ML17320A312

Amend 48 to License DPR-74,approving Cycle 4 Reload & Increasing Power Level

ML17320A312
Person / Time
Site:	Cook
Issue date:	01/14/1983
From:	Lainas G Office of Nuclear Reactor Regulation
To:
Shared Package
ML17320A313	List: ML17320A312 ML17320A315 ML17320A316
References
NUDOCS 8301240170
Download: ML17320A312 (34)

Text

gy,S Racy c>

P0 I'y I

O

'+~ ++*++

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

COOK NUCLEAR PLANT UNIT 'NO.."'

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 48 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 April 7, 1982, as supplemented by letters dated June ll and June 30, 1982, July 8, 1982, September 30,

1982, December 9 and December. 22, 1982 and January 12, 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 assur ance (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.

8301240170 8301 14 PDR ADQCK 05000316 P

PDR I

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:

~

~

(2)

Technical S ecifications The Technical Specifications contained in Appendices A and 8, as revised through Amendment No. 48, are hereby incorporated in the license.

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

3.

The license is also amended by the addition of paragraph 2.C.3 (p) to Facility Operating License No.

DPR-74 to read as follows:

"Operation during Cycle 4 with Exxon Nuclear Company 17x17 fuel assemblies is permitted subject to:

"(1) the satisfactory completion by the licensee of the following activities on or before the timesindicated:

i.

Complete and submit an analysis within one year from the issuance of this amendment using NRC approved methodology to comply with fuel assembly structural acceptance criteria in Appendix A to SRP-4.2 for the design seismic event.

ii.

Continue to comply with the operating restrictions imposed by the rod drop accident analysis until such time as the generic review of this event has been completed and any analyses required as a result of that review are performed.

iii.

Following NRC approval of the RODEX 2 thermal analysis

code, and prior to 10,000 SJD/MTU average fuel assembly burnup of the ENC 17xl7 fuel assemblies during Cycle 4 operation, resubmit the cladding strain, oxidation, and pellet/cladding interaction calculations with an approved version of the RODEX 2 code, and (2) the following conditions pending receipt and approval of confirmatory and other information on transients and accidents as noted in the Safety Evaluation'nd Environmental Impact (Report) issued with Amendment No.

48:

The PTS-PMR2 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.

ii.

The maximum value of Fg(Z) for the reactor core is to be limited to a max>mum 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 in respect of moderato@

temperature coefficients.

In addition to the conditions set forth above, the licensee is not authorized to operate in Cycle 5, modes 1

and 2, until it has satisfactorily resolved the issues identified in the Safety Evaluation and Envirormental Impact Appraisal (Report) issued with Amendment No.

48 and other Cycle 5 regulatory requirements."

4.

Within 30 days after the effective date of this amendment, or such other time as the Commission may specify, the licensee shall satisfy any applicable requirement of P.L.97-425 related to pursuing an agreement with the Secretary of Energy for the disposal of high-level radioactive waste and spent nuclear fuel.

5.

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

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

january 14, 1983 pQ's C, Lainas, Assistant Director for Operating Reactors Division of Licensing

~

~

ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. 48 TO FACILITY OPERATING LICENSE NO." DPR"74 DOCKET NO.

50-31'6 Revise Appendix A as follows:

A l-l 2-7 5 2"8 2-9 3/4 2-5 thru 3/4 2-Ba 3/4 2-9 thru 3/4 2-12 3/4 2-17 3/4 2-18 3/4 2-19 B2-1 5 B2-2 B3/4 2-1 B3/4 2-2 B3/4 2-4 2-1 5 2-2 2-3 5 2-4 P

1-1 2-7 Ei 2-8 2-9 3/4 2-5 thru 3/4 2-Ba 3/4 2-Bb 3/4 2-9 thru 3/4 2-12 3/4 2-17 3/4 2-18 3/4 2-19 B2-1 5 B2-2 B3/4 2-1 B3/4 2-2 B3/4 2-4 2-1 5 2-2 2-3 6 2-4 r

S

1. 0 DEFINITIONS DEFINED TEPMS

1. 1 The DEFINED TERMS of this section appear. in capitalized type and are applic-able throughout these Technical Specifications.

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

RATED THERMAL POWER 1.3

. RATED. THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3411 MWt.

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

ACTION 1.5 ACTION shall be those additional requirements specified as corollary state-.

ments tg each principle specification and shall be part of the specifications.

OP ERABLE -

OP ERABI LITY 1.6

.A -system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s).

Implicit in this definition shall be the assumption that all necessary attendant instru-mentation, controls, normal and emergency electrical power'ources, cooling or seal water, lubrication or other auxiliary equipment that are required for the

system, subsystem,

train, component or device to perform its function(s) are also capable of performing their related support function(s).

D.

C.

COOK " UNIT 2 AMENDMENT NO. 48

TABLE 2.2-1 (Continued REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION Overtemperature hT < dT I.KI-K2 1

(T-T )+K (P-P )-fl(hI)]

1+t S

'~

s T2 where:

hT0 Indicated hT at RATED THERMAL POWER T

Average temperature,

'F Indicated T

at RATED THERMAL POWER 57".O' avg Pressurizer

pressure, psig pr 1+el S 14v~s

& v2 2235 psig (indicated RCS nominal operating pressure)

The function generated by the lead-lag controller for T dynamic compensation avg Time constants utilized in the lead-lag controller for T.

~1

= 33 secs, avg

~2

= 4 secs.

Laplace transform operator

CD CD TABLE 2.2-1 (Continued REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION Continued M

Operation with 4 Loops Kl

= 1.267 K2

~ Q.01607 K3

-"0.000926 Operation with 3 Loops Kl

= 1.116

. K2

= 0.01607 K3

~ 0.000926 FOl CXI fTP CL B

TTP rt and f (hI) is a function of the indicated difference between top and bottom detectors of th) power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(i) for q

- q between - 40 percent and + 8 percent, fi (iil)

D (wher$ q knd q are percent RATED THERMAL PDHER in the top and bottom ha!ves of the c)lre respectively, end q

+ qb is total THERMAL PDHER in percent of RATED THERMAL PDIIER).

(ii) for each percent that the magnitucle of (q

- q) exceeds - 40 percent,'he aT trip setpoint sha!1 be automatical y reiiuced by 1.8 percent of its value at RATED THERMAL. POHER.

(iii) for each percent that the magnitude of (q

- q) exceeds

+ 3 percent, the hT trip setpoint shall be automat<cally reduced by 2.2 perceht of its value at RATED THERMAL POWER.

C7 TABLE 2.2-1 Continued REACTOR TRIP SYSTEH INSTRUHENTATION,TRIP SETPOINTS NOTATION Continued

~3S Note 2:

Overpo'eer eT c eT [K<-Ks )+

> T K< (T-T")-22(eI)]

"3'here:

hT

=

Indicated hT at rated power 0

T

=

Average temperature,

'F T"

=

. Indicated T

at RATEO THERHAL POMER

< 574.0'.F avg K4

=

1. 078 K5 K6 0.02/'F for increasing average temperature and 0 for decreasing average temperature 0.00197 for T > T"; K6 = 0 for T T"

z3S

~+

=

The function generated by the rate lag controller for T 3

dynamic compensation T3 Time constant utilized in the rate lag controller for Tav v3 = 10 secs.

avg Laplace transform operator O

Note 3:

f2(hI) =

0 for all hl The'channel's maximum trip point shall not exceed 4 percent.

its computed trip point by more than e

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 FO(Z) < ~1.97

[K(Z)]

Fq(Z) <

I.3 g4] EK(Z)]

THERMAL POWER FO(Z) < ~2.04 (K(Z))

F<(Z) < [4.08]

t:K(Z)]

p ) 0.5 p < 0.5 and K(Z) is the function obtained from Figure 3.2-2 for Westing-house fuel and Figure 3.2-2( a) for Exxon Nuclear

.Company fuel.

APPLICABILITY:

MODE 1

ACTION:

With F~(Z) exceeding its limit:

a.

Comply with either of the following ACTIONS:

2.

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

The Over power bT Trip Setpoint reduction shall be performed with the reactor in at least HOT STANDBY.

Reduce THERMAL POWER as necessary to meet the limits of Specification 3.2.6 using the APDMS with the latest incore map and updated R.

b.

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

D.

C.

COOK.- UNIT 2 3/4 2-5 AMENDMENT NO. 48

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

4.2.2.2 a.

b.

c ~

F (Z) shall be determined to be within its limit by:

Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5X of RATED THERMAL POWER.

Increasing the measured p (3) component of the power distribution map by 3X to account for liianufacturing tolerances and further increasing the value by 5X to account forpeasurement uncertainties.

This product defined is F~(Z).

Satisfying the following relationships at the time of the target flux determination.

Westin house Fuel p()(I) +

I: p j I:+li j

Fq(Z) < 13.943 t~y

]

where Exxon Nuclear Co. Fuel Fq(Z) < L-' j r.

]

P V

FM(Z)

P OB] t'K(Z)]

P >.5 P <.5 F<(Z) is the measured total peaking as a function of core height.

V(Z) is the function defined in Figure 3.2-3 which corresponds to the target

band, K(Z) is defined in Figure 3.2-2 for Westinghouse fuel and Figure 3.2-2( a) for Exxon Nuclear Co. fuel, P is the fraction of RATED THERMAL POWER.

d.

Measur ing F~(Z) in conjunction with the target flux difference and target band determination, according to the following schedule:

2.

Upon achieving equilibrium conditions after exceeding by 10% or more of RATED THERMAL POWER; the THERMAL POWFR at which F~(Z) was last determined*,

or r

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

• 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.

0.

C.

COOK - UNIT 2 3/4 2-6 AMENDMENT No.

48

POWER OISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS e.

With successive measurements indicating an increase in peak pin power, F>>, with exposure, either of the following additional actions shaW be taken.

1.

F (Z) shall be increased

.by 2% over that specified in 4.2.2.2.c, or 2.

Fg(Z) shall be measured and a target axial flux difference reestablished at least once per 7 effective full power days until 2 successive maps indicate that the peak pin

power, F<H, is not increasing.

f.

With the relationship specified in 4.2.2.2.c not being satisfied either of the following actions shall be taken.

1.

Place the core in an equilibrium condition where the limit in 4.2.2.2.c is satisfied and remeasure the target axial flux difference.'.

Comply with the requirements of Specification 3.2.2 for F (Z) exceeding its limit by the maximum percent calculated with the following expressions with V(Z) corresponding to the target band and P ).5:

max. over Z of F"(Z) x V(Z)

-1 x I'.K(Z)j 4

t F"(Z) x V(Z) max. over Z of

-1 x [K(Z)]

x 100 x 100 Westinghouse Fuel Exxon Nuclear Company Fuel 4.2.2.3 The limits specified in 4.2.2.2.c and 4.2;2.2.f above are not applicable in the following core plane regions:

1.

Lower core region 0 to 10% inclusive.

2.

Upper core region 90% to 100% inclusive.

When F~(Z) is measured for reasons other than meeting the requirements of Specification 4.2.2.2, an overall measured F0(Z) shall be obtained from a power distribution map and increased by 3% to account for manufactur ing tolerances and further increased by 5% to account for measurement uncertainty.

O.

C.

COOK - UNIT 2 3/4 2-7

.AMENDMENT NO.

48

0 1

1.2 1.0 (11.37,0.933) 0.8 12.0 0.754) 0.6 0.4 0.2 0

0 6

8 10 CORE HEIGHT (FT)~

12 C)

OCI FIGURE 3.2-2

0. C.

COOK UNIT 2, K(Z)-NORNLIZEO Fq(Z) AS A FUNCTION OF CORE HEIGHT FOR MESTINGHOUSE FUEL

1.2 C)

C3 1.0 (6.0. 1.0) 11.25 0

3

)

0.8 h4 0.6 QC C)

(12.0; 0.731) 0.4 0.2 m

rn CI 00 0.0 12 10 2

4 6

8 CORE HEIGHT (FT)

FIGURE 3.2-2(a)

0. C.

COOK UNIT 2, K(Z)

NORHALIZEQ fn (Z) AS A FUNCTION OF CORE HEIGHT FOR EXXON NUCLEAR CO.

FlJEL

1.18

~

~ j I

~

l

1. 16

1. 14

~ ~

e

~

4I

(

i ~

I

~

~

~

~

(11.25, 1.15) 1.12

+5~ Target;Band (9.25, 1.11) 11.25,1.12 1;10

+3~ parget',Band i'9.25, I;08) 1.06

~

~ r 1.04

1. 02 1.00 0

I I

a I

4 6

8 Axial Height (feet) 10

~ ~

I 12 Figure 3.2-3 V(Z) As A Function of Core Height D. C.

COOK - Unit 2 3/4 2-8(b )

AHENDMEHT NO. 48

POMER DISTRIBUTION LIMITS RCS FLOW RATE AND R LIMITING CONDITION FOR OPERATION 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.

Where:

b.

P=

Mestinghouse Fuel N

F~H

~

+

~

~

0 THERMAI POWER H

ALPW Exxon Nuclear Company Fuel N

+..

- p APPLICABILITY:

MODE 1.

ACTION:

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 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

1.

Either restore the combination of RCS total flow rate and R to within 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

<55K of RATED THERMAL POMER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b.

c ~

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 the combination of R and RCS total flow rate are restored to within the above limits, or reduce THERMAL POWER to less than 5X of RATED THERMAL POWER within the next 2

hours.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER Limit required by ACTION items a.2 and/or b above; subsequent POMER OPERATION may proceed provided that the combination of R and indicated RCS total flow rate are demonstrated, through incore flux mapping and RCS total D.

C.

COOK - ONIT 2 3/4 2-9 AMENDMENT NO.

48

POWER DISTRIBUTION LIMITS ACTION:

(Continued) flow rate comparison, to be within the region of acceptable operation shown on Figure 3.2-4 or 3.2-5 (as applicable} prior to exceeding the following THERMAL POWER levels:

l.

A nominal 50% of RATED THERMAL POWER, 2.

A nominal 75% of RATED THERMAL POWER, 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

~ 95% of RATED THERMAL POWER.

SURVEILLANCE RE UIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 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.

Where:

Westinghouse Fuel F~

N Exxon Nuclear Company Fuel NF~

~

+

~

- p

~

+

~

- p F<H

= Measured values of F~ obtained by using the movable N

N incore detectors to obtain a power distribution map.

The measured values of F~ shall be used to calculate R since N

Figures 3.2-4 and 3.2-5 include measurement uncertainties of 3.5% for flow and 4% for i~core measurement of F>.'.2.3.3 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.

48

0

Measurement Uncertainties of 3.5X for Flow and 4X for Incore Measurement of F>H are Included in this Figure.

-(1.074,43.9)

C5 42 ACCEPTABLE OPERATION REGION 38 UNACCEPTABLE OPERATION REGION (1.0,37.5)

(0.98,36.64) 34 0.90 1.10

. 1.14 0.94 0.98 1.02 1.06 R=FgH/1.48[1.0+0.2(1.0-P) j WESTINGHOUSE FUEL R=F~H/1.49f.1.0+0.2(1.0-P) ] EXXON NUCLEAR CO.

FUEL FIGURE 3.2-4 RCS TOTAL FLOWRATE VERSUS R - FOUR LOOPS IN OPERATION 0.

C.

COOK UNIT 2 3/4 2-11 Amendment No.48

38 36 Measurement Uncertainties of 3.5X for F1ow and 4C for Incore Measurement of F~>< are Inc1uded in this Figure.

CL CO 32 IcC 30 I

CD I

28 ACCEPTABLE OPERATION REGION (1.074,31,7)

UNACCEPTABLE OPERATION REGION (1.0,27.13)

(0.971,26.15 24.

0. 90 0.94 0.98 1.02

1.06 1.10

'=F~H/1.48I1.0+0.2(l.O-P)]

WESTINGHOUSE FUEL R=F~H/1.49t.1.0+0.2(1.0-P)]

EXXON NUCLEAR CO.

FUEL FIGURE 3.2-5 RCS TOTAL FLOWRATE VERSUS R - THREE LOOPS IN OPERATION 1.14 0.

C.

COOK UNIT 2 3/4 2-12 Amendment No. 4F

0 POWER DISTRIBUTION LIMITS AXIAL POWER DISTRIBUTION LIMITING CONDITION FOR OPERATION 3.2.6 The axial power distribution shall be limited by the following relationship:

Westinghouse Fuel Exxon Nuclear Company Fuel

)F (Z))

1.97 K Z

)F (Z))

2.04 K Z j

s (xj)(PL)(1.03)(l + aj)(1.07) j s

(Kj)(PL)(1.03)(1 + Qj)(1.07)

Where:

a.

F>(Z) is the normalized axial. power distribution from thimble j at core elevation Z.

b.

PL is the fraction of RATED THERMAL POWER.

c.

K(Z) is the function obtained from Figure 3.2-2 for Westinghouse Fuel and Figure 3.2-2(a) for Exxon Nuclear Company Fuel for a given core height location.

d.

%., for thimble j, is determined from at least n=6 in-core flux maps covering the full configuration of permissible rod patterns above 100% or APL (whichever is less) of-RATED THERMAL POWER in accordance with:

Where:

FMeas gi 1j max and [F,.(Z}]

is the maximum value of the 'normalized axial distribution at elevation Z from thimble j in map i which has a

measured peaking factor without uncertainties or densification allowance of F

D.

C.

COOK - UNIT 2 3/4 2-17 AMENDMENT NO.

48

POWER DISTRIBUTION LIMITS LIMITING CONDITIONS FOR OPERATION Continued

a. is the standard deviation associated with thimble j, expressed as a fraction or percentage of )(.,

and is derived from n flux maps from the relationship be'low, or

.02, (2%)'hichever is greater.

a>

1 2

1I)'2 f~T i

~gRi)3 The factor 1.07 is comprised of 1.02 and 1.05 to account for the axial power distribution instrumentation accuracy and the measure-ment uncertainty associated with F using the movable detector system respectively.

The factor 1.03 is the engineering uncertainty factor.

APPLICABILITY:

Mode 1 above the minimum percent of RATED THERMAL POWER indicated Eybb tip

. I APL min over Z of

)

x 100K, APL ~ min over Z of 100X-Westinghouse Fuel Exxon Nuclear Company Fuel v

where F~(Z) is the measured F~(Z), including a

3X manufacturing tolerance uncertainty and a

5X measurement uncertainty, at the time of target flux determination from a power distribution map using the movable incore'etectors.

V(Z) is the function defined in Figure 3.2-3 which corresponds to the target band.

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

1)

Lower core region OX to lOX inclusive.

2)

Upper core region 90K to 100K inclusive.

ACTION:

a.

With a F (Z) factor exceeding

[F.(Z)jS by

<4 percent, reduce THERMAL (OWER one percent for every percent by which th F.(Z) factor exceeds its limit within 15 minutes and within the next two hours either reduce the F.(Z) factor to within its limit or reduce THERMAL POWER to AP( or less of RATED THERMAL POWER.

e AP M

may be out o

service when surveillance for determining power distribution maps is being performed.

0-C.

COOK - UNIT 2 3/4 2-18 AMENDMENT NO. 48

POWER DISTRIBUTION LIMITS LIMITING CONDITIONS FOR OPERATION Continued b.

With a F (Z) factor exceeding [F.(Z)]

by

>4 percent, reduce THERMAL )OWER to API. or less of kATEO THERMAL POWER within 15 minutes.

SURVEILLANCE RE UIREMENTS 4.2.6.1 a.

F (Z) shall be determined to be within its limit by:

Either using the APDMS to monitor the thimbles required per Specification 3.3.3.7 at the following frequencies.

I.

At least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and 2.

Inmediately and at intervals of 10, 30, 60, 90, 120, 240 and 48Q minutes following:

a)

Increasing the THERMAL POWER above APL of RATED THERMAL POWER, or b)

Movement of control bank "D" more than an accumulated total of 5 steps in any one direction.

b.

Or using the movable incor e detectors at the following frequencies when the APDMS is inoperable:

l.

At least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and 2.

At intervals of 30, 60, 90,

120, 240 and 480 minutes following':

a)

Increasing the THERMAL POWER above APL of RATED THERMAL

POWER, or b)

Movement of control bank "D" more than an accumulated total of 5 steps in any one direction.

r 4.2.6.2 When the movable incore detectors are used to monitor F (Z), at least 2 thimbles shall be monitored and an F.(Z) accuracy equivalent to that obtained rom the APOMS shall be maintained.

C.

COOK - UNIT 2 3/4 2-19 AMENDMENT NO. 48

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'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 boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (ONB) and the resultant sharp reduction in heat transfer coefficient.

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

This relation has been developed to predict the ONB flux and the location of ONB for axially uniform and non-uniform heat flux distributions.

The local ONB heat flux ratio, ONBR, 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 ONB.

The DNB design basis is as follows:

ther e 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 ONBR limit of the ONB correlation being used (the XNB correlation in this applica-tion).

The correlation ONBR limit is est'ablished based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that ONB will not occur when the minimum ONBR is at the ONBR 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 ONBR 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 toler ances have been included in the analyses from which Figures 2.1-1 and 2.1-2 are derived.

O.

C.

COOK - UNIT 2 B 2-1 AMENDMENT NO 48

SAFETY LIMITS BASES The curves are based on a nuclear enthalpy'rise hot channel

factor, F, of 1.49 and a reference cosine with a peak of 1.]5 for axial power st pe.

An allowance is included for an increase in F

H at reduced power based on the expression:

a F<H 1.48 [1 + 0.2 (1-P)]

(Westinghouse Fuel)

N F>H 1.49 I.l + 0.2 (1-P)l (Exxon Nuclear Company Fuel) where P,is the fraction of RATED THERMAL POWER.

These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allow-able control rod insertion assuming the axial power imbalance is within the limits of the f (aI) function of the Overtemperature trip.

When the.axial power imbalance is not within the tolerance, the axial power imbalance 'effect on the Overtemperature aT trips will reduce the set-points to provide protection consistent with core safety limits.

2.1.2 REACTOR COULAfiT SYSTEM PRESSURE The restriction, of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

4 The reactor pressure vessel and= pressurizer are designed to Section III of thejASME. 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 qadi tion, 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 psi g, 125%

of design pressure, to demonstrate integrity prior to initial operation.

0.

C.

COOK - UNIT 2 B 2-2 Amendment No. 48

3/4.2 POMER 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 calculated DNBR in the core at or above design during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical proper ties 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 2200oF is not exceeded.

The definitions of certain hot channel and peaking factors as used in these specifications 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 manufacturing tolerances on fuel pellets and rods.

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.

The limits on F~(Z) and F>

for Mestinghouse supplied fuel at a core average power of 3411 NWt are 1.37 and 1.48, respectively, which assure con-sis'tency with the allowable heat generation rates developed for a core average thermal power of 3391 Wt.

The limits on F (Z) and FqH for ENC supplied fuel have been established for a core thermal poler of 342b4NWt and are 2 04 a.nd.

1.49, respectively.

3/4.2.1 AXIAL FLUX DIFFERENCE AFD)

The limits on AXIAL FLUX DIFFERENCE assure that the F~(Z) upper bound envelope is not exceeded during either normal operation or in the event of xenon'edistribution fo Ilowing'power changes.

The p (2) upper bound envelope is 1.97 times the average fuel rod heat flux for Wes inghouse supplied fuel and 2.04 times the average fue'I rod heat flux for Exxon Nuclear Company supplied fuel.

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 D.

C.

COOK UNIT 2 8 3/4 2-1 AMENDMENT NO. 48

POWER DISTRIBUTION LIMITS BASE target flux difference obtained under these conditions divided by the fraction of RATED THERNAL 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 THERMAI POWER level.

The periodic updating of the target flux difference value is necessary to reflect core burnup consi-derations.

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 AFO 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 tat get band) provided the time duration of the deviation is limited.

Accordingly, a

1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> penalty deviation limit cumulative during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is provided for operation outside of the target band but within the limits of Figure 3.2-1 while at THERMAL POWER levels above 5(C of RATED THERMAL POMER.

For THERMAL POWER levels between 15% and 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> actual time reflects this reduced significance.

Provisions for monitoring the AFO 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 AFO for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the target band and the THERNAL POMER is greater than 90K or 0.9 x API of RATED THERMAL POWER (whichever is less).

During operation at THERMAL POWER levels between 50% and 9(5 or 0.9 x APL of RATED THERMAL POWER (whichever is less) and between 15K and 50%

RATED THERMAL PQMER, the computer outputs an alarm message when the. penalty deviation accumulates beyond the limits"of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, respectively.

Figure B 3(4 2-1 shows a typical monthly target band.

The basis and methodology for establishing these 'limits is presented in topical report XN-NF-77-57, "Exxon Nuclear Power Distribution Control for PMRs - Phase II" and Supplements 1 and 2 to that report.

P.

C.

COOK - UNIT 2 B 3(4 2-2 AMENDMENT NO. 48

POWER DISTRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR RCS FLOWRATE AND NUCLEAR H

N L

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.

c.

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

L 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 ONBR are not exceeded and 2) in the even) of a LOCA the peak fuel c'lad temperature wil'I not exceed the 2200 F

ECCS acceptance criteria limit.

d.

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

F~H 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<H may be "traded off" against one anot)er (i.eee a low measured RCS flow rate is acceptable if the measured F~< is also low) to ensure that the calculated DNBR wi'll not be below thne design ONBR value.

The relaxation of Fa.

as a function of THERMAL POWER allows changes in the radia'I power shape fo all permissible rod insertion limits.

When an F

measurement is taken, both experimental error and man-ufacturing tolerance must be allowed for 55

.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~H are measured, no additional allowances N

are necessary prior to comparison with the limits of Figures 3.2-4 and 3>2-5.

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.

0.

C.

COOK UNIT 2 B3/4 2-4 AMFNOMENT NO.

48

2.0 SAFETY LIMITS AND LIMITING <."-ETY SYSTEt1 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 Figures 2.1-1 and 2.1-2 for 4 53d 3 loop operation, respectively.

APPLICABILITY:

MODES 1

and 2.

ACTION:

l 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 <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

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 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

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 - ijNIT 2 2-1 Amendment No.

48

660 UNACCEP.TABLE RATION O

UJ M I

CO 640 620

~ '00 580 ACCEPTABLE OPERATION

2400 PS ZA 250 PSIA =

2000 PSIA

. 1860 PSIA =-

U K D D O 2:

~ O UOD UJ W UJ OR I O Cl t

UJ UJ MCl UJ UJ OO O C 560

~ 2

,4

.6

.8 1.0 1.2 UJ UJ I

I FRACTIor< OF RATED THERMAL POVER Figure 2.1-1 Reactor "ore Safety Limits-Four Loops in Operation D.

C.

COOK - UNI'i 2 2~2 A~~ m.

48

660 UNACCEPTABLE OPERATION 540 2250 PSZA = 2400 PSIA o

620 2009 PSZA Z O cC 5nO 1860 PSIA O CQ I

I CL Ch LaJ 4J I

580 ACCEPTABLE OPERATION 560 0,

.2

~ 5 1.0 FRACTION OF RATEO THERMAL POWER Fiaure 2.1-2 Reactor Core Sa=ety Limit-Thre loops in Operation D.

C.

COOR - UNIT 2 2-3 NO ~

48

SAFETY LIMITS AND LIMITING SAFETY SYSTc.".; -T,'lGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor 'trip system instrumentation setpoints. shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY:

As shown for each channel in Table 3.3-1.

ACTION:

With a reactor trip system instrumentation setpoint less conservative than the value shown in the Allowable Values. column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION state-ment requirement of Specification 3.3.1.1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

D. C.

COOK - UNIT 2 2-4