ML20070R005

Proposed Ts,Reflecting Relocation of Response Time Limit Tables 3.3-2 & 3.3-5 to UFSAR

ML20070R005
Person / Time
Site:	Sequoyah
Issue date:	05/16/1994
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20070Q995	List: ML20070Q992 ML20070R005
References
NUDOCS 9405190346
Download: ML20070R005 (30)

Text

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' J ENCLOSURE 1 1

l PROPOSED TECHNICAL SPECIFICATION CHANGE l SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 j

! DOCKET NOS. 50-327.AND 50-328 (TVA-SQN-TS-94-03)'-

i i

! . LIST OF AFFECTED PAGES Unit 1 3/4'3-1  !

3/4 3-9 3/4 3 3/4 3-14 3/4 3-29 3/4 3-30 3/4 3-31 3/4-3-32 3/4 3-33 3/4-3-33a B3/4 3-2 Unit'2 )

3/4 3-1 3/4 3-9 3/4 3-10 3/4 3-14 3/4 3-29 3/4 3-30 3/4 3-31 3/4 3-32 3/4 3-33 3/4.3-33a B3/4 3-2

~ 9405190346940516 PDR .ADOCK- 05000327

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3/4.3 INSTRUMENTATION l

3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION v

3.3.1.1 As a minimum, the reactor trip system instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE.with RESPONSE TIMES as showr. m

-Tab' A _A- --

APPLICABILITY: As shown in Table 3.3-1.

ACTION:

l As shown in Table 3.3-1.

i l

SURVEILLANCE REQUIREMENTS l

4.3.1.1.1 Each reactor trip system instrumentation channel and interlock shall R16 be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-1.

4.3.1.1.2 The logic for the interlocks shall be demonstrated OPERABLE prior to each reactor startup unless performed during the preceeding 92 days. The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.

4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shallbedemonstratedtobewithinitslimitatleastonceper18 months.Q4 Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the " Total No. of Channels" column of Table 3.3-1.

v s Neunce bencroas A RE E venwr Faom #csawse 7;,c TE5Yl!JG. p -

MAR 251982 SEQUOYAH - UNIT 1 3/4 3-1 Amendment No. 12

l .

I o, f$ THis TAELE TABLE 3.3-2 8 2s DELETED

! g I * -REACTOR TRIP SYSTEf4 INSTRtMENTATION-RESPONSE-TWICS l- c- -FUNCTI L UNIT x s -

Z l. Manual ctor Trip RESPONSE TIME --

e NOT APPLICABLE

2. Power Range, utron Flux

3. $ 0.5 seconds

• Power Range, Neutro Flux, liigh Positive Rate NOT APPLICABLE

( 4. Power Range, Neutron Flux, liigh Negative Rate 5.

5 0.5 seconds *

• Intermediate Range, Neutron Flux i

• T

6. Source Range, Neutron Flux  % NOT APPLICABLE NOT APPLICABLE

7. Overtemperature Delta T

$ 8.0 seconds *

8. Overpower Delta T R145 9.

$ 8.0 seconds Pressurizer Pressure--Low i

5 2.0 seconds i 10. Pressurizer Pressure--liigh 5 0 seconds

11. Pressurizer Water Level--liigh NOT APP CABLE 2, 12. Loss of Flow - Single Loop

@ (Above P-8)

~

g -< 1.0 seconds s

g

j. s o x a

, R& Neutron detectors are exempt from response time testing. Response time of the

~~_, neutron flux signal portion of the' channel shall be measured from detector O cutput or input cf first clcctronic cci,,cncat in don..cl.

J

~

N .

g .

.o TABLE 3.3-2 (Continued)

E sf REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES x

E FUNCTIONALUNIT\

N RESPONSE TIME M 13. LossofFlow-Tkloops s

(Above P-7 and belon4-8)

14. N -< 1.0 seconds Main Steam Generator Wat Level--

Low-Low  ;

' A. RCS Loop AT (P5 50% RTP: P> 50% RTP) 5 8.0 seconds B. Steam Generator Water R145 Level -- Low-Low 5 2.0 secondsII) w (Adverse, EAM)

C. 3 Containacnt Pressure

~ T (EAM) 1 2.0 seconds (1)

E$

15. Deleted 16.

Undervoltage-Reactor Coolant Pumps /

17. _ l.2 seconds Underfrequency-Reactor Coolant Pumps

18. Turbine Trip 5 O. seconds

!. A. Low Fluid Oil Pressure B. Turbine Stop Valve NOT APPLICAB NOT APPLICABLE

.y 19. Safety Injection Input from ESF NOT APPLICABLE i

-< k 20. Reactor Trip Breakers  !

=

cb 21. Automatic Trip Logic NOT APPLICABLE

22. NOT APPLICABLE Reactor Trip System Interlocks

= .2 e -

NOT APPLICABLE ca ru ~

(1) Does not include Trip Time Delays.

O protection cabinets, solid state protection cabinets, and actuation devices. Response times R145noted time necessary for TilERHAL POWER in excess of 50% RTP. This reflects the response

~

.o INSTRUMENTATION ,

3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION f l (

LIMITING CONDITION FOR OPERATION

/

/

Y Y '

3.3.2.1 The Engineered Safety Feature Actuation Syst m (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be' 0PERABLE with their trip setpoints set consistent with the values shown in the Table 3.3-4. cad with RESPONSE TIMES c; ;hcun in3.3-S. Tchkl Trip Setpoint column of

~ A ~ A APPLICABILITY: As shown in Table 3.3-3. j ACTION:

I

/

a. With an ESF/S instrumentation channel or inteilock trip setpoint less conservativt than the value shown in the Al}bwable Values column of Table 3.3-4, declare the channel inoperable'and apply the applicable ACTION requirement of. Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value.

b.

With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3.

SURVEILLANCE REQUIREMENTS 4.3.2.1.1 Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCilVNAL TEST operations for the MODES and at the frequencies shown in Table 4.3-2.

4.3.2.1.2 The logic for the interlocks shall be demonstrated OPERABLE during 1 the automatic actuation logic test. The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.

4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS functio ~

shall be demonstrated to be within the limit at least once per 18 months.

Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per fur.ction such that all channels are tested at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the " Total No. of Channels" Column of Table 3.3-3.

SEQUOYAH - UNIT 1 3/4 3-14 '

I

~

~

Itus RbLE Ls D ELETeT) l ENGIMERED CAIETY-FEATURES RESPONSE TI"ES RITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 1.\ Manual N

a. Safety Injection (ECCS) Not Applicable edwater Isolation Nct Applicable

Reabtor Trip (SI) Not Applicable Containpent Isolation-Phase "A" Not Applicable Containm t Ventilation Isolation Not Applicable Auxiliary dwater Pumps Not Applicable Essential Raw ooling Water System Not Applicable j Emergency Gas Tr tment System Not Applicable 4

b. Containment Spray Not Applicable Not Applicable ContainmentIsolation-(hase"B" Containment Ventilation 1 olation Not Applicable Containment Air Return Fan d Not Applicable 4

c'

c. Containment Isolation-Phase " Not Applicable 9

Emergency Gas Treatment System h Not Applicable

.l Containment Ventilation Isolation Not Applicable

d. Steam Line Isolation Not Applicable

2. Containment Pressure-High R59

a. Safety Injection (ECCS) s 32. (1) l b. Reactor Trip (from SI) 13.0

) c. Feedwater Isolation < 8.0(2) i d. Containment Isolation-Phase "A"(3) 18.0(8)/28. (9)

M

e. Containment Ventilation Isolation 5.5(8)(13)

f. Auxiliary Feedwater Pumps 1 60(11) lggi

g. Essential Raw Cooling Water System Emergency 00: T reatment System 1 65.0(8)/75.0(9) \\

39.0(9) i i

, SEQUOYAH - UNIT 1 3/4 3-29 Amendment No. 55, 77, 106 March 13, 1989

, TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATIN ASIGNAL AND FUNCTION RESPONSE TIME IN SECONOS

3. Pressurizer essure-Low i

j a. Safety Inject (ECCS) 1 32.0(1)/28.0(7)~

i b. Reactor Trip (froa I) < 3.0 8.0(2) i c. Feedwater Isolation

d. Containment Isolation-Pha "

A"(3) 18.0(8)

e. Containment Ventilation Is & on 5.5(8)(13)

f. Auxiliary Feedwater Pumps d 60(11) i g. Essential Raw Cooling Water System k 65.0(8)n5.0(9)

h. Emergency Gas Treatment System k28.0(8)

! 4. Deleted i

5. Fecative Steam Line Pressure Rate - High

a. Steam Line Isolation 1 8.0 i R145 i

J _

I 1

i i

+

j i

i SEQUOYAH - UNIT 1 3/4 3-30 Amendment No. 55, 77, 106 , 141 iO}9%

g .m. -

s.,,,.--.,m., --.<.-r.n.~ , - - ,..y ..,,n nr,n.,......,..,e. ,.,--,m,, gw.7,,.w,-v.

- og ,- w.r,,, sg4.g e - g,,-w,r- -

i .

TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES

\

INITIATIN'.GSIGNAL.AND FUNCTION RESPONSE TIME IN SECONOS s

R145

6. Steam L 'e Pressure-Low

\

a. Safety $hjection(ECCS) 1 28.0(7)/28.0(1) l

b. Reactor Tr'ip (from SI) 13.0 i c. Feedwater Is ation < 8.0(2)

d. 18.0(8)/28.0C9)

ContainmentIso(ation-Phase"A"(3) Not Applicable

e. Containment Venttiation Isolation

f. Auxiliary Feedwate umps < 60(11)

g. Essential Raw Coolin ater System 65.0(8)/75.0(9)

I h. Steam Line Isolation < 8.0

1. Emergency Gas Treatment 5 tem 1 38.0(9)

7. Containment Pressure--Hiah-High 1 Containment Spray 1 208(9) a.

b. Containment Isolation-Phase "B"(12 - 1 65(0)/75(9) -

c. Steam Line Isolation 1 7.0

d. Containment Air Return Fan > 540.0 and $ 660

8. Steam Generator Water level--Hich-Hich

a. Turbine Trip i 2.5

b. Feedwater Isolation i 11. (2)

9. Main Steam Generator Water Level -

Low-Low

a. Motor-driven Auxiliary 1 60.0(14) ,

Feedwater Pumps (4)

b. Turbine-driven Auxiliary < 60.0(14)

Feedwater Pumps (5)(11)

SEQUOYAH - UNIT 1 3/4 3-31 Amendment No. 55, 59, 63, 77, 82 , 141 1990 k /.(}'.{ $ g May 16, ,

.+re , . -,rw..-3,m- - - . . , , , , - _.-e.-,, , ...,,n.wim, ,..,.-,ewm.---,-w--y- ,--.nw,----we,n..y-.m--.,,m.,,y-.g.,w- ,v. .*i

's l

TABLE 3.3-5 (Continued)

I ENGINEERED SAFETY FEATURES RESPONSE TIMES l INITIATIM SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

10. Stati lackout

a. Auxili Feedwater Pumps #81 5 60(11)

11. Trip of Main Fee %3ter Pumps i

a. Auxiliary Feedwa r Pumps #31 1 60(11) l

12. Loss of Power

a. 6.9 kv Shutdown Board - D raded < 10(10) g31 Voltage or Loss of -

Voltage

13. RWST Level-Low Coincident with Containme bmp Level-High and Safety Injection P

a. Automatic Switchover to '

Containment Sump i 250

14. Containment Purge Air Exhaust Radioactivity - High

a. Containment Ventilation Isolation i 10(6)

~

/ R172 SEQUOYAH - UNIT 1 3/4 3-32 Amendment No, fB/,/JJ,168 June 25, 1993

~

STRUMENTATION TABLE 3.3-5 (Continued)

TABLE TATION (1) Die generator starting and sequence loading delays included. Response time it includes opening of valves to establish SI path and attainment '

of disc rge pressure for centrifugal charging pumps, SI and RHR pumps.

(2) Using air erated valve. The.ESFAS instrumentation channel RESPONSE TIME require ent for specific feedwater air-operated valve (s) can also be met when the sociated air-operated va'1ve is either. closed with air- gi supply (s) isola ed, isolated by a closed manual valve, or isolated by a closed feedwater ' solation valve with power removed. When using one of these provisions r satisfying the air-operated valve response time, the-closed or isolated c ndition described above will be verified at least once per 7 days.

(3) The following valves are exceptions to the response times shown in the table and will have the v signalsandfunctionindic{ueslistedinsecondsfortheinitiating a ed:

Valves: FCV-26-240, -243 Response' times: 0) 2.d.

3.d. 2221((8)/ (9) 8 j )

l Valves: FCV-61-96, -97, -110, -122, -1 , - -192, -193, -194 l Response times:

2.d. 31(8) l 3.d. 32(8) k 4.d. 31((0) 8) 5.d.

6.d. 31 34(8) -

Valve: FCV-70-143 Response times: 8) 2.d. 61(8)/71(9) 3.d. 61(

4.d 62(8)j (9) a 5d 64(8) (9) 6.d. 61(8)j j (9)

(4) On 2/3 any Ster- Generator l (5) 0n 2/3 in 2/4 Steam Generator  :

I (6) Radiation detectors for. Containment Ventilation Isolation may be excl ed i from Response Time Testing.

[. .

SEQUOYAH - UNIT 1 3/4 3-33 Amendment No. 17, 55, 165 December.8, 1992 1

, _n..,,, , , , ,,.. - ,, - -n, m v- r*='^ * * ' ' ' "' ' ' ' ~ ' " ' ' ' '

j >:.

STRUMENTATION TABLE 3.3-5 (Continued)

TABLE N ATION l (7) Diese enerator starting and sequence loading delays not included.

Offsite ower available. Response time limit includes opening and R59 closing o valves to establish SI path and attainment of discharge pressure fo centrifugal charging pumps.

(8) Diesel generat starting and sequence loading delays not included.

Response time 11 it includes operating time of valves.

(9) Diesel generator st ting and sequence loading delays included. Response time limit includes o erating time of valves.

l (10) The response time for lo of voltage is measured from the time voltage is lost until the time fu voltage is restored by the diesel. The response time for degraded 1tage is measured from the time the load i

shedding signal is generated, ither from the degraded voltage or the SI 7 enable timer, to the time full oltage is restored by the diesel. The response time of the timers is c ered by the requirements on their setpoints.

(11) The. provisions of Specification 4.0.4 re not app 1'icable for entry into MODE 3 for the turbine-driven Auxiliary eedwater Pump.

(12) The following valves are exceptions to the esponse times shown in the Table and will have the values listed in sec ds for the initiating signals and the function indicated:

Valves: FCV-67-89, -90, -105, -106 Response times: 7.b, 75(8)/85(9) 4 Valve: FCV-70-141 Response times: 7.b, 70(8)/80(9) ,

(13) Containment purge valves only. Containment radiation monito valves have )

a response time of 6.5 seconds or less. /

(14) Does not include Trip Time Delays. Response times noted include e transmitters, Eagle-21 process protection cabinets, solid state R145 protection cabinets, and actuation devices (up to and including pump .

This reflects the response times necessary for THERMAL POWER in excess f 50% TP.

SEQUOYAH - UNIT 1 3/4 3-33a Amendment No. 29, 77, 82, 55 106, 141

.May 16, 1990

,C .

INSTRUMENTATION BASES i

f 1

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 x ' Tat LAPbero Fmet. 5 AFET)/ 8Ndlyns k@ ORT ~

as not applicable TN A -

Response time may be demonstrated by any s +eries of sequential, overlapping  ;

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

be demonstrated by either 1) in place, onsite or offsite test measurements or l 2) utilizing replacement sensors with certified response times.

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 RSSl 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 l

WCAP-10271, Supplement 1, which determines bypass breaker availability.

l l l 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that 1) the I radiation levels are continually measured in the areas served by the individual I channels and 2) the level trip setpoint alarm or automatic action is initiated when the radiation is exceeded.

3/4.3.3.2 MOVABLE INCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum I 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 l

each detector used and determining the acceptability of its voltage curve. l ForthepurposeofmeasuringF(Z)orFNHafullincorefluxmapisused.

q Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in 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 Power Range Channel is inoperable. '

l l

3/4.3.3.3 SEISMIC INSTRUMENTATION l

The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event l

' 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 March 16, 1987 l SEQUOYAH - UNIT 1 B 3/4 3-2 Amendment No. 54

is l

3/4.3 1NSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION I

LIMIT 1HG CONDITION FOR OPERATION A

. ^ --- '

3.3.1 As a minimum, the reactor trip system instrumentation channels and Tinterlocks L1- 9 'S _ of 9 Table 3.3-1 shall be OPERABLE with PI T 0 E E TIMES x h = i-suwss sw a. .

APPLICABILITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

SURVEILLANCE REQUIREMENTS 4.3.1.1.1 Each reactor trip system instrumentation channel and interlock i shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-1.

4.3.1.1.2 The logic for the interlocks shall be demonstrated OPERABLE prior to each reactor startup unless performed during the preceeding 92 days. The I total interlock function shall be demonstrated OPERABLE at least once per i 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.  ;-

4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be demonstrated to be within its limit at least once per 18 months.

Each test shall include at least one logic train such that both logic trains i are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the " Total No. of Channels" column of Table 3.3.1.

Ngtagog DET EcToEs hR.E EXGm PT Faom RE5PONsE ,

Timt 7E3TI N G. A ^- #

% f I

l 4

SEQUOYAH - UNIT 2 3/4 3-1 )

TABLE 3.3-2 y,

m TaGas % Das '

E REACTOR TRIP-SY& TEM INSTRUMENTAT40N-RESPONSE TIMES E

i FUNCTION UNIT C RESPONSE TIME 5 1. ' Manual Re or Trip w Not Applicable

2. Power Range, Ne on Flux 1

5 0.5 seconds *

3. Power Range, Neutron x, '

High Positive Rate Not Applicable \

4. Power Range, Neutron Flux, \

High Negative Rate 5 0.5 seconds *

5. Intermediate Range, Neutron Flux Not Applicable 1 6. Source Range, Neutron Flux Not Applicable

7. Overtemperature AT

$ 8.0 seconds *

8. Overpower AT R132 5 8.0 seconds

9. Pressurizer Pressure--Low < 2.0 seconds *

10. Pressurizer Pressure--High 1 2. econds

11. Pressurizer Water Level--High Not Applic le

> 12. Loss of Flow - Single Loop

,;p2 (Above P-8)

C ! E 5 1.0 second /

2

, - E

- . x

?

t'

• Neutron detectors are exempt from response time testing. Response time of the neutron flux signal por " n

~ Of the chcnnel sh: 1! bc ecc:urcd fr r dctcctor output er i.put of first clectronic cc penent i charncl.

, M w

A

._._.m..__.m. _-..-..._m _.__.~.__.__....-..__..._._m_._.. ._...._......_._.-.~._=__._m __

~

t.

• + .

i m TABLE 3.3-2 (Continued) ~ '

E REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES 8 i' i Y I

FUNCTIONAL UN RESPONSE TIME E 13. Loss of Flow - wo Loops Z (Above P-7 and b w P-8)  !

i ro 5 1.0 second i

14. Main Steam Generator er Level--Low-Low 4

' A. RCS Loop AT (P $ 50% RTP; P > 50% P) i 8.0 seconds (I)

B. Steam Generator Water 4 < 2.0 secondsg Level--Low-Low t (Adverse, EAM)

C. Containment Pressure (EAM) $ 2.0 seconds (1) R132 .

=

15. Deleted b

16. Undervoltage-Reactor Coolant Pumps  % +

Y 5 1.2 seconds g 17. Underfrequency-Reactor. Coolant Pumps i 5 0.6 seconds .!

18. Turbine Trip  !

j A. Low Fluid Oil Pressure Not Applicable  ;

B. Turbine Stop Valve Not Applicable j 19. Safety Injection Input from ESF.

No pplicable L 20. Reactor Trip Breakers j i Not App ' able i

. p 21. Automatic Trip Logic

/

- <a Not Applicab -

l hi h 22. Reactor Trip System Interlocks

. p' g Not Apolicable j

l

~l o[ (1) Does not include Trip Time Delays.

Response times noted include the: transmitters, Eagle- '

'" process' protection' cabinets, solid' state protection cabinets and actuation devices. This- )

' (# U

^

reflects the response time necessary for THERMAL POWER in excess of- 50% RTP. R132 I w A A A

_ _ _ _ ___r _ _ _ . _

i . l i

J

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INSTRUMENTATION 5

)

3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION i

! LIMITING CONDITION FOR OPERATION i e m

i 3.3.2 The Eng;.7eered Safety Feature Actuation System (ESFAS) instrumentation

{ channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their trip

{

4-setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4, nd with RESPON;E TIMES es she.n in Teble 0.0 5.

! ~  % .

APPLICABILITY: As shown in Table 3.3-3.

j ACTION: i

a.

j With an ESFAS instrumentation channel or interlock trip setpoint i less conservative than the value shown in the Allowable Values

column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted con-

{ sistent with the Trip Setpoint value.

s

b. With an ESFAS instrumentation channel or interlock inoperable, take

the ACTION shown in Table 3.3-3. ,

5 SURVEILLANCE REQUIREMENTS e

i 4.3.2.1.1 i Each ESFAS instrumentation channel and interlock shall be demon-j strated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies j shown in Table 4.3-2.

4.3.2.1.2 The logic for the interlocks shall be demonstrated OPERABLE during j

the automatic actuation logic test. The total interlock function shall be i

demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION

, testing of each channel affected by interlock operation.

4.3.2.1.3 i The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS funct shall be demonstrated to be within the limit at least onceEper 18 months.

{ Each test shall include at least one-logic train such that both logic trains i

are tested at least once per 36 months and one channel per function such that all channels are tested at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in i

the " Total No. of Channels" Column of Table 3.3-3.

i 1

i j SEQUOYAH - UNIT 2 3/4 3-14 J

(

, .- . . . ~ - - - - ~ - , - , - , , , .-,,..,,.c ,, ,-s + - - - , - - , , , - . -

TABLE 3.3-5 TOs TAot E Ls Det.e i eu ENCINEERED SAFETY FEATURES-RE4PONSE TUiES INITI ING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS l 1. Mahal

a. Sa ty Injection (ECCS) Not Applicable Feedw ter Isolation Not Applicable

Reactor rip (SI) Not Applicable

Containmen Isolation-Phase "A" Not Applicable

{

Containment ntilation Isolation Not Applicable j Auxiliary Feed ter Pumps Not Applicable Essential Raw Coo 'ng Water System Not Applicable ,

l Emergency Gas Treatm t System Not Applicable '

b. Containment Spray Not Applicable Containment Isolation-Pha "B" Not Applicable Containment Ventilation Iso tio Not Applicable Containment Air Return Fan Not Applicable

c. Containment Isolation-Phase "A" Not Applicable Emergency Gas Treatment System Not Applicable Containment Ventilation Isolation Not Applicable

d. Steam Line Isolation t t Applicable i 2. CcatainmentPressure-Hig

a. Safety Injection (ECCS) ) R47 5,32.0

b. Reactor Trip (from SI) <3.0

c. Feedwater Isolation 8.0(2)

d. Containment Isolation-Phase "A"(3) 18.0(8)/28.0(

e. R96 Containment Ventilation Isolation 5.5(8)(13) )

f. Auxiliary Feedwater Pumps <60(11) lR6

g. Essential Raw Cooling Water System 65.0(8)/75.0(9) h Emergency Cc: Treatment-Sys4cm -38. 0(9)

[ /

' SEQUOYAH - UNIT 2 3/4 3-29 Amendment No. 47, 68 , 96 March 13, 1989

TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL ND FUNCTION RESPONSE TIME IN SECONDS

3. Pressurizer Pressu -Low R47

a. Safety Injection ECCS) 132.0(1)/28.0(7) l b. Reactor Trip (from 1 3.0 l

c. Feedwater Isolation 1 8.0(2)

d. Containment Isolation-Phas "A"(3) 118.0(8) i e. containment Ventilation Isola 'on R96 5 5.5(8)(13) l

! f. Auxiliary Feedwater Pumps 160(11) R68 -

g. Essential Raw Cooling Water System @ 165.0(8)/75.0(9)

h. Emergency Gas Treatment System @ $28.0(8)

4. Deleted

5. Negative Steam Line Pressure Rate-High

a. Steam Line Isolation 58.0 R132 SEQUOYAH - UNIT 2 3/4 3-30 Amendment No. 47, 68, 96, 132 Q {jT [ '[ 'iC C )

i TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING IGNAL AND FUNCTION RESPONSE TIME IN SECONDS

6. Steam Line Pressure-Low R13

a. Safety jection (ECCS) 5 28.0(7)/28.0(1)' R47

b. Reactor T p (from SI) $ 3. 0

c. Feedwater Is lation < 8.0(2)

d. Containment Is lation-Phase "A"(3)' 5 18.0(8)/28.0(9)

e. Containment Vent ation Isolation Not Applicable R68

f. Auxiliary Feedwate Pumps 160(11)

g. Essential Raw Cooling ater System 5 65.0(8)/75.0(9) i h. Steam Line Isolation < 8.0-

i. Emergency Gas' Treatment S tem 38.0(9)

I

7. Containment Pressure--High-High dp R51

a. Containment Spray 1 208(9) \

b. Containment Isolation-Phase "B"(12 R73 1 65(8)/75(9) l c. Steam Line Isolation 3 7.0

d. Containment Air Return Fan

> 540.0 and 1660

8. Steam Generator Water Level--High-High R55

a. Turbine Trip d' 2.

b. Feedwater Isolation )

5 11.0

9. Main Steam Generator Water Level -

Low-Low

a. Motor-driven Auxiliary R132 1 60.0(14)

Feedwater Pumps

b. Turbine-driven Auxiliary R132 1 60.0(14)

Feedwater Pumps (5)(11)

SEQUOYAH - UNIT 2 3/4 3-31 Amendment No. 47, 51, 55, 68, 73 , 132 October 31, 1990 e ,; - - -

ENGINEERED SAFETY FEATURES RESPONSE TIMES

. 1 INITIATIN SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS i

10. Statio lackout

a. Auxili Feedwater Pumps 5 60(11) i

11. Trip of Main Fee ter Pumps

a. Auxiliary Feedwa er Pumps 5 60(11)

12. Loss of Power

a. 6,9 kv Shutdown Board - raded Voltage or Loss of -

< 10(10)

Voltage

13. RWST Level-Low Coincident with Containm t Sump Level-High and Safety Injection

a. Automatic Switchover to Containment Sump 5 250

14. Containment Purge Air Exhaust Radioactivity - High

a. Containment Ventilation Isolation 5 10(6)

R158

- 3 A

SEQUOYAH - UNIT 2 3/4 3-32 Amendment No. 18//68,158 June 25, 1993

~

.hSTRUMENTATION TABLE 3.3-5 (Continued)

TABLE TATION

[ (1) Diese generator starting and sequence loading delays included. Response time 1 it includes opening of valves to establish SI path and attainment I of disc rge pressure for centrifugal charging pumps, SI and RHR pumps. J (2) Using air erated valve. The ESFAS instrumentation. channel RESPONSE TIME require nt for specific feedwater air-operated valve (s) can also be gi, met when the sociated air operated valve is either closed with air supply (s) isola ed, isolated by a closed manual valve, or isolated by a closed feedwater solation valve with power removed. When using one of these provisions r satisfying the air-operated valve response time, the closed or isolated ndition described above will be verified at least /

once per 7 days.

(3) The following valves are exceptions to the response times shown in the table and will have the v ues listed in conds for the initiating signals and function indica ed:

Valves: FCV-26-240, -243 Response times: 2.d. 21 f /

(9) R8 5

3.d. 22 '

4.d. 21(8)j )

5.d. 24(8)j (9 i 6.d. 21(8)j (9)

Valves: FCV61-96, -97, -110, -122, -19 -192, -193, -194 Response times 0) 2.d.

3.d. 32( 31(8) j 4.d. 31(8) p 6d: 3$

Valve: FCV-70-143 Response times: 2.d. 61 /71(9) 3.d. 62(8)j (9) 4.d. 61(8)j (9) 5.d.

6.d. 61 64(8)j (9)

(4) On 2/3 any Steam Generator (5) On 2/3 in 2/4 Steam Generator (6) Radiation detectors for Containment Ventilation Isolation may be excl ed from Response Time Testing.

A ~ '

SEQUOYAH - UNIT 2 3/4 3-33 Amendment No. 8, 47. 155 December 8, 1992

l i

.- V

. STRUMENTATION i

TABLE 3.3-5 (Continued) ,

i l

TABLE N ATION i

)

(7) Diese generator starting and sequence loading delays not included.

Offsite ower available. Response time limit includes opening and g47 closing valves to establish SI path and attainment of discharge pressure f centrifugal charging pumps.

(8) Diesel genera r starting and sequence loading-delays not included, j Response time 'mit includes operating time of valves.

(9) Diesel generator s arting and sequence loading delays included. Response time limit includes perating time of valves.

(10) The response time for ss of voltage is measured from the time voltage R68 is lost until the time f 1 voltage is restored by the diesel. The

, response time for degrade voltage is measured from the time the load shedding signal is generate either from the degraded voltage or the SI

-enable timer, to the time ful voltage is restored by the diesel. The

~

response time of.the timers is overed by the requirements on their setpoints.

(11) The provisions of Specification 4.0. .are not applicable for entry into R68 MODE 3 for the turbine-driven Auxilia Feedwater Pump.

(12) The following valves are exceptions to t e response times shown in the table and will have the values listed in conds for the initiating /

signals and the function indicated:

"3 \

Valves: FCV-67-89, -90, -105, -106 (p Response times: 7. b , 75(8)/85(9)

+

Valve: FCV-70-141 Response times: 7.b, 70(8)/80(9) 1 (13) Containment purge valves only. Containment radiation mon' tor valves have R96 a response time of 6.5 seconds or less.

(14) Does not include Trip Time Delays. Response times noted inc1 e the transmitters, Eagle-21 process protection cabinets, solid state R132 protection cabinets, and actuation devices (up to and including p mps). /

This reflects the response times necessary for THERMAL POWER in ex ss of 50% RTP. f l

SEQUOYAH - UNIT 2 3/4 3-33a Amendment No. 18, 47, 68, October 31, 19h' '

s -

INS'TRUMENTATION i

i BASES l

i REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM -

{ INSTRUMENTATION (Continued) j The measurement of response time at the specified frequencies _provides assurance that the protective anti the engineered safety feature actuation associated with each channel is completed within the time limit assumed in the ,

i~ accident analyses. No credit was taken in the analyses for those channels-  !

! with response times indicated as not applicablegTN THE UPDATm b N AL.

MPETy AN%fsts, REPORr. _

l ' Re'sponse '

monstrated by any series of sequential, overlapping 1

or total channel test measurements provided that such tests demonstrate the

. total channel response time as defined. Sensor response time verification may I be demonstrated'by either 1) in place, onsite or offsite test measurements or

! 2) utilizing replacement sensors with certified response times.

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 R46 i 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 1, which determines bypass breaker availability.

3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION '

4

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

3/4.3.3.2 MOVABLE INCORE DETECTORS

! The OPERABILITY of the movable incore detectors with the specified minimum l complement of equipment ensures that the measurements obtained from use of

, this system accurately represent the spatial neutron flux distribution of the i 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 H a full incore flux' map is used.

9 l Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in 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 j POWER TILT RATIO when one Power Range Channel is '@erab h j 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 i 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 j R72

} SEQUOYAH - UNIT 2 B 3/4 3-2 Amendment No. AU , 72 j September 1, 1988

4 4

ENCLOSURE 2 PROPOSED TECHNICAL SPECIFICATION CHANGE

! SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-94-03)

DESCRIPTION AND JUSTIFICATION FOR RELOCATION OF RESPONSE TIME LIMIT TABLES

Deacriplion of Citange TVA proposes to modify the Sequoyah Nuclear Plant (SQN) Units 1 and 2 technical specifications (TSs) to relocate the response time limits for reactor trip and engineered safety feature functions. This change will revise TS Items 3.3.1.1, 3.3.2.1, and 4.3.1.1.3 for Unit 1 and Items 3.3.1, 3.3.2, and 4.3.1.1.3 for Unit 2. These revisions remove references to TS Tables 3.3-2 and 3.3-5 and incorporate a footnote that exempts neutron detectors from response time testing for Surveillance 4.3.1.1.3. TS Tables 3.3-2 and 3.3-5 have been deleted. A clarification has been added to the bases section indicating that the response time limits are maintained in the Updated Final Safety Analysis Report (UFSAR).

Egason for Change Relocation of the response time limits from TSs to the UFSAR will allow

TVA to administrative 1y control changes to these limits in accordance with 10 CFR 50.59 without the need to process license amendment requests. Placing these limits in the UFSAR will provide the ability to add additional discussions related to the existing response time footnotes that will clarify how the limits are applied. The proposed change will significantly reduce the time and cost associated with processing changes to the response time limits. The requirements to ensure such changes are adequately evaluated for the impact to nuclear safety will be controlled by the 10 CFR 50.59 process.

duS11fication for Change The proposed change is consistent with the guidance provided by NRC in Generic Letter 93-08 dated December 29, 1993, for the relocation of response time limit tables to the UFSAR. The TS surveillance requirements that verify the response times are within limits are unchanged and only the location of the limits is being changed. Changes ,

to the response time limits, after relocation to the UFSAR, will be controlled by the 10 CFR 50.59 process to ensure the SQN design basis is I maintained and no unreviewed safety question is created. Required plant procedure revisions will be subject to the requirements that control changes to plant procedures in accordance with the administrative controls section of the TSs. Changes to the response time limits will be submitted to NRC with the periodic updates to the UFSAR in accordance with 10 CFR 50.71(e). TVA will incorporate the reactor trip system and engineered safety feature actuation system response time limits in the next update of'the Final Safety Analysis Report following issuance of the TS amendment by NRC.

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

1. Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as l

,e .

modified by the staff'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 effluents or power levels.

'. Result in matters not previously reviewed in the licensing basis for SQN th. sit may have a significant environmental impact.

t

c* ..- ,

Enclosure'3.

)

i PROPOSED TECHNICAL' SPECIFICATION CHANGE i

1. . . .

j- SEQUOYAH NUCLEAR PIANT UNITS 1 AND -2" I iDOCKET NOS 50-327 AND 50-328 a

l . (IVA-SQN-TS-94-03) .

I DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION' i

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l Significant Hazards Evaluation l

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

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

The proposed change does not alter the response time limit requirements for the reactor trip or engineered safety feature actuation systems or surveillance testing and frequency. Placing ,

these limits in the Updated Final Safety Analysis Report (UFSAR) will I ensure the plant design basis is maintained in accordance with '

l 10 CFR 50.59. Since no actual changes to response time limits or surveillance requirements are involved, the probability or consequences of an accident are not increased.

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

The proposed change does not affect any plant equipment, functions, or setpoint by relocating response time limits to the UFSAR.

Therefore, the possibility of a new or different kind of accident is not created.

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

The proposed change will continue to require SQN to maintain the  !

plant functions at the required setpoints necessary for the design  !

basis and to support the accident analysis. The margin of safety is l not reduced because there is no change to plant functions and the l 10 CFR 50.59 process will continue to ensure the plant design basis is appropriately maintained. I

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1 1

Enclosure 4 i

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PROPOSED TECHNICAL SPECIFICATION CHANGE e

! SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2

' DOCKET NOS. 50-327 AND 50-328 5-1- (TVA-SQN-TS-94-03)~

+

COMMIIMENT '

SUMMARY

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I . . _ . _ _ _ . _ . _ . - = _ _ _ . _ . ._ _ - . . . . , . . . ~ - . . , . , _ -.

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e l Commitment TVA will incorporate the reactor trip system and engineered safety feature actuation system response time limits in the next update of the Final Safety Analysis Report following issuance of the technical specification amendment by NRC.

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