ML20085C221

Proposed Tech Specs Re Plant Safety Limits,Limiting Condition for Operation 3.0.3,ECCS Accumulators & Auxiliary Feedwater Sys

ML20085C221
Person / Time
Site:	Sequoyah
Issue date:	08/23/1991
From:	TENNESSEE VALLEY AUTHORITY
To:
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ML20085C219	List: ML20085C218 ML20085C221
References
NUDOCS 9109030241
Download: ML20085C221 (18)

Text

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ENCLOSURE 1 PROPOSED TECliNICAL SPECIFICATION CHANGE l

SEQUOYAll NUCLEAR PLANT UNITS 1 AND 2 DOCKE;' NOS. 50-327 AND 50-328 (TVA-SQN-TS-91-11) t I

LIST OF AFFECTED PAGES Unit 1 I

B 2-1 B 3/4 0-1 B 3/4 3-1 B 3/4 7-2 4

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.*. 2:1 ' SAFETi L1H1T5 Basts 2.1.1 REACTOR CORE possible cladding perforation which woyld result in the relea pr0 ducts to the reactor coolant.

y restricting fuel operation to within the nucleate boiling regime wher t(at transfer coefficient is large and the claccing surface temperature is slightly above the coolant saturstion temperature.

result in excessive claddinOperation above the upper boundary of the nucleate boi from nucleate boiling (DNB)g temperatures because of the onset of caparture and the resultant sharp reduction in heat transfer coefficient.

DNB is not a directly measurable parmeter during operation and therefore THEf#AL PCWER and Reactor Coolant TemperatLre and Pressure hav related to DNB through the WRB 1 correlation and the W 3 correlation for a

a conditions outside the range of WRB 1 correlation.

The ONB correlations have R142 b,

uniform and non-uniform heat flux distributions.been developed to pr The local DNS heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The DNB design basis is as follows:

probability that the minimum DNSR of the limiting red during Condition I and gr II events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-1 or W-3 correlation in this application).

The R142 correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 DNBR lim'st. percent confidence that DNS will not occur when the minimum DNBR is at the The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum ONBR is no less than the safety snalysis ON11R limit, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

5 F"*

t044 O fit /A DM, L tany The curves are based on an enthalpy het channel factor, FAH, [ofZJ5]and

"# b N

a reference cosine with a peak of 1,55 for axial power shape.

An allowance is % s.

included for an increase in f g at reduced power based on the expression:

RTP fu P%

y F33 = *E (1+ tM (1-P))

rwe m P" d where P h no < x,4ics of asigg iE'?R %

• m r-me A""

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ffg = TM f$ L

• s '~ er AMru ryu;,

,q,w,ca (prP) sace gento in me cas. ie, An o P Fan r e a m n l,a,nu a,a,e p a

& O8][00 Ffg M c wie mm ua, i

SEQUOYAH - UNti 1 B 2-1 Amendment No. 19, 114, 133

b-W 3/4.O APpt!CABILITY l

The specifications of this section provide the general requirements applicable to each of the Limiting Conditions for Operation and Surveillance Requirements within Section3/4.

3. 0.1 This specification defines the applicability of each specification in terms of defined OPERATIONAL MODES or other srsecified conditions and is provided to delineate specifically when each specification is applicable.

3.0.2 This specification defines those conditions necessary to constitute compliance with the terms of an individual Limiting Condition for Operation and associated ACTION requirement.

3.0.3 This specification delineates the ACTION to be taken for circumstances not directly provided for in the ACTION statements _and whose occurrence would violate the intent of the specification.

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'3.0.4 This specification provides that entry into an OPERATIONAL MODE or other specified applicability condition aust be made with (a) the full complement of required systems, equipment or components OPERABLE and (b) all other para -

meters as specified in the Limiting Conditions for Operation being met without

-regard for allowable deviations and out of service provisions contained in the ACTION statements.

The intent of this provision is to insure that facility operation is not initiated with either required equipment or systems inoperable or other spet.i' led limits being exceeded.

Exceptions to this provision have been provided for a limited number of specifications when startup with inoperable equipment would not affect plant safety. These exceptions are stated in the ACTION statements of the V

appropriate specifications.

(

SEQUOYAH

• UNIT 1 B 3/4 0-1 SEP 17 080

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~..

=

0 INSERT C For example, Specification 3.5.2 requires two independent ECCS subsystems to be OPERABLE and provides expi.icit ACTION requirements if one ECCS subsystem is inoperable.

Under the requirements of Spe-ification 3.0.3. If both of the required ECCS subsyst2ms are inoperable, within one hour measures must be initiated to place the unit in at least !!OT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

and in at.least il0T SilUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

As a further example. Specification 3.6.2.1 requires two containment spray subsystems to be OPERABLE and provides explicit ACTION requirements if one spray subsystem is inoperable.

Under the requirements of Specification 3.0.3 if both of the required containment spray subsystems are inoperable, within one hour measures must be initiated to place the unit in at least HOT STANDBY withiti the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, in at least 110T SHUTD0WN within the f ollowing 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in COLD.

SilVTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3/4.5 EMERGENCY CORE COOLING SYSTEMS BASES 3L4j.1 ACCUMULATORS The OPERABILITY of each cold leg injection accumulator ensures that a sufficient volume of borated water will be immediately forced into the reactor R144l core in the event that the RCS pressure falls below the specified pressure of the accumulators.

For the cold leg injection accumulators, this condition occurs in the event of a large or small rupture, g36; The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are met.

The limits in the specirication for accumulator volume and nitrogen cover pressure are analysis limits and do not include instrument uncertainty.

The cover pressure limits were determined by Wcstinghouse to be 615 psia and 697.5 psia.

Since the instrument read-outs in the control room are in psig, the TS valves have been converted to psig and rounded to the nearest whole R144 numbers.

The actual nitrogen cover pressure safety limits in SQNfs desian documents are 600.3 psig and 682.8 psig.

IJhepfnimufDorc

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tor 9 re wi)4,rfe,aig subtritipel dur/yconcegratip epf,ur tpa't tp re 4

no the/accuu iatorf Anj ioK per/od a snuni brpik LWAJ 7;cww egib doD m Ide r /4 The accumulator power operated isolation valves are considered to be

" operating bypasses" in the context of IEEE Std. 279-1971, which requires that

• • !!Y bypasses of a protective function be removed automatically whenever permissive conditions are not met.

In addition, as these accumulator isolation valves fail to meet single failure criteria, removal of power to the valves is required.

The limits for operation with an accumulator inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures, if a closed isolation valve cannot be immediately opened, the full capability of one accumulator is not available and prompt action is required to place the reactor in a mode where this capability is not required.

E4.5.2and3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration.

Either subsystem operating in conjunction with the accumulators is capable Cf supplying sufficient core cooling to limit tne peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward.

In addition, each ECCS subsystem provides long term core cooling capability in the recircu-lation mode during the accident recovery period.

(

SEQUOYAH - UNIT 1 B 3/4 5-1 Amendment No. 140 VM'11L990

1 i

1 INSERT A l

4 i

Thrs minimum boron concentration ensures that the reactor core will remain suberitical durl..n the post-LOCA (loss of coolant accident) recirculation phase based upon the cold leg accumulators' contribution to the post-LOCA swnp

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mixture concentration.

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QANTSYSTEMS BASES I

X =

Total relieving capacity of all safety valves per steam line in lbs/ hour, 4.75 x 105 lbs/ hour at 1170 psig.

Maximum relieving capacity of any one safety valve in Y

• Ibs/ hour, 950,000 lbs/ hour at 1170 psig.

3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 350*F from norN 1 operating conditions in the event of a total loss of off-site power.

The steam driven auxiliary feedwater pump is capable of delivering 880 gpm (total feedwater flow) and each of the electric driven auxiliary feedwater pumps are capable of delivering 440 gpm (total feedwater i'Ic-) to the entrance of the steam generators at steam generator pressures of 11Cs psia.

At R119 1100 psia the open steam generator safety valve (s) are capable of relieving at least 11% of nominal steam flow.

A total feedwater flow of 440 gpm at pressures of 1100 psia is sufficient to ensuae that adequate feedwater g[19 flow is available to romove decay heat and reduce the Reactor Coot-et System temperature to less than 350*F where the Residual Heat Removal Sys

. may be placed into operation.

The surveillance test values ensure that each pump will provide at least 440 gpm plus pump rceirculation flow against a steani R119 generator pressure of 1100 psia.

Each motor-driven auxiliary feedwater pump (one Train A and one Train B) supplies flow paths to two steam generators.

Each flow path contains an automatic air-operated level control valve (LCV).

Ths LCVs have the $dme train designation as the associated pump and are provided trained air.

The I

turbine-driven auxiliary feedwater pump supplies flow paths to all four ster generators.

Each of these flow paths contains an automatic air-operated LCV, two of which are designated as Train A receive A~ train air, and provide flow to the same steam generators that are supplied by the B-train motor-driven auxiliary feedwater pump.

The remaining two LCVs are designated as Train B, BR-1 receive B-train air, and provide flow to the same steam generators that are supplied by the A-train motor-driven pump.

This design provides the required redundancy to ensure that at least two steam generators receive the necessary flow assuming any single failure.

It can be seen from the description provided above that the loss of a single train of air (A or B) will not prevent the auxiliary feedwater system from performin-} its intended safety function and is no more severe than the loss of a single auxiliary feedwater pump.

Therefore, the loss of a single train of auxillary air only affects the capability of a single motor-driven auxiliary feedwater pump because the turbine-driven pump is still capable of providing flow to two steam generators that are separate from the other m riv n ump.

Aoo sv < a T*

t 3/4.7.1.3 CONDENSATE STORAGE TANK

=- ~ -- -

The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT l

STANDBY conditions for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with steam discharge to the atmosphere concurrent I

with total loss of off-site power.

The contained water volume limit includes an allowance for water not useable because of tank discharge line location or other physical characteristics.

3 SEQUOYAH - UNIT 1 B 3/4 7-2 Amendment No. 115 Revised:

March 23, 1990

INSERT B Two redundant steam sources are required to be operable to ensure that at least one source is avaliable for the steam-driven auxiliary feedwater (AFW) pump operation following a feedwater or main steam line break.

This requirement ensures that the plant remains within its design basis (i.e.. ATW to two intact steam generatore) given the event of a loss of the No. I steam generator because ci. main steam line or feedwater_line break and a single f ailure of the B-train motor driven AFW pump.

The two redundant sources must be aligned such t!.at No. I steam generator source is open and operable and the No. 4 steam generator source is closed and operable.

For instances where one train of emergency raw cooling water (ERCW) is declared inoperablo in accordance with technical specifications, the AFW turbine-driven pump is considered operable since it is supplied by both trains of ERCW. This position is consistent with American National Standards institute /ANS 58.9 requirements (i.e.. postulation of the failure of the opposite train is not required while relying on the TS limiting condition for operation).

1 f~

2.1 $AFETY LIMITS BASES 2.1.1 REACTOR CORE Ti.. restrictions of this Safety Limit prevent overheating of the fuel and possib'e cladding perforation which w'ould result in the release of fission i

/

products to the reactor coolant.

Overheating of the fuel cladding is prevented f&

f, by restricting fuel operation to within the nucleate boiling regime where the Q

heat transfer coefficient is large and the cladding surface temperature is l3

}

slightly above the coolant saturation temperature.

3 Operation above the upper boundary of the nucleate boiling regime could p result in excessive cladding temperatures because of the onset of departure i

from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer q{

DNB is not a directly measurable parameter during operation and, k" coefficient.

therefore, THERMAL POWER and Reactor Coolant Temperature and Pressure have been v

related to DNB through the WRB-1 correlation and the W-3 correlation for condi-R120

)

t tions outside the range of the WRB-1 correlation.

The DNB correlations have g

'1 Q been developed to predict the DNB flux and the location of DNB for axially

(

uniform and non-uniform heat flux distributions, lha local DNB heat flux ratio, ONBR, defined as the ratio of the heat flux t, 't would cause DNB at a V{ particular core location to the local heat flux, is indicative of the margin to J

W4 '

9 I

The DNB design basis is as follows:

there must be at least a 95 percent 7

h

{

probability that the minimum DNBR of the limiting rod during Condition I and L

11 events is greater than or equal to the DNBR limit of the ONB correlation

}

lh R13 j

being used (the WRB-1 or W-3 correlation in this application).

The correlation g

DNBR limit is established based on the entire applicable experimental data set Q

L j

/*

such that there is a 95 percent probability with 95 percent confidence that DNB j ( will not occur when the minimum DNBR is at the DNBR limit, p

g g

N L

a 1

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reac-

'. tsd tor Coolant System pressure and average temperature for which the minimum DNBR 7

R104 -

1 3 ~i is no less than the safety analysis DNBR limit or the average enthalov at tha

-I Vessel exit is equal to the enthalpy of s4turated liquid.

RI k k

. J ej 3

r4 N

k These curves are based on an enthalpy hot channel factor, F aH and I

'f g

s

)h ? d b4 a reference cosine with a peak of 1.55 for axial power shape.

An allowance 43 included for an increase in F at reduced power based on the expression:

R21/ j g

e i FE f%

N g% 9 6H,,;_ g,[y 4_; (1 p))

g t

F n&

where P is the f raction of RAT [0 Tll[RMAL POWER- = ge wg% <it 1-Hm m.

k, 3

A g

g These limiting heat flux conditions are higher than those calculated for

• N r' the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the When the axial power BR 1 (delta 1) function of the Overtemperature Delta T trip.

f SEQUOYAH - UNIT 2 B 2-1 Amendment No. 21, 104, 130 Reused 08/18/87

___ -._____ _ _ _ b I 4,.

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I 3/4.0 APPLICABILITY BASES i

ihe specifications of this section provide the general requirements applicable to each of the Limiting Conditions for Operation and Surveillance Requirements within Section 3/4.

j l

3.0.1 This specification defines the applicability of each specification I

in terms of defined OPERATIONAL MODES or other specified conditions and is i

l provided to delineate specifically when each specification is applicable.

3.0.2 This specification defines those conditions necessary to constitute l

compliance with the terms of an individual Limiting Condition for Operation and associated ACTION requirement.

3 me s ys rem i

s ua m ysrrm s l

3.0.3 This specification delineates the measu es t be taken for.

t I

those circumstances not directly provided for in t e AC ON statements and whose occurrence would violate the intent of a specific tion For example, i

l, Specification 3.5.2 requires two independent ECC sub ystems to be OPERABLE and provides explicit ACTION requirements if on ECC subsystem is inoperable.-

iw Under the requirements of Specification 3.0.3, if b th of the required ECCS sub-f systems are inoperable, within one hour measu es st be initiated to place the unit in at least HOT STANDBY within the ext hours, and in at. least-HOT SHUT 00WN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

As a fu er example, $pecifica-tion 3.6.2.1requirestwoContainmentspraf gy/M#plto be OPERABLE and provides i

explicit ACTION requirements if one spray W ifyyis incperable.

Under the

_ requirements.of Specification 3.0.3 if both of ~ tie required Containment spray gue sysres [Meel are inoperable, within one hour measures must be initiated to place the uni't in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, in at least HOT 4

SHUT 00WN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in COLD SHUTOOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, t

3.0.4 This specification provides that entry into an OPERATIONAL H0DE or other specified applicability condition must be made with (a) the full complement of required systems, equipment or components OPERABLE and (b) all other parameters as specified in the Limiting Conditions for. Operation being met without regard for allowable deviations and out of service provisions contained in the ACTION l

statements.

The intent of this provision is to insure that facility operation is not l

l

-initiated with either required equipment or systems inoperable or other specified limits being exceeded.

Exceptions to this provision have been provided for a limited number of l

specifications when startup with inoperable equipment would not affect plant safety.

These exceptions are stated in the ACTION statements of the appropriate-specifications, i

SEQUOYAH - UNIT 2 B 3/4 0-1 L

4 3/4.5 FMERGENCY CORE COOLING SYSTEMS 1

\\

BASES l

3/4.5.1 ACCUMULATORS The OPERABILITY of each cold leg injection accumulator ensures that a g33' sufficient volume of borated water will be immediately forced into the reactor l

core in the event the RCS pressure falls below the pressure of the accumulators.

For the cold leg injection accumulators this condition occurs in the event of a l

large or small rupture, g33 l

The limits on accumulator volume, boron concentration and pressure onsyre that the assumptions used for accumulator injection in ite saffty and ysis are met.

The limits in the specific /1 tion for accumulator volse and nitrogen cover pressure are analysis limits and do not include instrument uncertainty.

The cover pressure limits were determined by Westinghouse to be 615 psia and 1

g33 697.5 psia.

Since the instrument read-outs in the control room are in psig, the TS values have been converted to psig and rounded to the nearest whole numbers.

The actual nitrogen cover pressure safety limits _in JONLs desian documents are 6_00.3 psig and 682.8 psig. PT 7mTninfn bor concF,ntratiptf /

i riod of/act6r corywil1jtm ih _

crit / cal no thef aceutpdlat4r enydres/ hat e re l

)6jection smalybreaVLO w.m A e 9 0 m554r A The accumulator power operated isolation valves are considered to be

" operating bypasses" in the context of IEEE Std. 279-1971, which requires that l

bypasses of a protective function be removed automatically whenever r?rmissive conditions are not met.

In addition, as these accumulator isolation valves

<~

fail to meet single failure criteria, removal of oower to the valves is required.

The limits for operation with an accumulator inoperable for any reason i

except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator I

which may result in unacceptable peak cladding temperatures.

If a closed isolation valve cannot be immediately opened, the full capability of one accumulator is not available and prompt action is required to place the reactor in a mode where this capability is not required.

3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems ensures th e sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration.

Either subsystem operating in conjunction.with the accumulators is capable of supplying sufficient core cooling to limit the peak cladding temperatures l

within acceptable limits for all postulated break sizes ranging from the l

double ended break of the largest RCS cald leg pipe downward.

In addition, each ECCS subsystem provides long term core cooling capability in the recirculation mode during the accident recovery period, i

With the RCS temperature below S50'F, one OPERABLE ECCS subsystem is acceptable without' single failure (.onsideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.

SEQUOYAH - UNIT 2 B 3/4 6-1 Amendment No. 131 00T 29 E0

a INSERT A The minitoum boron concentration ensures that the reactor core will remain suberitical during the post-LOCA (loss of coolant aceldent) recirculatlon phase based upon the cold leg accumulators' cont ribution to the pos t-LOCA sun.,

mixture concentratlon.

I 1

l 3

r..,.-.-.__,.___...

- ~ - - - - - - -

PLANT SYST[M5 r

BASES

(

SAflTY VALUt'$ (Continued)

Total relieving capacity of all safety valves per steam X =

6 line in Ibs/ hour, 4.75 x 10 lbs/hr at 1170 psig Y

= Maximum relieving G,apacity of any one safety valve in lbs/ hour, 9.5 x 10 lbs/hr at 1170 psig.

3/4.7.1.2 AUXILI ARY FEEDn'ATER SYSTEM The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 350 F from normal opetating conditions in the event of a total loss of off-site oower.

The steam driven auxiliary feedwater pump is capable of delivering 880 gpm (total feedwater flow) and each of the electric driven auxiliary feedwater pumps are capable of delivering 440 gpm (total feedwater flow) to the entrance of the steam generators at steam generator pressures of 1100 psia.

At 1100 psia the open steam generator safety valve (s) are capable of relieving at least 11% of nominal steam flow.

A total feedwater flow of 440 gpm at pressures of 1100 psia is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350 F where the Residual Heat Removal System may be placed into operation.

The surveillance test values ensure that each pump will provide at least 440 gpm plus pump recirculation flow against a steam generator pressure of 1100 psia.

Each motor-driven auxiliary feedwater rump (one Train A and one Train B) supplies flow paths to two steam generators.

Each flow path contains an D

automatic air operated level control valve (LCV).

The LCVs have the same train designation as the associated pump and are provided trained air.

The turbine-driven auxiliary feedwater pump supplies flow paths to all four steam generators.

Each of these flow paths contains an automc.ic air-op.;4ted LCV, two of which are designated as Train A, receive A-train air, and provide flow to the samt steam generators that are supplied by the B-train motor-driven auxiliary feedwater pump.

The remai_ning two LCVs are designated as Train B.

receive B-train air, and provide flos to the same steam generators that are supplied by the A-train motor-driven pum).

This design provides the required redundancy to ensure that at least two steam generators receive the necessary flow assuming any single failure.

It can be seen from the description HR-1 provided above that the loss of a single train of air (A or 8) will not prevent the auxiliary feedwater system from performing its intended safety function and is no more severe than the loss of a single auxiliary feedwater purp.

Therefore, the loss of a single train of auxiliary air only af fects the capability of a single motor-driven auxiliary feedwater pump because the turbine-driven pump is still capable of providing flow to two steam generators that are separate f rom the other motor-driv >

. ump, 3/4.7.1.3 CONDENSATE STORAGE TA h b The OPERABILITY of the condensate storage tank with the minitum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with steam discharge to the atmosphere concurrent with total loss-of off-site power.

The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics, i

SEQUOYAH - UNil 2 B 3/4 7-2 Amendment No 105 Revised:

March 23, 1990

+

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=

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INSERT B j

Two redundant steam sources are required to be operable to ensure that at-least one source is available for the steam-driven aaxiliary feedwater (AFW)

!=

pump operation following a feedwater or main steam line break.

This

=

requirement ensures that the plant remains within its design basis (i.e.. ATV t

to two intact steam generators) given the event of a loss of the No. I steam

,l generator because of a main steam line or feedwater line break and a single failure of the B-train motor driver AFW pump.

The two redundant sources must i

.be aligned such that No. I st9am generator source is open and operable and the No. 4 steam generator source is closed and operable.

t For instances where one train of emergency raw cooling water (ERCW) is declared inoperable in accordance with technical specifications, the AFW i

turbine-driven pump is considered operable since it is supplied by both triins of ERCW. This position is consistent with American National Standards Institute /ANS 58.9 requirements (i.e., postulation of the failure of the opposite train is not required while relying on the TS limiting condition for i

operation).

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r ENCLOSURE 2 l'ROPOSED TECilNICAL SI'ECIFICATION CHANGE SEQUOYAH NUCLEAR Pl. ANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-91-11)

DEFCRIPTION AND JUST1FICAT10N FOR REVISIONS TO Tile BASES

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Description of Change f

r TVA proposes to modliy the bases for Sequoyah Nuclear Plant (SQN)

Units 1 and 2 technical specifications (TSs) with regard to the safety

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limits, applicability, emergency core cooling system. and plant system j

sections.

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in the bases section' of the saf ety limits. 2.1.1 REACTOR CORE. the value:

of the enthalpy hot channel factor and values contained in the enthalpy hot channel factor equation have been replaced by factors that have values defined in the Core Operating Limit Report (COLR).

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In the bases section of TS 3.0.3 the word " system" assoelated with the l

containment spray specification has been revised to " subsystem." This change is being made ior both units.

Additionally, the Unit 1 TS 3.0.3 bases is being rewritten to agree with the Unit 2 75.3 0.3 bases and to correct the TS 3.0.3 action duscription.

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In the bases section of TS 3/4.5.1. ACCUMU1.ATCRS. the last sentence of the second paragraph has been deleted.

Two new sentences have been added to properly reflect the basis for the minimum accumulator boron concentration in the current analyses.

- In the bases section of TS 3/4.7.1.2. AUXILIAkY FEEDWATER SYSTEM. two i

paragraphs have been added to clarify the operability requirements of the I

turbine-driven pump with regard to steam and essential raw cooling water I

(ERCW) supplies.

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ii Reason for Change TS Change 91-08 submitted May 24, 1991, incorporated the recommendations of Generic letter 88-16. " Removal of Cycle-Specific Parameter Limits From Technical Specifleations." A COLR was created and associated TS i

parameters were replaced by reference to values contained in the COLR.

The enthalpy hot channel factor and equation in the safety limits bases are being revised to:similarly reference the COLR and to theref ore be I

consistent-with proposed TS Change 91-08.

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TS change 90-16 revised Containment Spray Specification 3.6.2.1 to i

reference subsystems rather than systems to prevent. confusion when determining system operability.

Subsequently; it was noted that the

. containment-spray system was used as an example for the-bases of l

- TS 3.0.3.

This proposed change will revise TS Bases 3.0.3 to agree with the terminology utilized in TS 3.6.2.1.

Additionally, a comparison of the Unit 1 and Unit 2 TS 3.0.3 bases found that the Unit 1 bases used a (1)

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- different emer6ency core cooling system (ECCS) example.'(2) contained an f

incorrect description of the actions required in TS 3.0.3. and (3) had a i

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superfluous last sentence added.

Following the processing of TS Change 91-01-(which provided a temporary change to the required baron concentration for a single cold leg i

-accumulator). it was noted that the Bases 3/4.5.1 may have been l

inconsistent with the.Information provided by Westinghouse Electric Corporation for the TS change.

Subsequent conversations with Westinghouse have clarified that the cold leg accumulator minimum boron concentration i

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s 2-is critical to the post-LOCA (loss of coolant accident) sump mean mixture boron concentration in ensuring that tne core renoins subetitical with all rods stuck outside the core.

(Note that the "all rods out" assumption is based on the Westinghoure position that the asymmetric large break LOCA loads will prevent the rods from falling into the core.) Thus, replacing the last sentence of the second paragraph for T5 Bases 3/4.5.1 with a revised explanation describing the purpose of the minimum boron concentration in the accumulators clarifies this bases relative to the current accident analyses.

While reviewing NRC Information Notice 89-58, " Disablement of Turbine-Driven Auxiliary Feedwater Pump Due to Closure of One of the Parallel Steam Supply Valves." SQN discovered that a potential existed to inconsistently interpret the attendant equipment relationship of the steam-supplies to the turbine-driven auxiliary f eedwater (TDAFW) Inunp.

TS Bases 3/4.7.1.2 is being revised to reflect that both steam supplies are required operable to consider a TDAFW pump operable.

This will provide documentation that ensures consistent interpretation of the TS.

A second paragraph to TS Bases 3/4.7.1.2 is being proposed that will also ensure consistent interpretation of the TSs relative to the A and B train ERCW supply to the TDAFW pumps.

Justification for Change The revision to the enthalpy hot channel factor and equattor. in the safety limits bases is being made to provide agreement between TS sections incorporating the provisions of Generic Letter 88-16.

TS 3/4.2.3,

" Nuclear Enthalphy llot Channel Factors," is being revised by TS Change 91-08.

This proposed revision is consistent with TS Change 91-08.

Additionally, no safety-related equipment, safety function, or plant operations will be_ altered as a result of'this change.

By changing the wording in Bases TS 3.0.3 from system to subsystem. SQN is making the referenced example consistent with TS 3.6.2.1.

To provide consistency, the Unit 1 bases was rewritten to provide the duplicate ECCS-example and correct the TS 3.0.3 action description error.

This change is also.in agreement with Westinghouse Standard TS, Revision 4 (NUREG-0452 Revision 4)..

The last sentence of paragraph two of TS Bases 3/4.5.1 is being replaced with a correction to the basis for the minimum boron concentration in the four cold leg accumulators. The addition of boron to the cold leg accumulators and.the refueling water storage tank is to ensure long-term subcriticality after a LOCA for the recirculation mode.

This new description provides an accurate account of that purpose relative to the current accident analysis.-

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s The first proposed change to the bases for the auxiliary feedwater system is consistent with the information provided to utilities by NRC Information Notice 89-58.

The notice addressed adherence to the safety

- analysis for a main steam line break on for a feedwater line break that requires at least two steam generators for postaccident cooldown.

Py placing this clarification in the bases, a consistent interpretation of the requirement to have two operable and redundant steam supplies will be provided to the operating staff.

This will enhance plant operation and reduce the possibility of violating the TS because of unclear requirements. Additionally, this chatge is consistent with a similar request submitted by Pacific Gas and Electric Company for Diablo Canyon on September 11. 1990.

The second paragraph is being provided to clarify the TDAFW pump operablid ty.

The TDAFW pump la supplied by both trains of ERCWt-therefoie, if one train of'ERCW is inoperable, the other train will be available to supply water in an accident condition.

Further. American National Standards institute (ANSll ANSI /ANS 58.9 states that "If one train of a redundant safety-related fluid system or its safety supporting systems is temporarily rendered inoperable due to short-term maintenance as allowed by the-unit technical 17ecifications, a single failure need not be assumed in the other train." tor this reason, only one train of ERCW is required to ensure that the TDAFW pump is operable as no other single failure is required to be. considered.

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