABLE status or close their associated block valves and remove power from the block valves and be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. Block valves inoperable:*

1. With one block valve inoperable, within 1 hour either (1) restore the block valve to OPERABLE status, or (2) close the block valve and remove power from the block valve, or (3) close the associated PORV and remove power from the associated solenoid valve; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

PORVs isolated to limit RCS leakage through their seats and the block valves shut to isolate this leakage are not considered inoperable.

D. C. COOK - UNIT 1 3/4 4-35 AMENDMENT NO.

REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION Continued

2. With two or more block valves inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either (1) restore a total of at least two block valves to OPERABLE status, or (2) close the block valves and remove power from the block valves, or (3) close the associated PORVs and remove power from their associated solenoid valves; and apply the portions of ACTION a.2 or a.3 above for inoperable PORVs, relating to OPERATIONAL MODE, as appropriate.

With PORVs and block valves not in the same line inoperable,

within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either (1) restore the valves to OPERABLE status or (2) close and de-energize the other valve in each line.

Apply the portions of ACTION a.2 or a.3 above, relating to OPERATIONAL MODE, as appropriate for two or three lines unavailable.

d. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.4.11.1 Each of the three PORVs shall be demonstrated OPERABLE:

a. At least once per 31 days by performance of a CHANNEL FUNCTIONAL TEST, excluding valve operation, and

b. At least once per 18 months by performance of a CHANNEL CALIBRATION.**

4.4.11.2 Each of the three block valves shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel. The block valve(s) do not have to be tested when ACTION 3.4.ll.a or 3.4.11.c is applied.

4.4.11.3 The emergency power supply for the PORVs and block valves shall be demonstrated OPERABLE at least once per 18 months by operating the valves through a complete cycle of full travel while the emergency buses are energized by the onsite diesel generators and onsite plant batteries. This testing can be performed in conjunction with the requirements of Specifications 4.8.1.1.2.b and 4.8.2.3.2.d.**

PORVs isolated to limit RCS leakage through their seats and the block valves shut to isolate this leakage are not considered inoperable.

+~The provisions of Specification 4.0.6 are applicable.

D. C. COOK - UNIT 1 3/4 4-36 AMENDMENT NO.

3 4 5 EMERGENCY CORE COOLING SYSTEMS ECCS ACCUMULATORS LIMITING CO DITION FOR OPERATION 3.5.1 Each reactor coolant system accumulator shall be OPERABLE with:

a. The isolation valve open,

b. A contained borated water volume of between 929 and 971 cubic feet,

c. A boron concentration of between 2400 ppm and 2600 ppm, and

d. A nitrogen cover-pressure of between 585 and 658 psig.

APPLICABILITY: MODES 1, 2 and 3.*

ACTION:

a. With one accumulator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within one hour or be in HOT SHUTDOWN within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b. With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in HOT STANDBY within one hour and be in HOT SHUTDOWN within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

SURVEILLANCE RE UIREMENTS 4.5.1 Each accumulator shall be demonstrated

a. 't least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by:

OPERABLE:

1. Verifying the water level and nitrogen cover-pressure in the tanks, and

2. Verifying that each accumulator isolation valve is open.

Pressurizer Pressure above 1000 psig.

DE C. COOK - UNIT 1 3/4 5-1 AMENDMENT NO.

EMERGENCY CORE COOLING SYSTEM 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 80 0 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 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.

D. C. COOK - UNIT 1 3/4 5-11 AMENDMENT NO.

3 4.7 PLANT SYSTEMS 3 4.7.1 TURBINE CYCLE SAFETY VALVES

'I LIMITING CONDITION FOR OPERATION 3.7 '.1 All main steam line code safety valves associated with each steam generator shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a. ~ With 4 reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves inoperable, operation in MODES 1, 2 and 3 may proceed provided,-that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Setpoint trip is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With 3 reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves associated with an operating loop inoperable, operation in MODE 3 may proceed provided, that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either the inoperable valve is restored to OPERABLE status or the reactor trip breakers are opened; otherwise, be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.7.1.1 Each main steam line code safety valve shall be demonstrated OPERABLE, with lift settings and orifice sizes as shown in Table 4.7-1, in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition.

D. C. COOK - UNIT 1 3/4 7-1 AMENDMENT NO.

PLANT SYSTEMS This page intentionally left blank.

D. C. COOK '- UNIT 1 3/4 7-3 AMENDMENT NO.

a TABLE 4.7-1 oO STEAM LINE SAFETY VALVES PER LOOP I

g VALVE NUMBER LIFT SETTING 14

ORIFICE SIZE a ~ SV-1 .1065 psig 16 in.

b. SV-1 1065 psig 16 in.

c SV-2 1075 psig 16 in.

d. SV-2 1075 psig 16 in.

e. SV-3 1085 psig 16 in.

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

at

PLANT SYSTEMS AUXILIARY FEEDWATER SYSTE LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:

a. Two feedwater pumps, each capable of being powered from separate emergency busses, and

b. One feedwater pump capable of being powered from an OPERABLE steam supply system.

APPLICABILITY: MODES 1, 2, and 3.

ACTION'ith one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater 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 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With two auxiliary feedwater pumps inoperable, 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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c ~ With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible.

SURVEILLANCE RE UIREMENTS 4.7.1.2 Each auxiliary feedwater pump shall be demonstrated OPERABLE:

At least once per 31 days by:

1. Verifying that each motor driven pump develops an equivalent discharge pressure of greater than or equal to 1240 psig at 60 0 F on recirculation flow.

2. Verifying that the steam turbine driven pump develops an equivalent discharge 0

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.

D. C. COOK - UNIT 1 3/4 7-5 AMENDMENT NO.

PLANT SYSTEMS STEAM GENERATOR STOP VALVES LIMITING CONDITION FOR OPERATION 3.7.1.5 Each steam generator stop valve shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

MODES 1 - With one steam generator stop valve inoperable but open, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; otherwise, reduce power to less than or equal to 5 percent of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

MODES 2 - With one steam generator stop valve inoperable, subsequent and 3 operation in MODES 2 or 3may proceed provided:

a. The stop valve is maintained closed.

b. The provisions of Specification 3.0.4 are not applicable.

Otherwise, be in HOT SHUTDOWN within the next 12 hours.

SURVEILLANCE RE UIREMENTS 4.7.1.5.1 Each steam generator stop valve that is open shall be demonstrated OPERABLE by:

a. Part-stroke exercising the valve at least once per 92 days, and

b. Verifying full closure within 5 seconds on any closure actuation signal while 0 in HOT STANDBY with T greater than or equal to 541 F during each reactor snutdown except that verification of full closure within 5 seconds need not be determined more often than once per 92 days.

4.7.1.5.2 The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.

4.7.1.5.3 The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 when performing PHYSICS TESTS at the beginning of a cycle, provided the steam generator, stop valves are maintained closed.

D. C. COOK - UNIT 1 3/4 7-10 AMENDMENT NO.

ELECTRICAL POWER SYSTEMS SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE'.

One circuit between the offsite transmission network and the onsite Class 1E distribution system, and

b. One diesel generator with:

1. A day tank containing a minimum of 70 gallons of fuel,

2. A fuel storage system containing a minimum of 42,000 gallons of fuel, and

3. A fuel transfer pump.

APPLICABILITY: MODES 5 and 6.

ACTION:

With less than the above minimum required A.C. electrical power sources OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes* until the minimum required A.C. electrical power sources are restored to OPERABLE status'URVEILLANCE RE UIREMENTS 4.8.1.2 The above required A.C. electrical power sources shall be demonstrated OPERABLE by the performance of each of the Surveillance Requirements of 4.8.1.1.1 and 4.8.1.1.2 except for requirement 4.8.1 1.2a.6.**

~

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.

- The provisions of Specification 4.0.6 are applicable.

D. C. COOK - UNIT 1 3/4 8-5 AMENDMENT NO.

3 4 9 REFUELING OPERATION BORON CONCENTRATION LIMITI G CONDITION FOR OPERATION 3.9.1 With the reactor vessel head unbolted or removed, the boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions is met:

a. Either a K.eff of 0.95 or less, which includes ff a 1% d k/k conservative allowance for uncertainties, or

b. A boron concentration of greater than or equal to 2400 ppm, which includes a 50 ppm conservative allo~ance for uncertainties.

APPLICABILITY: MODE 6*

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes** and initiate and continue boration at greater than or equal to 10 gpm of 20,000 ppm boric acid solution or its equivalent until K ff is reduced to less than or equal to 0.95 or the boron concentration is restored to greater than or equal to 2400 ppm, whichever is the more restrictive. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:

a. Removing or unbolting the reactor vessel head, and

b. Withdrawal of any full length control rod in excess of 3 feet from its fully inserted position.

4.9.1.2 The boron concentration of the reactor coolant system and the refueling canal shall be determined by chemical analysis at least 3 times per 7 days with a maximum time interval between samples of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

The reactor shall be maintained in MODE 6 when the reactor vessel head is unbolted or removed.

- 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.

D. C. COOK - UNIT 1 3/4 9-1 AMENDMENT NO.

V REFUELING OPERATIONS INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 As a minimum, two source range neutron flux monitors shall be operating, each with continuous visual indication in the control room and one with audible indication in the containment.

APPLICABILITY: MODE 6.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes.* The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.2 Each source range neutron flux monitor shall be demonstrated OPERABLE by performance of:

a. A CHANNEL FUNCTIONAL TEST at least once per 7 days, and

b. A CHANNEL FUNCTIONAL TEST within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to the initial start of CORE ALTERATIONS, and

c. A CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during CORE ALTERATIONS.

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.

D. C. COOK - UNIT 1 3/4 9-2 AMENDMENT NO.

REFUELING OPERATIONS 3 4 9 8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal loop shall be in operation.

APPLICABILITY: MODE 6.

ACTION:

a. With less than one residual heat removal loop in operation, except as provided in b. below, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System.* Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b. The residual heat removal loop may be removed from operation for up to 1 hour per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs.

c. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.8.1 A residual heat removal loop shall be determined to be in operation and circulating reactor coolant at a flow rate of greater than or equal to 2000 gpm at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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.

D. C. COOK - UNIT 1 3/4 9-9 AMENDMENT NO.

SPECIAL TEST EXCEPTIONS GROUP HEIGHT INSERTION AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The group height, insertion and power distribution limits of Specifications 3.1.3.1, 3.1.3.4, 3.1.3.5, 3.2.1, and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:

a. The THERMAL POWER is maintained less than or equal to 85% of RATED THERMAL POWER, and

b. The limits of Specifications 3.2.2 and 3.2.3 are maintained and determined at the frequencies specified in Specification 4.10.2.2 below.

APPLICABILITY: MODE 1 ACTION:

With any of the limits of Specifications 3.2.2 or 3.2.3 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.4, 3.1.3.5, 3.2.1 and 3.2.4 are suspended, either:

a. Reduce THERMAL POWER sufficient to satisfy the ACTION requirements of Specifications 3.2.2 and 3.2.3, or

b. Be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE RE UIREMENTS 4.10.2.1 The THERMAL POWER shall be determined to be less than or equal to 85% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.

4.10.2.2 The Surveillance Requirements of Specifications 4.2.2.2 and 4.2.3 shall be performed at the following frequencies during PHYSICS TESTS:

a. Specification 4.2.2.2 - At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. Specification 4.2.3 - At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D. C. COOK - UNIT 1 3/4 10-2 AMENDMENT NO.

SPECIAL TEST EXCEPTIONS PRESSURE TEMPERATURE LIMITATION - REACTOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.10.3 The minimum temperature and pressure conditions for reactor criticality of Specifications 3.1.1.5 and 3.4.9.1 may be suspended during low temperature PHYSICS TESTS provided:

a. The THERMAL POWER does not exceed 5 percent of RATED THERMAL POWER,

b. The Reactor Trip Setpoints for the OPERABLE Intermediate Range, Neutron Flux and the Power Range, Neutron Flux, Low Setpoint are set at less than or equal to 25% of RATED THERMAL POWER, and

c. The Reactor Coolant System temperature and pressure relationship is maintained within the region of acceptable operation shown on Figures 3.4-2 and 3.4-3.

APPLICABILITY: MODE 2 ACTION'.

With the THERMAL POWER greater than 5 percent of RATED THERMAL POWER, immediately open the reactor trip breakers.

b. With the Reactor Coolant System temperature and pressure relationship within the region of unacceptable operation on Figures 3.4-2 and 3.4-3, immediately open the reactor trip breakers and restore the temperature-pressure relationship to within its limit within 30 minutes; perform the analysis required by Specification 3.4.9.1 prior to the next reactor criticality.

SURVEILLANCE RE UIREMENTS 4.10.3.1 The Reactor Coolant System shall be verified to be within the acceptable region for operation of Figures 3.4-2 and 3.4-3 at least once per hour.

4.10.3.2 The THERMAL POWER shall be determined to be less than or equal to 5%

of RATED THERMAL POWER at least once per hour.

D. C. COOK-UNIT 1 3/4 10-3 AMENDMENT NO.

SPECIAL TEST EXCEPTIONS PHYSICS TESTS LIMITING CONDITION FOR OP RATION 3.10.4 The limitations of Specifications 3.1.1.4, 3.1.3.1, 3.1.3.4 and 3.1.3 ' may be suspended during the performance of PHYSICS TESTS provided:

a ~ The THERMAL POWER does not exceed 5% of RATED THERMAL POWER, and The Reactor Trip Setpoints for the OPERABLE Intermediate Range, Neutron Flux and the Power Range, Neutron Flux, Low Setpoint are set at less than or equal to 25% of RATED THERMAL POWER.

APPLICABILITY: MODE 2.

ACTION:

With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immediately open the reactor trip breakers.

SURVEILLANCE RE UIREMENTS 4.10.4.1 The THERMAL POWER shall be determined to be less than or equal to 5% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.

4.10.4.2 Each Intermediate and Power Range Channel shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to initiating PHYSICS TESTS.

D. C. COOK - UNIT 1 3/4 10-5 AMENDMENT NO.

SPECIAL TEST EXCEPTION NATURAL CIRCULATION TESTS LIMITING CONDITION FOR OPERATION 3.10.5 The limitations of Specification 3.4.1.1 may be suspended during the,,

performance of PHYSICS TESTS and Thermal-Hydraulic Tests, provided:

a. The THERMAL POWER does not exceed the P-7 Interlock Setpoint, and

b. The Reactor Trip Setpoints for the OPERABLE Intermediate Range, Neutron Flux and the Power Range,'eutron Flux, Low Setpoint are set at less than or equal to 25% of RATED THERMAL POWER.

APPLICABILITY: During operation below the P-7 Interlock Setpoint.

ACTION:

With the THERMAL POWER greater than the P-7 Interlock Setpoint, immediately open the reactor trip breakers.

SURVEILLANCE RE UIREMENTS 4.10.5.1 The THERMAL POWER shall be determined to 'be less than the P-7 Interlock Setpoint at least once per hour during PHYSICS TESTS.

4.10.5.2 Each Intermediate, Power Range Channel and P-7 Interlock shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to initiating PHYSICS TESTS.

D. C. COOK - UNIT 1 3/4 10-6 AMENDMENT NO.

ADMINISTRATIVE CONTROLS PEAKING FACTOR LIMIT REPORT 6.9.1.11 The Peaking Factor Limit Report shall be provided to the NRC Regional Administrator with a copy to Director of Nuclear Reactor Regulation, Attention:

Chief, Core Performance Branch, U.S. Nuclear Regulatory Commission, Washington, D. C. 20555, containing V(Z) functions for the new cycle at least 15 days prior to each cycle initial criticality unless otherwise approved by the Commission by letter. In the event that the limit should change, a new or amended Peaking Factor Limit will be submitted 15 days prior to the date the limit would become effective unless otherwise approved by the Commission by letter. Any information needed to support the content of the Peaking Factor Report will be by request from the NRC and need not be included in this report.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference Specifications:

a. Inservice Inspection Program Review, Specification 4.4.10.

b.'CCS Actuation, Specifications 3.5.2 and 3.5.3.

C. Inoperable Seismic Monitoring Instrumentation, Specification 3.3.3.3.

d. Inoperable Meteorological Monitoring Instrumentation, Specification 3.3.3.4.

e. Seismic event analysis, Specification 4.3.3.3.2.

f. Sealed Source leakage in excess of limits, Specification 4.7.7.1.3.

g. Fire Detection Instrumentation, Specification 3.3.3.7.

h. Fire Suppression Systems, Specifications 3.7.9.1, 3.7.9.2, 3.7.9.3 and 3.7.9.4.

D. C. COOK - UNIT 1 6-19 AMENDMENT NO.

2 1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation, which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime, where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures, because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in the heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore, THERMAL POWER and Reactor Coolant Tempera-ture and Pressure have been related to DNB. This relation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The DNB design basis is as follows: there must be at least a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the design DNBR limit.

In meeting this design basis, uncertainties in plant operating para-meters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically, such that there is at least a 95 percent confi-dence that the minimum DNBR for the limiting rod is greater than or equal to the applicable design DNBR limit for each fuel type (as defined below).

For 4 loop operation, the improved thermal design procedure is used. The uncertainties in the plant parameters are used to determine the plant DNBR uncertainty. This DNBR uncertainty, combined with the correlation DNBR limit (as defined below), establishes a design DNBR limit value, which must be met in plant safety analyses, using values of input parameters without uncertainties.

The table below indicates the relationship between the correlation limit DNBR, design limit DNBR, and the safety analysis limit DNBR values used for this design.

D. C. COOK - UNIT 1 B 2-1 AMENDMENT NO.

2 1 SAFETY LIMITS BASES 4 Loop Operation Exxon Westinghouse Fuel Nuclear Co. Fuel (15x15 OFA) (15xl5)

(WRB-1 Correlation) (W-3 Correlation)

Typical Thimble Typical Thimble Cell* Cell** Cell* Cell**

Correlation Limit 1.17 1.17 1.30 1.30 Design Limit DNBR 1 32

~ 1.31 1.58 1.50 Safety Analysis Limit DNBR 1.69 1.69, 1.58 1.50 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 D. C. COOK - UNIT 1 B 2-1(a) AMENDMENT NO.

LIMITING SAFETY SYSTEM SETTINGS BASES Over ower T The Overpower hT reactor trip provides assurance of fuel integrity, e.g., no melting, under all possible overpower conditions, limits the required range for Overtemperature 5T protection, and provides a backup to the High Neutron Flux trip. The setpoint includes corrections for axial power distribution, changes in density and heat capacity of water with temperature, and dynamic compensation for piping delays from the core to the loop temperature detectors. No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System. If axial peaks are more severe 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 ~

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. No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.

DE C. COOK - UNIT 1 B 2-5 AMENDMENT NO.

SAFETY LIMITS BASES Loss of Flow The Loss of Flow trips provide core protection to prevent DNB in the event of a loss of one or more reactor coolant pumps.

Above will occur ll if percent of RATED THERMAL POWER, the flow in any an automatic reactor two loops drops below 90% of nominal trip full loop flow. Above the P-8 setpoint, less than or equal to 31% of RATED THERMAL POWER, automatic reactor loop drops below 90% of nominal trip will occur full loop flow.

if the flow in any single Steam Generator Water Level The Steam Generator Water Level Low-Low trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity. The specified setpoint provides allowance that there will be sufficient water inventory in the steam generators at the time of trip, to allow for starting delays of the auxiliary feedwater system.

Steam Feedwater Flow Mismatch and Low Steam Generator Water Leve The Steam/Feedwater Flow Mismatch in coincidence with a Steam Generator Low Water Level trip is not used in the transient and accident analyses, but is included in Table 2.2-1 to ensure the functional capa-bility of the specified trip settings and thereby enhance the overall reliability of the Reactor Protection System. This trip is redundant to the Steam Generator Water Level Low-Low trip. The Steam/Feedwater Flow Mismatch portion of this trip is activated when tge steam flow exceeds the feedwater flow by less than or equal to 0.71 x 10 lbs/hour. The Steam Generator Low Water level portion of the trip is activated when the water level drops below 25 percent, as indicated by the narrow range instrument.

These trip values include sufficient allowance in excess of normal operating values to preclude spurious trips but will initiate a reactor trip before the steam generators are dry. Therefore, the required capacity and starting time requirements of the auxiliary feedwater pumps are reduced and the resulting thermal transient on the Reactor Coolant System and steam generators is minimized.

D. C. COOK - UNIT 1 B 2-6 AMENDMENT NO.

LIMITING SAFETY SYSTEM SETTINGS BASES Undervolta e and Underfre uenc - Reactor Coolant Pum Busses The Undervoltage and Underfrequency Reactor Coolant Pump bus trips provide reactor core protection against DNB as a result of loss of voltage or underfrequency to more than one reactor coolant pump. The specified set points assure a reactor trip signal is generated before the low flow trip set point is reache'd. A O.l second time delay is incorporated in each of these trips to prevent spurious reactor trips from momentary electrical power transients.

Turbine Tri A Turbine Trip causes a direct reactor trip when operating above P-7.

Each of the turbine trips provide turbine protection and reduce the severity of the ensuing transient. No credit was taken in the accident analyses for operation of these trips. Their functional capability at the specified trip settings is required to enhance the overall reliability of the Reactor Protection System.

Safet In'ection In ut from ESF If a reactor trip has not already been generated by the reactor protective instrumentation, the ESF automatic actuation logic channels will initiate a reactor trip upon any signal which initiates a safety injection.

This trip is provided to protect the core in the event of a LOCA. The ESF instrumentation channels which initiate a safety injection signal are shown in Table 3.3-3.

Reactor Coolant Pum Breaker Position Tri The Reactor Coolant Pump Breaker Position Trips are anticipatory trips which provide reactor core protection against DNB resulting from the opening of any one pump breaker above P-8 or the opening of two or more pump breakers below P-8. These trips are blocked below P-7. The open/close position trips assure a reactor trip signal is generated before the low flow trip set point is reached. No credit was taken in the accident analyses for operation of these trips. Their functional capability at, the open/close position settings is required to enhance the overall reliability of the Reactor Protection System.

D. C. COOK - UNIT 1 B 2-7 AMENDMENT NO.

'I REMOVE THIS PAGE D. C. COOK - UNIT 1 B 2-8 AMENDMENT NO.

3 4 1 REACTIVITY CONTROL SYSTEMS BASES 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 for increased load events occurs at EOL, wrth T ave'ondition at no load operating temperature, and is associated with a postulatei Steam line break accident and resulting uncontrolled RCS cooldown. In the analysis of this accident, a minimum SHUTDOWN MARGIN of 1.60% Ak/k is initially required to control the reactivity 0transient and automatic ESF is assumed to be available.

With T less than 200 F, the reactivity transients resulting from a postulated steam Pi5e break cooldown are minimal and a 18 Ak/k SHUTDOWN MARGIN provides adequate protection for this event.

In shutdown MODES 4 and 5 when heat removal is provided by the residual heat removal system, active reactor coolant system volume may be reduced.

Increased SHUTDOWN MARGIN requirements when operating under these conditions is provided for high reactor coolant system boron concentrations. to ensure sufficient time for operator response in the event of a boron dilution transient.

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 + 100 cubic feet in approximately 45 minutes. The reactivity change rate associated with boron reductions will therefore be within the

'apability for operator recognition and control.

3 4 1 1 4 MODERATOR TEMPERATURE COEFFICIENT MTC 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 D. C. COOK - UNIT 1 B 3/4 1-1 AMENDMENT NO.

3 4 1 REACTIVITY CONTROL SYSTEMS BASES 3 4 1.1 4 MODERATOR TEMPERATURE COEFFICIENT MTC Continued principally to the reduction in RCS boron concentration associated with fuel burnup. The confirmation that the measured and appropriately compensated MTC value is within the allowable tolerance of the predicted value provides additional assurances that the coefficient will be maintained within its limits during intervals between measurement.

3 4 1 1 5 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 541 0 F. This limitation is required to ensure 1) the moderator temperature coefficient is within its analyzed temperature range, 2) the protective instrumentation is within its normal operating range, 3) the pressurizer is capable of being in a OPERABLE status with a steam bubble, and 4) the reactor pressure vessel is above its minimum temperature.

Administrative procedures will be established RTNDT to ensure the P-12 blocked functions are unblocked before taking the reactor critical.

3 4 1.2 BORATION SYSTEMS The boron injection system ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include 1) borated water sources, 2) charging pumps, 3) separate flow paths, 4) boric acid transfer pumps, 5) associated heat tracing systems, and 6) an emergency power supply from OPERABLE diesel generators.

With the RCS average temperature above 200 F, a minimum of two separate and redundant boron injection systems are provided to ensure single functional capability in the event an assumed failure renders one of the systems inoperable. Allowable out-of-service periods ensure 'that minor component repair or corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair period.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps and safety injection pumgs, except the required OPERABLE charging pump, to be inoperable below 170 F, unless the reactor vessel head is removed, provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.

The boration capability of either system is sufficient to provide the required SHUTDOWN MARGIN from all operating conditions after xenon decay and cooldown to 200 0 F. The maximum expected boration capability, usable volume requirement, is 5641 gallons of 20,000 ppm borated water from the boric acid storage tanks or 99,598 gallons of 2400 ppm borated water from the refueling water storage tank. The minimum contained RWST volume is based on ECCS considerations. See Section B 3/4.5.5.

D. C. COOK - UNIT 1 B 3/4 1-2 AMENDMENT NO.

REACTIVITY CONTROL SYSTEMS BASES BORATION SYSTEMS Continued With the RCS average temperature below 200 0 F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable.

The boration capability required below 200 F is sufficient to provide 0the required 0

MODE 5 SHUTDOWN MARGIN after xenon decay and cooldown from 200 F to 140 F. This condition requires usable volumes of either 2890 gallons of 20,000 ppm borated water from the boric acid storage tanks or 76,937 gallons of 2400 ppm borated water from the refueling water storage tank, The boration source volumes of Technical Specification 3.1.2:7 have been conservatively increased to 4300 gallons from the boric acid storage tank and 90,000 gallons from the RWST These values were

~

chosen to be consistent with Unit 2. The Unit 2 value for the boric acid storage tank volume includes sufficient boric acid to borate to 2000 ppm.

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'he OPERABILITY of boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6.

3 4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) limit the potential effects of rod ejection accident.

OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.

The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original criteria are met. Misalignment of a rod requires measurement of peaking factors or a restriction in THERMAL POWER; either of these restrictions provide assurance of fuel rod integrity during continued operation. The reactivity worth of a misaligned rod is limited for the remainder of the fuel cycle to prevent exceeding the assumptions used in the accident analysis for a rod ejection accident.

The maximum rod drop time restriction is consistent with the assumed rod drop time used in the accident analyses. Measurement with T greater than or equal to 541 0 F and with all reactor coolant pumpsavg operating ensures that the measured drop times will be representative of insertion times experienced during a reator trip at operating conditions.

D. C. COOK - UNIT 1 B 3/4 1-3 AMENDMENT NO.

3 4 2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (a) maintaining the minimum DNBR in the core greater than or equal to 1.69 during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200 F is not exceeded.

The definitions of hot channel factors as used in these specifi-cations are as follows:

F (Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for man-ufacturing tolerances on fuel pellets and rods'>H Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power.

3 4 2 1 AXIAL FLUX DIFFERENCE AFD Target flux difference is determined at equilibrium xenon conditions. The full length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady state operation at high power levels. The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.

D. C. COOK - UNIT 1 B 3/4 2-1 AMENDMENT NO.

POWER DISTRIBUTIO LIMITS BASES Although it is intended that the plant will be operated with the AXIAL FLUX DIFFERENCE within the target band about the target flux difference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target band at reduced THERMAL POWER levels. This deviation will not affect the xenon redistribution sufficiently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POWER (with the AFD within the target band) provided the time duration of the deviation is limited. Accordingly, a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months penalty deviation limit cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is provided for operation outside of the target band but within the limits of Figure 3.2-1 while at THERMAL POWER levels above 50% of RATED THERMAL POWER. For THERMAL POWER levels below 50% of RATED THERMAL POWER, deviations of the AFD outside of the target band are less significant. The penalty of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months actual time reflects this reduced significance.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer determines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the target band and the THERMAL POWER is greater than 90% or 0.9 x APL of RATED THERMAL POWER (whichever is less). During operation at THERMAL POWER levels between 15$ and 908 or 0.9 x APL of RATED THERMAL POWER (whichever is less), the computer outputs an alarm message when the penalty deviation accumulates beyond the limits of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , respectively.

The upper bound limit (90% or 0.9 x APL of RATED THERMAL POWER (whichever is less)) on AXIAL FLUX DIFFERENCE assures that the F (Z) envelope is not exceeded during either normal operation or in th9 event of xenon redistribution following power changes. The lower bound limit (50% of RATED THERMAL POWER) is based on the fact that at THERMAL POWER levels below 50% of RATED THERMAL POWER, the average linear heat generation rate is half of its nominal operating value and below that value, perturbations in localized flux distributions cannot affect the results of ECCS or DNBR analyses in a manner which would adversely affect the health and safety of the public.

Figure B 3/4 2-1 shows a typical monthly target band near the beginning of core life.

The bases and methodology for establishing these limits is presented in topical report WCAP - 8385, "Power Distribution Control and Load Following Procedures."

D. C. COOK - UNIT 1 B 3/4 2-2 AMENDMENT NO.

5% 5%

100 90~

80 70 p .L i t arget Fluy Differen ce

!6O 1 L

l

]

I a >0~

o40I 0

~ 30F

.L i

I 20 10I I

OL i l.

-30 -10 0 10 20 30 INDICATED AXIAL FLUX DIFFERENCE (PERCENT)

Figure B 3/4 2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL POWER AT BOL D. C. COOK UNIT 1 B 3/4 2-3 AMENDMENT NO.

POWER DISTRIBUTION LIMITS BASES 3 4 2.2 and 3 4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTORS The limits on heat flux and nuclear enthalpy rise hot channel factors ensure that 1) the design limits on peak local power density and minimum DNBR are not exceeded and 2) in the event of a LOCA, the peak fuel clad temperature will not exceed the 2200 0 F ECCS acceptance criteria limit.

Each of these hot channel factors are measurable, but will normally only be determined periodically, as specified in Specifications 4.2.2.1, 4.2.2.2, 4.2.3, 4.2.6.1 and 4.2.6.2. This periodic surveillance is sufficient to insure that the hot channel factor limits are maintained provided:

a. Control rods in a single group move together with no individual rod insertion differing by more than + 12 steps from the group demand position.

b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.5.

c. The control rod insertion limits of Specifications 3.1.3.4 and 3.1.3.5 are maintained.

d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE is maintained within the limits.

N The relaxation in F~ as a function of THER1AL POWER allows changes ig the radial power shape Eor all permissible rod insertion limits, F will be maintained within its limits, provided conditions (al through

( above are maintained, When an F measurement is taken, both experimental error and manu-facturing tolerance must be allowed for. 5% is the appropriate allowance for a full core map taken with the incore detector flux mapping system, and 3% is the appropriate allowance for manufacturing tolerance.

N When F is measured, experimental error must be allowed for, and 4%

is the appropriate allowance for a full core map taken with the incore detection system. This 4% measurement uncertainty has )een included in the design DNBR limit value. The specified limit for F also contains an additional 4% allowance for uncertainties. The total allowance is based on the following considerations:

D. C. COOK - UNIT 1 B 3/4 2-4 AMENDMENT NO.

POWER DISTIBUTION LIMITS BASES

a. abnormal perturbations in the radial power shape, such as from rod misalignment, affect F H

more =directly than F

b. although rod movement has a direct influence upon limiting F to within its limit, such control is not readily available to lz it F and

c. errors in prediction for control power shape detected during startup PHYSICS TESTS can be compensated for in F , by restricting axial flux distributions. This compensation for F H is less readily available.

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

The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x-y plane power tilts. A limiting tilt of 1.025 can be tolerated before the margin for uncertainty in F is depleted. The limit of 1.02 was selected to provide an allowance for tPe uncertainty associated. with the indicated power tilt.

The two hour time allowance for operation with a tilt condition greater than 1.02, but less than 1.09, is provided to allow identification and correction of a dropped or misaligned rod. In the event such action does not correct the tilt, the margin for uncertainty on F is reinstated by reducing the power by 3 percent for each percent of tilt i8 excess of 1.0.

D. C. COOK - UNIT 1 B 3/4 2-5 AMENDMENT NO.

POWER DISTRIBUTION LIMITS BASES 3 4 2 5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated to be adequate to maintain the applicable design limit DNBR values for each fuel type (which are listed in the bases for Section 2.1.1) throughout each analyzed transient. The indicated values of T and flow include allowances for instrument errors. Measurement uncertaintzef ave have been accounted for in determining the DNB parameters'imit values.

The 12 hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. The 12-hour surveillance of the RCS flow measurement is adequate to detect flow degradation. The CHANNEL CALIBRATION performed after refueling ensures the accuracy of the 12-hour surveillance of the RCS flow measurement. The total flow is measured after each refueling based on a secondary side calorimetric and measurements of primary loop temperature.

3 4 2 6 ALLOWABLE POWER LEVEL - APL Constant Axial Offset Control (CAOC) operation manages core power distributions such that Technical Specification limits on F (Z) are not violated during normal operation and limits on MDNBR are no9 violated during steady-state, load-follow, and anticipated transients. The V(Z) factor given in the Peaking Factor Limit Report and applied by the Technical Specifications provides the means for predicting the maximum F (Z) distribution anticipated during operation using CAOC taking into account the incore measured equilibrium power distribution. A comparison of the maximum F (Z) with the Technical Specification limit determines the power level (APL) below which the Technical Specification limit can be protected by CAOC. This comparison is done by calculating APL, as defined in specification 3.2.'6.

D. C. COOK - UNIT 1 B 3/4 2-6 AMENDMENT NO.

INSTRUMENTATION This page intentionally leEt blank.

D. C. COOK - UNIT l B 3/4 3-3 AMENDMENT NO.

3 4 4 REACTOR COOLANT SYSTEM BASES 3 4 4 1 REACTOR COOLANT LOOPS The plant is designed to operate with all reactor coolant loops in operation, and maintain DNBR above 1.69 during all normal operations and anticipated transients. A loss of flow in two loops will cause a reactor if trip. 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 170 0 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 t'e water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into or (2) by restricting starting of the RCP's to when 0 the secondary water temperature of each steam generator is less than 50 F above each of the RCS cold leg temperatures.

D. C. COOK - UNIT 1 B 3/4 4-1 AMENDMENT NO.

REACTOR COOLANT SYSTEM BASES 3 4 4.2 and 3 4 4 3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 420,000 lbs per hour of saturated steam at the valve set point. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown.

In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.

During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip set point is reached (i.e., no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves or steam dump valves.

Demonstration of the safety valves'ift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code,, 1974 Edition.

3 4 4 4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a hydraulically solid system and is capable of accomodating pressure surges during operation. The steam bubble also protects the pressurizer code safety valves and power operated relief valves against water relief. The power operated relief valves and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the power operated relief valves minimizes the undesirable opening of the spring-loaded pressurizer code safety valves. The requirement that 150 kW of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation conditions.

D. C. COOK - UNIT 1 B 3/4 4-2 AMENDMENT NO.

REACTOR COOLANT SYSTEM BASES 3 4.4 11 RELIEF VALVES The power operated relief valves (PORVs) operate to relieve RCS pressure below the setting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should the relief valve become inoperable. The electrical power for both the relief valves and the block valves is supplied from an emergency power source to ensure the ability to seal this possible RCS leakage path.

3 4.4.12 REACTOR COOLANT VENT SYSTEM The Reactor Coolant Vent System is provided to exhaust noncondensible gases and/or steam from the primary system that could inhibit natural circulation core cooling. It has been designed to vent a volume of Hydrogen approximately equal to one-half of the Reactor Coolant System volume in one hour at system design pressure and temperature.

The Reactor Coolant Vent System is comprised of the Reactor Vessel head vent system and the pressurizer steam space vent system. Each of these subsystems consists of a single line containing a common manual isolation valve inside containment, splitting into two parallel flow paths. Each flow path provides the design basis venting capacity and contains two lE DC powered solenoid isolation valves, which will fail closed. This valve configuration/redundancy serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a remotely-operated vent valve, power supply, or control system does not prevent isolation of the vent path. The pressurizer steam space vent is independent of the PORVs and safety valves and is specifically designed to exhaust gases from the pressurizer in a very high radiation environment. In addition, the OPERABILITY of one Reactor Vessel head vent path and one Pressurizer steam space vent path will ensure that the capability exists to perform this venting function.

The function, capabilities, and testing requirements of the Reactor Coolant Vent System are consistent with the requirements of Item II.B.1 of NUREG-0737, "Clarification of TMI Action Plan Requirement," November 1980.

The minimum required systems to meet the Specification and not enter into an action statement are one vent path from the Reactor Vessel head and one vent path from the Pressurizer steam space.

D. C. COOK -. UNIT 1 B 3/4 4-13 AMENDMENT NO.

EMERGENCY CORE COOLING SYSTEMS BASES 3 4.5 5 REFUELING WATER STORAGE TANK The OPERABILITY of the RWST as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. 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 mixing of the RWST and the RCS water volumes with all control rods inserted except for the most reactive control assembly. 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 0 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. The value of the minimum RWST temperature in Technical Specification 3.5.5 has been conservatively changed to 80 F to increase the consistency between Units 1 and 2. The lower RWST temperature results in lower containment pressure from containment spray and safeguards flow assumed to exit the break. Lower containment pressure results in increased flow resistance of steam exiting the core thereby slowing reflood and increasing PCT.

D. C. COOK - UNIT 1 B 3/4 5-3 Amendment No.

3 4 PLANT SYSTEMS BASES 3 4 7 1 TURBINE CYCLE 3 4 7 1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within its design pressure of 1085 psig during the most severe anticipated syst'm opera-tional transient. The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).

The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code, 1971 Edition. The total relieving capacity for all valves on all of the steam lines is 17,153,800 lbs/hr which is approximately 121 percent of the total secondary steam flow of 14,120,000 lbs/hr at 100$

RATED THERMAL POWER. A minimum of 2 OPERABLE safety valves per operable steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-1.

STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Range Neutron Flux channels. The reactor trip setpoint reductions are derived on the fol'lowing bases:

For 4 loop operation SP x (109)

X Where:

SP reduced reactor trip setpoint in percent of RATED THERMAL POWER V - maximum number of inoperable safety valves per steam line 1, 2 or 3.

X - Total relieving capacity of all safety valves per steam line 4,288,450 lbs/hour.

Y Maximum relieving capacity of any one safety valve

- 857,690 lbs/hour.

(109) Power Range Neutron Flux-High Trip Setpoint for 4 loop operation.

D. C. COOK - UNIT 1 B 3/4 7-1 AMENDMENT NO ~

PLANT SYSTEMS BASES 3 4 7 1 2 AUXILIARYFEEDWATER SYSTEM The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 350 0 F from normal operating conditions in the event of a total loss of off-site power.

Each electric driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 450 gpm at a pressure of 1065 psig to the entrance of the steam generators. The steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 900 gpm at a pressure of 1065 psig to the entrance of the steam generators. This capacity is sufficient to ensure that adequate feedwater flow's available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350 0 F when the Residual Heat Removal System may be placed into operation.

The acceptance discharge pressures for the auxiliary feedwater pumps are based on a fluid temperature of 60 0 F. Water density corrections are permitted to allow comparison of test results which vary depending on ambient conditions.

In addition to its safety design function, the AFW system is used to maintain steam generator level during startup (including low power operation). During this time, the system design allows for automatic initiation of the auxiliary feedwater pumps and their related automatic valves in the flow path.

D. C. COOK - UNIT 1 B 3/4 7-2 AMENDMENT NO.

3 4 9 REFUELING OPERATIONS BASES 3 4 9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that:

1) the reactor will remain subcritical during CORE ALTERATIONS, and 2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. These limitations are consistent with the initial conditions assumed for the boron dilution incident in the accident analyses. The value of 0.95 or less for K includes a 1 percent delta k/k conservative allowance for uncertain5fes.

Similarly, the boron concentration value of 2000 ppm or greater includes a conservative uncertainty allowance of 50 ppm boron. The boron concentration requirement of specification 3.9.1.b has been conservatively increased to 2400 ppm to agree with the minimum concentration of the RWST.

3 4 9 2 INSTRUMENTATION The OPERABILITY of the source range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

3 4.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor pressure vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products. This decay time is consistent with the assumptions used in the accident analyses.

3 4 9.4 CONTAINMENT BUILDING PENETRATIONS The requirements on containment building penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE.

3 4.9 5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATIONS.

D. C. COOK - UNIT 1 B 3/4 9-1 AMENDMENT NO.

670 UNACCEPTABLE OPERATION 650 2A'00 ps 2290 p

~lg 630 22g0

) ~lg 2040 0

ps

194'0 610 (Z

f-ps'8<0 I

590 ps'70 ACCEPTABLE OPERATION 550

0. 0 0.2 0.4 0.6 0.8 1.0 1.2 FRACTIGN GF RATED THERNRL PQNER PRESSURE (PSIA) BREAKPOINTS (FRACTION RATED THERMAL POl!ER, T AVG, DEG F) 1840 (0.00,616.2) , (0.98,585.1) (1.20,556.5) 1940 (0.00,623.8) , (0.93,594.7) (1.20,563.5) 2040 (0.00,631.0) , (0.88)603.8) (1.20,569.6) 2250 (0 00,645.9)

F

, {0.80,622.3) ) {1.20,580.9) 2290 (0.00,647.9) , (0.80,624.5) (1.20,586.5) 2440 (0.00,657.4) , (0.77,635.6) (1.20,597.2)

FIGURE 2.l-l Reactor Core Safety Limits Four Loops in Operation C. COOK - U'HIT 2 2-2 A,4EHDIKNT NO ~

REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.1.2

~ ~ The SHUTDOWN MARGIN shall be:

a. In MODE 4:

1. Greater than or equal to 1.68 Ak/k when operating with one or more Reactor Coolant Loops in accordance with Specification 3.4.1.3.

2. Greater than the value shown in Figure 3.1-3 when operating with no Reactor Coolant Loops but one or more Residual Heat Removal Loops in accordance with Specification 3.4.1.3.

b. In MODE 5:

1. Greater than or equal to 1.0$ Ak/k when operating with one or more Reactor Coolant Loops in accordance with Specification 3.4.1.3.

2. Greater than the value shown in Figure 3 '-3 when operating with no Reactor Coolant Loops but one or more Residual Heat Removal Loops in accordance with Specification 3 '.1.3.

APPLICABILITY: MODES 4 and 5 ACTION:

With SHUTDOWN MARGIN less than the above limits, immediately initiate and continue boration at greater than or equal to 10 gpm of 20,000 ppm boric acid solution or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE RE UIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to the above limits:

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 SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable control rod(s).

D. CD COOK - UNIT 2 3/4 1-3 AMENDMENT NO.

REACTIVITY CONTROL SYSTEMS BORON DILUTION LIMITING CONDITION FOR OPERATION 3.1.1.3 The flow rate of reactor coolant through the reactor coolant system shall be greater than or equal to 2000 gpm whenever a reduction in Reactor Coolant System boron concentration is being made.*

APPLICABILITY: ALL MODES.

ACTION:

With the flow rate of reactor coolant through the reactor coolant system less than 2000 gpm, immediately suspend all operations involving a reduction in .

boron concentration of the Reactor Coolant System.

SURVEILLANCE RE UIREMENTS 4.1.1.3 The flow rate of reactor coolant through the reactor coolant system shall be determined to be greater than or equal to 2000 gpm within one hour prior to the start of and at least once per hour during a reduction in the Reactor Coolant System boron concentration by either:

a. Verifying at least one reactor coolant pump is in operation, or

,b. Verifying that at least one RHR pump is in operation and supplying greater than or equal to 2000 gpm through the reactor coolan't I system.

For purposes of this specification, addition of water from the RWST does not constitute a dilution activity provided the boron concentration in the RWST is greater than or equal to the minimum required by specification 3.1.2.8.b.2 (MODES 1, 2, 3, and 4) or 3.1;2.7.b.2 (MODES 5 and 6).

D. C. COOK - UNIT 2 3/4 1-4 AMENDMENT NO.

REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.4 The moderator temperature coefficient (MTC) shall be:

a. Within the region of acceptable operation in Figure 3.1-2, and

-4

b. Less negative than -3.9 x 10 hk/k/o F for the all rods, withdrawn, end of cycle life (EOL), RATED THERMAL POWER condition.

APPLICABILITY: Specification 3.1.1.4.a - MODES 1 and 2* only¹ Specification 3.1.1.4.b - MODES 1, 2 and 3 only¹ ACTION'ith the MTC more positive than the limit of 3.1.1.4.a above:

1. Establish and maintain control rod withdrawal limits sufficient to restore the MTC to within its limits within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6.

2. Maintain the control rods within the withdrawal limits established above until subsequent measurement verifies that the MTC has been restored to within its limit for the all rods withdrawn condition.

3. Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 10 days describing the value of the measured MTC, the interim control rod withdrawal limits and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.

With the MTC more negative than the limit of 3.1.1.4b above, be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ' With K efff greater than or equal to 1.0

¹ See Special Test Exception 3.10.3 D. C. COOK - UNIT 2 3/4 1-5 AMENDMENT NO.

REACTIVITY CONTROL SYSTEMS SURVEILLANCE RE UIREMENTS 4.1.1.4 The MTC shall be determined to be within its limits during each fuel cycle as follows:

a. The MTC shall be measured and compared to the BOL limit of Specification 3.1.1.4.a, above, prior to initial operation above 5%

of RATED THERMAL POWER, after each fuel loading.

b. The MTC shgll be measured at any THERMAL POWER and compared to

-3.0 x 10 Ak/k/ F (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching a RATED THERMAL POWER equilibrium boron concentration of 300 ppm. In the event thip comparison indicates the MTC is more negative than -3.0 x 10 hk/k/ F, the MTC shall be remeasured, and compared to the EOL MTC limit of Specification 3.1.1.4.b, at least once per 14 EFPD during the'remainder of the fuel cycle.

D. C. COOK - UNIT 2 3/4 1-6 AMENDMENT NO.

FIGURE 3.1 2 A

oO Moderator Temperature Coefficient (MTC)

I MTC x 10 hk/k/deg.F g 1.00 Una eptabe Oper tion 0.50 0.00

-0.50

-1.00 Acc ptable Opera on

-1.50

-2.00

-2.50

-3.00 0 10 20 30 40 50 60 70 80 90 100 PERCENT RATED THERMAl POWER

REACTIVITY CONTROL SYSTEMS 3 4.1.2 BORATION SYSTEMS FLOW PATHS - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.1 As a minimum, one of the following boron injection flow paths shall be OPERABLE:

a. A flow path from the boric acid tanks via a boric acid transfer pump and charging pump to the Reactor Coolant Syst: em if only the boric acid storage tank in Specification 3.1.2.7a is OPERABLE, or

b. The flow path from the refueling water storage tank via a charging pump to the Reactor Coolant System if only the refueling water storage tank in Specification 3.1.2.7b is OPERABLE.

APPLICABILITY: MODES 5 and 6.

ACTION:

With none of the above flow paths OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes* until at least one injection path is restored to OPERABLE status.

SURVEILLANCE RE UIREMENTS 4.1.2.1 At least one of the above, required flow paths shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that the temperature of the heat0 traced portion of the flow path is greater than or equal to 145 F when a flow path from the boric acid tanks is used.

b. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

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.

D. C. COOK - UNIT 2 3/4 1-8 AMENDMENT NO.

REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION

3. 1. 2. 3 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.

APPLICABILITY: MODES 5 and 6.

ACTION:

a. With no charging pump OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.*

b. With more than one charging pump OPERABLE or with a safety injection pump(s) OPERABLE0 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.

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 discharge pressure of greater than or equal to 2390 psig when tested pursuant to Specification 4.0.5.

4.1.2.3.2 All charging pumps and safety injection pumps, excluding the above required OPERABLE charging pump, shall be demonstrated inoperable by verifying that the motor circuit breakers have been removed from their electrical power supply circuits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , except when:

a. The reactor vessel head is removed, or

b. The temperature of all RCS cold legs is greater than 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.

D. C. COOK - UNIT 2 3/4 1-11 AMENDMENT NO.

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.7 As a minimum, one of the following boxated water sources shall be OPERABLE:

a ~ A boric acid storage system and associated heat tracing with:

l. A minimum usable borated water volume of 4300 gallons,

2. Between 20,000 and 22,500 ppm of boron, and 0

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 minimum solution temperature of 80 F.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS 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 shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1. Verifying the boron concentration of the water,

2. Verifying the contained borated water volume, and

3. 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 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.

D. CD COOK - UNIT 2 3/4 1-15 AMENDMENT NO.

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.8 Each of the following borated water sources shall be OPERABLE:

a. 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 0

3. A minimum solution temperature of 145 F.

b. 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 0

3. A minimum solution temperature of 80 F.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. 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% hk/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:

D. C. COOK - UNIT 2 3/4 1-16 AMENDMENT NO.

REACTIVITY CONTROL SYSTEMS SURVEILLANCE RE UIREMENTS Continued

a. At least once per 7 days by:

1. Verifying the boron concentration in each water source,

2. Verifying the contained borated water volume of each water source, and

3. Verifying the boric acid storage system solution temperature.

b. At least once per 24 hours by verifying the RWST temperature.

D. C. COOK - UNIT 2 3/4 1-17 AMENDMENT NO.

REACTIVITY CONTROL SYSTEMS 3 4 1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3 ' All full length (shutdown and control) rods shall be OPERABLE and positioned within + 12 steps (indicated position) of their group step counter demand position.

APPLICABILITY: MODES 1* and 2*

ACTION:

a. With one or more full length rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With more than one full length rod inoperable or misaligned from the group step counter demand position by more than + 12 steps (indicated position), be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c ~ With one full length rod inoperable due to causes other than addressed by ACTION a, above, or misaligned from its group step counter demand height by more than + 12 steps (indicated position),

POWER OPERATION may continue provided that within one hour either:

1. The affected rod is restored to OPERABLE status within the above alignment requirements, or

2. The affected rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that:

a) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions, and b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and

See Special Test Exceptions 3.10.2 and 3.10.3 D. C. COOK - UNIT 2 3/4 1-18 AMENDMENT NO.

REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION Continued c) A power distribution map gs obtained from the movable incore detectors and F (Z) and F> are verified to be within their limits within 79 hours9.143519e-4 days 0.0219 hours 1.306217e-4 weeks 3.00595e-5 months , an%

d) Either the THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within one hour and within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the high neutron flux trip stepoint is re-duced to -less than or equal to 85% of RATED THERMAL POWER, or e) The remainder of the rods in the group with the inoperable rod are aligned to within + 12 steps of the inoperable rod within one hour while maintaining the rod sequence and inser-tion limits of Figure 3.1-1; the THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during sub-sequent operation.

SURVEILLANCE RE UIREMENTS 4.1.3.1.1 The position of each full length rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Rod Position Deviation 'Monitor is inoperable, then verify the group positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.1.3.1.2 Each full length rod not fully inserted shall be determined to be OPERABLE by movement of at least 8 steps in any one direction at least once per 31 days.

D. CD COOK - UNIT 2 3/4 1-19 AMENDMENT NO.

II REACTIVITY CONTROL SYSTEMS ROD DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full length (shutdown and control) rod drop time from the fully withdrawn position (228 steps) shall be less than or equal to 2.2 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:

0

a. T avg greater than or equal to 541 F, and

b. All reactor coolant pumps operating.

APPLICABILITY: MODES 1 and 2.

ACTION:

With the drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

SURVEILLANCE RE UIREHENTS 4.1.3.4 The rod drop time of full length rods shall be demonstrated through measurement prior to entering MODE 2:

a. For all rods following each removal of the reactor vessel head,

b. For specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and

c. At least once per 18 months.

D. C. COOK - UNIT 2 3/4 1-23 AMENDMENT NO.

REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be fully withdrawn (228 steps).

APPLICABILITY: MODES 1* and 2*¹ ACTION:

With a maximum of one shutdown rod not fully withdrawn, except for surveil-lance testing pursuant to Specification 4.1.3.1.2, within one hour either:

a. Fully withdraw the rod, or

b. Declare the rod to be inoperable and apply Specification 3.1.3.1.

SURVEILLANCE RE UIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be fully withdrawn:'.

Within 15 minutes prior to withdrawal of any rods in control banks A, B, C or D during an approach to reactor criticality, and

b. At least once per 12 hours thereafter.

See Special Test Exceptions 3.10.2 and 3.10.3

¹With K eff .greater than or equal to 1.0 C. COOK - UNIT 2 3/4 1-24 'MENDMENT NO.

REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as shown in Figure 3.1-1.

APPLICABILITY: MODES 1* and 2*¹.

ACTION'ith the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, either:

a. Restore the control banks to within the limits within two hours, or

b. Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the group position using the above figure, or cd Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE RE UIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

See Special Test Exceptions 3.10.2 and 3.10.3

¹ With K efff greater than or equal to 1.0.

D. CD COOK - UNIT 2 3/4 1-25 AMENDMENT NO.

REMOVE THIS PAGE D. C. COOK - UNIT 2 3/4 1-27 AMENDMENT NO.

3 4 2 POWER DISTRIBUTION LIMITS AXIAL FLUX DIFFERENCE AFD LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the target band (+5% or +3% flux difference units) about a target flux difference.

APPLICABILITY: MODE 1 above 50% RATED THERMAL POWER*

ACTION:

a. With the indicated AXIAL FLUX DIFFERENCE outside of the target band about the target flux difference and with THERMAL POWER:

l. Above 90% of 0.9 x APL (whichever is less) of RATED THERMAL POWER, within 15 minutes:

a) Either restore the indicated AFD to within the target band limits, or b) Reduce THERMAL POWER to less than 90% or 0.9 x APL (whichever is less) of RATED THERMAL POWER.

2. Between 50% and 90% or 0.9 x APL (whichever is less) of RATED THERMAL POWER:

a) POWER OPERATION may continue provided:

1) The indicated AFD has not been outside of the target band for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months penalty deviation cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and

2) The indicated AFD is within the limits shown on Figure 3.2-1. Otherwise, reduce THERMAL POWER to less than 50%

of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b) Surveillance testing of the Power Range Neutron Flux Channels may be performed pursuant to Specification 4.3.1.1.1 provided the indicated AFD is maintained within the limit of Figure 3.2-1. A total of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months operation may be accumulated with the AFD outside of the target band during this testing without penalty deviation.

See Special Test Exception 3.10.2 D. C. COOK - UNIT 2 3/4 2-1 AMENDMENT NO.

FIGURE 3.Z I ALLO'IIIIABLE DEVIATION

- FROM TARGET FLUX DIFFERENCE 8120 I I i

g

! $ E R

5100 I

'nocceptable,'( 0,90) (I+10 90) Unocce 'oble

-Operatian- '

A

'(+8, go)

' P X 80

+3%! Target

.L ..

t t,

I

'290)

+5% i4'arget Bond=I= j nd I

~ 60l I AcceptableL Oper'ation !

I 5,50) I+25,50)

(( ) ('+(23,50 g 40 t

tQ l

20 l I

I 0L I

(

20 ego 30 -10 0 10 20 Deviation from Target Flux Difference O

TABLE 3 2-1 DNB PARAMETERS LIMITS PARAMETER 4 Loo s in 0 eratio Reactor Coolant System T ( 0 576.3 F. (indicated) avg Pressurizer Pressure > 2205 psig* **

Reactor Coolant System Total Flow Rate > 138.6 x 10 6

lbs/hr Limit not applicable during either a THERMAL POWER ramp in excess of 5S RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10%

RATED THERMAL POWER.

Indicated average of at least three OPERABLE instrument loops.

3.5% penalty for measurement uncertainty included in this value.

D. C. COOK - UNIT 2 3/4 2-16 AMENDMENT NO.

TABLE 3 2-2 DNB PARAMETERS

~PAEAMETE LIMIT Reactor Coolant System T < 549.2 0 F. (Reactor Subcritical) avg Reactor Coolant System T avg

< 576.3 0 F. (Reactor Critical)

Pressurizer Pressure > 2176 psig Reactor coolant loop operational requirements are contained in Specifications 3.4.1.1, 3.4.1.2.c and 3.4.1.3.c.

Indicated average of at least three OPERABLE instrument loops.

D. C. COOK - UNIT 2 3/4 2-18 AMENDMENT NO.

POWER DISTRIBUTION LIMITS ALLOWABLE POWER LEVEL - APL LIMITING CONDITION FOR OPERATION 3.2.6 THERMAL POWER shall be less than or equal to ALLOWABLE POWER LEVEL (APL), given by the following relationships:

Westin house Fuel P

Exxon Nuclear Co Fuel P

> F (Z) is the measured hot channel factor, including a 3%

m9nufacturing tolerance uncertainty and a 5% measurement uncertainty.

~ V(Z) is the function defined in Figure 3.2-3 which corresponds to the target band.

~ F 1.00 except when successive steady-state power distribution mRps indicate an increase in peak pin power, F , with exposure.

Then either of the following penalties, F , shoal be taken:

P F 1.02 or, P

F 1.00 provided that Surveillance Requirement 4.2.6.2 is sRtisfied once per 7 Effective Full Power Days until 2 successive maps indicate that the peak pin F H is not increasing.

~ The above limit is not applicable in the following core regions.

1) Lower core region 0% to 108 inclusive.

2) Upper core region 90% to 100% inclusive.

APPLICABILITY: MODE 1 D. C. COOK - UNIT 2 3/4 2-19 AMENDMENT NO.

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION Continued ACTION'ith THERMAL POWER exceeding APL:

'a ~ Reduce THERMAL POWER to APL or less of RATED THERMAL POWER within 15 minutes. Then reduce the Power Range Neutron Flux-High Trip Setpoints by the same percentage which APL is below RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced the same percentage which APL is below RATED THERMAL POWER.

b. THERMAL POWER may be increased to a new APL calculated at the reduced power by either redefining the target axial flux difference or by correcting the cause of the high F (Z) condition.

SURVEILLANCE RE UIREMENTS 4.2.6.1 The provisions of Specification 4.0.4 are not applicable.

4.2.6.2 APL shall be determined by measurement in conjunction with the target flux difference and target band determination+ above 15$ of RATED THERMAL POWER, according to the following schedule:

a. Upon achieving equilibrium conditions after exceeding by 10% or more of RATED THERMAL POWER, the THERMAL POWER at which APL was last determined**, or

b. At least once per 31 effective full power days, whichever occurs first.

+APL can be redefined by remeasuring the target axial flux difference in accordance with ACTION statement b of Specification 3.2.6.

- During power escalation at the beginning of each cycle, the design target may be used until a power level for extended operation has been achieved.

D. C. COOK - UNIT 2 3/4 2-20 AMENDMENT NO.

TABL 3-1 Continued REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

9. Pressurizer Pressure-Low 1, 2

10. Pressurizer Pressure High 1i 2

11. Pressurizer Water Level High 1( 2

12. Loss of Flow Single Loop 3/loop 2/loop in 2/loop in (Above P-8) any opera- each opera-ting loop ting loop

13. Loss of Flow Two Loops 3/loop 2/loop in 2/loop in (Above P-7 and below P-8) two opera- each opera-ting loops ting loop

14. Steam Generator Water 3/loop 2/loop in 2/loop in Level 1, 2 Low-Low any opera- each opera-ting loop ting loop

15. Steam/Feedwater Flow 2/loop-level 1/loop-level 1/loop-level 1, 2 Mismatch and Low Steam and coincident and Generator Water Level 2/loop-flow with 2/loop-flow mismatch in 1/loop-flow mismatch or same loop mismatch in 2/loop-level same loop and 1/loop-flow mismatch x

O

TABLE 3.3-1 Continued REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE oO FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION I

16. Undervoltage-Reactor Coolant 4-1/bus Pumps

17. Underfrequency-Reactor 4-1/bus Coolant Pumps

18. Turbine Trip A. Low Fluid Oil Pressure B. Turbine Stop Valve Closure 6

19. Safety Injection Input 1, 2 from ESF

20. Reactor Coolant Pump Breaker Position Trip A. Above P-8 1/breaker 1/breaker 10 B. Above P-7 and below P-8 1/breaker 1/breaker 11 per oper-ating loop

21. Reactor Trip Breakers 1, 2 1, 13 3* 4* 5* 14

22. Automatic Trip Logic 1, 2 1 3* 4* 5* 14

TABLE 4.3-1 Continued REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS no CHANNEL MODES IN WHICH O CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST RE UIRED I

13. Loss of Flow-Two Loops R(8) N.A.

14. Steam Generator Water Level- 1, 2 Low-Low

15. Steam/Feedwater Flow Mismatch and S 1, 2 Low Steam Generator Water Level

16. Undervoltage-Reactor Coolant N.A.

Pumps

17. Underfrequency-Reactor Coolant N.A.

Pumps

18. Turbine Trip A. Low Fluid Oil Pressure N.A. N.A. S/U (1) 1, 2 B. Turbine Stop Valve Closure N.A. N.A. S/U(1) 1, 2

19. Safety Injection Input from ESF N.A. N.A. M(4) 1, 2

20. Reactor Coolant Pump Breaker N.A. N.A. N.A.

Position Trip

21. Reactor Trip Breaker A. Shunt Trip Function N.A. N.A. M(5) (ll) and S/U(l) (ll) 3* 4* 5*

O B. Undervoltage Trip Function N.A. N.A. M(5) (ll) and S/U(l) (ll) 3* 4* 5*

~ ~

22. Automatic Trip Logic N.A. N.A. M(5) 1 2 3* 4* 5*

23. Reactor Trip Bypass Breaker N.A. N.A. M(12) and S/U(l)(13) ] 2 3* 4* 5*

TABLE 3.3-5 Continued ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

6. Steam Line Pressure--Low

a. Safety Injection (ECCS) < 12.0¹/24.0¹¹

b. Reactor Trip (from SI) < 2.0

c. Feedwater Isolation < 8.0 Containment Isolation-Phase "A" < 18.0¹/28.0¹¹

e. Cont'ainment Purge and Exhaust Isolation Not Applicable

f. Motor Driven Auxiliary Feedwater Pumps < 60 '

g. Essential Service Water System < 14.0¹/48.0¹¹

h. Steam Line Isolation < 8.0

7. Containment Pressure--Hi h-Hi h

a. Containment Spray < 45.0

b. Containment Isolation-Phase "B" Not Applicable

c. Steam Line Isolation < 7.0

d. Containment Air Recirculation Fan < 600.0

8. Steam Generator Water Level--Hi h-Hi h

a. Turbine Trip Not Applicable

b. Feedwater Isolation Not Applicable

9. Steam Generator Water Level--Low-Low

a. Motor Driven Auxiliary Feedwater Pumps < 60 '

b. Turbine Driven Auxiliary Feedwater Pumps < 60.0

10. 4160 volt Emer enc Bus Loss of Volta e

a. Motor Driven Auxiliary Feedwater Pumps < 60.0

11. Loss of Main Feedwater Pum s

a. Motor Driven Auxiliary Feedwater Pumps < 60.0

12. Reactor Coolant Pum Bus Undervolta e

a. Turbine Driven Auxiliary Feedwater Pumps < 60.0 D. C. COOK - UNIT 2 3/4 3-28 AMENDMENT NO.

REACTOR COOLANT SYSTEM HOT STANDBY LI ITI G CONDITION FOR OPERATION 3.4.1.2 a. The reactor coolant loops listed below shall be OPERABLE and in operation as required by items b, c, and d:

1. Reactor Coolant Loop 1 and it's associated steam generator and reactor coolant pump,

2. Reactor Coolant Loop 2 and its associated steam generator and reactor coolant pump,

3. Reactor Coolant Loop 3 and its associated steam generator and reactor coolant pump,

4. Reactor Coolant Loop 4 and its associated steam generator and reactor coolant pump.

b. At least two of the above coolant loops shall be OPERABLE and at least one loop in operation if the reactor trip breakers are in the open position, or the control rod drive system is not capable of rod withdrawal.*

At least three of the above coolant loops shall be OPERABLE and in operation when the reactor trip system breakers are in the closed position and the control rod drive system is capable of rod withdrawal.

d. At least three of the above coolant loops shall be OPERABLE and in operation above P-12. (Refer to Technical Specification 3.3.2.1, Table 3.3-3 for instrumentation requirements.)

APPLICABILITY: MODE 3 ACTION:

a. With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With less than the number of operating coolant loops required by item c above, restore the required number of coolant loops within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or open the reactor trip breakers.

With less than the number of operating coolant loops required by item d above, restore the required number of coolant, loops within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or lower the reactor coolant system temperature below P-12.

D. C. COOK - UNIT 2 3/4 4-2 AMENDMENT NO.

REACTOR COOLANT SYSTEM HOT STANDBY LIMITING CONDITION FOR OPERATION Continued

d. With no reactor coolant loop in operation, suspend all operations involving a+geduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

SURVEILLANCE RE UIREMENTS 4.4.1 2.1 At least the above required reactor coolant pumps,

~ if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

4.4.1.2.2 At least one cooling loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

All reactor coolant pumps may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided (1) no operations are permitteg that would cause dilution of the reactor coolant system boron 0

concentration *, and (2) core outlet temperature is maintained at least 10 F below saturation temperature.

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

D. C. COOK - UNIT 2 3/4 4-2a AMENDMENT NO.

REACTOR COOLANT SYSTE SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 a. The coolant loops listed below shall be OPERABLE and in operation as required by items b and c:

1. Reactor Coolant Loop 1 and its associated steam generator and reactor coolant pump,*

2. Reactor Coolant Loop 2 and its associated steam generator and reactor coolant pump,*

3. Reactor Coolant Loop 3 and its associated steam generator and reactor coolant, pump,*

4. Reactor Coolant Loop 4 and its associated steam generator and reactor coolant pump,*

5. Residual Heat Removal - East, **

6. Residual Heat Removal - West **

b. At least two of the above coolant loops shall be OPERABLE least one loop in operation if the reactor trip breakers are in the open position, or the control rod drive system is not and't capable of rod withdrawal.***

At least three of the above reactor coolant loops shall be OPERABLE and in operation when the reactor trip system breakers are in the closed position and the control rod drive system is capable of rod withdrawal.

APPLICABILITY: MODES 4 and 5 ACTION:

With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; be in COLD SHUTDOWN within 20 hours.

b. With less than the number of operating coolant loops required by item c above, restore the required number of coolant loops within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or open the reactor trip breakers.

With no coolant loop in operation, suspend all operations involving a reggggion in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

D. C. COOK - UNIT 2 3/4 4-3 AMENDMENT NO.

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 + 1%.*

ACTION:

With no pressurizer code safety valve OPERABLE:

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

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

D. C. COOK - UNIT 1 3/4 4-4 AMENDMENT NO.

REACTOR COOLANT SYSTEM RELIEF VALVES - OPERATING LIMITING CONDITION FOR OPERATION 3.4.11 Three power operated relief valves (PORVs) and their associated block valves shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

PORVs inoperable:*

1. With one PORV inoperable, within 1 hour either restore the inoperable PORV to OPERABLE status or close the associated block valve and remove power from the block valve; otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

2. With two PORVs inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore at least one of the inoperable PORVs to OPERABLE status or close the associated block valves and remove power from the'lock valves; restore at least one of the inoperable PORVs to OPERABLE status within the following 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 and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3. With three PORVs inoperable, within 1 hour either restore at least one of the PORVs to OPERABLE status or close their associated block valves and remove power from the block valves and be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Block valves inoperable: *

1. With one block valve inoperable, within 1 hour either (1) restore the block valve to OPERABLE status, or (2) close the block valve and remove power from the block valve, or (3) close the associated PORV and remove power from the associated solenoid valve; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

PORVs isolated to limit RCS leakage through their seats and the block valves shut to isolate this leakage are not considered inoperable.

DE C. COOK - UNIT 2 3/4 4-32 AMENDMENT NO.

3 4 5 EMERGENCY CORE COOLING SYSTEMS ECCS ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1 Each reactor coolant system accumulator shall be OPERABLE with:

a. The isolation valve open,

b. A contained borated water volume of between 929 and 971 cubic feet,

c. A boron concentration between 2400 ppm and 2600 ppm, and

d. A nitrogen cover-pressure of between 599 and 644 psig.

APPLICABILITY:. MODES 1, 2 and 3.*

ACTION:

a. With one accumulator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within one hour or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in HOT STANDBY within one hour and be in HOT SHUTDOWN within the next.12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE RE UIREMENTS 4.5.1 Each accumulator shall be demonstrated OPERABLE:

a. At least once per 12 hours by:

1. Verifying the contained borated water volume and nitrogen cover-pressure in the tanks, and

2. Verifying that each accumulator isolation valve is open.

Pressurizer Pressure above 1000 psig.

D. C. COOK - UNIT 2 3/4 5-1 AMENDMENT NO.

EMERGENCY CORE COOLING SYSTEM 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 0

c. A minimum water temperature of 80 F.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION'ith 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 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.

D. C. COOK - UNIT 2 3/4 5-11 AMENDMENT NO.

ELECTRICAL POWER SYSTEMS SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

a. One circuit between the offsite transmission network and the onsite Class 1E distribution system, and

b. One diesel generator with:

1. A day fuel tank containing a minimum volume of 70 gallons of fuel,

2. A fuel storage system containing a minimum volume of 42,000 gallons of fuel, and

3. A fuel transfer pump.

APPLICABILITY: MODES 5 and 6.

ACTION:

With less than the above minimum required A.C. electrical power sources OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity*changes*.

SURVEILLANCE RE UIREMENTS 4.8.1.2 The above required A.C. electrical power sources shall be demonstrated OPERABLE by the performance of each of the Surveillance Requirements of 4.8.1.1.1 and 4.8.1.1.2 except for Requirement 4.8.1.1.2.a.5.**

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.

- The provisions of Specification 4.0.6 are applicable.

D. C. COOK - UNIT 2 3/4 8-5 AMENDMENT NO.

3 4 9 REFUELING OPERATION BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 With the reactor vessel head unbolted or removed, the boron concentration of all filled portions of the Reactor Coolant System and .the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions is met:

a Either a K ff of 0.95 or less, which includes a 1% hk/k conservativeeffallowance for uncertainties, or

~

A boron concentration of greater than or equal to 2400 ppm, which includes a 50 ppm conservative allowance for uncertainties.

APPLICABILITY'ODE 6*

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes** and initiate and continue boration at greater than or equal to 10 I gpm of 20,000 ppm boric acid solution or its equivalent until K is reduced to less than or equal to 0.95 or the boron concentration is restored to greater than or equal to 2400 ppm, whichever is the more restrictive. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:

a. Removing or unbolting the reactor vessel head, and

b. Withdrawal of any full length control rod in excess of 3 feet from its fully inserted position within the reactor pressure vessel.

4.9.1.2 The boron concentration of the reactor, coolant system and the refueling canal shall be determined by chemical analysis at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

The reactor shall be maintained in MODE 6 when the reactor vessel head is unbolted or removed.

- 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.

D. C. COOK - UNIT 2 3/4 9-1 AMENDMENT NO.

REFUELING OPERATIONS INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 As a minimum, two source range neutron flux monitors shall be operating, each with continuous visual indication in the control room and one with audible indication in the containment and control room.

APPLICABILITY: MODE 6.

ACTION'ith the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes.* The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.2 Each source range neutron flux monitor shall be demonstrated OPERABLE by performance of:

a. A CHANNEL FUNCTIONAL TEST at least once per 7 days, and

b. A CHANNEL FUNCTIONAL TEST within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to the initial start of CORE ALTERATIONS, and

c. A CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during CORE ALTERATIONS.

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.

D. C. COOK - UNIT 2 3/4 9-2 AMENDMENT NO.

REFUELING OPERATIONS 4.9 8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal loop shall be in operation.

APPLICABILITY: MODE 6.

ACTION:

With less than one residual heat removal loop in operation, except as provided in b. below, suspend all operations involving an increase in the reactor decay heat load or a,reduction in boron concentration of the Reactor Coolant System. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b. The residual heat removal loop may be removed from operation for up to 1 hour per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs.

c. The provisions of Specification 3.0.3 are not applicable.

SURVFILLANCE RE UIREMENTS 4.9.8.1 A residual heat removal loop shall be determined to be in operation and circulating reactor coolant at a flow rate of greater than or equal to 2000 gpm at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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.

D. C. COOK - UNIT 2 3/4 9-8 AMENDMENT NO.

SPECIAL TEST EXCEPTIONS PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specifications 3.1.1.4, 3.1.3.1, 3.1.3.5 and 3.1.3.6 may be suspended during the performance of PHYSICS TESTS provided:

a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER, and The Reactor Trip Setpoints for the OPERABLE Intermediate Range, Neutron Flux and the Power Range, Neutron Flux, Low Setpoint are set at less than or equal to 25% of RATED THERMAL POWER.

APPLICABILITY: MODE 2.

ACTION:

With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immediately open the reactor trip breakers.

SURVEILLANCE RE UIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be less than or equal to 5% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.

4.10.3.2 Each Intermediate and Power Range Channel shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to initiating PHYSICS TESTS.

D. C. COOK - UNIT 2 3/4 10-3 AMENDMENT NO.

SPECIAL TEST EXCEPTION REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION 3.10 ' The limitations of Specification 3.4.1.1 may be suspended during the performance of start up and PHYSICS TESTS provided:

a. The THERMAL POWER does not exceed the P-7 Interlock Setpoint, and

b. The Reactor Trip Setpoints for the OPERABLE Intermediate Range, Neutron Flux and the Power Range, Neutron Flux, Low Setpoint are set at less than or equal to 25% of RATED THERMAL POWER.

APPLICABILITY: During operation below the P-7 Interlock Setpoint.

ACTION:

With the THERMAL POWER greater than the P-7 Interlock Setpoint, immediately open the reactor trip breakers.

SURVEILLANCE RE UIREMENTS 4.10.4.1 The THERMAL POWER shall be determined to be less than the P-7 Interlock Setpoint at least once per hour during startup and PHYSICS TESTS 4.10.4 ' Each Intermediate, Power Range Channel and P-7 Interlock shall be subjected to a CHANNEL FUNCTIONAL TEST wi.thin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to initiating start up or PHYSICS TESTS.

D. C. COOK - UNIT 2 3/4 10-4 AMENDMENT NO.

ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference Specifications:

a. ECCS Actuation, Specifications 3.5.2 and 3.5.3.

b. Inoperable Seismic Monitoring Instrumentation, Unit No. 1, Specification 3.3.3.3.

c. Inoperable Meteorological Monitoring Instrumentation, Unit No. 1, Specification 3.3.3.4.

d. Fire Detection Instrumentation, Specification 3.3.3.8.

e. Fire Suppression Systems, Specifications 3.7.9.1, 3.7.9.2, 3.7.9.3 and 3.7.9.4.

f. Seismic Event Analysis, Specification 4.3.3.3.2.

g. Sealed Source leakage in excess of limits, Specification 4.7.8.1.3.

h. Moderator Temperature Coefficient, Specification 3.1.1.4 D. C. COOK - UNIT 2 6-19 AMENDMENT NO.

3 4 1 REACTIVITY CONTROL SYSTEMS BASES With the RCS average temperature above 200 0 F, a minimum of two separate and redundant boron injection systems are provided to ensure single functional capability in the event an assumed failure renders one of the systems inoperable. Allowable out-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair period.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps and safety injection pumgs, except the required OPERABLE charging pump, to be inoperable below 152 F, unless the reactor vessel head is removed, provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.

The boration capability of either system is sufficient to provide the required SHUTDOWN MARGIN from expected operating conditions after xenon decay 0

and cooldown to 200 F. The maximum expected boration capability usable volume requirement is 3700 gallons of 20,000 ppm borated water from the boric, acid storage tanks or 118,000 gallons of borated water from the refueling water storage tank, The required RWST volume is based on an assumed boron concentration of 2000 ppm. The minimum RWST boron concentration required by See Section B 3/4.5 ';

the post-LOCA long-term cooling analysis is 2400 ppm. The minimum contained RWST volume is based on ECCS considerations.

boration source volume from the boric acid storage tank has conservatively The been increased to 5650 gallons. This value was chosen to be consistent with Unit 1.

With the RCS temperature below 200 0 F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable.

The boron capability required below 200 0 F is. sufficient to provide the required MODE 5 SHUTDOWN MARGIN after xenon decay and cooldown from 200 F to 0

140 F. This condition requires usable volumes of either 4300 gallons of 20,000 ppm borated water from the boric acid storage tanks or 90,000 gallons of borated water from the refueling water storage tank. The value for the boric acid storage tank volume includes sufficient boric acid to borate to 2000 ppm. The required RWST volume is based on an assumed boron concentration of 2000 ppm. The minimum RWST boron concentration required by the post-LOCA long-term cooling analysis is 2400 ppm.

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 mechnical systems and components.

The OPERABILITY of boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6.

D. C. COOK - UNIT 2 B 3/4 1-3 AMENDMENT NO ~

REMOVE THIS PAGE D. C. COOK - UNIT 2 B 3/4 4-la AMENDMENT NO.

3 4 4 REACTOR COOLANT SYSTEM BASES 3 4 4 2 and 3 4 4 3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 420,000 lbs per hour of saturated steam at the valve setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE in MODES 4 and 5, an operating RHR loop, connected to the RCS, provides overpressure relief capability.

Additionally, if no safety valves are OPERABLE, then all Safety Injection pumps and all but one charging pump will be rendered inoperable to preclude overpressurization due to an inadvertent increase in the RCS inventory.

During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig.

The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip set point is reached (i.e., no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves or steam dump valves.

Demonstration of the safety valves'ift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.

3 4 4 4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a hydraulically solid system and is capable of accommodating pressure surges during operation. The steam bubble also protects the pressurizer code safety valves and power operated relief valves against water relief. The power operated relief valves and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the power operated relief valves minimizes the undesirable opening of the spring-loaded pressurizer code safety valves. The requirement that 150 kW of pressurizer heaters and their associated controls.

be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation at HOT STANDBY.

D. C. COOK - UNIT 2 B 3/4 4-2 AMENDMENT NO.

EMERGE CY CORE COOLING SYSTEMS BASES 3 4.5 5 REFUELING WATER STORAGE TA The OPERABILITY of the RWST as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on RWST minimum volume and boron concentration ensures that 1) sufficient water is available within containment to permit recirculation cooling flow to the core, and 2) the reactor will r'emain subcritical in the cold condition following mixing of the RWST and the RCS water volumes with all control rods inserted except for the most reactive control assembly. 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 0

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.

D. C. COOK - UNIT 2 B 3/4 5-3 AMENDMENT NO.

3 4 9 REFUELING OPERATIONS BASES 3 4 9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that:

1) the reactor will remain subcritical during CORE ALTERATIONS, and 2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. These limitations are consistent with the initial conditions assumed for the boron dilution incident in the accident analyses'he value of 0.95 or less for K includes a 1 percent delta k/k conservative allowance for uncertainFfes.

Similarly, the boron concentration value of 2000 ppm or greater includes a conservative uncertainty allowance of 50 ppm boron. The boron concentration requirement of specification 3.9.l.b has been conservatively increased to 2400 ppm to agree with the minimum concentration of the RWST.

3 4 9.2 INSTRUMENTATION The OPERABILITY of the source range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

3 4 9 3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor pressure vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products. This decay time is consistent with the assumptions used in the accident analyses.

3 4 9 4 CONTAINMENT BUILDING PENETRATIONS The requirements on containment building penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE.

3 4 9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATIONS.

D. C. COOK - UNIT 2 B 3/4 9-1 AMENDMENT NO.

Attachment 3 to AEP:NRC:0916W

SUMMARY

OF DONALD C. COOK UNIT 1AND UNIT 2 PROPOSED TECHNICAL SPECIFICATION CHANGES

AEP:NRC: ATZACHMENT 3

SUMMARY

DESCRIPZIONS H)R D- C. COOK UNIT 1 PBQKGED TEQiNICAL SPECIFICATIONS PAGE 1 SECTION +

DES CRIETION R1 MUG 1-7 Definition 1

001 APL made a defined tenn. Editorial change; definition included 1.39 for clarity.

2-1 2.1.1 2

002 Removed reference to Figure Ihree loop operation in Modes 1 and 2

2. 1-2 .and three loop operation. will be pmhibited.

2-3 Figure 2. 1-2 2

003 Figure is retmved. Khree loop operation in Modes 1 and 2 wi11 be prohibited.

2-8 Table 2.2-1 2

004 Parameters for three loop 'Ihree loop operation in Modes 1 and 2 operation are removed. will be prohibited.

2-9 Table 2.2-1 8

005 Words "dT span" are added. 'Ihis change reflects an analysis previously Notes 3 & 4 submitted. See page 2 of Attachment 1 to the letter dated 2hxymt 13, 1985 fram M. P. Alexich to H. R. Denton (Identifier AEP:NRC:0942D) .

To facilitate this review, we are re-transmitting the proprietary attactunent only as Attachment 4 to this letter.

3/4 1-1 3.1.1.1 1

006 ~CABIZZK chmped to MDDES Editorial change to move MDDE 4 SHUZ-1, 2, ard3. IXNN MARGIN Specification to Specification 3.1.1.2.

1 007 Mathematical symbols are written Editorial change for clarity.

out in words.

1 007a Specification title is charged. Editorial change; makes the specifications of both units more similar.

3/4 1-2 4.l.l.l.l.e 1

008 Surveillance changed to MDDE 3 Editorial change to move MDDE 4 SHVZ-only. DCWN MARGIN Surveillance to Surveill-ance 4.1.1.2.b.

NOTES: - 'lhe nuaher in the plus sign (+) column refers to applicable section of Significant Hazards in Attachrent 1.

An asterisk in the asterisk (*) column indicates that the proposed change had been previously approved for Unit 2 The nuaher in the pound sign (g) column is a sequential identifier for each proposed change.

AEP NRC $ ATI'ACHMENT 3

SUMMARY

DESCRIPZIONS FOR D. C. CDDK UNZIP 1 PBDKSED TEKSNICAL SPECIFICATIONS PAGE 2 PAGE SECTION +

DES CRIETION REMARKS 3/4 1-3 3.1.1.2 8

009 Revised to include MDDE 4 and Westinghouse Electric Corporation has 4.1.1.2 MDDE 5 in the same specific- performed. a new analysis for D. C. Cocik ation. Revised Technical Unit 1 similar to that described in the letter fram T. M. Anderson to V. Stello based on dilution acciderrt dated July 8, 1980 (Identifier NS XNK-2273) .

analysis in MDDES 4 and 5. '%his analysis is described in Attachment 14 to this letter. As indicated in Attadment 1 to the letter fram M. P. Alexich to H. R.

Denton dated March 27, 1986 (Identifier AEP:NRC:0916P) g the methodology of NS~-2273 has been in use on Unit 1 since beginning of Cycle 6. Attachment 1 to AEP:NRC:0916P was approved in the SER for Anendment 82 to DPR-74 To facilitate this review, we are also retxansrnitting Attachment 1 to AEP:NRC:0916P and NS IMA-2273 in Attachment 14.

1 010 Mathematical symbols are written Editorial change for clarity.

Gut 1I1 words+

1 010a Specification title is changed. Editorial change; makes the specifications of bath units more similar.

3/4 1-3a 4.1.1.2.b 1

011 Specification 4.1.1.2.b is Editorial change.

mved to new page 3/4 1-3a.

3/4 1-3b Figure 3.1-3 1

012 New figure is added. Editorial conge.

3/4 1-3ag 1

013 Pages added due to length of Editorial change.

3/4 1-3b new specification.

3/4 1-4 3.1.1.3 1 014 "reactor pressure vessel" is Editorial change; makes the Specific-4.1.1.3 changed to "reactor coolant ations of both Units more similar.

system" .

1 015 Mathematical symbols are written Editorial change for clarity.

out in words+

AEP:NRC: ATIACHMENT 3 SUMKQK DESCRIPZIONS FOR D. C. COOK UNIT 1 PBOPOSED TEXZNICAL SPECIFICATIONS PAGE 3 SECTION +

DESCRIPTION REMARKS 8 016 Flmr rate recpivwent reduced An analysis was performed to reduce to 2000 gpm. the ~~red reactor coolant flow rate to 2000 gpm. See Attachment 5 for discussion of heat removal, mixirg, and stratification consideratians. See Attachment 14 for dilution transient cansideratians.

4

017 Footnote added. 'Ihe Technical Specification boran con-centration in the HNST is sufficient to provide adequate shutdown margin fram mgxx:ted operating ccaxKtions.

3/4 1-5; 3.1.1.4 8 018 %he upper limit on MZC for To impmve operational flexibility.

3/4 1-5a Figure 3.1-2 operation above 70% HZP is Justification provided in Attachment 6, changed. %he uppm limit is Item Nznber 4.

near graphically displayed (see Item 020) .

1 019 Mathematical symbols are written Editorial change for clarity.

out 111 words+

1 020 %be new MPC limits proposed in Editorial change.

item 018 are nor graphically displayed in Figure 3. 1-2.

3/4 1-7 3.1.2.1 4 021 Footnote added. 'Ihe Technical Specification boron con-ctmtration in the HNST is sufficient to provide adequate s?mxtdlown margin fram expected operating conditions.

3/4 l-ll 3.1.2.3 4

022 Footnote added. 'Ihe Technical Specification boron can-I centratian in the HNST is sufficient to provide adequate shutdown margin fram expected ogmatmg cccxhtxons.

ACZIQN c 1 022a Editorial change; typographical error correction.

4.1.2-3-1 1 023 Mathematical symbols are written Editorial change for clarity.

out in words.

AEP:NRC: 8 ATrACHMENT 3

SUMMARY

DESCRIPTIONS FOR D. C. COOK UNIT 1 PROKSED TlXHNICALSHKIFICATIONS PAGE 4 SECTION +

DESCRIPTION REMAMS 4.1.2.3.2.b 1 023a Period is added. Editorial change; typographical error correction>>

3/4 1-13 3.1.2.5 4

024 Footnote added. 'Ihe Technical Specification boran concentration in the HNST is sufficient to provide adequate shutdown margin fram expected operating conditions.

4.1.2.5.b 1 025 Mathematical symbols are written Editorial charge for clarity.

out in wards>>

3/4 1-14 3.1.2.6 1 026 <<PIRXUS<< is changed to "status" Editorial ~e; defined term.

"status" is not a 4.1.2.6.b 1 027 Mathematical symbols are written Editorial changes for clarity.

out in words>>

3/4 1-15 3.1.2.7 4 028 Footnote added. le Technical Specification boran concentration in the EST is f

suf icient to provide adecgmte shutdown margin fram expected operating ~tions.

029 (No change for this identifier) .

3.1.2.7.a.l 8

030 Changed BAST and HNST miniman Boration source volumes have been adj-3.1.2.7.b.l volumes. usted to address the shutdown margin recpired for a dilution transient when operating on RHR at the beginning of cycle.

Both volumes are usable volumes. T/S values for volumes have been selected to bound both Units. Ihe wards <<borated water" are added for consistency with Unit 2. See Attachment 13. 9'ilution transient is discussed in Attachment 14.

3.1.2.7.b.2 8 031 EST mininaun boron concentration 'Ihe minhaum RHST boron concentration limit is changed. has been increased to provide additional margin for the IDCA long-term cooling criterion. See Attachment 13.

AEP:NRC: ATIACHMEÃZ 3

SUMMARY

DESCRIPTIONS FOR D. C. COOK UNIT 1 PROPOSED TECHNICAL SPECIFICATIONS PAGE 5

+

SECTION 3.1.2.7.b.3 11 DESCRIPTION 03la Ihe is

~dred RNST pmperature innwased to 80 F.

~

REQBKS meum RNST temperature is tively raised to the temperature required for operability as a safeguaxds in modes 1,2, 3 & 4. 'Ihe value of 80 fram the Unit 2 IDCA. analysis is con-servatively chosen.

4.1.2.7.b ll 03lb 'Ihe RNST temperature will be monitored regardless of outside is a conservative. increase in surveillance nxpdrermnts.

air taaperatuxe.

3/4 1-16 3.1.2.8.a.l 8 032 Changed EAST minimum volume. Eoration source volume has been adj-usted to address the shutdown margin yegllire5 for a 111Ut1on trans1ent when cperating on RHR at the beginning of cycle.

T/S

~

values volume is a usable volume.

for volumes have been sel~ to bound both Units. Cate words "borated water" are added for consistency with Unit 2. See Attachment

13. Ihe dilution transient is discussed in Attachment 14.

3.1.2.8.b.2 8 033 RNST minimum boron concentration '%he minimum RNST boron concentration limit is changed. has been increased to provide additional margin for the DXA lcmg-term coolirxy criterion. See Attachnent 13.

3.1.2.8.b.2 8 034 RNST boron limit is concentration upper Wtll I Ith added. changeover to hot-leg recirculation safe-guards analysis require an upper limit on the RNST concentration. See Attach-

, ment 13.

AEP:NRC: ATZACHMENT 3

SUMMARY

DESCRIPTIONS FOR D. C. COOK UNIT 1 PROPOSED TECHNICAL SPECIFICATIONS PAGE 6 PAGE SECTION +

DESCRIPTION RP6&KS 3.1.2.8.b.3 11 034a %he recpired RHp temperature is 'Ihe miniznzn EST taagerature is con-increased to 80 F. servatively increased to the value for the Unit 2 IDCA analysis. Ghe Unit 1 analysis was perfarmed with an RNST teqxzature of 70 F.

3/4 1-17 4.1.2.8.b ll 034b 1he RÃST teuaperature will be monitored regardless of outside is a conservative increase in surveillance requinments.

air temperature.

3/4 1-18; 3/4 1-19'/4 3.1.3.1 1 035 %he words Mu.ch are in the core" are ins~

namved.

Editorial change. Ihese words refer to part length rods inserted in the core.

1-19a See p 3/4 1-14 Rev. 4, STS. Shee are no part length rods in Cook Unit 1.

3.1.3.1 1 036 Ihe word "bank" is replaced by Editorial charge to clarify the Specifi-ACTIQN b the words "gmup step counter". cation. Unit 1 is ecpipped with group step counters not bank demand counters.

Makes the Specifications of both units more similar.

ACTION c 1 037 Ihe words "due to causes other Editorial change to clarify meaning than addressed by ACTION a, of Sgecif ication; makes Sgecificaticins above," are added. of both units more similar.

ACTION c. 1 038 '"Ihe rod" is changed to "%he Editorial change for clarity.

ACTION c.2 affects nxV'.

ACTIQN c.2.a 3 039 %his ACTION placed.

stabmmt is re- ~ analyses which would recg.dre re-evaluation if Unit 1 were to be agitated with an inoperable control rod are more numerous than those recguring re-evalua-tion in the current specifications.

~

more change makes the Unit sgecifications 2

1 like the Unit specifications.

Also see STS Rev. 4, pp 3/4 1-14 and 3/4 1

AEP:NRC: ATXACHMENT 3

SUMMARY

DESCRIPXXONS POR D. C. COOK UNIT 1 PROEQSED TJKSNICAL SPECIFICATIONS PAGE 7 SECTION +

DESCRIPTION REMARKS ACXXQN c.2.c 3 040 Ihis ACZIGN statenent is added. Additional pcver distribution aanitoring would be required if Unit 1 were to be operated with an inoperable control rod.

Ihe change makes the Unit 1 specifica-tions more like the Unit 2 specifica-tions. Also see 8IS Rev. 4, p 3/4 1-15.

ACXXON c.2.d 1 041 Current ACXXON statements c.2.c Editorial change made to reflect addition ACXXQN c.2.e and c.2.d are renumbered. of new ACXXQN c.2.c.

1 042 Words added to emphasize that Editorial change. Mesc clarifications when ACXXON c.2 is chosen that also makes the Specifications of both items a, b and c must be per units more similar.

formed plus the choice of either d or e.

ACXXQN c.2.d 1 043 Mathematical symbols are Editorial ch-mge for clarity.

written out in words.

ACXXQN c.2.e 2 044 Refexerme to Figure 3.1-2 Ihree loop operation in Modes 1 and 2 is removede will be prohibited.

Table 3. 1-1 1 045 Table is ref~ to in Item added.

039 Editorial change. See Item 039.

3/4 1-21 3.1.3.3 1 046 Mathematical symbols are written Editorial change.

out 111 words.

1 047 n(228 Editorial change; clarifies mew~ of fully withdrawn.

1 048 APPXZCABILITY changed to MODES Editorial change; 'Ihe current Tech-1 and 2. nical Specification incorxectly in-dicate the applicable MODE to be 3.

9he specification is applicable to plant operation in MODES 1 or 2.

Xhe surveillance test is performed in MODE 3. Hakes the Unit 1 Spe-cifications more like the Unit 2 Specifications.

~ ~

AEP:NRC: ATZACHMENT 3 SUMKQK DESCRIPZIONS FQR D. C. COOK UNIT 1 PBDKSED TlKHNICALSHXXFICATIONS PAGE 8 SECTION +

DES CRIPZION REMARKS 2 049 ACZION statement b removed. 'Ihree loop operation in Nodes 1 and 2 will be prohibited.

4.1.3.3 3 050 words "prior to entering MDDE 2<< Requiring the completion of this test replace "prior to reactor cri- prior to entering MDDE 2 is conserva-ticality<<. tive to requiring the test prior to criticality. MDDE 2 is entmmi with the reactor subcritical by 14. However, making the requirement mode dependent eases administrative control.

3/4 1-22 3.1.3.4 1 051 <<(228 steps)" is added. Editorial change, clarifies meaning of fully withdrawn.

1 052 Mathematical symbols are written Editorial change for clarity.

out in woxdso 3/4 1-23 3.1.3.5 2

053

~ed'Ihree Reference to Figure 3.1-2 is loop operation will be prohibited.

in Nodes 1 and 2 ACZION b 1 054 <<fjgures<<becomes <<figure<< Editorial change.

1 055 Mathematical symbols are written Editorial change for clarity.

Gut 111 words+

3/4 1-24'igure 3.1-1 1

056 Rod Group Insertion Limits figure Editorial change. See mam dated 3/4 1-25 Figure 3.1-2 3/4 1-26 for 4 Iaop Operation is redrawn February 26 I 1986 I frcKR F J Silva with labeled endpojnts. to J. C. Miller of Westinghouse Electric Corporation found in Attachment 7.

2

057 Rod Group Insertion Limit figure Ihree loop operation in Nodes 1 and 2 for 3 Ioop Operation is removed. will be prohibited.

1 058 Rod Group Insertion Limit figure Editorial change.

for 4 Ioop Operation is renamed Figure 3. 1-1.

1 059 Pages 3/4 1-25 and 3/4 1-26 are Editorial change; blank pages are un-removed e necessary at the end of a section.

AEP:NRC: ATI'ACHMENT 3 SOME'ESCRIETIONS FOR D. C. COOK UNIT 1 PBOKGED TlXSNICAL SPECIFICATIONS PAGE 9 PAGE SECZION +

DESCRIPTION REMARKS 3/4 2-1 3.2.1 p6p ~~3,4,2<> becomes ii3/4,2" in title Editorial chalge.

1 061 APL footnote is rented. Editorial charge. API is nm found in definitions.

1 062 Mathematical symbols are written Editorial charge for clarity.

Gut ln wordse 3/4 2-2 3.2.l.a.2.c

Distr~on Monitoring for AHNS calibration is

'%he Social Rarm removed. System (AKHS) is will operate Iavel (APL) .

~

not used. Ihe plant below the Allowable 3.2.l.d 1 064 Action d is removed. Editorial change. Ihis action referred to the IER section of Technical Speci-fications. Ihe IZR rules are now in-cluded in CFR.

3/4 2-3 4.2.1.3 10

065 F (Z) is changed to APL. Ihe combined F (Z) target flux 4.2.1.4 R8ferenced specification surveillance c8angtai rc ccmhinei ApL number has changed. target flux surveillance. See Technical Specification 3.2.6.

3/4 2-4 Figure 3.2-1 1 066 Figure is redrawn. Editorial change for clarity.

3/4 2-5 3.2.2 1

067 Description of F (Z) penalties Editorial change for clarity.

moved from surve91lance to IlG.

3/4 2-5 3.2.2 10 068 ~ F limit for Emu'uel is 'Ihis charge is based on the I~

chalg9d to fixed value of 2.04.

Rev. 2. ~

analysis presented in XN-NF-85-115(P),

report was transmitted to the NRC with a letter dated January 15, 1987 from Exxon Huclear Company, Inc. Ihe Exxon letter was identified as GNW:001:87. Ihis report was .placed on our docket by a letter dated January 29, 1987 from M. P. Alexich to the NRC Doculm'.nt Contxol Desk. (Identifier AEP:NRC: 0940E) .

Ihe new analysis does not reaQt in a

AEP: NRC: ATI'ACHMENT 3

SUMMARY

DESCRIPZIONS FOR D. C. COOK UN' PROPOSED TE'KSNICAL SPECIFICATIQNS PAGE 10 PAGE +

DES CRIPZION burnup dependence for Emzn fuel as discussed in Section 2.0 of XN-NF-85-115(P) .

'Ignis result is also discussed and supported in a letter fram H. G. Shaw of Advanced Nuclear Fuels to Mr. Rick Bennett dated March 5, 1987, identifier ENC/AEP 0556.

'Ihe letter fram Mr. Shaw is included, as Attachment 15. To facilitate this review we are retransmitting AEP:NRC:0940E and a proprietary version only of XN~ 115(P) with Attadment 15. In addition, we are retransmitting our letter AEP:NRC:

1018 and its Attachment 1 and its propri-etary Attachment 4 with Attachment 15 of this letter. %hase documents denxestrate our recognition of burlap limits based on mechanical design and our ~tment not to exceed those limits without performing recpured analyses.

1

069 Definitions for P, F (Z), and Editorial change for clarity.

K(Z) reworded with na change of meaning.

3.2.2.a 10

070 Modified existing ACZION state- In the current Technical Specification ment a.l to remove the reguiLre- 3.2.2, ACZION a.l requires that the ment to 1carer the Overpower hT OPbT trip setpoirrt reduction be per-(OPLT) in hot stamky. formed when the reactor is in hot standby. This has been deleted. %he change in the ACTION statetumt for specification 3.2.2 is consi:~ with Standardized Technical Specifications, Revision 5. Our evaluation irdicated

-that the reduction of the Ovexpmmr AT setpoint can be done while the reactor is in Mode 1.

1 071 nF " ~ changed to iiF (Z) Editorial change for clarity.

AEP:NRC: ATIACHMENT 3

SUMMARY

DESCfUXTIONS FOR D. C. OOOK UNIT 1'ROPOSED TECHNICAL SPECIFICATIONS PAGE 11

+

SECTION 10

072 DESCRIPTION ACTION 3.2.2.a.2 is removed.

REMARKS Jhe AKMS is not operate belch APL.

used. ~ plant will 3/4 2-5g 3.2.2 1 073 ACTION statement b is moved Editorial change for clarity.

3/4 2-6 fmm page 3/4 2-6 to page 3/4 2-5.

3/4 2-6 4.2.2.2 10

074 Much of this surveillance Specification is simplified.

3/4 3/4 2-7; 2-8 th tt APL Specification 3.2.6. specification are new incorporated in 3/4 2W specification 3.2.6. No provisions of cun~nt Technical Specifications other than those pertaining to the follmr-ing were deleted or substantially mxiified:

(1) AHOIS See items 63, 72, and 134.

(2) Exxon F limit, based on the revised IDCA anRlysis See item 68.

of

~

(3) Removal implied burnup dependencies for F~.

justification for burnup limit of removing the 42.2 MHD/8'g for Westirxgxvse fuel is contained in the group 10 of the significant

~

hazards evaluation, Attachment 1 of suke~sion. See items 68 (Exxon fuel) and group 10 of Attachment 1 (4) Resaoval of the V(Z) figure - See item 75.

(5) %he modification to existing ACTION state-ment a.l of Technical Specification 3.2.2 See items 70 and 96.

(6) Items 65, 74, and 96 describe the sim-plification. Item 189 adds the movement to submit the Peaking Factor Limit Report.

AEP:NRC: 0 ATI'ACHMENT 3 SUMS'ESCfCETIONS FOR D. C. COOK UNIT 1 PROPOSED TEKSNICAL SPECIFICATIONS PAGE 12

+

DES CRIZTION RIXMKS Ghe methodology which supgorts the F surveil-lance is described in Part B of the topical report, WCAP-10217-A "F Surveillance Technical Specification". Atta8hment 19 includes:

(1) A reviev of greased simplificatians by aur

. %A!Nb pellet hump.

R /awk 3/4 2-8(a) Figure 3.2-3 1 075 Ihe V(Z) function provided by Editorial changes; 'Ihe V(Z) curve in Exxan Nuclear Co. is removed Figure 3.2-3 is associated with the fram Technical Sgecifications. previous fuel- vendor's methadolagy.

This page is to be remaved fram 'Ihe ecpivalent penalty is supplied T/S. tr in the Peaking Factor Limit Report. Re-maval of this figure was previously proposed for Cycle 8 agemtion. Remaval of this page was inadvertmtly omitted fram Amendment 74 to DPR-58.

3/4 2-10 Figure 3.2-2 1

076 'Ihe figure is redrawn. Editorial change for clarity.

1 077 Ihe page number is changed to Editorial change.

3/4 2-7.

3/4 2-11 Figure 3.2-3 1 078 Ihe figure is redrawn. Editorial change for clarity.

1 079 Ghe page weber is changed to Editorial change.

3/4 2-8.

3/4 2-12 3.2.3 1 080 Mathematical symbols are written Editorial change for clarity.

aut in words.

1 081 "paver" xa changed to "PONIES~ Editorial is change.

a defined term.

Th~ paver

AEP:NRC: ATTACHMENT 3 SUMKQK DESCfGPXXONS FOR D- C. COOK UNIT 1 PROPOSED TEKSNICAL SPECIFICATIONS PAGE 13 SECTION +

DES CRIPZION REMARKS 1 082 %he page number is changed to Editorial change.

3/4 2-9.

3/4 2-13 4.2.3 1 083 4.2.3.1 is changed to 4.2.3. Editorial change.

1 084 The page number is changed to Editorial change.

3/4 2-10.

3/4 2-14; 3/4.2.4 1 085 lhe page numbers are changed to Editorial change.

3/4 2-15 3/4 2-11 and 12, respectively.

1 086 "IZMZIS" is added to title. Editorial change.

1 087 Mathematical symbols are written Editorial change.

alit in wards+

ACTION b.2 1 087a "trip" is changed to "Trip". Editorial change.

3/4 2-16 3/4.2.5 1 088 The page number is changed to Editorial change.

3/4 2-13.

4.2.5.2 * !

4.2.5.3 is expanded and clarified. add CHANNEL GKZBRATION and flew 18-11enth calibration and fleer measure-ment are required to ensure the accuracy of the 12-hour surveillance of RCS fleer arxi the accuracy of the lear fleer trips.

Monthly flmr surveillance is renaved as redundant to shiftly surveillance.

are consistent with Unit 2 Technical Specifications. See Attadment 6, Iten Nut~ 10.

4.2.5.4 7

090 Exemption fram Specification Primary flem surveillances must be 4.0.4 is added for primary fleer made in the applicable mcde.

surveillances.

AEP:NRC: ATZLCHMENT 3 SUMMA'ESCRIPTIONS FOR D. C. COOK UNIT 1 PROPOSED TECHNICAL SPECIFICATIONS PAGE 14 SECTION +

DES CRIETION REKQES 3/4 2-17 Table 3.2-1 2

091 Ihe parameters for three loop Ihree-loop operation in Nodes 1 and 2 operation are removed. will be pmhibited.

1 092 Ihe parameters for Design 'Ihezmal Editorial chmpe for simplification; Pc%Ax're ZBEKwedo these values cannot be used prior to cctopletion of power re-rating analy-sis o 1 093 Units used for pressure Editorial change for simplificaticn.

chmped fram psia to psig.

1 093a Exponent chmped fram 10 to Editorial ghange. 1.386 x 10 charged to 10 . 138.6 x 10 for consistency with the Unit 2 T/Ss.

8 094 Footnotes are added for KS T 'this change reflects an analysis previously and RCS Total Flow Rate. submitted. See page 3 of Attachment 1 to the letter dated August 13, 1985 fram M. P. Alexich to H. R. Dentan (Identifier AEP:NRC:0942D).

To facilitate this review, we are re-transrnitting the proprietary attachment anly as Attachment 4 to this letter.

See also Attadment 6, Item Number 9, for supplementary information ~lied by our calculation for Unit 2 are exhibited on page (vii) of Attachment 18.

1 095 lhis page number is clued to Editorial change.

3/4 2-14.

096 Ibis entire Technical 3/4 2-18 3/4 2-19t 3/4 2-20'/4 2-21 3/4 2-22 3.2.6 10 an Allowable ~

Specification is changed to Ievel (APL)

Technical Specification.

%he ASS, an option in current Tech-nical Specificatians, is not used. Ihe plant will operate below APX. this specification is added to satisfy the requirements of the westinghouse F 3/4 2-23; Surveillance Technical Specificatian 3/4 2-24 M'ethodology. No provisions of current Technical Specifications other than those

AEP:NRC: ATTACHMENTS 3 SUMMATE DESCRIPXXQNS FOR D. C. COOK UNZIP 1 PRO~ TEQKICAL SPECIFICATIONS PAGE 15

+

DES CRIETION REMARKS pertaining to the follmriLng were deleted (1) APCHS .See items 63, 72, and 134.

(2) Exxon F limit, based on the revised Exxon IHCA analysis See item 68.

(3) Removal of bur@up dependencies for F~.

'Ihe justification for remcving the implied burnup limit of 42.2 MHD/8'g for Westinghouse fuel is contained in grt:mp 10 of the significant hazards evalua-tion, Attachment 1 of this Submission.

See items 68 (Exxon fuel) and group 10 (4) Removal of the V(Z) figure - See item 75.

(5) Ghe modification to existirg ACTION state-ment a.l of Technical Specification 3.2.2 See items 70 and 96.

(6) Items 65, 74, and 96 describe the sim-plification. Item 189 adds the requhmoent to submit the Peaking Factor Limit Report.

uhe nethcdology which sumaorts the F surveil-lance is described topical report, in Part B WCAF-10217-A of the 'use "F Burueillance Technical Specification". Attachment 19 includes:

(1) A revier of pmgceed simplifications hy our fuel vendor, Westinghouse.

AEP:NRC: ATI'ACSMENT 3 SUM%6K DESCRIPTIONS FOR D. C. COOK UNIT 1 PBDPOSED TEKSNICAL SPECIFICATIONS PAGE 16

+ + DES CRIPPION REKQKS

%P>>W Westinghouse fuel to at least 60 AND~ peak pellet burnup.

specification 3.2.2 have been yf incorporated in this specification.

In the cu~~nt Technical Specification 3.2.2, ACTION a.l recpdzes that the OPDT trip setpoint reduction be per-formed when the reactor is in hot standby. Qua has been deleted. Ihe change in the action statement for specification 3.2.2 is consistent with Standardized Technical Specifications, Revision 5. Our evaluation indicated that the reduction of the Overpower hT setpoint can be done while the reactor is in Miode l.

1 097 lhe new APL Technical Specifi- Editorial change.

cation is on pages 3/4 2-15 and.

3/4 2-16.

3/4 2-17 through 1 098 Pages 3/4 2-17 through 3/4 2-24 Editorial change.

3/4 2-24 are deleted.

3/4 3-2 Table 3.3-1 1 099 ~

blank.

page is intentionally left Editorial condensed.

change; Table 3.3-1 is Resulting table is more similar to the Unit 2 Technical Specifications.

3/4 3-3 Table 3.3-1 3

100 Power Range, Neutron Flux 'Ihe Plant Transient Analysis retires Item 2 Functional Unit has an added the Arum Rarge, Neutmn Flux Funct-applicable mode:*. ional Unit to be opexable with the reactor trip breaker in the closed position and the control rod drive

AEP: NRC: 0 ATZACHMEÃZ 3 SUMhSRY DESCRIPTIONS FOR D. C. COOK UNIT 1 PROPOSED TEKSNICAL SPECIFICATIONS PAGE 17 SECTION +

DESCRIPTION M%ARES mechanism capable of rod withdrawal.

this is consistent with section 14.3.1 of Agperdix 14.C of the Unit 1 FSAR, as well as Table 4.3-1 of the Unit 1 Technical Specifications.

Item 3 1 101 Camma is added. Editorial change.

Item 5 1 102 "Intermediate Range, Neutron Editorial change.

Flux" is tyged onto two lines.

Item 7 2

103 References to three loop %tume loop ogeration in Nxhs 1 and 2 operation axe xemavedo will be prohibited.

Item 8 2

104 References to three loop three loop operation in Nx~ 1 and 2 operation are removed. will be prohibited.

3/4 3-4 Item 13 1 105 ninn added Editorial change for clarity.

Item 14 1 106 "loops" xs changed to "loop" Editorial change; somatical error correction.

3/4 3-5 Item 16 1 107 slash replaced by hyphen. Editorial chanc~e; typographical error correction.

Item 20B 7

108 Reactor Coolant Pump Breaker Gee Reactor Coolant Pump Breaker Position Trip Abave P-7 has Position Trip pmvides protection an added exemption fram 3.0.4 against DNB at reactor coolant flow applicability. rates abave the P-7 interlock. 'Ibis interlock is enabled between 0 and 114 rated thermal pawer. Technical Specif-ication 3.4.1.1 recg,~s all reactor coolant loops be in egexation for MODES 1 and 2. With all coolant loops in agexatian, there is more than enaegh flmr for DNB protection up to the P-7 interlock (114 RZP) and the ESF actua-tion for DHB pmt~ion is nat needed in K)DE 1 until after the P-7 is enabled. At that point, the Reactor

AEP:NRC: ATI'ACHMENT 3 SUMMMY DESCRIPZIQNS FOR D. C. COOK UNIT 1 PRQKSED TEKSNICAL SPECEFICATIQNS PAGE 18 SECTION +

DESCfUPI'ION KMQKS Coolant Pump Breaker Position Trip channel must be in operation. 'Ihe proposed change to exempt Section 3.0.4 will allow entry into Mode 1 withcxrt these channels raptured operable but will not aller operation above P-7 interlocks without meeting the appmp-riate action stateaents. %his proposed charge was also recognized in later revisions to the Standard Technical Specifications.

Item 22 1 109 Clarifications identify which made to properly ACTION state-Editorial change; this change cps a format error made in the issuance of ments apply to applicable mode. Aaerdment No. 99.

3/4 3-6 Table 3.3-1 2

110 Footnote ** is removed. 'Ihree loop operation in Modes 1 and 2 Notation will be prohibited.

ACZIQN 2.b 1 illWords "of the other channels" Editorial charge for clarification.

are added. Makes Specifications for both units more similar.

ACTION 1 112 Mathematical symbols are written Editorial charge for clarity.

2oc out in words+

3/4 3-8 Table 3.3-1 1

113 "OHRR Editorial change; typographical error ACZIQN 14 "OPERABIZ" correction.

ACZION 9 2

114 Action 9 is removed. Wree loop operation in H'odes 1 and 2 will be prohibited.

1 115 Mathematical symbols are written Editorial change for clarity.

out in words+

3/4 3-9 Table 3.3-1 1 116 Mathematical symbols are written Editorial change for clarity.-

out l11 woD3S, ll 117 Ihe value of P-8 is changed to To achieve greater consistency with 31% RIP. Unit 2 Technical Specifications. 314 is conservative relative to current 514.

AEP:NRC: ATZACHMENT 3

SUMMARY

DESCRIETIONS FOR D. C. COOK UNIT 1 PROPOSED TECHNICAL SHZIFICATIONS PAGE 19

+

~

PAGE SECTION DES CRIETION REKQGE 3/4 3-12 Table 4.3-1 3

118 Power Range, Neutron Flux %he Plant Transient Analysis requires Item 2 Functional Unit has an the Range, Neutron Flux Funct-additional Channe1 Functional ional Unit to be operable with the Test (S/U(l) ) . reactor trip system breakers in the closed position and the control rod drive mechanism capable of nd with-drawal. See section 14.3.1 of Agperdix 14.C of the Unit 1 FSAR.

3/4 3-12; Item 2, 5, 6, 7

119 Power, Intermediate, and Source Exemptions are provided for surveill-3/4 3 13 7g 8~ 12 & 13 Range Neutron Flux, loss of Flow ances which must be performed in the Single Zaop and Two Zoog Func- applicable mode. Nate that the

does tional Units have added not agply to loss of flow in two units exemptions fram Specification which was inadvertent1y amitted fram aur 4.0.4. Overpower hT and Over- Unit 2 submittal.

temperature hT Functional Units have added exemptions fram Specification 4.0.4 for fl(kl) and f (AI) penalties.

3/4 3-14 Table 4.3-1 1 120 Mathematical symbols are written Editorial change for clarity.

Notation out in woD3s ~

7 '* 121 Footnates (8) and (9) are added. See remarks for Item 119.

3/4 Table 3.3-3 2

122 References to three loop three loop operation in M'odes 1 and 2 Item l.e will be 3-16'/4 operation in Mades 1 and 2 prohibited.

3-17'/4 3-18; l.f axe xGNcvede 3/4 4.d

123 Reference to Ngl footnote for 3-20'/4 3-21 5 'Ihe Differential Pressure Between Differential Pressure Between Steam Lines~gh actuatian differs Steam Lines-High Functional fram os~ ESF AcI~tion signals in Unit changed to NNNN footnote. that a signal fram ane loop is comp-ared to signals in the ather loops.

Placing all channels associated with the idle loop in trip would result in an ESF actuation. this actuation would prec3.ude 3 loop operation.

lherefore, the agpxopriate channels to trip are the bistables which irdicate

AEP:NRC: ATI'ACHMENT 3

SUMMARY

DESCRIPZIONS FOR D. C. COOK UNIT 1 PROPOSED TECHNICAL SPECIFICATIONS PAGE 20 SECTION + + DES CRIPZION RI~RKS law active steam pressure relative to the idle loop. this action reduces the ESF actuation logic for the active loop differential pressures fram 2/3 to 1/2. An ESF actuation does not result because the three bistables, which indicate low idle loop steam pressure relative to the active loops, and which are in a 2/3 logic, are not tripped. See simplified logic dia-gram in Attachment 16.

Items 1.f & 12 124 References to Footnote ** are 'Ibis change reflects an analysis pre-4.d XGEcved>> viously submitted. See Attadment 4 of the letter dated August 13, 1985 fram M. P. Alexich to H. R. Denton (Identifier: AEP:NRC:0942D) . To facilitate this review we are re-transmittiinp the proprietary attadment anly as Attachment 8 to this lett~.

125 (No change for this identifier) 3/4 3-22 Table 3.3-3 5

126 Footnote /ANN is added. See remarks for Item 123.

12 127 Footnote ** is remcved. See remarks for Item 124.

3/4 3-23 Table 3.3-3 9

128 Rewarded Condition and Set- change clarifies the definitions point, Function description of the interlock and malines the defin-for P-12 interlock. ition less ambiguous. Patterned after STS, Rev.4.

1 129 Mathematical symbols are written Editorial change for clarity.

cut in words>>

3/4 3-29 Table 3.3-5 12 130 Reactor trip is removed fram Ihis wording is consistent with SIS, Item 8a description Revision 4. %he analysis of Rmessive Heat, Removal due to Feedwater System Malfunctions event is the only analysis which uses the ESF Steam

AEP:NRC: ATIACHM1.'NT 3 SUM%6K DESCRIPTIONS POR D- C. COOK UNIT 1 PROPOSED TEKSNICAL SPECIE'ICATIONS PAGE 21

+

DESCRIPTION REMARES Generator Water level-High High featuze. %his analysis uses the reactor trip on turbine trip as an anticipatory trip to terminate the event. Since the trip is not re-quired the only re@ense time needed is the response time for turbine trip.. %his event is discussed in greater detail in item 2 of Attac?unent II to Attachment 6 of this submittal.

3/4 3-31 Table 4.3-2 1 =131 Period is changed to comma. Editorial change; typographical exzar Item 1c co~iona 3/4 3-33 Table 4.3-2 1

132 loss of Main Feedwater Pumps Editorial charge; Mode 3 applicability Item 6d for Zoss of Main Feedmter Pumps deleted. was deleted fram Table 3.3-3 in Unit 1 Lice'mendment N92.

3/4 3-33a Table 4.3-2 1 133 "Zoss of Voltage" is charged to Editorial change to clarify difference Item 8.b "Degraded Voltage" .. between Item 8a & 8b.

3/4 3-49; 3.3.3.6 10

134 this entire Technical Ihe AKRON is not used. 'Ihe plant will 3/4 3-50 4.3.3.6 Specification is zerxved. opexate below APL.

3/4 4-2 3.4.1.2 3

135 Criterion for the operability of reactor coolant loops are estab-

~ Plant Transient Analysis these changes based on recg.aires the uncontzalled lished based on the status of control zad bank wit1xhawal fram the reactor trip system breakers subcritical. %he proposed Specification and/or the control zod system. conservatively requires 3 pumps for consistency with Unit 2. An appropriate ACrraN statenM nt has been p~ased to letter fram E. P. Rahe, Jr. to D. Eisenhut dated July 9, 1984 (Identifier NS ZL-84-003) and letter fzam M. P. Alexich to H. R. Denton dated July 30, 1984 (Identifier AEP:NRC:0895) .

To facilitate this review, we are retransmitting these letter as Attachment 17.

AEP:NRC: 0 ATIA'CEMENT 3

SUMMARY

DESCRIPTIONS KR D- C. COOK UNIT 1 PROPOSED TECHNICAL SPECIFICATIONS PAGE 22 SECTION +

DESCRIPTION REPAIRS ll 135a Criterion for the agerability of reactor coolant loops based Table 3.3-3 mgu.ires at least three loops operating above P-12. gris on P-12 is added. ensures flow through RID by-pass loops. 'Ibis pmvision is added for consistency with Table 3.3-3.

An appropriate ACHQN statement'as been proposed to correspczxl 1 136 Existing text reorganized for Editorial change.

convenience. ACTION b becomes ACTION d.

3.4.1.2 4

137 ** is added to ACTIQN d and 'Ihe Technical Specification boron ACTION d foat2ute *; footnate ** is added concentration in the INST is suff-Foatnate

to bottom of page. icient to provide adequate shutdown margin fmm expected agexating conditions.

3/4 4-2a 4.4.1.2.1 1

138 Surveillances, footnotes and Editorial ch;eye; additional text 4.4.1.2.2 ACTION. d moved fram previaus requires moving this material.

page.

3/4 4-3; 3.4.1.3 3 .* 139 Criteria for the agerability of See remarks for Item 135.

3/4 4-3a reactor coolant loops are estab-lished based on the status of the reactor trip system breakers and/or the contxol rod system.

1 140 Rcisting text reorganized for Editorial cha~e.

convenience. ACTION b beccues ACZION c.

****is 3.4.1.3 ACZION c Footplate ***

4 141 f~~ added to A(TION c and

- - ~ footnote ****

added on page 3/4 4-3a.

'Ihe Technical Specification boron concentration in the HNST is suff-icient to provide adequate shutdown margin fram expected operating conditions.

1

142 Foatnates are moved fram Editorial change; expanded specific-page 3/4 4-3 to page 3/4 4-3a. ation requ~ the movement of this material.

0 F

AEP:NRC: ATIACHMENT 3 h SUMMtQK DESCRIPXXONS H)R D. C. COOK UNIT 1 PROPOSED TECHNICAL SHXXFICATIONS PAGE 23 SECTION 1 142a Changed 62.00% to 624. Removed Editorial change.

underlining.

3/4 4-3bg 3.4.1.4 2

143 Jhe entire Technical Three loop operation in Modes 1 and 2 3/4 4-3c; 4.4.1.4 Specification is rented. will be prohibited.

3/4 4-3d 1 144 Pages 3/4 3m and 3/4 3d are Editorial change.

to be removed.

3/4 4-4 3.4.2 4 145 Footnote ~* added. We Technical Specification boron concentration in the HHST is sufficient to provide adequate shutxhmn margin fram expected operating cancKtions.

ll 146 Footnote

is added. this change clarifies the conditions to which the pressurizer code safety valve lift settings corresgczxl. this footnote is in the Unit 2 Technical Specifications and in practice accurately describes what is done currerrtly in Unit

l. %his change does not impact the operations of Unit 1 and is primarily administrative in nature.

3

147 ACTION statement added. Changed to make the Specifications of both Units more similar. 'Ihe analyses of overpressurization for Unit 2 de-scribed in XN-NF-85-28 (P), Supplement 1-"D. C. Cook Unit 2, Cycle 6 Safety Analyst Report" zdentxXzed the need for the proposed additional ACTION to prevent overpressurization with no safety valve parable. Since Unit 1

.and Unit.2 primary systems are essmtially identical, the additional ACTION is proposed for Unit l.

3/4 4-5 3.4.3 4 4' ll 148 Footnote

is added. See remarks for Item 146.

AEP:NRC. ATZACHMENT 3

SUMMARY

DESCRIPZIONS FOR D. C. COOK UNIT 1 PROPOSED TEKSNICAL SPECIFICATIONS PAGE 24 SECTION +

DESGUPZION 3/4 4-35 3.4.11

149 to only allcar 6 ACZION charged one PORV or block valve ~ Changed to make the Specification of bath units @axe similar. 'Ihe proposed able. Making more than one PORV changes are inhmRd to ensure that the inoperable without shutting dawn EORVs are available to assist in RCS the reactor is nat allawed. depressurization follcaring a steam generator tube rupture without offsite Power. See Section 14 2 4g "Steam

~

Generator Tube Rupture", of the Unit 1 FSAR.

1

150 Reference to Section 6.9.1.9 reference is no longer is deleted. appropriate. Section 6.9.1.9 of the Technical Specifications delineated tV P 50.72 and 10 CFR 50.73.

3/4 4-36 4.4.11.1 1

151 Portions of expanded ACZION Editorial change; surveillance statement and surveillance I repdzments moved to this'age.

p 3/4 4-36.

4.4.11.2 1

152 Refexence to Section 6.9.1.9 See remarks for Item 150.

is deleted.

1 153 Foatnate

is changed to **. Editorial change.

4.4.11.3 1

154 Refemme to Surveillance Editorial change; the current refer-4.8.2.3.2.c is changed to ence is 3Jlcoxxecto 4.8.2.3.2.d.

3/4 5-1 3.5.1.b 1 155 Text revised. Editorial change to make the specifi-cations of bath units more similar.

3.5.l.c 8 156 Mininaan accumulator boron 9he ma'am accumulator boron concen-concentration is changed. tration limit has been increased to provide additional margin for the DXA longb~ cooling criterion.

See Attachment 13.

AEP:NRC: ATI'ACHMENZ 3

SUMMARY

DESCRIPZZONS FOR D. C. COOK UNIT 1 PBOKSED TEXSNICAL SPECIFICATIONS PAGE 25 SECTION +

DES CECETEON E~KS 3.5.1.c 8 157 Accumulator boron concentration 'Ihe containment sump pH analysis and the upper limit is added. changeover to hat-leg recirculation safe-guards analysis recuire an ~mr limit on the accumulator concexdxatian. See Attachment 13.

3/4 5-11 3.5.5.b 1 158 Text revised. Editorial change to make the specifi-4.5.5.a.l cations of bath units more similar.

159 (No change for this identifier) .

3.5.5.b 8 160 Minimum HNST boron concentration 'Ihe minyan HNST boron cancentration is changed. limit has been increased to provide additianal margin for the IDCA long-term cooling criterion. See Attach-ment 13.

8 161 RNST boron concentration upper limit is added.

changeover to hat-leg recirculation safe-guards analysis rely.ore an upper limit on the HNST concentration. See Attach-ment 13.

3.5.5.c ll 161a The is required RNST ~+suture 'Ibe minim HNST temperature is conser-increased to 80 F.

Unit 2 IOCA analysis.

analysis was

~

vatively increased to the value for the Unit 1 perfarmed. with an HNST temperature at 70 F.

4.5.5.b ll 161b ~ RNST taagerature will be monitored regardless of outside

%his is a canservative iacnese in air temperature.

3/4 7-1 3.7.1.1 2

162 ACTION b is modified to remove three loop apemtion in hRxhm 1 and 2 three loop operation in Modes 1 will be prohibited.

and 2.

3/4 7-3 Table 3.7-2 2

163 Table is remcved. 'Ihzee loop operation in Modes 1 and 2 will be prohibited.

AEP:NRC: AITACHK"NT 3 SUMMA DESCRIPTIONS FOR D. C. COOK UNIT 1'ROPOSED TEKSNICAL SHKIFICATIONS PAGE 26 SECTION +

DESCRIPTION 3/4 7-4 Table 4.7-1 ll 163a Footnote

is added.

REMARKS

'Ihis change clarifies the conditians to which the pressurizer code safety valve liftsettings correspand. this footnote is in the Unit 2 Technical Specifications and in practice accurately describes what is dane currerrtly in Unit

l. %his change does not impact the operations of Unit 1 and is primarily administrative in nature.

3/4 7-5 4.7.1.2 8

164 Discharge pressures for aux- 'Ihe limiting accident for auxiliary iliary feechater pump flow feedwater pump performance is the feed-testiq changed. water line break. In Aaencb.nant 82 to DPR 74 (Unit 2), the auxiliary feedwater pump discharge pressures were lowered to the values being proposed for Unit l.

%his reduction was based on the feed-water line break analysis performed by Exxon Nuclear Co., which is found in Section 15.2.8 of XN-NF-8564 (P),

Rev. 1, "Plant Transient Analysis for D. C. Cook Unit 2 with 10% Steam Generator Tube Plugging". this new analysis allawed credit for ogemtor action after 10 minutes to isolate the faulted steam generator ancl ensure adequate auxiliary feedwater was delivered to the intact steam generators.

Res differed fram the-previous Unit 2 analysis, which asmnaed auxiliary feed-water was delivered within ane minute following the initiation of the break.

'Ihe new Exxon analysis rem&ted in reduced auxiliary feechmter discharge pressure regLllxGHRIlts~ which werB reflected in the Amendment 82 T/Ss.

AEP:NRC: ATI'ACHMEÃZ 3 SUMEQK DESCtKETIONS K)R D. C. CGOK UNIT 1 PROPOSED TECHNICAL SPEKXFICATIONS REMARKS PAGE 27

+

DESCZUPFION For Unit, 1, Feedwater Line Break is not part of the license basis, as natal in Chapter 14.2.8.1 of the Unit 1 UFSAR.

However, an evaluation of this accident was performed and included in Chapter 14.2.8.1 of the UESAR. 'Ihis analysis, like the Exxon analysis, assumed 10 minutes for operator action and an identical value for the amount of auxiliary feedwater delivered to the intact steam generators (600 gpm) .

Ihus, it supgorts the same value for auxiliary feedmter pump discharge pressure as that currently included in the Unit 2 T/Ss, and the change is recuested to maintain consi~mcy bellmen the Units.

1 165 Mathematical symbols are written Editorial change for clarity.

out in woxl9so 3/4 7-10 3.7.1.5 3

166 ACTION statements are revised. 'Ihe provision of the ACTION statenent for MODE 1 perrnittiLng operation in MODE 1 with a steam generator stop valve closed is deleted. Failure to restore the stop valve to agee~le status in MODE 1 results in MODE 2 instead of MODE 4 ogeration. Ihe reference in the MODE 2, 3 ACTION statement to continued operation in MODE 1 is deleted. Ihe SIS terminology is changed to be consist~ with Oook Plant terminology.

AEP:NRC: ATI'ACHMEÃZ 3 SUMQRY DESCRIPI'IONS FOR D. C. COOK UNIT 1 PROPOSED TECHNICAL SPECIFICATIONS PAGE 28 SECTION +

DESCRIPTION RE KGB The proposed Technical Specification y

consisterxy with SIS, Rev.4. The emangtion fram 3.0.4 permits entxy into MODES 2 and 3 with an inoperable stop valve because such operation is permitted in those modes.

4.7.1.5.1 1

167 Specification 4.7.1.5 is re- Editorial change.

numbered 4.7.1.5.1.

4.7.1.5.2 7

168 Bceqotion fram Specification Evans are pravided for surveil-4.0.4 is added for entry lances which must be performed in the into Mode 3. applicable mode.

7

169 Exertion fram Specification 'Ibis specification ensures that no 4.0.4 is provided for entry more than one steam generator will into Mode 2 with stop valves blawdawn in the event of stmm line closed for BiYSICS TESXS. rupture. If the valves are closed during PHYSICS TERR only the affect~

steam generator can blawdawn. this provision pravides added agerational flexibilityat BOC.

3/4 8-5 3.8.1.2 4 170 Foatnate added. 'Xhe Technical Specification boron concentration in the RAT is sufficient to provide adequate shutdown margin fram expected operatic conditions.

1 171 Footnote

is changed to **. Editorial change.

3/4 9-1 3.9.1 1 172 Mathematical symbols are written Editorial change for clarity.

aut in words.

4 173 Foatnate added. The Technical Specification boron con-centration in the HNST is sufficient to provide adequate shutdown margin fram expec~ operating conditions.

AEP: NRC: AITAQiMENT 3

SUMMARY

DESCRIPTIONS FOR D. C. COOK UNIT 1 PROPOSED TIXSNICAL SPECZFICATIQNS PAGE 29

+

DES CIUPZION 3.9.l.b ll 173a %he ra~eed boron cancentration for refueling is increased to RIDARKS She recpxired cxmmrtration is canser-ACZIQH vative1y inn~ed to agree with the 2400 ppm. RNST concentration. %he result is a which the core is shutdown during refue1ing.

3/4 9-2 3.9.2 4 174 Foatnote added. The Technical Specification boron can-centmtion in the RNST is sufficient to provide adequate shutdown margin fram ex'~ operating conditions.

3/4 9-9 3.9.8.1 1 175 Mathematical symbols are written Editorial change for clarity.

4.9.8.1 alIt in words+

An analysis was performed to reduce to 2000 gpm. the required reactor coolant flow rate to 2000 gpm. See Attachttent 5 for discussion of heat remval, mixing, and stratisfications can-siderations. See Attadunent 14 for dilution transient considerations.

4

177 Footnote added. '%he Tectnical Specification boron con-centration in the RNST is sufficient to provide adequate shutdawn margin fram expected operating conditions.

3/4 10-2 4.10.2.2 1

178 Referenced specifications are Editorial change; .reflects simplifica-renumbered.- tion of F and APL specifications, 3.2.2 and 3.2.6 respective1y. See page 3/4 2-6.

1 179 Mathematical synkmls are written Editorial change for clarity.

aut 1ll words+

1 180 Reference to the Augmented Editorial change; the Aucpnmted Startup Startup Test Program is rewed. Test Program has been campleted. See Attachment 6, Item Nmiber 11.

. AEP:NRC. ATE'ACHMEPZ 3 SUMMA'ESCRIPTIONS FOR D. C. COOK UNIT 1 PROPOSED TECHNICAL SPECIFICATIONS PAGE 30 3/4 10-3 3.10.3.b 12 181 Specification is rewarded. Clarifies intention of specification.

See Attachment 6, Item Number 14.

1 182 Editorial change for consistency with written as "Reactor Trip Set- specification 3.10.5.b.

points" .

3.10.3.b 1 183 Mathemtical symbols are written Editorial change for clarity.

ACZIQN GLIt 1I1 words.

4.10.3.2 3/4 10-5 3.10.4.b 12 184 Specification is rewozded. See Remarks for Item 181.

1 185 "reactor trip setpoints" is re- Editorial change for cansistt~ with written as "Reactor Trip Set- specification 3. 10.5.b.

points.

1 185a tiBKRMAl Editorial change; typographical error

"'IHBMAL". correction+

3.10.4.b 1 186 Mathematical symbols are written Editorial charge for clarity.

ACTION Gut in words~

4.10.4.1 3/4 10M 3.10.5.b 12 187 Specification is reworded. See Remarks for Item 181.

1 188 Mathematical symbols are written Editorial charge for clarity.

GUt in words 4.10.5.1 1 188a "the" is added. Editorial change.

6-19 6.9.1.11 10 189 Section added. 'Ihe Peaking Factor Limit Report will be submitted each cycle. %his achieves greater consistency with SIS, Rev. 4.

See Attachment 9 for reason for change fram 60 days to 15 days. Ihis item is specifically addressed in bath the cover letter of this submission and aur significant hazards evaluation in Attachment 3.

AEP:NRC: ATZLCHMEÃZ 3 SUM%6K IXSCRIETIONS POR D. C. COOK UNIT 1 PROPOSED TEKSNICAL SHKIFICATIQNS PAGE 31 PAGE SECTION +

DESCZCPZION REMAINS B 2-1 2.1.1 B

190 References to three loop oper 'Ihree loop operation in Modes 1 and 2 B 2-la (Bases) ation and Figure 2.1-2 are will be prohibited.

removed>>

B 191 Headings are clarified; Editorial change to clarify meaning of footnotes are added. text.

B 2-5 Overtemperature B

192 Paragraph referring to 'three loop operation in Modes 1 and 2 hT (Bases) three loop operation is removed. will be prohibited.

Overpower ~T (Bases)

B

193 Added reference penalty for OPAT.

to f(bI) Penalty is used for the cu~ analysis.

Included in the basis for completeness and consistency with Unit 2 Technical Specification Bases.

Pressurizer B

194 Added reference to the use of See section 14.C.3.6 of Unit 1 FSAR.

Pressure the pressurizer pressure high (Bases) trip in the loss of load event.

B 2-5 2.2. 1 B

195 Moved text fram page B 2-6 to Editorial change.

B 2W (Bases) B 2-5.

B 2W Zoss of Flow

196 Ihe value of the Editorial

(~) B is is changed reworked, to P-8 setpoint 314. 'Ihis sentence change.

B

197 References to three loop 'Ihree loop operation in Modes 1 and 2 operation are removed>> will be prohibited. See Attachment 6, Item Number l.

B 198 Mathematical symbols are written Editorial change for clarity.

olxt in words>>

B 2-6' 2.2.1 B

199 Mcved text fram page B 2-7 to Editorial change.

2-7 (Bases) B 2-6, and from page 2-8 to B 2-8 B 2-7. Page B 2-8 may near be deleted.

AEP:NRC: ATZKCKKNT 3 SUMEQK DESCRIETIONS FOR D. C. COOK UNIT 1 PROPOSED TEKSNICAL SPECIFICATIONS PAGE 32 PAGE SECTION +

DESCfG'ETION RE%GUS B 3/4 1-1 3/4.1.1.1 B

200 Revision to Shutdown Maxgin Bases revised to address dilution 3/4.1.1.2 Basis. transient when opexating on RHR at (Bases) beginning of cycle. See Attachnent 14.

B 200a 350 F is changed to 200 F. Editorial change; typograptd.cal error coxrec$ ion. 'Ihe upper limit to Mode 5 is 200 F.

B 201 Mathematical symbols are written Editorial change for clarity.

Gut in words+

3/4.1.1. 3 B 202 Flow rate xecpumoent rcxTuaad An analysis was perfox1%Kl to reduce (Bum) to 2000 gpm. the recpixed reactor coolant flow rate to 2000 gpm. See Attachment 5 for discussion of heat removal, mixirg, and stratification con-siderations. See Attachment 14 for dilution transient considerations.

B 202a Circulation time is increased Circulation time increased due to to 45 minutes. decreased flow rate. See Item 202.

B 3/4 1-2 MiImmum B

203 Revised discussion of inter- Bases vere revised to more accurately Temp. for action between minimum reflect the operation of P-12 reset Criticality for criticality

(~) tenperatuxe point; paragraph rewoxded for polIIto consistency with Unit 2.

Technical Specifications.

B 3/4 1-2; 3/4.1.2 B 204>>above>> is changed to Editorial change; tygxxpaphical error B 3/4 1-3 (Bases) correction.

B

205 Revisions were made to desc- Boxation source volumes were adjusted ription of the RNST and BAST to address dilution transient when as boration scauzes. operating on RHR at beginning of cycle.

'le higher boron concentration of the RNST is also xeflect~ in the basis. Volumes used in the Technical Specifications which bound Units 1 & 2 axe discussed.

See Attachment 13.

AEP NRC: ATZACHMENT 3 SUMKQK DESCRIPZXONS FOR D. C. COOK UNIT 1 PROPOSED TECHNICAL SPECIFICATIONS PAGE 33 SECTION +

DES CRIETION REPAS B 206 Mathematical symbols are written Editorial change for clarity.

out in words+

B 207 pH value limits are added. Limits reflect the analysis in AttachIIent 13.

B 3/4 2-1 3/4.2 B 208 "F (Z,~)" is changed to "F (Z)". ~ charge ~ based on the Ehazn analys~

presented in XN-NF-85-115(P), Rev. 2. this report was transmitted to the NRC with a letter dated January 15, 1987 fram Exxon Nuclear Company, Inc. %he Exxon letter was iderItified as GNW: 001: 87. Ktlis report was placed on our docket by a letter dated January 29, 1987 from M. P. Alexich to the NRC Document Control Desk. (Identifier

~t AEP:NRC:0940E.) Xbe new analysis does not in a burnup dependence for Duon fuel as discussed in Section 2.0 of XN-NF-85-115(P). This result is also discussed in a letter from H. G. Shaw to R. Bennett dated January 26, 1987.

letter fmm Mr. Shaw is included as Attach-merrt 15. To facilitate this review we are retransmitting AEP:NRC:0940E and a proprietary version only of XN-NF-85-115(P) with Attach-ment 15.

B 209 Updated minimum IMR limit. Editorial change. Updated to value in FSAR Table 3.6.3-1.

B 3/4 2-2 3/4.2.1 B 210 ~ word "of" is added. Editorial correction change; grammatical e

error B 210a "signifigance" is change to Editorial change; typographical error significance correction 211 iiF (p g,)" m changed to F (2) See remarks for Item 208.

212 Description of burnup dependent See renmks for Item 208.

F envelope zs remcved.

AEP:NRC. ATZKCHMENT 3

SUMMARY

DESCRIPTIONS FOR D. C. COOK UNIT 1 PROPOSED TECHNICAL SPECIFICATIONS PAGE 34 SECTION +

DESCRIPTION REEQUIP B 212a Period replaced by ccma. Editorial change.

B 3/4 2-3 Figure B 3/4 2-1 B 212b Figure is ~wn. Editorial change for clarity.

B 3/4 2-4 3/4.2.2 B 213 Gee words "nuclear enthalpy hot Editorial chape; makes Technical 3/4.2.3 channel factor" changed to Specifications for both units mre "nuclear enthalpy rise hot similar ~

channel factor".

B

214 Befezencea to F ani F Editorial change.

mp ~~ t ~ne p BFL Technical Bpecif ication.

B 215 (No change for this identifier) .

B 3/4 2-5 3/4.2.3 B 216 "physics tests" is changed to Editorial charge; physics tests is "PHYSICS TESIS". a defined term.

B 217 ~on on burnup dependent for Exxon fuel removed.

F See remarks for Item 208.

B 3/4 2-6 3/4.2.5 B

218 Discussion of flow rate (Bases) surveillances are included. add CRENEL CALIBRATION and flow y

is remcved as redundant to shiftly surveillance. Resulting surveillance Technical Specificatians. See Attachment 6, Item Number 10.

3/4.2.6 (Bases)

B

219 Allowable ~

%his section is changed level (APL)

Technical Specification.

to an ~ AKHS xs not used.

operate below APL.

'Ihe plant wx11 B 3/4 3-3 3/4.3.3.6 (Bases)

B

220 %his section is remcved. ~ AHOIS is not used.

operate below APL.

Ihe plant will B 3/4 4-1 3/4.4.1 B

221 References to three loop 'Ihree loop operation in Hodes 1 and 2 (Bases) operation are ~ed. will be pated.

AEP:NRC. MTACHMENT 3

SUMMARY

DESCRIPZIONS FOR D. C. COOK UNIT 1 PROPOSED TEKSNICAL SPECIFICATIONS PAGE 35 SECTION +

DES CRIPI'ION REMARKS B 222 Updated rtunimum GNHR limit. Editorial charge. Updated to value in FSAR Table 3.6.3-1.

B 223 P-8 is charged to 314 of RTP. Conservative change to make Unit 1 Technical Specifications more like the Unit 2 Technical Specifications.

B

224 Additional operable loops are 'Ibe Plant Transient Analysis rapdres rapxumi with control rods these charges based on the umxntrolled capable of withdrawal. control rod bank withdrawal fram subcritical. 'Ihe proposed specification conservatively recpu~ 3 guam for con-sistency with Unit 2. See letter fmm E.P. Rahe, Jr. to D. Eisenhut dated July 9, 1984. (Identifier NS ZK-84-003) .

To facilitate this review, we are tran-smitting this letter as Attac?anent 17.

B 3/4 4-lg 3/4.4.2 B 225 Text is moved from page Editorial change.

B 3/4 4-2 3/4.4.3 B 3/4 4-1 to page B 3/4 4-2.

(Bases)

B 3/4 4-13 3/4.4.11 B

226 Periods are corrverted to slashes. Editorial ch;age.

3/4.4.12 (Bases)

B 3/4 5-3 3/4.5.5 B 227 pH value limits are changed. Limits reflect the analysis in Attach-(Bases) ment 13.

B 227a Discussion of the difference %he minimum RNST temperature is conser-vatively incr~xi to the value for the between the analysis value and Technical Specification value of the RHST tetaperature is Unit 2 IDCA analysis. ~ Unit 1 analysis was perfarmed with an RHST added + temperature of 70 F.

B 3/4 7-1 3/4.7.1.1 B

228 References to three loop Three loop operation in H'odes 1 and 2 (Bases) operation are rermved. will be prohibited.

B

229 Reference to Table 3.7-2 Editorial change; incorrect table is changed to Table 3.7-1. reference.

This basis is condensed to one page.

AEP:NRC: ATZACHMENT 3 SUM%6K DESCRIPTIONS PDR D. C. COOK UNIT 1 PROPOSED TiKHNICALSPECIFICATIONS PAGE 36 SECVION +

DESCRIPTION EKHARHS B 3/4 7-2 3/4.7.1 B

230 Variable definitions are Editorial chape; impmve readability.

(Bases) maved to previous page.

B 3/4 9-1 3/4.9.1 B 230a 'Ihe basis section fram SIS is 216 xegQ1vad cQIlcpJICERMon xs conserva-substituted for ~ing and is aucpnented with a basis tively increased to agree with the HNST concentration. %he result is a subst~-

discussion of the inn~ase in tial increase in the anmnt by which the core is shutdown during zefueliay.

to 2400 ppm.

3/4.9.5 B 230b "CORE ALT1BNATIONS" is chaDged Editorial charge; typographical exror (Bases) to nCGRE ALT1XAXIONSn. correction.

AEP:NRC: ATI'ACSMEÃZ 3

SUMMARY

DESCRIPZXONS FOR D. C. COOK UNIT 2 PROPOSED TIXSNICAL SPECIFICATIONS PAGE 37 SECTION +

DESCRIPTION REMtQKS 2-2 Figure 2. 1-1 1 231 Curve for 2250 psia is added. Editorial change. See letter dated July 31, 1986, ENC-AEP/0511, H.G.Shaw to D.H.Malin found in Attac?maes 10.

3/4 1-3 4.1.1.2 1 232 Change "greater than<< to Editorial change. Makes Unit 1

. "greater than or eglbQ to and Unit 2 more-consistent.

1 232a Mathematical symbols are Editorial change for clarity.

written out in words.

3.1.1.2.b 1 232b Period added. Editorial change.

3/4 1-4 3.1.1.3 An'analysis was performed to reduce 4.1.1.3 to 2000 ggm. the required reactor coolant flow rate to 2000 gpm. See Attachment 5 for discussion of heat rental, mixiap, and stratisf ication considerations. See Attachment ll considerations.

for dilution transient 1 234 Mathematical symbols are written Editorial change for clarity.

out in words+

3/4 1-5g 3.1.1.4 8 235 'Ihe upper limit on MK for To improve operational flexibility.

3/4 1-6g Figure 3.1-2 operation above 70% RTP is Justification provided in Attachment 10.

3/4 1-6a changed. %he upper limit is now graphically displayed (see Item 238) .

4.1.1.4.b 1 236 Specified 300 pgm surveillance Editorial change; change made to at "RATED 'ZHERMAL HNER equil- clarify the intent of the surveillance librium boron concentration<<. requllXGHIBIlto 1 237 Mathematical symbols are written Editorial change for clarity.

out in words+

1 238 'Ihe new ÃIC limits proposed in Editorial change.

Item 235 are now graphically displayed in Figure 3.1-2 on new page 3/4 1-6a.

AEP:NRC. ATIRCHMEPZ 3 K2!IEGER DESCRIPXXONS FOR D. C. COOK UNIT 2 PRDKSED TECHNICAL SPECIFICATIONS PAGE 38

+

3/4 1-8 SECTION 3.1.2.1 4 DES CRIPPION 239 Footrmte added.. ~

RHKBKS Technical Specification concentration in the HNST is boron suf f1c1eIIt to pzav3.de adequate shutdown margin fram expected operating conditions.

4.1.2.l.a 1 240 t'> 145n is Editorial change.

than or ecpal to 145 F".

3/4 1-11 3.1.2.3 1 241 "the" is removed fram footnote. Editorial change; typoc~phical error correction.

1 24la "ar" is changed to "are". Editorial change; typographical error correction.

1 242 Mathematical symbols are written Editorial changes for clarity.

aut in words+

3/4 1-15 3.1.2.7.a.l 1 243 "of" is added. Word "contained" Editorial changes; typographical 1s removedo error correction; clarification of meaning o 3.1.2.7.b. 1 1 -243a Word "contained Editorial change; clarification of meaning+

3.1.2.7.b.2 8 244 HNST miniasm boron is changed.

concentration ~

limit lmnbaum HNST boron concentration has been increased to provide additional margin for the DXA lang-term cooling criterion. See Attach-ment 13.

~ tt Attachmerrt 13 includes the evaluations of impacts an Unit 2 performed by Advanced Nuclear Heels (Exxon) and MESC.

3.1.2.7.b.3 11 244a '%he recpired HNp temperature is 'Ih minimum EST temperature is. conserva-increased to 80 F. tively raised to the teIagerature Ixqu:ired for operability as a safeguards syspan in modes 1 2 3 & 4 g g ~ Kt18 vallle of 80 F fram the Unit 2 IDCA analysis is conservatively chosen o

AEP:NRC: ATI'ACKKFZ 3 SUMKQK DESCRIPXXONS FOR D. C. COOK UNIT 2 PROPOSED TIXSNICAL SHKZFICATIONS PAGE 39 SECTION +

RIMGRS 4.1.2.7.b ll DESCRIPTION 244b %he EST temperature monitored regardless of will be this is a conservative increase in outside air tetagm~ture.

4 245 Footnote added. %he Technical Specification boron concentration in the RNST is f

suf icient to provide adequate shutdawn margin fram ~~ted apexatizxy canditians.

3.1.2.8.a.l ll 246 Changed BAST aunimum volume.

Substituted "usable" for Boration sources are being charged to select the most conservative volume cantained. fram the Unit 1 and Unit 2 analyses.

For this value the Unit 1 analysis is more conservative.

3.1.2.8.b.1 1 246a Upend volume limit on RHST 'Ihe upper limit of 420,000 gallons is is remavedo the capacity of the tank. Ghe-limit has no effect.

3/4 1-16 3-1.2.8.b.2 8 247 EST minion boron co~ration See remarks for item 244.

is charged.

8 248 RNST boron concentration upper 'Ihe revised cont-~nerrt sump pH analysis limit is changed. and the changeover to hat-leg recircu-lation safeguards analysis recgCire a new upper limit on the RNST concentration.

See Attachment 13.

3/4 1-17 4.1.2.7.b 11 248a 'Ihe RNST temperature will be is a canservative increase in monitored regardless of outside air temperature.

U AEP:NRC: ATI'ACHMENT 3

SUMMARY

DESCRIPTIONS H)R D. C. COOK UNIT 2 PROPOSED TEKSNICAL SPECIFICATIONS PAGE 40 SECTION +

DES CBIPZION REMARKS 3/4 1-18 3.1.3.1 1 249 "Ihe rod" is changed to "Ihe affected'ocV'. Editorial change for clarity.

ACZION c. 1 ACZION c.2 3/4 1-18 'CTION 1 250 ACZION c.2.b is moved fram page Editorial change.

3/4 1-19 c.2.b 3/4 1-19 to page 3/4 1-18.

ACZION 251 Words added to eaphasize that Editorial change.

c>>2 when ACZION c.2 is chosen that items a, b and c plus the choice between items d and e must be performed>>

3/4 1-19 ACZION 1 252 Mathematical symbols are written Editorial change for clarity.

c.2.d out in words>>

ACZION 1 253 Reference to Figure 3.1-2 is re- Editorial change; three loop operation c>>2>>e moved>> in Modes 1 ard 2 was removed for Unit 2 in Amendmemt No. 82.

4.1.3.1. 1 1 254 References to part length rods Editorial change; part length rods are zGHKved>> are not used>>

4.1.3.1.2 1 255 'Ihe words "m the core" are Editorial change. Makes Specifica-removed>> tions of both units more similar.

3/4 1-23 3.1.3.4 1 256 "(228 steps)" is Editorial charge; clarifies meaning of fully withdrawn.

1 257 Mathematical symbols are written Editorial chnge for clarity.

out 3I1 words>>

4.1.3.4 3 258 words "prior to entering Mode 2" Requiring the completion of this test replace "prior to reactor cri- prior to entering MODE 2 is conservative txcalzty to requiring the test prior to cri-ticality. MODE 2 is entered with the reactor subcritical by 1%. However, malcing the requirement mode dependent eases administrative control.

AEP:NRC: ATE'ACHMEÃZ 3

SUMMARY

ITS(ZGPZIONS FOR D. C. COOK UNIT 2 PROKSED TECHNICAL SPECIFICATIONS PAGE 41 PAGE SECPION +

DES CRIETION 3/4 1-24 3.1.3.5 1 259 "(228 steps)" is added. Editorial change; clarifies meaning of fully withdrawn.

1 260 Mathematical symbols are written Editorial change for clarity.

aut in words+

3/4 1-25 3.1.3.6 1 261 iifigures<) is changed to Itfiguren Editorial ~gee 1 262 Mathematical symbols are written Editorial ch-age for clarity.

aut in words+

3/4 1-27 1 263 Page is removed. Editorial change. Blank page not necessary at encl of section.

3/4 2-1 1 264 APL footnote is reaxved. Editorial change; APL is a defined term.

3/4 2-4 Figure 3.2-1 1 265 Figure is redrawn. Editorial change for clarity.

3/4 2-16 Table 3.2-1 8 266 Footnote added to donunent flmr Omitted fram letter to H. R. Denton fram M. P. Alexich dated March 14, Analysis value reduced by the 1986 (Identifier AEP:NRC: 0916I) . To value of the allowance. facilitate this review we are re-transmitting Attachment 7 of AEP:NRC:

0916I as Attachment 12 to this letter.

See page 2 of Attachment 12.

Table 3.2-1 1 267 Footnote ~** is added. See Remarks for Item 266.

1 267a Asterisks moved to right hancl Editorial ch-ape.

column.

Footnote ** 8 268 Ihe wards "at least three" are %his change reflects an analysis previously added. submitted in Attachment 3 to AEP:NRC:0916I for RCS Tavg and AttaWnent 7 of AEP:NRC: 0916I for the pressurizer pressure.

To facilitate this. review, Attachments 3 ancl 7 to AEP:NRC:0916I are retranstnitted as Attachment 18 and 12, respectively, of this letter. See page (vii) of Attachment 18 and page 3 of Attachment 12.

AEP:NRC: AZTAma"m 3 SUMMATE DESCRIPZIQNS KR D. C. COOK UNIT 2 PROKSED TEKSNICAL SPECZFICATIQNS EARMARKS PAGE 42 SECZXON +

DESCRIPTION 3/4 2-18 Table 3.2-2 8 269 Allowance for readability Qmitted from letter to H.R. Denton included for RCS Tavg and fram M.P. Alexich dated March 27, 1986 Pressurizer Pressure. 'Ihe (Identifier AEP:NRC:0916P) . See Attacb-allowance was calculated ment 3 to AEP:NRC:0916I for RCS Tavg consistently with footr~ *. and Attachment 7 of AEP:NRC:0916I for the pressurizer pressure. To facilitate this reviev, Attachments 3 and 7 to AEP:NRC:0916I are retransmitted as Attachment 18 and 12, respectively, of this letter. See page (vii) of Attachment 18 and page 3 of Attactunent 12.

3/4 2-19 3.2.6 1 270 ALUNABZZ HMK ZEUS is Editorial change; AIZQK(BZZ H3NER ZZVZL capitalized (APL) is a defined term.

1 271 Expression for APL is revised to Editorial change; APL cannot be greater more accurately reflect the than 100% of Rated Bmrmal Rm~.

mear~ of APL.

1 272 Second "F (Z)" is replaced by Editorial change for clarity.

"measureAat channel factor".

3/4 1 273 ACZION statements are moved from Editorial change.

to 2-19'/4 2-20 page 3/4 2-19 page 3/4 2-20.

3/4 3-3 Table 3.3-1 1 274 <tinn Editorial change; typ~phical Items 13 & 14 error correction.

3/4 3-4 Table 3.3-1 1 275 Clarifications made to properly Editorial change; this change Items 21 & 22 identify which ACXXON statements corrects a format ermr made in apply to each applicable mode. the issuance of Amendment No. 86.

3/4 3-12 Table 4.3-1 7 276 Zoss of FlowDm Zoops Functio- Ibis was omitted fram letter fram Item 13 nal Unit has an added eumiption M. P. A16xlch to H. R. Denton fram Specification 4.0.4. dated Hach,27, 1986

AEP:NRC: ATZACSMENT 3 SUMMA'ESCRIPZIONS H)R D. C. COOK UNIT 2 PROPOSED TEXXENICAL SHKXFICATIQNS PAGE 43 SECTION +

DES CRIPZION EKMAERS (Identifier AEP:NRC:0916P) . Ex-emption is provided for surveil-lance which must be performed in the applicable mode. Ghe change was apgmved for "loss of Flmr Single Loop" in Amendment 82 to DPR-74.

3/4 3-28 Table 3.3-5 1 277 Reactor trip is removed fram Editorial change; to make the proposed Item 8a description. Technical Specifications between Units more similar. %be response time for ESF Steam Generator Water Ievel-High High turbine trip is not mdeled in the current analysis of record.

3/4 4-2 3.4.1-2.d ll 277a Criterion for the operability of reactor coolant loops based Table 3.3-3 recgures at least t?u~

loops ogemting abave P-12. '%his on P-12 is added.

loops. ~

ensures flmr through RH) by-pass provision is added for consistency with Table 3.3-3.

An appropriate ACTION statement has been proposed to correspond ACZION b c

ll 277b ACZION statements added to address too few reactor coolant Proposed to maintain similarity to ACZION Unit 1. See Item 135.

loops when control rods are capable of withdrawal. Old.

ACZION b becames ACZIQN d.

3/4 4-2a ACTIQN d 1 277c ACZION d and footnotes Editorial charge; additional text moved fram previous page. requires moving this material.

3/4 4-3 3.4.1.3 11 277d ACTION statement added to Proposed to maint~ similarity to ACZION b address too few reactor coolant Unit 1. See Item 135.

loops when control rods are capable of withdrawal. Old ACTION b becomes ACZION c.

3/4 4-4 3.4.2 4 278 Footnote added. The Technical Specification boron concentration in the RNST is sufficient to provide adequate shutdown margin fram expected

AEP:NRC: ATI'ACHMENT 3

SUMMARY

DESCRIPZIONS POR D. C. COOK UNIT 2 PBOPOSED TZXZNICAL SPECIFICATIONS PAGE 44 PAGE +

DES CRIETION &HARES operatirg conditions.

3/4 5-1 3.5.l.c 8 279 Minimum accumulator boron 'Ihe minimum accumulator boron concen-concentration is change. txation limit has been incr~ed to provide additional maxgin for the IOCA lorg-term coolirg criterian.

See Attachment 13.

3.5.l.c 8 280 Maxirmm accuaailator boron t tll concentration is changed. and changeover to hat-leg recirculation safeguaxds analysis establish a nev ~mr limit on accumulator boron cxexentratian.

See Attactuaent 13.

3.5.5.a 1 280a Qpper volume limit on RWST 'Ihe ~mr limit of 420,000 gallons is is zBEI3vede the capacity of the tank. %be limit has no effect.

3/4 5-11 3.5.5.b 8 281 Minimum RWST boron concentration %he minhnum RWST boron concentration is charged. limit has been inn~sed to pmvide additional margin for the IDCA, lang-term cooling criterion. See Attach-ment 13.

8 282 Maximum RNST boron is concentration Wtt t changed. and chargeover to hot-leg recirculation safeguard analysis establish a new ~xz limit on RWST boron concerrtratian. See Attachment 13.

4.5.5.b ll 282a Ihe RWST teraperature wi11 be monitored regardless of outside this is a consexvative m~e in surveillance requirements.

air temperature.

3/4 8-5 3.8.1.2 4 283 Foatnote added. 'Ihe Technical Specification boran concentration in the RWST is sufficient to provide adequate shutdawn margin fram m@ected operating corditions.

AEP:NRC: ATZLCSMENT 3

SUMMARY

DESGGPZIONS FOR D. C. COOK UNIT 2 PROKSED TECHNICAL SHKXFICATIONS PAGE 45 SECTION +

DESCRIPTION REKQRS 1 284 EÃlGting footnote

is changed Editorial change.

to footrmte **.

3/4 9-1 3.9.1 1 285 Mathematical symbols are written Editorial change for clarity.

out iI1 words+

3.9.l.b ll 285a Ihe required boron concentxation for refueling is inca~ed to

%he required concentration vatively inn~ed to is conser-ACZION agree with the 2400 pgm. HNST concentration. The result is a substantial increase in the amount by which the core is shutdown during refueliag.

4 286 Footnote added. 'Ihe Technical Specification boron concentration in the EST is sufficient to provide adequate shutdown margin fram expected operating conditions.

3/4 9-2 3.9.2 4 287 Footnote added. We Technical Specificatian boron concentration in the RNST is sufficierrt to provide adequate shutdown margin fram expected operative conditions.

1 288 Footnote rewed. Editorial charge; the 1984 Refuelirg Outage has been completed.

3/4 9-8 3.9.8.1 An analysis was performed to reduce to 2000 gpm. the required reactor coolant flmr rate to 2000 gpm. See Attachment 5 for discussion of heat removal, mixup, and stratisfications considerations. See Attach-ment 11 for dilution transient considerations.

1 290 Mathematical symbols are written Editorial change 'for clarity.

out in words.

3/4 10-3 3.10.3.b 12 291 Specification is rewarded. Clarifies intention of specifica-tion. See letter fram Nestirghouse found in Attachment 6,- Item Number 14.

AEP:NRC: ATZACHMEÃZ 3

SUMMARY

DESCRIPZIQNS FOR D. C. COOK UNIT 2 PBOKGED TECHNICAL SPECIFICATIONS PAGE 46 SECTION +

DES CRIPZION REKGRS 1 292 "reactor trip setpoints" is Editorial charge for consistency changed to "Reactor Trip Set- with 3.10.4.b.

points" .

3.10.3.b 1 293 Mathematical symbols are written Editorial change for clarity.

ACZION out in worse 4.10.3.1 3/4 10-4 3.10.4.b 12 294 Specification is reworded. Clarifies intention of specifica-found in Attaclment 6, Item Number 14.

4.10.4.1 1 294a Editorial charge.

1 295 Mathematical symbols are written Editorial change for clarity.

out in words+

6-19 6.9.2.h 1 296 Moderator Ter~rature Goefficient Editorial charge; A Special Report is added to the Special Reports is to be submitted to the NRC within list. 10 days of exceeding the limit of Figure 3. 1-2.

6.9.2.e 1 296a Gamma is rented. Editorial change; typographical error correction.

B 3/4 1-3 3/4.1.2 B 297 Revisions made to the ~iption 9he higher boron concentration of (Bases) of the RNST as a boration source. the HNST is reflect~ in the basis.

Volumes used in the Technical Specifica-tions which bound Units 1 and 2 are discussed. See Attachment 13.

B 298 pH value limits are charged. Limits reflect the analysis in Attactuxnt 13.

B 3/4 4-la 3/4.4.2 B 299 Text is combined to one page; Editorial charge; remakes dupli-B 3/4 4-2 3/4.4.3 B 3/4 4-la is to be removed. .cation of text that was included 3/4.4.4 with License Amendment No. 82.

(Bases)

AEP:NRC: ATI'ACHKM' SUMMATE DESCRIPTIONS PQR D. C. COOK UNIT 2 PROKSED TEKSNICAL SHXXFICATIONS PAGE 47 SECTION +

DESCRIPTION &RMES B 3/4 5-3 3/4.5.5 B 300 pH value limits are changed. Limits reflect the analysis in (Bases) Attachment 13.

B 3/4 9-1 3/4.9.1 B 301 the basis section fram SIS is She required coaaer~tion is conserva-substituted for existing basis tively increased to agree with the EST and is augmented with a concentration. 'Ihe remit is a substan-discussion of the increase in tial increase in the anent by which the 1 core is shutdown during refuelizg.

to 2400 pgm.

Attachment 5 TO AEP:NRC:0916W ANALYSIS OF HEAT REMOVAL AT 2000 GPM PRIMARY FLOW AND EVALUATION OF MIXING AND STRATIFICATION

Attachment 5 to AEP:NRC 0916W T/S 3/4.1.1.3 (Reactivity Control Systems - Boron Dilution) presently requires an RCS flow rate of '3000 gpm whenever a reduction in RCS boron concentration is being made. As discussed in the Bases for this T/S, the purpos'e of this requirement is to provide adequate mixing, prevent boron stratification, and ensure that reactivity changes will be gradual during boron concentration reductions in the RCS. Similarly, T/S 3/4.9.8.1 (Refueling Operations - Residual Heat Removal and Coolant Circulation) requires 3000 gpm of RHR flow during Mode 6 operation. According to the Bases for this T/S, its purpose is to (1) ensure sufficient cooling capacity is available to remove decay heat and (2) maintain sufficient coolant circulation through the reactor core to minimize the effect of a boron dilution incident and prevent boron stratification. In practice, however, the 3000 gpm requirement can present severe operational difEiculties because of the possibility of pump vortexing. This concern exists during RHR system operation in Mode 5 (cold shutdown), when the RCS may be partially drained down to facilitate various maintenance operations (half-loop operation) ~ Problems with vortexing of the RHR pumps during halE-loop operation has been a recurring problem in the industry, and was the subject of IE Information Notice No.86-101. Because loss of the RHR pumps due to vortexing could conceivably cause the loss of decay heat removal capability, we have performed analyses to demonstrate that 2000 gpm flow is 'sufficient for the purposes of T/Ss 3/4.1.1.3 and 3/4 '.8.1.

Analyses addressing the boron dilution incident concerns are discussed in 4 for Unit 1 and in Attachment 11 for Unit 2. Analyses for mixing and boron stratiEication concerns and decay heat removal capability are discussed below. ReEerences for these analyses and a nomenclature list are included at the end of thi.s attachment.

1. Mixin and Boron StratiEicatio

~Summa'oron stratification was assessed by comparing the turbulence and core crossflow that would exist at 2000 gpm to what exists at 3000 gpm.

At 3000 gpm (the current technical specifications limit) boron stratification would not occur. The evaluation showed that the Reynolds number in all RCS piping, RHR piping and the reactor vessel downcomer is in the turbulent region. Turbulence in the downcomer would promote mixing, thereby reducing any concentration gradients that may have existed when the fluid entered the downcomer. Upon entering the lower plenum of the reactor vessel, the momentum of the fluid combined with the effects of a sudden expansion would tend to entrain surrounding fluid, further reducing concentration gradients. Finally, crossflow in the core would promote additional mixing. The crossflow is a function of the Reynolds number to the 0.9 power (References 2 and 3). The crossElow at 2000 gpm would be 69% of that at 3000 gpm. Thus, a significant amount of crossflow would exist. (See analysis section.)

Once the flow exits the core, the RHR piping turbulence would be very high, and considerable mixing would continue, especially as the fluid flows through the pump.

Based on the forgoing mixing evaluation, it is concluded that boron stratification is no more of a concern with 2000 gpm RHR flow than with 3000 gpm RHR flow.

Details of Anal sis There are several places in the RCS where mixing could occur. These include the reactor coolant system piping, the reactor vessel downcomer,

--the reactor vessel 'upper and lower plenum, the core region, and the RHR system piping. For all of these regions except the plenums, Reynolds numbers were calculated. (Although a calculation was not made for the plenums, the core area was determined to be the least turbulent flow region and thus would bound the plenums.) These Reynolds numbers are listed below.

Reynolds Numbers Location Re nolds Number

~3000 m ~2000 m Reactor Inlet 151,600. 101,000 Reactor Outlet 144,100. '6,000.

Reactor Downcomer 49,400. 33,100.

Reactor Core 840. 560.

RHR Piping 1,271,000. 847,000.

As can be seen in the table above, all areas of interest except the core had Reynolds numbers well in excess of 4000, at 2000 gpm flow, and thus it was concluded that the flow in these regions would be turbulent and that adequate mixing would occur.

In the core and plenum regions, flow is laminar. However, there is mixing due to crossflow within the core region. A mixing parameter B G/G K Re

-0.1 exists which ratios cross flow in the core to the average core flow (Ref.

2). Since the Reynolds number is directly proportional to the flow in the system, the equation can be modified to give G K'(G)

From this, the crossflow of two different flow rates can be compared by using:

0.9 G2/Gl [G2/Gl)

Using this equation, it can be seen that the cgo~sflow at 2000 gpm RHR flow compared to 3000 gpm would be (2000/3000) ' or approximately 69% of the value at 3000 gpm. This remaining crossflow, together with the mixing that would occur in the piping and the downcomer, is judged to be sufficient to prevent significant boron stratification.

N

2. Deca Heat Removal

~Summa'alculations were performed to determine the minimum RHR flow which would be required to remove decay heat. The assumptions used in the analysis were that the lake 'water temperature is0 85 0 F and the maximum temperature of the reactor coolant water is 200 F. The lake serves as the ultimate heat sink for the decay heat generated in the core. This temperature ultimately determines the required coolant flow to the reactor core. The maximum reactor coolant temperature is set by the technical specification limit of 200 0 F in Mode 5, cold shutdown. To account for uncertainties, additional calculations were performed with a margin of0 20%

added to the decay heat value and the lake temperature increased to 95 F.

It was also assumed that the product of overall heat transfer coefficient and surface area (UA) was constant and equal to the design value in the CCW heat exchanger. A ratio was computed for UA as a function of reduced flow .for the RHR heat exchanger. Flow rates other than RHR flow rate (such as CCW and ESW loops) were also assumed to remain constant and equal to 0 design values. Constant pressure specific heat was taken as 1.0 Btu/ibm/ F for all flow streams. Decay heat was calculated using Reference 1 methodology.

These calculations demonstrated that RHR flow of 2000 gpm would be more than sufficient to remove decay heat, even with the reactor drained to the half-loop condition.

Details of Anal sis This section provides details of calculations to determine the minimum flow rate required to remove the decay heat from the reactor. The RHR system was modeled using the flow diagram shown in Figure 1. The problem involves six equations and six unknowns (the temperature of each stream). The basic equations to be solved are:

(1) Q i i C p

DT and 1

Equation (1) describes the sensible heat gain or loss in the coolant.

Equation (2) describes the heat transfer between the fluids flowing on the shell side and the tube side of the heat exchanger. The log mean temperature difference, hT in equation 2, compensates for the fact that the temperature difference between the hot and cold fluid may change as both fluids traverse the heat exchanger.

The product of the overall heat transfer coefficient, U, and the heat exchanger surface area, A, was determined from the design condition given in Table 1. This was accomplished by rearranging equation 2 to give (3) UA (6 T~) (F)

The calculated values of UA are given in Table l.

The decay heat used in the equation was determined by (4) P/P 0 (t0 ,t S ) P/P 0 (

,t S ) - P/P 0 ( , t0 + t S

)

where P/P 0 - Power to full-power ratio t0 - Effective full power seconds at 3411 MW ts .

Number of seconds since shutdown (5) P/P (~, t) -At Where A and a are values obtained from Reference 1.

Based on 1202 effective full-power days (Reference 5), the decay heats were calculated for decay times of 2.5 to 6.0 days. These results are given in Table 2.

The component cooling water has heat loads other than the decay heat from the core. The total amount of these )eat loads was obtained from the design values and was found to be 34.9 (10 ) Btu/hr. For the6calculation of the minimum low flow, the decay heat at 2.5 days, 40.4 (10 ) Btu/hr, was used. This made the total heat load 75.3 (10 ) Btu/hr.

Mass flows in the system, (other than RHR flow which will be calculated), were obtained from the design values. The component cooling water flow which is diverted to the auxiliaries is summarized in Table 3.

The mass flows used in the calculation are summarized in Table 4.

The minimum mass flow rate required to remove decay heat after 2.5 days with a lake temperature of 85 F was determined by iteration to be approximately 1000 gpm.

To account for uncertainties in the decay heat value, a margin of 20%

was conservatively added, the lake temperature was conservatively increased to 95 0 F and the calculation repeated. When this was done, the minimum required flow was determined to be approximately 1450 gpm.

TABLE 1 Heat Exchanger Design Conditions RHR Heat Exchanger Design Heat Load, Btu/hr 41.1 x 10 6 Shell Side Inlet Temperature, 0 0 F 95.

Tube Side Inlet Temperature, F 140.

Shell Side Outlet Temperature, 0 0 F 111.6 Tube Side Outlet Temperature, F 112.3 Calculated UA, Btu/hr F 1.836 x 10 CCW Heat Exchanger Design Heat Load, Btu/hr 76 x 10 6 Shell Side Inlet Temperature, F 114.

Tube Side Inlet Temperature, 0 F 76.

Shell Side Outlet Temperature, F 95.

Tube Side Outlet Temperature, 0 F 92.

Calculated UA, Btu/hr F 3.71 x 10 References 5, 7

TABLE 2 Decay Heat as a Function of Time Time after Degay Heat, Shutdown, Days 10 Btu/hr 2.5 40 '

3.0 38.1 3.5 36.2 4.0 34.6 4.5 33.3 5.0 32.1 5' 31.1 6.0 30.2

TABLE 3 CCW Auxiliary Cooling Water Flows Component Flow,, gpm Reactor Coolant Pump 560 Sealwater Heat Exchanger 38 Letdown Heat Exchanger 300 Spent Fuel Heat Exchanger 1500 RHR Pump 10 SI Pump 40 Spray Pump 20 Charging Pump 90 Penetrations 300 Gas Compressor 13 Reactor Support 40 TOTAL 2911 Reference 6

TABLE 4 Mass Flows Used In Analysis 6

Flow Stream Mass Flow, 10 lb/hr (Refer to Figure 1)

M To be calculated M 2.56 M

3 4.67 M4 4.0 References 4, 5, 6

REACTOR

~

VESSEL T4 M4 RESIDUAL HEAT HOT LEG OLD LEG REMOVAL (RHR) AUXILIARY HEAT EXCHANGER HEAT LOADS COMPONENT COOLING T3 WATER (CCW) LAKE HEAT TS WATER T2 EXCHANGER T3 gag)

M2 RHR PUMP CCW PUMP FIGURE 'I. RHR LOOP CONFIGURATION

References (1) NUREG/CR-2507 "Background and Derivation of ANS 5.4 Standard Fission Product Release Model," January, 1982.

(2) XN-NF-81-73, "Turbulent Mixing in Rod Bundles," October, 1982.

(3) McCabe and Smith, "Unit Operations of Chemical Engineering," 1956, p.

53.

(4) FSAR, Table 9.8-5.

(5) FSAR, Table 9.5-3.

(6) FSAR, Table 9.5-2.

(7) FSAR, Table 9.3-2.

NOMENCLATURE Ql

=

heat flow Btu/hr M.

i mass flow rate for lb/hr stream i C heat capacity Btu/lb F P

dT.i temperature difference for stream i F

U.

1 overall heat transfer Btu/hr ft2 0 F

coefficient for heat exhanger i heat transfer area for heat exchanger i LM, i log mean temperature F difference for heat exchanger i P

0 actual reactor power rated reactor power effective full-power Days operating time ts time since shutdown Days cross-flow lb/ft2 sec average coolant flow lb/ft2 sec mixing parameter dimensionless F.

i configuration correc- dimensionless tion factor for heat exchanger i K, K'roportionality constants

Attachment 6 TO AEP:NRC:0916W EVALUATIONS OF PROPOSED TECHNICAL SPECIFICATIONS PERFORMED BY WESTINGHOUSE ELECTRIC CORPORATION

Energy Systems Westinghouse Power Systems Service Oivision Electric Corporation Box 355 Pittsburgtt Pennsylvania 15230 0355 AEP-87-135/REV. 2 March 6, 1987 Mr. E. G. Lewis Nuclear Materials and Fuel Management Section American Electric Power Service Corporation One Riverside Plaza Columbus, OH 43216 AMERICAN ELECTRIC POWER SERVICE CORPORATION Res onse to D. C. Cook Unit 1 Tech S ec Chan es

Dear Mr. Lewis:

The purpose of this letter is to revise our previous transmittal of the attached Technical Specification Changes. The revisions are minor in nature and involve items numbers 4 and 9.

In the "response" section of item number 9, reference to "Reference 4" was deleted. In item number 4, the Technical Specification originally supplied to Westinghouse relative to the MTC change was replaced by AEP with another Technical Specification page. At AEP's request, Westinghouse reviewed this change and found it acceptable. It is now a part of the, attachments.

Please note these changes in the attachments to this letter. If you have questions, please do not hesitate to contact us.

Very truly yours, Q=(

H. C. Walls, Project Manager Mid-America Area U. ST Nuclear Projects cc: J ~ G. Feinstein E. G. Lewis V. Vanderburg J. M. Cleveland W. G. Smith, Jr.

8981 f/fc

ITEM NUMBER 1 AEP Comment: Review the revised basis (specification 2.2.1 reactor trip system instrumentation setpoints) for loss of flow. Verify that the revisions made by AEPSC in removing reference to 3-loop operations are consistent with Westinghouse methodology.

Response: The reference to three loop operation may be deleted. In addition to what was deleted in the basis by AEP, the following statement (Page B 2-6) should also be deleted since it is applicable for three loop operation:

"'Ibis latter trip will prevent the minimum value of the DNBR from going below the applicable safety analysis design limit DNBR value for each fuel type, (as listed in the bases for (Section 2. 1. 1) during normal and operational transients and anticipated transients when three loops are in operation and the overtemperature Delta-T trip setpoint is adjusted to the value specified for all loops in operation."

ITEM NUMBER 9 AEP Comment: Confirm which parameters in DNB specification (3.2.5, Table 3.2-1) have readability allowances. What is the accurate manner to address the error penalty in flow? (Analog of 3.5$ penalty in standardized technical specifications). Confirm the treatment of measurement allowances in the draft DNB basis (3/4.2.5 DNB parameters) is correct.

Response: The value for reactor coolant system T-avg (570.4 degrees F) was verified to include measurement uncertainties and is the indicated value as read in the control room. The T-avg indicator for at least three loops is read, added together, and divided by the number, of loops measured (three or four), to obtain the reactor coolant system average temperature. It is recommended that the footnote in the proposed tech. specs. (Table 3.2-1),

"indicated average of operable instrument loops" be changed to "indicated average of at least three operable instrument loops". The value for pressurizer pressure was verified to be the safety analysis bounding value.

We value for reactor coolant system total flow rate ( 1.386 times ten to the eighth power pounds per hour) in Table 3.2-1, (DNB parameters) is an "indicated" value to which the flow rate must be compared to, to demonstrate compliance with this specification.

It is acceptable to add the statement "the indicated values of T-avg and flow include allowances for instrument errors." To the basis of specification 3/4.2.5, DNB parameters. It is recommended that the last statement in the first paragraph be revised as follows: "Measurement uncertainties have been accounted for in determining the DNB parameters limit values.

ITEM NUMBER 10

3. AEP Comment: "Review new primary flow surveillance requir ements (specification 3.2.5 DNB parameters and basis for 3/4.2.5). Monthly surveillance removed per discussion with R. Jansen in connection with Unit 2 T/S revisions.

Response: The monthly total flow rate surveillance (specification 4.2.5.2 in the current D. C. Cook Unit- 1 tech specs) may be removed since the total flow rate is verifi:ed once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . It is acceptable to add the

'surveillances on the channel calibration of the flow indicators and the total flow rate measurement. The revised basis for specification 3/4.2.5 adding the discussion on the new surveillances added, and the deletion of the discussion on the monthly flow surveillance is acceptable.

ITEM'UMBER 11 AEP Caanent: Confirm that the augmented startup test progr am is canpleted.

Response: The augmented startup test program is canplete and the proposed tech spec change in specification 3.10.2 may be implemented.

ITEM NUMBER 4 Provide justification for changing MTC fran a step to a ramp function of power as proposed by Westinghouse.( 18)

Response: A safety evaluation of the proposed change to the D. C. Cook Unit 1 moderator temperature coefficient (MTC) Technical Specification 3/4.1. 1.4, has been canpleted. Specifically, American Electric Power has expressed an interest in changing the form of the PC spec fran a "step>> of

+5 t8 0 pcm/ F at 70$ power to a "ramp" of +5 pcm/ F at 70$ power to 0 pcm/ F at 1004 power.

The following accidents, determined to be sensitive to a positive MTC, were analyzed in support of the OFA transition for the Cycle 8 reload transition:

RCCA Bank Withdrawal fran Subcritical RCCA Bank Withdrawal at Power Loss of Reactor Coolant Flow (including Locked Rotor Analysis)

Loss of External Load Excessive Heat Removal Due to Feedwater System Malfunction RCCA 'Ejection

4 0

With two exceptions, the current safety analyses were based on a +5 pcm/ F MTC, which was assumed $o remain constant for variations in temperature. The assumption of a +5 pcm/ F MTC existing at full power is conservative, since the proposed Technical Specification requires that the coefficient be linearly ramped to zero above 70$ power. Therefore, the conclusions presented in the cycle 8 reload transition safety analyses (the current analyses) remain valid.

The RCCA ejection and RCCA Bank Withdrawal fran'ubcritical analyses performed in support of the Cycle 8 reload transition were based on a coefficient which was at least +5 pcm/ F at, the appropriate zero or full power nominal average temperature, and which became less positive for higher temperatures. This was necessary since the 'IWINKLE computer code used 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 conclusions of the Cycle 8 r eload transition analyses remain valid.

Since this proposed Tech Spec change does not alter the previous Tech Spec limits for MTC at 0$ power and at 100$ power, the results of the large break and small break LOCA analyses and long term core cooling calculation will not be affected by this change.

A copy of the proposed D. C. Cook Unit 1 Technical Specification 3/4.1.1.4 is attached, incorporating the MTC change. A Nuclear Safety Evaluation Checklist has been completed for this evaluation and is attached.

ITEM NUMBER 14

5. AEP Comment: Documentation may be needed that states that the high flux low setpoint is sufficient during physics tests. (Specifications 3. 10.4 physics tests and 3.10.5 natural circulation tests) we have received the interpretation verbally from R. Jansen on Westinghouse.

Response: Westinghouse recommends that 3.10.4 B. and 3. 10.5 B. be changed to read as follows:

The reactor trip setpoints for the operable intermediate range, neutron flux and the power range, neutron flux, low setpoint are set at less than or equal to 25$ of rated thermal power.

The justification for this change is for clarification purposes, the intent of the spec is not changed.

NAR 03 '87 16:Z5 AEPSC-NUC OPER 2004 P02 ATTACHMENT I AC V 3.1.1.4 The moderator temperature coefficient (MTC) shall be'.

a. Within the region of acceptable operation in Figure 3.1-2, and

b. Less negative than -3.5 x 10 bk/k/ F at RATED THERMAL POWER.

dKIQH'ith the moderator temperature coefficient outside any one of the above limits, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURV 4.1.1.4.1 The MTC shall be determined to be within its limits by confirmatory measurements. MTC measured values shall be extrapolated and/or compensated to permit direct comparison with the above limits.

4.1.1.4,2 The MTC shall be determined at the following frequancias and THERMAL POWER conditions during each fuel cycle:

a. Prior to initial operation above 3% of RATED THERMAL POWER, after each fuel loading, b, At any THERMAL POWER, within 7 EFPD after reaching a RATED THERMAL POWER equilibrium boron concentration of 300 ppm.

With K than or equal to 1.0 ff greater off Saa Special Test Exception 3, 10,4 D, C. COOK - UNTT 1 3/4 1-3 AMENDMENT NO.

ATTACHMENT lI WESTINGHOUSE PROPRIETARY CLASS 2 The following discussion pertains to a discrepancy between a statement in the D. C. Cook Unit 1, Cycle g reload transition safety report Rupture of a Steam Pipe write-up and the plantt s actual conf jguration.

Appendix C.3.11 of the D. C. Cook Unit 1, Cycle 5 reload transition safety report discussed the analysis of a Rupture of a Steam Pipe. Condition B f th rite-up stated that <<Since the steam generators are provided with integral flow restrictors with a 1,4 square foot throat area, any rupt ure with a break ares greater than 1 A square feetf regardless of location, would have the aune effect on the NSSS as the 1,4 square foot break," In actuality, the steam generators for D. C. Cook Unit 1 are not equipped with integral flow restrictors. However, the reanalysis perf~cd did assume the correct plant oonfiguration. The most limiting ateamline break scenario was assumed in the analysis. The case analyzed fcr a Rupture of a Steam Pipe was a oomplete severance of a pipe at the outlet of the steam generator (break area <<4.6 square teat) upstream of the flar rastriotor, with the plant initially at no-lead oonditions, full reactor ooolant flar with offsite power available. As such, oondition B of Appendix C.3.11 of the Cycle 8 reload transition safety report should be replaoed with the f ollowing: ~

K. The most limiting case of a rupture of a steam pipe wss analyzed. This was determined to be a break at the outlet of the steam generator (br eak area a 4.6 square feet) upstream of the flow restrictor, with the plant initially at no-load oonditions, full reactor coolant flar with offsite power available. This case has been oonsidered in determining the core power and RCS transients,

ATTACHMENT II

%5TZRGBOUSE PROPRIETARY CLASS 2 2~ Th e t 011 ofing provides a discussion of a Feehrater System Malfunction transient assumption, regarding thee modelling mcd ng of reactor trip on turbine trip, fcc the Cycle 8 reanalysis.

Discussion The ourrent R<<eh>ster System Malfunction analysis (presented in~<< Cycle d t sition safety analysis report) vas yerfcrmed assuming a fully oyen feedvater oontrol valve and is tenainated by a steam g enerator hi-hi level trip ign a 1 vhich closes all PV oontr ol and isolation ralvci, trips s

the Fit pumps, and trips the turbine. The feechrater system mal function event is the only RSAR accident that assumes a turbine trip on steam generator hi-hi level. A r<<actor trip on turbine trip @as then assumed to prevent reactor ooolant heatup oonsistent arith the oooldmrn oharacteristics of the feedwater malfunction event. The reactor trip on turbine trip vas assumed as an antic1patory trip. If the reactor triy @as not assumed, the transient would turn into a heatup <<vent - in part 1 c ul ar, a Ioss s of normal feehrater due to the feedvater isolation hi h C occurs on steam generator hi-hi level. A reactor trip mould v then be provided by a lmlav steam generator eater level signal. Further, Further the reactor trip on turbine trip is not raquired for oore yrotection for the feedvater aalfirction event. The results (rLLnimta DNBR) of the feechrater malfunction accident mould be essentially unchanged was not assumed to occur if the reactor trip on turbine trip. Therefore, a reactor trip on turbine trip is not required in any non-LOCA transient for oor e protection.

SECL NO NS-SECL-B7~42 Customer Reference No(s) 0 o C 00 Westinghouse Reference No(s)

AF No. A- <4 WESTINGHOUSE NUCLEAR SAFETY EVALUATION CHECK LIST 2 > NUCLEAR PLANT (S) . C. ook ni

2) CHECK LIST APPLICABLE TO! han e o TC ec (Subject afChange)

'5) The written safety evaluation of the revised procedure, design change or modification required by 10CFR50.59 has been prepared to the extent required and is attached. If a safety evaluation is not required or is incomplete for any reason, explain on page 2.

Parts A and B of this Safety Evaluation Check List are to be completed anly an the basis af the safety evaluation performed.

CHECK LIST PART A (5. 1) Yes No g A change to the plant as described in the FSAR?

(5.2> Yes No g A change to procedures as described in the FSAR?

(5.5) Yes No g A test ar experiment not described in the FSAR?

(3.4> Yes g No A change to the pl ant technical specifications (Appendix A to the Operating License)?

4) CHECK LIST PART B (Justification for Part B answers must be included on Page 2.)

(4. 1) Yes Will the prabability af an accident previously evaluated in the FSAR be increased?

(4 2) Yes Mill the consequences af an accident previously evaluated in the FSAR be increased' (4. 5) Yes No }( Hay the probability of an accident which is different than any already evaluated in the FSAR be created?

(4.4> Yes No g Will the probability af a malfunction of equipment important to safety previously evaluated in the FSAR b inct eased' (4. 5> Yes Will the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR t increased?

(4 b) Yes Hay the possibility of a malfunctian af equipment important to safety different than'ny already evaluated in the FSAR be created?

(4.7) Yes Will the margin af safety as defined in the bases to any technical specification be reduced?

Page 1 af 2

~ ~ NS-SECL-87~1 lf the answers to any of'he above questions are unknown, indicate under g) REMARKS and explain below.

lf the answer to any of the above questions in 4) cannot be answered in the negative, based an written safety evaluation, the change cannat be appraved without an applicatian far license amendment submitted to NRC pursuant to 10CFR50.90.

5)

REMARKS'he follawing summarizes the Justification,upon the written safety evaluation, (1) f or answers given in Part B of the Safety Evaluatian Check Lists An evaluation has determined that the current safety analyses far D. C.

Cook Unit 1 support a Tech Spel: change for the moderator temperature coefficient (spec 5/4. 1. 1.4) ~ The spec will be changed from a "step" af

+5 ta 0 pcm/>F at 70/ power to a "ramp" af +5 pcm/iF at 70% power to 0 pcm/>F at 100% power.

1) Reference to document(s) containing written safety evaluation!

t j

S-OP 8-TA- l<<B7-0>B j

FOR FSAR UPDATE Sections Page(s)l Table (s) I Fi gurc (s)!

'easan for/Description af Change!

6) APPROVAL LADDER

6. 1) Prepared by (Nucl ear Saf ety) t i/z.z

<<<<<<<<<<<<<<<<<<<<<<ates <<<<<p<<

6.2) Coordinated with Engineer(s)l <<<<<<<<<<<<<<<<<<Date I <<g I 6.5) Coordinating Group Manager(s)t--

6 4) Nuclear Saf ety Qt oup Manage I -~-jWZ== =Date> C- -

Page 2 af 2

1 FIGURE 3.1 2.

Moderator Temperature Coefficient (MTC)

~ ~

MTC x 10ie4hk/k/dag J'.00 F.)

r.g

~

I Unac eptab Ope eton 0.50 0.00

-0.50

-1.00 Acc phabte Opera OA

-$ .50

-2.00 p -2.50 fr]

-3.00

~

a 0

0

~ 0 10 20 30 40 50 60 70 80 SO fOD PERCENT RATED THERMAL POSER

Attachment 7 TO AEP:NRC:0916W ROD INSERTION LIMIT INTERCEPTS SUPPLIED BY WESTINGHOUSE ELECTRIC CORPORATION

BAR 5 Ig8, Nr".lear Fuel Oivis!on Westinghouse Water Reactor Electric Corporation Divisions :"ox 39i2 P!!rsourePernsy van!a e5230 33r2 Feb~ 27, 1986 86&:*W-0019 Indiana and Michigan Elect ic Co. W-AEP/0243 c/o Joseph L. Bell Engineer, Nuclear lhteria's and Fuel Keywords: AEP RIL Management 'Tech-Specs Arrerican Electric Poacher Service Corp.

One Riverside Plaza, 20th Floor Columbus, OH 43215

Dear Mr. Bell:

AMERICAN ELECTRIC RX~ SERVICE CORPORATION D.C. COOK i'NIT 1 TECHNICAL SPECIF1CATIONS ROD INSERTION LIMITS Attached are change pages to be incorporated in the D.C. Cook Unit 1 Technical Specifications. 'Ihe RIL limit lines being submitted here for 3-loop and 4-loop operation are no different from the ones in the current Tech Specs.

At your request, Westinghouse is incorporating on those limit lines the actual endpoints in steps withdrawn at both HZP and HFP for control banks D and C.

If you have any questions, please call.

Very truly yours,

.C. k.lier Project Engineer NFD Fuel Projects cc: M.P. Alexich J.M. Cleveland D.H. Malin w/enc.

V.D. Vanderburg W.L. Zimmermann

WESTINGHOUSE PROPRIETARY CLASS 2 CDC 058 February 26, 1986 F. J. Silva, 412/374-2189 Westinghouse Nuclear Fuel Division Core Engineering MMOB-301 MS 3-28 P. 0. Box 3912, Pittsburgh, Pa. 15230 MEMO TO: J. C. Miller NFD Fuel Projects CC: B. M. Bowman B. J. Johansen

SUBJECT:

D. C. Cook Unit 1 Tech Specs Rod Insertion Limits KEYWORDS: AEP TECH-SPECS RIL Attached please find change pages to be incorporated in the D. C. Cook Unit 1 Technical Specifications. The RIL limit lines being submitted here for 3-Loop and 4-Loop operation are no different, from the ones in the current Tech Specs.

At AEP request we are incorporating on those limit lines the actual endpoints in steps withdrawn at both HZP and HFP for control banks D and C.

Please send this information to AEP to be submitted to the NRC together with other Tech Specs changes.

F. J. Silva CE Core Design C APPROVED: W. L. Orr, Manager CE Core Design C

(FULLY WITHDRAWN) 228 (0.24,228)

I~ ,

b b

~ i

-i BANK C ~ ~ t ~

200 (O.v 1 84)

(o,182) 0 150 i, ~ ~

'8ANK D b

~ b 100 6

O 0 (0,54) .::

I b

~ ~ b so b

b ~

b ~ b ~

P

, ~

(FULLY INSERTED) 0' 0.2 0.4 O.e 0.7 FRACTION OF RATED THERMAL POWER FIGURE 3.1-1 ROD GROUP INSERTION I.IMITS VERSUS THERMAL POWER 3 LOOP OPERATION

(FULLY WITHDRA WN) 228 (0. 66,2 2 8) 200

'- ~ -'."."I B A NK C 82):.

'I ~

~ i ~

R 0

~ ~

'I ~

150 )

i ~

I ~

~ ~

Co 0

O.

L, (O,1 UJ I-Co e BANK D:--

d 0

60 4 ~ ~

(0, 0 o)'.2 0.0 0.4 0.8 0.8 1.0 (FULLY INSERTED) FRACTION OF RATED THERMAL POWER 3.1-2 e FIGURE ROD GROUP INSERTION LIMITS VERSUS THERMAL POWER 4 LOOP OPERATION

Attachment 9 to AEP:NRC:0916W EVALUATION OF PEAKING FACTOR LIMIT REPORT TECHNICAL SPECIFICATION PERFORMED BY WESTINGHOUSE ELECTRIC CORPORATION

Sax 3912 Westinghouse Nuclear Fuel Pinsnurgn Pennsylvania l 5230 3912 Electric Corporation Oivisions January 21, 1987 86AE*~0008 W-AEP/0322 Indiana ~ Zeus Michigan Electric Campany KZZWORDS:

AEP c/o Eric G.

Engineer, Nuclear Materials ~

American Electric Fca~ Service Corporation Fuel Management PEAHING FACZOR REPORT TECH SPEC One Riverside Plaza, 20th Floor Columbus, OH 43215

Dear Mr. Lewis:

AMERICAN ELECIRIC POWER SERVICE COREGRATION D. C. COOK UNIT 1 PEAKING FACIOR ZZKVZ'EPORT American Electric Power (AEP) representatives have asked Westinghouse to support a proposed Tech Spec ch age to reduce the notification time to the NRC for the Peaking Factor Limit Report (PFZR) from the current 60 days to 15 days. AEP plans to submit the Tech Spec change request by February 1, 1987 so that the change will be in place for the upcaning D. C. Cook Unit 1, Cycle 10.

D. C. Cook 1 has been using PQ Surveillance Tech Specs and has been supplying a V(Z) based PFZR to the NRC for the last cycles. Due to the significantly reduced reload autage time and the cycle design time continually beirut reduced closer to the aperating cycle shutdown, Westinghouse recammencis and supports AEP's decision to change their Tech Specs on PFZR notification time fram 60 to 15 days.

A reduction to 15 days before planned criticality would enable AEP to submit the peaking factor report af~ the previous cycle shutdown.

Attached is the information requested by AEP.

Very truly yours, Project Engineer, NFD Projects NEC:mid Attachment cc. M. P. Alexich J. M. Cleveland D. H. Malin w/Enclosure~

V. D. Varxterburg

WESTINGHOUSE PROPRIETARY CLASS 2 PEAKING FACTOR LIMIT REPORT FOR D. C. COOK UNIT 1, CYCLE 9 F SURVEILLANCE EXAMPLE This Peaking Factor Limit Report is provided in accordance with Paragraph 6.9.1.11 of the D. C. Cook Unit 1 Technical Specifications.

D. C. Cook Unit 1, Cycle 9 evaluation dependent V(Z) values as a function of burnup are shown in the attached table. This information is sufficient to determine V(Z) versus core height for Cycle 9 burnups in the range of 0 MWD/MTU to 15,750 MWD/MTU through the use of interpolation.

The V(Z) function is used to confirm that the heat flux hot channel factor, F (Z), will be limited to the Technical Specification values of:

F LIMIT Q

F (Z) tK(Z)j for P > 0.50 and P

. LIMIT 2 10 Q

F LIMIT Q

F (Z) (K(Z)j for P ( 0.50 0.50 The appropriate elevation dependent V(Z) values, when applied to a power distribution measured under equilibrium conditions, demonstrates that the initial conditions assumed in the LOCA are met, along with the ECCS acceptance criteria 'of 10CFR50.46.

(1) WCAP-10216-P-A, Relaxation of Constant Axial Control - F Surveillance Technical Specification

VIEStINGIIOUSE PIIOPIIIEtARY CLASS 2 AEP CYCLE 9 - BURNUP DEPENOENT V(Z) FUNCTION CORE BURNUP (IOe/MTU)

HEIGHT (FEET) 150 12000

0. 13 1.07979 1.07220 '1.0762S ~ 1.08443 ~ I. 09011 ~ 1. 05319 1.09486 'I . 09$ 26 1. 118$ 6 0.3$ $ .08002 1.07737 OS 16O ~ $ .09051 ~ 1.09S63 ~ 1.06100 I. 10056 1. 10399 1. 12527

~ 0.63 1.07827 1.07S03 $ .08625 1.09345 '

1.09901 ~ 1.06681 1. 10338 1. 10757 I. 12799 0.8$ 1.08440 1.0$ 530 1.09036 $ .09885 1. 10261 ~ 1.07377 1. 106S2 I I 1028 1. 13074

t. 13 1.$ 0022 1.08967 1,09323 1. 10229 1. 10739 1.08256 1. 11090

~

1. 11557 1. 13450 t.3$ $ .$ 3063 1.09$ 10 1.09957 1. 10243 1. 10481 1.08421 'I . 10707 1. $ 1171 1. 12936 1.63 I. 13470 1.09424 $ .09554 1.09767 I . 10094 1.08698 1. 10637 1. 11232 1. 12860 1.88 1. 11401 1.09111 1.09049 1.09231 1. 09619 1.08652 I. 10424 $ . 10893 1. t2431

2. 13 1.08956 1.08421 $ .08730 1.08867 1.09155 1.08713 I . 10056 1. 10589 $ . 1933 2.38 1.07877 $ .08046 1.08457 1.08619 I 08772 1.08787 1.09826 1. 10231 1.

14'l4 2.63 I.OBO44 $ .07984 1.08303 1.08314

~

1.08537 1.08818 1.09525 1.09969 1.10890 2.88 1. 08 132 1.08018 $ .08084 1.08021 1.08245 1.08722 1.09174 '1.09596 1. 10244

3. 13 1.07924 1.07847 1.07844 1.07770 1.07914 1.08773 1.08747 I . 09 176 1.09597 3.3S 1.07877 1.07656 1.07708 '1.07522 1.07642 'I . 088 19 1.08528 $ .08962 '1.09602 3.63 1.07850 1.07504 1.07508 1.07394 1.07439 1.09048 1.08454 1.09123 .09935 3.$ S 1.07830 1.07519 'I . 07381 $ .07358 1.07518 1.09495 1.08580 1.09288

1. 10595

4. 14 1.07765 $ .07489 1.07376 1.07484 'I . 07510 1.09823 1.08687 1.09530 1. 11256 4.39 $ .07722 1.074 15 1.07354 1.07490 1.07543 1. 10072 1.08758 1.09632 1. 11816 4.64 1.07568 1.07443 1.07244 1.07511 1.07562 1. 10253 1.08780 'I .097 14 1. 12295 4.89 1.07358 1.0728S 1.07185 1.07464 1.07467 1. 10389 1.08746 1.09641 $ . 126 tO

5. 14 1.07304 1.0723t 1.07024 '1.07344 I . 07391 1. 10369 I'.08650 I . 09740 t. 12813 5.39 1.07221 1.07135 1.06918 1.07222 1.0719$ 1. 10244 1.08551 1.09943 1. 12S5$

5.64 1.0702$ 1.06940 $ .06725 1.07030 $ .07025 I. 10007 1.08642 I . 10055 1. 12663 5.89 1.06740 1.06717 1.06504 1.06825 1.06732 1.09732 1.08666 1. 10075 1. 12282

6. 14 1.06493 1.06443 1.06184 1.06663 1.06706 1.09717 1.08573 1.09877 1. 11739 6.39 '1.06146 '1.06073 1.05841 $ .06487 1.06616 1.09505 I . 08319 1.09575 1. 11020 6.64 1.05723 1.05735 1.05663 1.06297 1.06465 1.09307 1.07986 1. 091 12 1. 10123 6.89 1.0534 I 1.05565 1.05391 1.05957 1.06122 1.08909 1.07443 1.08490 1.091$ 0

7. 14 1.05131 1.05294 1.05226 1.05803 1.05906 1.08307 1.07101 1.07848 1.08908 7.39 1.05023 $ .05265 I . 05105 1.05640 1.05694 1.07651 1.06662 1.07260 1.08934 7.64 1.04917 1.05009 1.04964 $ .05331 1.05489 1.06776 1.06153 1.06590 I.oas44 7.89 1.04823 1.04749 1.04678 $ .04944 1.05022 1.06245 1.05434 1.05777 1.08569 8.15 1.05234 1.04417 1.04321 1.04497 1.0450$ 1. 05514 1.04710 1.04$ 09 t.os ta2 8.40 1.05543 1.03884 $ .03830 1.03928 '1.03877 1.04594 1.03860 1.03772 $ .07687 8.65 1.05836 $ .03231 1.032SB 1.03110 1.03084 1.03707 $ .02743 1. 0294 I 1.07163 8.90 1.06137 1.03272 1.02908 1.02758 1.02774 1.03419 1.02824 1.03243 1.07815

9. 15 t.06465 1.03439 1.03057 1.03034 1.03104 $ .03471 1.03191 1.03460 I.OS413 9.40 1.06745 '1.03509 $ .03230 1.03308 1.03465 1.03530 1.03479 1.03716 1.09017 9.65 1.06490 .t.03533 1.03262 $ .03530 1. 03S 12 1.0356$ 1.03670 1.03$ 66 1.09535

9. 90 1.06032 '1.0294$ 1.03376 1.03785 1.04 153 1.03553 1.03870 1.03985 1. $ 0024

10. 15 1.05996 1.02919 $ .03096 1.03935 1.04456 1.03392 1.03999 1.04067 1. 10413

'Io. 40 1.05964 1.02890 '!.02577 1.03866 1.0451$ 1.03t96 $ .04054 1.0412$ 1. 10616 10.65 1. 06319 I . 03192 1.03287 1.04450 $ .051 $ 5 1.03232 1.04540 1.04529 1. I $ 194 IQ. 00 I . 06907 1.04168 $ .04783 ~ 1.05132 ~ 1.05735 ~ 1.03384 1.05011 1.047$ I 1. $ 15$ 8 g ~

1.07523 $ .04480 1.04692 ~ 1.06071 '

I . 07017 ~ 1.04482 1.06676 1.06543 1. $ 36$ 7

~

AC) 1.07097 I 04506 $ .05253 ~ 1.07060 ~ $ .08402 ~ 1.05712 0 1.08509 '1.08299 1. 15243

~

t ~ .r 1.08375

~

1.05781 1.06680 1.08773 ~ 1. $ 0107 ~ 1.07185 '

t. 10264 1. 10213 1. 17076 I 1.90 I . 08491 1.05867 1.06868 ~ 1.09092 '

I . 10222 ~ 1.06720 I $ 0076 1.09762 1. $ 6193 TOP BOTTOM $ 0$ ( EXCLUDEO AS PER TECHNICAL SPECIFICATION 4 .2.2.2

Attachment 10 to AEP:NRC:0916W EVALUATION OF PROPOSED MTC LIMIT TECHNICAL SPECIFICATION AND SAFETY LIMIT CURVE FOR 2250 PSIA FOR UNIT 2 SUPPLIED BY EXXON NUCLEAR COMPANY, INC.

SUPPLEMENT TO EVALUATION OF PROPOSED MTC LIMIT TECHNICAL SPECIFICATION

AUG 0 4 1986.

E+ON NUCLEAR COMPANY, INC 800 I 08TH AVENUENE. PO BOX 90777. BELLEVUE.WA98009 g08) 453-4300 July 31, 1986 ENC-AEP/0511 Mr. D. H. Malin, Sr. Engineer Nuclear Material & Fuel Management Indiana 8 Michigan Electric Company c/o American Electric Power Service Corp.

One Riverside Plaza Columbus, OH 43216-6631

Subject:

Technical S ecification Chan es to the MTC Limit and Safet Limit Curves Ref.: (I) Letter, Douglas H. Malin (AEP) to H. G. Shaw (ENC), "D. C. Cook Unit 2, Cycle 6 Required Exxon Fuel Support Activities," dated May 29, 1986 (AEP-ENC/0231)

(2) XN-NF-85-64(P), Revision I', Suppiement I, "Plant Transient Analysis for D. C. Cook Unit 2 with 10 Percent Steam Generator Tube Plugging,"

Exxon Nuclear Company, Inc., March 1986.

Dear Doug:

Items 10 and I I of Reference I requested that Exxon Nuclear provide support for Technical Specification (T.S.) changes for< D. C. Cook Unit 2 for both the moderator temperature coefficient (MTC) limit and the safety limit curves. The current analyses supporting D. C; Cook Unit 2 Cycle 6 operation, presented in Reference 2, have been reviewed with respect to supporting these changes.

Moderator Tem erature Coefficient Limit The current T.S. gives a MTC limit of +5 pcm/F for all powers less than 70 percent of rated thermal power (RTP) and a limit of 0 pcm/F for all powers of 70 percent or greater. Item 10 of Reference I indicates that fuel management flexibility can be gained by replacing the step change in the MTC limit at 70 percent of RTP with a linear ramp rate from +5 pcm/F at 70 percent RTP to 0 pcm/F at 100 percent RTP. Review of the analyses presented in Reference 2 indicates that five transients were performed with a positive MTC at power levels that would potentially be affected by this T.S. change. These five transients are:

15.2.1 Loss of External Load 15.3. I Loss of Primary Coolant Flow 15.3.3 Locked Primary Pump Rotor I 5.4.2 Uncontrolled Rod Withdrawal at Power 15.4.3 Single RCCA Withdrawal AN AFFII.IATEOF EXXON CORPORATION

D. H. Malin

~ ~ July 31, l986 Review of the first three transients (I5.2.I, I5.3.I, and I5.3.3) indicated that they had all been performed at I00 percent of RTP with a conservatively high positive MTC value. The review of the I5.4.2 transient analyses showed that the event had been analyzed at three power levels: 9, 60, and l00 percent of RTP. Here, again, the l00 percent RTP case had been performed with a conservatively high positive MTC value consistent with the first four transients. The 9 and 60 percent RTP cases were found to have been performed with temperature-dependent MTC curves, as shown in Reference 2. Both of these cases, however, were adjusted to an initial MTC nominal value of +5 pcm/F for the thermal hydraulic conditions at the start of the transient calculations. These MTC temperature-dependent curves were then conservatively biased for the actual transient calculations.

Review of the fifth transient, I5.4.3, indicated that -it had been performed as a bounding analysis of the results obtained in the 15.4.2 transient analyses accounting for the increase in the augmentation factor for a single rod withdrawal. Thus, it supports the same MTC values that are supported by the event I5.4.2.

From the above review, it is apparent that conservatively high positive MTC values have been used in all the transients where it is conservative 'to do so. Since the positive MTC values used in these analyses either support or exceed the value at the respective power level in the proposed T.S. change,.it is concluded that the analyses presented in Reference 2 will support the proposed T.S. change.

Safety Limit Line at 2250 sia Reference 2 and the current T.S. have safety limit lines (SLL) at pressures of I840, 1940, 2040, 2290, and 2440 psia. Item I I of Reference I indicates that a SLL is desired at the nominal D. C. Cook Unit 2 operating pressure of 2250 psia. A SLL at 2250 psia has, consequently, been conservatively interpolated from the data that was used in generating the SLLs reported in both Reference 2 and the current T.S.

This SLL is shown in the attached figure, and the points defining the SLL are given on the figure.

If you have any questions concerning this analysis, please feel free to contact our Jerry Holm at (509) 375-8 I 42, Since ely, l

C i H. G. Shaw Contract Administrator HGS/wjj xc: MP Alexich JM Cleveland V Vanderburg

670 UNACCEPTABLE OPERATION 650 0 pg1 2290 ppsst w 630 22~0 C3 p~

Sled 2040 ps

1940 610 PS1g CL 1840 I

I 590 ACCEPTABLE OPERATION 570 550

0. 0 0. 2 0. 4 0. 6 0. 8 1. 0 1. 2 FRRCTIUN QF RATED THERHRL PCLNER PRESSURE (PSIA) BREAKPOINTS (FRACTION RATEO THERMAL POWER, T AVG, DEG F) 1840 1940 (0.00,616.2)

(0.00,623.8)

(0.98,585 (0.93,594.7)

') (1.20,556.5)

(1.20,563.5) 2040 (0.00,631.0) (0.88,603.8) (1.20,569.6) 2250 (0.00,645.9) (0.80,622.3) (1.20,580.9) 2290 (0.00,647.9) (0.80,624.5) (1.20,586.5) 2440 (0.00,657.4) (0.77,635.6) (1.20,597.2)

jl,f~~

q 1987 ADVANCEDNUCLEAR FUELS CORPORATION 800:08th AVENUE NE PO 8OX 90777. 8ELLEVUE WA 983094777 908> 493 4300 March 5, 1987 ANF-AEP/0557 Mr. Richard B. Bennett, Engineer Nuclear Materials & Fuel Management Indiana 6 Michigan Electric Company Ref.:

c/o American Electric Power Service Corp.

One Riverside Plaza, 20th Floor Columbus, OH 43216-6631 (1) Letter, H.G. Shaw (ANF) to D;H. Malin (AEP), "Technical Specification Changes to the MTC Limit and Safety Limit Curves," dated July 31, 1986 (ENC-AEP/0511)

(2) XN-NF-85-64, Rev. 1, "Plant Transient Analysis for D.C. Cook Unit 2 with 10% Steam Generator Tube Plugging," Exxon Nuclear Company, March 1986 (3) XN-NF-85-64, Rev. 2, Supp. 1, "Plant Transient Analysis for D.C. Cook Unit 2 with 10% Steam Generator Tube Plugging," Exxon Nuclear Company, September 1986 (4) Letter, G.N. Ward (ANF) to H.R. Denton (NRC), "Response to NRC Questions on XN-NF-85-28(P)," dated April 14, 1986 (GNW:053:86)

Dear Mr. Bennett:

This letter is in response to your request in a telephone conversation with Jerry Holm on February 26, 1987 for an additional evaluation of the proposed Technical Specification (T.S.) change in the D.C. Cook Unit 2 moderator temperature coefficient (MTC). Specifically, an evaluation of the T.S. change on events 15.1.1 Decrease in Feedwater Temperature 15.1.2 Increase in Feedwater Flow 15.1.5 Steam Line Break 15.4.1 RCCA Withdrawal from Subcritical 15.4.6 Boron Dilution 15.4.8 RCCA Ejection was requested. The, initial evaluation of the proposed T.S. change was reported in Reference 1 and the evaluation of these additional events is presented in the following paragraphs.

The proposed T.S. change in the MTC involves the replacement of a step change in MTC at 70% rated thermal power (RTP) from +5 pcm/'F to 50 pcm/'F for all

' 'FtLfALECF KRASTWERK UH(OH Q~ KkVV

Mr. R. Bennett (AEP) March 5, 1987 RTP greater than 70% to a ramp change from +5 pcm/'F at 708 RTP to XO pcm/'F at 100% RTP. This proposed T.S. change will allow a positive MTC over the power range from 70% to 100% of RTP, whereas only a 0 or negative MTC was allowed before. A positive MTC is only a concern for heatup events since for these events the potential for an increase in power is aggravated by a positive reactivity contribution from the MTC. Events 15.1.1, 15.1.2 and 15.1.5 are all cooldown events, and are consequently limiting only for negative MTCs. Therefore,, these three events are unaffected by the T.S.

change and will continue to be bounded by the current analysis presented in Reference 2.

The event 15.4.1, RCCA Withdrawal from Subcritical or Low Power, is not affected by the proposed T.S. change. The limiting case is initiated from a low initial power level (approximately 1.0E-9 RTP), which bounds the hot shutdown and startup modes of operation. This low initial power level yields the maximum margin to trip, and hence the maximum time for rod withdrawal.

These two conditions produce the largest prompt multiplication which maximizes the power overshoot past trip. Since the proposed T.S. change only affects operation at or above 70% RTP, the limiting event presented in Reference 2 remains bounding.

The Boron Dilution event (15.4.6) was evaluated for the full range of operating modes, that is, for all modes from 1 to 6. Modes 2 all restricted to power levels less than 5% of RTP, and are through 6 are consequently unaffected by the proposed T.S. change. Mode 1, which is power operation with powers greater than 5% of RTP, is bounded by Event 15.4.2, RCCA Withdrawal at Power, which was addressed in Reference 1. It is bounded by Event 15.4.2 from a DNB standpoint because the reactivity insertion rates considered in 15.4.2 bound the maximum rate achievable by boron dilution. The time to lose shutdown margin in Mode 1 is unaffected by the T.S. change since it is only a function of the shutdown margin, primary coolant system volume, and the maximum boron dilution rate. Since none of these parameters are altered by the T.S. change, the analysis presented in Reference 3 remains bounding.

The limiting RCCS Ejection event (15.4.8) was found to occur at end of cycle (EOC) from HFP conditions. The EOC conditions were found to be limiting over the BOC conditions due to a larger rod- worth and a smaller delayed neutron fraction at EOC. Both these conditions result in an increase in the calculated return to power for the event. The proposed T.S. change will not affect the results of the EOC analysis from HFP conditions because the MTC is negative at EOC. Furthermore, the MTC has only a small effect on the results of this e'vent because the extremely rapid nature of the event does not allow sufficient time for the heat to be transferred from the fuel. Thus, the current analysis for this event presented in Reference 4 will not be altered by the proposed T.S. change and will continue to bound current operating conditions.

Mr. R. Bennett (AEP) March 5, 1987 If you have any further questions regarding this MTC Technical Specification review, please feel free to contact our Mr. Jerry Holm (509-375-8142).

Sincerely, H. G. Shaw Contract Administrator gf cc: Mr. J.M. Cleveland Mr. D.H. Malin Mr. V. VanderBurg Mr. J.S. Holm (ANF)

Attachment 11 to AEP:NRC:0916W EVALUATION OF THE IMPACT OF 2000 GPM PRIMARY FLOW ON THE UNIT 2 DILUTION TRANSIENT PERFORMED BY EXXON NUCLEAR COMPANY, INC.

i JUL i.v t;o6 C/PN NUCLEAR COMPANY, INC.

6 00 1067HAVENUENE.POBOX90777BELLEVVE'+496009 July 11, 1986 (206) 453-4300 ENC-AEP/0505 Mr. D. H. Malin, Sr. Engineer Nuclear Material 8 Fuel Management Indiana un Michigan Electric Company c/o American Electric Power Service Corp.

One Riverside Plaza Columbus, OH 43216-6631

Subject:

Boron Dilution Analysis During RHR Operation for D.C. Cook Unit 2 Ref.: (1) Letter, D. H. Malin (AEP) to H. G. Shaw (AEP), "D.C. Cook Unit 2, Cycle 6 Required Exxon Fuel Support Activities,"

dated May 29, 1986 (AEP-ENC/0231)

(2) XN-NF-85-64(P), Rev. 1, Supp, 1, "Plant Transient Analysis for D.C. Cook Unit 2 with 10% Steam Generator Tube Plugging," Exxon Nuclear Company, Inc., March 1986

Dear Doug:

Item 6 o f Reference 1 requested that Exxon Nuclear perform a boron dilution analysis to support operation of D.C. Cook Unit 2,with a residual heat removal (RHR) system flow rate of 2000 gpm. The analyses presented in Reference 2 wer e performed to support an RHR flow rate of 3000 gpm, which is the minimum flow rate specified in the D.C. Cook Unit 2 Technical Specifications.

The RHR analyses described in Reference 2 were performed using a dilution front method since the RHR flow rate is potentially insufficient to assure a completely mixed primary coolant volume. This dilution front method assumes a step boron concentration reduction at the charging inlet which migrates through the core and the remainder of the non-stagnant primary coolant and RHR system. When this dilution front completes one transit time, the entire volume of the non-stagnant coolant system is at the reduced boron concentration and a second step reduction begins to transit the system.

A detailed review of the calculations which have been performed indicates that the analysis presented in Reference 2 will bound RHR flow rates 2000 gpm or greater. The RHR flow rate is not specified in Reference 2. A revision to this report will be issued which specifies the minimum flow rate.

6 'F Ex'k0v CQ4POsariQw

Hr. D. H. Malin (AEP) July 11, 1986 If you have any questions concerning this analysis, please feel free to contact our Mr. Jerry Holm, telephone 509-375-8142.

Sincerely, H. G. Shaw Contract Administrator gf C

CC: Hr. M. P. Alexich Hr. J. H. Cleveland Hr. V. Vanderburg

Attachment 12 to AEP:NRC:0916W PRESSURIZER PRESSURE READABILITY ALLOWANCE AND RCS FLOW MEASUREMENT ALLOWANCE FOR UNIT 2 PREVIOUSLY SUBMITTED WITH AEP:NRC:0916I

Attachment 7 is provided as an aid to assist the reviewer in understanding the development of certain values cited in the Technical Specifications. The included calculations supplement information provided in XN-NF-85-64(P), XN-NF-85-64(P) Rev. 1, and WCAP 11080. Reference to is indicated in the Remarks column of Attachment 10 for those Technical Specification items which require the additional explanation so provided.

Item A of this attachment demonstrates the development of the Reactor Coolant System (RCS) Analysis Flow Value.

Item B of this attachment demonstrates the derivation of the required minimum indicated RCS Flow in ibm/hr.

Item C of this attachment demonstrates the conversion of the minimum indicated RCS flow obtained in 'Item B from ibm/hr to gpm.

Item Dl provides the minimum indicated pressurizer pressure indication value in psig for Mode 1 operation.

Item D2 provides the minimum indicated pressurizer pressure indication value in psig for Modes 2 & 3 operation.

ATTACHMENT 7 A. ANALYSIS VALUE OF REACTOR COOLANT SYSTEM (RCS) FLOW Nominal RCS Flow with 10% Steam Generator 6

Tube Plugging 141.3 X 10 ibm/hr 6

Flow Measurement, Uncertainty (3.5%) 5.0 X 10 ibm/hr Flow Measurement Repeatability 3.4 X 10,6 ibm/hr Analysis Flow:

141.3 E6 - 5.0 E6 - 3.4 E6 132.9 X 10 6 ibm/hr B. TECHNICAL SPECIFICATION MINIMUM INDICATED REACTOR COOLANT SYSTEM (RCS)

FLOW (ibm/hr)

Nominal RCS Flow with 10% Steam Generator Tube Plugging 141.3 X 10 6 ibm/hr 6

Flow Measurement Repeatability 3.4 X 10 ibm/hr Correction to Flow Measurement Repeatability to Support Larger Pressure Allowance 6 (Section 15.0.2, XN-NF-85-64(P) Rev. 1) 0.7 X 10 ibm/hr Modified Flow Measurement Repeatability:

3.4 E6 - 0.7 E6 2.7 X 10 6 ibm/hr Minimum Indicated RCS Flow:

141.3 E6 - 2.7 E6- 138.6 X 10 6 ibm/hr C. TECHNICAL SPECIFICATION MINIMUM INDICATED REACTOR COOLANT SYSTEM (RCS)

FLOW (gpm)

Minimum Indicated RCS" Flow 138.6 X 10 6 ibm/hr RCS Pressure 2250 psia T

cold 542.3 F 1 Gallon 0.13368 cu.ft.

Specific Volume of Water at Stated Pressure and Temperataure Conditions (1967 ASME Steam Tables) 0.021119 ft3 /ibm RCS Flow - (13).6 E6 ibm/hr) X (1 hr/60 min)

X (0.021119 ft /ibm) X (1 gal/0.13368 ft.)

Minimum Indicated RCS Flow- 364,940 gpm Minimum Indicated RCS Flow/Loop 91,240 gpm

D. INDICATED PRESSURIZER PRESSURE DNB LIMIT (TABLES 3.2-1 AND 3.2-2)

The method of determining the allowance for pressure readability is similar to that provided in WCAP 11080 for the indicated T Actual values for the terms used in the calculation, with the ave exception of the rack calibration allowance and the indicator readability, were also obtained from WCAP 11080 Page viii. The value used for the rack calibration allowance was obtained from the pressurizer pressure channel calibration procedure; the value used for indicator readability was determined from a review of the indicator span and scale.

The total pressurizer pressure channel allowance was determined to be 3.41% of span which equates to 27.29 psia.

Assuming a minimum of 3 channels available for averaging, the allowance may be reduced by the square root of 3. This yields a final pressurizer pressure readability allowance of 15.8 psia.

1) Minimum Indicated Pressure in Mode 1 Nominal Pressure 2250 psia Pressure Control Allowance (WCAP 11080, Page 3) Proprietary Indication Allowance 15 ' psi Allowance assumed in Analysis- 40 psi Additional Pressure Allowance accounted for by .5% increase in minimum RCS Flow (Section 15.0.2 XN-NF-85-64(P) Rev. 1) 7.5 psi Analysis Pressure:

2250 7.5 2202.5 psia Minimum Indicated Pressurizer Pressure:

2202.5 + 15.8- 2218.3 psia Table 3.2-1 Value for Minimum'ndicated Pressure in Mode 1:

2220 psia- 2205 psi.g

2) Minimum Indicated Pressure in Modes 2 & 3 Analysis Pressure 2175 psia Minimum Indicated Pressure:

2175 + 15.8 << 2190.8 psia Table 3.2-2 Value for Minimum Indicated Pressure in Modes 2 & 3:

2191 psia- 2176 psig

Attachment 13 to AEP:NRC:0916W

SUMMARY

TO ATTACHMENT 13 ANALYSIS TO JUSTIFY AN INCREASE IN BORON CONCENTRATION IN THE REFUELING WATER STORAGE TANKS AND ACCUMULATOR TANKS PERFORMED BY WESTINGHOUSE ELECTRIC CORPORATION EVALUATIONS OF THE IMPACT ON AN INCREASE IN BORON CONCENTRATION ON THE UNIT 2 ANALYSES PERFORMED BY EXXON NUCLEAR CORPORATION AND BY AMERICAN ELECTRIC POWER SERVICE CORPORATION

SUMMARY

OF ATTACHMENT 13 The purpose of Attachment 13 is to provide justification for increasing the minimum boron concentration in the RWST and accumulators for D. C. Cook Units 1 and 2. The minimum boron concentration is being increased to 2400 ppm to provide fuel management flexibility. The changes impact T/Ss 3.1.2.7 (Borated Water Sources - Shutdown), 3.1.2.8 (Borated Water Sources Operating), 3.5.1 (Accumulators), and 3.5.5 (Refueling Water Storage Tank).

Included in this attachment is a safety evaluation performed by Westinghouse in support of this change. The Westinghouse evaluation considers the impact that raising the minimum boron concentration has on the LOCA and non-LOCA safety analyses, as well as LOCA related design considerations. The Westinghouse discussion of LOCA related design considerations references WCAP 11020, entitled "Spray Additive Tank Deletion Analysis for the Donald C. Cook Nuclear Plant". This analysis was submitted to the NRC via our letter AEP:NRC:0914C, dated February 28, 1986, in support of our proposal to remove the NaOH spray additive tank and its associated T/S (T/S 3/4.6.2.2). Although this submittal is still under NRC review, reference was made to it since operation without spray additive is bounding with respect to those issues considered in the LOCA related design considerations section of the Westinghouse evaluation contained in this attachment.

The Westinghouse evaluation also contains a discussion of post-LOCA long term core cooling. This discussion demonstrates that for D. C. Cook Unit 1 during the present and upcoming fuel cycles, the boron concentration in the sump following a LOCA would be sufficient to maintain the reactor subcritical.

Analogous evaluations for the present Unit 2 fuel cycle were performed by the Americ'an Electric Power Service Corporation, using methodology similar to that described by Westinghouse. These calculations are described in our letter AEP:NRC:1008, which was submitted to the NRC on November 17, 1986.

Related to the change in RWST and accumulator boron concentrations are changes to the boric acid storage tank and RWST volumes required by T/Ss 3.1.2.7 and 3.1.2.8. These changes are described by Westinghouse in 3. The changes to the required tank volumes ensure the capability to bring the core from hot, full power to Mode 4 and 6 shutdown conditions, including allowing for the increased shutdown margin requirements based on the boron dilution event. (Reference our proposed Unit 1 T/Ss 3/4.1.1.2).

Similar changes were made for Unit 2 and approved in Amendment 82 to DPR-74.

Lastly, Attachment 13 contains' letter from Advanced Nuc'lear Fuels Corporation (ANF, formerly Exxon Nuclear Co.). This letter documents ANF's concurrence with the increase in the RWST and accumulator boron concentration.

SAFETY EVALUATION FOR INCREASE IN THE BORON CONCENTRATION LIMITS FOR THE RWST AND ACCUMULATOR LIMITS FOR D. C. COOK UNITS 1 AND 2

1.0 INTRODUCTION

It must be demonstrated, each cycle, that the core can be maintained subcritical via boron addition from the ECCS. in the unlikely event of a Large Break LOCA. However, evaluations of future fuel cycles show that subcriticality may not be assured with the present minimum RHST/Accumulator boron concentration. In order to provide adequate post-LOCA shutdown =margin for future cycles, increasing the accumulator and RHST boron concentration into the range of 2600 ppm is proposed.

2.0 SCOPE OF EVALUATION Both Mestinghouse Electric Corporation and American Electric Power Service Corporation (AEPSC) have assessed the impact of increasing the RMST and accumulator boron concentration from a minimum of 1950 ppm into the range of 2600 ppm. This assessment identified the following areas in which the boron concentration increase must be shown to have a favorable or non-detrimental impact on the D. C. Cook design basis:

1. Non-LOCA Safety Analysis

2. LOCA Analysis (10 CFR50.46) o Small Breaks o Large Breaks o Long-Term Core Cooling o Boron Precipitation 029lv:1D/022787

3. LOCA Related Design Consideration o Radiological Consequences o Hydrogen Production o Equipment gualifications Evaluation summaries for each of the above areas are provided in the following section.

3.0 SAFETY EVALUATION 3.1 FSAR NON-LOCA SAFETY ANALYSIS The proposed increase in RWST boron concentration has been evaluated and the impact of this change on each of the non-LOCA FSAR transients which model the RWST and/or accumulators follows.

3. 1.1 Uncontrolled Boron Dilution The refueling and startup cases are impacted by the RWST boron concentration change. The increased concentration increases the time to reach criticality which increases the available operator action time. This is a benefit in the analysis.

3. 1.2 Ma'or Secondar S stem Pi e Ru ture

a. Ru ture of a Main Steamline Core Res onse and Mass/Ener Release Inside Containment - The current safety analyses for Units 1 and 2 assumes that boron concentration of 20,000 ppm in the Boron Injection Tank (BIT) would be available to provide negative reactivity to shut down the reactor, Although an increase in boron concentration in the RWST and accumulators would generally be a benefit for this transient, the impact would be negligible when compared with the available BIT boron concentration .

(20,000 ppm) which would be purged before the RWST water reaches the core. As such, the current safety analyses provided in Chapter 14 of the FSAR for the core response and the mass and energy release inside containment rema'in valid.

0291 v:1D/022787

b. Ru ture of a Main Steamline Mass/Ener Release Outside Containment - The recent outside containment mass/energy release data following a steamline break provided in WCAP-10961 Revision 1 (Steamline Break Mass/Energy Releases for Equipment Qualification Outside Containment) assumed a BIT boron concentration of 0 ppm to bound the other similar Westinghouse units. .An increase in the minimum boron concentration in the RWST and accumulators would be a benefit for this transient because it would shut down the reactor sooner. The boron concentration increases would give less limiting results for the mass/energy releases outside containment provided in WCAP-10961.

3.1.3 Accidental De ressurization of the Main Steam S stem The current safety analyses for Units 1 and 2 assumes a boron concentration of 20,000 ppm in the Boron Injection Tank (BIT) would be available to provide negative reactivity to shut down the reactor. Although an'ncrease in boron concentration in the RWST and accumulators would generally be a benefit for this transient, the impact would be negligible when compared with the available BIT boron concentration (20,000 ppm) which would be purged before the RWST water reaches the core. As such, the current safety analysis provided in Chapter 14 of the FSAR remain valid.

3. 1. 4 Conc 1 us i ons The above discussions demonstrate that the proposed RWST and accumulators boron concentration increase does not adversely impact the conclusions of non-LOCA transient analyses. Accident reanalysis is not required, therefore there are no FSAR changes associated with this evaluation.

3.2 FSAR LOCA ANALYSIS The following evaluation discusses the impact of the increase from 1950 ppm to 2400 ppm in RWST/Accumulator boron concentrations for D. C. Cook Units 1 and 2 on the Large and Small break LOCA analyses, Long Term Core Cooling and Hot Leg Switchover Time. The time when hot leg recirculation should be initiated to 0291 v:1O/022787 3

prevent boron precipitation in the core was determined to be 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months following' LOCA. FSAR section 6.3 was revised to reflect the hot leg switchover time for both units.

3.2. 1 Hot Le Recirculation Switchover Time An analysis has been performed to determine the maximum boron concentration in the reactor vessel following a hypothetical LOCA. This analysis considered D.

C. Cook Units 1 and 2 with a proposed maximum boric acid concentration of 2600 ppm in the RHST, accumulators, and RCS..

The analysis considers the increase in boric acid concentration in the reactor vessel during the long term cooling phase of a LOCA, assuming a conservatively small effective vessel volume. This volume includes only the free volumes of the reactor core and upper plenum below the bottom of the hot leg nozzles.

This assumption conservatively neglects the mixing of boric acid solution with directly connected volumes, such as the reactor vessel lower plenum. The calculation of boric acid concentration in the reactor vessel considers a cold leg break of the reactor coolant system in which steam is generated in the core from decay heat while the boron associated with the boric acid solution is completely separated from the'steam and remains in the effective vessel volume.

The results of the analysis show that the maximum allowable boric acid concentration of 23.53 weight percent established by the NRC, which is the boric acid solubility limit less 4 weight percent, will not be exceeded in the vessel if hot leg injection is initiated 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the LOCA inception.

This switchover time is applicable to both units. The operator should reference this switchover time against the reactor trip/SI actuation signal.

The typical time interval between the accident inception and the reactor trip/SI actuation signal is negligible when compared to the switchover time.

Procedures philosophy assumes that it would be very difficult for the operator to differentiate between break sizes and locations. Therefore one hot leg switchover time is used to cover the complete break spectrum.

0291v:1D/022787

3.2.2 Small Break LOCA D. C. Cook Unit 2 The current FSAR small break analysis for D. C. Cook Unit No. 2 employs the Westinghouse WFLASH Evaluation Model and is based on a full core of Westinghouse fuel. Since the time that the FSAR small break LOCA analysis for D. C. Cook Unit No. 2 was performed, the Westinghouse fuel has been almost completely replaced with fuel provided by the'xxon Nuclear Corporation (ENC). The Peak Clad Temperature results of small break LOCA analyses employing this Evaluation Model will not be altered by the changes in boron concentrations for the RWST and accumulators. Confirmation of the applicability of the FSAR small break LOCA analysis will be required by the current fuel vendor.

3.2.3 Small Break LOCA D. C. Cook Unit 1

-Small break LOCA analyses performed by Westinghouse assume that the reactor core is brought to a subcritical condition by the trip reactivity of the control rods. There is no assumption requiring the presence of boron in the ECCS water or needing the negative reactivity provided by the soluble boron.

Thus the changes to the, RWST and Accumulator Tech-Specs covering boron concentrations do not alter the conclusions of the FSAR small break LOCA analysis.

3.2.4 Lar e Break LOCA D. C. Cook Unit No. 1 Large break LOCA analyses performed by Westinghouse do not take credit for the negative reactivity introduced by the soluble boron in the ECCS water in determining reactor power level during the early phases of the hypothetical large break LOCA, The large break LOCA analyses performed by Westinghouse analyze the LOCA transient to a time just beyond the time at which Peak Cladding Temperature is calculated to occur. During this time period the reactor is kept subcritical by the voids present in the core. Thus the changes to the RWST and Accumulator Tech-Specs covering .boron concentrations do not alter the conclusions of the FSAR large break LOCA analyses.

0291 v:10/022787

3.2.5 Lar e Break LOCA D. C. Cook Unit No. 2 It is the responsibility of the current fuel vendor to address the impacts that the proposed Tech-Spec changes may have on the fuel, LOCA model, LOCA methodology, and LOCA assumptions employed for this unit.

3.2.6 Lon Term Core Coolin D. C. Cook Unit No. 1 The Westinghouse licensing position for satisfying the requirements of 10CFR Part 50 Section 50.46 Paragraph (b) Item (5) "Long-term cooling" is defined in.

WCAP-8339 (page 4-22). The Westinghouse commitment is that the reactor remain shutdown by the borated ECCS water. Since credit for the control rods is not taken for large break LOCA, the borated ECCS water provided by the RMST and Accumulators must have a concentration that, when mixed with other sources of water, will result in the reactor core remaining subcritical assuming all control rods out (ARO). The attached figure (Figure 1) shows the effect on the post-LOCA RCS/Sump boron concentration as a result of changing the minimum Tech-Spec boron concentration from 1950 to 2400 for the RMST and from 1950 to 2400 for the Accumulators. The result is an increase of over 200 PPM in the RCS/Sump boron concentration which would provide enough negative reactivity to keep the cycle 9 core subcritical with a margin of about 204 PPM. Thus the long-term core cooling requirement that the reactor remain subcritical is satisfied by the new proposed Technical Specifications for D. C. Cook Unit No.

l. It is here noted that the ability to maintain core subcriticality following a hypothesized LOCA is highly dependent on cycle specific core conditions, and an evaluation of Long Term Core Cooling capability is routinely performed before the start-up of each cycle.

3.2.7 Lon Term Core Coolin D. C. Cook Unit No. 2 The Westinghouse licensing position for satisfying the requirements of 10CFR Part 50 Section 50.46 Paragraph (b) Item (5) "Long-term cooling" is defined in MCAP-8339 (page 4-22). The assumptions employed by Westinghouse to satisfy these requirements have been stated above (LONG TERM CORE COOLING D. C. COOK UNIT NO. 1), The assumptions employed by ENC for the satisfaction of the 0291 v:10/022787

requirements of 10CFR Part 50 Section 50.46 Paragraph (b) Item (5) may differ from those employed by Westinghouse. At the request of American Electric Power, Westinghouse has performed a calculation to determine the minimum Post-LOCA RCS/Sump boron concentration for a range of pre-trip RCS boron concentrations for D. C. Cook Unit No. 2. This calculation is based on the current Westinghouse assumptions and methodology for Westinghouse fuel using the most recent available input sources for D. C. Cook Unit No. 2. The attached figure (Figure 2) shows the minimum post-LOCA RCS/Sump boron concentration as a function of pre-trip RCS boron concentration with a minimum Tech-Spec boron concentration of 2400 PPM for the RWST and 2400 PPM for the Accumulators based on the above-stated assumptions. The adequacy of these limits to ensure core subcriticality following a postulated large break LOCA is. dependent on the limiting RCS boron requirements for criticality ad dictated by the core design for a specific cycle. Confirmation of the applicability of these limits and that Long Term Core Cooling requirements will be satisfied must be provided by American Electric Power.

3.2.8 Conclusions The increase in the minimum RWST boron concentration from 1950 PPM to 2400 PPM and minimum Accumulator boron concentrations from 1950 to 2400 do not negatively affect the FSAR LOCA analysis for D. C. Cook Unit No. 1. The new proposed Technical Specifications provide an additional safety margin to ensure long-term cooling of the reactor core after a postulated large break LOCA for D. C. Cook Unit No. 1.

3.3 LONG TERM SUMP H The minimum calculated pH is 7.6. The assumptions used in calculating this sump pH are as follows:

1. The amount of boric acid that is transported to the sump was maximized.

The volumes of solution that were assumed to enter the sump are as follows:

a. RWST total tank volume as provided in chapter 6 of the FSAR, 0291v:1O/022787

b. maximum SI accumulator water volume as allowed by technical f

speci ications,

c. boron injection tank volume of 900 gallons,

d. RCS volumes and auxiliary piping volumes as indicated by Westinghouse ca 1 cul at i ons.

The boric acid concentration of solutions entering the sump was maximized. The following concentrations were assumed:

a. The maximum allowable RWST concentration was assumed to be 2600 ppm.

b. The accumulator and piping concentration was assumed equal to the RMST concentration.

c. The maximum boron injection tank concentration allowed by technical specifications was used.

d. The RCS concentration was conservatively chosen as 2400 ppm.

2. The amount of sodium tetraborate transported to the sump was minimized by assuming the minimum ice mass and ice pH allowed by the technical specifications.

These assumptions taken in total were aimed at determining a conservative lower bound for the long term sump pH.

3.4 RADIOLOGICAL, HYDROGEN, AND EQUIPMENT QUALIFICATION EVALUATIONS Increasing the boron concentration in the Refueling Mater Storage Tank (RWST) and accumulators decreases the pH of the recirculating core cooling solution.

A'decrease in pH can decrease the elemental iodine decontamination factor (DF), increase the rate of hydrogen production due to corrosion of zinc (galvanize and zinc based paint) and increase the potential for chloride induced stress corrosion cracking of stainless steel.

0291 v:1 0/022787

Based on the above considerations, 2600 ppm has been determined to be the maximum RHST and accumulator boron concentration. Details of the specific evaluations follow.

3.5 RADIOLOGICAL CONSEQUENCES The minimum calculated sump pH of 7.6 is sufficient to support the elemental iodine DF assumed in the Spray Additive Deletion Analysis (reference 1).

Hence, the radiological consequences will not change as a result of the boron increase, and the dose .analysis (reference 1) remains valid.

The reference analysis assumes a DF of,1000 (99.9 percent removal) for the combined elemental iodine. reduction effects of the ice condenser, sprays, surface deposition, and radioactive decay. The sump solution, with a minimum pH of 7.6, can retain approximately 98 percent (reference 2) of the elemental iodine that is assumed to be released from the core. The containment surfaces utilized for deposition have the capacity to retain 100 percent of the released iodine in the short term and greater than 70 percent in the long term. Hence, the DF assumption of the reference analysis, for the combined long-term iodine capacity of sump and surfaces, remains valid.

3. 5. 1 ~Sum H The calculation of the minimum equilibrium sump solution pH considers the following delivered tank volumes, ice mass, and boron concentrations:

RMST - 420,000 gal, 2600 ppm B Accumulators(4) - 29,052 gal, 2600 ppm B RCS (hot zero power, no xenon) - 88, 958 gal, 2400 ppm B Sodium tetraborate ice - 2,372,000 lb, 1800 ppm B Boron injection tank - 900 gal, 22,500 ppm B The resulting pH is 7.6, which is sufficient to support a partition coefficient of approximately 600 (reference 2) which supports an elemental iodine DF of 78 (98%%d capacity) for the recirculating solution.

0291v:10/022787

1

3.6 HYOROGEN PRODUCTION Hydrogen produced by the corrosion of aluminum and zinc is a function of solution pH. The corrosion rates incorporated in the FSAR Chapter 14 combustible gas analysis were assumed to be based on a spray pH of 9.3 and 2000 ppm B.

The evaluation of hydrogen production presented in reference 1 concludes that aluminum corrosion decreases with decreasing pH and zinc corrosion increases.

P Specifically, the zinc corrosion rate at pH 5 was found to be as much as 20 percent greater than the pH 9.3 rate for the temperature range of 110 to 175 degrees F (see attached Figure 6-1 of the referenced report). However, it was further concluded that this low temperature increase would have a negligible impact on the aggregate hydrogen generation rate since the solution pH would be quickly raised into the caustic range by the melting sodium tetraborate ice.

Additionally, a corrosion rate constant comparison was made for the FSAR condition (pH 9.3, 2000 ppm B) versus the new reduced pH/increased boron condition (pH 7.6, 2400 to 2600 ppm 8) (reference 3). The comparison showed a rate constant change, for the nevi condition, of +1 to - 0.5 percent, depending on temperature. This variation is also concluded to have a negligible impact on the aggregate hydrogen generation rate.

To summarize, the rates of hydrogen generation due to corrosion of aluminum and zinc, for the increased boron/decreased pH condition, are enveloped by the analysis presented in the FSAR.

I

3. 7 EQUIPMENT QUALIFICATION The primary concerns of equipment qualification are protection of the stainless steel components of the emergency core cooling system from chloride induced stress corrosion cracking, failures of electrical components required to operate post-accident, and failures of containment coatings which could jeopardize the ECCS by flaking or pealing off, clogging the emergency sump and other flow paths, and thus restrict the flow of emergency core cooling water.

0291v:1D/022 787 10

3.8 PROTECTION OF STAINLESS STEEL To minimize the occurrence of chloride stress corrosion cracking of stainless steel, Hestinghouse recommends maintaining the equilibrium sump solution pH above 7.5 (Reference 4). The NRC recommends a solution pH in the range of 7 to 9.5 (Reference 5). The minimum calculated sump solution pH of 7.6 is consistent with these recommendations.

3.9 ELECTRICAL COMPONENTS Electrical equipment is tested to determine the ability of component seals to exclude the containment environment from the interior of the component. To maximize the challenge to the seal materials, high pH sprays, in the range of 8 to 11, have traditionally been used.

For all modes of ECCS operation, the solution pH with increased boron concentration will always be less than the corresponding pH with reduced boron. Hence, components qualified at higher pH may actually have a longer p ost-acci d8 nt service life in a lower pH (in the caustic range) environment.

3. 10 CONTAINMENT COATINGS Coatings are used in the containment to provide corrosion protection for metals and to aid in decontamination of surfaces during normal operation.

Like electrical equipment, coatings are tested with a high pH solution to maximize the potential deterioration of the coating, and may show better resistance to lower pH solutions.

4.0 IMPACT ON D. C. COOK UNITS 1 5 2 TECHNICAL SPECIFICATION The D. C. Cook Technical Specification that were affected by increasing the RMST and accumulator allowable boron concentrations are presented here via marked up technical specification pages.

0291 v:10/022787

5.0 FINAL SAFETY ANALYSIS REPORT FSAR /TECHNICAL SPECIFICATION CHANGES Please find attached the FSAR/Technical Specification changes that were modified as a result of the RWST and accumulator boron concentration increase. The ph limits in the basis of Specification 3.5.5 were also revised to reflect the increase in the boron concentration.

Changes to the Boration Systems basis (3/4.1.2) resulted from recalculating RWST volumes based on a boron concentration of 2400 ppm and bounding boron requirements for D. C. Cook Unit 1 extended fuel cycles. These changes include the additional borated water source volumes required to consider a boron dilution event during cooldown from HFP to 200 degrees-F (Mode 4) and cooldown from 200 degrees-F to 140 degrees-F (Mode 6).

Changes to the Boric Acid Tank (BAT) and RWST volumes in Specifications

3. 1.2.7 and 3.1.2.8 and in the Boration Systems basis (3/4. 1.2) are associate with the above boron dilution event.

6.0

~

SUMMARY

AND CONCLUSIONS

~ ~

The proposed 1ncrease 1n the RWST and accumulator allowable boron concentration limits to 2600 ppm has been assessed from a safety standpoint.

Based on these results, it is concluded the proposed boron concentration increases will have no adverse impact on the non-LOCA Accident Analysis, the LOCA Analysis or LOCA Related Design Considerations and is thus acceptable for implementation at D. C. Cook.

0291v:1D/022787 12

Confirmation that the boron concentration increases will provide enough margin to meet post-LOCA shutdown requirements will be concluded through the normal Westinghouse RSAC evaluation process.

7.0 REFERENCES

1. "Spray Additive Tank Deletion Analysis for the Donald C. Cook Nuclear Plant", MCAP-11020 (WCAP-11021, non-proprietary), December, 1985.

2. "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants", NUREG-0800, Section 6.5.2.

3. "The Relative Importance of Temperature, pH and Boric Acid Concentration on Rates of H2 Production From Galvanized Steel Corrosion",

NUREG/CR-2812, November, 1983.

4. "Stress Corrosion Testing", MCAP-7628, non-proprietary, December, 1970.

5. Branch Technical Position MTEB 6-1, "pH for Emergency Coolant Mater for PMR's".

0291 v:10/022787 13

FI POST-LOCA RCS/SLIP BGRCN C VS f%E-TRIP RCS BMyq gag D. C. cOrK uuIT I

"~ amia/

gtNT m/aieag borat cr~

+f~0 op p&l. Acceev]

~'p<IMAf L PJ/jr'f 4+ddrpA.

c INS IC lON pcs

+

I f!%

ias fIj >fAFJ E

~

~ [ceo (Jap /70le+

i~ iltS g Ill A>T mrs(~gyp peg dna <<~ ~~<L nf II~0 pp+

N 7 ~t7) ~<Araarlik> w/ rt.wurst T

0 bison &W. y/r W5O a ItN PAt.

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INO

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(gdpgl8 1 t)

INS QH 5& HI IA fN 7'%N ISO %Q NO l000 f0$ 0 I IM I I& tlat I~ y~ g7$ Q /ipse li$0 tlat-Tilt IC5 MCHI COCEN1llTIQI )at%)

Fi3ST-LOCA RCS/SUMP 8%% CONC VS PRE:-TRIP RCS BM3N CDg

0. C. COCK LNIT 2 IOT5 0

f <aso I

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I' FIGURE 6-1 HYDROGEN PRODUCTION RATE CONSTANTS FOR ZINC CORROSION RATE COHSTNI iOR 2IHC CORROSIN pH 9.3 - 22ee ppe B t p85 4 phe le 2 8 COHTAIHNKNT THP. (DKG. i) 6-5

ATTACHMENT A (Safety Evaluation)

FSAR TECHNICAL SPECIFICATION CHANGED PAGES

Al] active components of the safety '.-.jec='. cysts.".. w.i-:. -'--.-"-:. -""'ng the injection phase of a loss of coolant accident are lcca " outside the containment system. The safety injection pumps, centri ugal charging pumps< and residual heat removal pumps are located in the auxiliary building.

Recirculation Phase Spilled coolant and injection water which is collected in the containment recirculation sump following the injection phase is recirculated back to the reactor coolant system by the residual heat removal pumps. The containment spray pump suction is also supplied directly from the con-tainment recirculation sump. The reactor coolant system is supplied directly from the discharge of the residual heat removal heat exchangers, and from each of the heat exchanger outlets to the suction of the centrifugal charging and safety injection pumps which in turn pump into the coolant system.

The recirculation phase of operation has two modes, cold leg recirculation ~

and hot leg recirculation.'nitially, the discharge from the RHR pumps flows directly, and via the safety injection and charging pumps, to the same cold leg injection points used during the injection phase of opera-tion. Later in recirculation, the discharge of each safety injection pump is, along with the RHR pump discharge, switched to two individual hot leg injection points. The switch to hot leg recirculation is made in order to mhnimixe the potential for boron precipitation.

Hot leg injection may begin during the recirculation phase of operation whenever the reactor coolant system and secondary coolant system are cooled down. The changeover to hot leg injection is specified to occur tR, ours after the accident. At this time the residual approximately~2 heat generation rate has decayed to less than 1% of the nominal, the sensible heat in the stean generatcr secondary side will have been removed and the containment atmosphere and sump ligid temperature will have been reduced.

$ ~ 2a4 July, IPE"

Qf ACT IVITY 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:

boric acid storage system and associated heat tracing with:

ae A 1;;, 0 A.SL 1 f22f >>

of boron, and 229

2. Between 20,000 and 22,500 ppm

3. A minimum solution temperature of 145'F.

b. The refue'ling water storage tank with: yf qgp MS A.GL.C,

1. A minimum eagled volume of gallons,

2. A-efrniessswboron concentration of ppe, and

~oo

3. A minimum solution temperature of 35'F.

APPLICABILITY: MODES 5 and 6.

ACTION:

With.no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS 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 shall be demonstrated OPERABLE:

a. At least once per 7 days by:

l. . Verifying the boron concentration of the water,

2. Verifying the water level volume of. the tank, and

3. Verifying the boric acid storage tank solution temperature whenI 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 and the outside air temperature is

< 35 F.

- 3/4 1-15 AMENDMENT NO. 52 D. C. COOK .UNIT 1

REACTIVITY CONTROL SVSTcMS BORATED,~ATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION

~ ~

3.1.2.8 Each of the following borated wat r sources shall be OPEPASLE:

boric acid storage systan and as'sociated heat tracing with:

~

a. A us~LE. +H l.

1. '

minintum @esca-H~ volume of gallons, 2.'etween 20,000 and 22,500 ppg of boron, and

3. A minimum solution tanperature of 145'F.

b. The refueling water storage tank with:

1.' minimum contained volume of 350,000 gallons of water, Se4 we~ zloo ~n WoO toPM,

2. 2 3.. A minimum solution temperature of 70'F. i APPLICABILIiY: NODES I, 2, 3 and a.

ACTION:

a. With the boric acid storage system inoperable, restore the storage systen 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'orated to a SHUTDOWN MARGIN equivalent to at least 1" ak/k at 200'F; restore the boric acid s-orage systetn to OPERABLE status within the next 7 days or be in COLO 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 s orage tank inoperable, restore the tank to OPERABLE s:atus 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-OOHN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REOUIREMENTS 4.1.2.8 Each borated water source shall be denonstrated OPERABLE:

0. C. COOK - UNIT 1 3/4 1-16 Anen&ent .".o. 40

3/4 5 E! cPI:-'~CY CO..c'l Qi .:,'>T '", 'EPC ACCUSE'!ULATORS 3.5.1 Each reactor coolant syst m accumulator shall be OPERA".-LE "'.-:.::

a. T!le lsolat'ion valve c!) n, b.

c~

d.

A.

A

~:, ~-"d 8eti;een '-'29 ard 971 curio .eet of orate" nitrogen cover-pressure o,

+~4 be+;;een

~v

'L'a

+qoO c 585 and 653 c

psig.

APPLICAB;L 'TY: !'.OCES 1 q 2 and 3.*

ACT IOi"l:

a. With one accumulator inop rable, except as a ; suit OT a isolation valve, restore th nop rc Ie accumulat r status <<i thin one hour or be in HOT S!,"UTDOl:l ui .! in -.he .".e..:

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b. !Aith owe accumul tor inoocra~le due to the isolat cn va" .-

closed, eiiner i.-,".edia.ely o~en the isolation .alve or -.=-

ST~~;BEY,vithin one hour and '"e in O' i,-;UT."""'.."..':;ith-~n "- ..'

!10urs.

SURVE ILLA".CE oEOUIR"-.".E")TS 4.5.1 Each accumulator shall be demonstrated OP=RAiLE.

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by:

l. Verifying the:eater level and nitroaen cover-pressu'n the tan):s, and

2. Verifying that each accumulator isolation valve is c,-'n.

"Pressurizer Pressure above 1000 psig.

D.C.COOK-UNIT 1 3/4 5-1

E.,'( ( CORE CQL,",'Q SvST-"5 R FJ s 8 AA I EIs ST<'RAGE I At+ K LTg 1 "" .""~T ..l' rnR nP.""(-.

3.5.5 The refuel;ng water storage 'an!< (R!:SI) s >a 1 be . ~ J.P a.. A minimum contain d volume of 350,0:0 gallons of bora.ed

b. A %~iir boi on co>>cen:"a" i cn of ~~. ~

2-+GO A~D wlooo pphh cI.~ >

c. A minimum water temperature oi 70'F.

APPLlCA"-iLiTY: 4iOOES 1, 2, 3 ard 4.

ACT!G'".

'i!ith the refuel ng wa'er storace tank inooe. able, rest re -he .-.n'< ~ v OPERABLE status'within 1 hour -r be in at least HOT Tnt GB'( wl n'.n hours and in C"LQ Si UTCO'.lH within tne fo'l'cw rg 30 ncurs.

SURVE jl I "',Cc Qcf' T QcMc",TS 4.5.5 The R'EST sl all be demonstrated OPEiQSLE:

a. At leas. once per 7 days by:

1. Verifying the water level in the tank, and

2. Verifying the boron concentration of .he wat r.

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RMST temoerature when the outside air temperature is less;,".an 7 F.

O.C.COOK-UilIT 1 3]4 5-10 Amendment "io

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EMERGENCY CORE COOLING SYSTEMS t

BASES

'he contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

7 c7 The limits on contained water volume and boro, concentration of the RMST also ensure a pH value of between and . for the solution recirculated within containment af.er a LOCA. This pH band minimizes

the evolution of iodine and minimizes the effec 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. The temperature value of the RslST 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.

4

0. C. COOK - UNIT I B 3/4 6-3 Anendment Ho. 40

ADVAHCKONUCLEAR FUELS CORPORATION FFB O 6 1987 600 108(n AVENUE NE. PO BOX 90777. BELL El'UE WA 9&009 0777 (208) 453 4300 January 30, 1987 ANF-AEP/0550 Hr. Richard B. Bennett, Engineer Nuclear Materials 5 Fuel Management Indiana 8. Michigan Electric Company c/o American Electric Power Service Corp.

One Riverside Plaza, 20th Floor Columbus, OH 43215

Dear Hr. Bennett:

In response to your telephone request, Advanced Nuclear Fuels (ANF) has performed a review of the transient and LOCA analysis performed in support of D.C. Cook Unit 2. The review included the current work being performed for the steam line break and the analysis of record for the small break LOCA presented in the UFSAR. This review indicates that increasing the boron concentration in the refueling water storage tank (RWST) and the accumulators (ACC) to 2400 ppm would not adversely affect any of the ANF analysis.

If you have any questions regarding the above review, please feel free to contact our Hr. Jerry Holm (telephone 509-375-8142).

Sincerely, za os%~

H. G. Shaw Contract Administrator gf cc: H.P. Alexich J.M. Cleveland D.H. Malin V. VanderBurg Ah) Ai<i i( 8 Oi ii'wA i886< V'ii0N O~ Kwu

Attachment 14 to AEP:NRC:0916W

SUMMARY

OF ATTACHMENT 14 EVALUATION OF THE IMPACT OF 2000 GPM PRIMARY FLOW ON THE UNIT 1 DILUTION TRANSIENT PERFORMED BY WESTINGHOUSE ELECTRIC CORPORATION RESPONSE TO QUESTION 9 OF REACTOR SYSTEMS BRANCH TRANSMITTAL OF JANUARY 8, 1986 PREVIOUSLY SUBMITTED WITH AEP:NRC:0916P LETTER, NS-TMA-2273, FROM T. M. ANDERSON (WESTINGHOUSE)

TO V. STELLO (NRC) DATED JULY 8, 1980

Summary of Attachment 14 This attachment is divided into three parts. The first part entitled, "Revision of Figure A-1 of NS-TMA-2273" describes a new analysis for D. C.

Cook Unit 1 similar to that described in the letter from T. M. Anderson to V. Stello dated July 8, 1980 (Identifier NS-TMA-2273). The analysis was performed by our contractor, Westinghouse Electric Corporation. The curve from this calculation which corresponds to our maximum dilution flow rate of 225 gpm was used to prepare Unit 1 T/S Figure 3.1-3, Required Shutdown Margin.

The second part of this attachment is Attachment 1 to AEP:NRC:0916P.

As indicated in Attachment 1 to AEP:NRC:0916P, the methodology of NS-TMA-2273 has been in use on Unit 1 since beginning of Cycle 6. to AEP:NRC:0916P was approved in the SER for Amendment 82 to DPR-74.

The third part of this attachment is a copy of NS-TMA-2273. This document and Attachment 1 to AEP:NRC:0916P are being retransmitted to facilitate your review.

Revision of Fi ure A-1 of NS-TMA-2273 This discussion pertains to a revision of Figure A-1 provided in NS-TMA-2273.

The scales of Figure A-1 have been extended to account for increased RCS boron concentrations Modes 4 and 5 (hot and cold shutdown). American Electric Power has indicated that the Mode 5 maintenance level minimum RHR flow rate is 2000 gpm. This is more limiting than the current minimum Mode 4 RHR flow rate of 3000 gpm. As such, the Mode 5 RHR flow rate of 2000 gpm was assumed for this revision. The maximum dilution flow rate is given as 225 gpm on page 14.C-21 (Unit 1) of D. C. Cook FSAR. The D. C. Cook Unit 1 plant specifics, as noted above, have been incorporated in the development of the revised curve.

American Electric Power has decided to incorporate in the Technical Specifica-tions shutdown margin protection to ensure adequate operator response time for the mode 4 and 5 dilution transient. This is being done by applying the Westinghouse methodology described in NS-TMA-2273. In the process of generating a revised curve, which describes the shutdown margin requirements as a function of RCS boron concentration and possible dilution flow rate, certain assumptions of NS-TMA-2273 may no longer be applicable. In particular, the assumptions stating that in all cases a shutdown margin of 5%

delta-k/k (K eff-ff < 0.95) is considered sufficient for continued operation without a requfrement for control rod bank withdrawal is no longer valid. Due to increased RCS boron concentrations and the assumed minimum RHR flow rate of 2000 gpm, the revised curves show that a shutdown margin greater than 5S delta-k/k is required for dilution flow rates greater than 250 gpm.

Figure'1 provides the shutdown margin requirements as a function of initial Reactor Coolant System concentration and maximum possible dilution flow rate.

Figure 1 is based on DE C. Cook Unit 1 plant conditions as listed below:

1. The Reactor Coolant System effective volume is limited to the vessel and the active portions of the hot and cold legs when on RHR, i.e., steam generator volumes are not included.

2. The plant is borated to a shutdown margin greater than or equal to 1%

delta-k/k.

3. Uniform mixing of clean and borated RCS water is not assumed, i.e.,

mixing of the clean, injected water and the affected loop is assumed but instantaneous, uniform mixing with the vessel, hot leg, and cold leg volume upstream of the charging lines is not assumed. Thus a "dilution front" moves through the cold legs, downcomer, and lower plenum to the core volume as a single volume front.'his results in subsequent decreases in shutdown margin due to dilution fronts moving through the active core region with a time, constant equal to the loop transit time when on RHR. The RHR flow rate assumed for this D. C. Cook Unit 1 figure is 2000 gpm.

Figure 1 notes areas of acceptable operation of different dilution flow rates as a function of the RCS boron concentration and borated shutdown margin ff).

(Keff For a given dilution flow rate, if the RCS boron concentration and shuNown margin result in a point placed to the left of the flow rate line, no control rod bank withdrawal is necessary. If the results place the plant to

the right of the line, then either the shutdown margin must be increased such that the plant is moved to the area of acceptable operation, or 1% delta-k/k in control rods must be withdrawn to provide additional shutdown margin. The tripping of the withdrawn rods provides positive operator indication that a dilution event is in progress and additional time for operator termination of the event.

Figure 1 is based on best estimate calculations for the "all rods in" configuration.

Use of Figure 1 is applicable any time there is boration/dilution capability from the normal boric acid blending system. The above procedure is not required if boration and/or makeup during cold and hot shutdown is performed utilizing water from the RWST. This requires that the normal dilution/

boration path is isolated from the charging path.

03 I

8 Pl

( V I

A i XI Vl Ci

,500 Vl j QAf .

B

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Acceptabl e Operation colo Bank iltthdraaeaX 1X gk/k Bank itithdreaea1 Required 0

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AEP:NRC:0916P

~uestio 9 The times required for loss of shutdown margin from boron dilut'on are provided on Page 188 of X."a-1NF-85-64. These times are significant for.

providing operating reaction times only following the initiation of an alarm.

For each reactor condition given 'n Table 15.4.6.1, provide the time following initiation of the boron dilution event to the time when the alarm would function. Discuss diversity and redundancy of available alarms.

R~es onse 9 A) The time from initiation of dilution to the time of alarm has not been specifically calculated for the analysis presented in XN-NF-85-64, Rev.

1, Supp. 1. Instead, the analysis in XN-NF-85-64 (P), Rev. 1, Supp. 1, was performed in a similar manner to the analysis presented in Section 14.1.5 oF tke Unit 2 Donald C. Cook Nuclear Plant ~Udated ~FS Additional detail on the FSAR analysis which bounds operation in Modes 4 5 and 6 is provided in a letter (AEP:NRC:0860I) from M. P. Alexich to Harold R. Denton dated May 17, 1984. The analysis is also described in a letter (NS-TMA-2273) from T. M. Anderson of Westinghouse Electric Corporation to Victor Stello dated July 8, 1980. The results have been in use on Unit 1 since the beginning of Cycle 6 and on Unit 2 since the beginning of Cycle 3.

Both the FSAR analysis and the XN-NF-85-64(P), Rev. 1, Supp. 1 analysis for Modes 4, 5 and 6 ensure that 15 minutes are available from the initiation of dilution to the loss of shutdown margin. Volumes used in these analyses are limited to those assumed to have active flow.

As indicated in the updated FSAR and XN-NF-85-64(P), Rev. 1, Supp. 1, substantially longer times are available for operator response for the cases of dilution during startup and dilution during full power operation. The FSAR Mode 3 analysis is performed for startup from a reactor coolant system boron concentration of 2000 ppm.

B) Indications available to the operator include:

1) Status indication of the Chemical and Volume Control System and Reactor Makeup Water System with, a ~ Indication of boric acid and clean makeup flow rates including alarms on deviation from setpoint for both of these flows.

These alarms would be expected to occur at the initiation of any inadvertent dilution involving the blender.

b. CVCS valve position status lights, and C. Reactor Makeup Water Pump "running" status light.

AEP:NRC:0916P

~Res onse 9 ~Cont <l)

2) Source Range Neutron Flux with,

a. High Flux at Shutdown Alarm set at half a decade above background. This alarm is expected to occur after the dilution transient has been in progress for a period of time.

b. Use of the audible count rate indication to distinguish significant changes in flux, i.e., a doubling of the count rate.

c. Periodic, i.e., frequent surveillance of the Source Range meters and continuous strip chart recorder performed by the operator.

During startup operations, the high flux at shutdown alarm is not available. Additional indications available during startup operations include pressurizer and volume control tank levels. During power operations, the high flux at shutdown alarm and audible source range indications are not available. Source range meters and continuous strip chart indication are replaced by power range and intermediate range meters and a continuous strip chart which selectively displays these indications. When the rods are in automatic, rod insertion low and low-low alarms are available. When rods are in manual, Overtemperature

,Delta Temperature trip, alarm, and turbine runback are available.

1 Qtt ii

.nghouse Eiectric Corporation Power Systems PWR Systems Gvistett Bu355 Pittsaug Peasytneia 152K July 8, 1980 iVS-TMA-2273

. Mr. Victor Stello Office of NucIear Reactor Regulation U.S. Nuclear Regulatory Conmission Phillips Building 7920 Norfolk Avenue Bethesda, MD 20014

SUBJECT:

Boron Dilution Concerns at Cold and Hot Shutdown

Dear Mr. Stello:

On June 27, 1980, Ed Jordan of your staff was notified by Westinghouse of an Unreviewed Safety guestion under 10CFR50.59. This notification concerned the potential for an inadvertent boron dilution event while shutdown and operating on the Residual Heat Removal System. Attachment 1 is the text of the written notification supplied Co our customers on July 8, 1980 which outlines potential Westinghouse concerns and the basis for recomended interim actions which address these concerns, These interim actions are somewhat modified frotti those previously reported. If there are any questions regarding the attached, please contact D. M. Call at 412/373-5074.

Very truly yours, T. M. Anderson, Manager Nuclear Safety Department Attachment cc; E. Jordan Moods

ATTACHMENT I On June 27, 1980. you were notified of. certain Westinghouse concerns and recom-mended actions regarding .he potential for an inadvertent boron dilution event at cold or hot shutdown conditions while on the Residual Heat Removal System.

This notification was in accor d with Mestinghouse determination that these con-cerns constitute an Unreviewed Safety guestion under IOCFR Part 50.59. The NRC Office of inspection and Enforcement was also notified on June 27, 1980 that these concerns have generic applicability to Westinghouse-supplied nuclear power plants. further clarification was made to the NRC Office of Inspection and Enfold cement on June 30, 1980 that Westinghouse concerns are not applicable while the plant is greater than 5> shutdown.

This letter is intended to formally document these concerns and to provide ad-ditional relevant information. This letter also modifies the ear lier recommend-ed actions by a more detailed specification of applicable plant operating conditions.

Inadvertent boron dilution at shutdown has been generally regarded as an event

~ ~

which can be identified and terminated by operator action prior to a return to

~ ~

critical. Automatic protection has not been a standard feature for Westinghouse

~

plants. Westinghouse has recently been conducting a general investigation of

~ ~ ~

~

this potential event relative to the licensing requirements imposed on newer

~

plants not yet in oper ation. This investigation is not yet complete. However, it has been determined that under certain shutdown conditions and with certain assumed dilution rates, adequate time for operator action to prevent a return to critical may not be available.

The current Mestinghouse evaluations are based on plant conditions as noted below:

1. The Reactor Coolant System effective volume is limited to the vessel and the active portions of the hot and cold legs when on RHR, i.e., steam gen-volumes are not included. 'rator

2. The plant is borated to a shutdown margin greater than or equal to 1%

ak/k.

3. Uniform mixing of clean and borated RCS water is not assumed, i.e., mixing of the clean, injected water and the affected loop is assumed but instan-taneous, uniform mixing with the vessel, hot legs, and cold leg volumes upstream of the charging lines is not assumed. Thus a "dilution front" moves through the cold legs, downcomer, and lower plenum to the core vol-ume as a single volume front. This results in subsequent decreases in shutdown margin due .o dilution fronts moving through the active core region with a time constant equal to the loop transit time when on RHR (five to seven minutes).

If a return to critical occurs as a'result of an inadvertent dilution, the fol-lowing potential concerns have been identified:

A rapid,,uncontrolled power excursion into the Iow and intermediate power ranges occurs, resulting in a power/flow mismatch due to the Iow flow (approximately I - 2" of nominal) provided by the RHR pumps;

2. The potential exists for significant system overpressurization. Pressure increases above the RHR cut off head (approximately 600 psig) further ac-centuate the effects of a power/flow mismatch when all RCS (RHR) flow is lost. An investigation of the adequacy of existing cold overpressurization protection systems is necessary in order to assess the full impact of this potential problem.

3. The potential exists for limited fuel damage. This is not currently a significant concern. Preliminary evaluation indicates that the potential for exceeding ONB limits is Iow due to the cold initial operating condi-tions. Further investigation of this problem is underway.

The recommended interim actions to prevent or mitigate an inadvertent boron di-lution at shutdown conditions are detailed in Appendix A. If no cocked control rods are required, as specified in Figure A-l, the plant operator has fifteen minutes from the initiation of dilution event to terminate the event before a return to critical occurs. It is the Westinghouse position that a fifteen min-ute time interval from the initiation of the dilution to the time shutdown mar-gin is lost is sufficient time for operator action. lf cocked control rods are required, the sour ce range reactor trip provides positive indication f'r iamed-iate operator action to terminate dilution.

1t is expected that the operator has available the following information for determination that a dilution event is in progress:

I. Source Range neutron Flux with,

a. High Flux at Shutdown Alarm set at half a decade above background.

b. Use of the audible count rate indication to distinguish significant changes in flux, i.e., a doubling of the count rate.

C. Periodic, i.e., frequent surveillance of the Source Range meters per-formed by the operator.

Status indication of the Chemical and Yolume Control System and Reactor Makeup Mater System with,

a. Indication of boric acid and blended (total) flow rate, or

b. Indication of boric acid and clean makeup flow rate,

c. CVCS valve position status lights, and

d. Reactor Makeup Mater Pump "running"'status light.

The operator action necessary upon determination that .a dilution event is in pro-gress (by High Flux at Shutdown Alarm, Source Range Reactor Trip, "P-6 Available" indication, high indicated or audible count rates, or make up flow deviation alarms) is:

1. Immediately open the charging/SI pump suction valves from the RMST (that open on receipt of an "S" signaI). (For 312 plants these are LCY-116-8, 0.

For 412 plants these are LCV-112-0, E.)

2. Immediately close the charging/SI pump suction valves from the VCT (that close on receipt of an "S" signal). (For 312 plants these are LCY-IIG-C, E.

For 412 plants these are LCV-112-8, C.)

3. For two-loop plants, iamediately open the charging suction valves from the RMST. (For 212 plants these are LCY-113-8 and LCV-112-C.) . Also inmediate-ly close the charging suction valves from the VCT. (For 212 plants these are LCY-113-A and LCV-112-8.)

r Through the use of Appendix A and the above noted operator action requirements, Mestinghouse is attempting to minimize the operational burden placed on the plant to prevent or mitigate an inadvertent dilution event while maintaining adequate safety margin. Our investigation of this event is continuing. A detailed ana-lytical model of the system response to a dilution event at shutdown conditions is being developed and the potential for system overpressurization and fuel fail-ure will subsequently be assessed. The Mestinghouse investigation is expected to be compIeted by September 15, 1980. Me will keep you informed as to the re-sults of our efforts.

APPENDIX A Figure A-l, attached, provides the shutdown margin requirements as a function of Reactor Coolant System boron concentration and maximum possible dilution flow rate. Prior to use of this figure, the plant must determine the maximum dilution flow rate of all charging pumps not rendered inoperable once the plant is placed on RHR. To cover all modes, it should be assumed that the flow rate is based on pump runout unless there are flow limiting devices in the system (orifices, pip-ing resistances, etc.). The Reactor Makeup Mater pump capacity may be limiting in the determination of the maximum possible dilution flow rate.

Figure A-I notes areas of acceptable operation of three different dilution flow rates as a function of RCS boron concentration and borated shutdown margin (K ff).

For a given dilution flow rate, if the RCS boron concentration and shutdown margin result in a point placed to the left of the flow rate line, no control rod bank withdrawal is necessary. If the results place the plant to the r ight of the line, then either the shutdown margin must be increased such that the plant is moded to the area of acceptable operation, or l~ 4k/k in control rods must be withdrawn to provide additional shutdown margin. The tripping of the withdrawn rods provides positive operator indication that a dilution event is in progress and additional time for operator termination of the event. In all'cases-.

a shutdown margin of 5" ak/k (K ) ( 0.95) is considered sufficient for contin-ued operation without a requirei t for control rod bank withdrawal.

Figure A-I is based on best estimate calculations for the "all rods in" configu-ration. It is recomended that the Mestinghouse Nuclear Design Report for your plant be used a's a reference in determining the RCS boron concentration with the appropriate conservatism to be used in the figure. The Mestinghouse Nuclear Fuel Division is availab'te to provide assistance in meeting the constraints imposed by the Figure A-I requirements.

Use of Figure A-I is applicable any time there is boration/dilution capability from the normal boric acid blending system. The above procedure is not r equT'red if boration and/or makeup during cold and hot shutdown is performed utilizing water from the RMST. This requires that the normal dilution/boration path is isolated from the charging path. Two means of lockout to isolate the charging path are available:

1. Lock out Reactor Makeup Mater Supply.

This is accomplished by valve 8338 for 212 plants, valve 8457 for 312 plants, and valve 8455 for 412 plants.

OR:

Z. Lock out valves between the boric acid blender and the VCT.

These are FCV-111B, FCV-llOB, 8339, 8355, and 8361 for 212 plants; FCV-114A, FCV-113B, 8454, 8441, and 8439 for 312 plants; FCY-111B, FCV-110B, 8453, 8441, 8439 for 412 plants.

This recommendation precludes the occur rence of an inadver tent dilution while bor ating or making up water from the RMST under these conditions.

l

(

~ A I 4 4 elk4I 4 Jf L Ct Hg ~ pQ@

E~

ADVANCEDNUCLEAR FUELS CORPORATION MAR ' 1987 800 t08th AVENUE NE PO BOX 90777, BELLEVUE WA 9800947777 t 208) 483-4300 March 5, 1987 ENC/AEP-0556 Mr. Rick Bennett, Engineer Nuclear Materials & Fuel Management Indiana 6 Michigan Electric Company c/o American Electric Power Service Corp.

One Riverside Plaza, 20th Floor Columbus, OH 43216-6631

Dear Mr. Bennett:

Attached is a recommended change to the D.C. Cook Unit 1 Technical Specification on Fq to allow operation of ANF fuel to peak pellet exposures of 51 GWd/MT. A justification of this change is also attached for your use in obtaining NRC approval for this change. This is a revision to our letter ENC/AEP-0535 dated November 11, 1986.

If you have any questions regarding the attachment, please contact our Mr.

J.S. Holm (telephone 509-375-8142).

Sincerely, H. G. Shaw Contract Administrator gf Attachment cc: J. M. Cleveland D. H. Malin V. VanderBurg J. S. Holm (ANF)

' 4F UAtE OF nRAF @YERK UNiON Q~ Kivu

Attachment D C COOK UNIT 1 TECHNICAL SPECIFICATION CHANGE Ref: (1) XN-NF-85-115, Rev. 1, "D.C. Cook Unit 1 Limiting Break K(Z)

LOCA/ECCS Analysis," November 1986.

(2) XN-NF-85-68(P), Rev. 1, "Donald C. Cook Unit 2 Limiting Break LOCA/ECCS Analysis, 10% Steam Generator Tube Plugging, and K(Z)

Curve," April 1986.

(3) XN-NF-85-117, Supp. 1, "St. Lucie Unit 1 Revised LOCA/ECCS Analysis with 15% Steam Generator Tube Plugging Break Spectrum and Exposure Results," December 1985.

A LOCA/ECCS analysis justifying the operation of ANF fuel currently in the D.C. Cook Unit 1 reactor is presented in Reference 1. The analysis in that report supports a peak Fq of 2.04 with an axial dependence as shown in Figure

1. This analysis is applicable to the ANF fuel currently in the D.C. Cook Unit 1 reactor, with a minimum peak rod average exposure greater than 20 GWd/MT and anticipated to be less than 47 GWd/MT.

Justification for an exposure independent Fq for D.C. Cook Unit 1 is based on an exposure analysis for D.C. Cook Unit 2 (Reference 2). Peak cladding temperatures are dependent upon fuel rod initial stored energy, which for the EXEM/PWR models increases from 0 to about 2 GWd/MTM and then decreases with exposure. The analysis for D.C. Cook Unit 2 with 17xl7 fuel geometry demonstrated that over the exposure range of 0 to 47 GWd/MTM, the peak cladding temperature decreased with exposure for exposures beyond the peak stored energy exposure. A similar trend was observed for St. Lucie Unit 1 with 15x15 fuel geometry (Reference 3). Similar results would be expected for D.C. Cook Unit 1 with 15x15 fuel geometry using EXEM/PWR models. Based on the trend of decreasing peak cladding temperature with increasing exposure, the analysis in Reference 1 is conservative and supports an exposure independent Fq of 2.04, along with the K(Z) curve shown in Figure 1, for ANF. fuel at peak rod average exposures between 20 and 47 GWd/MTM. A peak rod average exposure of 47 GWd/MTM is equivalent to a peak pellet exposure of 51 GWd/MTM.

1.2 (0 0 1.0) (6.0, 1.0) 1.0 (11.01, 0.936

~tL8 C3 (12.0, 0.490) 0.0 0 4 S 6 7 8 9 10 11 12 CORE HEIGHT (FT)

Figure 1 Hot Channel Factor Normalized Operating Envelope, Fq=2.04, K(Z)'Function

INDIANA 8 MICHIGAN ELECTRIC COMPANY P.O. BOX 16631 COLUMBUS, OHIO 43216 January 29, 1987 AEP:iVRC:0940E Donald C. Cook Nuclear Plant Unit iVo. 1 Docket No. 50-315 License No. DPR-58 D. C. COOK UNIT 1 LIMITING BREAK K(Z)

LOCA/ECCS ANALYSIS U. S. iVuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555

Dear Sirs:

The purpose of this letter is to inform you that Exxon Nuclear Company (ENC) has transmitted to you proprietary and non-proprietary copi.es of their report No. XiV-iVF-85-115, Rev. 2, entitled "D. C. Cook Unit 1 Limiting Break K(Z) LOCA/ECCS Analysis," via their letter No. GNW:001:8?, dated January 15, 1987. By this letter, we request that these documents be added to our Unit 1 docket, No. 50-315. The report documents the resu's of the LOCA/ECCS analysis performed by ENC to determine the K(Z) for the ENC fuel in Unit 1 of the D. C. Cook Plant. The analysis supports operation of D. C. Cook Unit 1 at its currently licensed thermal power rating of 3250 MW.

Revision 0 of this report was transmitted to you on February 5, 1986 via our letter AEP:iVRC:0940C. In that letter, (and in, the NRC staff's subsequent safety evaluation report dated February 21, 1986) it was indicated that the Fuel Cooling Test Facility (FCTF) reflood correlations which were used by ENC in their analysis were undergoing NRC review, and that the K(Z) curve presented in XN-iVF-85-115 Rev. 0 would be reexamined after completion of the NRC's review of the FCTF data. Subsequent to this, ENC has modified the FCTF correlations to resolve NRC concerns. ENC has received formal approval from the NRC to use the correlations as modified. The analyses presented in Revision 2 to XiV-NF-85-115 utilize the revi.sed FCTF correlations. We note, however, that the K(Z) curve presented in Revision 0 of XN-NF-85-115 remains unchanged in Revision 2 to that document. Revision 2 to XN-NF-85-115 also incorporates minor editorial changes to Table 2.1 of the document. These changes correct errors in the listed volumes for the reactor vessel and pressurizer and add a footnote to denote the amount of steam generator tube plugging used in the analysis. These changes are editorial only, and do not impact the K(Z) results.

AEP:NRC:0940E This document has been prepared following Corporate procedures which incorporate a reasonable set of controls to insure its accuracy and completeness prior to signature by the undersigned.

I Very truly yours Alexich President ll>

I'ice I ~

Attachment cc: John E. Dolan W. G. Smith, Jr. - Bridgman R. C. Callen G. Bruchmann G. Charnoff NRC Resident Inspector - Bridgman

INDIANA 8, MICHIGAN ELECTRIC COMPANY P.O. BOX 1663I COLUMBUS, OHIO 43216 February 20, 1987 AEP:NRC:1018 Donald C. Cook Nuclear Plant Unit No. 1 Docket No. 50-315 License No. DPR-58 PROPOSED TECHNICAL SPECIFICATION CHANGE REGARDING EXTENSION OF PEAK PELLET EXPOSURE FOR ADVANCED NUCLEAR FUEL CORPORATION FUEL U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555

Dear Sirs:

This letter and its attachments constitute an application for amendment to the Technical Specifications (T/Ss) for the Donald C. Cook Nuclear Plant Unit No. 1. Specifically, we propose to modify T/Ss 3/4.2.2 (Heat Flux Hot Channel Factor - F (Z)) and 3/4.2.6 (Axial Power Distribution) to allow an increase in the al)owed peak pellet exposure for Advanced Nuclear Fuel Corporation (ANF) (formerly Exxon Nuclear Company) fuel from its present

.value of 48.0 MwD/kg to 51.0 MwD/kg.

Predictions of fuel burnup made prior to the beginning of the current cycle indicated that the ANF fuel would not exceed the current peak pellet exposure limit of 48.0 MwD/kg. However, recent flux maps have indicated the potential for the ANF fuel to slightly exceed the limit prior to discharge at the end of cycle. According to our flux maps, the limit may be exceeded as early as May 3, 1987, approximately three ~eeks prior to the start of the upcoming Unit 1 refueling outage, currently scheduled to begin on May 24, 1987. Because this situation creates the potential for a required early shutdown of the unit, we request an expedited review of the proposed changes and a response by April 30, 1987. We are currently preparing proposed simplifications to the D. C. Cook Unit 1 power distribution monitoring T/Ss. These proposed changes, which are intended to provide consistency between the D. C. Cook Units 1 and 2 T/Ss, will most likely propose deletion of the burnup requirements from the T/Ss. However, because we will reach our peak pellet exposure limit in early May 1987, we have decided to submit the peak pellet exposure extension request separately to allow adequate time for NRC review.

The reasons for the proposed changes and our analysis concerning significant hazards considerations are contained in Attachment 1 to this letter. The proposed revised T/S pages are contained in Attachment 2.

Attachments 3 and 4 contain evaluations performed by ANF in support of the changes. These evaluations are discussed in more detail in Attachment 1.

AEP:NRC:1018 Since Attachment 4 contains ANF proprietary information, we have included an affidavit to that effect with it. Attachment 5 contains a non-proprietary version of the ANF document in Attachment 4.

The ANF analyses we have attached provide justification for an extension of the allowed peak pellet exposure for their fuel to 48.7 Mwd/kg, rather than the 51.0 Mwd/kg we have proposed in this submittal. As detailed in Attachment 1, it is our understanding that the additional analyses necessary to support the value of 51.0 Mwd/kg can be reviewed by us under the provisions of 10 CFR 50.59 and therefore will not require an additional submittal. The value of 48.7 Mwd/kg should be sufficient to allow operation to continue until the start of the Unit 1 refueling outage, currently scheduled for May 24, 1987. At this time, however, we are investigating the possibility of delaying the outage start date due to various system concerns, such as outages in other of our operating units. For this reason, we are considering having analyses performed to justify peak pellet exposure limits for ANF fuel greater than 48;7 Mwd/kg. ANF has informed us that these analyses may be extensive and involve several weeks preparation time.

In order to allow adequate time for NRC review of our proposed changes and for our own evaluation of our peak pellet exposure needs, we have chosen to submit analyses supporting peak pellet exposures of 48.7 Mwd/kg and to pursue exposures beyond this value via the 10 CFR 50.59 process. This approach was discussed with the NRC staff on February 12, 1987. Since at the present time we can only justify a value of 48.7 Mwd/kg, we would implement administration controls to prohibit operation above peak pellet exposures for ANF fuel of 48.7 Mwd/kg without appropriate analyses and 10 CFR 50.59 review.

We believe that the proposed changes will not result in (1) a significant change in the types of effluents or a significant increase in the amounts of any effluents that may be released offsite, or (2) a significant increase in individual or cumulative occupational radiation exposure.

These proposed changes have been reviewed by the Plant Nuclear Safety Review Committee (PNSRC), and will be reviewed by the Nuclear Safety and Design Review Committee (NSDRC) at their next regularly scheduled meeting.

In compliance with the requirements of 10 CFR 50.91(b)(1), copies of this letter and its attachments have been transmitted to Mr. R. C. Callen of the Michigan Public Service Commission and Mr. G. Bruchmann of the Michigan Department of Public Health.

Pursuant to 10 CFR 170.12(c), we have enclosed an application fee of

$ 150.00 for the proposed amendment.

11 I

AEP:NRC:1018 This document has been prepared following Corporate procedures which incorporate a reasonable set of controls to insure its accuracy and completeness prior to signature by the undersigned.

Very truly yours, .

/

M.'. Alexich :7 Vice President T \

cm Attachments cc: John E. Dolan W. G. Smith, Jr. - Bridgman G. Bruchmann R. C. Callen G. Charnoff NRC Resident Inspector - Bridgman J. G. Keppler - Region III

Attachment 1 to AEP:NRC:1018 Reasons and 10 CFR 50.92 Analyses for Changes to the Donald C. Cook Nuclear Plant Unit No. 1 Technical Specif ications

Attachment 1 to AEP:NRC:1018 Page 1 Back ound This letter proposes to increase the allowable peak pellet exposure for ANF fuel from its present value of 48.0 Mwd/kg to a higher value of 51.0 Hwd/kg.

Peak pellet'exposure is in general limited by either LOCA analysis considerations or fuel mechanical design characteristics. For ANF fuel in Unit 1, the value has been included in the T/S's specifically because of LOCA analysis considerations, which are discussed in more detail below. The limit of 48.0 Mwd/kg appears in the graphs of exposure-dependent F limit L

(F (E g,)) and normalized F limit (T (E R )) found in Figure 3.2-4 of the Unit 1 T/Ss (page 3/4 2-23). It also appears in the F uncertainty factors E (Z) (page 3/4 2-7) and F (page 3/4 2-20).

P P During the design phase of a fuel cycle, predictions of peak pellet exposure are made, and these predicted exposures are ensured to be within applicable limits (mechanical and LOCA, as well as T/S where applicable). For ANF assemblies in D. C. Cook Unit 1, we monitor burnup via flux mapping to ensure adherence to T/S limits. Recent flux mapping has demonstrated that the potential exists for several ANF fuel assemblies to slightly exceed their 48.0 Mwd/kg T/S limit by May 3, 1987, approximately three weeks prior to the scheduled start of the upcoming Unit 1 outage, which is currently scheduled to begin on May 24, 1987.

Currently, all new fuel for D. C. Cook Unit 1 is being supplied by Westinghouse Elec ric Corporation (Westinghouse). The present cycle (Cycle 9) uses only 34 ANF assemblies. Of these 34 assemblies, only 4 are expected to exceed the current peak pellet exposure limit of 48.0 Mwd/kg.

By May 24, 1987 none should have exceeded the limit by more than 0.7 Hwd/kg, which represents an excess of less than 2%. Current design plans for the Cycle 10 core do not call for any of the ANF assemblies to be reused, although these plans are subject to change should we encounter unanticipated fuel failures or damaged assemblies during refueling.

ANF has evaluated the saf< ty impact of operation up to 51.0 Hwd/kg for LOCA considerations (Attachment 3), but only to 48.7 Mwd/kg for mechanical design

Attachment 1 to AEP:NRC:1018 Page 2 considerations (Attachment 4). The mechanical design evaluation was limited to 48.7 Mwd/kg because this value could be supported in large part by extrapolations from existing analyses. ANF has informed us that analyses to support higher values of peak pellet exposure may be extensive and involve several weeks preparation time. Thus, we were unable to have these analyses performed in time to accompany this letter and still allow adequate time for NRC review. Additionally, as discussed in the cover letter, we are unsure at this time whether peak pellet exposures, beyond 48.7 Mwd/kg are even necessary. We are therefore unsure whether we want to undertake the expense and effort to have the analyses performed.

ANF has informed us that in meetings with the NRC staff held in August 1986, the staff explained that fuel mechanical design analyses could be reviewed under the provisions of 10 CFR 50.59 without NRC review provided that ANF followed the methodology outlined in their document XN-NF-82-06, Rev. 1, "Qualification of Exxon Nuclear Fuel for Extended Burnup", and if the batch average is below the approved high burnup level in this document. Should we decide to pursue peak pellet exposures beyond 48.7 Mwd/kg, which equates to batch average burnup considerably less than batch average burnups approved in XN-NF-82-06 Rev. 1, we propose"'to have ANF do so using the parts of XN-NF-82-06 Rev. 1 which are applicable to peak pellet exposure, and to review these analyses under the provisions of 10 CFR 50.59. (Since peak rod and peak assembly exposures are not being changed beyond that addressed in the currently approved mechanical design safety evaluation, XN-NF-84-25, not all aspects of the XN-NF-82-06 Rev. 1 methodology need to be addressed.)

Descri tion of Pro osed Chan es The ANF evaluations presented in Attachments 3 and 4 provide support for a peak pellet exposure limit of 51.0 Hwd/kg based on LOCA considerations, but only 48.7 Mwd/kg based on mechanical design considerations. These analyses L

allow the exposure-dependent peaking factor limit, F (ER ) of T/S Figure Q

3.2-4 (p. 3/4 3-23) to remain at 1.82 (its present value at 48.0 Mwd/kg peak pellet exposure). We have redrawn T/S Figure 3.2-4 to show the curve extending to an F L (E ' value of 1.82 at 51.0 Mwd/kg. T (E2 ), the L

normalized F (E 2 ), which is also contained in T/S Figure 3.2-4, has been

Attachment 1 to AEP:NRC:1018 Page 3 similarly redrawn. We have also modified the values of E (Z) in T/S 4.2.2.2 P

(p. 3/4 2-7) and F in T/S 3.2.6.g. (p. 3/4 2-20) to define these factors as P

1.0 from 48.0 to 51.0 Mwd/kg peak pellet exposure. E (Z) is an uncertainty P

factor to account for a reduction in the F (E R) curve due to an accumulation of exposure between flux maps. The quantity F is a similar P

factor for use with the Axial Power Distribution Monitoring System (APDMS).

The values of these factors are related to the slope of the F L

(E g, ) curve L

from T/S Figure 3.2-4. A flat slope for the F (E g) curve, as we have proposed between 48.0 and 51.0 Mwd/kg, results in no change in the allowable value of F L (E a) between flux maps and thus no penalty (penalty factor of 1.0). This is consistent with the value of 1.0 assigned to these factors between peak pellet exposures of 0.0 and 17.62 Mwd/kg where the slope of F L Q

(Eg, ) is also flat. Since at the present t'ime we can only justify a peak pellet exposure of 48.7 Mwd/kg, we would implement administrative controls to prohibit operation beyond 48.7 Mwd/kg without an analysis which uses the methodology from the appropriate sections of XN-NF-82-06 Rev. 1 and a subsequent review of these analyses under 10 CFR 50.59.

Justification for Pro osed Chan es The following justifications address LOCA considerations up to 51.0 Mwd/kg and mechanical design considerations up to 48.7 Mwd/kg. As discussed previously, we propose that any additional mechanical design analyses which may be performed in support of higher peak pellet burnups will be performed using the approved methodology of XN-NF-82-06 Rev. 1 and will be reviewed under the provisions of 10 CFR 50.59.

1. LOCA Considerations F does not vary as a function of burnup for Westinghouse fuel in either the D. C. Cook Units 1 or 2 T/Ss. For ANF fuel, it varies as a function of burnup only in the Unit 1 T/Ss. The reason the burnup dependence is included for ANF fuel in Unit 1 is that the limits were based on ANF LOCA analyses dating back to the mid-1970s, which used a burnup-dependent F . More detailed and modern ANF LOCA analyses do not require F to be burnup-dependent. For example F for ANF fuel in I

Q to AEP:NRC:1018 Page 4 D. C. Cook Unit 2 is a constant at 2.10, with no exposure dependence or limits found in the T/Ss. The newer ANF analyses have determined the limiting exposures with regard to peak clad temperature concerns to be at relatively low exposures (less than 10 Mwd/kg). Similarly, Westinghouse LOCA models assume a constant value for F throughout the cycle.

ANF has recently performed a new limiting break K(Z) LOCA/ECCS analysis for Unit 1. This analysis, which is contained in XN-NF-85-115 Rev. 2, was sent to you directly by ANF in their letter GNW:001:87, dated January 15, 1987 (as noted in our letter AEP:NRC:0940E, dated January 29, 1987). This analysis used the modern ANF evaluation methods including the Fuel Cooling Test Facility (FCTF) reflood heat transfer correlations. The document discusses analyses performed for peak pellet exposures of 2 Mwd/kg and 9 Mwd/kg, which ANF has determined to be bounding with regard to,peak clad temperature. These analyses assumed an F value of 2.04 peaked at the core midplane at 2 Mwd/kg and 1.95 peaked at the core top at 9 Mwd/kg. Both of these values are conservative with respect to the value of 1.82 required by Unit 1 T/S Figure 3.2-4 at 48 Mwd/kg.

As discussed in Attachment 3, ANF has informed us that the analyses they performed for XN-NF-85-115 Rev. 2 are applicable up to a peak rod average exposure of 47 Mwd/kg, which corresponds to a peak pellet exposure of 51 Mwd/kg. This is based on comparisons of exposure analyses ANF performed for their fuel in D. C. Cook Unit 2 and St. Lucie Unit 1. The analyses for both of these units demonstrated maximum values of peak clad temperature occurring in the very low exposure range. For D. C. Cook Unit 2, the peak temperature occurred at an exposure of only 2 Mwd/kg. Since all the ANF assemblies have undergone significant burnup, we did not need an F value as high as Q

that supported by the ANF analyses and have thus conservatively proposed to maintain F at a value of 1.82, which corresponds to its present limit at 48.0 Mwd/kg.

to AEP:NRC:1018 Page 5

2. Mechanical Design Considerations The analysis supporting the current peak pellet exposure of 48.0 Mwd/kg is contained in ANF report XN-NF-84-25 (P), entitled "Mechanical Design Report Supplement for D. C. Cook Unit 1 Extended Burnup Fuel Assemblies." This document was submitted directly to you by ANF with their letter JCC:113:84, dated August 21, 1984. It was referenced by us in our letter AEP:NRC:0745M, dated August 23, 1984, which proposed to increase peak pellet exposure for ANF fuel in D. C. Cook Unit 1 from 42.2 Mwd/kg to its present value of 48.0 Mwd/kg. The changes were approved by the NRC via Amendment 82 to the D. C. Cook Unit 1 T/Ss, which is dated November 29, 1984.

Attachment 4 to this letter contains an evaluation by ANF to support extending the peak pellet burnup to 48.7 Mwd/kg. This evaluation demonstrates that applicable ANF mechanical design criteria would be satisfied with a peak pellet exposure limit of 48.7 Mwd/kg.

Of these criteria, which are discussed in Attachment 4, ANF has determined that all criteria except steady-state strain, corrosion, hydrogen absorption, and fuel rod internal pressure are essentially independent of the peak pellet exposure limit. For steady-state strain, corrosion, and hydrogen absorption, ANF performed extrapolations of their analyses reported in XN-NF-84-25 (P). The results of these extrapolations, reported in Attachment 4, demonstrate significant margin to the ANF design limits. For fuel rod internal pressure, ANF performed a new analysis using their RODEX2 code. The peaking factor was increased by 2% at the maximum axial region from that used for the XN-NF-84-25 analysis to bound the increased peak pellet burnup. The results of this analysis demonstrated a peak internal pressure well below the ANF design criteria limit of 2250 psia specified in XN-NF-84-25.

Attachment 1 to AEP:NRC:1018 Page 6 Si nificant Hazards Considerations Per 10 CFR 50.92, a proposed amendment will not involve a significant hazards con'sideration if the proposed amendment does not:

(1) involve a significant increase, in the probability or consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety.

Criterion 1 We have presented analyses which demonstrate that operation up to 48.7 Hwd/kg peak'ellet exposure will not violate any applicable safety limits or design criteria. In addition, we would implement administrative controls to prohibit operation beyond 48.7 i4wd/kg unless analyses are performed using methodology that is known to be acceptable to the NRC. Therefore, we conclude that the proposed changes will not significantly increase the probability of occurrence or consequences of a previously evaluated accident, nor will they involve a significant reduction in a margin of safety.

Criterion 2 LOCA analyses and fuel mechanical design limits are the principal areas of concern regarding peak pellet exposure. We have presented evaluations which conclude that applicable criteria with regard to these issues will continue to be met for exposures up to 48.7 i4wd/kg, and have committed to not exceed that limit without analyses which use methodologies acceptable to the NRC.

Thus, we conclude that the proposed changes will not create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated.

to AEP:NRC:1018 Page 7 Criterion 3 See Criterion 1, above.

Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing certain examples (48 FR 14870) of amendments considered not likely to involve significant hazards consideration. The sixth of these examples refers to changes which may result in some increase to the probability of occurrence or consequences of a previously analyzed accident or may reduce in some way a safety margin, but the results of which are clearly within limits established as acceptable. Because these proposed changes involve extension of a limit contained in the T/Ss, they may be perceived as involving a reduction in safety margin; however, for reasons previously presented, we do not believe that any. reductions would be significant.

Attachment 4 to AEP:NRC:1018 ANF Evaluation (Proprietary) of Mechanical Design Considerations for Peak Pellet Exposures Up to 48.7 Mwd/kg

FEB 1 1 )g87 ADVANCEDNUCLEARFUELS '~r";5 5i'?AiiC'i February 10, 1987 HGS-87-055(P)

Indiana 6 Michigan Electric Company c/o Richard B. Bennett Engineer, Nuclear Materials & Fuel Mgmt.

American Electric Power Service Corp.

One Riverside Plaza, 20th Floor Columbus, OH. 43216-6631

Dear Rick:

Subject:

D. C. Cook 1 Peak Pellet Burnu Extension Attached is a summary report of the D. C. Cook Unit 1 peak pellet burnup extension analysis. This review was conducted to provide an increase in the peak pellet burnup limit from 48,000 to 48,700 MWd/MTU. The peak assembly burnup remains unchanged at 41,000 MWd/MTU. The extension of the peak pellet exposure will not result in the vi.olation of any design criteria.

Advanced Nuclear Fuels Corporation considers information contained in the enclosed technical report to be proprietary. Also enclosed is a non-proprietary version of the report. The Affidavit enclosed provides the necessary information to allow the withholding of the proprietary version from public disclosure as required by 10 CFR 2.790(b).

Very tr ly yours,

~a g.r H. G. Shaw Contract Administrator Attachment ric: M. P. Alexich J. M. Cleveland D. HE Malin V. Vanderburg

HGS-87-055 (P), Attachment 1 Page 1 of 2 DC Cook Unie 1 - Peak Pellet Burnu Extension

~Back round:

The lase reload of ANF (Eovmevly ENC) fuel supplied for the DC Cook Unit 1 reactor. is cuvven" ly in its last cycle of operation. A burnup extension analysis had been pelformed fov this Euel in 1984 in order eo support buvnup levels of 41.0. 43.7, and 48.0 GWD/MtU respectively for peak assembly. peak rod.;lnd peak pellet. Reactor operating condieions since that time have resulted in higher axial peaking than originally pvojeceed. Conseque>>ely. ehe peak pellet burnup is now expected to appvoach a level of 48.5 GWD/<kieU. The peak rod and peak assembly burnup levels are not affeceed. A review of the 'original analyses supporting the burnup extension has been conducted in order to determine the consequences of an inc vease in peak pellet exposure. The revi'ew considered an additional increase in peak pellet exposure eo 48.7 GWD/MeU to provide margin for. a potential end of cycle coastdown.

Summar Y of Buviru Ex tens ion AnalYsis Review:

The oviginal burnup extension analvsis. reported in XN-NF-84-25, Rev. 0

<,Reference 1). addvessed the following aspects of design: (1) Steady St;iee Serfs:;. (2l Steady State Strain. (3) Cladding Corrosion and Ilvdrog(>> Absovpi i<a<<. < ~) TLansient Sevess .nid Strain and Cladding

} aii<;<<(. < ) ) Ci;id(lin<. <'ve(-p Collapse. iG) Fuel Rod Internal Gas Pressure,

) I:ui I. Iio(I ("r(aw<<li. < '>> ~pa(<-v Spri<<L; Fovce. 'i!id (9) Fuel Assembly

('<)wi I1, < <,.'<>rpt L<ni. <Ild I'<<<, I I;<)(I Lne<. L'lliiI. l L<. ss<lL 6 'lre,n L,'.ilLI: LC'lntI v af feet(d by

i<( <<xi.:ll I<('<>I'i,}(')I i I<au('I L'(<d. "I'he L'(mali!(ler of the items 're IIL i.a I I y I,n(I( p<<t<(le nr of the pe'ik pe Ilet <axposuve . The resul es L'( pore<*(I Ln X.< -NF 84- .> . I4( v. 0 vema in vci lid Eov these items .

The power hiseovy us(-d Eov ehe oviginal I)uvnup extension analysis was I)used oi! u conservacive best-estimate oE the maximum discharge exposuve vo<l. assumi.iig full pow(-r opevaeion. In reality ehe operation of the veaceoL has been limited to 90 percent of Cull. powev. Thevefove, the original. power history projection represents a bounding case for ehis Eue I..

I hP L ev is<a<I cli'ia I< s Ls <h<)ws a hrl e <-'- la(l s tL'a ill, corrosion and hydvogen

<<bsovpeio<< i erne in;:i "hii< eli( des igil l.imies. and ehe fuel rod pressuve vemains 1)elow syse"n! Dvk ssuve.

HGS-87-055 (P), Attachment 1 Page 2 of 2 Steady Scate Strain. Cladding Corrosion and H dro en Absor tion:

The maximum cladding strain, corrosion and hydrogen absorption were determined to occur at the peak axial region in the original burnup extension analysis. Review of this analysis showed the results from the previous analysis co have been taken for a peak pellet exposure of 48.3 GWD/HtU. Because oE the substantial mavgin fov these design criteria a simple extrapolation was used to project the conditions for a peak pellet exposure of 48.7 GWD/HtU. Extvapolating che results of the original analysis and including an uncertainty of five percent yields the following results:

~Pro'ecLed Cviteria Total Positive Scvain. (8) [. none 1.0 j Haximum Positive Stvain Increase. (%) 0.31 1.0 ]

Cladding Corrosion. ( inch) 0.00073 0.002 j Hydvogen Absorption. (ppm) [ 85. 300. ],.

Therefore, the fuel will vemain well within the criteria for these items.

Fuel Rod Internal Pvessuve:

A new RQDEX2 (Refevence 2) anal'sis was perfovmed using the approved methodology for incevnai gas pvessuve detevmination and the bounding powev history. The axial peaking factov from the oviginal extension snllysis was inure ~sed by 2~s at the maxiamun axial region in order to bound the 1.5~ inc) ease in buvnup fvom 48.0 to ~".7 GWD/HtU. The results of this analysis showed a'eak incernaL pvessure of (1825] psia over the design l.ife of the fuel.. This value is well within the cvitevia limit of the 2250 psia reactor operating pvessuve as given in XiV-NF-84-25, Rev. 0.

==

Conclusion:==

Review oE the analysis Eov the ANF fuel supplied co che DC Cook Unit 1 reactor has shown che fuel capable oE meeting all design critevia at a peak pellet exposuve of 48. 7 GWD/HtU. The vesults pvesented in the extended buvnup vepovv XN-NF-84-25, Rev. 0 with che addition of che 1 esul ts presented in ch is Le c tea vemain @a 1.id for the fuel .

Ref: (1) XN-NF-84-.!3. Re':ision 0. Hechanical Desi n Re ort Su lement fov DC Cook Unit l. Extended Buvnu Fuel Assemblies, April 1984.

(2) XN-NF-81.-58 < P) (A') . Revision 2. RODEX2 Fuel Rod Thevmal-Hechanica1 Re::ponse Evaluation Hode1. Harch 1)'84.

DATA iN BRACKETS iS PROPRIETARY TO ADVANCED NUCLEAR FUELS CORPORATION

AFF I DAY IT STATE OF WASHINGTON )

) ss.

COUNTY OF BENTON I, H. E. Williamson being duly sworn, hereby say and depose:

l. I am Manager, Licensing and Safety Engineering, for Advanced Nuclear Fuels Corporation ("ANF"), and as such I am authorized to execute this Affidavit.

2. I am familiar with ANF's detailed document control system and policies which govern the protection and control of information.

3. I am familiar with the Letter HGS-87-55(P) entitled "DC E

Cook Unit I Peak Pellet Burnup Extension" referred to as "Document."

Information contained in this Document has been classified by ANF as proprietary in accordance with the control system and policies established by ANF for the control and protection of information.

4. The document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by ANF and not made available to the public, Based on my experience, I am aware that other companies regard information of the kind contained in the Document as proprietary and confidential.

5. The Document has been made available to the U.S. Nuclear Regulatory Commission in confidence, with the request that the information contained in the Document will not be disclosed or divulged.

6. 'he Document contains information which is vital to a competitive advantage of ANF and would be helpful to competitors of ANF when competing with ANF.

7. The information contained in the Document is considered to be proprietary by ANF because it reveals certain distinguishing aspects of PWR Fuel Design methodology which secure competitive advantage to ANF for fuel design optimization and marketability, and includes information utilized by ANF in its business which affords ANF an opportunity to obtain a competitive advantage over its competitors who do not or may not know or use the information contained in the Document.

8. The disclosure of the proprietary information contained in the Document to a competitor would permit the competitor to reduce its expenditure of money and manpower and to improve its competitive position r

by giving it extremely valuable insights into PWR Fuel Design methodology and would result in substantial harm to the competitive position of ANF.

9. The Document contains proprietary information which is held in confidence by ANF and is not available in public sources.

10. In accordance with ANF's policies governing the protection and control of information, proprietary information contained in the Document has been made available, on a limited basis, to others outside ANF only as required and under suitable agreement providing for non-disclosure and limited use of the information.

ll. ANF policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

12. This Document provides information which reveals PWR Fuel Design methodology developed by ANF over the past several years. ANF has invested thousands of dollars and several man-months of effort in developing the PWR Fuel Design methodology revealed in the Document.

Assuming a competitor had available the same background data and incentives as ANF, the competitor might, at a minimum, develop the information for the same expenditure of manpower and money as ANF.

THAT the statements made hereinabove are, to the best of my knowledge, information, and belief, truthful and complete.

FURTHER AFFIANT SAYETH NOT.

SWORN TO AND SUBSCRIBED before me this ~Sa- day of I 9 )r'rt I

NOTARY PUBLIC

Attachment 16 to AEP:NRC:0916W EXPLANATION OF STEAMLINE DIFFERENTIAL PRESSURE ENGINEERED SAFETY FEATURE ACTUATION INSTRUMENTATION LOGIC

LOOP LOOP 2 LOOP 5 LOOP g c2 Ic3 1<4 7 cl 2c3 2<< 3cJ pic 4~1 9~2. 4<3 K E I III Il H ]Jj I zr 2 2/

J I This figure represents the Steamline Differential Pressure logic. The small circles represents individual bistable signals and the enclosed roman numerals indicate the protection channels from which the signal is derived. All bistables are shown in the untripped condition.

LOOP L QOP 2 LOOP 3 LQQP g <? lc) 1c+ Z c l 2c3 2~4 3cl 3cZ 3c4 $ < ] peg. $ c$

I It E E HI IE 2 3K K XL 2 2 2/

This figure represents the Specifications are interpreted toconditions which would occur if the currently approved Technical mean bistables for the isolated loop should be placed condition. In this example, Loop l is the isolated loop and the in the tripped soon as the second bistable is tripped, the Loop l logic for a safety injection bistables are is satisfied shown tripped., As and a SI will occur.

FIG. 2

LOOP LOOP 2 LOOP 3 LOOP g <<2 l<<3 1<<+ ? <<1 2<<3 2<<k ? 3cq <<) peg.

1K I III If If W I Z Q. I 2 2/ 2/

5 This figure represents the conditions established I

by the correct interpretation of the Technical Specifications for three loop operation. This interpretation is clarified in the proposed ggg0 of Table 3.3-3 which was approved for Unit 2. Again, Loop is assumed to be the isolated loop. footnote l

indicated, only the operating loop bistable which compares the operating loop's pressure relative As to the isolated loop's pressure is placed in the tripped condition. This action reduces the Steamline Differen-tial Pressure SI logic to a one per steamline in any operating loop. This is what the Technical Specifications require.

Tripping the indicated bistables does not diminish the protection available for a steamline Should the break occur in one of the operable steamlines, the protective action will occur as soonbreak.

as the pressure of the affected steam generator falls sufficiently below that of either of the two remaining operable. steam generators. If the break be present to provide protection.

occurred in the isolated loop, the normal protective logic would There has been no compromise of the isolated loop's logic.

Attachment 17 to AEP:NRC:0916W COPY OF LETTER DATED JULY 9, 1984 FROM E. P. RAHE (WESTINGHOUSE) TO D. EISENHUT (NRC) (NS-TA-84-003)

COPY OF INDIANA AND MICHIGAN ELECTRIC COMPANY LETTER AEP:NRC:0895

NS-EPR-2935 westinghouse Water Reactor HtcM Tcctnology 0<wsion Bectric Corporation Divisions Box 355 PitTamgnPennsylvania 1523'uly 9, 1984 NS-TA-84-003 Mr. 0. Eisenhut, of Licensing Director'ivision U.S. Nuclear Regulatory Comission 2920 Norfolk Avenue Washington, D.C. 20555

Dear Mr. Eisenhut:

NUMBER OF OPERATING REACTOR .COOLANT PUMPS IN MODE 3 This letter formalizes the material presented on June 15, 1984, with respect to the consistency between the Technical Specifications and the safety analysis for the nunber of operating reactor coolant pumps in Mode 3. This meeting was held at the request of the NRC staff in order to discuss the Westinghouse deter-mination of a potential unreviewed safety question for three and four loop plants for this issue. Enclosed are ten (10) proprietary copies of the slides and ten (10) non-proprietary copies. Also enclosed are one (1) copy of Application 'for Withholding, AW-84-63 (non-proprietary) and one (1) copy of Affidavit (non-proprietary).

As part of an informal review of a utility's Tech Specs by the NRC Reactor

.Systems Branch, the staff asked what the safety analysis assunptions were con-cerning the nunber of operating reactor coolant pumps, particularly at or near zero power. Although the question was never formally asked, Westinghouse reviewed the analysis assumptions with respect to the Tech Specs.

The requirement for operating reactor coolant pumps under these conditions is contained in Specification 3.4.1.2 of the Standard Tech Specs. In non-Standard Tech Specs, the requirement is contained in Specification 3.1. These Specs state that when the plant is subcritical by the shutdown margin between 350'F (RHR cut-in) and 547'F or 557'F (no-load conditions), there must be two loops operable, but only one loop has to be actually operating.

However, the safety analysis in the FSARs assumes that either two or all of the reactor coolant pumps are operating, not just one. (At the staff's request, the assunptions made concerning the nwnber of operating pumps have been noted for those plants within Westinghouse scope in the attachment). The accidents which are limiting at zero power are steamline break, rod ejection, and bank withdy'awal from subcritical. Westinghouse has reviewed these accidents under the reduced flow conditions of one puap. For the rod ejection and steamline break events, Westinghouse has determined that the inconsistency between the safety analysis

Mr, D. Eisenhut, Oirector &2% NS-EPR-2935 and the Tech Spec will not impact the conclusions presented in the FSAR. For the bank withdrawal from subcritical event, Westinghouse has performed calcu-lations which show that the ONB design basis may not be met when only one pump is in operation. Thus, the margin of safety as defined in the basis of the Tech Specs is reduced.

Westinghouse has also performed calculations for one pump operation assuming more realistic, but still conservative, reactivity insertion rates. The results of these calculations show that the ONB design basis is met. Other assumptions and models used in these analyses are identical to the FSAR methods of analysis for this event. Thus, Westinghouse feels that no significant safety hazard exists.

Westinghouse is currently considering long term analytical solutions to this issue which will show that the ONB design basis can be met when only one reactor coolant pump is in operation so that the Tech Specs will not need to be changed.

t However, in the shor term, Westinghouse recommends that the plants be operated with the same number of reactor coolant pumps in operation as was assumed in the analysis. Note that this is not a realistic requirement when the plant is cooling down prior to going into Mode 4 (RHR operation), particularly for those plants for which the analysis assumes all pumps in operation. Thus, an alternative to having more than one pump in oper ation is to prevent rod withdrawal. This will preclude the accident from taking place. Although physical prevention of with-

~

drawal will accomplish this, administrative procedures may be preferable.

~

~ The ability to cock the rods partway out of the core during Mode 3 provides desired operating flexibility. Furthermore, there is no mechanism by which the control e ~ ~

~

rods can be automatically withdrawn in Mode 3 due to a control system error.

Increased operator awareness. during this time and adherence to procedures will also prevent the accident from occurring.

Finally, while Westinghouse feels that it is appropriate to consider bank withdrawal when in Mode 3, Westinghouse does not intend to address this event in other modes of operation (Standard Tech Spec Modes 4 and 5). Bank withdrawal from subcritical is a valid scenario when going from Mode 3 to Mode 2. However, consideration of bank withdrawal in Modes 4 and 5 is unrealistic and it is

, questionable as to whether it is applicable or if it is a Condition II event.

Again, increased operator awareness must be considered when evaluating the appropriateness of the event.

Mr. D. Eisenhut, Director & 3& NS-EPR-2935 Correspondence with respect to the Westinghouse affidavit or application for withholding should reference AW-84-63, and should be addressed to Mr . R. A. Wiesemann, Manager, Regulatory and Legislative Affairs, P.O. Box 355, Pittsburgh, Pennsylvania 15230. Other correspondence or questions should be directed to Mr. J. L. Little, Manager, Operating Plant Licensing Support, 412/374-5054.

Very truly yours, WESTINGHOUSE ELECTRIC CORPORATION E.. Rahe, Jr.

uclear Safety Department M. P. Osborne/ds Enclosures

' , OPERATING STS PLANTS NON-OPERATING D. C. Cook 1 Seabrook 1 & 2 Salem 1 & 2* Catawba 1 & 2 Beaver Valley 1* Byron/Braidwood Diablo Canyon 1 & 2 Beaver Valley 2 McGuire 1 & 2 Vogtle 1 & 2 Sumer* Millstone 3 Farley 1 & 2* Comanche Peak 1 & 2 Sequoyah 1 & 2* Watts Bar 1 & 2*

Trojan* South Texas 1 & 2 Shearon Harris 1 & 2 Marble Hill 1 & 2 NON-STS PLANTS Turkey Point 3 & 4*

Zion 1 & 2*

Indian Point 2 & 3*

Assumes al pumps operating PLANTS OUTSIDE W SCOPE D. C. Cook 2 Yankee Rowe Robinson'2 Surry 1 & 2 Haddam Neck North Anna 1 & 2

IND.IANA 8 M)CHIGAN ELECTRIC COMPANY P.O, BOX 16631 COLUMBUS, OHIO 43216 July 30, 1984 AEP:NRC:0895 Donald C. Cook Nuclear plant Unit Nos. 1 and 2 Docket Nos. 50-315 and 50-316 License Nos. DPR-58 and DPR-74 NUMBER OF REACTOR COOLANT PUMPS OPERATIONAL IN MODE 3 Mr. Harold R. Denton, Director Office of Nuclear Reaotor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Dear Mr. Denton:

By letter dated June 6y 1984, Indiana E Michigan Electrio Company (IMECo) was notified by Westinghouse Electric Corporation (g) that several Final Safety Analysis Report (FSAR) analyses performed at Hot Zero Power (HZP) assumed the operation of two (2) Reactor Coolant Pumps (RCPs). The limiting analyses at HZP, i.e., steam line break, rod egection, and bank withdrawal from subcritical conditions, are assumed to bound postulated Operational Mode 3 accidents and transients. The Donald C. Cook Nuclear Plant Unit Nos. 1 and 2 Appendix "A" Technical Specification (T/S) 3.4.1.2, however, requires that only one (1) RCP be operating during Operational Mode 3, and that at least one ( 1) additional RCP be available to meet single failure criteria.

The attachment to this letter contains a copy of the notification which we received from g. As noted in this letter, g has determined that the inconsistency between the FSAR and the T/S will not impact the FSAR conclusions for the steam line break accident and the rod e)ection transient. For the bank withdrawal from subcritical conditions transient, g calculations indicate that the departure from nucleate boiling (DNB) design basis may not be met when only one (1) RCP is running. On a best estimate basis, however, g believes that

. the DNB design basis can be met. The FSAR licensing basis analysis inoludes conservatisms (such as high reactivity insertions rates) which when removed, show that r,departure from nucleate boiling ratio] DNBR is above the limit value. Thus, no significant safety hazard exists.

We are currently preparing a proposed amendment to the T/S to deal with this situation. In the interim period until the modified T/S is approved by your staff, we have instituted a tempo! ary procedural change to ensure that plant operations are consistent. with the FSAR analysis assuaptions. That instruction requires that we operate with at least two. (2) reactor coolant pumps while in Mode 3 unless the reactor trip breakers are disconnected.

AEP: N RC:0895 Mr. Harold R. Denton We are noti ying you consistent with 10CPR50.36. This matter was disc with your staff upon notification from ge This document has been prepared following Corporate procedures which incorporate a reasonable set of controls to ensure its accuracy and completeness prior to signature by the undersigned.

M..P. Al xich >>p4 Vice President 1)

MPA/dam Attachment cc: John E. Dolan W. G. Smith, Jr. - Bridgman R. C. Callen G. Charnoff E. R. Swanson - NRC Resident Inspector, Bridgman

'I KcNÃ ~CC geg~fiyo~g. Water Reactor uneysm Dms e Bectrlc Corporatlntf

Il bet 2r26 pmsaoy pemsyt~e 15230 2128

-; Juvks-~)-, i, June 6, 1984 Mr. g. Q. Smith, Plant Manager--.""-

~

~ '~iJ U AEP"84"612 D. C. Cook Nuclear Plant ERA'(AGEItlAL Indiana and Michigan Power Company P. 0. Box 458 Brfdgman, Michigan 49106 Oear Mr. Smith:

American Electric Power Service Corporation 0, C. Cook Unit 1 CONSISTENCY BETWEEN SAFETY ANALYSIS ANO TECHNICAL SPECIFICATIONS CONCERNING NUNBER OF REACTOR COOLANT PUMPS IN OPERATION This letter fs to notify you of a potential unrevfewed safety question concerning the consistency between the safety analysis and the Technical eciffcatfons. According to 10CFR50.36, the assumptions in the safety alysfs and the plant Tech Specs must be consistent. This ensures that the ant fs operatea in a manner such that ft is bounded by the FSAR accident analysis; As part of an informal review of a utility' Tech Specs fn the NRC Reactor Systems Branch, the staff asked what the safety analysis assumptions were concerning the number of operating reactor coolant pumps, particularly at or near zero power. This information fs stated fn the FSAR for the zero power accidents. Although the question was never formally asked, Westinghouse reviewed the analysis assumption with respect to the Tech Specs.

The issue fn questfon concerns the number of operating reactor coolant pumps when in Node 3, which fs defined fn the Tech Specs as between 350'F and the no-load temperature (either 547 or 557'F). The reactor is also subcrftfcal as required by the Shutdown Margfn Spec, Standard Tech Spec 3. l. l. 1. The STS Spec number (which should correspond to your Spec number which contains the requirement for the number of operating loops fs Spec . .1. This Tech Spec

~ states that fn Node 3, there must be two loops ooerable (whfch means that the reactor coolant pump must be operable), but only one loop must be actually

~aoeratfn However, the safety analysis fn the FSAR assumes that either two or all of the reactor coolant pumps are actually operating, not just one. In the FSAR, nalyses performed at Hot Zero Power (H7P) are assumed to bound Mode 3 opera-ion. The accidents which are limiting at HZP are steamline break, rod

June 6, 1984 Page 2 ejection and bank withdrawal from subcr itfcal. Westinghouse has reviewed these accidents under the reduced flow conditions of one pump. For the rod ejectfon and steamlfne break events, Westinghouse has determined that the inconsistency between the safety analysis and the Tech Spec will not impact the conclusions presented 1n the FSAR. However, for the bank withdrawal from subcrftfca1 iccfdent, 'Astfnghouse has performed calculations which show that the QNB des1gn basis for this Condition II event may not be met when only one pump fs fn operation. Thus, the margin for safety as defined fn the basis for the Tech Specs fs reduced and this may be an unrevfewed safety quest1on according to'0CFR50.59.

Note that on a best estimate basis,. the ONB design basis can bi met. The FSAR licensing -bas1s analysis 1ncludes conservatisms (such as high reactfv1ty insertions rates) which when removed, show that the QNBR fs above the limit value. Thus, no significant safety hazard ex1sts.

Westinghouse recommends that you review your FSAR analysis for the bank withdrawal from subcrftfcal event for consistency with your Tech Specs.

Furthermore, Westfnghouse recommends that you require the number of operating pumps fn Mode 3 to be cons1stent with the analysis. Alternat1vely, you should ensure that rod withdrawal wf11 not occur when in Mode 3 if the requirement for pump operation cannot be met in Mode 3. This wfl'nsure that the safety analysfs fs consfstent wf:h plant operatfon.

If you have any questions, please contact me.

Very truly yours, l4 '.

W. ~ nson, Manager Projects Oepartment Central Area HT/387L cc: M. P. Alexfch W. G. Smith J. Waleko W

Attachment 19 to AEP:NRC:0916W REVIEW OF THE PROPOSED POWER DISTRIBUTION TECHNICAL SPECIFICATION SIMPLIFICATIONS PERFORMED BY WESTINGHOUSE ELECTRIC CORPORATION LETTER FROM WESTINGHOUSE ELECTRIC CORPORATION SUPPORTING A BURNUP INDEPENDENT F FOR WESTINGHOUSE FUEL TO AT LEAST 60 MWD/KG PEAK PELLET BURNUP

Box 3912 Westinghouse Nuclear Fuel Pir'sour'ennsrrivania 15230 3912 Electric Corporation Divisions 87AE*~0010 January 23, 1987 W-AEP/0324 M%WORDS:

Indiana and Michigan Electric Gztrpany AEP c/o Eric G. Lewis TECH-SPEC Engineer, Nuclear Materials and Fuel Management American Electric PoWer Service Corporation One Riverside Plaza, 20th Floor Columbus, OH 43215

Dear Mr. Lewis:

AMERICAN ELECZRZC POWER SERVICE CORPORATION D. C. COOK UNIT 1 TECHNICAL SPECIFICATION SIMPLIFICATION As requested by American Electric Power Service Corporation (AEPSC) in AEP-W/0151, Westinghouse has reviewed your proposed simplification of Sections 3.2.2 and 3.2.6 of the D. C. Cook Unit 1 Technical Specifications. The changes include removal of burnup dependence in the heat flux hot channel factor limit and allowable power level for EXXON fuel.

Westinghouse has found the proposed changes to be consistent with the design basis for D. C. Cook Unit 1 and the Westinghouse reload methodology.

Very truly yours,

. E. Campkkll

Project Engineer, NFD Projects NEC:mid cc: M. P. Alexich J. M. Cleveland D. H. Malin V. D. Vandeztau~

MAR 'j 19M ttuneer Fuel Oinsinn Westinghouse Water Reactor Eiectric Corporation Divisions Rex "912 P neu.igu Peeusyiveun:5290 "9I2 March 3, 1986 86AE*~0020 Indiana and Michigan Electric Co. F.-AEP/0244 c/o Joseph L. Bell Engineer, Nuclear Materials and Fuel Keywords: AEP Management Tech-Spec American Electric Power Service Corp.

One Riverside Plaza, 20th Floor Columbus, OH 43215

Dear Mr. Bell:

. AMERICAN ELECTRIC HXKR SERVICE CORPORATION D.C.. COOK UNIT 1 AEP TECH SPEC CHANGE Please find attached pages of the D.C. Cook Unit 1 Tech. Spec. which have been marked up to reflect the extension of the PQ exposure dependent limit to 60 MWD/Kg. This was informally given to you at our meeting on February 28, 1986.

As per your request, the cuIxent Tech. Spec format has been maintained with Ep(Z) = 1.0, T(El) = 1.0, and PQ (El) = 2.10 for a peak pellet exposure exterding from 0.0 to 60.0 MND/Kg.

If you have any questions, please call me.

truly yours,

.C. lier Project Engineer NFD Fuel Projects cc: M.P. Alexich J.M. Clevelard D.H. Malin - w/enc.

V.D. Vanderburg W.L. ZiIllsexmann

I I ~ I I I I I II I I II I

I e r re J s

Attachment 20 to AEP:NRC:0916W LIST OF RETRANSMITTED PROPRIETARY DOCUMENTS WHICH ARE REQUESTED BE WITHHELD

List of Resubmitted Proprietary Documents AEP:NRC:0916W Pro rietar Document Attachment Number Previous Submittal

l. AEP-D.C. Cook Unit 1 Indiana & Michigan RdF RTD Installation Letter AEP:NRC 0942D, Safety Evaluation dated August 13, 1985 August 6, 1985

2. Safety Evaluation for Indiana & Mighigan Operation Between the Letter AEP:NRC:0942D Time RTD Cross Calibra- dated August 13, 1985 tion Data is Obtained and Calibration is Updated

3. XN-NF-85-115(P) Rev. 2 15 Exxon (Now Advanced D. C. Cook Unit 1 Nuclear Fuels) Letter Limiting Break K(Z) GNW'001:87, dated LOCA/ECCS Analysis January 15, 1987

4. Advanced Nuclear Fuels 15 Indiana & Michigan Evaluation of Mechanical Letter AEP:NRC:1018 Design Considerations dated February 20, for Peak Pellet Exposures 1987 up to 48.7 MWd/kg

5. American Electric Power 18 Indiana & Michigan D. C. Cook Unit 2 RdF Letter AEP:NRC:0916I RTD Installation Safety dated March 14, 1986 Evaluation

hf

.

ML17324B291

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Dear Mr. Lewis:

Dear Mr. Bell:

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Dear Mr. Lewis:

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Dear Doug:

Dear Mr. Bennett:

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Dear Doug:

SUMMARY

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1.0 INTRODUCTION

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7.0 REFERENCES

Dear Hr. Bennett:

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Dear Mr. Stello:

Dear Mr. Bennett:

Dear Sirs:

Dear Sirs:

Dear Rick:

Subject:

Dear Mr. Eisenhut:

Dear Mr. Denton:

Dear Mr. Lewis:

Dear Mr. Bell:

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