ML20212K325

Proposed Tech Specs,Consisting of Tech Spec Change 77,adding Requirement to Limiting Condition for Operations 3.3.1.1 & 3.3.1 Re Operability of New Shunt Trip Attachment to Reactor Trip Breakers

ML20212K325
Person / Time
Site:	Sequoyah
Issue date:	01/20/1987
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20212K304	List: ML20212K301 ML20212K325
References
NUDOCS 8701290049
Download: ML20212K325 (23)

Text

.

TABLE 3.3-1 (Continued) m

@ REACTOR TRIP SYSTEM INSTRUMENTATION S

N

. MINIPUM c TOTAL NO. CHANNELS CHANNELS APPLICABLE y FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERA 3 E MODES ACTION

19. Safety Injection Input from ESF 2 1 2 1, 2 12

20. Reactor Trio Breakers 2 1 2 1, 2, and

• 12,15 t 121';' '-~ 0"~' ~

z , , s <, n =, .a s- n

21. Automatic Trio logic 2 1 2 1, 2, and

• 12 o

22.

I. E7[1'fl I"" W'"

Reactor Trip System Interlocks 2 i 3:,94 , ,,a ss ;g j A. Intermediate Range R Neutron Flux P-6 2 1 2 2, and* 8a

• Power Range Neutron 8.

Y Flux - P-7 4 2 3 1 8b-

• C. Power Range Neutron Flux - P-8 4 2 3 1 8c D. Power Range Neutron Flux - P-10 4 2 3 1, 2 8d i E. Turbine Impulse Chamber l

ga E Pressure - P-13 2 1 2 1 8b D F. Power Range Neutron h ~ Flux - P-9 4 2 3 1 8e

) G. Reactor Trip - P-4 2 1 2 1, 2, and* 14 i q s 58m

• 2 l r9 I

T gpag%$TsNkl* e P

F O 9

?

TABLE 3.3-1 (Continued)

ACTION 8 - With less than the Minimum Number of Channels OPERABLE,' declare the interlock inoperable and verify that all affected channels of the functions listed below are OPERABLE or apply the appro-priate ACTION statement (s) for these functions. Functions to be evaluated are:

a. Source Range Reactor Trip -

b. Reactor Trip

. Low Reactor Coolant loop Flow (2 loops)

Undervoltage Underfrequency Pressurizer Low Pressure Pressurizer High Level

c. Reactor Trip Low Reactor Coolant Loop Flow (1 loop)

d. Reactor Trip Intermediate Range

. Low Power Range Source Range

e. Reactor Trip ,

Turbine Trip ACTION 9 -

Deleted ACTION 10 - Deleted ACTION 11 - Deleted ACTION 12 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be bypassed for up to 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing per Specification 4.3.1.1.1 I

, provideQhe other channel is OPERABLE.

I 2

SEQUOYAH - UNIT 1 3/4 3-7 #seedment ;o _ _ - - - . . _ _ - _ .. ,_ _ _. _ _

?,

TABLE 3.3-1 (Continued)

ACTION 13 - With the number of OPERABLE channels one less than the Total -

Number of Channels and with the THERMAL POWER level above the P-7 (Block of Low Power Reactor Trips) setpoint, place the inoperable channel in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, operation may continue until performance'of the next required CHANNEL FUNCTIONAL TEST.

ACTION 14 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Ad%d /~5~-

WsTM odd of tH' E b:yERSc TRIP j~fArtdCJ ( uunCRnrgcc 02 SHuNr TRIP *ATTdcHmKMT-) 1po7cepnis , RESrces Z7 To o?ERABLE STHTuS Virnia yg //oug3 g DEctAff THE 'Bf64Kf'R ENofEEABLE ff/D Affty

, Action /2. -rRE 'Besarcr. .sHnu por DE BYP S.SEb A WHLEI ONE o f T H E p iv'Ef f& TIP FferttPE.5 ES ZN O'tYO60l6. EtcEFT f~ot UP 7b 'l HauR $ (o g TreFogetidG ' MAIN 75 NANCE To TESf0RE 77/E

'BREAKEE 7b OPERA 6LE SrdmS.

ACr/odIt" WITH THE Mutn 6 E2 of OWRA6lf C/M/MELS ode LGI -HM 1k6 A11N1pium CH4NNGLS Offggg g REQ 4tREn!ENr, VE.STOfE THE *ENo?g,ggpl5 gggggg[

To 07t*2A6L E STRTt)S WIfHiN 'f9 RouRS of OTVH fMS TEACTof. Trip ffEsit'GRS wIritia 7-HE ttE >c T* Ao d -

SEQUOYAH - UNIT 1 3/4 3-8 AundmentWh

( , ,

( M ~

TABLE 4.3-1 y 8 .

'~

3 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS >

[. E

. c CHANNEL MODES IN WHICH

5 CHANNEL i CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED f ] 4

1. Manual Reactor Trip N.A. N.A. S/U(1) Ano f(9) 1, 2, and *

2. Power Range, Neutron Flux S . D(2),M(3) Q 1, 2 and Q(6) ,

.! 3. Power Range, Neutron Flux, N.A. R(6) Q 1, 2 High Positive Rate i

4. Power Range, Neutron Flux, N.A. R(6) Q 1, 2 High Negative Rate

• 5. Intermediate Range, S R(6) S/U(1) . 1, 2, and

• j Neutron Flux

{

6. Source Range, Neutron Flux S(7), R(6) M and S/U(1) 2, 3, 4, S, and *

7. Overtemperature Delta T S Oti!' R H 1, 2

]. f p' ' F

8. Overpower Delta T S R M 1, 2 l 9. Pressurizer Pressure--Low S R Q 1, 2 P-:

$ N 10. Pressurizer Pressure--High S R Q 1, 2 p ;c

11. Pres'surizer Water Level--High S R Q 1, 2

) i :;o

12. Loss of Flow - Single Loop S R Q 1

[ 13. Loss of Flow - Two Loops S R N.A. 1 se i

" TABLE 4.3-1 (Continued) c"s h REACTCR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE RtQUIREMENTS N MODES IN milch

' CHANNEL CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE

& CHECK CALIBRATION TEST REQUIRED q FUNCTIONAL UNIT R Q 1, 2

~

14. Main Steam Generator Water 5 Level--Low-Low S R Q. 1, 2

15. Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level N.A. M- 1

16. Undervoltage - Reactor Coolant R Pumps M 1

17. Underfrequency - Reactor Coolant N.A. R Pumps

18. Turbine Trip A. Low Fluid Oil Pressure N.A. N.A. S/U(1) 1 w

Turbine Stop Valve Closure N.A. h.A. S/U(1) I 2 B. .

N.A. M(4) 1, 2 w 19. Safety injection Input from ESF N.A.

5 Reactor Trip Breaker N. A. ,', N.A. M(5) and S/U(1) 1, 2, and *

20. a N.A. N.A. M(5) 1, 2, and ^

21. Automatic Trip Logic -

22. Reactor Trip System Interlocks 2, and

• A. Intermediate Range N.A. R S/U(8)

Neutron Flux, P-6

6. Power Range Neutron N.A. R S/U(8) 1 Flux, P-7 C. Power Range Neutron N.A. R S/U(8) 1

? Flux, P-8 1, 2

?ower Ra'ge n Neutron N.A. R S/U(8) d' :- "

O.

Flux, P-10 g g , E. Turbine Impulse Chamber N.A. R S/U(8) 1

<+ - Pressure, P-13 F. Power Range Neutron N.A. R S/U(8) i 4: ;;

Flux, P-9 S/U(8) 1, 2, and

• 0 (f G. Reactor Trip, P-4 N.A. R

23. REner** rit! BYF4j.S pts;;ce y,A.

y 'j G4 y p g' I At(to)]{IIh

I l

f INSTRUMENTATION k

TABLE 4.3-1 (Continued)

NOTATION With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.

(1) -

If not performed in previous 7 days.

(2)

• fleat balance only, above 15% of RATED THERMAL POWER.

Adjust channel if absolute difference greater than 2 pe'rcent. , , , .

s . . . . ,

(3) -

Compare incore to excore axial ffux' difference ab'ove 15% of RATED THERMAL POWER. Recalibrate if-the absolute difference greater than or equal to 3 percent.

(4) -

Manual ESF functional input check every 18 months.

(S) -

Each train or logic channel shall be tested at least every 62 days on a STAGGER T Tdsv- snow mostmWar v=QsgTBQ,gQ{ST e, , BASIS. THEV"O*'ouw C huo Au ro ~^s-Qn

;q,'t Of W (6) -

utron detectors,may be excluded from CHANNEL CALIBRATION.

(7) -

Below P-6 (Block of Source Range Reactor Trip) setpoint.

(

(8) -

Logic only, each startup or when required with the reactor trip k

- system breakers closed and the control rod drive system capable of rod withdrawal if not performed in previous 92 days.

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i[lAi\20IUN b SEQUOYAH - UNIT 1 3/4 3-13 --Amend.wt L. 42 1

l I

INSTRUMENTATION BASES The measurement of response time at the specified frequencies provides assurance that the protective and ESF action function associated with each channel is completed within the time limit assumed in the accident analyses.

No credit was taken in the analyses for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential, overlapping .

or tot'al~ ~ channel ' test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or

2) utilizing replacement sensors with certified response times.

ZNSERT~ @ (por Aty) 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION

/

The OPERABILITY of the radiation monitoring cnannels ensures that 1) the radiation levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded. ~~'

3/4.3.3.2 MOVABLE INCORE DETECTORS The OPERABILITY of the ,vable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core. The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve.

For the purpose of measuring F (Z) or-F g a full incore flux map is used.

9 Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibrati6n of the excore neutron flux detection system, and full incore flux maps or symmetric incore thimbles may be used for nonitoring the QUADRANT POWER TILT RATIO when one Power Range Channel is inoperable.

3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features i.mportant to safety. This capability is required to permit comparison of the measured response to that used in the SEQUOYAH - UNIT 1 B 3/4 3-2

e- .-

INSERT A U-1, SEQUOYAH NUCLEAR PLANT, PAGE B 3/4 3-2 Action 15 of Table 3.3-1, Reactor Trip System Instrumentation, allows the breaker to be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for the purpose of performing maintenance. The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is based on a Westinghouse analysis performed in WCAP 10271, Supplement I which determines bypass breaker availability.

t

- - I 'i . ,-

'~%( ,

u  ;,

O O

O e ,

e UNIT 2 PROPOSED CHANCES e

i i

f I

TABLE 3.3-1-(Continued) .

g REACTOR TRIP SYSTEM INSTRUMENTATION.

?

c3 MINIMUM TOTAL NO. CHANNELS CHAhNELS , APPLICABLE 7 FUNCTIONAL UNIT OF C!!ANNELS TO TRIP OPERABLE MODES ACTION 2 19. Safety injection Input -

G from ESF 2 1 2 _1, 2 12 m

20. Reactor Trio Breakers 2 *

s. sroeruP nofewett arresrm 1 2 '

1, 2, and

• 12,sS s.s m oawn 1 a n 3,> p> w gnr ,g

21. Automatic Trip Loci 2 1 2 1, 2, and ^ 12 a srnrur ANO vowen AENM p- snerco t N. 2- 0 I

22. Reactor e rip System Interlocks 3 *, 4 *1 aaa Sh(

A. Intermediate Range Neutron Flux, P-6 2 1 2 2, and" 8a B. Power Range Neutron Flux, P-7 4 2 3' I 8b a

E C. Power Range Neutron o flux, P-8 4 2- 3 1 8c b

0. Power Range Neutron Flux, P-10 4 2 8 3 1, 2 8d E. Turbine Impulse Chamber Pressure, P-13 2 1 2 1 8b F. Power Range Neutron Flux, P-9 4 2 3 1 8e G. Reactor Trip, P-4 2 1 2 1, 2, and

• 14 Ff .  :

• TABLE 3.3-1(Continuedj ACTION 8 - With less than the Minimum Number of Channels OPERABLE, declare -

the interlock inoperable and verify that all affected channels of the functions listed below are OPERABLE or apply the appro-priate ACTION statement (s) for those functions. Functions to be evaluated are:

~

a. Source Range Reactor Trip.

b. Reactor Trip Low Reactor Coolant Loop Flow (2 loops)

Undervoltage -

Underfrequency Turbine Trip Pressurizer Low Pressure Pressurizer High Level

c. Reactor Trip Low Reactor Coolant Loop Flow (1 loop)

, d. Reactor Trip -

Intermediate Range Low Power Range Source Range ACTION 9 - Deleted ACTION 10 - Deleted ACTION 11 - Deleted ACTION 12 - With the number of OPERABLE channels one less than required by the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be bypassed for up to Ehour for surveillance testing per Specification 4.3.1.1.1 provided the other channel is OPERABLE.

2 SEQUOYAll - UNIT 2 3/4 3-7 Amendent--No . - _ . . _ - . . . - - . . . - . , _ - . -- --- . _ - _ - - . - - - - - ._.

TABLE 3.3-1 (Continued)

ACTION 13 - With the number of OPERABLE channels one less than the Total Number of Channels and with the THERMAL POWER level above the P-7 (enable reactor trips) setpoint, place the inope.rable channel in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, operation may continue until performance of the next required. CHANNEL FUNCTIONAL TEST.

I. ,

ACTION 14 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

$$7/00 /S' WITM oAE of t'HE bivCRSC .7751P /~fdTidLS ( trtt0CRtot. TAC 6 O'R SNuNr. ~fRIP fflYcHMEdr) -TNOTCedBLG , f6STbtS ZT To OTERABLE SfdTP.S VITl//n/ 'f8 //ott23 g DEcures -rys Se6sys? rnorcessis gio gypiy Accod u . T/IE 3RG M M ftMLL por 86 BYPp.UGb Wntt E DNS of" -rHE DIVEg3E TRsP FE67kPE3 ES 7907t~fdBl6 EycGtY foK UP' 7b V Aba!J (~o g 9tERFogm/dG ' MAIN 75NANc2 70 TESTOR6 THE 13RGAKER 7b OP$9Adl6 SfdMS.

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REQnoREn1EMr, 26.$70RE f//E ZNo'?sPAdlE CHppNEL To .076f/16iE Sf/1ftJ.S Wt rHiA! 'fB /louBS of OTE8 f//E TEAC70f. Trip ffE4KGRS WITHis

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SEQUOYAH - UNIT 2 3/4 3-8 -A5.end"9t--No c .-__.- - - -- - _ - - -

e: .

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~

i z.

v, g TABLE 4.3-1 REACTOR' TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS

i CHANNEL MODES FOR WHICH F s ,

CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS c'

L

$ FUNCTIONAL UNIT CHECK CALIBRATION, TEST REQUIRED f 1. Manual Reactor Trip N.A. N. A.~ S/U(1) Awo R@) 1, 2, and * '

c

{ 2. Power Range, Neutron Flux 5 D(2), M(3) Q 1, 2 4 and Q(6) ,

3. Power Range, Neutron Flux, N.A. R(6) Q 1, 2 High Positive Rate h 4. Power Range, Neutron Flux, N.A. R(6) Q 1, 2 High Negative Rate I 5. Intermediate Range, S R(6) S/U(1) -

1, 2, and

• y Neutron Flux

6. Source Range, Neutron Flux S(7) R(6) M and S/U(1) 2, 3, 4, 5, and *

7. Overtemperature AT S' R H 1, 2

> B. Overpower AT -

H 1, 2 I Sf,i.,,N,IR

,p 9. Pressurizer Pressure--Low S R Q 1, 2

, ,E 10. Pressurizer Pressure--High S R Q 1, 2

• E 11. Pressurizer Water Level--High .S R Q 1, 2 1 .--

ir y 12. Loss of Flow - Single Loop S R Q 1 Y

13. Loss of Flow - Two Loops S R N.A. 1

14. Steam Generator. Water Level-- S R Q 1, 2 Low-Low '

~

o 4

T'ABLE 4.3-1 (Continued) r M ~

S REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIf6MENTS

. f$

MODES FOR WHICH E CtfAMEL SURVEILLANCE IS-

CHANNZL CHANNEL FUNCTIGAAL c TEST REQUIRED Ch2CK CALIBRATICM

{

FUNCT!GNAL UN!T S R Q 1, 2 i 15. Steam /Feedwater Flow Mismatch and ,

Low Steam Generator Water Level M 1 Undervoltage - Reactor Coolant N.A. R "

j 16.

Pu=ps '

H 1 i

17. Underfrequency - Reactor Coolant N.A. R i

I Pumps e

18. Turbine Trip S/U(1) 1

' w N.A. N.A.

i > A. Low fluid Oil Pressure 1 h.A. N.A. S/u(1)

Y 3. Turbine Stop Valve Closure M(4) 1, 2

! U Safety Infection Input from ESF N.A. N.A 19.

N.A. M(S) and S/U(1) 1, 2, and

• Reactor Trip Breaker N.A.

20.

N.A. h(5) 1,,2, and

• 1 Automatic Trip Logic N.A. ,

21.

22. Reactor Trip System Interlocks 2, and

• N.A. R S/U (8) l j

?

6

) A. Intermediate Range Neutron Flux, P-6 S/u (S) 1 Power Range Neutron N.A. R l 9 j, B.

l t o Flux, P-7 S/U (S) 1 Power Range Neutron N.A. R Nh C.

ll g Flux, P-8 Power Range Neutron N.A. R S/u (6) 1, 2

y 3.

Flux, P-10 S/U (6) 1 1

Turbine Impulse Chamber N.A. R i

E.

Pressure, P-13 i

F. Power Range Neutron S/U (S'i 1 .

Flux, P-9 N.A. R 1, 2, and

• l' N.A. R S/u (8)

G. Reactor Trip, P-4 '

23. TGA:fbE fA'# 8YDr.tf SEEMEA N 4- ., . N.4. Pflia)T[lI) l 2) ad

• i

INSTRUMENTATION -

TABLE 4.3-1 (Continued) -

NOTATION With the reactor trip system breakers closed and the control rod.

drive system capable of rod withdrawal.

(1) -

If not performed in previous 7 days.

(2) -

Heat balance only, above 15% of RATED THERMAL POWER. Adjustchannel if absolute diff,erence greater than'2 percent.

.y . -

(3)

Compare incore to excore axial ' flux differenc'e above 15% of RATED THERMAL POWER. Recalibrate if the absolute difference greater than or. equal to 3 percent.

(4) -

Manual ESF functional input check every 18 months.

(5) -

Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST BASIS googgy.,rv oc rac u. 7w resr mas.

noctws.s ierow~ocurw

^<c A*'o wascy Aur*" Ar'<- $"uur (6) -

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(7) -

Below P-6 (Black of Source Range Reactor Trip) setpoint.

(8) -

Logic only, each startup or when required with the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal if not performed in previous 92 days.

(9)- 77M Q UANH et- f u n c.rros] r x s r* ::r y g t n.

/nxAwrl.y vaur1 rec ommiewy se ruc winwourmc j18o .S &HN 7 TW/P d1/2d u sr3. /@rl ryg fdyt<sqg_.

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l (10 ) - Lock. M4NaAt. $Nasf Tf2/P TRlbf To TIAC//G Bff/M'2A 1$ SGNicC. [ACH 1XAlf C//kL BE TEST &O AT' tEdJT'

(*y2eY f2 CAYJ cal A STAG 6G26D T6ST BASIS.

0 0 - AahAkrr AND Ndrant taloEMcWM6 11'IP l

v SEQUOYAll - UNIT 2 3/.4 3-13

.- 6 INSTRUMENTATION BASES REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM -

INSTRUMENTATION (Continued)

The measurement of response time at the specified frequencies provides assurance that the protective and the engineered safety feature actuation associated with each channel is completed within the time limit assumed in the accident analyses. No credit was taken.in the analyses for those channels with' response times indicated as not applicable.

' Response time may be demonstrated by any series of sequential, overlapping

. or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may

_ be demonstrated by either 1) in place, onsite or offsite test measurements or

2) utilizing replacement sensors with certified response times.

_TNSE/tr @ (mr- Face) 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPIRABILITY ~of the radiation monit'oring channels ensures that 1) the radiation levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.

-  ?

s- (%

~

. . '[5

'NT

~ ~

3/4.3.3.2' MOVABLE'IllCORE" DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of

, this system accurately represent the spatial neutron flux distribution of the reactor core. The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve.

N For the purpose of measuring F a full incore flux map is used.

Quarter-core flux maps, as defined 9n(Z) or FWCAP-8d8, June 1976,maybeused i recalibration of the excore neutron flux detection system, and full incore flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Pcwer Range' Channel is inoperable.

3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient capability is availeble to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. .This capability' is required to permit comparison of the measured response to that used in the design basis for the facility to determine if plant shutdown is required pursuant to Appendix "A" of 10 CFR Part 100. The instrumentation is consistent with the recommendations of Regulatory Guide 1.12, " Instrumentation for Earthquakes,"

April 1974 -

SEQU0YAli - UNIT 2 - B 3'/4 3-2

INSERT B U-2, SEQUOYAH NUCLEAR PLANT, PAGE B 3/4 3-2 ,

Action 15 of Table 3.3-1, Reactor Trip System Instrumentation, allows the breaker to be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for the purpose of performing maintenance. The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is based on a Westinghouse analysis perform?d in WCAP 10271, Supplement I which determines bypass breaker availability.

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ENCLOSURE 2 PROPOSED TECHNICAL SPECIFICATION CHANGES SEQUOYAH NU& CLEAR PLANT UNITS 1 AND 2 (TVA SQN TS 77)

Justification to add requirements to LCO 3.3.1.1 (unit 1) and LCO 3.3.1 (unit 2) technical specifications for operability of the new shunt trip attachment and also incorporate testing requirements for the reactor trip bypass breakers SR 4.3.1.1.1 (unit 1 and unit 2).

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Description of Channe This technical specification change will add requirements to LCO 3.3.1.1 ,

(unit 1) and LCO 3.3.1 (unit 2) for operability of the new shunt trip attachment to the reactor trip breakers (RTB). Further additions are for surveillance of these attachments and consequential inoperability of the RTBs.

Additionally, this change will add requirement to SR 4.3.1.1.1 (unit 1 and unit 2) to require shunt trip attachment operability of the bypass breakers be '

ver.fied before being placed in service. The surveillance requirement will also independently verify the operability of the undervoltage trip attachment at least once per 18 months. (See enclosure 1.)

Reason for Channe Item 4.3 of Generic Letter 83-28, " Required Actions Based on Generic Implications of Salem Anticipated Transient Without Scram (ATWS) Events,"

established NRC requirements (Generic Letter 85-09) for automatic actuation of the shunt trip attachment for Westinghouse plants. NRC concluded that technical specification changes should be proposed to explicitly require independent testing of the undervoltage (UV) and shunt trip attachments during power operation and independent testing of the control room manual switch contacts during each refueling outage. These tests are necessary to ensure reliable RTB operation.

Present technical specifications allow one channel for the RTB to be bypassed for up to one hour for the purpose of surveillance testing provided the other channel is operable. Due to the additional surveillance requirement for the auto shunt trip attachment and current problems in meeting the one-hour requirement, Sequoyah Nuclear Plant (SQN) is requesting that this time limit be extended to two hours.

NRC has also required adding surveillance testing on the bypass breakces. TVA had originally declined to provide these technical specification changes on the basis that Westinghouse Owners Group (WOG) calculations have shown no significant reliability improvement from including periodic surveillance tests of the bypass breakers in the technical specifications.

In a July 28, 1986 letter from B. J. Youngblood to S. A. White, NRC rejected TVA's proposal to omit bypass breaker testing in the technical specifications. Since that time TVA has agreed to include bypass breaker testing in the technical specifications to comply with Generic Letter 85-09.

Justification for Channe The auto shunt trip feature will upgrade the RPS to automatically trip the RTB in a similar manner that is accomplished by the UV output circuit.

The UV output circuit is designed so that in the absence of a reactor trip signal from the universal logic boards or switched inputs, Dariington Pair Transistors in the UV output circuit will conduct current and energize the UV

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i trip coil of the associated RTB. A reactor trip input will result in the i turning off of the pair transistors, thereby interrupting the current flow to

, the UV trip coil of the associated RTB and causing the RTS to open.

1 Similarly, with the addition of the automatic shunt trip feature, a reactor trip input will close the contacts on the shunt relay in the shunt trip coil ,

circuit. This energizes the shunt trip coil and trips the same RTB by a diverse mechanism.

The RTB surveillance test will independently verify the operability of the 4

shunt and UV trip features of the RTBs as part of a single sequential test procedure. Therefore, the surveillance test which identifies a failure of one i diverse trip feature also confirms the operability of the other trip feature.

I As a consequence, there is a high degree of confidence that the operable trip

feature will be capable of initiating a reactor trip in the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

! Accordingly, an additional action statement will be included in the technical i specifications for the RTBs to permit continued plant operations for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> with one of the diverse trip features inoperable before further action needs to be taken. If unable to restore operability within the 48-hour time f

{' requirement, the breaker must be declared inoperable and the plant taken to hot standby within six hours. During this 48-hour time period, the breaker can only be bypassed for up to four hours for the purpose of performing maintenance. Another action statement will be added for Modes 3*, 4*, and 5*

j (asterisk means with the reactor trip system breakers in the closed position, the control rod drive system capable of rod withdrawal, and fuel in the reactor vessel) for functional units 20 and 21 of table 3.3.1. This action statement will require the reactor trip breakers to be opened under certain i inoperability conditions. This surveillance testing and additional action

, statements meet the request by NRC as proposed by item 4.3 of Generic Letter j 83-28.

l The surveillance testing for the automatic shunt trip attachment will take an additional 10 minutes above the 40 minutes it presently takes to complete the i

test. This 40-minute. time period does not allow for any delays that may be encountered during testing. Should the logic test fail to meet its acceptance l criteria, the stringent one-hour does not allow sufficient time to evaluate i the problem due to the complex electronics involved with the SSPS. There have been three potential reportable occurrences (PRO) written as a result of i exceeding the one-hour time limit. SQN is requesting this time interval for

surveillance testing of the RTBs be extended to two hours, provided the other i j breaker and associated channel is operable. Current Westinghouse Standard

} Technical Specifications (STS) allow two hours for this testing.

The generic letter also required periodic independent testing of the control room manual switch contacts during each refueling outage. The procedures for testing the breakers will be revised to require voltage measurements in the i

RTB cabinets to verify operability of the reactor trip switch contacts and l wiring to the circuit breakers. precautions will be taken to ensure that the

" block auto shunt trip" switch will be used to preclude sensing the application of power to the shunt trip coil via the automatic shunt trip i.

feature. Additionally, with the breaker in a tripped condition, voltage j

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will be measured across the combination of the shunt trip coil and series "a" auxiliary contact, due to the presence of the breaker closed position ~

status light located in parallel with the normally open manual reactor trip switch contacts. This indicating light will be removed to avoid ambiguous voltage measurements.

Finally, the generic letter required periodic testing of the bypass breakers. This is to verify bypass breaker operability before being placed in service. The required testing will be performed by moving the breaker to the test position and verifying operability of the shunt trip attachment via the local shunt trip switch at the breaker. A manual reactor trip will also verify operability of the shunt trip attachment at least once per 18 months. The undervoltage trip. attachment will also be tested at least once per 18 months both manually and automatically. This surveillance testing for the bypass breakers meets the request by NRC as proposed by itee 4.3 of Generic Letter 83-28.

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ENCLOSURE 3 PROPOSED TECHNICAL SPECIFICATION CHANGES SEQUOYAH NUCLEAR PLANT (TVA SQN TS 77)-

Determination of no significant hazards considerations for proposed changes for operability of the new shunt trip attachment and required testing.

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'O Significant Hazards Considerations

1. Is the probability of an occurrence or the consequences of an accident ,

previously evaluated in the safety analysis report significantly increased?

No. The RPS is fundamental to plant safety because all transient and accident analyses are predicted on the basis that the RPS operate to terminate the energy released by the fission process. If the reactor fails to shut down following a transient or accident, an ATWS will exist. This change is intended to further minimize the possibility that an ATWS will occur. The additional testing requirements will not preclude any present testing procedures. The RTBs and UV attachment will be subject to the same testing and perform as originally intended. The _

additional testing and maintenance times do not alter the manner in which protection is afforded.

2. Is the possibility for an accident of a new or different type than previously evaluated in the safety analysis report created?

No. The intent of the proposed change is to increase the reliability of the RPS. The auto shunt trip attachment will meet all electrical and physical separation requirements as defined by IEEE 279-1971 and will be seismically and environmentally qualified. Surveillance testing and maintenance activities are still performed in a manner to ensure equipment reliability. The bypass breakers will be physically tested before being put in service. The proposed changes do not result in a change in the manner in which the RPS provides plant protection; therefore, this change will not create the possibility of occurrence of a new or different type accident.

3. Is the margin of safety significantly reduced?

No. The proposed changes are expected to increase the overall margin of safety. The independent testing of the UV and shunt trip attachment ensures that each RTB can be automatically tripped by two diverse mechanisms. The reliability of manual trip from the control room is retained by the periodic testing of the switch contacts, and the bypass breakers are ensured to be operable before service. The proposed changes

do not significantly alter the manner in which safety limits, limiting i safety system setpoints, or limiting conditions for operation are determined.

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