Text

Proposed Tech Specs Clarifying Testing of Log Power Level Trip & Allowing Reduction of Shutdown Cooling Flow in Mode 6 to Minimize Potential for Loss of Shutdown Cooling Due to Vortexing
	ML20149M825
Person / Time
Site:	Waterford
Issue date:	01/28/1988
From:	; LOUISIANA POWER & LIGHT CO.
To:	;
Shared Package
ML20149M822	List: ML20149M825; W3P88-0160, Application for Amend to License NPF-38,changing Tech Specs to Clarify Testing of Log Power Level Trip & to Allow Reduction of Shutdown Flow in Mode 6 to Minimize Potential for Loss of Shutdown Due to Vortexing.Fee Paid;
References
	NUDOCS 8802290113
	Download: ML20149M825 (21)
	v • d • e

4 8 6 i

NPF-38-73 ATTACHMENT A l

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l 8802290113 880128 PDR ADOCK 05000382 l

P PDR l

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NS41376 l

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t TABLE 3.3-1 D

REACTOR PROTECTIVE INSTRUMENTATION E

l MINIMUM B

TOTAL NO.

CHANNELS CHANNELS APPLICABLE

)

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION C

i 5

1.

Manual heactor Trip 2 sets of 2 1 set of 2 2 sets of'2 1 2 1

A i

2 sets of 2 1 set of 2 2 sets of 2 3, 4*, 5*

8 u

2.

Linear Power Level - High 4

2 3

1, 2 2#, 3#

3.

Logarithmic Power Level-High a.

Startup and Operating 4

2(a)(d) 3 1 2 2#, 3#

4 2

3 3, 4*, 5*

8 b.

Shutdown 4

0 2

3, 4, 5 4

4.

Pressurizer Pressure - High 4

2 3

1, 2 2#, 3#

I 5.

Pressurizer Pressure - Low 4

2(b) 3 1, 2 2#, 3#

w 6.

Containment Pressure - High 4

2 3

1, 2 2#, 3#

J, 7.

Steam Generator Pressure - Low 4/SG 2/SG 3/SG 1, 2 2#, 3#

8.

Steam Generator Level - Low 4/SG 2/SG 3/SG 1, 2 2#, 3#

9.

Local Power Density - High 4

2(c)(d) 3 1, 2 2#, 3#

10. DNBR - Low 4

2(c)(d) 3 1, T 2#, 3#

)

11. Steam Generator Level - High 4/SG 2/SG(g) 3/SG 1, 2 2#, 3#

12. Reactor Protection System Logic 4

2 3

1 2 5

4 3, 4*, 5*

8 1

13. Reactor Trip Breakers 4

2(f) 4 1 2 5

<k E

3,

4*, 5*

8 z

14. Core Protection Calculators,

4 2(c)(d) 3 1, 2 2#, 3# and 7 5

15. CEA Calculators 2

1 2(e) 1, 2 6 and 7

,8

16. Reactor, Coolant Flow - Low 4/SG 2/SG 3/SG 1, 2 2#, 3#

w

TABLE 3.3-1 (Continued)

TABLE NOTATION 3

A With the p~otective system trip breakers in the closed position, the CEA drive system capable of CEA withdrawal, and fuel in the reactor vessel.

The provisions of Specification 3.0.4 are not applicable.

(a) Trip may be manually bypassed above 10'h of RATED THERMAL POWER; bypassshal]4be automatically removed when THERMAL POWER is less than or equal to 10 % of RATED THERMAL POWER.

(b) Trip may be manually bypassed below 400 psia; bypass snall be auto-matically removed whenever pressurizer pressure is greater than or equal to 500 psia.

(c) Trip may be manually bypassed below 10 h of RATED THERMAL POWER;

~

bypass shall bg automatically removed when THERMAL POWER is greater than or equal to 10 of RATED THERMAL Pa.'ER.

During testing pursuant to Special Test Exception 3.10.3, trip may be manually bypassed below 5%

of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is. greater than or ecual to 5% of RATED THERHAL POWER.

(d) Trip may be bypassed during testing pursuant to Specis' Test Exception 3.10.3.

1 (e) See Special Test Exception 3.10.2.

(f) Each channel shall be comprised of two trip breakers; actual trip logic shall be one-out-of-two taken twice.

(g) High steam generator level trip may be manually bypassed in Modes 1 and 2, at 20% power and below.

l ACTION STATEMENTS ACTION 1 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the protective system trip breakers.

With the number of channels OPERABLE one less than the Total ACTION 2 Number of Channels, STARTUP and/or POWER OPERATION may continue provided the inoperable channel is placed in the bypassed or tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

If the inoperable channel is bypassed, the desirability of maintaining this channel in the bypassed condition shall be reviewed in accordance with Specification 6.5.1.6k.

The channel shall be retumed to OPERABLE status prior to STARTUP following the next COLD SHUTOOWN.

WATERFORD - UNIT 3 3/4 3-4 AMENDHENT NO 14

i TABLE 4.3-1 REACTOR P'!OTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4

o CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE l

iE FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED

'Q 1, 2, 3*, 4*, 5*

w 1.

Manual Reactor Trip N.A.

N.A.

R and S/U(1) 2.

Linear Power Level - High S

D(2,4),M(3,4), M 1, 2' Q(4) 3.

Logarithmic Power Level - High 5

2(4)

M and S/U(1) 1,2,3,4,5 4.

Pressurizer Pressure - High 5

R M

1, 2 5.

Pressurizer Pressure - Low S

R M

1, 2 w

i1 6.

Containment Pressure - High S

R M

1, 2 w*

7.

Steam Generator Pressure - Low 5

R M

1, 2 8.

Steam Generator Level - Low S

R M

1, 2 l

9.

Local Power Density - High 5

D(2,4),R(4,5) M,R(6) 1, 2 I

i 10.

DNBR - Low S

S(7),0(2,4),

M,R(6) 1, 2 M(8),R(4,5) 11.

Steam Generator Level - High S

R M

1, 2 1

12.

Reactor Protection System Logic N.A.

N.A.

M and S/U(1) 1, 2, 3*, 4*, 5*

1

i TABLE 4.3-1 (Continued)

TABLE NOTATIONS

With the reactor trip breakers in the closed position, the CEA drive system capable of CEA withdrawal, and fuel in the reactor vessel.

(1) Each startup or when required with the reactor trip breakers closed and the CEA drive system capable of rod withdrawal, if not performed in the previous 7 days.

(2) Heat balance only (CHANNEL FUNCTIONAL TEST not included), above 15%

of RATED THERMAL POWER: adjust the Linear Power Level signals and the CPC addressable constant multipliers to make the CPC AT power and CPC nuclear power calculations agree with the calorimetric calculation if absolute difference is greater than 2%.

During PHYSICS TESTS, these daily calibrations may be suspended provided these calibrations are performed upon reaching each major test power plateau and prior to proceeding to the next major test power plateau.

(3) Above 15% of RATED THERMAL POWER, verify that the linear power sub-channel gains of the excore detectors are consistent with the values used to establish the shape annealing matrix elements in the Core Protection Calculators.

(4) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(5) After each fuel loading and prior to exceeding 70% of RATED THERMAL POWER, the incore detectors shall be used to determine the shape annealing matrix elements and the Core Protection Calculators shall use these elements.

(6) This CHANNEL FUNCTIONAL TEST shall include the injection of simulated process signals into the channel as close to the sensors as practicable to verify OPERABILITY including alarm and/or trip functions.

(7)

Above 70% of RATED THERMAL POWER, verify that the total RCS flow rate as indicated by each CPC is less than or equal to the actual RCS total flow rate determined by either using the reactor coolant pump differential pressure instrumentation or by calorimetric calculations and if necessary, adjust the CPC addressable constant flow coefficients such that each CPC indicated flow is less than or equal to the actual flow rate.

The flow measurement uncertainty is included in the BERR1 ters in the CPC and is equal to or greater than 4%.

~

(8) Above 70% of RATED THERMAL POWER, verify that the total RCS flow rate as indicated by each CPC is less than or equal to the actual RCS total flow rate determined by calorimetric calculations.

(9) The monthly CHANNEL FUNCTIONAL TEST shall include verification that the correct values of addressable constants are installed in each OPERABLE CPC, (10) At least once per 18 months and following maintenance or adjustment of the reactor trip breakers, the CHANNEL FUNCTIONAL TEST shall include independent verification of the undervoltage trip function and the shunt trip function.

WATERFORO - UNIT 3 3/4 3-12 AMENDMENT NO. 5 l

I NPF-38-73 ATTACHMENT B

)

I i

NS41376

TABLE 3.3-1 g-g REACTOR PROTECTIVE INSIRUMENTATION o

MINIMUM E

TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION C5 1.

Manual Reactor Trip 2 sets of 2 1 set of 2 2 sets of'2 1 2 1

A 2 sets of 2 1 set of 2 2 sets of 2 3, 4*, 5*

8 w

2.

Linear Power Level - High 4

2 3

1, 2 2#, 38

,h I

, ~3 d

- ~M d' 3.

Logarithmic Power Level-High a.

Startup and Operating 4

2(a)(d) 3 f Q1 2#, 3#

A 4

2 3

3, 4*, 5*

8 b.

Shutdown 4

0 2

3, 4, 5 4

~

4.

Pressurizer Pressure - High 4

2 3

1, 2 2#, 3#

5.

Pressurizer Pressure - Low 4

2(b) 3 1, 2 2#, 3#

ws*

6.

Containment Pressure - High 4

2 3

1, 2 2#, 3#

w J,

7.

Steam Generator Pressure - Low 4/SG 2/SG 3/SG 1, 2 2#, 3#

8.

Steam Generator Level - Low 4/SG 2/SG 3/SG 1, 2 2#, 3#

9.

Local Power Density - High 4

2(c)(d) 3 1, 2 2#, 3#

10. DN8R - Low 4

2(c)(d) 3 1, 2 2#, 3#

11. Steam Generator Level - High 4/SG 2/SG(g) 3/SG 1, 2 2#, 3#

l

12. Reactor Protection System Logic 4

2 3

I 2 5

3g, 4 *, 5*

8 1

13. Reactor Trip Breakers 4

2(f) 4 1 2 5

k A

3, 4*, 5*

8 h

14. Core Protection Calculatcrs,

4 2(c)(d) 3 1, 2 2#, 3# and 7 k

15. CEA Calculators 2

1 2(e) 1, 2 6 and 7

,8

16. Reactor, Coolant Flow - Low 4/SG 2/SG 3/SG 1, 2 2#, 3#

1 TABLE 3.3-1 (Continued)

TABLE NOTATION AWith the protective system trip breakers in the closed position, the CEA drive system capable of CEA withdrawal, and fuel in the reactor vessel.

p $The-provisions _of. Specification 3.0.4. are not.

licable.

<h2 Nw lyMAf4 by: ~/O' X id-w '%41 I; n'.

j (a) ' Trip may be manual 1y bypassed above 10 of RATED THERMAL POWER; bypass shal]h of RATED THERMAL POWER.be automatically removed when THERMAL P

[j equal to 10 (iC (b) Trip may be manually bypassed below 400 psia; bypass shall be auto-matically removed whenever pressurizer pressure is greater than or equal to 500 psia.

~4 (c) Trip may be manually bypassed below 10 % of RATED THERMAL POWER; bypass shall bg. automatically removed when THERMAL POWER is greater than or equal to 10 '% of RATED THERMAL POWER.

During testing pursuant to Special Test Exception 3.10.3, trip may be manually bypassed below 5%

of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is greater than or equal tc 5% of RATED THERMAL POWER.

(d) Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3.

(e) See Special Test Exception 3.10.2.

(f) Each channel shall be comprised of two trip breakers; actual trip logic shall be one-out-of-two taken twice.

(g) High steam generator level trip may be manually bypassed in Modes 1 and 2, at 20% power and below.

ACTION STATEMENTS ACTION 1 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the protective system trip breakers.

ACTION 2 With the number of channels OPERABLE one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may continue provided the inoperable channel is placed in the bypassed or tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

If the inoperable channel is bypassed, the desirability of maintaining this channel in the bypassed condition shall be reviewed in accordance with Specification 6.5.1.6k.

The channel shall be returned to OPERABLE status prior to STARTUP following the next COLD SHUT 00VN.

WATERFORD - UNIT 3 3/4 3-4 AMENDMENT NO. 14

TABLE 4.3-1 y

REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS 2

CHANNEL M00ES Frit MtICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE j

FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED 1.

Manual Reactor Trip N.A.

N.A.

R and S/U(1) 1, 2, 3*, 4*, 5*

2.

Linear Power Level - High S

D(2,4),M(3,4), M 1, 2' Q(4)

Afw-dro 3.

Logarithmic Power Level - High S

R(4)

M and S/U(1) ' J 4, 5 4.

ressurizer PreIs r-) - High S

R M

1, 2 5.

Pressurizer Pressure - Low S

R M

1, 2 6.

Containment Pressure - High S

R M

1, 2 y

7.

Steam Generator Pressure - Low 5

R M

1, 2 8.

Steam Generator Level - Low S

R M

1, 2 9.

Local Power Density - High S

D(2,4),R(4,5)

M. R(6) 1, 2

10. DNBR - Low S

S(7),D(2,4),

H,R(6) 1, 2 M(8),R(4,5) 11.

Steam Generator Level - High 5

R M

1, 2 12.

Reactor Protection System Logic N.A.

N.A.

M and S/U(1) 1, 2, 3*, 4*, 5*

v

TABLE 4.3-1 (Continued)

TABLE NOTATIONS

With the reactor trip breakers in the closed position, the CEA drive system capable of CEA withdrawal, and fuel la the reacter vessel.

(1) Each startup or when required with the reactor trip breakers closed and the CEA drive system capable of rod withdrawal, if not performed in the previous 7 days.

(2) Heat balance only (CHANNEL FUNCTIONAL TEST not included), above ?.5%

of RATED THERMAL POWER: adjust the Linear Power Level signals and the CPC addressable constant multipliers to make tne CPC AT power and CPC nuclear power calculations agree with the calorimetric calculation if absolute difference is greater than 2%.

During PHYSICS TESTS, these daily calibrations may be suspended prcvided these calibrations are performed upon reaching each major test power plateauandpriortoproceedingtothenextmajortestpowerplatea.s.

(3) Above 15% of RATED THERMAL POWER, verify that the linear power sub-channel gains of the excore detectors arc consistent with the values used to establish the sh' ape annealing matrix elements in the Core i

Protection Calculators.

(4) Neutroit detectors may be excluded from CHANNEL CALIBRATION.

(5) After each fuel loading and prior to exceeding 70% of RATED THERMAL POWER, the incore detectors shall be u,ed to determir.e the shape annealing matrix elements and the Core ?rotection Calculators shall use these elements.

l (6) This CHANNEL FUNCTIONAL TEST shall include the injection of simulated process signals into the channel as close to the sensors as practicable j

to verify OPERABILITY including alarm and/or trip functions.

(7), Above 70% of RATED THdidL POWER, verify that the total RCS flow rate as indic3ted by each CPC is less than or equal to the actual RCS total flow rate determined by either using the reactor coolant p'

l pump differential pressure instrumentation or by calorimetric calculations and if necessary, adjust the CPC addressable constant

~

'k flow coefficients such that each CPC indicated flow is less than or equal to the actual flow rate.

The flow measurement uncertainty is N

included in the BERR1 term in the CPC and is equal to or greater t

than 4%.

(8) Above 70% of RATED THERMAL POWER, verify that the total RCS flow rate as indicated by each CPC is less than or equal to the actual RCS total flow rate determined by calorimetric calculations.

i l

(9) The mon'thly CHANNEL FUNCTIONAL TEST shall include verification that the correct values of addressable constants are installed in each j

OPERABLE CPC, (10) At least once per 18 months and following maintenance or adjustment of the reactor trip breakers, the CHANNEL FUNCTIONAL 1EST shall include independent verification of the undervoltage trip function and the shunt trip function.

-'cj:i'c';w.': A S 02% O IAL^ C~ Il Y ' NA

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E b.em, hm. p e w re b 7 % e:. */ J ' S u n m h a luKymapw bwlGw & & fied wpi W.)

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WATERFORD - UNIT 3

~

3/4 3-12' MiENDMENDO. 5 l

I l

4 6

4 s

NPF-38-74

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0 DESCRIPTION AND SAFETY ANALYSIS 4

0F PROPOSED CHANGE NPF-38-74 3

This is a request to revise Technical Specifications 3.9.8.1 and 3.9.8.2,-

Refueling Operations, Shutdown Cooling and Coolant Circulation.

Existina Specifications See Attachment A Proposed Specifications See Attachment B t

Description The Surveillance Requirements of Technical Specifications 3.9.8.1 and 3.9.8.2 require at'least one shutdown cooling (SDC) train be in operation and circulating reactor coolant at an initial flow rate of at least 4000 gpm. At 175 hours0.00203 days 0.0486 hours 2.893519e-4 weeks 6.65875e-5 months following reactor shutdown, the minimum SDC flow may be reduced to i

3000 gpm. The proposed change would allow an. additional SDC flow reduction to 2000 spm, 375 hours0.00434 days 0.104 hours 6.200397e-4 weeks 1.426875e-4 months after reactor shutdown.

Backaround The impetus for the proposed change was the Waterford 3 review of Generic Letter 87-12, "Loss of Residual Heat Removal (RHR) While the Reactor Coolant l

System (RCS) is Partially Filled". The. Genaric Letter, and previous industry and NRC publications, noted the effect of SDC flow rate upon the potential for vortexing at the ' connection of the SDC suction line to the RCS when the RCS is partially drained.

In general, as the SDC flow rate increases, the potential fur vortexing (and subsequent 1.oss of SDC through air-binding the SDC puep) also incresses.

i The need to reduce SfC flow to avoid vortexing is balanced by the SDC System design basis to remove decay heat. As noted in the Technical Specification Bases, SDC flow serves two purposes:

1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140'F during Mode 6, and 2) sufficient coolant circulation is maintaineA through the reactor core to minimize the effects of a bcron j

dilution 1:11 dent and prevent boron stratification.

1 i

l l

l 1

l NS20707 i

l -

r In response to Generic Letter 87-12 (W3Pb7-1775 dated September 21,1987),

Waterford 3 committed to reduce the potential for vortexing by conducting analyses to determine the minimum acceptable SDC flow rate. These analyses are discussed below.

Impact of Reduced SDC Flow on Decay Heat Removal Analyses were performed to determine the SDC flow rates required to remove decay heat during Mode 6, at several different times following reactor shutdown.

For purposes of this evaluation, the required SDC flow rate was defined as the minimum flow necessary to maintain RCS temperature below 140'F (i.e., the Technical Specification definition of Mode 6) and maintain the temperature difference across the reactor core to less than 60'F - the full power core temperature difference (see Waterford 3 FSAR Section 9.3.6.2.1).

A heat balance was performed on the SDC System heat exchanger under steady-state RCS and SDC System conditions to determine the heat exchanger power (heat remeval rate) at various SDC flow rates. Decay heat curves were then used to identify the times after reactor shutdown when the decay heat rate equaled beat exchanger power for the selected SDC flow rates. At each flow rate the reactor core temperature difference was calculated to ensure it remained less than 60*F.

Heat exchanger power was calculated based on the heat exchanger effectiveness method, a standard heat exchanger analysis technique (see J.P. Holman, Heat Transfer, 5th Ed., McGraw-Hill, 1981). The parameters of interest in the effectiveness method include the heat exchanger tube and shell side flow ratis, tube and shell side inlet temperatures, and overall heat transfer coefficient.

The tube side flow rate is the same as the SDC flow rate and the tube side inlet temperature is the same as the RCS temperature. The remaining parameter values were taken as the FSAR design values.

Certain conservatisms were assumed in the analyses. The design value for the overall heat transfer coefficient was reduced for flow rates lower than design and held constant for flow rates higher than design. Decay heat curves, which contain a 10% conservatism, were used from NUREG 0800, Branen Technical l

Position ASB 9-2, Revision 2.

Heat losses from the RCS and SDC System piping l

were neglected, l

The analyses resulted in the following flow reductions:

Time After Reactor Required SDC l

Shntdown (hrs)

Flow Rate (gpm) 265 2500 323 2200 375 2000 2

NS20707 L

Based on the usual progression of a refueling outage, the proposed change only incorporates the 2000 gpm/375 hour result. The remaining figures are included for completeness and to shorten the review time should additional changes be necessary in the future.

Impact of Re,duced SDC Flow on Boron Dilution The boron dilution analysis for Waterford 3 is described in Section 15.4.1.4 of the FSAR.

In reviewing the potential of reduced SDC flow on the boron dilution event, the primary concern is that the flow rate be large enough to avoid boron stratification which could lead to reduced times to criticality.

Adequate boron mixing (i.e., no boron stratification) will occur if the flow in the RCS cold leg is turbulent and the fluid loop transit time through the RCS and SDC System is less than the calculated time to criticality for the dilution event. Therefore, calculations were performed to determine the degree of turbulence and loop transit time for a reduced SDC flow rate of 2000 gPm.

For a flow rate of 1000 gpm in each RCS cold leg of the operating SDC train (a total SDC flow rate of 2000 gpm), the Reynolds number value is well into the turbulent flow regime. The fluid loop transit time is less than half of the minimum time to criticality for the most limiting boron dilution event (Mode 5 drained). The velocity of the unborated charging flow as it enters the cold leg, the location of the charging nozzle relative to the safety injection nozzle, the high ratio of SDC flow rate to charging flow rate, and rotation of flow in the reactor vessel downcomer provide additional assurance that adequate boren mixing will occur. Therefore, SDC flow rates at 2000 gpa and greater will ensure negligible impact on the boron dilution analyses.

Implementation Date The second refueling outage for Waterford 3 is presently scheduled to commence during the first week of April, 1988. Mode 6 will be entered early in the outage.

In order to support this schedule and minimize the potential for loss c f SDC due to vortexing an effective date of late March,1988 is requested for

+

the proposed change.

Safety Analysis The proposed change described above shall be deemed to involve a significant hazards consideration if there is a positive finding in any of the following areas:

3 NS20707

m 1.

Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequence of any accident previously evaluated?

Response: No The purpose of the proposed change is to reduce the potential for an inadvertent loss of SDC due to vortexing during Mode 6 operation.

The reduction in SDC flow rate has been shown to have no adverse effect on RCS mixing while maintaining sufficient flow to remove core decay heat. Therefore, the proposed change will not increase the probability or consequence of any accident previously evaluated.

2.

Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed change affects only the SDC flow rate during Mode 6.

No new equipment, connections, modes of operation, etc., have been introduced through the change. Therefore, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3.

Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?

Response: No In the context of the proposed change, margin of safety is defined by the SDC System design bases - i.e., ensure sufficient cooling capacity to remove decay heat and prevent boron stratification.

Analyses performed by Waterford 3 have demonstrated that the proposed reduction in SDC flow preserves the design bases. The matgin of safety to a loss of SDC flow event is increased due to lowering the potential for vortexing. Therefore, the proposed change vill not involve a reduction in the margin of safety.

Sai'ety and Significant Hazards Determination Based on the ebove Safety Analysis, it is concluded that:

(1) the proposed change does not constitute a significant hazards consideration as defined by 10CFR50.92; and (2) there is a reasonable assurance that the health and safety of the public will not be endangered by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station on the enviror. ment as described in the NRC Final Environmental Statement.

4 NS20707

r a

@l NPF-38-74 ATTACHMENT A

REFUELING OPERATIONS 3/4.9.8 SHUT 00VN CCOLING AND COOLANT CIRCULATION HIGH WATER LEVQ LIMITING CONDITION FOR OPERATION l

3.9.8.1 At least one shutdown cooling train shall be OPERA 8LE and in operation."

APPLICABILITY: H00E 6 when the water level above the top of the reactor pressure vessel flange is greater than er equal to 23 feet.

ACTION:

With no shutdown cooling train OPERABLE and in operation, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System and immedi.2tely initiate corrective action to return the required shutdown cooling train to OPERA 8LE and operating status as soon as possible.

Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVE!LLANCE REQUIREMENTS 4.9.8.1 At least one shutdown c niing train shall be verified to be in operation and circulating reactt coolant at a flow rate of greater than or equal to 4000 gpm** at least or.-u per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

"The shutdown cooling loop may be removat from operation for up to I hour per 8-hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor p. assure vessel hot legs.

- The minimum flow may be reduced to 3000 gpm af ter the reactor has been shut down for greater than or equal to 175 hours0.00203 days 0.0486 hours 2.893519e-4 weeks 6.65875e-5 months or by verifying at least once per hour that the RCS temperature is less then 135'F.

WATERFORD - UNIT 3 3/4 9-8

REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.2 Two# independent shutdown cooling trains shall be OPERABLE and at least one shutdown cooling train shall be in operation.*

APPLICABILITY:

MODE 6 when the water level above the top of the reactor pressure vessel flange is less than 23 feet.

ACTION:

With less than the required shutdown cooling trains OPERABLE, a.

immediately initiate corrective action to return the required trains to OPERABLE status, or to establish greater than or equal to 23 feet of water above the reactor pressure vessel flange, as soon as

possible, b.

With no shutdown cooling train in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required shutdown cooling train to operation.

Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.9.8.2 At least one shutdown cooling train shall be verified to be in operation and circulating reactor coolant at a flow rate of greater than or equal to 4000 gpm** at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

nly one shutdown cooling train is required to be OPERABLE provided there are 0

no irradiated fuel assemblies seated within the reactor pressure vessel.

"The shutdown cooling loop may be removed from operations for up to I hour per 8-hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs.

- The minimum flow may be reduced to 3000 gpm after the reactor has been shut down for greater than or equal to 175 hours0.00203 days 0.0486 hours 2.893519e-4 weeks 6.65875e-5 months or by verifying at least once per hour that the RCS temperature is less than 135'F.

WATERFORD - UNIT 3 3/4 9-9

Y

l 2

NPF-38-74 ATTACHMENT 3

O REFUELING OPERATIONS 3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION

(

NIGH WATER LEVEL LIMITING CON 0! TION FOR OPERATION 3.9.8.1 At least one shutdown cooling train shall be OPERABLE and in operation.*

APPLICABILITY:

MODE 6 when the water level above the top of the reactor pressure vessel flange is greater than or equal to 23 feet.

ACTION:

With no shutdown cooling train OPERA 8LE and in operation, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System and inmediately initiate corrective action to return the required shutdown cooling train to OPERABLE and operating status as soon as possible.

Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.9.8.1 At least one shutdown cooling train shall be verified to be in operation and circulating reactor coolant at a flow rate of greater than or equal to 4000 gpm** at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The minimum flow may be reduced to 2000 gpm after the reactor has been shut down for greater than or equal )

to 375 hours0.00434 days 0.104 hours 6.200397e-4 weeks 1.426875e-4 months .

g

The shutdown cooling loop may be renoved from operatio for up to I hour per 8-hour period during the performance of CORE ALTERATIO 5 in the vicinity of the reactor pressure vessel hot legs,

- The minimum flow may be reduced to 3000 gpm after the reactor has been shut d

down for greater than or equal to 175 hours0.00203 days 0.0486 hours 2.893519e-4 weeks 6.65875e-5 months or by ver1 fying at least once per hour that the RCS temperature is less then 135*F. A WATERFORD - UNIT 3 3/4 9-8

m REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3:9.8.2 Two# independent shutdown cooling trains shall be OPERABLE and at least one shutdown cooling train shall be in operation.*

APPLICABILITY:

MODE 6 when the water level above the top of the reactor pressure vessel flange is less than 23 feet.

ACTION:

With less than the required shutdown cooling trains OPERABLE, a.

immediately initiate corr 2ctive action to return the required trains to OPERABLE status, or to establish greater than or equal to 23 feet of water above the reactor pressure vessel flange, as soon as

possible, b.

With no shutdown cooling train in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the i

required shutdown cooling train to operation.

Close all containment penetrations providing direct access from the containment atmosphere tc the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.9.8.2 At least one shutdown cooling train shall be verified to be in operation and circulating reactor coolant at a flow rate of greater than or equal to 4000 gpm** at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

I e minimum flow may be reduced to 2000 gpm after the reactor has been shut down for greater than or equal to 375 ] hours.

nly one shutdown cooling train is required to be OPERA LE provided there are 0

no irradiated fuel assemblies seated within the reactor pressure vessel.

The shutdown cooling loop may be removed from operatio s for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8-hour period during the performance of CORE ALTERATIO S in the vicinity of the reactor pressure vessel hot legs.

j

- The minimum flow may be reduced to 3000 gpm after the freactor has been shut down for greater than or equal to 175 hours0.00203 days 0.0486 hours 2.893519e-4 weeks 6.65875e-5 months or by ver: fying at least once per hour that the RCS temperature is less than 135'F.A WATERFORD - UNIT 3 3/4 9-9

ML20149M825

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