ML20078Q506

Proposed Tech Specs TS 3.7.1.1,Tables 3.7-1 & 3.7-2 & Bases 3/4.7.1.1 Associated W/Max Applicable Power Levels for Operation W/One or More MSSVs Inoperable

ML20078Q506
Person / Time
Site:	Sequoyah
Issue date:	12/16/1994
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20078Q504	List: ML20078Q500 ML20078Q506
References
NUDOCS 9412220222
Download: ML20078Q506 (38)

Text

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ENCLOSURE 1  !

l PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE . ,

i SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 1  :

DOCKET NOS. 50-327 AND 50-328  !

(TVA-SON-TS-94-07)  :

i

.h I

f LIST OF AFFECTED PAGES

?

Unit 1  !

3/4 7-1 -

3/4 7 2 I 3/4 7-3  !

3/4 7-4 l B 3/4 7 1 }

B 3/4 7-2 l I

Unit 2 3/4 7-1 3/4 7-2 {

3/4 7-3 3/4 7-4 8 3/4 7-1 i B 3/4 7-2 ,

1 941222O222 941216 PDR ADOCK 05000327 P POR

s e- ,, .*

3/4.7 PLANT SYSTEMS 3/4.7.1~ TURBINE CYCLE SAFETY VALVES LIllITING CONDITION FOR OPERATION  :

i (mssvd .

3.7.1.1 in steam line ccdc safety valves ::cciated ith cach stea-  !

g:r,: rater shall be OPERABLE with lift settings as specified in Table 3.7-APPLICABILITY: MODES 1, 2 and 3. I ACTION:

a. Wi 4 reactor olant loops d associate team genera rs in  !

ope tion and wi one or more in steam lin code safety alves t

- [g e, inope able, operat n in MODES 1, 2 and 3 may p ceed provi , that  !

within hours, eith r the inopera e valve is re tored to OP BLE  ;

a fc *~p status o the Power R ge Neutron F1 High Setpoi trip is re ed per Table' 7-1; othtrw e, be in at 1 st HOT STAND within the i A next 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD UTDOWN withir the following 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. ,

. b. W h 3 reactor oolant loops nd associate team generato in

,,igh ope ation and wi one or more main steam lin code safety v ves a;y asso 'ated with an perating lo inoperable, o eration in MOD 3 R11 i may pr eed provide that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, eithe the inoperable 3 valve is restored to ERABLE stat or the Power nge Neutron F1 High Trip etpoint trip is reduced p Table 3.7-2; therwise, be in  ;

t least H STANDBY wit n the next hours and in C D SHUTDOWN

' thin the f lowing 30 ho s.  :

c. The rovisions Specificati 3.0.4 are n t applicable.

SURVEILLANCE REQUIREMENTS ,

l 4.7.1.1 No additional Surveillance Requirements other than those required by l Specification 4.0.5.  ;

l i

i i

' i SEQUOYAH - UNIT 1 3/4 7-1 Amendment No.114 May 5, 1989 l

,. f n ger Y

a. With one or more MSSVs' inoperable, operation may proceed .

provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve.

is restored to OPERABLE status or the Power Range Neutron Flux High Setpoint trip is reduced per Table 3.7-1.- The provisions of Specification 3.0.4-are not applicable, i

b. With the requirements of ACTION a., not mot or with one or  :

more' steam. generators with less than'two MSSVs OPERABLE'be  !

in at least HOT' STAND 3Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN in the following.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, i

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m TABLE 3.7-1 E

g MAXIMUM ALLOWABLE POVER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM y LINE SAFE 1Y VALVES "L;':';: " LOO" CP:P":cM x

c- Maximum Allowable Power Range 5

Maximum Number of Inoperable Safety Neutron Flux High Setpoint Valves on Any Operating Steam Generator (Percent of RATED THERMAL POWER) 2 ps' VS pf 2. 8 R.

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SEQUOYAH - UNIT 1  :  ;

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SEQUOYAH - UNIT 1 3/47-[3 w

1

• e 3/4.7 PLANT SYSTEMS I

BASES 3/4.7.1 TURBINE CYCLE ,

3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures that '

the secondary system pressure will be limited to within 110% (1104 psig) of  :

the system design pressure during the most severe anticipated system operational r transient. The maximum relieving capacity is associated with a turbine trip .

from 100% RATED THERMAL POWER coincident with an assumed loss of condenser hea sink (i.e., no steam bypass to the condenser).

& g . -+

f,,,se f The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code, 1971 Edition. The total relieving capacity for all valves on all of the steam lines is 1.9 x 107 lbs/hr at 1170 psig which is 127 percent of the total secondary steam flow of 1.493 x 107 lbs/hr at 100% RATED THERMAL POWER. A minimum of 2 OPERABLE safety valves per steam generator ensures that sufficient relievi a-pareci available for the allowable THERMAL POWER restriction in Tabl '. 7 2. 3.7- I STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction <

- in secondary system steam flow and THERMAL POWER required by the reduced l' reactor trip settings of the Power Range Neutron Flux channels. The reactor trip setpoint reductions are derived on the following bases: t v

6c " .g For loop operati n dCH ;# 3p , (X) - ( )(V) x (109) x or 3 loop ope tion 3p ,(X) (Y)(U)

(76 X

here:

SP = red ed reactor tri setpoint in pe cent of RATED ,

THE MAL POWER V = ma imum number of inoperable saf y valves per s eam line U = ximum number inoperable sa ety valves per operating .

team line.

109 = Power Range N utron Flux-Hig Trip Setpoint or 4 loop '

) operation.

76 Maximum per ent of RATED T ERMAL POWER pe issible by P-8 Setpoi t for 3 loop o eration.

BR-1 SEQUOYAH - UNIT 1 B 3/4 7-1 Revised: March 23. 1990

4 PLANT SYSTEMS BASES X = tal rel capacit f all safe v valves steam [e.e-li in lbs/ho r, 4.75 x 1 6 lbs/ hour

• 1170 psig. .

= Maxim , relieving apacity of ny one safe valve in Ibs/hou 950,000 1 / hour at 1 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 normal 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 flow) to the entrance of the steam generators at steam generator pressures of 1100 psia. At gli' 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 git:

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 Rll' generator pressure of 1100 psia.

Each motor-driven auxiliary feedwater pump (one Train A ar.d one Train B) supplies flow paths to two steam generators. Each flow path contains an 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 automatic air operated LCV, two of which are designated as Train A, receive A-train air, and provide flow

,g_d 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 S, l receive B-train air, and provide flow to the same steam generators that are l supplied by the A-train motor-driven pump. This design provides the required j 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 performing 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 auxiliary air only affects the  ;

capability of a single motor-driven auxiliary feedwater pump because the l turbine-driven pump is still capable of providing flow to two steam ger.erators (

that are separate f rom the other motor-driven pump. 1 Two redundant steam sources are required to be operable to ensure that at least one source is available for the steam-driven auxiliary feedwater (AFV) Rh pump operation following a feedwater or main steam line break. This require- l ment ensures that the plant remains within its desien basis (i.e. , AFV to two  !

intact steam generators) given the event of a less of the No. I steam generztor l l

SE000YAH - UNIT 1 B 3/4 7-2 Amendment No.115 '.55

ngrY- -

.. .. 1 In Mode 1 above 28% RTP, the number of MSSVs per steam generator  :

required to be operable must be according to Table 3.7-1 in the '

accompanying LCO. At or below 28% RTP in Modes 1, 2, and 3, only I two MSSVs per steam generator are required to be operable.

I In Modes 4 and 5, there are no credible transients requiring the MSSVs. The steam generators are not normally used for heat t removal in Modes 5 and 6, and thus cannot be overpressurized; e there is no requirement for the MSSVs to be operable in these [

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• IT6CJ A$(y$r$4r To calculate this setpoint, the governing equation is the relationship q = m Ah, where q is the heat input from the primary side, m is the steam flow rate and Ah is the heat of vaporization at the steam relief pressure (assuming no subcooled feedwater). Thus, an algorithm for use in defining the revised Technical Specification table setpoint values would be:

h Hi 4 = (100/Q) (w' ,,N)

K where:

Hi $ = Safety Analysis power range high neutron flux setpoint, percent Q = Nominal NSSS power rating of the plant (including reactor coolant pump heat), Mwt K = Conversion factor,947.82 W sec)

Mwt w, = Minimum total steam flow rate capability of the operable MSSVs on any one steam generator at the highest MSSV opening pressure including >

tolerance and accumulation, as appropriate,in Ib/sec. For example,if the maximum number ofinoperable MSSVs on any one steam generator is one, then w, should be a summation of the capacity of the operable MSSVs at the highest operable MSSV operating pressure, excluding the highest capacity MSSV. If the maximum number ofinoperable MSSVs per steam generator is three then w, should be a summation of the capacity of the operable MSSVs at the highest operable MSSV operating pressure, excluding the three highest capacity MSSVs.

i h3 = herd of vaporization for steam at the highest MSSV opening pressure including tolerance and accumulation, as appropriate, Btu /lbm 4

. N = Number ofloops in plant ,

The values calculated from this algorithm must then be adjusted lower f6 S ii.?= Muvuf4 to account for instrument and channel uncertainties, N M 4d>I4&P M M sp***WEA%@iitf@6dMes#44*pb tB4ts.wpf;tultpg4 5 .

l . y 3/4.7 PLANT SYSTEMS

_ 3/4.7.1 TURBINE CYCLE SAFETY VALVES MAITINGCONDITIONFOROPERATION

^'l ,: in steam lin ::d safety valves (mssvs) 3.7.1.1 .

!:ted with ::ch :t ::

gr.:r:tershallbeOPERABLEwithliftsettingsasspecifiedinTable3.7-/

APPLICABILITY: Modes 1, 2 and 3.

ACTION:

s

a. With4reactortolantloopsandhsociatedstea enerators in eration and with e or more main eam line code afety valves

{g &D i aerable, operation 'n Modes 1, 2 an 3 may procee wit

• 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either rovided, that e inoperable va e is restore o OPERABLE s# statu or the Power Range eutron Flux Hig Trip Setpoint 's reduced A per Tab 3.7-1; otherwise, e in at least H STANDBY with the y next 6 ho and in COLD SHUT WN within the fo lowing 30 hou

b. ith 3 reactor colant loops and sociated steam g erators in

, op ation and w one or more main team line code s ety valves asso iated with an perating loop inop rable, operation 'n MODE 3 R 4 s may pr eed provide , that within 4 hou.- eithea the ino rable valve is restored to 0 RABLE status or th Power Reage Neu on '

Flux High rip Setpoint reduced per Table .7-2; otherwise, NbeinatleatHOTSTANDB ithin the next 6 h rs and in COLD t HUT 00WN with1 the followin 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. ,

l

c. The rovisions of ecification 0.4 are not applic le.

l l

r SURVEILLANCE REQUIREMENTS s l l

4.7.1.1 No additional Surveillance Requirements other than those required by Specification 4.0.5.  :

e SEQUOYAH - UNIT 2 3/4 7-1 Amendment No. 104 May 5, 1989

• . . f nSeyt

a. With one or.more MSSVs inoperable, operation may proceed provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve is. restored to OPERABLE status or the Power Range Neutron l Flux High Setpoint trip is reduced per Table 3.7-1. The J provisions of Specification 3.0.4 are not applicable.  !

b. With the requirements of ACTION a., not met or.with one or more steam generators with less than two MSSVs' OPERABLE be in at least HOT STANDBY within the ne:rt 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> r.nd in HOT .

SHUTDOWN in the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

L 5

b f

f 6

4

TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES """:MC " LO^^ ^?:":0N y Maximum Number of Inoperable Maximum Allowable Power Range j

. Safety Valves on Any Neutron Flux High Setpoint .

Operating Steam Generator (Percent of RATED THERMAL POWER) {

I 1 ^ $ 63 2 f YS .l 3 [ 78 .

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I SEQUOYAH - UNIT 2 3/4 7-2  ;

f

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• s BLE 3.7-2

1. IMUM ALLOVA POWER RAN NEUTROF FL X HIGH O!NT WITH INLNDERABLE STEAM \ tine SAFETY \ VALVE 5 00Rit$ 3 LOOP CRERATION Hax' mum Number o Inoperable Ma imum Allowa le Power nge Sa ety Vaives o Any .

utron Fiux igh Setpoi t Oper tinc Steam G nerator" Perc nt of RATED HERMAL PO ER I 60 45 3 30 "At least two 5a ety valves 5. 11 be OPERABL. on the non- erating Stea.

gen rator, v

8.:h v

Sta nnr.nn - un a i c. 3 / - 7 .3

TABLE 3.7D is:L 5 STEAM llNE SAFETY VALVES PER LOOP gj LIFT SETTING (1 1 N0ZZLE SIZE i

VALVE NUMBER  %)^_

I toop 1 Loop 2 Loop 3 Loop 4 l EE l

2-1-517 2-1-512 2-1-527 1064 psig 16 sq. in.

[#, 2-1-522 1077 psig 16 sq. in.

2-1-523 2-1-510 2-1-513 2-1-528 2-1-514 2-1-529 1090 psig 16 sq. in.

2-1-524 2-1-519 2-1-515 2-1-530 1103 psig 16 sq. in.

2-1-525 2-1-520 2-1-516 2-1-531 1117 p>'g 16 sq. in.

2-1-526 2-1-521 R

l l

sT' LP

'The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating terperature and pressure.

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4 3/4.7 PLANT SYSTEMS BASES I

'3/4.7.1 TURBINE CYCLE l

3/4.7.1.1 SAFETY VALVES '

l The OPERABILITY of the main steam lin'e code safety valves ensures that ,

the secondary system pressure will be' limited to within 110% (1194 psig) of '

the system design pressure during the most severe anticipated system operational i transient. The maximum relieving capacity is associated with a turbine trip .

from 100% RATED THERMAL POWER coincident with an assumed loss of condenser  !

ink (i.e., no steam bypass to the condenser).

z. . & ,

g t b The specified valve lift settings and relieving capacities are'in fr5 accordance with the requirements of Section III of the.ASME Boiler and i Pressure Code, 1971 Edition. The potal relieving capacity for all valves on- '

all of the steam lines is 1.9 x 10 lbs/hr at 1; l of the total secondary steam flow of 1.493 x 10}70lbs/hr psigatwhich is 127 percent 100% RATED THERMAL

, POWER. A minimum of 2 OPERABLE safety valves per steam generator ensures that -

sufficient relieving ilable for the allowable THERMAL POWER i restriction in Tabl . 7 ^. 3.7 - / l STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction 1 in secondary system steam flow and THERMAL POWER required by the; reduced  ;

J reactor trip settings of the Power Range Neutron Flux channels. The reactor trip setpoi.nt reductions are derived on the following bases:

l For 4 loop o eration 3p , (X) - (Y) ) x 109 X

hc-4 F r 3 loop opera on i j

3p , (X) - (Y)(U) x 76 l

I W ere:

, SP = Reduc reactor trip setp int in percent of RATED THERMAL WER V = Maxi um number of inope able safety valve per steam lin k U = Ma imum number of ino erable safety val es per operati g s eam line 109 = ower Range Neutro Flux-High Trip S point for 4 1 op operation T 76 Maximum percent RATED THERMAL P ER permissib1 by

,1 P-8 Setpoint f r 3 loop operati n.

BR-1 SEQUOYAH - UNIT 2 B 3/4 7-1 Revised: March 23, 1990 l

i . .

PLANT SYSTEMS BASES

$AFETY VA nued j X = tal relievi - capacity f all sa ty valve per steam g lin in lbs/ hour 4.75 x 1 lbs/hr a 1170 pst

& u ,t

= Maximu. relieving c acity of ' one sa ty valve 'n

(

5 lbs/ hour, 9.5 x 10 lb /hr at 117 psig.

3/4.7.1.2 AUXILIARY FEE 0 WATER stM l

The OPERABILITY of the auxiliary feedwater system ensures that the Reactor 8

Coolant System can be cooled down to less than 350 F from normal 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 ppm (total feedwater flow) to the entrance of the stese 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 pria 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 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). The LCVs have the same train designation as the associated pump and are provided trained air. The turbine-driven auxiliary feedaater pump supplies flow paths to all four steam 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, 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 feed.<ater system from performing 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 auxiliary 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 f rom the other motor-driven pump.

Two redundant steam sources are reouired to be operable to ensure that at least one source is available for the steam-driven auxiliary feedsater ( AFW) pump operation following a feedwater or main steam line breat This require-ment ensures that the plant remains within its design basis (i.e., AFV to two intact steam generators) civen tne event of a loss of the No. 1 steam generator SEQUDYAH - UNIT 2 B 3/4 7-2 Amencment No. 105 By letter 10/25/E1

A 9 , - -

'1_ i In Mode 1 above'28% RTP, the number of MSSVs per steam generator required to be operable must be according to Table 3.7-1 in the accompanying LCO. At'or below 28% RTP in Modes 1, 2, and 3, only two MSSVs per steam generator are required to be operable.

1 In Modes 4 and 5, there are no credible transients requiring the

.MSSVs. The steam generators are not normally used for heat removal in Modes 5 and 6, and thus cannot be.overpressurized; there is no requirement for the MSSVs to be operable in these ,

modes.

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n.sec t C

@ To calculate this setpoint, the governing equation is ths relationship q = m Ah, where q is th heat input from the primary side, m is the steam flow rate and Ah is the heat of vaporization at the steam relief pressure (assuming no subcooled feedwater). Thus, an algorithm for use in defining the revised Technical Specification table setpoint values would be:

Hi $ = (100/Q)

K where:

Hi$ = Safety Analysis power range high neutron flux setpoint, percent Q = Nominal NSSS power rating of the plant (including reactor coolant pump heat) Mwt K = Conversion factor,947.82 (BWsec) 3 Mwt w, = Minimum total steam flow rate capability of the operable MSSVs on any one steam generator at the highest MSSV opening pressure including tolerance and accumulation, as appropriate, in Ib/sec. For example,if the maximum number ofinoperable MSSVs on any one steam generator is one, then w, should be a summation of the capacity of the operable MSSVs at the highest operable MSSV operating pressure, excluding the highest capacity MSSV. If the maximum number ofinoperable MSSVs per steam generator is three then w, should be a summation of the capacity of the operable MSSVs at the highest operable MSSV operating pressure, excluding the three highest capacity MSSVs.

h,f = heat of vaporization for steam at the highest MSSV opening pressure including tolerance and accumulation, as appropriate, Btu /lbm

. N = Number ofloops in plant ,

'Ihe values calculated from this algorithm must then be adjusted lower ftce(phpBanids SptWibiWtW3$th to account for instrument and channel uncertaindes,QZM f

Nsrumtree tW18.wpf:1t>o11994 5 .

i- ..

ENCLOSURE 2 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE s i

SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2  !

DOCKET NOS. 50-327 AND 50-328 (TVA-SON-TS-94-07)

DESCRIPTION AND JUSTlFICATION FOR l REDUCTION OF THE MAXIMUM APPLICABLE POWER LEVEL FOR l OPERATION WITH ONE OR MORE MAIN STEAM SAFETY VALVES INOPERABLE P

k k

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Descriotion of Chanae >

1 TVA proposes to modify the Sequoyah Nuclear Plant (SON) Units 1 and 2 Technical Specifications (TSs) to incorporate more conservative setpoints for maximum allowable .

power level with one or more main steam safety valves (MSSVs) inoperable. The current l setpoints are based on the assumption that the relationship between relief capacity and  !

allowable power level is linear; this assumption may not be valid for all plant conditions.' ,

i To ensure that secondary side pressures do not exceed 110 percent of the system design  ;

pressure during the most severe anticipated transient, the maximum allowable power levels j

.found in TS Table 3.7-1 would be revised to a level below the heat removing capability of the remaining operable MSSVs. For example, if one MSSV is inoperable in one steam generator (S/G) then the relief capacity of that S/G has been reduced. To maintain the  !

ability of the MSSVs to relieve pressure, the energy transfer to that S/G must be reduced by l the same amount. This is accomplished by reducing thermal power, which has the.

conservative effect of limiting the energy transfer to all S/Gs to a value equal to the relief >

capacity of the S/G with the inoperable MSSV. TS Bases Section 3/4.7.1.1 would be t revised to reflect the calculation recommended by Westinghouse Electric Corporation for use in determining reactor power levels. ,

This change would also combine the separate action statements in TS 3.7.1.1.a, for 4-loop -  ;

operation witn one or more MSSVs inoperable and TS 3.7.1.1.b, for 3-loop operation while  ;

in Mode 3 with one or more MSSVs inoperable. Table 3.7-2 would also be removed. l 1

The proposed change we'.:!d &c rsvise the action statement (TS 3.7.1.1.a) to reflect that I the plant is to be in hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after entering hot standby. Current TSs require that the plant be in cold shutdown within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> after entering hot standby.

The proposed change would also add an action to require the unit to go to hot standby I within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and hot shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> if 1..a requirements for restoring the valve to operable status or reducing power are not met or if any S/G has less than 2 MSSVs operable.

TS Bases Section 3/4.7.1.1 would be revised to correct an error in which Table 3.7 2 is incorrectly referenced. The correct reference should be to Table 3.7-1.

Other changes have been made to TS 3.7.1.1 for consistency with NUREG-1431.

Reason for Chanae Westinghouse identified a potential problem with the allowable rated thermal power levels in l conjunction with one or more inoperable MSSVs. The relationship between available MSSV relief capacity and allowable rated thermal power (RTP) level may not be linear under all plant conditions. Westinghouse recommended either reanalysis of the plant specific loss-of-load turbine trip transient to maximize the allowable power levels, which may validate RTP I levels in current TS, or revision of TSs to reduce the maximum power level with inoperable {

MSSVs to a level below the heat removal capability of the operable MSSVs. SON proposes to revise TS Table 3.7-1 and the corresponding basea to reflect Westinghouse recommendations to reduce the maximum power levels when one or more MSSVs ere inoperable.

The proposed change will continue to require that the power range HNF trip setpoint be utilized as the means of controlling the RTP Although NUREG-1431 uses administrative control of the RTP withou;the use of the power range high neutron flux (HNF) trip setpoint, TVA chose to retain use of the HNF trip setpoint because administrative controls would require further analysis of the power levels due to additional uncertainties associated with the instrumentation that would be used for administrative control.

The proposed change to combine TSs 3.7.1.1.a, and 3.7.1.1.b,into a single action statement is being requested to simplify the specification by eliminating the conflict with TS 3.4.1. This conflict regards the control of the number of loops required to be in operation. TSs 3.4.1.1 and 3.4.1.2 control the number of loops required to be in operation while TS 3.7.1.1 controls the number of MSSVs required to be opereble. Combining the actions under TS 3.7.1.1 into a single step allows control of the number of operable reactor coolant system (RCS) loops to revert to TS 3.4.1 while continuing to maintain control of the number of operable MSSVs in TS 3.7.1.1. This will also provide consistency with NUREG-1431.

The change to revise the action statement on 4-loop operation (TS 3.7.1.1.a)is being proposed so that the required mode changes are consistent with the modes of applicability.

The addition of Action b., is to clarify the actions to be taken when less than two MSSVs are operable on any S/G. This is consistent with NUREG-1431 and should prevent any potential confusion regarding what actions are to be taken when less than two MSSVs are operable on any S/G. Other changes to TS 3.7.1.1 are made for consistency with NUREG 1431.

The change to Bases Section 3/4.7.1.1 will correct an error in the current TS. Table 3.7-2 is incorrectly referenced, the correct reference should be to Table 3.7-1.

Justification for Chanaes The MSSVs provide overpressurization protection for the secondary system by ensuring that the system pressure will be limited to 110 percent of the design pressure during the most severe anticipated system operational transient. TS Table 3.7-1 establishes maximum allowable power levels when one or more MSSVs are inoperable. These power levels are  ;

established using a linear relationship defined in the Bases Section 3/4.7.1.1. Westinghouse indicated by letter (Reference 2) that the basis for the TS may not be sufficient to ensure that the main steam system pressure remained below 110 percent of design pressure for all conditions associated with the loss-of-load turbine trip transient, which is the most limiting j secondary side pressurization transient. '

At higher power levels the linear relationship is valid. However, at lower power levels, pressurizer pressure /overtemperature delta T reactor trips are not actuated early in the transient; instead, the reactor will trip on low S/G water level later in the transient. Generic analyses performed by Westinghouse have shown that the additional heat generated as a result of the longer reactor trip interval can cause secondary side pressures to exceed 110 percent of design pressure with the current setpcints of TS 3.7.1.1. In Reference 2,  ;

Westinghouse recommended the use of a different calculation to derive the maximum power

3-levels for use with one or more MSSVs inoperable. A calculation has been performed (Raference 3) to determine the maximum power levels applicable to the SON facility and to adjust these values for plant specific instrument and channel uncertainties, in addition, i Westinghouse has evaluated in Reference 4 the potentialimpact of the trip time delay (TTD) and the anticipated transient without scram mitigating system actuation circuitry (AMSAC) and determined no impact from these features on the algorithm or on the conclusions presented in the nuclear safety advisory letter (Reference 2).

Retention of the HNF trip setpoint reduction is acceptable since it will ensure that the necessary power level reductions associated with one or more MSSVs inoperable will not be exceeded. Replacement of the HNF trip setpoint reduction with administrative controls could require a further reduction of the RTPlimits because of the additionalinstrumentation uncertainties, which would have to be accounted for.

i The proposed change to combine TSs 3.7.1.1.a, and 3.7.1.1.b,into a single action .

statement is acceptable since TS 3.4.1.1 continues to require that all four RCS loops be operable in Modes 1 and 2 and TS 3.4.1.2 continues to require that at least two RCS loops be operable in Mode 3. Therefore, it is not necessary for TS 3.7.1.1 to repeat the requirement that the unit be in Mode 3 if only three RCS loops are operable. Also, the RTP levels in the proposed change are designed to be dependent upon the number of operable MSSVs per required operable S/G, rather than upon the number of RCS loops operable. This also explains the rationale for removing Table 3.7-2, in conjunction with the fact that this table would require the power range HNF trip setpoint be reduced for a unit that is already in Mode 3.

The addition of Action b., to define the steps to take if the actions under Action a., are not complete or when any S/G has less than two operable MSSVs is acceptable because:

(1) the requirements, if Action a., is not completed, have not changed; and (2) defining what actions to take, should less than two operable MSSVs exist on any S/G, is appropriate and the actions are more conservative than those required by TS 3.0.3, which would otherwise be invoked. The proposed revision to limit the action requirements to drive unit operation to hot shutdown (Mode 4)instead of cold shutdown (Mode 5)is actmptable because the TS limiting condition of operation is no longer appi; cable upon reaching hot shutdown, in Modes 4 and 5, there are no credible transients requiring the MSSVs. The S/Gs are not normally used for heat removalin Modes 5 and 6, and thus cannot be overpressurized; there is no requirement for the MSSVs to be operable in these modes. This change is consistent with NUREG-1431. Other editorial changes are consistent with NUREG-1431.

The change to Bases Section 3/4.7.1.1 corrects an error in the TS that has previously gone undetected.

References

1. Final Safety Analysis Report, Sections 10.3 and 15.2.7

2. Westinghouse Nuclear Safety Advisory Letter NSAL-94-001

3. SON Calculation SON-01-D053-EPM-TVL-041594, Revision 1

4. Westinghouse Letter TVA-94-156 (Enclosure 4)

y 1 b; .

e ,

4 EnvironmentalImoact Evaluation The proposed change does not involve an unreviewed environmental question because ,

operation of SON Units 1 and 2 in accordance with this change would not:

1. Result in a significant increase in any adverse environmentalimpact previously evaluated in the Final Environmental Statement (FES) as modified by NRC's testimony to the Atomic Safety and Licensing Board, supplements to the FES, environmentalimpact appraisals, or decisions of the Atomic Safety and Licensing Board,

2. Result in a significant change in effluents or power levels.

3. Result in matters not previously reviewed in the licensing basis for SON that may have a significant environmentalimpact.

5 P

F t

t 4

1 6' e ENCLOSURE 3 PROPOSED TECHNICAL SPECIFICATION CHANGE.  !

i SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2 l DOCKET NOS,50-327 AND 50 328 (TVA-SON-TS-94-07)

DETERMINATION OF NO SIGNIFICANT HAZAHDS CONSIDERATION j l

i i

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Significant Hazards Evaluation TVA has evaluated the proposed technical specification (TS) change and has determined that it does not represent a significant hazards consideration based on criteria established in 10 CFR 50.92(c). Operation of Sequoyah Nuclear Plant (SON) in accordance with the proposed amendment will not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated.

This change reduces the power level at which the reactor may be operated with one or more main steam safety valves (MSSVs) inoperable, to ensure that the secondary system is not overpressurized during the most severe pressurization transient of the secondary side. Additionally, this change will combine the TS action statements for 3- and 4-loop operation with one or more MSSVs inoperable, revise the mode requirements and times of Action Statement 3.7.1.1.a, and correct a reference in the bases section to Table 3.7-1. Reduction of the high neutron flux (HNF) trip setpoint will continue to be used as the means to ensure that the required reactor power level reductions are met.

There is no change to the function of the MSSVs by the proposed change. This change will not alter any accident analysis assumptions or results for SON. The proposed change will reduce the amount of relief capacity required to mitigate the consequences of the transient by reducing the total amount of energy in the pnmary system. Therefore, this change will not increase the probability of an accident.

This change is consistent with current SON accident analysis assumptions for the MSSVs and does not change the containment response for any design basis event. Therefore, no change in the mitigation of an accident will result from this proposed change and no change will occur in the consequences of any accident currently analyzed.

2. Create the possibility of a new or different kind of accident from any previously analyzed.

Inadvertent opening of a MSSV is currently analyzed as an initiating event for accidental depressurization of the main steam system. The proposed change does not alter the valves or any other plant component. The valves will continue to perform as analyzed in current accident analyses. The proposed change will not create the possibility for any new or different kind of accident.

By retaining the use of the HNF trip setpoint reduction, no change is being proposed in the methodology used to ensure that power reductions are carried out; therefore, this will not create the possibility of placing the plant into any new unanalyzed condition. The existing accident analysis is still bounding.

)

Combining the separate action statements for 3 and 4-loop operation into a single action does not create the possibility for a new or different kind of accident. Operation with 4 loops will continue to be required in Modes 1 and 2 by TS 3.4.1.1, Operation with less than 4 loops will continue to be governed by TS 3.4.1.2 in Mode 3 and TS 3.4.1.3 in Mode 4. This change will not place the plant in a configuration not currently bounded by existing accident analysis.

Revising the mode requirements and their associated times, consistent with the requirements in NUREG-1431, will continue to ensure that if the unit is unable to comply with the limiting condition for operation, the unit will begin an orderly shutdown until a mode is reached where the specification is not applicable.

3. Involve a significant reduction in a margin of safety.

The proposed change reduces the total energy of the reactor coolant system that will ensure the ability of the MSSVs to perform their intended function as assumed in current accident analyses. This change has been evaluated on a generic basis for Westinghouse Electric Corporation designed 4-loop nuclear steam supply systems. SON plant specific features have been evaluated including power limit calculations and the interaction of the reactor protection system trip time delay and the anticipated transient without scram mitigating system actuation circuitry. Correcting this nonconservatism restores the margin of safety to what was originally envisioned. Therefore, the margin of safety assumed in the accident analysis is not reduced by this change.

Combining the separate action statements for 3- and 4-toop operation into a single action has no effect on the margin of safety for 4-loop operation with one or more MSSVs inoperable. Under the revised TS,3-loop operation with one or more MSSVs inoperable would only be allowed in Mode 3, and 4-loop operation will be required in Modes 1 and 2 in accordance with current TSs 3.4.1.1 and 3.4.1.2.

Revising the mode requirements and their associated times, consistent with the requirements ia NUREG-1431, will not reduce the safety margin since the new requirements will continue to place the unit in a mode where the TS is no longer applicable. The new completion times for mode changes are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner without challenging unit systems.

1; 4

.c r ENCLOSURE 4 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2-DOCKET NOS,50-327 AND 50-328 (TVA-SON TS 94-07)

Wti3TINGHOUSE ELECTRIC CORPORATION EVALUATION OF PROPOSED TS CHANGE 1

1 (m)

Westinghouse Energy Systems 8fghremoniamnm Electric Corporation Mr. Mark J. Burrynski, Manager TVA-94-156 Department of Nuclear Engineering NTD-NSRLA OPL-94-230 Tennessee Valley Authority August 29,1994 sequoyah Nuclear Power Plant, DSC-A P. O. Box 2000 Soddy Daisy, TN 37379 Ref.: 1) N94-030 '

2) NSAL-94 001 Tennessee Valley Authority Sequoyah Nuclear Plant Ooeration At Radntad Power Levels With Inonerable Main Stam Safety Valves Review of TVA Marked-un Tehnical Seecification 3.7.1.1 l

1

Dear Mr. Burzynski:

In response to your request, an evaluation is attached addressing the proposed revised Technical '

Speeltication 3.7.1.1 (also attached) for Operation At Reduced Power Levels With Inoperable MSSVs. This evaluation was previously transmitted to you via fax on August 19,1994 to meet your schedular requirement.

If you have any questions, please do not hesitate to contact us.

Very truly yours, D. W. Salak, ager j Sequoyah Project Manager j TVA Projects l LVT/bbp cc: D. M. Lafever NB11.A230 LAVA 1SW8/2416 I

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Operation At Reduced Power Levels With Inoperable Main Steam Safety Valves Pranaamd TVA Ravland Technical Sneifh*Ian 3.7L1 Format and TTD/AMSAC Intaraction Evaluation l

Introduction l

l Westinghouse identitled in Nuclear Safety Advisory Letter NSAL 94 001 a deficiency in the buss for

{

Technical Specification 3.7.1.1. This Technical Specification allows the plant to operate at a reduced power level with a reduced number of operable Main Steam Safety Valves (MSSVs). The reduction i

in the power level is ensured by the Technical Specification requirement to reduce the maximum allowable High Neutron Flux (HNF) setpoint. The HNF setpoint varios depending upon the number i of inoperable MSSVs. The deficiency identiflod in NSAL 94 001 is that the maximum allowable  ;

HNF setpoint is not a linear function of the available MSSV capacity, as noted in the bases of the i current Technical Specifications. l To address the deficiency, NSAL-94-001 presents a conservative algorithm that can be utilized to '{

calculate revised HNF setpoints for 1,2 or 3 inoperable MSSVs per loop. However, TVA wishes to modify th6 Technical Specification format to be more consistent with that recommended by the MERITS Technical Specification format. That is, rather than present the HNF setpoints for 1,2, or 3 inoperable MSSVs per loop, designated and administratively controlled reduced power levels, which are also based on the conservative algorithm presented in NSAL-94401, would be required without requiring a reduction in the HNF trip setpoint.

NSAL-94-001 provides sufficient information to allow for the calculation of the revised power levels for Technical Specification 3.7.1.1 with a reduced number of operable MSSVs. However, an issue specific to the Sequoyah units, related to the potential interaction between the Trip Time Delay (TTD) system and the ATWS Mitigating System Actuation Circuitry (AMSAC), may invalidate the assumptions used in the algorithm presented in NSAL-94-001. This issue and the acceptability of the modified Technical Specification to a MERITS type format, i.e., use of acceptable power leveh rather than HNF setpoints, based on the algorithm provided in NSAL 94 001 are addressed in the evaluation presented below.

Evaluation The Technical Specifications for the Sequoyah units allow operation at a reduced power level as dletated by the reduced HNF setpoint with inoperable MSSVs. The algorithm provided in NSAL-94-001 was developed to provide a conservative basis for determining the maximum power level from which a plant could experience a loss of load transient without exceeding 110% of the main steam system design pressure. Currently, the Technical Speelf! cations require that the HNF serpoint be reduced for operation with 1,2, or 3 Inopernble MSSVs on any loop. This requirement was originally built into the Technical Specifications as a convenient means to force a corresponding power reduction. The HNF setpoint conveniently places a cap on the maximum core power generation. In the presence of a positive moderator temperature coefficient (MTC), a loss ofload event would result in a power increase above the initial power. For plants, such as the Sequoyah units, which are not licensed to operate with a positive MTC, no power increase above the initial NSRLADOL/IVAl*I"#96 ra rs93 sn3 rom N3 tem u-m mn t o n:' m

. ., m 3 _ ,

level would result during the limiting loss of load event. This ensures that the power levels associated with inoperable MSSVs will not be violated, thus negating the need for a corresponding reduction in h the HNF reactor trip setpoint. Therefore, the modified Technical Speelfication proposed for the Sequoyah plant is acceptable.

With respect to the AMSAC and TTD interaction, the following is presented. The TTD is a system of predetermined programmed steam generator low low level reactor trip and auxiliary feedwater delay times that are based upon 1) the prevailing power level at the time a low-low level trip setpoint j l

is reached, and by 2) the number of steam generators that are affected. In the Sequoyah TfD design, trip time delays are only actuated below 50% RTP and are determined by two delay curves (single or multiple steam generators with low inventories). The design of AMSAC, as described in i WCAP-10858-P-A, provides an independent backup to the existing reactor protection system to I initiate a turbine trip and actuate auxillary feedwater flow fbilowing an ATWS event above 40% RTP. q In the 40% to 50% RTP range, if the water levels in 3 or 4 steam generators drop below the AMSAC )

setpoint, the AMSAC and TfD systems will both be actuated. Since, in this power range, the  ;

AMSAC delay time is shorter than the 'ITD delay time, the turbine will be tripped and the auxiliary I feedwater system will be started by AMSAC before reactor trip occurs. A Loss of Normal Feedwater l (LONF) or Feedline Break (FLB) transient initiated in this power range could experience a loss of steam load transient prior to reactor trip in addition to the loss of main feedwater flow. l A turbine trip will result in a rapid reduction in the heat removal capability of the secondary system l due to a loss of steam flow. With fbli MSSV capacity, the secondary system pressure limit (110% of design) would not be challenged at this reduced power level. However, if operating in the 40 to 50%

power range with inoperable MSSVs, and therefore reduced capacity, the main steam system pressure limit may be challenged for the LONF and FLB cvents, The proposed modification to Technical Specifications would require that the plant reduce the operating power level to less than or equal to 45% RTP with 2 inoperable MSSVs on any loop.

Therefore, operation in the 40% to 50% RTP range, with inoperable MSSVs, will be allowed by the Technical Specifications. This condition has been considered in an analysis of the LONF and FLB transients to confirm that AMSAC actuation and turbine trip, prior to reactor trip, will not invalidate the results of the NSAL-94-001 algorithm. The analysis conservatively assumes a maximum initial power level of 51 % RTP to allow for uncertainties in the measurement of the reduced power level and a reduced MSSV relief capacity corresponding to 2 inoperable MSSVs per loop. The results demonstrate that a turbine trip actuated by the AMSAC system during either a LONF or FLB event {

will not result in overpressurization of the main steam system. In the case of the LONF event, the relief capacity is sufficient to maintain system pressures and RCS cooling without an immediate reactor trip. Reactor trip occurs after the trip time delay expires. The FLB transient results in a reactor trip on a low steam pressure signal in the faulted loop prict to the time that the cooling capacity of the secondary is significantly reduced. Therefore, the calculated steam system pressure does not seriously challenge the !!mit.

Plant operation with 3 inoperable MSSVs on any loop requires a power reduction to less than 40% RTP, therefore, no TfD/AMSAC interaction would occur. The plant may operate in the 40%

to 50% RTP range with 1 inoperable MSSV on any loop, however, this condition is bounded by the evaluation above for 2 inoperable valves.

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Conclusions The analysis and evaluations described within demonstrate that a modified Technical Specification requiring a power reduction only, without a corresponding reduction in the HNF setpoint, is acceptable for the Sequoyah units. As noted within, this assumes that the plants do not operate with a positive moderator temperature coemclent. An analysis also demonstrated that interaction of the TTD and AMSAC systems do not invalidate the conclusions of NSAL-94-001 nor the application of the 1 algorithm to determine the required reduced power levels for inoperable MSSVs. Thorofore, the ,

recommended Technical Specification changes to 3,7.1.1 can be supported and do not invalidate the l conclusions of the Sequoyah FSAR.

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• may proceed p ,,,idad, v that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />,. either the inoperable valve is re, stored OPERA 8L!

status or 1;t.. T 2;r "e.

. per Table 3.7-11 otherwlsel,L.^ r^ T I '-- " ' ' ' - ^ ' - ' ' ' - i s re duce d be in at least H07 STANOBY 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 SHUT ghours.

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h. The provisions of 5pecification 3.0.4 are not applicable.

SURVEILLANCE REQUIR_EHEHTS sSpectficatton

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O S TABLE 3.7-d -

k' 1

E STEAM tlNE SAFETY VALVES PER LOOP o E

VARVE InsEtt LIFT SEITING (1 1%)* tt0ZZLE 512E E loop 1 I.oop 2 Loop 3 Loop 4

[m a

1-1-522 1-1-517 1-1-512 1-1-527 1964 psig 16 sq. in. 3 1-1-523 1-1-518 l-1-513 1-1-528 1977 psig 16 sq. in. {

1-1-524 1-1-519 1-1-514 1-1-529 1090 psig -

~

15 sq. in.

1-1-525 1-1-52D 1-1-515 1-1-530 1103 psig 45 sq. In.

1-1-526 1-1-521 1-1-516 1-1-531 1117 psig 15 sq. In.

g 3

0 Y E

- 8 1

o \u .

5 2

3

• The ilf t selling pressure shall correspond Le ambient condf Liens of the valve at nominal operating E

[ O r tenyerature and pressure.

2 S

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3 x

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g 08/10/94 08:18 C818 843 $s03 SEQUong MECHNUC 3014 ,,

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. l' 3/2.7 PLANT SYSTEP.5_

l j' BASES --

j, 3/4.7.1 TURSINE CYCLE i'

3/4.7.1.1 SAFETY VALVil5 The OPERABILITY of the main steam line code safety vaivas ensuras that ,

the secondary system pressure will be limited to within 110% (1194 pstg) of '

the system casign pressure during the cost severt anticipat'ed systam operational transient. The maximum ralinving capacity is associated with a turbina trip from 100 RATED THERMAL POWER coincident with an assumed loss of condanser haat sink (i.e. , no steam bypass to the condenser).

The specified valve lift settings and relieving capacities are in accordance with the requirements of Section !!! of the A5ME Boilar and Pressure Code, 1971 Edition. The total relieving capacity for all valves on all of the steam ifnes is 1.9 x 107 lbs/hr at 1170 secon.dary steam f)ow of 1.493 x lbs/hr 10psig which at 100% RATEDis 127 percent THERMAL of theA total POWER.

minimum of 2 ' OPERABLE safety valvas par steam ganarator ensures that sufficient relievi writgle available for the allowabia THERMAL POWER rastriction in Table l ' 1 3 /7- 9 STARTUP and/or POWER OPERATION is allowabla with safety valves inoperable

{ within the limitations of the ACTION requiramants n th R th reduction fe,'

.- d a ve.,  % ..rs. :., = tcd. w "p ,

r::.:tn tri; 5th.;;: .. .... . ;c ..=;; =v. fM ;hrt:h. 'h: 7'.::t:r  :

--tr4; . e tpe k.t . e.~ e t kr.: ei; driv;d L' 'h- SII;Uh- 5;;^";

• For 4 loop a ration  ;

sp = N f # x (105)

X .

1 l

for loop operatio  :

"I 5 = x (75)

X i

k are: l SP = reduce reactor trip s point in parcant f RATED THERMAL OWER V = maximum n .cer of inopera la safety valves r steam lina U maximum numb r of inoperabi safety valvas par operating tamm i!ne. ,

109 = Po er Range Hau on Flux-High T p 5etpoint for 4 op i opa. tion. l 1

= Maximu. percent of r 'TED THEF. MAL Pos{;; pennissible by P-S Set int for 3 lo operation. \ ,

IR-1 SECW F_AH - UNIT 1 3/Q-Q e mnu on,cnhi"d'r . f*Gc,h,13 1990 l

.wi o8/lo/84 04:1e Cell 843 8333 SEQUoYAE MECINUC , 4 01s I y4 i

P!. ANT $YSTEW.5 I

' \

BASES 1

x/ y v v =

X a tal relie ng capacit f all saf valves p if in 1bs/h 4.75 x 8 lbs/ hour 1170 psi .

steam (88 0

= Maxim reifevin apacity a ny one safa valve in ga r lbs/hou 950,000 / hour at 1 psig. A  !

/1 i 3/4.7.1.2 AUXILIARY FEEDVATER 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 operating conditions in the event of a total loss of off sita power. '

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

'gt

,1100 psia the open steam generator safety valve (s) are capable of relieving at least 11% of nominal steam flow. A total feedwater flew of 440 spa at .

pressures of 1100 psia is sufficient to ensure that adequate feedwater flow is availabla to remova decay heat and reduce the Rasctor Coolant System u

- tamperature to less than 350'F whara the Residual Hast Ramoval System may be placed into operation. -

will provide at laast 440 gpa plus pump recirculation flow against ua generator pressura of 1100 psia. i a.

Each motor-driven auxiliary feedwater pump (ona Train A and one Train 8) l l

supplies flow paths to two steam generators. Each flow path contains an N automatic air-operated level control valva (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 genera tors. Each of these flow paths contains an automatic air oparatad 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 8-train motor-drivan 3g, auxiliary feedwater pump. The remaining two LCVs are designated as Train 5, ,

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 i redundancy to ensure that at Isast 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 5) will not  !

prevent the auxiliary feedwater systas from performing its intended safety functicn and is no more severs than the loss of a single auxiliary feedwater ,

pump. Therefore, the loss of a single train of auxiliary air only affects the  !

capability of a single motor-driven auxiliary feedwater pump becausa the turbine-drivan pump is still capable of providing flow to two steam generators t, hat are separata from the other motor-driven pump. .

Two redundant staam sources are required to be operable to ensure that at '

least one sourca is available for the steam-driven auxiliary feedwater (AW) l pump operation following a feedwater or main steam line break. This require-ment intact ensures that the p)lant steam generators givenremains the event witnin of aits lessdesign of thebasis No. I(f.e.,

steam ANgenerator to two j

_