ML17325B246

Amend 126 to License DPR-58,revising Tech Specs to Allow Operation of Future Reload Cycles at Reduced Primary Coolant Sys Temp & Pressure Conditions to Decrease Steam Generator U-tube Stress Corrosion Cracking Observed in Unit 2

ML17325B246
Person / Time
Site:	Cook
Issue date:	06/09/1989
From:	Yandell L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML17325B247	List: ML17325B246 ML17325B248
References
NUDOCS 8906220285
Download: ML17325B246 (34)

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UNITEDSTATES NUCLEAR REGULATORYCOMMISSION WASHINGTON, D. C. 20555 INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50"315 DONALD C.

'COOK NUCLEAR PLANT UNIT NO.

1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

126 License No.

DPR-58 l.

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

A.

The application for amendment by Indiana Michigan Power Company (the licensee) dated October 14, 1988 as supplemented December 30,

1988, and June 5, 1989, 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 reghlations 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'egulations;

~

4 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 5l of the Commission's regulations and all applicable requirements have been satisfied.

8906220285 890609 PDR ADOCK 050003l5 p

pNU

C

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-58 is hereby amended to read as follows:

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

Z26", are hereby incorporated in the license.

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

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date'f Issuance:

Junc 9.', Z9.89.

Lawrence A. Yandell, Acting Director Project Directorate III-1 Division of Reactor Projects III, IV, V 8 Special Projects Office of Nuclear Reactor Regulation

t E

ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. 126TO FACILITY OPERATING LICENSE NO.

DPR-58 DOCKET NO. 50-315 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.

The revised pages are identified by amendment number and contain marginal lines indicating the area of change.

REMOVE 1-7 1-10 2-2 2-5 2-7

'2-8

'-9 B 2-1(a)

B 2-2 B 2-4 B 2-5 3/4 1-6 3/4 2-5 3/4 2-7 3/4 2-8 3/4 2-9 3/4 2-14 3/4 2-15 3/4 3-10 3/4 3-24 3/4 3-26 3/4 4-6 3/4 5-1 3/4 5-5 3/4 5-6 3/4 7-5 B 3/4 2-1 B 3/4 6-2 INSERT 1-7 1-10 2-2 2-5 2-7 2-8 2-9 B 2-1(a)

B 2-2 B 2-4 B 2-5 3/4 1-6 3/4 2-5 3/4 2-7 3/4 2-,8 3/4 2-9 3/4 2-14 3/4 2-15 3/4 3-10 3/4 3-24 3/4 3-26 3/4 4-6 3/4 5-1 3/4 5-5 3/4 5-6 3/4 7-5 B 3/4 2-1 B 3/4 6-2

1

MEMBER S OF THE PUBLIC 1.35 MEMBER(S) OF'THE PUBLIC shall include all persons who are not occupationally associated with the plant.

This category does not include employees of the utility, its contractors or its vendors.

Also excluded from this category are persons who enter the site to service equipment or to make deliveries.

This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the plant.

SITE BOONDARY 1.36 The SITE BOUNDARY shall be that line beyond which the land is not

owned, leased or otherwise controlled by the licensee.

UNRESTRICTED AREA 1.37 An UNRESTRICTED AREA. shall be. any area at or beyond the SITE BOUNDARY

'o which access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or any area within the site boundary used for residential quarters or industrial, commercial, institutional and/or recreational purposes.

ALLOWABLE POWER LEVEL APL 1.3& APL means "allowable power level" which is that power level, less than or equal to 100%

RATED THERMAL POWER, at which the plant may be operated to ensure that power distribution limits are satisfied.

COOK NUCLEAR PIANT UNIT 1 1-7 AMENDMENT NO 7g,Egg, 126

THIS PAGE INTENTIONALLYLEFT BLANK

~

~

COOK NUCLEAR PLANT UNIT 1 ANENDMENT NO,7As 126

870 650 CQ a 830 UNACC PTABLE OP ERA TION 24 QP SIA 2

50PSI 2100 SIA 0'10 200 PSI 590 570 17 ACCEPTABLE OP ER TION 75PSIA PRESSURE OPSIA 1775 2QQQ 2100 2250 2400

0. 0 0.2

-. 0.4 0.6 0.8 1.0 F RACTlON OF RATED THERMAL POWER

~$

CTGN RATED PKRMAL PO T AVG N DEGREES (0. 0, 617. 1), (1. 16, 584. 5), (1. 20, 579. 7)

(0. 0, 633. 5), (1. 11, 603. 9), ('1. 20, 593. 1)

(0. 0, 840. 3), (1. 09, 611. 5), (1. 20, 598. 3)

(0. 0, 65Q. Q), (1. 06, 623. 2), (1. 20, 806. 0)

(0. 0, 659. 0), (1. 02, 634. 8), (1. 20, 613. 0)

1. 2 Figure

2. 1-1 Reactor Core Safety Limits Four Loops ln Operation D.

C.

COOK - UMT 1

2

~ 2 Amendment No. 7b,~2

TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT 1.

Manual "Reactor Trip 2.

Power Range, Neutron Flux 3.

Power Range, Neutron Flux, High Positive Rate 4.

Power Range, Neutron Flux, High Negative Raite 5.

Intermediate

Range, Neutron Flux TRIP SETPQINT Not Applicable Low Setpoint - < 25K of RATED THERMAL POWER High Setpoint

< 109K of RATED THERMAL POWER

< 5X of RATED THERMAL POWER with a time constant

> 2 seconds 5X of RATED THERMAL POWER with a time constant

> 2 seconds

< 25K of RATED THERMAL POWER ALLOWABLE VALUES

'Not Applicable Low Setpoint - < 26K of RATED THERMAL POWER

.High Setpoint - < lllOX of RATED

'HERMAL POWER

-< 5.5X of RATED THERMA POWER with a time constant 2 seconds

.< 5.5X of RATED THERMAL POWER with a time constant 2 seconds

< 30K of RATED THERMAL POWER 6.

Source

Range, Neutron Flux

< 10s counts per second 7.

Overtemperature hT See Note 1

8.

Overpower hT See Note 2 9.

Pressurizer Pressute-L~ow

> 1875 psig 10.

Pressurizer Pressure-High

< 2385 psig 11.

Pressurizer Water LevelHigh < 92% of Instrument span 12.

Loss of Flow

> 90K of 'design flow per loop*

<<1.3 x 10s counts per second See Note 3

- See Note 4 1865 psig

.< 2395 psig

.< 93K of Instrument span

.> 89.1% of design flow per loop*

(0

• Design flow is 91,600 gpm per loop.

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION Note 1:

Overtemperature dT 4 dT f/'K o

1 2

l+~ S 1+v S

(T-T'+K (P-P ' "f (AI)3 3

1 where:

hT0 pl 1+VIS 1+T~S TlgT2

= Indicated b T at RATED THERMAL POWER

~ Average temperature, P

0

/

Indicated T at RATED THERHAL POWER (~ 567.8 F) avg

~ Pressurizer

pressure, psig Indicated RCS nominal operating pressure (2235 psig or 2085 psig)

~ The function generated by the lead-lag controller for T dynamic compensation avg

~ Time constants utilized kn the lead-lag controller for T 22 secs avg

~1 4

SPECS e

'2

= Laplace transform operator

O TABLE 2.2-1 Co ti d

EA OR P

S STEM NSTR EN A 0

NOTATIONS Conti ued Operation with 4 Loops K1 ql K2

l. 32 0.0230 K

~.

0 00110 3

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

(i)

For q-q between

-37 percent and +2 percent, f (nI) 0 (where q and q are p5rcenk RATED THERMAL powER in the top and bhttoa halves of She cork respectively, and qt + qb is-total THERMAL POWER in percent of RATED THERMAL POWER).

(ii)

For each percent: that the magnitude oi (q - q ) exceeds

-37 percent, the hV trip setpoint shall be autonatlcally red hudebII o.33 percent of its value at RATED THERMAL POWER I

(iii) por each percent that the naqnitude of (q -:q ) erceeds

+E percent, ths AT trip setpoint shall be autoaatlcally rsdu~ b 2.17 percent of its value at RATED THERMAL POWER.

TABLE 2.2-1 Continue 0

R S ST NS UMEN A IO 0 AT 0 Cont nue Note 2:

Overpower hT < hT tK -K o

4 5

~3S T - K (T-T")-f (dZ)]

where:

hT0 Indicated hT at MEED THERMAL POWER Average temperature, F

Indicated T at amED THERMAL POWER

(~ 56~.8

~>

K4 K5 K6 1-083 0

0.0177/

F for increasing average temperature And 0 for decreasing average temperature o.oa15 for T > T '6 0 for T < T" 1+~

S Y3 The function generated by the dynamic compensation Time constant utilized in the

~3 ~ 10 secs.

Laplace transforg operator rate lag controller for Tavg rate lag controller ior Tavg 2(hI) ~

0 Note 3:

The channel's maximum gore than 3.2 percent Note i:

The channel's gaxigug gore than 2,1 percent trip point shall not exceed 'its computed trip point bY AT span.

trip point shall not, exceed its coaputed trip point bY hT span.

2.1 SAFETY LIMITS BASES 4 Loop Operation Westinghouse Fuel (15x15 OFA)

(WRB-1 Correlation)

Correlation Limit Design Limit DNBR Safety Analysis Limit DNBR Typical Cell 1.17 1.33 1.45 Thimble Cell 1.17 1.32 1.45 The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than the applicable design DNBR limit, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

• +gepresents typical fuel rod

• represents fuel rods near guide tube Cook Nuclear Plant Unit 1 B 2-1(8)

Amendment No. 7$,egg

SAFETY I.IMITS BASES The curves are based on an enthalpy hot channel factor

>, of 1.49 for Qestinghouse fuel and a reference cosine axial power shape with a peak of 1.55.

An allowance is included for an increase in F+ at reduced power, based on the expression:

F 1.49 fl + 0.3 (1-P))

H where P is the fraction of RATED THERMAL POVER

Note, do not include a 4S uncertainty value, since this measurement uncertainty has been included in the design DNBR limit values, which are listed in the bases for Section 2.1.1.

These limiting heat flux conditions are high'er than th'ose calkul'ate'd for

'he 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 fl (delta I) function of the Overtemperature trip.

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

2.1.2 REACTOR COOLANT 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.

D.

C.

COOK - UNIT 1 8 2-2 Amendment No.7W, 126

SAFETY LIMITS BASES The Power Range Negative Rate Trip provides protection for control rod drop accidents.

At high power, a rod drop accident could cause local flux peaking which could cause an unconservative local DNBR to exist.

The Power Range Negative Rate Trip will prevent this from occurring by tripping the reactor.

No credit is taken for operation of the Power Range Negative Rate Trip for those control rod drop accidents for which the DNBR's will be greater than the applicable design limit DNBR value for each fuel type.

Intermediate and Source Ran e

Nuclear Flux The Intermediate and Source

Range, Nuclear Flux trips provide reactor core protection during reactor startup.

These trips provide redundant protection to the low setpoint,trip of the Power Range,. Neutron. Flux" channels<5 The source Range Channels will initiate a reactor trip at about 10 counts per second, unless manually blocked when P-6 becomes active.

The Intermediate Range Channels will initiate a reactor trip at a current level proportional to approximately 25 percent of RATED THERMAL POWER unless manually blocked when P-10 becomes active.

No credit was taken for operation of the trips associated with either the Intermediate or Source Range Channels in the accident analyses; however, their functional capability at the specified trip settings is required by this speci.fication to enhance the overall reliability of the Reactor Protection System.

Overtem erature delta T The Overtemperature delta T trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds),

and pressure is within the range between the High and Low Pressure reactor trips.

This setpoint includes corrections for changes in density and heat capacity of water with temperature and dynamic compensation for piping delays from the core to the loop temperature detectors.

The reference average temperature (T') and the reference operating pressure (P') are set equal to the full power indicated Tavg and the nominal RCS operating pressure, respectively, to ensure protection of the core limits and to preserve the actuation time of the Overtemperature delta T trip for the range of full power average temperatures assumed in the safety analyses.

With normal axial power distribution, this reactor trip limit is always below the core safety limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the notations in Table 2.2-1.

COOK NUCLEAR PLANT UNIT 1 B 2-4 AMENDMENT NO.

7$ ~

126

LIMITING.SAFETY SYSTEM SETTINGS BASES Ove ower delta T The Overpower delta T reactor trip provides assurance of fuel integrity, e.g.,

no melting, under a11 possible overpower conditions, limits the required range for Overtemperature delta T protection, and provides a

backup to the High Neutron Flux trip.

The setpoint includes corrections for changes in density and heat capacity of water with temperature, and dynamic compensation for piping delays from the core to the loop temperature detectors.

The reference average temperature (T") is set equal to the full power indicated Tavg to ensure fuel integrity during overpower conditions for the range of full power average temperatures assumed in the safety analysis.

No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to.et@ance, the overall reliabil'ity of the Rea'ctor'rotection System.

Pressurizer Pressure The Pressurizer High and Low Pressure trips are provided to limit the pressure range in which reactor operation is permitted.

The High Pressure trip is backed up by the pressurizer code safety valves for RCS overpressure protection, and is therefore set lower than the set pressure for these valves (2485 psig).

The High Pressure trip provides protection for a Loss of External Load event.

The Low Pressure trip provides protection by tripping the reactor in the event of a loss of reactor coolant pressure.

Pressurizer Water Level The Pressurizer High Water Level trip ensures protection against Reactor Coolant System overpressurization by limiting the water level to a volume sufficient to retain a steam bubble and prevent. water relief through the pressurizer safety valves.

The pressurizer high water level trip precludes water relief for the Uncontrolled RCCA Withdrawal at Power event.

COOK NUCLEAR PLANT UNIT 1 B 2-5 AMENDMENT NO.728>>

126

REACTIVITY CONTROL SYSTEMS MINIMUMTEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.1.1.5 The Reactor Coolant System lowest operating loop temperature (Tavg) shall be > 541 F when the reactor is critical.

APPLICABILITY: Modes 1 and 2*.¹ ACTION:

With a Reactor Coolant System operating. loop temperature.(Tavg)..(-541 F.,

0 restore (Tavg) 'to withi'n its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes.'URVEILLANCE RE UIREMENTS 4.1.1.5 The Reactor Coolant System temperature (Tavg) shall be determined to be > 541 F:

a.

Within 15 minutes p=..or to achieving reactor criticality, and b.

A least once per 30 minute; who.i the reactor is critical and the Reactor Coolant System Ta g i.- less than 545 F or when the low Tavg 0

alarm is inoperable.

• See Special Test Exception 3.10.3

¹With K

> 1.0.

D. C.

COOK - UNIT 1

~KNDhKNT NO. 126

POWER DISTRIBUTION LIMITS HEAT PLOX HOT CHANNEL PACTOE - P (Z)

LIMITING CONDITION FOR OPERATION 3.2.2 F (Z) shall be limited by the following relationships:

~2.15 [K(Z)]

F (Z) <

P F (Z) < [4.30) [K(Z)]

P > 0.5 P < 0.5

'P - THERMAL POWER RATED THERMAL POWER

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

m9nufacturing tolerance uncertainty and a 5% measurement uncertainty.

'K(Z) is the function obtained from Figure 3.2-3.

APPLICABILITY'ODE1 ACTION'ith F (Z) exceeding its limit:

a.

Reduce THERMAL POWER at;least 1% for each 1% F (Z) exceeds the limit within 15 minutes and similarly reduce tke 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 provided the Overpower delta T Trip Setpoints have been reduced at least 1% for each 1% F (Z) exceeds the limit.

b.

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

COOK NUCLEAR PLANT UNIT 1 3/4 2-5 AMENDMENT NO. 825iT285 126

POWER DISTRIBUTION LIMITS This page intentionally left blank.

COOK NUCLEAR PLANT - UNIT 2 3/4 2-7 AMENDMENT NO 78.728. 1~6

I o 2 O

- IU C9R

1. 0 0

A lLI h4

0. 8 tK OR 001.0 (6. 0, 'I. 0) 0, 0. 926 0

4 6

8 CORE HEIGHT (FT)

'IO FIQURE 3. 2-3 K(Z) - Normalized F sub Q(z) as a function of Core Height D.

G COOK - LNT 1

3/4 2 t.Na gl,

egg, 126

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

shall be limited by the following relationship:

PAH < 1.49

[1 + 0.3 (1-P))

Where P is the fraction of RATED THERMAL POWER APPLICABILITY:

MODE 1 ACTION:

With I exeeedlsg its limit:

a.

b.

C.

Reduce THERQQ. POWER to less than 50% of RATED THERMAL POWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, Demonstrate through in-core mapping that F is within its limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding the Pimit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER'within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and Identify and correct tHe cause of the out-of-limit condition prior to increasing THERMAL gOWER; subsequent POWER OPERATION may proceed, provided that F" is demonstrated through in-core mapping to be within its limR at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL POWER and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining 958 or greater RATED THERMAL POWER.

COOK NUCLEAR PLANT UNIT 1 3/4 2-9 AMENDMENT NO. 78,728. 126

TABLE 3.2-1 DNB PARAMETERS PARAMETER LIM1TS 4 Loops in Operation at RATED THERMAL POWER Reactor Coolant System Tavg Pressurizer Pressure Reactor Coolant System Total Flow Rate

( 570.9 F

> 2050 psig 44*

w 366,400 gpm Indicated average of at lea'st three OPERABLE instrument loops.

Limit not applicable during either a THERMAL POWER ramp increase in excess of 5 percent RATED THERMAL POWER per minute or a THERMAL POWER step increase in excess of 10 percent RATED THERMAL POWER.

Indicated value..

COOK NUCLEAR PLANT UNIT 1 3/4 2-14 AMENDMENT NO. g,728. ~26

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

~2.15 K Z x 100%, or 1OOe, whichever is less.

F (Z)xV(Z)xF P

'F (Z) is the measured hot channel factor including a 38 manufacturing tolerance uncertainty and a 58 measurement uncertainty.

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

F - 1.00 except when successive steady-state power distribution maps indicate an increase in max over Z of F Z

(Z) with exposure.

Then either of the penalties, F

, shall be taken:

P F

1 ~ 02 i or P

F - F 00 provided that Surveillance Requirement 4 '

iR satisfied once per 7 Effective Full Power Days until two successive maps indicate that the max over Z of F

Z K Z is not increasing.

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

1)

Lower core region 08 to 108 inclusive.

2)

Upper core region 908 to 1008 inclusive.

APPLICABILITY:

MODE 1 COOK NUCLEAR PLANT UNIT 1 3/4 2-15 AMENDMENT NO. Ns7lP.

126

TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT 1.

Manual Reactor Trip 2.

Power Range, Neutron Flux a.

High Setpoint

. b.

Low Setpoint

RESPONSE

TIME NOT APPLICABLE

< 0.5 seconds+

< -0.5 seconds*

3; Power Range, Neutron Flux, High Positive Rate NOT APPLICABLE 4.

Power Range, Neutron Flux, High Negative Rate 5.

Intermediate

Range, Neutron Flux 6.

Source

Range, Neutron Flux 7.

Overtemperature Delta T 8.

Overpower Delta T

< 0.5 seconds*

NOT APPLICABLE NOT APPLICABLE 6.0 seconds NOT APPLICABLE 9.

Pressurizer Pressure--Low

10. Pressurizer Pressure--High

11. Pressurizer Mater Level--High

< 1.0 seconds

< 1.0 seconds

.< 2.0 seconds

• Neutron detectors are exempt from response time testing.

Response

time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.

n TABLE 3.4-4 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS CD CD I

O FUNCTIONAL UNIT 1.

SAFETY INJECTION, TURBINE TRIP, FEEDWATER ISOLATION, AND MOTOR DRIVEN FEEDWATER PUMPS a.

Manual Initiation b.

Automatic Actuation Logic c.

Containment Pressure High d.

Pressurizer Pressure Low e.

Differential Pressure Between Steam LinesHigh f.

Steam Flow in Two Steam LinesHigh Coincident with T Low-Low or Steam 3'free Pressure--Low TRIP SETPOINT Not Applicable Not Applicable

< 1.1 psig

> 1815 psig

< 100 psi

< 1.42 x 106 lbs/hr from OX load to 20K load.

Linear from 1.42 x 106 lbs/hr at 20K load to 3.88 x 106 lbs/hr at 100K load Tayq

> 541'F

> 500 psig steam line pressure ALLOWABLE VALUES Not Applicable Not Applicable

< 1.2 psig

> 1805 psig

< 112 psi

< 1.56 x 106 lbs/hr from OX load to 20K load.

Linear from 1.56 x 106 lbs/hr at 20K load to 3.93 x 106 lbs/hr at lOOX load.

Tayg ) 539oF

> 480 psig steam line pressure

TABLE 3.3'l (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEH INSTRVHENTAT TION TRIP SETPOINTS FUNCTIONAL UNIT I

. 2.

Containment Radioactivity g

~

High Train A (VRS-1101, M

ERS-1301'RS-1305) 3.

Containment Radioactivity High Train B (VRS-1201, ERS-1401, ERS-1405)

TRIP SETPOINT See Table 3.3-6 See Table 3.3-6 ALlQMABLEVALUES Not Applicable Not Applicable 4.

STEAH LINE ISOLATION WI cn a.

Hanual b.

Automatic Actuation Logic c.

Containment Pressure High-High d.

Steam Flow In Two Steam Lines High Coincident vith T Lou-Law or Steam Line ir8ssure'ow Not Applicable

~ >

Not Applicable 2'

psig 1.42 x 10 lbs/hr from Oi 6

load to 2I)l load.

Linear-from 1.42 x 10 1I>s/hr at 20'oad to 3.88 x 10 lbs/hr at lppi load.

T 541 F

avg

> 5pp psig steam line pressure t]at Applicable Nat Api>licable C 3 psig l.56 x 10 lbs/hr from 0%

6 load to 20% logd linear fram 1.56 x 10 1bs/hg at 201 load to 3 93 x 10 lbs/

hr at 1001 load.

T 539 F ave) 480 psig steam line pressure 5 ~

TURBINE TRIP AND FEEDMATER ISOLATION a.

Steam Generator Mater Level <<-

High"High 67% of narra-range instiument span each steam generator 681 of narrow-range instrument span each steam generator

REACTOR COOIANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizer shall be OPERABLE with a ~ater volume less than or equal to 92% of span and at least 150 kW of pressurizer heaters.

APPLICABILITY:

MODES 1,2, and 3.

ACTION'ith the pressurizer inoperable due to an inoperable emergency power supply to the pressurizer heaters either restore the inoperable emergency power supply within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6

hours and in HOT SHUTDOWN within the, following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />..

With the.

pressurizer otherwise inoperable, be in at least HOT SHUTDOWN with the reactor trip breakers open within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE RE UIREMENTS 4.4.4.1 The pressurizer water volume shall be determined to be within its limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.4.4.2 The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE at least once per 18 months by transferring power from the normal to the emergency power supply and energizing the required capacity of heaters.

COOK NUCLEAR PLANT UNIT 1 3/4 4-6 AMENDMENT NO. Q s 126

3 4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS ACCUMULATORS LIMITING CONDITION "OR QPERAT'"'<

3.5.1 Each reactor coolant system accumulator shall be OPERABLE with:

a.

The isolation valve open, b.

A contained bora-ed water volume of between 921 and 971 cubic

feet, c.

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

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

APPLICABILITY:

MODES l, 2 and 3.>>

ACTION:

a.

b.

With one ecc..u!ator inoperable, except as a result of a closed isolation vaiv

, restore the inoperable'cumulator to OPERABLE status within cn hour or be in HOT SHU",rOWN within the next..

8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

With one accumulator inoperable due to the isolation valve being cIosed, either imediateiy open the isolation valve or be in HOT STANDBY within one,'hour and be in HOT SHUTDOWN within tne next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

SURVEILLANCE RE UIREMENTS 4.5.1 Each accumulator shall be demonstrated OPERABLE:

a.

At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by:

1.

Verifying the water level and nitrogen cover-pressure in tile tanks, and 2.

Verifying that each accumulator isolation valve is open.

Pressurizer Pressure above 1000 psig.

D.

C.

COOK - UNIT 1

3/4 5-1 Amendment No.

777~

126

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued}

d.

At least once per 18 months by:

1.

Verifying automatic isolation and interlock action of the RHR system from the Reactor Coolant System when the Reactor Coolant System pressure is above 600 psig.

e.

2.

A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks,

screens, etc.)

show no evidence of structural distress or abnormal corrosion.

At.least once per 18 months, during shutdown, by:*

1.

Verifying that each automatic valve in the flow path actuates to its correct position on a Safety Injection test signal.

2.

Verifying that each of the following pumps start automatically upon receipt of a safety injection test signal:

a)

Centrifugal charging pump b)

Safety injection pump c)

Residual heat removal pump By verifying that each of the following pumps develops the indicated discharge pressure on recirculation flow when tested pursuant to Specification 4.0.5 at least once per 31 days on a STAGGERED TEST BASIS'.

Centrifugal charging pump

> 2405 psig, 2.

Safety Injection pump

> 1345 psig 3.

Residual heat removal pump

> 165 psig g.

By verifying the correct position of each mechanical stop for the following Emergency Core Cooling System throttle valves:,

1.

Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE.

The provisions of Specification 4.0.6 are applicable.

Cook Nuclear Plant Unit 1 3/4 5-5 Amendment No. 797> 3:26

4 EMERGENCY GORE COOLING SYSTEMS SVRVEILIANCE RE UIREMENTS Continued 2.

At least one<< per 1B months.

Boron In]<<ction Throttle Valves Safety In)acti.on Throttle Valves V<<lvo Number Valve Numb<<r l.

1-SI-141 L1 1.

1-SI-121 N 2.

1-SI-141 L2 '.

1 SI-121 S

-3.

1 Sl-141 L3 4.

1-SZ-141 L4 h.

By performing <<Claw balance test during shutdown. fo11owing completian af modifications ta the ECCS subsystem that alter the subsystem flaw characteristics and verifying the.following flaw z'ates; Ior'on Ingestion System 8<<f<<ty Inject&an System

~*

Leap 1 Boron I+action Flow 117.5 gpm Loop 1 <<nd 4 Cold Leg Flow > 300 gpm.

Loop 2 Boron In)ection Flaw 117.5 gpm Loap 4 Baron In)ection Flow 117.5 gpm Laap 2 and 3 Cold Leg Flow > 300 gpm

• s'aop 3 Baron Zn)ection Combined Loop 1,2,3 and 4 Cold Flaw 117. 5 gpm Leg Flow (single pump) less than or equal to 640 gym.

Total V

SIS (single pump) flow, including miniflow, shell not exceed 700 gpm.

The flow r<<t<< in each Baron In)eccl.on (BI) line should~be ad)usted to provide 117,5 gpm (nominal) flow in each loop, Under these conditions there is zero miniflaw and 80 gpm plus or minus 5 gpm simulated RCP seal !.n]ection line flow.

The <<ctual flow in <<<<ch BI line may deviate from the nominal so long as:

<<)

lhe difference between the highest and lowest flow is 25 gpm or less.

b) the tot<<1 Claw to the four branch 11nes does not exceed 470 gpm.

c) t'h<<minimum flaw (total flow) through the three most conservative (lowest flaw) branch lines must not be less than 300 gpm.

d)

The charging pump d/schargo resistance (2.31xPd/Qd

) must not be21ess than 4.73K-3 ft/gpm and must not be greater than 9.27K-3 ft/gpm, (Pd is the pump discharge pressure at runout; Qd is the total pump flaw rate.)

COOk NUCLEAR PLANT UNIT 1 3/4 5-6 ZmNDMENT NO Pg~

126

PLANT SYSTEMS AUXILIARYFEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1-2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:

a.

Two feedwater

pumps, each capable of being powered from separate emergency

busses, and b.

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

'I APPLICABILITY:

MODES 1, 2'and 3.

ACTION:

a.

With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or 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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b.

With two auxiliary feedwater pumps inoperable, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c.

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

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

a.

At least once per 31 days by:

1.

Verifying that each motor driven pump develops an equivalent discharge pressure of > 1375 psig at 60 F in recirculation flow.

2.

Verifying that the steam turbine driven pump develops an equivalent discharge. pressure of > 1285 psig at 60 F and at a flow of > 700 gpm when the secondary steam supply pressure is greater than 310 psig.

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

D.

C.

COOK - UNIT 1 3/4 7-5 Amendment No.gg,/gal, 126

3 4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:

(a) maintaining the minimum DNBR in the core greater than or equal to the safety limit DNBR during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria.

In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200 F is not exceeded.

The definitions of hot channel factors as used in these 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.

3 4.2.1 AXIALFLUX DIFFERENCE AFD Target flux difference is determiwed at equilibrium xenon conditions.

The full length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady state operation at high power levels.

The value of the target flux difference obtained under these condtions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions, Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level.

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

D.C.

COOK - UNIT 1 B 3/4 2-1 Amendment No. gg,gS,N~72H~

>26

CONTAINMENT SYSTEMS BASES 3 4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that 1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 8 psig and 2) the containment peak pressure does not exceed the design pressure of 12 psig during LOCA conditions.

The maximum peak pressure resulting from a LOCA event is calculated to be 11.89 psig, which includes 0.3 psig for initial positive containment pressure.

3 4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that 1) the containment air mass is limited to an initial mass sufficiently low to prevent exceeding the design pressure during LOCA conditions and 2) the ambient air temperature does not exceed that temperature allowable for the continuous duty rating specified for equipment and instrumentation located within containment.

The containment pressure transient is sensitive to the initially contained air mass during a LOCA.

The contained air mass increases with decreasing temperature.

The lower temperature limit of 60 F will limit the peak pressure to 11.89 psig which is less than the containment design pressure of 12 psig.

The upper t'emperature limit influences the peak accident temperature slightly during a LOCA; however, this limit is based primarily upon equipment protection and anticipated operating conditions.

Both the upper and lower temperature limits are consistent with the parameters used in the accident analyses.

3 4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment steel vessel will be maintained comparable to the original design standards for the life of the facility.

Structural integrity is required to ensure that (1) the steel liner remains leak tight and (2) the concrete surrounding the steel liner remains capable of providing external missile protection for the steel liner and radiation shielding in the event of a LOCA.

A visual inspection in con]unction with Type A leakage tests is sufficient to demonstrate this capability.

COOK NUCLEAR PLANT UNIT 1 B 3/4 6-2 AMENDMENT NO.

126