ML20072U457

Proposed Tech Specs Requesting Withdrawal of 910122 Request to Delete wide-range Containment Pressure & Reactor Vessel Level Instrumentation from Table 3.3-10 & Original Proposed Change to Bases to Tech Spec 3/4.3.3.7

ML20072U457
Person / Time
Site:	Sequoyah
Issue date:	04/12/1991
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20072U435	List: ML20072U431 ML20072U457
References
NUDOCS 9104190147
Download: ML20072U457 (37)

Text

- _ _ _ _ - _ _ _ _

ENCLOSURE 1 PROPOSED TECIINICAL SPECIFICATION CilANGE SEQUOYAll NUCLEAR P1 ANT UNITS 1 AND 2 ,

DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-89-30 Rev. 1)

LIST OF AFFECTED PAGES Unit 1 3/4 3-56 3/4 3-56s 3/4 3-57 3/4 3-57a 3/4 6-19 B 3/4 3-3 B 3/4 6-3 6-17 Unit 2 3/4 3-57 3/4 3-57a 3/4 3-58 3/4 3-56a 3/4 6-19 5 3/4 3-3 B 3/4 6-3 6-16

'$,[h41h }

p lj 7 P

~

TABLE 3.3-10 .,

N o ACCIDENT MONITORING INSTRUMENTATION E MINIPUf 5

I TOTAL NO. CHANNELS OF CHANNELS REQUIRED ACTION INSTRUMENT - .

-q c 4(1/RCS Loop) 1

1. Reactor Coolant THot (Wide Range) 4(1/RCSLe g

g (Instrument Loops68-001,-024,-043,-065) 4(1/RCS Loop) y(1/RCSLoop) 1

2. Reactor Coolant TCold (Wide Range)

(Instrument Loops68-018,-041,-060,-083) 2 2 1

3. Containment Pressure (Wide Range)

(Instrument Loops30-310,-311) 2 2 1

4. Containment Pressure (Narrow Range)

(Instrument Loops30-044,-045) 2 1

5. Refueling Water Storage Tank Level 2 R

• (Instrument Loops63-050,-051) 3 3 2 l

6. Reactor Coolant Pressure (Wide Range) e (Instrument Loops68-062,-066,-069) l l

3 2

7. Pressurizer Level (Wide Range) 3 ( )

(Instrument Loops68-320,-335,-339) 2/ steam line 1 2/ steam line

8. Steam Line Pressure (Instrument Loops 1-002A,-002B,-009A,-0098,

-020A,-0208,-027A,-0278)

I

9. Steam Generatoe Level - (Wide Range) ef(1/steamgenerator hstea.agenerator) g (Instrument Loops 3-043,-056,-098,-111) w 2/ steam generator 2/ steam generator 1

10. Steam Gene-ator Level - (Narrow Range) h g (Instrumert Loops 3-039,-042,-052,-055, r -094,-097,-107,-110) 5 11. Auxiliary Feedwater 5

C3 1/ steam generator 1/ steam generator mg a. Flow Ra'e (Instrvrent Loops 3-163,-155,-147,-170)

C7 - 5 3/ steam generator 3/ steam generator .

C h b. Valve Pesition Indication

~1 m I (Instrument Loops 3-164,-164A,-172,-156,

-156A,-173,-148,-148A,-174,-171,-171A,-175) n O

O

')

y .: .q .. . .,

TABLE 3.3-10'(Continued)- ,

E g- . ACCIDENT MONITORING INSTRUMENTATION f, y.

! z- MINIMUM

=

TOTAL NO. CHANNELS g INSTRUMENT. OF CHANNELS REQUIRED ACTION Q

y ' -12. Reactor Coolant System Subcooling Margin

. Monitor (Instrument Loops94-101,-102) 2 2 #j l

13. Containment Water Level (Wide Range) 2 2 1 (Ir.strume t Loops63-173,-179) y ,g j l Inp A '

14. In Core Thermocouples -

65 1/ core quadrant / train 3 i

15. Reactor Vessel . Level Instrumentation 2 2 1 2- ' System (Instrument Loops68-367,-368, gg i -369,-370,-371,-372) 4- w A

f 16. Containment Area. Radiation Monitors

. .w j E a.' Upper Compartment' 2 1 4 i E (Instrument Loops90-271,-272)

{

~

b. Lower Compartment- , 2 1 4 (Instrument Loops90-273,-274) j' 17. Neutron Flux L a

j. a. Source Range . 2 2 1

! (Instrument Loops 92-5001,-5002) i @ b. Intermediate Range 2 2 ,

1

$ (Instrument Loops 92-5003,-5004) i-- E

g r

see LSeA 3 1

'c 3 g

.m-OU c) " # Source Range outputs may be disabled above the P-6 (Block of $ource Range Reactor Trip) setpoint.

. Rl*!

--a t

co

.c e

- _ . . . = . .+. - . , . .a -- . ~ __

u

~

r i-

{.

.L - INSERTA 1

MINIMUM TOTAL NO. CHANNELS INSTRUMENT OF CHANNELS REQUIPED ACTION

14. Incore Thermocouples 65

a. Core Quadrant (1) 2(1/ Train) I

, b. Core Quadrant (2) 2(1/ Train) I

c. Core. Quadrant (3) 2(1/ Train) 1

< d. Core Quadrant (4) 2(1/! rain) 1

15. Reactor Vessel Level Instrumentation 6 a Dynamic Range 2 , 1 (Instrument Loops68-367. 370)

b. Upper Range 2 1 (Instrument Loops68-368, 371) -

c. Lower Range 2 1

! (Instrument Loops68-369, 372) l l

l

\

= p. r M , - - . - - _ -

INSERT B MINIMUM TOTAL NO. CHANNELS REQUIRED ACTION OF CHANNELS INSTRUMENT

18. ERCW to AFW Valve Position 1

1/ TRAIN / PUMP 1/ TRAIN / PUMP a) Motor Driven Pumps (2 VALVES / TRAIN)

(Instrument Loops 3-116A, (2 VALVES / TRAIN)

-116B, -126A, -126B) 1 2 TRAINS 1 b) Turbine Driven Punp 2 TRAINS (2 VALVES / TRAIN) (2 VALVES / TRAIN)

(Instrument Loops 3-136A,

-136B, -179A, -179B) 3 1/ VALVE 1/ VALVE

19. Containment Isolation Valve Position (Panels TR-A XX-55-6K

& TR-B XX-55-6L)

1. 1. Not required for isolation valves that are closed and deactivated. .

l

TABLE 3.3-10 (Continued)

ACT 0 STATEMENTS LCO 3.3.3.5 ACTION 1 - NOTE: Also refer to eapplicayeactionrequirementsfrom Tables 3.3-1s 3.3-3, and 9-li-9-since they may contain more restrictive actions.  ;

a. With the number of channel one less than the minimum channels required, restor the inoperable channel to OPERABLE status within f' ys o be in H0 0"JTPO# within the next,1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />,g klaa cmdinHor5Hu.70%w truoar [gg,g Me nea4 sa kour,3,

b. Wi number of channels two less ~tlian thE iiiin a channels required, restore at least one inoperable chann n' to OPERABLE status withi " h: ggt GNT90WN .57ANDB(J within the ne 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> go7 sHu7Nu> w%%

c. The provisions of Specification 3.0.4 are no I g ,-

ACTION 2 - NOTE: Also refer to the_ applicable action requirements from Tables 3.3-1 since it may contain more restrictive actions. p

a. With the number of channels one less than the minimum channels required, restore the inoperable channel to OPERABLE status within days or be in t ast " ^

4 k H OT S d O T D ou)M 1-hCd HOT -6Httf00W within the neh,12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> QIANDBh WWW ne n ed /A heut6.

b.. With the n~ umber of channels two less t an-the m~inimum channels required, restore at least one inoperable channel to OPERABLE status within 7 days orgc'e in 3 Wet FQ

-4WT00WNtwithin the next f hoursf - M N H o7 A H uT DowAl '

LSNMU G ksN Ah ,, e reAi 8A 6 cur.s '

c . -- With the number of channels three less than t e m liitTm- ~

, channels-required, restore one channel-to OPERABLE status -

H within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in_at_ east HOT 4WT90W within the  ;

hours Wis Hof 3//fa7 . JfypBO  !

o * % f%e ne A t 1.2. h eu 3 ..

d. The provisions o p'ecTTT8tTon 3. . are not applicable.

rep lue W + h I' V'3 # A A

• M ' ^ '3

-ACTION 3 - D#dek r-- mmh the nc.-tcr ef-channets-4tss-than-the-mWaum chann(45-l req"ir:d, renoce-the inoperab4ewhannel(:)-to-OPEftABLE statu: with' @ our: Or be in- et lee i HOT-SWT00WN-l -within the next---12 hcur:.--

( l A -The-peev4s4 ens-of-Spoo444 cation-3-k4 are-not-eppiicbla - 1 l-I 3/4 3-57 Amendment No. 46,149 SEQUOYAH - UNIT 1 DEC 0 'l 1980

1 l

I ACTION 3 - NOTE: Also refer to the applicable action requirements from LCO 3.6.3 since it may contain more restrictivo actions.

4

1. 1. 1. a. With the accident monitoring indication for one of the penotration inboard or outboard valvo(s) inoperabic, restore the inoperable valve (s) accident indication to OPERABLE status within 30 days, or isolato each affected penetration within 30 days by ,

use of at least ono deactivated automatic valvo secured in the isolated position, or isolato each affected penetration within 30 days by use of at least one closed manual valvo or blind flango, or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

1. 1. 1. b. With the accident monitoring indication for both an inboard and-outboard valvo(s) on the same penetration inoperable, restore at least the inboard or outboard inoperable valvo(s) indication to OPERABLE status within 7 days, ce isolate cach affected penetration within 7 days by uso of*at least one deactivated automatic valvo secured in the isolated positlon, or isolato each affected penotration within 7 days by use of at least one closed manual valvo or blind flange, or be in at least HOT SHUTDOWN within the i next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c. The provisions of Specification 3.0.4 are not applicable.

l

-### On a penetration where accident indication is declared INOPERABLE on a valvo but on the opposito sido of the penetration an <

accident indication valvo does not exist (such as with a cloa'*

system or a check valve), only ACTION 3(a) must be entered.

However, valves FCV-63-158 & -172 are both inboard ponotration valves,-but if both valves have inoperable accident indication, ACTION 3(b must be ontored until at least one of the valvo's accident in)dication is restored to OPERABLE status. Valvos FCV-30-46 & VLV-30-571, FCV-30-47 & VLV-30-572, and FCV-30-48 &

VLV-30-573 are all outboard penetration valves, but if both valves have inoperabic accident indication, ACTION 3(b) must be entered until at least one of the valvo's accident indication is restored to OPERABLE status, i

~

TABLE 3.3-10 (Continued)

ACTION STATEMENTS (Continued)

ACTION 4 - a. With the number of channels less than the minimum channels required, initiate an alternate method of anonitoring containment area radiation within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and either i restore the inoperable channel (s) to OPERABLE status o 80 withiii)J' days, or prepare and submit a special report to the Commission pursuant to Specification 6.9.2.1 within 4ke next 14 days that provides actions taken, cause of the ,

inoperability, and plans and schedule for restoring the i channels to OPERABLE status. .

b. The provisions of Specification 3.0.4 are not applicable.

ACTION 5 ' NOTE: Also refer to the applicable action requirements from

-Tetrlw 3. 3 Jn contain more restrictive actions, gLc.0Js.3.]

a. With the number of channels on one or more/ steam generators less than the minimum channels / required for either flow rate or valve position. 1 channeltoOPERABLEstatuswithin?pestoretheinoperable dap or be n t , leas HOT -SHUTOOWN within the nc Jf'hoursfon M Ho (4 jus ne oui . JR shut DDiad GSTMDEd

~

_2 Aew With the' number of channe s on one or more sGim~

it b. Ltd generators less than the minimum channels required for..

either flow rate or valve position, restore thejinoperabl channel (s) to OPERABLE status within 46'.hou eMtp -

least HOT 4 HUT within the n t fh u in No7 3Mu7Dow' 3TANDB n'.a 4Ae nerf /.t kom ,

c. The provis ons of Specification .0.4 are no app EP '

- 4

' ACT! 6- a. With the number of channels less than the mini channels required restore the inoperable hnel to OPERABLE status wlthin 7 days or increas y one the minimum shift crew per Table 6.2-1. Jh additional shift rwmembershallbededicatedtojndcapableof de(teTm ning the subcooling marg n during an accident using Delete.

( existing rumentation.

1

b. With the number els two less than the minimum channels required, jr re at least one inoperable channel r to OPERABLE statgVwithi 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or increase by one the <

l minimum shift g ew per Tab'e 1. The additional shift I

crew memberf(hall be dedicated to nd capable of determipffg the subcooling margin durhg.an accident using exist ffg instrumentation,

c. e provisions of Specification 3.0.4 are not appl cahg.

(

SEQUOYAH - UNIT 1 3/4 3-57a Amendment No. 112, 149 DEC 0 71090

y _

'il .

• if TABLE 3.6-2 . .

M T CONTAIDMENT ISOLATION VALVES SE -

E

• ' VALVE NUMBER FUNCTION- MAXIMUM ISOLATION TIME (Secoads)

E y A. PHASE "A" ISOLATION

1. FCV-1-7 SG Blow Dn 10*

2. FCV-1-14 SG Blow Dn 10*

3.. FCV-1-25 SG Blow Dn 10* R41

4. FCV-1-32 'SG Blow Dn 10*

5. FCV-1-101 50 31= h 15*

G. FCV-1-102 SC 51= h 15*

7. FCV + i&5 SG Bio. On 15*

0. FCV-1-100 SC Bl= 5 15*

9. FCV-26-240 Fire Protection isol. 20

10. FCV-26-243 Fire Protection isol. 20

, 11. FSV-30-134 Cntat Bldg Press Trans 4*

g Sense Line

, 12. FSV-30-135 Cntat Bldg Press Trans 4*

• Sense Line 5 13. FCV-31C-222 CW-Inst Room Cirs

~

10* R74

14. FCV-31C-223 CW-Inst Room Clrs 10*

15. FCV-31C-224 CW-Inst Room Clrs 10*

16. FCV-31C-225 CW-Inst Room Cirs 10*

17. FCV-31C-229 CW-Inst Room Cirs 10*

18. FCV-31C-230 CW-Inst Room Cirs 10*

og 19. FCV-31C-231 CW-Inst Room Cirs 10*

Rg 20. FCV-31C-232 CW-Inst Room Clrs '10*

gg 21. FSV-43-2 Sample Przr Steam Space 10*

3g 22. FSV-43-3 Sample Przr Steam Space 10* -

o -4 23. FSV-43-11 Sample Przr Liquid 10*

C g 24. FSV-43-12 Sample Przr Liquid 10* <149

25. FSV-43-22 Sample RC Outlet Hdrs 10*

, 26. FSV-43-23 Sample RC Outlet Hdrs 10*

w 27. FSV-43-34 Accum Saaple 5*

(o 28. FSV-43-35 Accum Sample- 5*

29. FSV-43-55 SG Blow Dn Sample Line 10*

M LD

.n.

INSTRUMENTATION BASES design basis-for the facility to determine if plant shutdown is required pursuant to Appendix "A" of 10 CFR Part 100. All specified measurement ranges represent the minimum ranges of the instruments. This instrumentation is R85 consistent with the recommendations of Regulatory Guide 1.12. " Instrumentation for Earthquakes," April 1974.

3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23, "Onsite Meteorological Programs," February 1972.

3/4.3.3.5 REMOTE SHUT 00WN INSTRUMENTATION The OPERABILITY of the remote shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT STANDBY of the facility and the potential capability for subsequent cold shut- BR s;m down from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR 50.

3/4.3.3.6 CHLORINE DETECTION SYSTEMS R66 This specification deleted.

4 3/4.3.3.7 ACCIDENT HONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, Revision 2, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During ggc and Following an Accident," December 1980.

For Sequoyah, the redundant channel capability for Auxiliary Feedwater (AFW) flow consists of a single AFW flow channel for each Steam Generator with the second channel consisting of three AFW valve position indicators (twe 'evel control valves for the motor driven AFW flowpath and one level control valve for the turbine driven AFV flowpath) for each steam generator. Two containment hydrogen monitoring channels are designated as accidant monitoring instrumenta-tion (Type A, Category 1) in accordance with Regulatory Guide 1.97. Operability

, and Surveillance Requirements for the purpose of accident monitoring is governed 4 by Specificatio 4.1 for containment hydrogen monitors.

Revised 08/18/87 B 3/4 3-3 Amendment No. 62, 81, 149 SEQUOYAH - UNIT 1 DEC 07 30

INSERIC The postaccident monitering instrumentation limiting condition for operation provides the requirement of Type A and Category 1 monitors that provide information required by the control room operators to

• Permit the opera: to take preplanned manual actiens'to accomplish safe plant shutdown.

• Determine whether systems important to safety are performin6 their intended functions.

• Provide information to the operators that will enable them to determine the likelihood of a gross breach of the barriers to radioactivity release and to determine if a gross breach of a barrier has occurred.

CONTAINMENT SYSTEMS BASES 3/a.6.1.8 EMERGENCY GAS TREATMENT SYSTEM (EGTS)

The OPERABILITY of the EGTS cleanup subsystem ensures that'during LOCA conditions, containment vessel leakage into the annulus will be filtered through the HEPA filters and charcoal adsorber trains prior to discharge to g the atmosphere. This requirement is necessary to meet the assumptions 'used in the accident analyses and limit the site boundary radiation doses to within the limits of 10 CFR 100 during LOCA conditions. Cumulative operation of the system with the heaters on for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> over a 31 day period is sufficient to reduce the buildup of moisture on the absorbers and HEPA filters. ANSI N510-1975 Ril8 will be used as a procedural guide for surveillance testing.

3/4.6.1.9 CONTAINMENT VENTILATION SYSTEM Use of the containment purge lines is restricted to only one pair (one supply line and one exhaust line) of purge system lines at a time to ensure that the site boundary dose guidelines of 10 CFR Part 100 would not be exceeded in the event of a loss of coolant accident during purging operations. The analysis of this accident assumed purging through the largest pair of lines (a 24 inch inlet line end a 24 inch outlet line), a pre-existing iodine spite in -

the reactor coolant and four second valve closure times.

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of the containment spray system ensures that containment depressurization and cooling capability will be available in the event of a LOCA. The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the accident analyses.

' 3/4.6.2.2 CONTAINMENT COOLING FANS The OPERABILITY of the lower containment vent coolers ensures that ade-quate heat removal capacity is available to provide long-term cooling following a non-LOCA event. Postaccident use of these coolers ensures containment tem-peratures remain within environmental qualification limits for all safety

• R71 related equipment required to remain functional.

3/4.6.3 CONTAINMENT ISOLATION VALVES Qeplace M .Tn Sed D The OP dof tWo'ntain tisolati a ves ensur te ated fjr vtle outside p co jntin mojseme1Till b

• me the tvent of a y.elerse of r ctive ma rifal to the cjo t6nment .a phere or pre jss th &' tion of containmeJt Containment 1rolationwJt the tim fe HTnitsspecjf4 ensures Jhat'the relea jse radioactiv34aterial to Jh6 envirpnmeht will beJonfistent with the assumptiongsed in the a3ahses for a LOCA.

report oBythF letyt operating r (datedexMarch science f3M1, and of the Apryi pJafft no2,1981, later t TV(Astartup wit 1 submit aft (r 4 R8

( t f refueling ht information M1 be used to p de a bas 6 to

/

r-evaluatethepeuacyofthepurse- nd vent timed mits.

1 SEQUOYAH - UNIT 1 8 3/4 6-3 Amendment No. 67, 114 May 5, 1989

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

4 INSERID The valves identified in Table 3.6-2 are containment isolation valves as defined per 10 CFR 50. The operability of these containment isolation valves ensurua that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive mater.ial to the containment atmosphere or pressurization of the containment. Containment )

isolation within the time limits specified ensures that the releaso of 1 radioactive material to the environcent will be consistent with the assumptions used in the analyses for a loss of coolant accident.

Additional valves have been identified as barrier valves, which in addition to the-containment. isolation valves discussed above, are a part of the accident monitoring instrumentation in Technical Specification 3/4.3.3.7 and are designated as Category 1 in accordance with Regulatory Guide 1.97, Revision 2

" Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant ,

Conditions During and Following an Accident." December 1980. l 9e.--m,,---we'r' -

we "- w *+v------ 'wm--- - * *s- - - - - m'-*"-9M('- "--

-8'%we 'l-T-T* tt""*---T

i . j l

ADMINISTRATIVE CONTROLS ,

wyyTVV y y T vYve

d. -Beekup-Hethod for=0etermining- Sut,c00 liny "aiyin DEL E 7ED f _

j

-A program-which will ensure the-espatrH+ty-to eccurate+y-conitor-the~  !

-Reactor-Coohmt-System .,um vui,,,v 1mrgin. Thi >,,vv. m ehell inc1ude- )

1- -the-foHow+ngt-- .

l (i) iteining-Of personnel, ond- [

iii)- I'recedstres for : aitcringe ,

i

e. Postaccident Sampi ng A program which will ensure the capability to obtain and analyze 3

reactor-coolant, radioactive iodines and particulates in plant 1 gaseous effluents, and containment atmosphere samples under accident R16 <

conditions. The program shall include the following:

(i) Training of personnel, (ii) Procedures for sampling and analysis, (iii) Provisions for maintenance of sampling and analysis equipment. i

f. Radioactive Effluent Controls Program j A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to 6., MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably

• • achievable. The program (1) shall be contained in the 00CM, (2) shall i

be implemented by operating procedures, and (3) shall include remedial 3

, actions to be taken whenever the program limits are exceeded. The l program shall include the following elements:

1) Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and set- ,

point determination in accordance with the methodology in the l 00CM,

2) Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS conforming to 10 CFR Part 20, Appendix B, Table II, Column 2,

3) Mcnitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.106 and with the methodology and parameters in the 00CM,

4) Limitations on the annual and quarterly doses or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid l- effluents released from each unit to UNRESTRICTED AREAS conforming to Appendix I to 10 CFR Part 50,

5) - Determination of cumulative and projected dose contributions from-radioactive ef fluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the 00CM at least every 31 days, R7 SEQUOYAH - UNIT 1 6-17 Amendment Nos. 12, 32, 58, 74, 148 NOV 1 ^ O

^

_ g .

EF - -

E TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION MINIMUM h

x TOTAL NO. CHANNELS OF CHANNELS _ REQUIRED _ ACTION

• INSTRUMENT 4(1/RCS Loop) [ 1 E 1.' Reactor Coolant THot (Wide Range) m (Instrument Loops68-001,-024,-043,-065) ___ _

4(1/RCS Loop) 1 RCS Loop 1

2. Reactor Coolant TCold (Wide Range) ~

(Instrument Loops68-018.-041,-060,-083) 2 2 1

3. Containment Pressure (Wide Range) R13!

(Instrument Loops30-310,-311) 2 2 1

4. Containment Pressure (Harrow Range)

(Instrument Loops30-044,-045) 2 2 1 y 5. Refueling Water Storage Tank Level

• (Instrument Loops63-050,-051) 3 3 2 T 6. Reactor Coolant Pressure (Wide Range) 5 (Instrument Loops68-062,-066,-069) 3 3 2

7. Pressurizer Level (Wide Range) -

(Instrument Loops68-320,-335,-339) 2/ steam line 1

8. Steam Line Pressure 2/ steam line (Instrument Loops 1-002A,-002B,-009A,-0098, _-

-020 A,-020B ,-027A,-0278) _

(1/steamgenerator) I f g 9. Steam Generator Level - (Wide Range) 4(1/

j g (Instrument Loops 3-043,-056,-098,-111) 2/ steam generator 2/ steam generator 1 l @ 10. Steam Generator Level - (Narrow Range) -

l $ (Instrunnt Loops 3-039,-042,-052,-055,

-094,-097,-107,-110) l [

! O y 11. Auxiliary Feedwater 1/ steam generator 1/ steam generator 5

$g -

a. Flow Rate (Instrument Loops 3-163,-155,-147,-170) 3/ steam generator 3/ steam generator 5

~t y b. Vaive Position Indication i,- (Instrument Loops 3-164,-164A,-172,-156,

1. - 156 A ,- 173 ,- 148 ,- 148 A , - 174 ,- 171,- 171 A ,-175 )

e > 7; w

g TABLE 3.3-10 (Continued) ,

E g ACCIDENT MONITORING INSTRUMENTATION sI MINIMUM TOTAL NO. CHANNELS g INSTRUMENT OF CHANNELS REQUIRED ACTION Z

m 12. Reactor Coolant System Subcooling Margin 2 2 fi Monitor (Instrument Loops94-101,-102)

13. Containment Water Level (Wide Range) 2 2 1

()&ce,l

g (Instrument Loops63-178,-179). __ _

_ w _- _ __ __ - - - - - .

Jn9d S 14. In Core Thermocouples

_ _ 65 1/ core quadrant / train 3 u3

15. Reactor Vessel Level Instrumentation 2 2 1 l

System (Instrument Loops68-367,-368, _ ,/

-369 _370,-371,-372)_

' D 16. Containment Area Radiation Monitors w

h a. Upper Compartment 2 1 4 m (Instrument Loops90-271,-272)

b. Lower Compartment 2 1 4 (Instrument Loops90-273,-274)

17. Neutron Flux

a. Source Range 2 2 1 (Instrument Loops 92-5001,-5002)

[ b. Intermediate Range 2 2 ,

1

] g (Instrument Loops 92-5003,-5004) 3

um i (sa x-s i)

) lel ?

. # Source Range outputs may be disabled above the P-6 (Block of Source Range Reactor Trip) setpoint. "3 u, -

g. m ri 5

s INSERIA MINIMUM TOTAL NO. CHANNELS REQUIRED ACTION OF CHANNELS INSTRUMENT Incore Thermocouples 65 14.

2(1/ Train) I

a. Core Quadrant (1) 2(1/ Train) I

b. Core Quadrant (2) 2(1/ Train) 1

c. Core Quadrant (3) 2(1/ Train) 1

d. Core Quadrant (4) 6

15. Reactor Vessel Level Instrumentation 2 1

a. Dynamic Range (Instrument Loops68-367, 370) 2 1

b. Upper Range (Instrument Loops68-368, 371) 2 1

c. Lower Range (Instrument Loops68-369, 372)

i l INSERT B i

I 1

MINIMUM i

TOTAL NO. CHANNELS j- INSTRUMENT OF CHANNELS REQUIRED ACTION

18. ERCW to AFW valve Position 1

a)' Motor Driven Pumps 1/ TRAIN / PUMP 1/ TRAIN / PUMP 1 (Instrument Loops 3-116A, (2 VALVES / TRAIN) (2 VALVES / TRAIN)

-116B,f-126A, -126B) b) Turbine Driven Pump 2 TRAINS 2 TRAINS 1 (Instrument Loops 3-136A, (2 VALVES / TRAIN)' (2 VALVES / TRAIN)

-136B, -179A, -179B)

1. f

19. Containment Isolation 1/ VALVE . 1/ VALVE 3 Valve Position (Panels.TR-A XX-55-6K i & TR-B XX-55-6L) i E

4 1

1. 1. Not required for' isolation valves that'are closed and deactivated. ,

__.._m.. .

,..__.%. .,.uu m- - -- m_--- - - . _ _ - _ . _ _---m-s--.- . - -

TABLE 3.3-10 (Continued)

ACTION STATEMENTS an YO 3h ACTION 1 - NOTE: Also refer to,the applicable action requirements from Tables 3.3-1,13.3-3, and @ 4 since they may contain more restrictive actions,

a. Withthenumberofchannels)onelessthantheminimum channels required, restorg)the inoperable channel to 0PERABLE status within r . days or be in at least within the next ,1 oursy POT T tr [boWU

b. With the number of channels two less than the minimum channels required, restore at least one inoperable channe g' to OPERABLE status within '^ W e or be in HOT SHUT DU 00 M,1 " N within the next hoursw i 1' s

c. The provisions of Specification 3.0.4 are not applicabli.

ACTION 2 - NOTE: Also refer to the applicable action requirements from Tables 3.3-1 since it may contain more restrictive actions. '

3g

a. With the number of channels one less than the minimum channels required, restore the inoperable channel to OPERABLE status within -31 days or be in at least - - -

D. far.d. A" 6 urow M

'the ner+ I2 hors HOTfM00W_DM k.sthTDB within the next 6 ,1r hoursy%tL;n ~

b. With the number of channels two less than the minimum channels required, restore at least one inoperable channel to OPERABLE status within 7 days or be ~in at_le_ast H0_T _._

ME0aWM within the next .lf hoursy en 7n~nT hurbowiJ m_DBD & m % :n 4lq n o,Q I2 t,ours

~

, c. With the number of channels three les's than the minimum channels required, restore one channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> _or be in at _least_H_0T SHUT 00WM within the YLft_ _? u

d. The provisions of Specification 3.0.4 are not applicable, mlc+e ~ Replue a l %, revh eb AcH h 3 ACTION 3 - a. Wi th-the-n umbe r-o f-c ha nnel s-l e s s-tha n-the-mi ni mum-cha nne l s-

-req ui red r-re s to re-the-i nope ra bl e-c ha nnel (s }-to4 P ERAB LE-

-s t a t us-w i t hi n-40-hou rs-o r-be -4n-e t-l e a st-HOT-6WT00WN -

-wMM n-the-next-12-houest-

-b. T he-p rov 4s4 ens-of-Spec 444sa t4on-brGr4-a r-e-noupp44ca ble r i

\-

3/4 3-58 Amendment No. 38,135 SEQUOYAH - UNIT 2 Oh.O b -

ACTION 3 - NOTE: Also refer to the applicable action requirements from LCO 3.6.3 since it may contain more restrictive actions.

1. 1. 1. a. With the accident monitoring indication for one of the penetration inboard or outboard valve (s) inoperable, restore the inoperable valve (s) accident indication to OPERABLE status within 30 days, or isolate cach affected penetration within 30 days by use of at least one deactivated automatic valvo secured in the isolated position, or isolate cach affected penetration within 30 days by use of at least one closed manual valvo or blind flange, or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

1. 1. 1. b. With the accident monitoring indication for both an inboard and outboard valvo(s) on the same penetration inoperable, restore at least the inboard or outboard inoperablo valve (s) indication to OPERABLE status within 7 days, or isolate cach affected penetration within 7 days by use of.at least one deactivated automatic valve secured in the isolated position, or isolato each affected ponotration within 7 days by use of at least one closed manual valvo or blind flange, or be in at least HOT SHUTDOWN witnin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />,

c. The provisions of Specification 3.0.4 are not applicable.

1. 1. 1. On a penetration where accident indication is declared INOPERABLE on a valve but on the opposito side of the penetration an accident indication valve does not exist (such as with a closed system or a check valvo), only ACTION 3(a) must be entered.

However, valves FCV-63-158 & ~172 are both inboard penetration valves, but if both valves have inoperable accident indication, must be enterod'until at least one of the valvo's ACTION accident 3in (b) dication is restored to OPERABLE status. Valves FCV-30-4 6 & VIN-3 0-571, FCV-30-47 & VLV-30-572, and FCV-30-48 &

VLV-30-573 are all outboard penetration valves, but if both valves have inoperable accident indication, ACTION 3(b) must be entered until at least one of the valvo's accident indication is restored to OPERABLE status.

TABLE 3.3-10 (Continued)

ACTION STATEMENTS

-(Continued)

ACTION 4 - a. With the number of channels less than the minimum channels required, initiate an alternate method of monitoring containment area radiation within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and either M restore the inoperable channel (s) to OPERABLE status 30 '

days, or prepare and submit a special report t W wiUiTii/'

the Comm ission pursuant to Specification 6.9.2.1 within & iWt 14 days that provides actions taken, cause of the inoperability, and plans and schedule for restoring the channtis to OPERABLE, status,

b. The provisions of Specification 3.0.4 are not applicable.

ACTION 5 - NOTE: Also refer to the applicable action requirements from

-Tactions.

abie-3QQtic@ g contain more restrictive O R135

a. With the number of channels on one or more steam generators less than the minimum channels required for either flow rate or valve position,frestore the inoperable channel to OPERABLE status withind'd ,.orJL inJL JepjLt f HOT S WTOOWN within the next g ho ed in NOT (J/WrDod ~

Qr4{D fp  ; h the re d a b r.s b.

Withthenumberofchannels(g) on one of mo E sfelim hW generators less than the minimum channels required for either flow rate or valve position, restore thf/InoperMM3 71 channel (s) to OPERABLE status within-48-hou W or be in at least H0 -SHIJT00WM within the ne ,1f hou JMd in HOT .5Nu7pA 4 S TM)D D Y co w i% +Ae A eX r a k sur,s ,

.0.4 ar 'not appIIc5El E " " '

~

c. The provis ons of Specification ACT'IDtk -

a. With the number of channels less than the minim #

channelsreQuired,restoretheinoperablecpannelto OPERABLE status within 7 days or increasjpby one the inimum shift crew per Table 6.2-1. Jhe additional shift crewqember shall be dedicated to And capable of hW determink1g the subcooling mar W during an accident using existing instrumentation.

b. With the number of a is two less than the minimum channels required st least one inoperable channel to OPERABLE stat s within 4 rs or increase by one the minimum shift' crew per Table 6. he additional shift crew member shall be dedicated to an -apable of determifning the subcooling margin during'linsaccident using extiiting instrumentation.

he provisions of Specification 3.0.4 are not applicab e g M

3/4 3-58a Amendment No. 102, 135 SEQUOYAH - UNIT 2 0h.C h 7 . )

b .

( il s' ~

' ~

~

TABLE 3.6-2 -

M E CONTAINMENT ISOLATION VALVES l S MAXIMUM ISOLATION TIME (Seconds)

E VALVE NLHBER FUNCTION

( .

A. PHASE "A" ISOLATION f g SG Blow Dn 10*

[ 1. FCV-1-7 10*

2. FCV-1-14 SG Blow Dn R29 SG Blow Dn 10*

3. FCV-1-25 10*

4. FCV-1-32 SG Blow Dn SC S?cw On 15*

5. FCV-1-181 15*

G. FCV 1 1G2 50 01cw Ca SC S?cw On IS*

7. FCV 1'100 15*

0. FCV-1-104 50 O!cw On Fire Protection Isol. 20

9. FCV-26-240 20

10. FCV-26-243 Fire Protection Isol.

Cntmt Bldg Press Trans 4*

4 *

11. FCV-30-134 Sense Line Cntat Bldg Press Trans 4* R62

[ 12. FCV-30-135 w Sense Line CW-Inst Room Cirs 10*

13. FCV-31C-222 10*

14. FCV-31C-223 CW-Inst Room Clrs CW-Inst Room Cirs 10*

15. FCV-3IC-224 10*

16. FCV-31C-225 CW-Inst Room Clrs CW-Inst Room Cirs 10*

17 FCV-31C-229 10*

18. FCV-31C-230 CW-Inst Room Clrs CV-Inst Room Cirs 10*

19. FCV-31C-231 10*

20. FCV-31C-232 CW-Inst Room Clrs lR136 Sample Przr Steam Space 10*

E 21. FSV-43-2 Sample Przr Steam Space 10*

3 22. FCV-43-3 10" lR136 Sample Przr Liquid E 23. FSV-43-11 10*

3 24. FCV-43-17 Sample Przr Liquid

" Sa-:ple RC Outlet Hdrs 10* lR136 O 25. FSV-43-22 10*

m E 26. FCV-43-23 Saeple RC Outlet Hdrs Accum Sample 5* lR136 O ~ 27. FSV-43-34 Accum Sample 5*

'" @ 28. FCV-43-35 SG Blow Dn Sample Line 10* r

' ')

29. FSV-43-55 10* )R136 SG Blow Dn Sample Line

-j y 30. FSV-43-58 ~

0C m

~

1 .

' INSTRUMENTATION l

BASES 3/4.3.3.3 SEISMIC INSTRUMENTATION (Continued) design basis for the facility to determine if plant shutdown is required pursuant to Appendix "A" of 10 CFR Part 100. All specified measurement ranges represent the minimum ranges of the instruments. The iniitrumentation is R72 consistent with the recommendations of Regulatory Guide '1.12. " Instrumentation for Earthquakes," April 1974.

3_/4.3.3.4 METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the ublic and is consistent with the recommendations of Regulatory Guide 1.23, p'Onsite Meteorological Programs," February 1972.

3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION The OPERABILITY of the remote shutdown instrumentation ensures that suf-ficient capability is available to permit shutdown and maintenance of HOT (fg STANDBY of the facility and the potential capability for subsequent cold shut-down from locations outside of the control room. This capability is required gg in the event control room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR 50.

3/4.3.3.6 CHLORINE DETECTION SYSTEMS g This specification deleted.

3/4.3.3.7 ACCIDENT HONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with the reccmmendations of Regulatory Guide 1.97, Revision 2, " Instrumentation for light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1980.

fnsed C For Sequoyah, the redundant channel capability for Auxiliary Feedwater (AFW) flow consists of a single AFW ';ow channel for each Steam Generator with R13 the second channel consisting of three AFW valve position indicators (two level control valves for the motor driven AFW flowpath and one level u>ntrol valve for the turbine drive AFW flowpath) for each steam generator. Two containment hydrogen monitoring channels are designated as accident monitoring instrumenta-tion (Type A, Category 1) in accordance with Regulatory Guide 1.97. Operabil-ity and Surveillance Requ ements for the purpose of accident monitoring is governed by Specification 6.4.1 for containment hydrogen monitors, Revised 08/18/87 SEQUOYAH - UNIT 2 8 3/4 3-3 Amendment Hos. 35,46,54,72,135 ObO . )

, INSERT 0 The postaccident monitoring instrumentation limiting condition for operation provides the requirement of Type A and Category 1 monitors that provide information required by the control room operators to

• Fermit the operator to take preplanned manual actions to accomplish safe plant shutdown.

• Determine whether systems important to safety are performing their intended functions.

• Provide information to the operators that will enable them to determine the likelihood of a gross breach of the barriers to radioactivity release and to determine if a gross breach of a barrier has occurred.

e 1 .

.. s

. CONTAINMENT SYSTEMS

( BASES 3/4.6.1.8 EMERGENCY GAS TREATMENT SYSTEM (EGTS)

The OPERABILITY of the EGTS cleanup subsystem ensures that during LOCA conditions, containment vessel leakage into the annulus will be filtered through the HEPA filters and cnarcoal adsorber trains prior to discharge to the atmosphere. This requirement is necessary to meet the assumptions,used in the accident analyses and limit the sitt boundary radiation doses to within the limits of 10 CFR 100 during LOCA conditions. Cumulative operation of the system with the heaters on for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> over a 31 day period is sufficient to reduce the buildup of moisture on the absorbers and HEPA filters. ANSI H510-1975 will be'used as a procedural guide for surveillance testing.

3/4.6.1.9 CONTAINMENT VENTILATION SYSTEM Use of the containment purge lines is restricted to only one pair (one supply line and one exhaust line) of purge system lines at a time to ensure that the site boundary dose guidelines of 10 CFR Part 100 would not be exceeded in the event of a loss of coolant accident during purging operations. The analysis of this accident assumed purging tnrough the largest pair of lines (a 24 inch inlet line and a 24 inch outlet line), a pre-existing iodine spike in the reactor coolant and four second valve closure times, 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS ggggp 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of the containment spray system ensures that containment depressurization and cooling capability will be available in the event of a LOCA. The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the accident analyses.

3 /4. 6.' 2. 2 CONTAINMENT COOLING FANS o

The OPERABILITY of the lower containment vent coolers ensures that ade-quate heat removal capacity is available to provide long-term cooling following R$-

a non-LOCA event. Postaccident use of these coolers ensures containment tem.

peratures remain within environmental qualification limits for all safety-related equipment required to remain functional.

3/4.6.3 CONTAINMENT ISOLATION VALVES Rep kee g'd . 45cN D The OPERABILITY poe-75iitainment isol t rrT3T~esv ensures that e containment spfFere will be iso e rom the outsido e3y.ironm n in the even a release of radica ve material to the nesinment atmospher ge o .

pressurizat jionatrGTo'ntainment. Co 4o isolation with fine-tTme limJls- we~Eified ensures that th e ease of radioa environment wips emis~Tstent with the assumption used in the[ctive aatfihal analyses for a to the LOCA.

\

l B 3/4 6-3 Amendment No. 59 SEQUOYAH - UNIT 2 February 11, 1988

TNSERTD The valves identified in Table 3.6-2 are containment isolation valves as defined per 10 CFR 50. The operability of these containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurisation of the containment. Containment '

isolation ~within the time limits specified ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a loss of coolant accident.

Additional valves have been identified as barrier valves, which in addition to the containment isolation valves discussed above, are a part of the accident '

monitoring instrumentation in Technical Specification 3/4.3.3.7 and are designated as Category 1 in accordance with Regulatory Guide 1.97. Revision 2

" Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident." Der ^mber 1980.

I

, , , _ - . , . _ e 2.,,. _-,-c.- . - . - , . m.s.,,,-,- . . . , , , , , , , , , , , , u-,_,,ww.,.+.m.. .,r-e.-..-,,,ww e, ,.,- -. .w.w ,,r--- .-r-v.-,,,

1 A6 MIN!$TRATIVECONTROLS

) I

b. In-Plant Radiation Monitorina  !

l A program which will ensure the capability to accurately determine I the airborne-iodine concentrations in vital areas under accident conditions. This program shall include the following:i (i) Training of personnel. I (ii) Procedures for monitoring, and I (iii). Provisions for maintenance of sampling and analysis equipment.

c. Secondary Water Chemistry A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation. This program shall include:

(i) Identification of_a sampling schedule for the critical variables, and control points for.these variables, },

(ii) Identification of the procedures used to measure the values of '

s the critical variables, * '

n _

(iii) Identification of process sampling points, (iv) Procedures for the recording and management of data, ~6.K j (v) -Procedures defining corrective actions for off-control point chemistry conditions,.

(vi) Procedures identifying (a) the authority. responsible for the I interpretation of_the data; and (b) the sequence and timing.of 4 administrative events required to initiate corrective action,

.and- ,

j

,(vii) Monitoring of = the condensate at the ' discharge of the condensate pumps _for evidence of condenser in-leakage. When condenser in-leakage is confirmed, the leak shall be repaired, plugged, or-isolated. .

I

d. "::ke: "ethed-f+e-GeterriMac Subec011 c "scin be)6el -

A pregram which wi44-ensttre-the-espabH4ty to eeetteetely monit+r the _ l

^ :: tor Coolant-System Subce;1ing ",argin. This program shall includc

-th; fcilcwing.

-(4 ) T reitting-of-pectonnt i , and ,

(44-) Procedurc; fee-mest+ ring.

l SEQUOYAH - UNIT 2_ 6-16 AmendmentNo+hk66 R66 June 30, 1988

j.^ ..

, i ENCLOSURE 2 PROPOSED TECHNICAL SPECIFICATION CHANGE .

SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-89-30 Rev. 1)

DESCRIPTION AND JUSTIFICATION FOR POSTACCIDENT MONITORING (PAM) INSTRUMENTATION '

TECHNICAL SPECIFICATION (TS) 89-30. REVISION 1 I

i 1

f I

, , - . . . . ~ . .

ENCLOSURE 2 Description of Change Tennessee Valley Authority proposes to modify the Sequoyah Nuclear Plant (SQN)

Units 1 and 2 technical specifications (TSs) to revise TS sections involving postaccident monitoring (PAM) instrumentation and containment isolation valves '

(CIVs).

For the PAM instrumentation, the essential raw cooling water (ERCW) to auxiliary feedwater (AFW) valve position and CIV position indication has been added to Limiting Condition for Operation (LCO) 3.3.3.7. The ERCW to AFW valve positiota indication requires one train per motor-driven AFW pump and two trains per steam-driven AFV pump. Each ERCW to AFW injection line contains two isolation valves, each with its own position indication channel.

Thus position indication on both valves on an injection line is necessary. In a similar fashion, the CIV position indication requires one channel per CIV.

Clarification has been added to the minimum channels required for the reactor coolant Tnot and Teoio indication instrumentation. Clarifications have also been added to both the total number of channels and the minimum channels required for the steam generator level (wide-range) indication, reactor vessel level instrumentation, and incore thermocouples instrumentation.

The reactor coolant system (RCS) subcooling margin monitor instrumentation action requirement has been changed to reference Action Statement 1, while Action Statement 6 has been deleted. Additionally, the incore thermocouples action has been changen to reference Action Statement 1.

Action Statements 1 and 5 notes have been revised to reference LC0 3.3.3.5 as opposed to Table 3.3-9. rn addition, the allowable outage times (A0Ts) have been changed in Action Statements 1, 2, and 5 to be consistent with those currently delineated in the Technical Specification Improvement Program (TSIP). These same action statements have also been revised to first require a reduction to HOT STANDBY (Mode 3) within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and then to HOT SHUTDOWN (Mode 4) within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The A0T in Action Statement 4 has been changed from 7 to 30 days, and a clarification has been added to the required timeframe for the completion of the special report.

Additionally, four valves associated with steam generator blowdown were deleted from CIV Table 3.6-2.

A typographical error has been corrected in the basis for LCO 3.3.3.7, and additional information has been added on PAM. The basis section for LCO 3.6.3 has been revised to delete information regarding purge and vent times and to clarify the difference in the scope of valves required to be stroke-time tasted and those additional valves required to have PAM position indication.

Administrative Section 6.8.5, Item d, which describes the program requirements for a backup method for determinind subcooling margin, has been deleted.

Reason for Change The basic changes presented in this proposed TS change are a result of the effort to incorporate the balance of RG 1.97, Revision 2, Category 1, parameters. NRC's letter to TVA dated December 7, 1990, provided approval of Revision 0 to TS 89-30 on PAM instrumentation and requested various changes based upon the NRC staff's position that all RG 1.97, Category 1, instrumentation, not just the Type A, be included in the TSs. Since the wide-range containment pressure and reactor vessel level instrumentation are Category 1 and had been proposed for deletion, the NRC staff agreed that these two parameters would remain in Table 3.3-10 of LCO 3.3.3.7.

TVA's original proposal to include only the Type A Category l, parameters was based upon our understanding of NRC's position on the inctusion of PAM instrumentation in TSs. Subsequent conversations with the NRC staff indicated that the NRC position had been provided in a letter dated May 9, 1988, to the Babcock & Wilcox Owners Group as a part of the split report for the TSIP.

This position indicates that those Category 1 parameters, other than Type A, which have been shown to be not risk significant, may be excluded from the TS. TVA does not currently want to pursue elimination of any Category 1 parameters on the basis of the parameter not being risk significant.

As indicated in the cover letter to this proposal. TVA is withdrawing its original proposal to the basis of TS 3/4.3.3. 7. This withdrawal is being made because it provided a licensing position that indicated only RG 1.97, Revision 2, Type A, Category 1, variables were required tn be in the TS. A new proposed revision has been provided in Enclosure 1.

The clarifications provided for the reactor coolant Tno, and Tcoin, incore thermocouples, steam generator level (wide range), and reactor vessel level instrumentation channel requ!.rements were made to enhance the understanding and thus the application of the specification.

The NRC letter to TVA dated December 7, 1990, also requested that TVA propose action statements for inoperable subcooling margin monitors consistent with the guidance in Generic Letter (GL) 83-37 or provide justification that this guidance is not applicable. GL 83-37 provided a format for PAM instrumentation in the T3. With respect to the subcooling margin monitors, this f ormat assumed that there were two channels available and that actions would be taken, upon loss of a channel, similar to that which was taken for the other two-channel PAM instrumentation. With the upgrade to two PAM channels provided to the subcooling margin monitors during the Cycle 4 autages, TVA could comply with the GL 83-37 TS format.

1 Although it would appear that depending upon the availability of PAM-qualified RCS information (i.e. , pressure and temperature), an acceptable alternate means of determining subcooling margin would still be to dedicate a qualified individual to this calculation, this proposed TS complies with the TS format in GL 83-37. With this proposed TS change, there is no longer an action to add an additional shif t crew member for subcooling margin calculations; thus the program requirements from Administrative Section 6.8.5.d have also been deleted.

The incore thermocouple action reference has been changed from Action Statement 3 to 1. This change was nede to relax the A0T for this parameter.

TVA was overly conservative in its original proposal that required shutdown after 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> upon failure to maintain at least one required PAM channel per core quadrant per train.

Action Statements 1 and 5 have proposed relaxations to the A0T while Action Statement 2 decreases the A0T by one day for loss of one required channel to provide consistency with the A0Ts currently in the TSIP. For the same reason, these same action statements and Action Statement 2 have been re'tised to first require a fixed time to reach Mode 3 and then a fixed time to reach Mode 4.

The A0T in Action Statement 4 was revised for consistency with the above action statements and to clarify the timeframe by which a special report must be submitted to the NRC.

As indicated above, the PAM requirements for CIV position indication have been added in this proposed TS change. In preparation for this submittal, it became cicar that in addition to the need to clarify the basis for LCO 3.6.3 on CIVs, there were four valves associated with the steam generator blowdown system (FCV-1-181, 182, 183, and 184) that are not required to have PAM position indication nor required as CIVs per 10 CFR 50 (i.e., piping inside containment is considered to be a closed system-reference Updated Final Safety Analysis Report Figure 6.2.4-1). To lessen the potential for confusion with the addition of PAM CIV position indication, this TS change proposes to delete these valves. In addition, the basis for LCO 3.6.3 has been revised to delete superfluous information and to clarify that PAM requires position indication for CIVs that receive a containment isolation signal (and are therefore included in TS 3.6.3 for stroke-time testing) as well as those that have been identified as containment barrier valves but do not receive a containment isolation signal. An example of a barrier valve would be a main feedwater isolation valve, main steam isolation valve, containment vacuta relief check valve, or auxiliary f eedwater isolation valve. Those valves that require PAM position indication have been identified as a part of SQN's engineering calculations.

Justification for Change The ERCW to AFW pump valve and CIV position indication is designated Category 1 in accordance with RG 1.97, Revision 2.

The Category 1 variables are required to be in the TS since these variables are of prime importance in limit.ng risk. To limit the risk

c

}

of operator recovery actions, a knowledge of Category 1 variables is required. Furthermorec.recent NRC severe-accident studies have shown significant potential for risk reduction from accident management. This position has been provided in an_NRC letter to the Babcock & Wilcox Owners Group dated May 9, 1988. ,

The clarifications provided for the channel requirements for the Reactor Coolant Twoi and Teoia, incore thermocouples, steam generator level (wide range), and- reactor vessel level instrumentation are justified on the basis that these are administrative enhancements that have not changed the intent or requirements of the TSs. i The change for the RCS subcooling margin monitor parameter to reference Action Statement l will require unit shutdown after exceeding the A0T as-opposed to the assignment of a dedicated shift crew member for subcooling margin calculations and brings this parameter into agreement with GL 83-37. Thus this change can be viewed as conservative relative to the current TS.

The change for the parameter of the incore thermocouples to reference Action Statement 1 allows for a relaxation-in the original, overly conservative ,

proposal that required plant shutdown upon failure to maintain at least one channel per core quadrant per train fo: g'reater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The proposal to reference Action Statement 1 allows the action requirements to be consistent.with those proposed in the current TSIP.

-The A0T has been relaxed in Action Statements 1 and 5, which'provides

• consistency with the A0Ts currently proposed in the TSIP. The A0T of 30 days for the loss of one channel is based on operating experience and takes into y account the remaining operable channel (s) and the low probability of an event.

requiring PAM instrumentation during this interval. With two required channels inoperable in_one or more functions, at least one. channel in each function should be restored to OPERABLE status within seven days. The A0T of seven days is based-on.the relatively low probability of an event requiring PAM instrument-operation and the availability of alternate means to_obtain the required information or the presence of a third required PAM channel.

Continuous operation with two-required channels inoperable (and a third PAM channel not_available) is not acceptable because the alternate indications-may not: fully meet all performance qualification requirements applied to the PAM instrurentation. Therefore, requiring restoration of at least one inoperable channel limits the risk that the PAM function will be degraded should an accident occur.

Also for Action Statements 1, 2, and 5, the proposed change to allow a transition to Mode 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to Mode 4 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is l Justified _on its minimal impact to the cumulative A0T while allowing for a

! more controlled and orderly shutdown from full power without challenging y safety systems and/or operators. This change is also consistent with timeframes proposed in the current TSIP,

s o

Action Statement 4, which still requires an alternate means of radiation monitoring to be established in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, was changed from 7 to 30 days to require the radiation monitor to be returned to operable status. This relaxation from 7 to 30 days was made for consistency with the changes provided to Action Statements 1 and 5 and allows for a greater amount of time to ascertain the extent of the failure and its estimated returnito-service date before filing the special report. Since operation with the RM out of service for an undetermined duration is allowed by Action Statement 4, this change has no effect on the safe operation of the plant. The addition of the words "the next" clarifles that the special report must be submitted 14 days after exceeding the 30-day A0T. This change is considered administrctive in nature and does not affect plant operation.

Action Statement 3 was provided specifically for the PAM CIV position indication parameter. The intent is consistent with that proposed in the current TSIP for loss of any one channel. However, it is recognized that for penetrations with both inboard and outboard PAM-required CIVs. a more expeditious action should be taken if both the inboard and outboard valve position indication is lost. Thus a 7-day A0T is proposed for this condition.

The four inboard steam generator blowdown valves have been proposed for deletion on the basis that inside containment'the plant secondary side is considered a closed system. In accordance with Criterion 57 in 10 CFR 50 Appendix A, the requirement for a closed system inside containment and "cach line that penetrates primary reactor containment and is neither part of the reactor coolant pressure boundary nor connected directly to the containment atmosphere shall have at least one containment valve that shall be either automatic, or locked closed, or capable of remote manual operation- This

-valve shall be outside containment and located as close to the containment as practical. A single check valve may not be used as the automatic isolation valve." The steam generator blowdown line complies with this requirement via outboard containment isolation valves FCV-t-7, 14, 25, and 32 and FSV-43-55, 58, 61, and 64 without the inboard penetration valves. Therefore, these valves are not considered containment isolation valves nor do they require position indication in accordance with RG l.97, Revision 2. This position was provided to the NRC in a letter dated January 2, 1987. The deletion of these valves from Table 3.6-2 had been overlooked subsequent to the establishment of this position. After deleting these valves, all the valves in Table 3.6-2 require PAM position indication monitoring.

O i:_ _,;

c.- .

Environmental Impact Evaluation The proposed change request docs- not involve an unreviewed environmental question because operation of SQN Units 1 and 2 in accordance with this change would not: .

1. Result in a-significant increase in any adverse environmental impact previously evaluated in the Final. Environmental Statement (FES) as modified by1 the. staf f 's ' testimony to the Atomic Safety and Licensing Board, supplements to the FES.-environmental impact appraisals, or-decisions of'the Atomic Safety and Licensing Board.

2. Result in a significant change in effluentslar power levels.

3. Result in matters.not previously reviewed in the licensing basis for SQN that may have a significant environmental impact.

l l-h .

I f

I l

l

{.

l l

,g. -..--.a ...n..sn., .ns, a > .- a ~a--nuax---.-.a. w.~-8 s +.a u. sa a .- - - +> - , - . -,~..s-. ~ ~ . . a- w - .+.

n ,

4 ENCLOSURE 3 PROPOSED TECHNICAL SPECIFICATION CHANGE .,

SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 ,

DOCKET NOS. 50-327 AND 50-328 -

(TVA-SQN-TS-89-30 Rev. 1)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION 2

e e

1

.=

i l.

f; t

?'

l ;-

. _ - ._.. . - _ .- , .._. . _ _ . ~ . . . _ _ . , . - _ - , _ . - _ . _ . __ - _ _ _ - . . ___ . .

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

.t

.. *L. o ENCLOSURE 3 Significant flazards Evaluation l 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 (SQN) in accordance with the proposed amendment will not:

(1) Involve a significant increase in the probability or consequences of an accident previously eva.uated.

The changes proposed do not involve a significant increase in the probability i or consequences of an accident previously evaluated. The changes proposed cannot increase the probability.oi-nn accident since-the proposed changea

. cannot affect components that can cause an accident. The addition of the ERCW to AFW valve and CIV position indication can decrease the consequences of an accident 1 previously evaluated in thnt- the TSs provide for a limited out-of-service time and their availability can help in mitigating the consequences'of an accident. The change for the action requirements for the lose of'subcooling margin monitoring does not increase the consequences of any event since plnut shutdown will be required in lieu of adding an additional crew member. The consequences of an event are not-significantly increa rd by

- the change to the.A0T for Action Statements 1, 2, or 5 since channel redundancy and/or' the relatively short A0T ensures that suf ficient infor .ation

- existsLto. mitigate the consequences of an accident and that there is a relatively low =$robability of an event requiring PAM instrumentation dueing

' the A0T.

'The deletion of the four inboard steam generator. blowdown valves does not increase the consequences of an accident since adequate isolation capability remains-with th- outboard isolation valves and the closed inboard system.

(2)' Crease the possibility of a new or dif ferent kind of accident f ron. any

=previously analyzed. .

. The changes proposed cannot increase the possibility of a new or dif fet, it ~

klnd of accident from any previously analyzed. The PAM (ndicators'and C(Vs

'themselvesicannot create an accident. In a postaccident condition, the J1 indicators-serv < to help the operator mitigate the event. The CIVs.that were deleted are redundant.and not. required by 10 CFR 50. Loss of function of these valves would not increase the possibility of a 'new or dif ferent f ind of accident. ,

t

-.w. -

. . - . < , - . ~ . - __ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ _

5

?

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

The proposed changes do not involve a significant reduction in any margin of safety. The proposed changes to the PAM instrumentation do not affect the design of the safety-related components relied upon to automatically mitigate the consequences of any design basis event occurring while in Modes 1, 2, or 3. Since there is not a significant increase in the consequences of any accident previously evaluated, a significant reduction in any margin of safety cannot exist.

The margin of. safety is not significantly reduced for the proposed change to delete the four inboard steam generator blowdown valves since adequate containment isolation exists with the presence of the outboard isolation valves and the inboard closed system.

l-l l

l