ML17321A109

Amend 64 to License DPR-74,approving Cycle 5 Reload,Changing Surveillance Requirements for Ice Condenser Inlet Doors, Revising Containment Isolation Valve List,Correcting RCS Indicated Temp & Changing Requirements for Rod Position

ML17321A109
Person / Time
Site:	Cook
Issue date:	06/18/1984
From:	Varga S Office of Nuclear Reactor Regulation
To:
Shared Package
ML17321A110	List: ML17321A109 ML17334A529
References
NUDOCS 8406280035
Download: ML17321A109 (26)

Text

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

COOK NUCLEAR PLANT UNIT NO.

2 AMIENDMENT TO FACILITY OPERATING LICENSE Amendment No.'4 License No.

DPR-74 1.

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

A.

The application for amendment by Indiana and Michigan Electric Company (the licensee) dated March 1, 1984, as supplemented by letters dated March 5, 23, 28, April 19, May 4, 11, 17, 21, 23, June 1 and 4, 1984, and the license condition supporting letters dated September 9,

1983 and November 11, 1983, complies with the standards and requirements 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 Commission; 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 Commission's regulations and all applicable requirements have been satisfied.

2.

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:

8406280035 840618 PDR ADOCK 05000316, P

PDRi i

(2)

Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 64

, 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 before entry into the applicable mode for the Technical Specification.

4.

The license condition 2.C.3(p) is amended to read as follows:

2.C.3.(p)

"Operation during and subsequent to Cycle 5 with Exxon Nuclear Company 17x17 fuel assemblies is permitted subject to the following conditions pending receipt and approval of confirmatory and other information on transients and accidents as noted in the Safety Evaluation issued for Cycle 5:

The PTS-PWR2 model, and its adjunct thermal-hydraulic

models, cannot be used by the licensee to justify changes to the set points and related uncertainties, and instrumentation response and delay time, for Reactor Protection System (RPS) and Engineered Safeguards Features (ESF) initiation and actuation functions.

The maximum value of F ('Z) for the reactor core is to be limited to a maximum value of 2.04 irrespective of any subsequent changes to this value permitted by revisions to LOCA calculations.

iii.

No change is allowable to the current Technical Specifications with respect to moderator temperature coefficients.

In addition to the conditions set forth above, the licensee is not authorized to operate in Cycle 6, modes 1 and 2, unti 1 it has satisfactorily resolved the issues identified in the Safety Evaluation issued for Cycle 5 and other Cycle 6 regulatory requirements."

5.

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

FOR THE NUC A

EGULATORY COHl1ISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

June 18, 1984 even

. Varga, ie Operating Reactor ranch Pl Division of Licensing

t ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO.

64 FACILITY OPERATING LICENSE NO.

DPR-74 DOCKET NO. 50-316 Revised Appendix A as follows:

Remove Pa es 8-21*

82-2 3/4 2-9 3/4 2-10 3/4 2-11 3/4 2-12 3/4 2-16 3/4 4-14 3/4 5-6 3/4 6-20 3/4 6-21 3/4 6-29 3/4 6-31 3/4 6-39 3/4 6-40 3/4 10-5 8 3/4 2-1 8 3/4 2-4 8 3/4 2-4a 8 3/4 2-4b 8 3/4 2-5 Insert Pa es 8-21*

-.82-2 3/4 2-9 3/4 2-10 3/4 2-11 3/4 2-12 3/4 2-16 3/4 4-14 3/4 5-6 3/4 6-20 3/4 6-21 3/4 6-29 3/4 6-31 3/4 6-39 3/4 6-40 3/4 10-5 8 3/4 2-1 8 3/4 2-4 8 3/4 2-4a 8 3/4 2-4b 8 3/4 2-5

• Included as convenience copy only.

., ~ 'v.>>

~

~ p 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 res'ult in the release of fission products to the reactor coolant.

Overheating of the fuel cladding is prevented by restricting fuel'peration to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant satura-tion temperature.

Operation above the upper boundary of the nucleate boi ling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient.

OHB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temper-ature and Pressure have been related to ONB.

This relation has been.

developed to predict the DNB flux and the location of ONB 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 OHB at a particular core location to the local heat flux, is indicative of the margin to ONB.

The ONB design basis is as follows:

there must be at least a 95 percent probability that the minimum ONBR of the limiting rod during Condition I and II events is greater than or equal to the DNBR limit of the OHB correlation being used (the XNB correlation in this applica-tion).

The correlation ONBR limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that OHB will not occur when the minimum ONBR is at the OHBR limit.

The curves of Figures 2.1-1 and 2.1-2 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature below which the calculated DHBR is no less than the correlation ONBR limit value or the average enthalpy at the vessel exit is less than the:.

enthalpy of saturated liquid.

Uncertainties in primary system pressure, core temperature, core thermal power, primary coolant flow rate, and fuel fabrication tolerances have been included in the analyses from which Figures

2. 1<<1 and 2. 1-2 are derived.

O.

C.

COOK - UNIT 2 8 2-1 AMEHOMEHT HO.

48

a t$

N The curves are based on a nuclear.enthalpy rise hot channel factor, F

or 1.tp and a reference ooaine" with a peak or J.55 ror axial power shape.

hn allowance is included for an inorease in F< > at reduced power based

~H'n the expression:

F~ H -

1.48 [1 + 0.2 (1-P)1 (Westinghouse Fuel)

N F~H 1.49

[1 + 0.2 (1-P)]

(Exxon, Nuclear Company Fuel) where P is the fraction of RATED THERMAL POWER.

e These limiting heat flux conditions are higher than those calculated ior the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f

( h I) function of the Overtemperatur e trip.

When the axial power imbalance ik not within the tolerance, the axial power imbalance effect on the Overtemperature 6 T trips will reduce the setpoints to provide protection consistent with core safety limits.

For Exxon Nuclear Company. supplied fuel, an additional limitation on F

is applied t'o ensure compliance with ECCS acceptance criteria.

This limitation is di..ussed in basis section.3/4.2.2 and 3/4,2.3 and does not affect the safety limit curve.

2.1.2 P

0 The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and ther eby prevents the release of radionuclides contained in the r eactor coolant from reaching the containment atmosphere.

The reactor pressure vessel and pressurizer are designed to Section IIIof the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110$

(2735 psig) of design pressure.

The Reactor Coolant System piping, valves and fittings, are designed to ANSI B 31.1 1967 Edition, which permits a maximum transient pressure of 120$

(2985 psig) of component design pressure.

The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydrotested at 3107 psig, 125$ of design pressure, to demonstrate integrity prior to initial operation.

D.C.

COOK - VNIT 2 8 2-2 AMENDMENT NO, 64

3.2.3 The combination of indicated Reactor Coolant System (RCS) total flow rate and R shall be maintained within the region of allowable operation shown on Figures 3.2-4 and 3.2-5 for 4 and 3 loop operation, respectively.

For: Westinghouse Fuel N

AH R

=

1,48 [1.0 + 0.2 (1.0-P) ]

And F

A H X

1 36/P for Exxon N

for: Exxon Nuclear Company Fuel FN aH R

=

~\\

~ 0 1.49 [1.0 + 0.2 (1.0 - P) ]

Nuclear Company Fuel where:

RATED THERMAL POWER and F

A H N

= measured values of F N>

obtained by using the movable incor e detectors to ob$aik a power distribution map.

The measured values'f F

A:

and flow, without additonal

'ncertainty allowance, shall be used to compare with limits.

EKkaIUZX:

&ZIQK' With F g H above the allowable limit or with the combination of RCS total flow rate and R outside the region of acceptable operation shown on Figure 3.2-4 or 3.2-5 (as applicable):

a ~

Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />; l.

Either restore F g and the combination of RCS total flow N

rate and R to withL the above limits, or 2.

Reduce THERMAL POWER to less than 50$ of RATED THERMAL POWER and reduce the'Power Range Neutron Flux - High trip setpoint to g 55$ of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

D.C.

COOK - UNIT 2 3/4 2-9 AMENDMENT NO.6"

~XQK: (Continued) b.

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of initially being outside the above limits, verify through incore 'flux mapping and RCS total flow rate comparison that F )R and the comhination of R and ROE total flow rate are restored 6o within the above limits, or reduce THERMAL POWER to less than 5$ of RATED THERMAL POWER within the next 2

hours.

c ~

Identify and correct the cause of the out-of-limit cordition prior to increasing THERMAL POWER above the reduced THERMAL POWER Limit required by ACTION items a.2 and/or b above; subsequent POWER OPERATION may proceed provided that F )

and the combination of R and indicated RCS total flow rate are demonstrated, through incore flux mapping and RCS total flow rate comparison, to be within the region of acceptable operation as defined above for FAN and as shown on Figure 3.2-4 or 3.2-5 (as applicable) for RCS Row rate and R prior to exceeding the following THERMAL POWER levels:

1.

A nominal 50$ of RATED THERMAL POWER, 2.

A nominal 75$ of RATED THERMAL POWERR and 3.

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of attaining g 95$ of RATED THERMAL POWER.

4.2.3.1 The provisions of Specification 4.0.4 ar e not applicable.

N 4.2.3.2 F <H shall be determined to be within the above limits and the combination of indicated RCS total flow rate and R shall be determined to be within the region of acceptable operation of Figure 3.2-4 or 3.2-5 (as applicable):

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.

4.2.3e3 The RCS total flow. rate indicators shall be subJected to a CHANNEL CALIBRATION at least once per 18 months.

4.2.3.4 The RCS total flow rate shall be determined by measurement at least once per 18 months.

D.C.

COOK - UNIT 2 3/4 2-10 AMENDMENT NO,64

46

&asurenent Uncertaint)es of 3.5f.for F1m and 4%

LH d'or FHH are accounted

~or in the analysis which supports this Figure.

C) 42 ACCEPTABLE OPERATION REGION 40 CO

~'I

~ 4 UNACCEPTABLE OPERATIGN REGION

'.0,37.63 0'96,36.77~

, 34 0.90 0.94 0.98 1.02 1.06 R>FNN/1 48LI.O 0.2(1.0-P)]

WESTINGHOUSE FUEl R~FNH/1.49[1.0 0.2(1.0-P)]

EXXON NUCLEAR CO.

FU

'IGURE 3.2-4 RCS TO AL FLOMRATE YERSUS R -

FOUR LOOPS IN OPERATION 0

~

0.

C.

COOK UNIT 2 3/4 2-11 Anendment No.64

~

~

~

~

0 36 Neasurteent Uncirta)ntfes,'f 3.5% fw f'les and IS

.:.Eor faH are accounted for in the analysis which supports this Figure.

>> ~ 0

~

~

O ~

OIO>>>>

'0 0

~ ~

~

~

i ~

~

~ 0

~

~~

~

0\\>>

~ >> t ~ ~ ~ ~,Qt>>

~ >>>>

~

I O~

I Zt should be noted hat three'ooo ooeration using this curve is not, curren ly=-.

al3,owed.

.he changes contained in this table are for Reference only.

34.

32 ACCEPTASLE OPERAT}ON REGlON

~ I

~

~

~

~

~

~

~

~

8 30

~

~

~

l

~

~

~

~

UNACCEPT>>BLE 3"ERAT'.ON R G}GN

>> ~

~

a ~

~

~

~

26

,", "(1.0, 22.13).

~

=,:.. (.0.971," 26.15)i:..-::.."-

24

0. 90

0. 94 0.98 1.02

}.06 1.'10 R<FNH/}.48{},0+0.2(}.O-P)] MES

'IGHOUSE FUEL RiFNH/}.49I }.0 0.2(}.O-P) ] EXXON NUCLEAR CO.

FUEL FlGURE 3.2-5 RCS TOTAL FLOXRATE YERSUS R -

THREE LOOPS l N.OP ERAT lON I

0.

C.

COOK UklT 2 3/I 2-12 Amendment N0.64

n00 TABI,E 3. 2-1 DHB PABAMI'.TERS) tQH I.IMITS PARAMETER 3 Loo s in 0 eration'*

Reactor Coolant System Ta'g Pressurizer Pressure

~ 576-7 I'indicated)

) 2220 I)sea'70oF

> 2220 psia~

40 Limit not aPP1 icable during ei ther a TllEBHAL POWER ramP in excess of 5%

RATED TIIERHAL POWER per mi.nute or a

TIIERMAL POWER step in excess of 10%

BATED TllEBMAL POWER.

I

• ~ Indicated average of OPERABLE instrument loops.

" It should be noted that three loop operation using this curve is not currently allowed.

changes contained in this table are for Reference only.

The

3.4.6.1 The following Reactor Coolant System leakage detection systems shall be OPERABLE:

a

~

One of the containment atmosphere particulate radioactivity monitoring channels (FRS-2301 or FRS-2401),

The containment sump level ard.'o" mor'tor'".g sy"t"=, "rd C ~

Either the containment humidity monitor or one of the contairnt atmosphere gaseous radioactivity monitoring channels (ERS-2305 or ERS-2405).

MODES 1, 2,

3 and 4.

~T~Thf ~

With only two of the above required leakage detection systems

OPERABLE, operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the reouired gaseous and/or particulate radioactivity monitoring channels are inoperable; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SUl Z~5LW~CMUSSM~S.

4.4.6.1 The leakage detection systems shall be demonstrate OPERABLE by:

a

~

Containment atmosphere particulate and gaseous (if being used) monitoring system-performance of CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the fr equencies speci fied in Table

4. 3-3, b.

Containment sump level and flow monitoring system-performance of CHANNEL CALIBRATION at least once per 18 months, c ~

Containment humidity monitor (if being used) - performance of CHANNEL CALIBRATION at least once per 18 aanths.

D. C.

COOK UNXT 2 I

3/4 4-14 Amendment No. 64

EMERGEHCY CORE COOLIHG SYSTEMS SURVEILLANCE RE UIREMENTS Continued 2.

At, least once per 18 months.

Boron Injection Throttle Valves Val ve Number 1.

.2-SI-141 L1 2.'-SI-141 L2 3.

2-S I-141 L3 4.

2"SI-141 L4 Safety Injection Throttle Valves Valve Number 1.

2-SI-121 N

2.

2-SI"121 S

h.

By performing a flow balance test during shutdown following completion of modifications to the ECCS subsystem that alter the subsystem flow characteristics and verifying the following flow rates:

Boron Injection System Sinale Pumo" Loop 1 Boron Injection Flow 117.5. gpm Loop 2 Boron Injection Flow 117.5 gpm Loop 3 Boron Injection Flow 117.5 gpm Loop 4 Boron Injection Flow 117.5 gpm Safety Injection System Sin le Pumo""

Loop 1 and 4 Col d Leg Flow. > 300 gpm Loop 2 and 3 Col d Leg Flow > 300 gpm

• • Combined Loop 1,2,3 and 4 Cold Leg Flow (single pump) (640 gpm.

Total SIS (single pump) flow, including miniflow, shall not exceed 700 gom.

"The flow rate in each Boron Injection (BI) line should be adjusted to provide 117.5 gpm (nominal) flow into each loop.

Under these conditions there is zero mini-flow and 80 gpm simulated RCP seal injection line flow.

The actual flow in each BI line may deviate from the nominal so long as the difference between the highest and lowest flow is 10 gpm or less and the total flow to the four branch lines does not exceed 470 gpm.

Minimum flow (total flow) required is 345.8 gpm to the three most conservative (lowest flow) branch 1 ines.

0.

C.

COOK - UNIT 2 3/4 5-6 Amendment No. 64

TABLE 3.6-1 (Continued)

COHTAIHMEHT ISOLATION VALVES VAI.VE NUMBER FUHCTIOH ISOLATION TIME IH SECONDS h.

PllASE "h" ISOLATIOH (Cont lnued 67.

HCR-252 68.

(gM-250 69.

QCM-350 70.

QCR-300 71.

QCR-301 72.

(gR-919 73.

gCR-920 74.

PCR-140 75.

RCR-100 76.

RCR-101 77.'CR-10 78.

VCR-I1 79.

VCR-20 80.

VCR-2 1 81.

XCR-100 82.

XCR-lOl Primary Water to Pressurizer Relief Tank RCP Seal Water Discharge RCP Seal Water Discharge Letdown to Letdown llx.

Letdown to Letdown Ilx.

Demin Wtr. Supply for Refueling Cavity Demon Wtr. Supply for Refueling Cavity Containment Service Air PRZ Relief Tank to Gas Anal.

PRZ Relief Tank to Gas Anal.

Glycol Supply to Fan Cooler Glycol Supply to Fan Cooler Glycol Supply from Fan Cooler Glycol Supply from Fan Cooler Control hir to Containment Control Air to Contafnment Isolation g 10 g 15 g 15 g 10 Q )0 C 10

+10 Q 10

+10 Q 10 Q 10 Q 10 g 10 g 10 g 10

/10

a VALVE NUMBER FUNCT IOH TAB).E 3.6-1 (Continued)

CONTAINMENT ISOLATION VALVES ISOLATION TIME IN SECONDS nOOpt I

5 A.

P))ASE "A" ISOLATION (Continued) 83.

XCR-102 84.

XCR-103 B.

Pl)ASE "B" ISOLATION Control Air to Containment Isolation Control Air to Containment g 10 g 10 o

1.

CCM-451 2.

CCM-452 3.

CCM-453 4.

CCM-454 5.

CCM-458 6.

CCH-459 7.

ECR-31 8.

ECR-32 9.

ECR-33 10.

ECR-35 ll.

ECK-36 CCW from RCP Oil Coolers CCW from RCP Oil Coolers CCW from RC)'hermal Barrier CCW from RC1'hermal Barrier CCW to RC)'il Coolers

& Thermal Barrier CCW to RCP Oil Coolers

& Thermal Barrier Containment Airborne Rad Monitor Contairnnent A)rborne Rad Monitor Containment Airborne Rad Monitor Containment Airborne Rad Monitor Containment Airborne Rad Monitor g 60 g 60 g 30 g 30 g 60 g 60 g10 g 10 g 10 g 10 Q 10

TABLE 3. 6-1 (Continued).

COHTAItkMENT ISOLATION VALVES VALVE ttUMBER PUNCrlOH ISOLATION TIME IH SECONDS E.

OTHER (Continued) 10.

PA-342 Containment Service Air NA 19.

NPX-151 VI 20.

H-160 21.

SM-1 Dead Weight Calibrator N. to R.

C. Drain Tank 2

Air Particle/Radio Gas Detect Return NA 22.

N-102 23.

SI-171 24.

SI-172 25.

SI -194 26.

PW-275 27.

CS-321 N

to Accumulators Safety Injection Test Line Safety Injection Test Line Safety Injection 'l'est Line Primary Wtr. to Pre. Relief Tank R.C. S. Cliarging NA NA NA NA NA

TABI.E 3.6-1 Continued CONTAINMENT ISOLATION VALUES VALVE NUMBER FUNCTION ISOLATION TIME IN SECONDS E.

OTllER Continued 40.

PPP-300 41.

PPP-301 42.

PPP-302 43.

PPP-303 Instrument Penetration Instrument Penetration Instrument Penetration Instrument Penetration NA NA NA NA 44.

PPA-310 and PPA-311 Instrument Penetration'5.

PPA-312 and PPA-313 46.

Blind Flange 47.

Blind Flange 48.

Blind Flange 49.

Blind Flange Instrument Penetration Fuel Transfer Penetration

~ice Condenser Ice Supply Ice Condenser Ice Return In-Core Flux Thimble Acc NA NA NA NA

CONTAIlmEHT SY ICE CollOEl5B OOORS I141TIlC CQNOITIOH FOR OPERATION n

3.6.5.3 The ice condenser inlet door s, intermediate deck doors, and top deck doors shall be closed and OPERABLE.

APPLICABILITY:

MODES 1, 2, 3 and 4.

ACT?ON:

Mfth one or more ice condenser doors open or otherwise inoperable, POWDER OPERATION may continue for up to 14 days provided the ice bed temperature is monitor'ed at least once per '4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and the maximum ice bed temperature is maintained

~ 27'F: otherwise, restore Che doors to their closed posi-tions or OPERABLE status (as applicable) within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HAT STANDBY within Che next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLO SHUTOON within the followino 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLAHCE REOV IREHEHTS 4'.6.5.3.1 Inlet Ooors - Ice condenser inleC doors shall be:

a.

Continuously monitored and determined closed by the inlet door position monitoring

system, and

b. 'emonstrated OPERABLE during shutdown (MODES 5 and 6) at least once oer 9 months by:

l.

Verifying Chat the torque rcquircd to initially, open each door is c

675 inch pounds.

2.

Verifying that opening of each door is not impaired by ice, frost or debris.

3.

Testing a

sample of at least 50% of the doors and verifying that the torque required to open each door is less znan 195 inch-pounds when the door is 40 degrees open.

This torque is defined as the "door opening torque" and is equal to the nominal door torque plus a frictional toroue component.

The doors selected for determination of'the "door opening torque" shall be selected to ensure that all doors are tested at least once during two test intervals.

0.

C.

COOK - UNIT 2 3/4 6-39 Amendment tlo 64

CONTAIHHEHT SYSTEMS SURVEILLANCE REQUIREMENTS (Ccntinu d

4.

Teitfng a sample af at least 50% of the doors and verify-ingg that the torque required to keep each door from closing is greater than 78 inch-pounds when the daar is 40 degrees open.

This to~que is defined as tne "door closing torque" and is equal to the nominal door torque minus a f~ictional torque component.

The doors selected for determination of the "door closing torque" shall be selected to ensure that all doors are tested at least ance during two test intervals.

5.

Calculation of the frictional torque of each door es ed in accordance with 3 and 4, above.

Thc calculated fric-tional torque shall be 40 inch-pounds.

4.6.5.3.2 Intermediate Deck Doors -

Each icc condcnscr intermediate deck door shall be; a.

Verified closed and that onenin'g of each door is not impaired by ice, frost or debris by a visual inspection at least once ner 7 days, and b.

Oemonstr'ated OPERABLE at least once per 18 months by visually veri ying no structural deterioration, by verifying free movement of the vent assemblies, and by ascertaini ng free movement when lifted with the applicable force shown below:

Door l.

Ad.'acent to Crane Mall 2.

Paired with Door Adjacent, to Crane Mall c4% nc 3/.4

>bs 33.8

'.bs 3.

Adjacent ta Containment Mall Paired wi th Door Adjacent to Con ta i nmen t Ma 1 1 31.8 lbs.

31.0 lbs.

4.6.5.3.3 Top Deck Coors

- Each icc condenser top "eck.'oor shall be determined closed and QPERABL: at least once per 52 days ay visual'ly veri fying; 0.

C.

COOK - UNIT 2

3/4 6-40 jlmendment No.

64

3.10.5 The limitations of Specification 3.1.3.3 may be suspended during the per formance of individual full length (shutdown and control) rod drop time measur ements provided; a.

Only one shutdown or control bank-is-withdrawn from the fully inserted position at a time, and b.

The demand position indication system is OPERABLE>> during the withdrawal of the rods, and c.

The rod position indicator is OPERABLE<< during the withdrawal of the rods.

~>>/JAN~~

MODES 3, 4 and 5 during performance of rod drop time measurements.

With the rod position indicator channels or the demand position indication system not OPERABLE", or more than one bank of rods withdrawn, immediately open the reactor trip breakers.

S UiRVga,~PJCJ Jg QU~Rj',P<EN3'S 4.10.5.1 The demand position indication system shall be determined to be OPERABLE>> by verifying the demand position indication system is responsive to a rod movement demand signal during withdrawal:

4.10.5.2 The rod position indicator channels shall be determined to be OPERABLE>> by verifying the rod position indicator channels indicate rod movement during withdrawal.

<<OPERABlLITY for this Technical Specification is defined by the above Surveillance Requirements.

D.

C.

COOK - UNIT 2 3/4 10-5 Amendment No.

64

The speoifioations of th1s section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moder ate Frequency) events by: (a) maintaining the calculated DNBR in the core at or above design during normal operation and in short term tt ans1ents, 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 certain hot channel and peak1ng factors as used in these specifications are as follows:

FQ(Z)

Heat Flux Hot Channel Factor, is defined as the maximm local heat flux on the sur face of a fuel rod at core elevation Z

divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods.

N AH 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.

N The limits on F (Z) and F

for Westinghouse supplied fuel at a core average power of 3431 MWt are 1.$7 and 1.48, respectively, which assure consistency with the allowable heat generation rates developed /or a core average thermal power of 3391 MWt.

The limits on F (Z) and F

for ENC supplied fuel have been established, for g core thermal power ol 3411 MWt.

The limit on F (Z) is 2.04.

The limit on F>

is 1.36 for LOCA/ECCS analysis and 1.49 for DR3 analyses.

The analyses supporting the Exxon Nuclear Company limits are valid for an average steam generator tube plugging of up to 5$ and a

maximum plugging of one or more steam generators of up to 10$.

In establishing the limits, a plant system description, with improved accuracy was employed during the reflood portion of the LOCA Transient.

With respect to the Westinghouse supplied fuel the minimum projected excess margin of at least 10S to ECCS limits will more than offset the impact of increase steam gener ator tube plugging.

The limits on AXIAL FLUX DIFFERENCE assure that the F (Z) upper bound envelope is not exceeded during either normal operation oF in the event of xenon redistribution follow1ng power changes.

The F (Z) upper bound envelope is 1.97 times the avet age fuel rod heat flux for Westinghouse supplied fuel 'and 2.04 times the average fuel rod heat flux for Exxon Nuclear Company supplied fuel.

Target flux difference is determined at equil1brium xenon cond1tions.

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 D.C, COOK - UNIT 2 8 3/4 2-1 MENDMENT NO. 64

~ l

~

~

e The limits on heat, flux hot channel factor, RCS flowrate, and nuclear enthalpy rise hot channel factor ensure that 1) the design limits on peak local power density and minimum DNBR are not exceeded and 2)

$n the event'of a LOCA the peak fuel clad temperature will not exceed the 2200 F ECCS acceptance crater ia limit.

Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3.

This periodic surveillance is sufficient to ensure that the 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. 1.3.6.

I c.

The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained.

d.

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

H F<M will be maintained within its limits provided conditions a. through d.

above are maintained.

As noted on Figures 3.2-4 and 3.2-5, RCS flow rate and F >

may be "traded off" against one another (i.e.,

a low measured RCS flow H

ra@ is acceptable if the measured F<

is also low) to ensure that the calculated DNBR will not be below the%esign DNBR value.

The relaxation of F<M H

as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.

The form of this relaxation for DNBR limits is discussed in Section 2. 1. 1 of the basis.

An ditional limitation on F applies to Exxon Nuclear Company fuel.

H This F

< < limit, in combination witk the F~(Z) limit, ensures compliance with the ECCS acceptance criteria.

An allowance is included for an increase in F

> > at reduced power based on the following expression:

H F

g 1.36gp (Exxon Nuclear Company Fuel) where:

P is the fraction of RATED THERMAL POWER.

The power dependence of this allowance is 1/P because the associated F

H limit of 1

36 results from the LOCA analysis.

The more restrictive of the flow dependent DNBR F

> N limit and the LOCA H

F

~

limit for Exxon Nuclear Fuel Company fuel must be applied.

D.C.

COOK - VNIT' 83/4 2-4 AYENDMENT N0,64

2LiEG.: (Continued)

Figure B 3/4 2-2 illustrates the implementation of the limits as a function of power.

A measured flow will r esult in a limiting value for R which must be obtained from Figure 3.2-4 or Figure"3.2-5.

From this limiting R, a limiting F

~< H can be obtained because:

Westinghouse Fuel F

A N=1.48 X

R X L1.0+0.2(1.0-P) le Exxon Nuclear Company Fuel P

A H-1.49 X

R X L1.0+0.2(1.0>>P) ]

THERMAL POWER Whet e:

P RATED THERMAL POWER Figure B 3/4 2-2 displays two limitip DNBR F curves fear Exxon Nuclear N

Company fuel for flows of 36.77 X 10

cpm, an 37.63 X 10 gpm.

Also displayed on Figure B 3/4 2-2 is the limit op F< H which results from the LOCA analysis for Exxon Nuclear Company fue3 F

a8st be maintaiyd below and to the left of both the applicable DNBR F < H 1 mkt and the LOCA F< H U.mit.

For Westinghouse fuel there is only one F

limit. lt must be obtained N

from the applicable relationships among R,

F < H, P,

and flow.

When an F

measurement is taken, both experimental error and manufacturing 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.

When RCS flow rate and F<

are

measured, no additional allowances are N

necessary prior'o comparison Lith the limits of Specification 3.2.3.

Measurement errors of 3.5$ for RCS flow total flow rate and 4$ for F)H have been allowed for in determination of the design DNBR value and in the determination of the LOCA/ECCS limit.

D.C.

COOK - UNIT 2 B 3/4 2-4a AMENDMENT NO. 64

4 1

55 i

1.45 I

I I

l'l t"

.35 0

20 40

~

~

60 80 100 PERCENT OF RATED THERMAL POWER FIGURE B 3/4 2-2 ILLUSTRATIVE EXAMPLE OF F~H LIMIT VERSUS PERCENT THERMAL POHER FOR EXXON FUEL B 3/4 2-4b Amendment.No.

64

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.4 UADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis.

Radial power distribution measurements are made during startup testing and periodically during power operation.

The limit of 1. 02 at which corrective action is required provides DNB and linear heat generation rate pr otection with x-y plane power tilts.

The two hour time allowance for operation with a tilt condition greater than 1:02 but less than 1.09 is provided to allow identification and correction

'of a dropped or misaligned rod.

In the event such action does not correct the tilt, the margin for uncertainty on F

is 'reinstated by reducing the power by 3 percent from RATED THERMAL POWER for each per cent of tilt in excess of 1. 0.

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 adequate to maintain design DNBR throughout each analyzed transient.

"The four loop Tavg (Indicated) valve of 576.7o F i's the equivalent of 578 F less the instrument inaccuracies."

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters thru instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

3/4. 2. 6 AXIAL POWER DISTRIBUTION The limit on axial power distribution ensures that F~ will be controlled and monitored on a more exact basis through use of the APDMS when operating above APL of RATED THERMAL POWER.

This additional limitation on F

is necessary in order to provide assurance that peak clad temperatures will remain below the ECCS acceptance criteria limit of 2200'F in the event of a LOCA.

D.

C.

Cook - Unit 2 B 3/4 2-5 AMENDMENT NO. 64