Text

Amends 78 & 59 to Licenses DPR-58 & DPR-74 Respectively, Revising DHR Capability & Refueling Water Cavity Level
	ML17334A497
Person / Time
Site:	Cook
Issue date:	11/22/1983
From:	Varga S; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML17334A498	List: ML17334A497; ML17334A499;
References
	NUDOCS 8401100171
	Download: ML17334A497 (36)
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4y V/+n gO UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 INDIANA AND MICHIGAN ELECTRIC COMPANY DOCKET NO. 50-315 DONALD C COOK NUCLEAR PLANT UNIT NO.

1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 78 License No.

DPR-58 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Indiana and Michigan Electric Company (the licensee) dated August 14, 1981, as supplemented by. letter dated August 19, 1983, complies with the standards and requit emeIIts of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the COIImIission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the ComIission's regulations; D.

The issuance of this amendment will not be inimical to the coIInIon defense and security or to the health and'afety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the CoInIIission's regulations and all applicable requirements have been satisfied.

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Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment.

and paragraph 2.C.(2) of Facility Operating License No. DPR-74 is hereby amended to read as follows:

.(2)

Technical S ecifications The Technical Specifications contained in Appendices A and B, as revisgd through Amendment No. 78

, are hereby incorporate'd in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

The change in Technical Specifications is to become effective within 30 days of issuance of the amendment.

In the per iod between issuance of the amendment and the effective date of the new Technical Specifications, the licensee shall adhere to the Technical Specif-ication for the systems, components, or operation existing at the time.

The period of time during changeover of systems, components or operation shall be minimized or compensated for by suitable temporary alternatives.

4.

This license amendment is effective as of the date of its issuance.

FOR'HE NUCLEAR REGULATORY COMMISSION ven Va ief Operating Reactors ranch No.

1 Division of Licensing

Attachment:

Changes to the Techical Specifications Date of Issuance:

November 23, 1983

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASMINGTON,O. C. 20555 INDIANA AND MICHIGAN ELECTRIC COMPANY DOCKET NO. 50-316 DONALD C.

COOK NUCLEAR PLANT UNIT N0..2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 59 License No.

OPR-74 1.

The Nuclear Regulatory ComIission (the Comission) has found that:

A.

The application for amendment by Indiana and Michigan Electric Company (the.licensee) dated August 14, 1981, as supplemented by'.letter dated August 19, 1983, complies with the standards and requirements of the Atomic Energy Rct of 1954, as amended (the 'Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the CoImission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2-2.

Accordingly, the Iicense is amended by changes to the Technica1 Specifications as indicated in the attachment to this license amendment.

and paragraph 2.C.(2) of Facility Operating License No. DPR-74 is hereby amended to read as fo11ows:

(2)

Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 59

, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

The change in Technical Specifications is to become effective within 30 days of issuance of the amendment.

In the period between issuance of the amendment and the effective date of the new Technical Specifications, the licensee shall adhere to the Technical Specif-ication for the systems, components, or operation existing at the time.

The period of time during changeover of systems, components or operation shall be minimized or compensated for by suitable temporary alternatives.

4.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION even A.

arga, Chi f Operating Reactors nch No.

1 Division of Licensing

Attachment:

Changes to the Techical Specifications Date of Issuance:

November 23, 1983

ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. 78 TO FACILITY OPERATING LICENSE NO.

DPR-58 AMENDMENT NO. 59 TO FACILITY OPERATING LICENSE NO.

DPR-74 DOCKET NOS. 50-315 AND 50-316 Revise Appendix A as follows:

Remove Pa es Unit 1 3/4 4-1 3/4 4-2 3/4 4-3 3/4 9-9 3/4 9-11 3/4 9-12 8 3/4 4-1 8 3/4 9-1

. 8 3/4 9-2 8 3/4 9-3

Included for convenience Unit 2 3/4 4-1 3/4 4-2 3/4 4-3 3/4 9-7 3/4 9-8 3/4 9-10 8 3/4 4-1 8 3/4 9-1 8 3/4 9-2 8 3/4 9-3

Included for convenience Insert Pa es 3/4 4-1 3/4 4-2 3/4 4-3 3/4 4-3a 3/4 4-3b 3/4 4-3c 3/4 4-3d 3/4 9-9 3/4 9-9a 3/4 9-11 3/4 9-12" 8 3/4 4-1 8 3/4 9-1*

8 3/4 9-2 8 3/4 9-3 3/4 4-1 3/4 4-2 3/4 4-3 3/4 4-3a 3/4 4-3b 3/4 4-3c 3/4 4-3d 3/4 9-7*

3/4 9-8 3/4 9-Sa 3/4 9-10 8 3/4 4-1 8 3/4 4-la 8 3/4 9-1*

8 3/4 9-2 8 3/4 9-3

I g) 1 3/4.4 PEACTOR COOLANT S STEH 3/4..4.1 RE~CTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTU? AND ?OWER O?ERATION LIMITING CONDITION FOR.OPEPATION 3.4.i.1 A11 reactor coolant loops sha11 be in operation.

APPLICABILITY:

MODES 1 and 2.*

ACTION:

With less'than the above required reactor coolant loops in operation, be in at least HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

SURVEILLANCE REQUIREMENT 4.4;1.1 The above required reactor coo1ant loops shall be verified.o be in operai.ion and circ"la ing reactor coolant at leas once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

See Speciai Test Exception 3.10.

5 D. C.

Cook - Unit 1

3/4

='.-1 Amendment No. 78

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REAC 3R COOLANT SYSTEM SHUTDOWN LIM'I:.ING CONDITION FOR OPERATION 3.4.1.3 a.

At least two of the coolant loops listed below shall be OPERABLE:

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 one of the above coolant loops shall be in operation.*~

APPLICABILITY:

MODES 4 and 5

ACTION:

a.

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 with-in 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b.

With no coolant loop in. operation, suspend all operations in-volving a reduction in boron concentration of the Reactor Cool-ant System and immediately initiate corrective a tion to return the required coolant loop to operation.

A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures less than or equal to 188 OF unless

1) the pressurizer water volume is less than 62.00% of span or 2) the secondary water temperature of each steam generator is less than 50 oF above each of the RCS cold leg temper-atures.

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.

- "Allreactor 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

-re pe!viitted that wou!d =ause dilution of the reac.or coolant system boron concentration, and 2) core out-let temperature is maintained at least 10 'oF below satura"on temperature.

D.

C.

Cook - Unit 1

3/4 4-3 Amendment No.

78

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

4.4.1.3.2 The required reactor coolant pump(s), if not ir, operation, shall be determined to be OPERABLE once per 7 days by 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 25K 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 .

D. C.

COOK - UNIT 1

3/4 4-3a

'mendment No. 7B

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS 7HREE LOOP OPERATION N-1 LIMITING CONDITION FOR OPERATION 3.4,1.4 All reactor coolant loops shall be in operation.

APPLICABILITY:

As noted below, but excluding MODE 6.*

ACTION:

Above P-7, comply with either of the following ACTIONS:

a.

b.

With one reactor coolant loop and associated pump not in operation, STARTUP and/o'r continued POWER OPERATION may proceed provided THERMAL POWER is restricted to less than 46K of RATED THERMAL POWER and the following ESF instrumentation channels associated with the loop not in operation, are placed in their tripped condition within 1 hour:

1.

T Low-Low channel used in the coincidence circuit w)5 Steam Flow - High for Safety Injection.

2.

Steam Line Pressure - Low channel used in the coincidence circuit with Steam Flow - High for Safety Injection.

3.

Steam Flow-High Channel used for Safety Injection.

4.

Differential Pressure Between Steam Lines - High channel used for Safety Injection (trip al 1 bi stabl es which indicate low active loop steam pressure with respect to the idle loop steam pressure).

With one reactor coolant loop and associated pump not in operation, subsequent STARTUP and POWER OPERATION above 46" of RATED THERMAL POWER may proceed provided:

1..The following actions have been completed with the reactor subcritical:

a)

Reduce the overtemperature hT.trip setpoint to the value specified in Specification 2.2.1 for 3

loop operation.

See Special Test Exception 3.10.5.

D.C.

COOK - UNIT 1

3/44-3b Amendment No. 78

REACTOR COOLANT SYSTEM ACTION Continued b)

Place the following reactor'rip system and ESF instrumentation

channels, associated with the loop not in operation, in their tripped conditions:=

1)

Overpower aT channel.

2)

Overtemperature hT channel.

3)

T Low-Low channel used in the coinci-d5Qe circuit with Steam'low - High for Safety Injection.

4)

Steam Line Pressure

- Low channel used in the coincidence circuit with Steam Flow - High for Safety Injection.

5)

Steam Flow-High channel used for Safety Injection.

6)

Differential Pressure Between Steam Lines - High channel used for Safety Injection (trip all bistables which indicate low active loop steam pressure with respect to the idle loop steam pressure).

c)

Change the P-S interlock setpoint from the value specified in Table 3.3-1 to < 76K of RATED THERMAL Power.

2.

Thermal Power is restricted to

< 71>> of RATED THERMAL POWER.

O.C.

COOK - UNIT 1

3/4 4-3c Amendment No. 78

REACTOR COOLANT SYSTEM ACTION. Continued Bel ow P-7: 0 a.

Startup and Power operation below P-7 may proceed provided at least two reactor coolant loops and associated pumps are in operation.

b.

Hot standby, hot shutdown, and cold shutdown operation may proceed provided at least one reactor coolant loop in operation with an associated reactor coolant or residual heat removal pump; however, operation for up to 15 minutes with no pump in operation is permissible to accommodate transition between residual heat removal pump and reactor coolant pump operation.

c.

The 'provisions of Specifications 3.0.3 and 3.0.4're not applicable.

SURVEILLANCE REOUIREMENTS 4.4.1.4.1 With one reactor coolant loop and associated pump not in operation, at least once per 7 days determine that:

a.

The applicable reactor trip system and/or ESF actuation system instrumentation channels specified in the ACTION statements above have been placed in their tripped conditions, and b.

If P-8 interlock setpoint has been re'set for 3 loop operation, its setpoint is

< 76K of RATED THERMAL POWER.

4.4.1.4.2 Within 30 minutes prior to the start of a reactor coolant pump when any RCS. cold leg temperature is

< 188'F, verify that:

a.

The <emperature of the secondary water of each steam generator is

< 50'F above the temperature of each of the RCS cold legs, or b.

The pressurizer water volume is less than 1116 cubic feet, equivalent to less than 62 indicated on the wide range..level indicator.

A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures less than or equal to 188 F. unless

1) the pressurizer water volume is less than 1116 cubic feet (621m of span or 2) the secondary water temperature of each steam generator is less than 50 p above each of the

RCS cold leg temperatures.

/

D.C.

COOK - UNIT 1

3/4 4-3d Amendment No. 78

REFUELING OPERATIONS 3/4. 9. 8 RESIDUAL HEAT REMO'r'AL AND COOLANT CIRCULATION LIMITING CONDITION FOR OP~:";ATION 3.9.8.1 At least one residual heat removal loop shall be in operation.

APPLICABILITY:

MODE 6.

ACTION:

a.

b.

c ~

With less than one residual heat removal loop in operation, except as provided in b. below, suspend all operations involving an incr ase in the reactor decay heat load or a reduction in boron concontration of the Reactor Coolant System.

Close all containment penetrations providing direct access from the contain;;;ent atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The residual heat removal loop may be removed from operation for up to l.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 th vicinity of the reactor pressure vessel hot legs.

The provisions of Specification 3.0..3 are not applicabIe.

SURVETLLANCE REOUIR"-MENTS 4.9.8.1 A residual heat removal loop shall be determined to be in operation and circulating reactor coolan at a flow rate of

> 3000 gpm at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

0.

C.

COOK - UNIT 1

3/4 9-9 Amendment No. 78

REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONG.'TICN FOR OPERATION 3.9.8.2 Two independent Residual Heat Removal (RHR) loops shall be OPERABLE.*

APPLICABILITY:

MODE 6 when the water level above the top of the reactor pressure vessel flange is less than 23 feet.

ACTION:

a.

With less than the required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible.

b.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.8.2

,he required Residual

Heat, Removal loops shall be determined OPERABLE per Specification 4.0.5.

The normal or emergency power source may be inoperable for each RHR loop.

0.

C.

COOK - UNIT 1

3/4 9-9a Amendment No.78

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'ErUELENG OPERATIONS f

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3/4.9r 10 MAT=R LEVEL -

REACTQR VESSEL LIHITING CQhQITICN

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3.9.10 At le st 23 feet of water shall be main ained over he top of the reactor pressure vessel flange.

APPLICABILITY: During movement of fuel assemblies or control rods within the reactor pressure vessel while in MODE S, ACTION:

f l

! kith the requirements of the above specifica ion nc. sa isfied, suspend all operations involving movement of fuel assemblies or control rods within tho pressure vessel.

The provisions of Specification 3.0.3 are not, applicable.

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SURVEILLANCE "EQUIRE.'tENTS 4.9.10 The water level shall be determined 'o be at least",.s min, m m

required d pth within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to tf e star'f an@ ai lees: cr" pe.

24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s

.",crea f ar duri"" -ove.,ent of fuel asse.,bl ies or cont ol

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C.

Cook - Unit 1

f 3/4 9-1'.

.Amendment No. 78

REFUELING OPERATIONS STORAGE POOL WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.

APPLICABILITY: Whenever irradiated fuel assemblies are in the storage pool.

ACTION'ith the requirements of the specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore water level to within its limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The provision of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.11 The water level in the storage pool shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the fuel storage pool.

D. C.

COOK - UNIT 1

3/4 9-12

3/4.4 REACTOR COOLANT SYSTEM eASES 3/>>".4.

1 REACTOR COOLANT LOOPS The plant is designed to operate with all reac cr coolant loccs in opera ion, and maintain ONBR above 1.30 during all normal operations and anticipated transients.

With one reactor coolant loop not in operation, THERMAL POWER is restricted to 51 percent of RATED THERMAL POWER until the Overtempera+ure AT trip is reset.

Either action ensures that the DNBR will be maintained above 1.30.

A loss of flow in two loops will cause a reac or trip if operating aoove P"7 (11 percent of RATED THERMAL POWER) while a loss of flow in one loop will cause a'eactor trip if operating abcve P-B (Sl percent cf 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.

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 cn star ing a

Reac or Coolant Pu;,." below P-7 wi.h one or

ore RCS co: o legs less than or equal tc 'BB F are prcvice" tc preven-

'RC~

pr essure ransients, caused bv energoy addi ions from the seconaary

svstem, which could exceed

.he limits of Apper dix G to 10 CFR Par+

50.

The RCS wi 1'.

be pT cte=ted against cverpressure transients an" vill nct exceed the limits c; A-.penc.x G bv either (:) restrict..".c the water vc'u.-e in the cress"-izer

=-:-."

therebv provicing a volume for tne primary coola.";: to expand

'.n:o c" (2) res.ricting star-ing o, the RCPs to wher. the secondary water temperat'e of eacsh steam gene". ator is less than 50 F above eacn of the RCS cold ;eg temperatures.

3/'. '.. 2 aac 3/a.a. 3 SAFETY VALVE"-

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e scare y v orcv',aes s'Cvvs is des',cnec

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RCS, essurizaticn.

pressurize.

cede safetv valves. coerate to prevent he RCS zed above i+s Sa e.y Limit of 2735 ps-'g.

Each sa=ety valve ve -'2",OOO lus ",er hear c.

sacer ad

-~a:

'O valve se=

ere caoac'.ty c

a lllcle sa.~~e 5 aoec"a e 'c relieve ccnditison whicn could occur our;n" shu.cown.

In the eve'-

alves are

OPERABLE, an cperat1ng RHR lccp. ccnnectea tc t.".e overpressure relisaf caoability anc will pre:en RCS ever"r D.

C.

COOK - UNIT 1

B 3/4 4-1 Amendment No. 78

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 limi-tations are consistent with the initial conditions assumed for the boron dilution incident in the accident analyses.

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 assump-tions 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 con-tainment will be restricted from leakage to the environment.

The OPERA-BILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of con-tainment 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

REFUELING OPERATIONS BASES 3/4.9. 6 MANIPULATOR CRANE OPERAB IL'.TY

'he OPERABILITY requirements for the manipulator cranes ensure that:

1) manipulator cranes will be used for movement or control rods and fuel assemblies

2) each crane has sufficient load capacity to lift a control rod or fuel assembly and 3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged dur-ing lifting operations.

3/4.9.7 CRANE TRAVEL - SPENT'UEL STORAGE BUILDING The restriction on movement of loads in excess of the nominal weight of a fuel assembly over other fuel assemblies ensures that no more than the contents of one fuel assembly will be ruptured in the event of a fuel handling accident.

This assumption is consistent with the activity release assumed in the accident analyses.

3 4.9.8'ESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirementthat at least one residual heat removal (RHR) loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain'he water in the reactor pressure vessel below 140oF as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor cor to minimize the effect of a boron dilution incident and prevent boron stratification.

The requirement to have two RHR loops OPEPABLE when the. e is less than 23 feei: of water above the reactor pressure vessel flange ensures that a

single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability.

Nith the reactor vessel head

. emoved and 23 feet of water above the reac.or pr ssure vessel

flange, a large neat sink is available for core cooling.

Thus, in the event of a railure of the operating RHR loop, adequate t;me is prov',ded to initiate emergency procedures to cool the core.

3/4.9.9 CONTAINMENT PURGE ANDi EXHAUST ISOLATION SYSTEM The OPERABILITY of this system ensures that the containment vent and purge penetrations will be automatically isolated upon detection of high radiation levels within the containment.

The OPERABILITY of thiis system is required to restrict the release of radioactive material from the contain-ment atmosphere to the envirorment.

D. C.

COOK-UNIT 1

B 3/4 9-2 Amendment No.

I I

I BASES 3/8, 9, 1 C

. d 3/4 9, 1 1

(eT R

L V L REACTOR I SSEL AsiD S l O.,."'.Gi PCOL The restrictions on minimum water level ensure thai sufficient water depth is available to remove 99,. of ine assumed 10'! iodine cap aciivity released from the rupture of an irradiated

.uel assembly.

lhe minimum water depth is consis ent with the assumptions of tne acciaent analysis.

Water level above the vessel flange in HOOE 6 will vary as the reactor vessel head and the sys em internal"=

re removed.

The 23 feet of wat are required before any subsequent movement fuel assemblies or control rods.

3/4.9.12 STORAGE POOL V 'lTILATIOH SYST:-."~

The limitations on the stcrace pool veniilaiion system ensure tha all radioactive material released rrom an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere.

The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assumptions of the accident analyses.

3/4.9.13 SPENT FUEL CASK i)0'I:-!E'EN The limitations of this speci.ica

sertion or removal of spent

= el casks ll fuel cask movement.vill ve constraine

'assumed in the Cask Drop ?roi ct-ion Svs

,:in" the spent fuel cas'

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... nt <<i=nin plotec'tlon fol

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'ael oool ano 07 a

s ue, casi urcp ac>>

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tion ensures

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t nese rec" I storea

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"'ur ing ir,-

1t fuel "col, 1= lysis.

c-::;e.

I The limitations on the use of spent fuel casks

<<eign n" in excess of 110 tons (nominal) provid s assurance that ihe soent

-.ue.

coo'.

would not be damaged by a orooped fuel casv. since t1is wei-.".= is consistent with the assumotions used in the safety analys->

-,or the.

l performance of the Cask Drop Protect',on System.

ll 0.

C. C"C"-U"IT 1 I

i 8 3/4 9-3

ig; dt".cnt Ho. 73

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REACTOR

".40LAt'(T SYSTEM

~SUTOOMN LIMITING CONDITION FOR OPERATION 3.4.1. 3 a.

At least two of the coolant loops lis.ed below shal':

be OPERABLE:

1.

Reactor Coolant Loop 1

and its associated steam generator and reactor coolant pump,*

2.

3.

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

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

generator and generator and 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 one of the above cooiant loops shall be in operation.'*~

APPLICABILITY:

iMODES 4 and 5

AiTION:

a.

With less than ttfe above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; be in COLD SHUTDOWN with-in 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b.

With no coolant loop in ooeration, suspend all operations in-volving a reduction in boron concentration of the R=actor Cool-an System and isemediately initiate corrective act:or. 'o return the required coolant loop to operation.

A reactor coolant pump shall not be starred with one or more of the PCS coId leg t mperatures less than or equal to 152 oF unless

1) the pressu~izer uat:-r volume is less than 62.00% of span or 2) the secondary water t mperature of each steam generator is less than SO oF above each of the RCS cold leg t sq:er-atures.

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 NODE 5.

-*All reactor coolant pumos and residual heat removal pumps may be de-energized

-or uo to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided 1} no operations are permit.ed :hat "ci ld cause dilution of the reactor coolant system boron concentration, ano 2) c"re out-let temperature is maintained at least sO oF below saturation

=:-;.pera=.re.

D. C.

Cook - Univ 2

3/4 4-3 Amendment No.

REACTOR COOLANT SYSTEM SURVE ILLA.'<CE RE UIREt/ENTS 4.4.1.3.1 The required residual heat removal loop(s) shall be determined OPER"BLE per Specification 4.0.5.

4.4.1.3.2 The required reactor coolant pump(s), if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignmen s

and indicated power availability.

4.4.1.3.3 The required steam generator(s) shall be determined OPERABLE by verifying secordary 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 op'eration and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

0 C

COOK UNIT 2 3/4 4-3a Amendment No. 5~

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.

1 REACTOR COOLANT LOOPS LIMIT'ING CONDITION FOR OPERATION

'.4.ll reactor coolant loops shall be'in operation.

APPLICABILITY:

As noted below, but excluding MODE 6."

ACTION:

Above P-7, comply with either of the following ACTIONS:

a.

With one reactor coolant loop and associated pump not in operation, STARTUP and/or continued POWER OPERATION may

'proceed provided THERMAL'OWER is restricted to less than 31K of RATED THERMAL POWER and the following ESF instrumentation channels associated with the loop not in operation, are placed in their tripped condition within 1 hour:

1.

T Low-Low channel used in the coincidence circuit wf9 Steam Flow - High for Safety Injection.

2.

Steam Line Pressure - Low channel used in the'coincidence circuit with Steam Flow - High for Safety Injection..

3.

Steam Flow-High Channel used for Safety Injection.

4.

Differential Pressure Between Steam Lines - High channel used for Safety Injection (trip all bistables which indicate low active loop steam pressure with respect to the idle loop steam pressure).

b.

With one reactor coolant loop and associated pump not in operation; subsequent STARTUP and POWER OPERATION above 32K of RATED THERMAL POWER may proceed provided:

1.

The following actions have been completed with the reactor in at least HOT STANDBY:

a)

Reduce the overtemperature hT trip setpoint to the value specified in Specification 2.2.l for 3 loop operation.

See Special Test xception 3. 10. 4.

D.C.

COOK - UNIT 2 Amendment No.69

REACTOR COOLANT SYSTEM ACTION Continued 2.

b)

Place the following reactor trip system and ESF instrumentation

channels, associated with the loop not in operation, in their tripped conditions:

1)

Overpower aT channel.

2),';Overtemper'ature LT channel.

3)

-- Low-Low channel used in the coinci-dMe circuit with Steam flow - High'or Safety Injectiori.

4)

Steam Line Pressure - Low channel used in the coincidence circuit with Steam Flow - High for Safety Injection.

5)

Steam Flow-High channel used. for Safety Injection.

. 6),

Oiffe'rential-Pressure Between Steam Lines - High:

channeT used for Safety Injection (trip all Mstables which incHcate Tow active loop. steam pressure with respect to the idle loop steam pressure).

r,

, ~

c)

Change the P-'8 interlock setpoint from the value specified in Table 3.3-1 to

< 76">> of RATED THERMAL POWER.

THERMAL POWER is restricted to ~ 71" of RATED THERMAL POWER.

Below P-7:

a.

STARTUP and POWER o"erat-Ion may proceed provided ac

-:-s -.

.A.

reactor coolant '.oops and associated pumps a,"

b.

HOT STANDBY, HOT SHUTDOWN, and COLD SHUTDOWN oper ation may proceed provided at least one reactor coolant loop is in operation with an assoc',ated reactor coolant or res-'dual heat

.remova',

pump.+

c.

The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

AMENDMENT NO. 59 D.C.

COOK - UNIT 2

" All reactor coolant pumps and residuaT 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 no operations. are permitted which could cause dilutiorr of the reactor coolant system boron concentration.

j 3/4

~

REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS 4.4.1.4.i With one reactor coolant loop and associated pump not in operation, at least once per 7 days determine that:

a.

The applicable reactor trip system and/or ESF actuation system instrumentation channels specified ip "the ACTION statements above have been placed in their tripped conditions, and b.

If the 'P-8 interlock setpoint has been reset for 3 loop operation, its setpoint is

< 76X of RATED THERMAL POWER.

Within 30 minutes prior to the start of a reactor coolant pump when any rcGS cold leg temperature is

< 152'F, verify that:

a.

b.

The temperature of the secondary water of each steam generator 'is

< 50'F above the 'temperature of each of the RCS cold legs, or The pressurizer water volume is less than 1116 cubic feet, equivalent to less than 62 indicated en the wide range level indicator.

0.

C.

COOK - UNIT 2 3/4 4~3d Amendment No.59

REFUELING OPERATIONS CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING>>

LIMITING CONDITION FOR OPERATION 3.9.7 Loads in excess of, 2,500 pounds shall be prohibited from travel over fuel assemblies in the storage pool.

Loads carried over the spent fuel pool and the heights at which they may be carried over racks containing fuel shall be limited in such a way as to preclude impact energies over 24,240 in.-lbs., if the loads are dropped from the 'crane.

APPLICABILITY: With fuel assemblies in the storage pool.

ACTION:

With the requirements of the above specification not satisfied, place the crane load in a.safe condition.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.7.1 Crane interlocks and physical stops which prevent crane travel with loads in excess of 2,500 pounds over fuel assemblies shall be demonstrated OPERABLE within 7 days prior to crane use and at least once per 7 days thereafter during crane operation.

4.9.7.2 The potential impact energy due to dropping the crane's load shall be determined to be ( 24,240 in.-lbs. prior to moving each load oyer racks containing fuel.

"Shared system with D.

C.

COOK - UNIT 1

D.

C.

COOK - UNIT 2 3/4 9-7

REFUELING OPERATIONS 3/4.9. 8 RES IOL'AL HEAT REt10VAL AND COOLANT CIRCULATION LIMITING CONOIITICN FOR OPERATIO,'l 3.9.8.1 At least one residual heat removal loop shall be in operation.

APPLICABILITY:

MOO 6.

ACTIO!l:

a.

c ~

)hth less than one residual heat r moval loop in operation, except as provided in b.

beloved, suspend all opera ions involving an increase in the reactor decay heat load or a reduction in boron conc ntration of the Reactor Coolant System.

Close all containm nt penetrations providirg direct access from the contain: ent atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The residual heat removal loop may be removed from operation for up to l.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 th vicinity of the reactor pressure vessel hot legs.

The provisions of Specification 3.0.3 are not applicable.

p$ /f7)

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4.9.8.1 A residual heat removal loop shall be determined to be in opera"cn-and circulating r"actor coolant at a flo v rate or

> 3000 gpm,'at leas once pe:

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3/4 9-S Amendment No. 59

REFUELING OPERATIONS

'3s'I HATER LE~/EL NITING CONDITION FOR OPERATION

.9.8.2 Two independent Residual Heat Removal (RHR) loops sha.l be OPERABLE.+

.-',PPLICABILITY:

t<ODE 6 when the water level above the top of the re ctor pressure vessel flange is less than 23 feet.

ACTION:

a.

>hth less than the required QHR loops OPEQABLE, immediately initia e corrective action to r turn the required RHR loops to OPERABLE status as. soon as possible.

b.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREflENTS

',.9.8.2 The required Residual Heat Removal loops shall be determired OPERABLE "er Specification 4.0.5.

The normal or emergency power sourc may be inoperable for each RHR loop.

D.

C.

COOK - UflIT 2 3/4 9-8a Amendment No.

59

REFUELIHG OPERATIONS 3/4.9.10 MATER LEVEL -

REACTOR VESSEL LIMITING CONDITION FOR OPERATIOH 3.9.10 At least 23 feet of water shall be maintained over the top of the reactor pressure vessel flange.

APPLICABILITY: During movement of fuel assemblies or control rods within the reactor pressure vessel while in MODE 6.

ACTION:

lith the requirements of the above specification not satisfied, suspend all operations involving movement of fuel assemblies or control rods within the pressure vessel.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREHENTS 4.9. 10 The water level shall be determineC to be at least its minimum required depth within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to 'ihe start of and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter Curing movement of fuel assemblies or control rods.

'D. C.

COOK - UNIT 2 3/4 9-;0 Amendment No. 59

3/-'.4 REACTOR COOLANT SYSTEM BAS=S 3/4.4.2 and 3/4.4.3 SAFETY 'lALVES 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.

Ouring 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'f the power operated relief valves or steam dump valves.

0. ".

CCOK - Jl)IT 2 B 3/4 4-la Amendment No. 59

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 limi-tations are consistent with the initial conditions assumed for the boron dilution incident in the accident analyses.

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 assump-tions 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 con-tainment will be restricted from leakage to the environment.

The OPERA-BILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of con-tainment 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

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 calculated DNBR above the design DNBR value during Condition I, and II events.

With one reactor coolant loop not in operation, THERMAL POWER is restricted to 51 percent of RATED THERMAL POWER until the Overtemperature 4T trip is reset.

Either action ensures that the calculated DNBR will be maintained above the design DNBR value.

A loss of flow in two loops will cause a reactor trip if'perating above P"7 (ll percent of RATED THERMAL POWER) while a loss of flow in one loop will cause a reactor trip if aoerating above P"8 (31 percent of RATED THERMAL POWER).

In MODE 3, a s;ngle re c.or coolant loop pr'ovides suffic',en'eat removal capability for removing decay heat; however, single iailure const-~rations require that two loops be OPERABLE.

In MODES 4 and 5,

a. sinale reactor coolant loop or RHR loop provides sufficient heat removal capabi1 ity =or removing decay hea.;

but sz nal e i ai lure

.considerations require that. at leas two loops be OPERABLE.

Inus if the reac or coolant loops are not 0?ERABLE, i>>his specification requires two RHR loops to be OPERABLE.

~

I The operation of one Reactor Coolant Pump or one RHR pump prides adeuate Floe to ensure mixing, pr vent stratiiic tion and oroduce gradual reactivity chang s during boron conc ntration reductions in.he Reactor Coolant Sys am.

he reactiivity ch nae ra:e.associated with boron reduction will, herefcre, be w thin t>>he capabil ity 'of operator recoc.".i ion and control.

The restrict-.:,.s on st'-.

-.-ng a Reac or Coolant

?ump w;-~ oe or more RCS cold lhgs 1ess th-n or equal to 152 o,

a-.e provide to -revent;..S i

oi ace iry +i =ni:. =n

=iig-r 'i - -ri i =r.i i.. ~e -

g q

i<<

r

~ ~

i>>J ess>>

~

i

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~ <<4 i'

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i QQ I I Iv Is i

~ <<

~

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~,s) killch co>> Ic x>>>>

o i use I ii~i3 i o oi

~.pp nc i.c 0

<<0 U

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o 0

a'>>>> i il be protec:

d aca>,".s ov=-.".ressur

-.r=-.",s n:s ana i i. i no..:-x=:

'ppendix G by either (I) restricting the wat r vo'~ume in the Pre sur;z=.

thereby providina a volu;e ior the primary coolant 0 expand in.o, or (") ".y restric ing st ing of " e "C?s '0 when the secc,.dary w ter each st am aenera "or is less than 5O above each of the RCS cold 1 ea tempera tures.

D.

C.

COOK " UNIT 2 B 3/4 4-1 Amendment No. >9 I~ <

')<<(<<'>>

~ '.'".'>>'9'~"

~ '

~ "'

'X+ 'i ~ c.'>> i ii>>>>ag'>> <<i <<:

>> i ','

o>>'~ ~ >>

~>>

~ '

REFUEL ING OPERATIONS BASES 3/4.9.6 i4IANIPULATOR CRANE OPERABILITY The OPERABILITY requirements for the manipulator cranes ensure that:

1) manipulator cranes will be used for movement of control rods and fuel assemblies,

2) each crane has sufficient load capacity to lift a control rod or fuel assembly, and 3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING The restriction on movement of loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel assemblies in the storage pool ensures that in the event this load is dropped

1) the activity release will be limited to that contained in a single fuel assembly, and 2) any possible distortion of fuel in the storage rocks will not result in a critical array.

This assumption is consistent with the activity release assumed in the accident analyses.

3/4.9.8 RESIDUAL HEAT RB10VAL AND COOLANT CIRCULATION The requirement tha. at least one residual heat removal (RHR) loop be in operation ensures that (1) sufficient cooling capacity is availab'ie to remove decay heat and maintain the water in the reactor press re vessel below 140oF as required during the,R EFUELING RODE, and

',2) sufficient coolant ci.-

culation is.;:aintained th oug'.". the reactor core to minimize the effect of a boron dilution incident and pr vent boron stratification.

The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor pressure vessel flange ensures that a

single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability.

Mith the reactor vessel head removed and 23 feet of water above the reactor pressure vessel

flange, a large heat sink is available for core cooling.

Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency proce-dures to cool the core.

D. C.

COOK - UN:.T 2 B 3/4 9-2 Amendment No. 59

REFUELING OP~" TICNS BASES 3/4, 9 ONTHIl~Ji'lE.li PURGE AND:('r.AUS I OLA i I N

YSTu<

The OPERABILITY of this system ensures that the containment vent and purge penetrations will be automatically isolated upon detection of high radiation levels within the containment.

The OPERABILI Y of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment.

3/4.9.10 and 3/4.9.11 MATER LE".

L - REACTOR >/ESSEL AND STOPPAGE POOL The restrictions on minimum water level ensure that suf icient uater depth is available to remove 99'.l of the assumed lC'.l iodine gap activity released from the rupture of an irradiated fuel assemble.

The minimum water deoth is consistent with the assumptions of the accident analysis.

Mater level above the vessel flanae

-'n

'<GDE

-". ~; '=~

reactor vessel head and the upper intervals are removed.

The 23 feet of water are required before -ny subsequent movement of fuel 3/4. 9.12 STORAGc'.POOL 'IENTILATION SYSTB<

The limitations on the storage pool ventilation system ersure that all radioactive ma erial released

-.rom an irradiated fuel assembly w.'ll be filtered through the HEPA filters and charcoal adsorber prior to dis-charge to the atmosphere.

The CPERABILITY of this system and the r suit-ing iodine removal capacity are consistent with the assumptions of the accident analyses.

SP N

'-"'ASK."CV=".".EN; The 1 imitations oi this s."ecifica".ion ensures t..a;, ",nc sertion or removal of soent fue casks from.he s"ent

=-:

= c.,

fuel cask movement wili '."e'constrainea to the ".atn anc 1;ft;;el=.-.t assumed in the Cask Drop 'Protection System saf2ty ana ysis.

Re=tr c".-

ing the spent fuel cask movement, within these reou'.rements prov'.des protec.ion or the spent fuel pool and stored uel from the effects of a fuel cask drop accident, 3/4.9.14 SPENT rUEE, CASK DROP aROTECTION SYSTE!l The limita~ions on t~h se o

spent

.uel casKs weighing in ex ess of lip "ons ~nomina-~~rovides assurance

.hat "he s"ent fuel =cc, wouVd not be damaged

."y a~ dropped fuel cask since this weight is

.i cons',stent w'.th

-:-h2 ssump ions us2d in :..2 s

" =n..>ys's ""r performance "f the Cask Drop Pr".tectio.l Syst 0.

C.

COCK - UNIT.2

/493 Amendment No.

59

ML17334A497

Text

Attachment:

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