ML17326A327

Proposed Tech Specs,Covering Reactor Trip Sys Instrumentation Trip Setpoints

ML17326A327
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Site:	Cook
Issue date:	11/02/1979
From:	INDIANA MICHIGAN POWER CO.
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Download: ML17326A327 (41)

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ATTACHMENT A TO AEP:NRC:00297 V 911090

'0 I CHANGE NO, 1 Revision to Table 2.2-1 "Reactor Tt'ip System Instrumentation and Tri Set oints" and Table 3.2-1 "DNB Parameters" - Unit No. 2 This revision involves changing the maximum Tavg. at Rated Thermal Power used for the overtemperature and overpower QT trips and DNB limits. The technical basis for this change will be provided as part of Amendment 85 to the Donald C. Cook FSAR which will include the reanalysis results of the various design basis transients affected by this change. A summary of the transient reanalyses is enclosed in Attachment 'C'f this letter. The results show conformance with established safety criteria. This change will not adversely affect the health and safety of the public.

CHANGE NO. 2 Revision to Item 3.2.2 Fi ure 3.2-2 and Basis Item 3 4 2. 1 - Unit'No. 2 This change involved lowering the maximum allowable Fq(Z) limits.

The maximum value is being reduced from 2.32 to 2. 11. At present, the

2. 11 limit is being used on an administrative basis in compliance with the NRC order which followed the discovery of a 'logic inconsistency'n the metal-water reaction calculation of the Westinghouse ECCS Evaluation Models. The reason for this revision is to change the status of the limit from an administrative limit to a Technical Specification limit.

The present administrative limits were established after a new Westinghouse ECCS analysis was performed in which the logic inconsistency was corrected.

These results were transmitted to you on April 28, 1978. This change will not affect the health and safety of the public.

CHANGE NO. 3 Revision to Section 4.2.2.2e - Unit No.,2 This change involves revising the F y Limits at Rated Thermal Power.

This change is based on a Donald C. Cook Fnit 2, Cycle 2 safety evaluation. performed by Westinghouse Electric Corporation. The data used in this evaluation is reported in WCAP-9566, "The Nuclear Design and Core Management of the Donald C. Cook Nuclear Power Plant Cycle 2". This change will not adversely affect the health and safety of the public.

CHANGE NO. 4 Revision to Item 3.2. 1 and Fi ure 3.2-1 Unit No. 2 This change involves increasing the upper power limit for the taking of action from'84/ to 90Ã Rated Thermal Power. The technical'.

basis for this change is the same as that of Change .No. 3. This change will not adversely affect the health and safety of the public.

CHANGE NO. 5 Revision to Items 3.2.6 and 4.2.6 - Unit No. 2 This change involves raising the APDMS turn-on point to 100K Rated Thermal Power . This change is based on the revisions to Item 3.2. 1 and Figure 3.2-1 as discussed in Change No. 4 since the APDMS turn-on point is defined as 105 above the upper limit of Item 3.2.1. This change will not adversely affect the health and safety of the public.

CHANGE NO. 6 Revision to Sections 3.5.5 4.5.5 and Bases Item B 3 4.5.5 - Unit 'No.'

This calls for increasing the minimum RWST temperature from 35 F to 70 F. The basis for this change is that the Unit 1 Reload Safety Analysis performed by Exxon Nuclear Company used a safety injection water temperature of 70oF. There are also editorial changes- to the Basis Item. These changes will make the Unit 1 specifications consistent with the Unit 2 specifications in this area. This change will not affect the health and safety of the public.

CHANGE NO. 7 Revision to Table 2.2. 1 and Basis Item B.2.2. 1 - Unit:Nos. L & 2 This change calls for the revision of the setpoints and allowable values for the "Power Range, Neutron Flux, Hi'gh Positive (and Negative)

Rate" functions. The revision involves changing the time constant for these functions from its present value of ~ 2 seconds to ~ 1 second, and reduction of the trip setpoint for the high negative rate ti ip-from

~ 5A/sec of Rated Thermal Power (RTP) to ~ 3! of RTP. We have been informed by Westinghouse Electric Corporation that this change will assure thatany or multiple dropped rods will generate a reactor trip via the negative rate trip -function. Bases Section 2.2. 1 has also 'been revised accordingly. This change will not affect the health and safety of the public.

0 '0 l E

I

ATTACHMENT B TO AEP:NRC:00297

CHANGE NO. 1 TABLE 2.2-1 Continued REACTOR TRIP SYSTB4 INSTRUHENTATION TRIP SETPOINTS NOTATION 1+t S NOTE 1: Overtemperature AT < aT 0 fK1-K2 1

2~S 1 (T-T )+K (P-P 3 )-f1(aI)]

1 "2

where: hT 0 .

Indicated aT at RATED THERi1AL POWER T Average temperature, 'F Indicated T avg at RATED THERMAL POWER < 573.8F Pressurizer pressure, psig pw 2235 psig (indicated PCS nominal operating pressure) 1+~lS The function generated by the lead-lag controller for T dynamic compensati I+~~s avg

& ~2 Time constants utilized in the lead-lag controller for T ~ = 33 secs,

= 4 secs. avg 1 w2 Laplace transform operator

TASLE 2.2-1 Continued)

REACTOR TRIP SYSTEM INSTRUMiiENTATION TRIP SETPOINTS NOTATION Continued)

Operation with 4 Loops Operation with 3 Loops K1=1.334 K1=1.116 K2 0.01607 K2 0 01607

~

= 0.000744 = 0.000744 K3 K3 and f> (hI) is a function of the indicated difference between top and bottom detectors of th6 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 + 3 percent, f (hI) = 0 (wher3 q )nd q are percent RATED TREINAL PORER in the top and bottom halves oI'he c5re respectively, and q + q is total THERMAL POMER in percent of RATED THERMAL PO!lER).

(ii) for each percent that the iaagnitude of (q - q ) exceeds- 40 percent, the aT trip setpoint shall be automatical)y relluced by 1.8 percent of its value at RATED THERMAL POWER.

(.iii) for each percent that the maqnitude of (q - q ) exceeds + 3 percent, the AT trip setpoint shall be automatical)y reIIuced by 2.2 percent of its value at RATED THERMAL PO)f!ER.

TABLE 2.2-1 (C~onttnued REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION Continued Note 2: Overpower AT < AT

[Ke-KB

'"3'here:

1

~3S B

T - KB (T-T") f2-(AI)]

hT = Indicated hT at rated power T = Average temperature, 'F T" = Indicated T at RATED THERMAL POWER < 573.8~F avg

= 1 078 K4 K5 0.02/'F for increasing average temperature and 0 for decreasing average tempera ture K

6

= '.00197 for T > T"; K = 0 for T < T" 6

v3S

~+~3S The function generated by the rate lag control er for 1 T dynamic compensation avg T3 Time constant utilized in the rate lag controller for T

= 10 secs. avg 3

S = Laplace transform operator f2(AI) = 0 for all hI Note 3: The channel's maximum trip point shall not exceed its computed trip point by more than 4 percent.

TABLE 3.2-l DNB PARAMETERS LIMITS PARAMETER 4L 00 3 Loo s In 0 eration Reactor Coolant System T 578 F= 570'F '

avg Pressurizer Pressure > 2220 psia* 2220 psia*

• Limit not applicbale during either a THERMAL POWER ramp increase in excess of 5X RATED THERMAL POWER per minute or a THERMAL POWER step increase in excess of lOX RATED THERMAL POWER.

CHANGE NO. 2 POWER 01 STRIBUTION L IMITS HEAT FLUX HOT CHANNEL FACTOR-F (2)

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

F (Z) < pp yQ [K(Z)] for P > 0.5 p

F~(Z) < D4.22)j [f(2)l for P c 0.5 THER!1AL POWER p

RATED THEPMAL POWER and K(Z) is the function obtained from Figure 3.2-2 for a given core height location.

APPLICABILITY: MODE 1 ACTION:

With F~(Z) exceeding its limit:

a . Comply with either of the following ACTIONS:

1. Reduce THERMAL POWER at least 1". for each 1/ F~(Z) exceeds the limit within 15 minutes and similiarly reduce the Power Range Neutron Flux-Hggh 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 OPERA .ON 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 />; subsequenc POWER OPERATION may proceed provided the Overpower dT Trip Setpoints have been reduced at least lil for each 1" F~(Z) exceeds the limit. The Overpower aT Trip Setpoint reduction shall be performed with the reactor in at least HOT STAN08Y.

2. 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 condiiton prior to increasing THERMAL POWER above the reduced. limit re-quired by a, above; THERMAL POWER may then be increased provided F~(Z) is demonstrated through incore'apping to be within its 1 imi t.

D. C ~ COOK - UNIT 2 3/4 2-5

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~ I FIGUfIE 3,2-2 K(Z) NOBMALIZED Fo(Z)'AS A FUNCTION OF COAL IIEIGII'I HEAT FLUX HOT CHAHfIEL FACTOR NOPJQLIZED OPERATING ENYELOPE FOUR-LOOP OPERATION Basis: F~(Z) x P ECCS limit of 2.11.

1.0  :-:. (6;0,1.000) i

":": (11.16, 0.935

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0. C. COOK - UNIT 2 3/4 2-8

3/4.2 POll" g O'STR'T1011 L1N1TS 8ASES The speci-,ications o this section provide assurance of fuel in eg-rity during Condition I (Normal Operation) and II (Incidents of Hoderat Frequency) events by: (a) maintaining the calculated DHBR in th corp at or above design during normal operation and in short term <<ransients, and (b) limiting the =ission gas release, fuel pellet t mperature and cladding mechanical proper.ies to within assumed design cri eria. In addition, limiting <<he peak linear power density during Condi:ion I events provides assuranc that the initial conditions assumed =or -.he LOCA analyses are met and the ECCS acceptance crit ria limiit of 2200'F is not exceeded.

The definitions of c rtain hot channel and peaking factors as used in these specifications are as ollows:

F<(Z) Heat Flux Hot Channel Factor, is defiried as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the averaoe suel rod heat, flux, allowing for man-ufacturing tolerances on fuel pellets and rods.

af Nuclear Enthalpy Rise Hot Channel Fac.or, is defined as the aH 'ratio of the integral of linear power along the rod with the highest integrated power to the average rod power.

F,(Z) Radial Peakino Fac.or, is defined as the rat o o peak power density to average power density in the horizontal plane, at core elevation Z.

3/4.2.1 AXIAL FLUX DIFFERENCE AFD The limits on AXIAL FLUX DIFFERENCE assure that the Fg(Z) upper bound envelope o= .'2.11 times the norma1i=ed axia1 peaking ac-.or is not exceeded during either normal operation or in the event of xenon redis- ]

tribution ,ollowing power changes.

Target 'flux diiferen e is determined at equilibrium xenon conditions.

The full leng.h rods may be positioned wi h n the core in accordance with their respective insertion limits and should be insert d near The their normal position =or s.eady state operation at high power levels.

value of the taroet flux difference obtained under these conditions divided by the fraction of RATED THERMAL POMER is the :arget fluxcon-difference a<<RATED THERMAL POMER for the associated core burnup are POWER levels ditions. Target flux differonces for other THERMAL obtained by r;.ul iply.'ng .he RATED THERtiAL POWER value by -.he appropria-e fractional THERMAL PO'nlER level. The periodic upda ing of consid .he "arget, ra,-.icns.

flux difference value is necessary to reflect core burnup

- UNIT 2 8 3/4 2-1 Amendment No. 1O D.C. COOK

POWER D<STRrgUTIOH LIl!ITS BASES Although it is in:ended that the plant will be operated with the AXIAL FLUX DIFFEREllCF. wiihin the -.'5~ target band about the target flux dif erence, during rapid plani Tl-;ERYAL POWDER reduciions, control rod motion wi-11 cause the AFD io deviaie outside of the iarget band ai re-duced THE.";Y~L POWER leve1s. This devia ion will noi arfect the xenon redis ribut-on sufficiently to chance the envelope', peaking f'ac:ors which may be r=ached on a subsequent reiurn to %TED THEP, VL POWER (with the AFD uiihin he target b nd} provided the time duraiion of the devi-ation 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 /> penal iy deviaiicn limit cumu-lative 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 =or oper tion ouis.de of t he targ i band but w i thin the 1 imi is o-, Fi gure 3. 2-1 whi 1 e at TH=RYartL POW=R levels between GG" and 'gpss of RATED THERqAI p0>;ERR. For TH P3'."L POWER levels be.~een lS~ and "t;.. or t<e) tu ttIt~c t vhcR, deviaiions of the AFD ouiside of the targei b nd are less signiric'nt. The pena', ty of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ac.ual time ref1ec-s this reduced significance.

Provisions for monitor tng the AFD on an automatic basis are derived from ti;e plan process computer.ihrough the AFD Yenitor Alarm. The determines the one minute average of each of the OPERABLE if'he

'omputer excore det=ctor outputs and provides an alare message in-.vediate'ly AFD for a= least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the ta,"g t band and the THER'lAL PGWER is greater than 90% of,~T=.D THERMAL POWER. During operation at THERMAL PGWER levels between ~G~ and 90/ and between 15 and.50" RATED THER'1AL PO'~ER, ihe computer ou puts an alarm message when the penal iy deviation accumulates beyond he 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.

D.C. COOK - Ui'lIT 2 B 3/4 2-2

CHANGE NO. 3

• ~

I~ ~l POWEP. 9ISTi? IBUTION LIi"ITS SURVE ILLANC"- R:gUIREl~.'ITS (Cont inued',

2. When the F C

i's less than or equal to the FxRTP limit for xy the appropriate measured core plane, additional power C

distribution maps shall be taken and Fcompared lo L

F xy "

and 'xy at least F once per 31 EFPO.

e. The F limits for RAieD TH"-?~lAL POW="R within specific core planes shall be:

F-RTP <

1.87 for all core planes containing control rods.

RTP z Fxy - 1.58 for all unrodded planes above 6.2 feet.

- 1.62 for all unrodded planes below 6.2 feet.

The F limits of e, above, are not applicable in the followino core planes regions as measured in percent of core height from the b'cttcrn of the fuel .

1, Lower core region "rom 0 to 15", inclusive.

2. Upper core region from 85 to 100~, inclusive.

3, Grid plane regions at 17.8 + 2"... 32.1 ~ 2', 46.4 + 2~,

60,6 + 2 and 74,9 + 2~ incIusive.

4. Core plane regions within + 2" of core height (+2.88 jnches) about the bank demand position of the bank "0" contro1 rods, g, Mith Fexceeding F: L.

'l. The F~(Z) limit shall be reduced at least 1 for each 1" c

F'xceeds zy F, xy',

and The effects of Fon Fq(Z). shall be evaluated to determine tf F~(Z) is within its limits.

,2.2.3 When Fq(Z) is mea" ured for other than F determinations a overall measured F~(Z) shall be obtained from a paver distribution map and increased by 3~ to account for manufacturing increased by 5~ to account for measurement uncertainty.

toleranc s and urther

0. C. COOK - UNIT 2 3/4 2-7 Pn'. nd;. nt <<o. 10

CHANGE NO. 4 3/4.2 POWER DISTRIBUTION L IMITS AXIAL FLUX DIFFERENCE (AFD LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within a

+5% target band (flux difference units) about the target flux difference.

APPLICABILITY: YiODE 1 ABOYE 50% RATED THERMAL POWER*

ACTION:

a. With the indicated AXIAL FLUX DIFFERENCE outside of the +5".

target band about the target flux difference and with THERMAL POWER:

1 . Above 90% of RATED THERMAL POWER, within 15 minutes:

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

2. Between 50% and 90% of RATED THERt1AL POWER:

a) POWER OPERATION may continue provided:

1) The indicated AFD has not been outside of the

+5% target band for more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> penalty deviation cumulative during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and

2) The indicated AFD is within the limits shown on Figure 3.2-1. Otherwise, reduce THERMAL POWER'o less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High Trip Setpoints 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 />.

b) Surveillance testing of the Power Range Neutron Flux Channels may be per formed pursuant to Specification 4.3.1.1.1 provided the indicated AFD.is maintained wi thin the limits of Figure 3.2-1. A total of 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> oper ation may be accumulated with the AFD out-side of the target band during this testing without penal ty deviation.

D. C. COOK - UNIT 2 E, i ..2 3/4 2-1

OWER DISTRIBUTION LIMITS ACTION: (Continued) c) Surveillance testing of the APDMS may be performed pursuant to Specification 4.3.3.7.1 provided the indicated AFD is maintained within the 1-imits of Figure 3.2--1. A total of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of operation may be accumulated with the AFD outside of the target

. band during this testing without penalty deviation.

b. THERMAL POWER shall not be increased above 90/ of RATED THERMAL POWER unless the indicated AFD is within the +5K target band and ACTION 2.a) 1), above has been satisfied.

c. THERMAL POWER shall not be increased above 50". of RATED THERMAL POWER unless the indicated AFD has not been outside of the

+5"" target band for more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> penalty deviation cumulative during the prev ious 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ~

SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AXIAL FLUX DIFFEPENCE shall be determined to be within its limits during POWER OPERATION above 15",. of RATED THERMAL POWER by:

a. Monitoring the indicated AFD for each OPERABLE excore channel:

- l. At least once 'per 7 days when the AFD Monitor Alarm is OPERABLE, and

2. At least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after restoring the AFD Monitor Alarm to OPEPABLE status.

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

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

D. C. COOK - UNIT 2 3/4 2-2

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-50 -40 -30 -20 -10 0 10 20 30 40 50 FLUX DIFFERENCE (Dl)  %

FIGURE 3.2-1 AXIALFLUX DIFFEREi'JCE LIMITS AS A FUNCTION OF RATED THERMAL PON/ER 1

0. C. COOK - UNIT 2 3/4 2-4

CHANGf NO. 5 POMER DISTRIBUTION Llf)ITS AXIAL POMER DISTRIBUTION L I Y(IT Ill G CONDITION FOR OPERATION 3.2.6 The axial power distribution shall be limited by the following relationship:

p,si][ v(z) 1

[Fj(Z)~S (R ) (PL) (1 03)(1 + Q ~

) (1 .07)

Where:

a. F.(2) is the normalized axial power distribution from thimble j at core elevation 2.

b. PL is the fraction of RATED THERltAL POMER.

C. K(Z) is the function obtained from Figure 3.2-2 for a given core height location.

d. R ~, for j thimble j, is determined from at least n=6 in-core flux maps coverirg the full configuration of permissible rod patterns above IDOLS of RATED THERtQL POWER in accordance with:

n R

J n i lj Mhere:

Peas 1

R.- =

1 j Max and [F..

ij (Z)]<Max is the maximum value of the normalized axial distribution at elevation Z from thimble j in map i which had a measured peaking factor without uncertainties or densification allowance of Peas. t~

0. C. COOK - UNIT 2 3/4 2-17

POWER DISTRIBUTION LIHITS LIHITING CONDITION FOR OPERATION Continued

a. is the standard deviation associated with thimble J

j, expressed as a fraction or percentage of R., and is derived from n flux maps J

from the relationship below, or 0 02, (2,) whichever is gr ater.

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R.

R ~ll R 2 1/2 The factor 1.07 is comprised of 1.02 and 1.05 to account 'or 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: HODE 1 above 1005 OF RATED THERllAL POWER~.

ACTION:

a. With a F.(Z) factor exceeding [F.(Z)]S by < 4 percent, reduce J J THERHAL POWER one percent for every percent by which the F.(Z) factor exceeds its limit kithin 15 minutes and within the next two hours either reduce the F .1.(Z) factor to within its limit or reduce THERHAL POWER to 1005 orless of RATED THERMAL POWER.

b. With a F.(Z) factor exceeding LF.(Z)]S by > 4 percent, reduce J J THERHAL POWER to 100K, or less of RATED THERHAL POWER within 15 minutes.

0 The APDHS may be out of service when surveillance for determining power distribution maps is being performed.

D. C. COOK - UNIT 2 3/4 2-18

POWER DISTP.IBUTION L.'MITS SURYEI LLANCE REOUI REMENTS 4.2.6.1 F.(Z) shall be determined to be within its limit by:

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a. Either using the APDflS to monitor the thimbles required per Specification 3.3.3.6 at the following frequencies.

1. 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., Immediately and at intervals of 10, 30, 60, 90, 120, 240 and 480 minutes following:

a) Increasing the THERMAL POWER above 1005 of RATED

'HERMAL POWER, or b) Movement of control banl "D" more than an accumulated total of 5 steps in any one direction.

b. Or using the movable incore detectors at the ,ollowing fre-quencies 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 100K of RATED THERMAL POWER, or b) Movement of control bank "D" more than an accumulated total of 5 steps in any one direction.

4.2.6.2 When the movable incore detectors are used to monitor F .(Z),

J at least 2 thimbles shall be monitored and an F.(Z) accuracy equivalent J

to that obtained from the APDMS shall be maintained.

D. C. COOK - UNIT 2 3/4 2-19

CHANGE NO. 6 EN".RGEi'ICY CORE COOLING SYSTEMS LIMITING CONDITION FOR OPERATION 3.5.5 The refueling water storage tank (RMST) shall be OPERABLE with:

a. A minimum contained volume of 350,000 gallons of borated water.

b. A minimum boron concentration of 1950 ppm, and

c. A minimum water temperature of 70oF APPLICABILITY: MODES 1, 2, 3 and 4.

ACTiON:

With the refueling water storage tank inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or 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 COLD SHUTDOWN within the follo&ing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.5.5 The RWST shall be demonstrated OPERABLE:

a. At least once per 7 days by:

l. Verifying the water level in the tank, and

2. Verifying the boron concentration of the water.

b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by veri,ying the RMST temperature .

when the outside air temperature is < 7poF, D.C.COOK-UNIT 1 3/4 5-10

~ ~

I rt EMERGEiVCY CORE CGOLiiVG SYSTEMS BASES e

Mith the RCS temperature below 350 F, one OPERABLE ECCS subsys em is acceptable without single failure consideration on the basis o the stable reactivity condition of the reactor and the limited core cooling requiremenis.

The Surveillance Pequirements provided to ensure OPERABILITY of each component ensures thai at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is ftzintained. Surveil-lance requi. ements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be ttaintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in ihe piping sysiem io each injection point is necessary to: (1) prevent total pumo flow from exce ding rurou. conditions when the system is in its minimum resistance configuration, (2) provide ihe proper flow split betwe n injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assured in the ECCS-LOCA analyses.

3/4.5.< BORON INJECTEON SYSTEM The OPERAB I ITY of the boron injection system as part of the ECCS ensures that sufficient negaiive reactivi .y is injected inio the core to counter-act any positive increase in reactivity caused by 'RCS system cooldown.

RCS cooldown c n be caused by inadvertent depressurization, a loss-of-coolant accident or a steam line rupture.

The limits on injection tank minimum volu...e and boron concentration ensure thai the assumptions used in the steam line break analysis are met. ~ ~

The OPERABILITY of the redundant heat, tracing channels associated with the boron injection sys iem ensure thai the solubility of the boron solution will be maintained above the solubility limit of 135'F at 21OGO ppm boron.

4 3/4.5.5 REFUELING WATER STORAGE TANK The OPERABILITY of the RWST as part of the ECCS ensures that a sufficient supply of borated wate~ is available =or injection by the ECCS in the event of a LOCA. The limi s on RWST minimum volume and boron concentra-tion ensure that 1) sufficient water is available within c'oniainmeni to permi", recirculation cooling -,low to the core, and 2) the reactor wil'I remain subcritical in the cold condition following mixing oi the RRST and the RCS water volumes with al'I control rods inserted except for the most reactive control assembly. These assumptions are consistent with the LOCA analyses.

D. C. COOK-UNIT 1 B 3/4 5-2 Amendment No. 23

Es~1ERGE.'sCY CORE COOL I'sG SYSTE'1S BASES The contair ed water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

The limits on contained water volume and boron concentration of the PAlST also ensure a pH value of between 8.5 and 11.0 for the solution recir-

~

culated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The ECCS analyses to determine F limits in Specifications 3.2.2 aod 3.2.6 assured a RIIST water temper)tore of 70op. This temperature value of the RWST water determines that of the spray water initially delivered to the containment following LOCA. It is one of the factors which determines the containment back-pressure in the ECCS analyses, performed in accordance with the provisions of 10 CFR 50.46 and Appendix K to 10 CFR 50.

a D. C. COOK - UNIT I B 3/4 5-3

CHANGE NO. 7 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOIHI ALLOMABLE YALUES

l. Hanual Reactor Trip Not Applicable Hot Applicable

2. Power Range, Neutron Flux Low Setpoint - < 25$ of RATED 'Low Setpoint - < 26K of BATED TIIERIQL PO'WER TIIERIIAL POWER 1

Iligh Setpoint - < 109K of fblTED Iligh Setpoint - < llOX of RATED TIIERNL I'OWER TIIERINL POWER

3. Power Range, Neutron Flux,. < 5X of RATED TIIERIIAL POWER with < 5.5$ of RATED TIIEINAL POWER Iligh Positive Rate a'ime constant > l second with a tinie constant > 1 second

~ j0

>'i

4. Power Range, Neutron Flux, of RATED TIIERIIAL POWER with of RATED TIIERIIAL POWER lligh Negative Rate a tioie constant > 1 second with a tiaia constant . I sacond I

5. Intermediate Range, Neutron < 25$ of RATED TIIERIQL POWER < 30K of RATED TIIERtlAL POWER Flux 5

6. Source Range, Neutron Flux < l05 counts per second < 1.3 x 10 counts per second

7. Overtemperatu're AT See Note 1 See Note 3

8. Overpower AT See Note 2 See Note 3

9. Pressurizer Pressure--Low -> 1865 psig > 1855 psig

10. Pressurizer Pressure lligh < 2385 psig c'2395 psig

11. Pressurizer Water Level Iligh < 92K of instrument span < 93K of instrument span

12. Loss of Flow > 90$ of design flow > 89K of design flow per loop* per loop"

<<Design flow is 88,5OO gpm per, loop.

TADLE 2. 2-1 Continued REACTOR TRIP SYSTEM INSTRUMENTATIOH TRIP SETPOIHTS FUNCTIONAL UNIT TRIP SETPOIHT ALLOWADLE YALUES

13. Steam Generator Water > 10$ of narrow range instrument > 9X of narrow range instrument Level Low-Low span-each steam generator span-each steam generator 6 6

14. Steam/Feedwater Flow < 0.71 x 10 lb/hr of steam flow < .0.73 x 10 lbs/hr of steam flow .

Mismatch and Low Steam at RATED TIIERtIAL POWER coincident at RATED THERMAL POWER coincident Generator Water Level wfth steam generator water level with steam generator water level

> 25K of narrow range fnstru- > 24% of .narrow range fnstru-ment span--each steam generator ment span--each steam generator

15. Undervoltage-Reactor > 2750 volts-each bus > 2725 volts-each bus Coolant Pumps

16. Underfrequency-Reactor > 57.5 )iz - each bus. > 57.4. iiz - each bus Coolant Pumps

17. Turbine Trip A. Low Trip System > 000 psfg > 750 psig Pressure D. Turbine Stop Yalve > lX open > 1% open Closure

18. Safety Injection Input Hot Applicable Hot Applicable from ESF

19. Reactor Coolant Pump Not Applicable Hot Applicable Breaker Position Trfp

2.2 L Iiti i ~,~-'.ST'=.(

)

-1 !'l"S BAS=S 2.2.1 R~ZCTQR TRIP SYSTEM INSTRUCT ')TAT The Reactor Trip Setpoint Limits specified in Table 2.2-1 at.e the values at which the Reacior Trips are set for each parameier. Tne Trip Setpoin"s have been selecied to ensure thai the reactor core and reactor coolant sys iem are prevented rom exceeding :heir safety limits. Opera-tion with a trip set less conservative ihan its Trip Setpoin~ but wi ihin its sp cified Allowable 'lalue is acceptable on the basis that each Allowable 'lalue is equal to or less than ihe drift allowance assumed for each trip in the safety analyses.

t1anual Reac or Trio The Hanual Reac or Trip is a redurdant chanrel to ihe autcmatic pro.e" ive instrumentation channels and provides manual reactor trip capability.

Power Rance, 'neutron Flux The Power Range, tleuiron Flux channel high setpoin provides reactor core protection acainst reactivi y excursions which are too rapid io be proiec.ed by temperature <<rd pressure proteciive circuitry. Tne low set point provides redundant pro"ec ion in the power range for a power excursion be~inning fr"m low "ower. The trip associated with the low setpoint ",.ay be manually bypassed when P-10 is active (two of the four power range .channels inCica:e a power level o= above approx;mately 9 percent of RAT=D TH:-R.'!AL PG'il:-R) and is automatically reinstat d when P-10 becomes inac ive (three of the our channels indicate a ,".ower level below approximately 9 percent of RAT:-0 TH:-R'AL POWER) ~

Power Rance, .')eutron Flux, Hich Ra es criteria The Power Rang Positive Rate trip provides prot==c.ion agains rapid flux increases whIch are characteris"ic of red e'ec"ion events from any oower level. Specifically,- this trip comple.-,.en s "ihe .-"

wer Range '(eutron Flux High a.".d Low trips to ensure that he >are met =or rod ejection =". m partial power.

0. C. COOY - JNIT 1 8 2-3

Llt<ITIhG SAF""TY S'(ST- 1 Sc I TIelGS BAS""S The Power Range t(egative Rate Trip provides protection to ensure that the minimum 0 lBR is maintained above i.."0 ."r od dr op accidents.

At high power a single or multiple rod drop accid. nt could cause local flux peaking which when in conjunction with nuclear power being main-tained equivalent to turbine power by action o the automatic rod control sys em could cause an unconservative local 0,'lBR to exist. The Power Range,",egative Rate trip will prevent this from occurring by tripping the reactor for all single or multiple dropped rods.

Intermedia.e and Source Rance, luclear Flux The Inter...ediate and Source Range, t(uclear Flux tr'ps pr"vide reac or core pro:ec ion during reactor startup. These trips provide redundant protection "o the low se:point trip o= the Power Range, lieu".ron Flux channels. The Sourc Rance Channels will initiate a reactor trip at about 10.~ counts per second unless manually blocked <<hen P-~ "e "..-.. s active. The inter;..edia e Range Channels will ini ia.e a reactor trip at a current level proportional to approxima;ely 25 percent o= ~"T=O TH~R".4L PG'~"-.. unless manually blocked wnen P-'" "eco.;.es a tive. Ho credit was taken 7or operation o= the trips associa=ed with either the Inter,, dia;e or Source Rance Channels in the accid n aralys s; hcw ver, their ,unctional capability a th specified tr'.p settings is re.uired by tnis spe ificat'ion to enhance the overall reliability of the Reac.or Protection System.

Overtemoerature aT The Overtemperature tT trip provides core protection o prevent Di"B for ail combinations o= pressure, power, coolant temperature, and ax'.al power dis ri'"ut cn, provided tha the transi nt is slow wi h respec. "o piping transit delays ;rom the core to the te. "era.ure de ec=ors (abou-4 seconds), and pressure is within the range between the Hign and L".w Pressure reac.or trips. This setpoint-includes c"rrec=ions for c."ang=-s

$ n density and hea. capacity of wa"er with 'temperat re and dyrami= c:;;.-

pensation =or piping delays rom the core -to the looo tempera=ure detectors. hi th normal axial power distribution, this reac:"r tr',J limit is always below the core sa ety l.imit as shown in Figure 2.1-1.

If axial pe ks are greater than design as indica-. d by .here'cdif==-r=.-.c between tcp and bot=cm oower range nuclear detectors, the or .r:p automatically reduced according to the notations in Table 2.2-1.

's

0. C. COOK - UNIT 1 B 2-4

TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES n

1. Manual Reactor Trip Not Applicable Not Applicable

2. Power Range, Neutron Flux Low Setpoint - < 25K of RATED Low Setpoint - < 26K of RATED TllERt1AL POWER TllEBtQL POWER High Setpoint - < 109Ã of RATED High Setpoint - < llOX of RATED THERMAL f'OWER THERMAL POWER

3. Power Range; Neutron Flux, < 5X of BATED THER11AL POWER with < 5.5$ of RATED THERMAL POWER High Positive Rate a time constant > 1 second with a'ime constant > 1 second

4. Power Range, Neutron'Flux, + 3X of RATED THERt1AL POWER with -3 " .of RATED THERMA'OWEP, High Negative Rate a time constant > 1 second with a time constant > 1 second

5. Intermediate Range, Neutron < 25" of BATED THEBt1AL POWER < 30K of RATED TllERMAL POWER Flux 5

6. Source Range, Neutron Flux < 10 counts per second < 1.3 x 10 counts per second

7. Overtemperature AT See Note 1 See Note 3

8. Over power hT See Note 2 See Note 3 .

9. Pressurizer Pressure Low > 1950 psig > 1940 psig

10. Pressurizer Pressure High < 2305 psig < 2395 psig ll. Pressurizer Water Level High < 92Ã of instrument span < 93K of instrument span

12. Loss of Flow > 90'5 of design flow > 89K of design flow per loop* per loop*

~Design flow is 93,750 gpm per loop.

C)

E TABLE 2.2-1 Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

13. Steam Generator Mater > 155 of narrow range instrument > 14~ of narrow range instrument Level - Low-Low span - each steam generator span - each steam generator

14. Steam/Feedwa ter Flow < 1.47 x 10 6 lb/hr of steam flow < 1.56 x 10 6 lbs/hr of steam floe ,

Mismatch and Low Steam at RATED TtlENlAL POWER coincident at RATED TIIERtlAL POWER coincident:

Generator Water Level with steam generator water level with steam generator water le

> 25K of narrow range instru- > 24K of narrow range instru-oient span - each steam generator ment span - each steam generator 1

15, Undervoltage '- Reactor . > 2750 volts - each bus > 2725 volts - each bus 1 Coolant Pumps l6. Underfrequency - Reactor > 58.2 Hz - each bus > 58.1 l)z - each bus Coolant Pumps

17. Turbine Trip A. Low Trip System > 58 psig > 57 psig Pressure B. Turbine Stop Valve > 15 open < IX open Closure

18. Safety Injection Input Not Applicable Not Applicable from ESF

19. Reactor Coolant Pump Not Applicable. Not Applicable Breaker Posit(on Trip

2.2 LI'BAITING SAFETY SYST=l'1 SE TINGS 8ASES 2 2 1 REACTOR TR I P S YST Et'1 'INSTRUCT'lENTAT10.'I SETPO I NTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the values at which the Reactor Trips are set for each parameter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented rom exceeding their safety limits. Opera-tion with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acc ptable on tho basis that each Allowable Yalue is equal to or less than the drift allowance assumed for each trip ir the safety analyses.

Manual R actor Trio The manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

Power Range,,"(eutron Flux The Power Range. Neutron Flux channel high setpoint provides reactor core protection against reactivity excursions which are too rapid to be protected by te'perature and pressure protective circuitry. The low set point provides redundant protection in the po:;er range for a power excursion beginning from low power. The trip associated with the low setpoint . ay be manually bypassed when P-10 is active (two of +he four power range cnarnels indicate a pc: e~ level of above approximat ly 9 percent of RA EO THER.AL PG'HER) and is automatically reinstat d when P-10 beco'es inactive (three of the four channels indicate a power level below approximately 9 percent of RATEO.THER'%L PO'HER).

Power Rance, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteris ic of rod ejection events from any power level. Specifically, this trip complements the Power Range Neu ran Flux High and Low trips to ensure that the criteria are met for rod ejection from partial power.

0. C. COOK - UNIT 2 B 2-3

Llt)ITItlG SA, E Y SYSTE'l SETTINGS BASES The Power Range llegative Pate Trip provides protection to censure that the mini.-..um 9:,'"-R is maintaired above 1.30 for rod drop " =iderts.

AC hig) power a single or multiple rod drop accident could cause local flux p=-a~ing whic!i when in conjunction with nuclear oower being main-tain d equivalent to turbine pc;Ier by action of the automatic rcd control syste.-.~ could cause an unconservative local "NBR to exist. The Power R~nge !.'ega'ive Rate trio will prevent this frc.".: occurring by tripping the reactor for all single or multiple dropped rods.

Intermediate and Source Range 'luclear Flux The Intermediate and Source Range, Nuclear Flux trips provide .

reactor core protection during reactor startup. These trips provide redundan'rot ction to the low setpoint trip of tne Power Range, Neutron Flux channels. Tie Source Range Channels will initiate a reactor trip at about 10 counts per second unless manually bIocked when P-6 becomes active. The ?ntelmediate Pange Channels will initiate a reactor trip at a current level proportional to approx'rateTy 2~

percent, of RAT 0 THERMAL PQ'HER unless manually blocked when P-10 becomes active. ~'lo credit was taken for operation of the trips associated with either the Intermediate or Source Range Chanrels in the accident analyses; however, ti:eir furctional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Protection System.

Overte.".oera tore 'T The Overte.-..perature aT trip provides core protection to prevent CNB for all co.,birations of pressure, power, coolant temperatur, and axial power distribution, provided that the transient is slow with respect to p'iping transit delays from the core to the temperature detectors (about 4 seconCs), and pressure is within the range between tne High and Low Pressure reactor trips. This setpoint includes corrections for changes in density and heat capacity of water with temperature and dynamic com-pensation for piping delays from the core to the loop te.,oerature detectors. This reactor trip limit is always below the core safety limit as shown in Figure 2. 1-1. If axial peaks are more severe than design, as indicated by the difference between top and bottcm power range nuclear detectors, the reactor trip is automatically reduced according to the notations in Table 2.2-1.

r

0. C. COOK - UNIT 2 B 2-4

ATTACHMENT C TO AEP:NRC:00297 '

TABLE 1: TRANSIENT REANALYSIS RESULTS*

Transient MDNBR MDNBR (original analysis) (reanalysis)

Loss of RCS Flow (Coastdown of 4 pumps) 2.1 2.1 Loss of RCS Flow (Coastdown of 1 pump) 2.7 2.3 Startup of Inactive Loop Peak Power = 1.25 nominal Peak Power =. 1.30 nominal Loss of Load w/pressurizer spray and PORV's, BOL 2.5 2.5 Loss of Load w/pressurizer spray and PORV's, EOL 2.7 2.7 Loss of Load w/o pressurizer spray or PORV's, BOL 2.7 2.7 Loss of Load w'/o pressurizer spray or PORV's, EOL 2.7 2.7 Feedwater Control Malfunction 2.20 2.15

'10K Load Increase, BOL Manual Reactor Control 2.70 2.70 104 Load Increase, EOL Manual Rod Control 2.50 2.50 10Ã Load Increase, BOL Automatic Reactor Control 2.50 2.45 10Ã Load Increase, EOL Automatic Reactor Control 2.50 2.45

• Uncontrolled Rod Withdrawal at Power, RCCA misalignment and Steam Line Break were also re-evaluated and shown to be in accordance with applicable safety criteria.