ML20141F468

Amend 74 to License DPR-61,changing Tech Specs to Eliminate Refs to Application to Permit Plant Coastdown at End of Cycle 14

ML20141F468
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	04/14/1986
From:	Charemagne Grimes Office of Nuclear Reactor Regulation
To:
Shared Package
ML20141F467	List: ML20141F464 ML20141F468 ML20141F814
References
NUDOCS 8604230031
Download: ML20141F468 (25)

Text

. .

. f(

y ' ')

p, UNITED STATES NUCLEAR REGULATORY COMMISSION 5 ' ,, i a j WASHINGTON, D. C. 20555 S, a V

%, ...../

CONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NECK PLANT DOCKET N0. 50-213 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 74 License No. DPR-61

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

A. The application for amendment by Connecticut Yankee Atomic Power Company (the licensee) dated December 11, 1985 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public; and ('li) that such activities will be conducted in compliance with the Commissiori's regulations; D. The issuance of this amendment evill not be inimical to the common 7 defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the technical specifications as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of Facility Operating License No. DPR-61 is hereby amended to read as follows:

g"n8u8%ha P

4 4

a i ,

4 (2) Technical Specifications The technical specifications contained in Appendix A as revised.through Amendment No. 74, 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 C MiyAAO1. '

Christopher I. Grimes, Director Integrated Safety Assessment i' i Project Directorate Division of PWR Licensing - B

Attachment:

Changes to the Technical

' Specifications Date of Issuance: April 14,1986 i

h A

. . . ._ .~ -- .. - - _ - - . . _ .

t , .

ATTACHMENT TO LICENSE AMENDMENT NO. 74 FACILITY OPERATING LICENSE NO. DPR-61 i

h DOCKET N0. 50-213 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages as indicated. The revised pages are identified by the l captioned amendment number and contain vertical lines indicating the area l of change.

REMOVE INSERT 1-4 1-4 2-5 2-5 2-7 2-7 3-28 3-28

! 3-29 3-29 3-30 3-30 3-31 3-31 3-31a 3-31a 3-32 3-32 3-33 3-33 l 3-34 3-34

• i 3-34a 3-34a 3-35 3-35 3-36 3-36 Figure 3.18-la Figure 3.18-la l Figure 3.18-lb Figure 3.18-Ib l Figure 3.18-1c Figure 3.18-1c Figure 3.18-2a 1 -

Figure 3.18-2b  !

Figure 3.18-2c 3-39 3-39  !

! 3-40 3-40 i

l,

, 'Ye _

DEFINITIONS PRESSURE BOUNDARY LE AK AGE 1.16 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall:

CONTROLLED LE AK AGE 1.17 CONTROLLED LEAKAGE shall be that seal water return flow from the reactor coolant pump seals.

OUADR ANT POWER TILT R ATIO 1.18 The QUADRANT POWER TILT R ATIO shall be the ratio of the maximum quadrant power to the average quadrant power as determbed by the excore detector outputs.

DOSE EQUIVALENT I-131 1.19 Not used 1.20 Not used FREQUENCY NOTATION 1.21 The frequency notation specified for the performance of Surveillance Requirements sha!! correspond to the intervals defined in Table 1.2.

1-4 Amendment No. 74

e e o

2.4 Maximum Safety Settings - Protective Instrumentation Applicability: Applies to trip settings for instruments monitoring reactor power and reactor coolant pressure, temperature, and flow.

Obiective: To provide for protective action in the event that the principle process variables approach a safety limit.

Specification: Protective instrumentation trip settings shall be as follows:

Four Reactor Coolant Three Reactor Coolant Pumps Operating Pumps Operating Pressurizer Pressure 5 2300 psig i 2300 psig (1) 6 86% of range 6 86% of range (2) Pressurizer Level' (3) Variable Low Pressure *" 217.4 (Tavg+I 176T)-8850 > 17.4(Tavg+1.174T)-8850 Nuclear Overpower" S 109% of rated power $74% of rated power ]

(4)

(5) Low Coolant Flow"* 290% of normal loop flow 290% of normalloop flow (6) Reactor Coolant Loop Valve-Temperature f 200F 6200F Interlock High Steam Flow 110% of fullload steam  !!O% of fullload steam (7) flow flow

• May be bypassed when the reactor is at least 1.5%ok subcritical.

" The nuclear overpower trip is based upon a symmetrical core power distribution. If any asymmetric power distribution should occur, resulting in the power in any quadrant being 2% greater than the average core power as indicated by the neutron detectors, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> either a) reduce the QUADRANT POWER TILT RATIO to within its limit, or b) reduce thermal power at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.0 and similarly reduce the Nuclear Overpower trip setpoint within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

When the reactor power is <10% the overpower trip setpoint is reduced to 25% of rated power.

- May be bypassed below 10%,pf rated power.

~

Basis: The reactor protegive system is designed and constructed such that no single failure in any of the instrument systems will prevent the desired safety action if an applicable parameter exceeds a safety >

setpoint.

AmendmentNo.pf,74 2-5

and shutdown.

It is safe to block this trip below 10% power since the protection af forded by this t' rip is not required at this low level.

Removal of unnecessary spurious trips. trip signals will reduce the number of (4) Nuclear Overpower.

As explained above, the nuclear the variable overpower reactor trip, in conjunction 'with overpower, low pressure reactor trip, provides overtemperature pro tection.. The nuclear overpower trip channels will respond first to rapid reactivity insertion rates, detected by the increase in flux, before there are any significant changes in the system process variables. A instrumentation, maximum error of 9% of full power due to setpo and calcrimetric determination (see Section establishing 4.3.6 of the FDSA) is considered the setpoint. in {

1 In order to reduce the time to trip for certain accidents occurring at low  !

power, the overpower setpoint is lowered to 25%

when reactor power is below 10E This low overpower trip would terminate the postulated large steamline break accident from the hot zero power condition. The lower setting for three level equivalent to that provided boperation provides pro four loop operation at full power. y the setting for The reduction in setting in the event of an asymmetric ::cwer distribution provides protection for the more adverse hot channel factors.

f detected by observation of changesThe asymmetry is in : neutron i l . detector ion chamber current readings.

(5) 1.ow Coolant Flow. The low coolant flow reactor trip protects the core against an increase in coolant

{ temperature resulting from a reduction in cool flow while the reactor is at substantial powergt This trip will prevent DNB in any loss-of-flow .

incident, which eliminates the possibility of clad damage.

Flow detection in each reactor coolant loop is from a measurement of pressure drop from inlet to outlet of each steam generator. The 90%

low flow signal is high enough to activate a trip in time to prevent DNB, and low enough to reflect that a loss-of-flow condition truly exists.

A maximum instrument and setpoint error of f ~' full flow is considered in determining the setpomt. Loss-of-flow protection is also-provided by reactor coolant

, pump breaker and from undervoltage P

2-7 AmendmentNo.[,74

r , .

3.16 ISOTHERM AL COEFFICIENT OF REACTIVITY Aeolicability: Applies to the isothermal coefficient of reactivity for the core.

Obiective: To limit the maximum positive moderator coefficient that can exist in the reactor.

Soecification: The isothermal temperature coefficient as measured at l zero power shall be such that when corrected to operating conditions the value of the moderator temperature coefficient shall be:

1) Less positive than a calculated +0.5 x 10-4 delta k/k/0F for the all rods withdrawn, beginning of life (BOL), hot zero thermal power condition.

2) Less positive than a calculated 0.0 delta k/k/0F for the all rods withdrawn, beginning of life (BOL),

rated thermal power condition.

3) Less negative than a calculated -2.9 x 10-4 delta k/k/0F for the all rods withdrawn, end of cycle life (EOL), rated thermal power condition.

Basis: The transient analyses described in Section 10 of the Facility Description and Safety Analysis (FDSA) include a range of moderator temperature coefficients between

-3.5 x 10-4 per OF and +1.0 x 10-4 per CF. The specified range of coefficients from -2.9 x 10-4 per OF to

+0.00 x 10-4 per OF is bounded by these analyses.

3-28 Amendment No. 74

Deleted .

1 3-29 Amendment No. 74

4 3.17 LIMITING LINEAR HEAT GENER ATION R ATE Applicability: Applies to measured peak linear heat generation rate

, (Kw/f t)in the reactor core.

Objective: To establish limits on linear heat generation rate (Kw/f t) which are based on the postulated loss-of-coolant accident (LOCA) with (1) appropriate allowances for fuel densification and (2) a reanalysis of LOCA considering upper head fluid temperature equal to reactor vessel outlet temperature and (3) reanalysis considering an increased fuel pellet / clad gap and fuel rod pr e-pressurization and (4) reanalysis considering coastdown and (5) reanalysis considering steam generator tube plugging.

Specification: A. During steady state power operation, the peak linear heat rate values shall not exceed those limits shown below as defined in Reference (2) and modified by References (4)-(7). l

1. Cycle residency ume less than 3000 EFPH:

14.3 Kw/f t ]

i

2. Cycle residency time greater than or equal to 3000 EFPH but less than 6000 EFPH:  ;

14.5 Kw/f t - k

- 3. Cycle residency time greater than or equal to 6000 EFPH

~

15.5 Kw/f t i

l B. (Deleted)

I C. Measured values of core power peaking factors used l

' in determining measured linear heat generation I rates in the Section A specification shall include I I

allowances for the following:

l

1. Normal power peaking. I

- 1 e

e 3-30 Amendment No. 58, t - - - -

O

2. Flux peaking augmentation factors (Power Spike), using Figure 3.17.1.

3. Measurement uncertainty (1.05).

4. Statistical density f actor (1.012).

5. Engineering factor (1.02).

6. Stack shortening / thermal expansion factor {

(1.007).

7. Power level uncertainty (1.02).

These factors are multiplicative and items 1 and 2 shall be chosen at a core height so as to maximize )

their product. l D. Three loop operation at above 65% of full licensed power shall not be permitted until additional analysis is performed and proposed technical specification changes submitted. During three loop operation, the (Kw/f t) limits of Specification 3.17.A shall be multiplied by 0.65.

Basis: Specification A sets limits that assure the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the 23000F limit specified in the Interim Acceptance Criteria 'UAC) issued in June,1971, cor.sidering the postulated effects of fuel densification.

These limits incorporate the results of a recalculation of the LOCA event considering upper head fluid temperature equal to the reactor vessel outlet temperature (THOT), and an increased

}

pellet / clad gap and fuel rod pre-pressurization incorporated in Batch XIV and subsequent reloads. ,

I I

The basis for the multiplicative factors in Specification C are described in Section 8.1 of Reference (2). The engineering factor has been reevaluated based on current manufacturing tolerances. This factor includes the effects of variations on pellet diameter, density and enrichment and fuel rod diameter.

The basis for Specification D is that previous analysis (Ref. 3) '

submitted December 5,1972, showed the acceptability of.75%. l cf rated power for three loop operation. An additional 10% , l margin in rated power was added as a conservative allowance i for fuel densification.

3-31 Amendment No. ,T, 74

References:

(1) WC AP-8213, Effects of Fuel Densification on the Connecticut Yankee Reactor, October,1973.

(2) Description and Safety, including the Effects of Fuel Densification on the Connecticut Yankee Reactor, Cycle V, CY APCO, Novernber,1973.

(3) Switzer, D. C.

to Assistant Director for Operating Reactors, USAEC, CYAPCO letter of December 5,1972.

(4) D. C. Switzer. letter to A. Schwencer (USNRC), dated May 2,1977. -

(5) D. C. Switzer letter to A. Schwencer (USNRC), dated October 31,1977.

(6) W. G. Counsil letter to D. M. Crutchfield (USNRC), dated December 14,1932.

(7) W. G. Counsil letter to D. M. Crutchfield (USNRC), dated March 30,1984.

(8) J. F. Opeka letter to C. I. Grimes (USNRC), dated Decem!>er 11, 1985.

l l

l 3-31a Amendment No. , 74

4 l 3.18 POWER DISTRIBUTION MONITORING AND CONTROL Aeolicability:

Applies to monitoring of the reactor power distribution as a function of reactor power level and control rod position. This specification is based upon the control rod bank "B" consisting of eight (S) RRC's 13,30,31,32 consisting of rod position numbers 10,11,12, and 33.

Obiective:

To assure operation within limits as specified in Technical Specification 3.17.

l Soccification: A.

Full Core Power Distribution Measurements

1. Full core power distributions shall be measured at I

least once per full power month during normal power operation, using the movable incore neutron detector system.

2. Prior to commencement of full power operation for each new operating cycle, denoted by a core refueling, a full core power distribution measurement will be performed before exceeding 80% of rated power using the movable incore neutron detector system. These measurements will be adjusted to 100% of rated power and evaluated for compliance with Specification 3.17 limits prior to increasing power to 100% of rated power.

3. These measurements shall be repeated at 100%

power equilibrium xenon conditions approximately 40 full power hours after initially reaching 100%

power in each new operating cycle. These measurements will be evaluated in a time period not to exceed 5 days after satisfactory completion of the measurements. If measured values are found to exceed Specification 3.17 limits, core thermal power shall be reduced by the percent exceeded until compliance can be demonstrated.

B. Power Distribution Monitoring

1. Periodic surveillance of the core power distribution shall be performed above 40% of rated power as described below:

3-32 Amendment No. M 74 I

l l

1.1 Determination of the axial offset by use of excore power range detectors.

1.1.a The incore axial offset shall be continuously monitored using at least two calibrated power range channels and verified, using incore/excore correlation to be within the power dependent envelope specified in Figure l

3.18-1 or 3.18-2 with an appropriate allowance made for excore calibration l uricertainty, which shall not be less than +3%

in excore axial offset units.

1 1.1.b Should the axial offset be found to exceed the operating limit curves for 4 loop operation (Figures 3.18-1) or 3 loop operation (Figures 3.18-2), corrective action via control I rods or power reduction shall be taken, and the axial offset returned to within the  ;

operating band, unless it can be verified by the Technical Specification 3.18.B.I .2 l procedure that the Technical Specification 3.17 limits are not being violated.

1.1.c The excore-incore correlations shall be )

verified every full power month using the results from the full core power distribution ,

map specified in 3.18.A.I. Every three full '

power months, the excore-incore axial offset correlation will. be verified anu adjusted based upon results from the movable incore neutron detector system. The excore-incore

• correlation shall be initially checked with each change in fuel configuration and af ter a major change in excore instrumentation using results of incore measurements specified in 3.18.A.2 and 3, and/or supplemental core power maps. The excore detectors will be calibrated / correlated within 7 days after the satisfactory completion of the incore measurements.

+

3-33 Amendment No. g 74 M g .

s . s .

, o , s l.l.d Deleted .

1.2 incore detector measurements in two (2) thimble locations may be used to monitor the power dis tribution:

1.2.a An appropriate correlation specific to the two selected thimbles is applied to their .

measured axial power distributions so as to determine a total core peaking factor to assure compliance with Specification 3.17.

The thimble location which yields the higher total core peaking factor shall be used for verification of the Specification 3.17 limits.

If these limits are found to be exceeded, the core thermal power shall be reduced by the percent exceeded until compliance with the 3.17 limits can be demonstrated.

1.2.b The correlation factor applied to these  ;

thimble measurements shall be based upon '

previous full core power distribution measurements within the cycle with the movable incore detector system and the correlation shall be checked every full power month during normal power operation.

1.2.c The frequency of incore detector power distribution monitoring shall be once per 8-hour shift for the steady state operation, and at least once per hour for periods of non-steady state . operation. For this specification, " steady state operation" shall be considered at those periods of operation for which the reactor has operated above 80% of rated power for a continuous period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, with control rod bank "B" at a height no less than 270 steps. "Non-steady state operation"is defined as those periods of operation not complying with, the above definition.

• e n

3-34 Amendment No. 74

. ~

b C. Control Rod Insertion Limits

1. Except for lower power physics tests at or below 10% of full power or determination of "just critical" rod positions, operation of the control group banks shall be maintained above the limits shown in Figure 3.10-1.

2. The monthly average position of control rod bank "B" shall be at least 280 steps withdrawn when above 20% of rated thermal power when weighted l by the dally average thermal energy generation.

This requirement shall be evaluated twice per month.

e e

/

O e

3-34a Amendment No. 23,

Bases: A.!

Use of the movable incore neutron detectors provides an accurate power means for determination distributions when evpluated of three dimensional using the Westinghouse INCORE program.(!J measurements are used to check Results of these compliance with Technical Specification 3.17 limits.

A.2 Full core power distribution measurement at or below 80% of rated power for the startup of each new operating cycle provides verification of design predictions to assure compliance 100% of rated with core thermal limits before proceeding to power.

A.3 An additional full core power distribution measurement at 100% of rated power just after reaching equilibrium xenon conditions for each new operating cycle provides further verification of the acceptability of the power distribution.

B.1 i Monitoring below 40% of rated power is considered unnecessary because of the substantial margin in local fuel rod heat flux at this reduced power.

B.1.1.a Provides continuous monitoring of the incore pcwer distribution by means of the out-of-con- power range detectors.

Operation within the axial offset envelopes of Figure 3.18-1 and Figure 3.18-2 assures that the local heat flux will not exceed the peak l linear heat . rate limits Specification 3.i7. Specific axialdefined in offset versus power curves for future reloads will be calculated and checked against Figure 3.18-1 and Figure 3.18-2 to insure continued applicability. l B.1.1.b An appropriate allowance for incore/excore calibration uncertainty is used by the reactor operator. The bases for the excore detector calibration and its uncertainty are described in Appendix B to Reference 3.

B.1.1.c Monthly checks and calibrations every three full power months assure maintenance of the excore detector calibration. Provides for axial offset monitoring calibration after a new core loadin instrumentation. g or changes in nuclear B.I.2 Use of two movable incere thirable measurements along with an appropriate correlation converting the measurement axial l 3-35 Amendment No. 74

~ . - .

peaking farers to F$ provides an alternate means of v ..~ication of compliance with the Technical Specification 3.17 limits. For conservatism, the highest F$ determined will be used. The correlation is based upon full core power distribution measurements and has allowances for measurement uncertainties and the spatial effects of control rod insertions. A frequency of once per 8-hour shift is considered adequate because core power distributions do not change substantially during this mode of operation. A frequency of at least once per hour during non-steady state operation is to monitor for control rod and xenon induced power peaking.

C.2 This specification limits the potential for unfavorable axial power distributions due to operation for long burnup intervals at deep control rod insertions.

References:

(1) WCAP-7149, Leggett, W. D. and Eisenhart, L. D., The INCORE Prorram, Westinghouse Electric Company, December,1967.

(2) Technical Report Supporting Cvele VI Operation and Proposed License Amendments, Docket 30-213, May, 1973.

(3) " Axial Offset Monitoring including Revised Control Rod Grouping for the Connecticut Yankee Reactor, Cycle V,"

Docket No.30-213, August 1974.

(4) D. C. Switzer (CYAPCO) letter to Director of Nuclear Reactor Regulation (NRC), dated October 23,1975.

e e

P a

h e

3-36 Amendment No. 74 w n -, . . . ,,- ,, - - - , , . - - - , - - e r-, , - - ,r,,- - - - - , . , - , - , , - , , , - , , - - - - -

CL O

O J

C D

O E

l i

__o 1 l

4 o

~ l s s ,

E o

N o

o

- o o

o E E m . .

__oa l

CD

r. o o o o e o e o a

~S H w

cn - E o ~ o = ,o n n - en ,

I t f f f f f f f g W 4 e i e 4 e i e i

• w EO 3g W3 mod % ---

o O

t 0 L -

EW o o

2*

D "

N & - *'

T 1 -

o o o O n w --w m d d H 9 9 o g _ _

U)

E E

O so C

m --

N O

cL . .

Amendment No.,5 74

POWER vs OFFSET,12p250 EFPD, FOUR LOO'P .

100 -

. l Fii166)/ (10,100)

. j -

- e0

-s0

(-30,70) -

-70 (20,70) iE--60 W

3

,O--50 n.

(-30,40) *- -40 (20,40)

--30

-- 20 g -

- 10 k

a i f f I I I I I I I i i i i i i i i i i 40 20 -10 0 10 20 30 40 50 E

% OFFSET O

=

. ~~

FIGURE 3.18-1c -

POWER vs OFFSET,250 EFPD-EOC, FOUR LOOP j ion

, (-20,100) (10,100) 4 I

90

--s0

6:10,76) -- 70 (20,7o)

i

g. - - 60 su .

I-- 50

a. '

(-30,40) *--40 (20,40)

) *

- 30

- 20 j ifm -

-10 s

y h* l l l l l l l l l l

, . . . . i i i i i i i i

5 40 20 -10 0 10 20 30 40 50

! = .

.  % OFFSET e

N b

FIGURE 3.18-2a . .

POWER vs OFFSET,0-125 EFPD , THREE LOOP '

- - 100 .

-n

- 80

-7'

(-7,es) (10,s5)

- 80

.E

.w-- 50

(-30,4s.s) 3 (20,4s.5) 2--40

- 30

(-30,2s) (20,26) '

-- to

-10 g i I e i

e i e t t

, i i i i i i i i i g -so 30 10 o to 20 30 40 so a

,  % OFFSET E

~

i .. N ,

4

P O

O -

L E

E 3i o s

R H

) 0 T s. ) ti 4

' s s

, 4, 2 0

" D )

s 0

2 0

2 Ii 3

'b P

( (

s, 2 F 0 0 1

fi 2

- E (

8 .

o 1 0 0 fi t T 5 0 ,

, 0 ' 0 0 0 0 O 0 E 3 1 0 ' 8 5 4 3 S 1 S E. 2- -

~

~

- i. -

~

~

0 F F

R 25  :

i:W3oa #

g

' O U 1 ~-

Ii 1 0

I G ,

~) -

0 F ,ET 5 5

Ii 2 i .

S i

. Ie 0

3 F ( .

.)

F 5

.)

.s , 0 O s 4,

2, 0

fi 4 s 0 3

- 0 v .(

3

( Ii 5 R

E W

O P

Fe3gge g ;

i ,i

! #  :!} . l $ :; ; < 1' .l Ii :?'.1

e FIGURE 3.18-2c POWER vs OFFSET,250-EOC, THREE LOOP '

1 4

- - 100

- e0 i

,t -

- s0

(-20,es)

_n

, (to,es) g- - 80 w

3- - s0

(-30,8.s) , 2 (20,4s.s)

1. --40 s

j

. .. .. -- 30

(-30,26) (20,26) '

- 20 I

- 10 i

! I I I I I l I f f I i g i i i i i i i i i i o.

a o

40 20 -10 0 10 20 30 40 50

! 3

% OFFSET g

• I y .

I 2

i

. ~

4 3.20 Reactor Coolant System Flow, Temocrature and Pressure

~

~ ' Aeolicability: Applies to MODE 1. steady state operation with four reactor coolant loops operating.

Obiective:

To set limiting conditions for operation for minimum nominal reactor coolant flow and pressure and maximum inlet temperature.

Soecification

A. Reactor Coolant Flow Rate

1) RCS Flow Rater 257,000 gpm i

B. Reactor Coolant Temperature

1) T inlet f 540.60F C. Reactor Coolant Pressure

1) Pressurizer Pressure 22000 psig+

' D.

The RCS flow rate shall be determined by a

.: heat balance within 7 EFPD of achieving 100% RATED THERMAL POWER after refueling.

E. Following the completion of Section D above, the above parameters shall be verified to be within the 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 />. If any of the

, above parameters exceeds its specified limh ,

restore the parameter to within its above specified limit within two fiours or reduce THERMAL POWER to less than 5% 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 />.

Basis:

The JJmiting conditions for operation have been expanded to include limits on flow, Inlet temperature and pressure.

The flow . rate requirements are based on a steam generator plugging / sleeving level consistent with a maximum of 500 equivalent plugged tubes per steam generator.

Additionally, an evaluation has been i

performed to reduce the core bypass flow fraction from 9% to 4.5%. The reactor vessel flow rate decreases due to the assumed steam generator equivalent pIugging bypass, flow. levels but *' the core flow increases due to the reduction

~

This limit is not appljeable during either a THERMAL POWER ramp in  !

POWER step in excess of 10% of RATED THERM AL POWER. exces{ ~

)

3-39 Arnendment No. 74

4 The core inlet temperature of $40.60F includes a 40F instrument error and deadband which would allow a maximum core inlet temperature of $44.60F at 100% power. The maximum inlet temperature of 544.60F is used in all current safety analyses, with the exception of the dropped rod analysis which used 533.90F Tin. Sensitivity studies show that ' increasing the steady state temperature 10.70F from 333.90F to 544.60F will result in a small reduction in minimum DNBR. Starting the rod dropped accident from a 10.70F hotter.

condition will yield an increase in the end point temperature. If the dropped rod accident had been analyzed for 544.60F, the maximum DNBR would be well above the fuel design limit of 1.3.

The minimum reactor coolant pressure of 2000 psig assumes + 30 psig for instrument error and deadband which would allow a minimum core pressure of 1970 psig at 100% power.

The limiting values of the parameters in this specification are equal to, or more conservative than those assumed as the initial conditions in the accident and transient analyses; therefore, operation must be maintained within the specified limits for the accident and transient analyses to remain valid.

3-40 Amendment No. , 74