Text

Proposed Tech Specs,Adding Requirements to Limiting Conditions for Operation 3.3.1.1 & 3.3.1 Re Operability of New Shunt Trip Attachment to Reactor Trip Breakers
	ML20198L547
Person / Time
Site:	Sequoyah
Issue date:	05/23/1986
From:	; TENNESSEE VALLEY AUTHORITY
To:	;
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ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION CHANGES SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 (TVA SQN TS 67)

Proposed changes to add requirements to LCO 3.3.1.1 (unit 1) and LCO 3.3.1 (unit 2) technical specifications for operability of the new shunt trip attachment. This changes also incorporates relief of surveillance and outage times as allowed by Westinghouse Standard Technical Specifications for applicable reactor protection system instrumentation and engineered safety features manual actuation circuitry.

List of effective pages UNIT 1 3/4 3-4 3/4 3-5 3/4 3-6 3/4 3-7 3/4 3-8 3/4 3-11 3/4 3-13 3/4 3-34 3/4 3-35 3/4 3-36 3/4 3-37a 3/4 3-38 B3/4 3-2 UNIT 2 3/4 3-4 3/4 3-5 3/4 3-6 3/4 3-7 3/4 3-8 3/4 3-11 3/4 3-13

! 3/4 3-34 3/4 3-35 3/4 3-36 3/4 3-37a 3/4 3-39 B3/4 3-2 8606040313 860523 fDR ADOCK 05000327 PDR aw s. -- _

1 T

UNIT 1 PROPOSED CilANGES

.,g

TABLE 3.3-1 (Continued)

N

@ REACTOR TRIP SYSTEM INSTRUMENTATION 9

E e

MINIMUM e TOTAL NO. CHANNELS CHANNELS APPLICABLE

}

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

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

20. Reacto'r Trip Breakers 2 1 2 1, 2, and

12;15

21. Automatic Trip Logic 2 1 2 1, 2, and

12 '

22. Reactor Trip System Interlocks A. Intermediate Range t' Neutron Flux P-6 2 1 2 2, and* 8a B. Power Range Neutron Y Flux - P-7 4 2 3 1 8b

C. Power Range Neutron t~

Flux - P-8 4 2 3 1 8c D. Pcwer Range Neutron Flux - P-10 4 2 3 1, 2 Bd E. Turbine Impulse Chamber E Pressure - P-13 2 1 2 1 8b l '

2

$ F. Power Range Neutron i . Flux - P-9 4 2 3 1 8e j

Mi C-G. Reactor Trip - P-4 2 1 2 1, 2, and* 14

, r .

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< C)

V) l

[ $

i i

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1

[ INSTRUMENTATION TABLE 3.3-1 (Continued)

TABLE NOTATION With the reactor trip system breakers in the closed position and the control

==

rod drive system capable of rod withdrawal, and fuel in the reactor vessel.

The channel (s) associated with the protective functions derived from the out of service Reactor Coolant Loop shall be placed in the tripped condition.

The provisions of Specification 3.0.4 are not applicable.

High voltage to detector may be de-energized above the P-6 (Block of Source Range Reactor Trip) setpoint.

ACTION STATEMENTS ACTION 1 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the reactor trip breakers.

ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and POWER OPERATION may proceed provided the following conditions are satisfied:

I( 3 k a. The inoperable channel is placed in the tripped condition within hours.

b. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to g urs for surveillance testing per Specification 4.3.1.1. 4

c. Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL and the Power Range, Neutron Flux high trip reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or, the QUADRANT POWER TILT RATIO is monitored at least or ce per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

d. The QUADRANT POWER TILT RATIO, as indicated by the remaining three detectors is verified consistent with the normalized symmetric power distribution obtained by using the movable incore detectors in the four pairs of symmetric thimble locations at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when THERMAL POWER is greater than 75% of RATED THERMAL POWER.

11 ? G Af 40 %

sys r.a s c v . w 6-SEQUOYAH - UNIT 1 3/4 3-5 ^ m, e.s . . ~. '2

- . ~ _ . ._

WI t

____._.__-_~

9

INSTRUMENTATION TABLE 3.3-1 (Continued) l ACTION 3 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

a. Below the P-6 (Block of Source Range Reactor Trip) setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint.

b. Above the P-6 (Block of Source Range Reactor Trip) setpoint, but below 5% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 5% of RATED THERMAL POWER.

c. Above 5% of RATED THERMAL POWER, POWER OPERATION may continue.

d. Above 10% of RATED THERMAL POWER, the provisions of Specification 3.0.3 are not applicable.

ACTION 4 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

a. Below the P-6 (Block of Source Range Reactor Trip) setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint.

b. Above the P-6 (Block of Source Range Reactor Trip) setpoint, operation may continue.

ACTION 5 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, verify compliance with the SHUTOOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed -

provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within hours.

b. The Minimum Channels OPERABLE requirement is met '

oneadditionalchannelmaybebypassedforupto[;however, hours for surveillance testing per Specification 4.3.1.1.1. Y ACTION 7 - With the number of OPERABLE channels one less thrn the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed until performance of the next required CHANNEL FUNCTIONAL TEST provided the inoperable channel is placed in the tripped condition within / hours.

SEQUOYAH - UNIT 1

[o 3/4 3-6

,,n.

-.-nn

(-

~

n .1

.~ - - -.

1 INSTRUMENTATION C' TABLE 3.3-1 (Continued)

ACTION 8 - With less than the Minimum Number of Channels OPERABLE, declare the interlock inoperable and verify that all affected channels '

of the functions listed below are OPERABLE or apply the appro-priate ACTION statement (s) for these functions. Functions to be evaluated are:

a. Scurce Range Reactor Trip

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

Undervoltage Underfrequency Pressurizer Low Pressure -

Pressurizer High Level

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

d. Reactor Trip Intermediate Range Low Power Range C e.

Source Range Reactor Trip Turbine Trip N LET's ACTION 9 -

h hannel associated with an operating loop inoperable, rest inoperable channel to OPERABLE status within s or be in DBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; howev ,o channel associated w an perating loop may be bypass o p to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surve an testing per Specif' i 4.3.1.1.1.

ACTION 10 - With one channel inoperab restor h noperable channel to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months sce THERMAL POWER to below the P-8 (Block Low Reactor ol mp low) setpoint breaker within the next 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . O ation b'el g P-8 (Block of Luw Reactor Coolant Pu setpoint breake ay ntinue pursuant to ACTION 11.

ACTION 11 - With 1 f 6n'the Minimum Number of Channels CPERABL , e 'on m jMnue provided the inoperable channel is placed in o

rJ ged condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 12 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for 2 up tor)' hour for surveillance testing per Specification 4.3.1.1.1 provided the other channel is OPERABLE.12

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m n n 6 y i v u._

SEQUOYAH - UNIT 1 3/4 3-7 f i.; r,emm . ; Nv. :'

w Ed

o

/

INSTRUMENIAilpf! '-

TABLE 3.3-1 (Continued)

ACTION 13 - With the number of OPERABLE channels one less than the Total Number of Channels and with the THERMAL POWER level above the P-7 (Block of Low Power Reactor Trips) setpoint, place the inoperable channel in the tripped condition within )Phour33 _0 operation may continue until performance of the next required CHANNEL FUNCTIONAL TEST.

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

A CTION IS w TH O H oi of THC D e VdRSE TRIP FEATURS S

( u rtog.g vourAGE oR.sHwi~ rRrP sqrrac HMGHT*)

IN OPERA BLC , DC S r o K G. ar To OPER /M u~. z StMus u i n+iN y 8 Now.f_ S oR o e c. c. +,c .c THC 756EAKst4 in oM/CA/$ us 4.ao f9.jo gg.

hcroo so IZ, 771%~ sf&iAW.A &A Lt- No r M BW%SS 60 w H en- E oh) e of TH4 D Msc TftP fa*Arwees is in opestn'3 t-E tikc.g w RiL up ro 4 Houses fait manoamang M As n re nnac.c. To 12q c rosec tyM rashis+/< scie Tu OrWi l Mw srarus. .

d

( .

I SEQUOYAH - UNIT 1 3/4 3-8 bEP1,! COO-e w

, l

= -

2 -

{ yc _ TABLE 4.3-1 P

S

~

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENIS

$. E

e c CHANNEL MODES IN WHICH 6' 5 ~

CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE l ] FUNCTIONAL UNIT CHECK CALIBRATION TEST

REQUIRED

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

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

Q and Q(6)

W

3. Power Range, Neutron Flux, N.A. R(6) 1, 2 Q

High Positive Rate a

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

High Negative Rate

-t'

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

Y Neutron Flux 5

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

4

7. Overtemperature Delta T S R M 1, 2

8. Overpower Delta T S R M 1, 2 l

i 9. Pressurizer Pressure--Low 5 R

) *

? Q 1, 2

!c U 10. Pressurizer Pressure--High S R Q 1, 2

-:e I

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

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

! tl 13. Loss of Flow - Two Loops S R N.A. I l

i

._i-w

INSTRUMENTATION TABLE 4.3-1 (Continued)

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

(1) -

If not performed in previous 7 days.

(2) -

Heat balance only, above 15% of RATED THERMAL POWER. Adjust channel if absolute difference greater than 2 percent.

(3) -

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

(4) -

Manual ESF functional input check every 18 months.

(5) - Each train or logic channel shall be tested at least every 62 days on a STAGGERED T TdL7- s ua w /N W N W ^#re-v4AigrggqQgST BASIS. ',W,'t Tf4Eorrac wooewa-r46 h~onum-dri(

(6) -

NutrondetectorsmaybeexcludedfromCHANNELCALIBRATION.

(7) - Below P-6 (Block of Source Range Reactor Trip) setpoint, lrk (8) -

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

(&~ 'T't+E c lM NHe s- S u ss w o* 4 ra s s- 3HeLu

/^' MAwo e st rL.y vwosy rds coPctA4*s-eq OP re-+c ONome,vsur4 GL po Sean TK u P Ciet-asVS Poit. 7 W C f // %ia nt. /2stncrvd.-

ggaP G odr*ON-lv$$h h D iddd SEQUOYAH - UNIT 1 3/4 3-13 " x r.J,.~ . . ; ~ . - d

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I M TABLE 4.3-2

! E i r S ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION E SURVEILLANCE REQUIREMENTS l 5 CHANNEL MODES IN WHICH

CHANNEL FUNCTIONAL SURVEILLANCE CHANNEL l ~

CALIBRATION TEST REQUIRED FUNCTIONAL UNIT CHECK

1. SAFETY INJECTION AND FEEDWATER ISOLATION

a. Manual Initiation N.A. N.A. '

1,2,3,4

b. Automatic Actuation Logic N.A. N.A. M% 1,2,3,4

c. Containment Pressure-High S R Q 1, 2, 3 w d. Pressurizer Pressure--Low S R Q 1, 2, 3

'a w e. Differential Pressure S R Q 1, 2, 3 d, Between Steam Lines--High a

f. Steam Flow in Two Steam S R Q 1, 2, 3 Lines--High Coincident with T,yg--Low-Low or Steam Line Pressure--Low k

l 2. CONTAINMENT SPRAY N.A. N.A. 1,2,3,4 1

{,r ,

a. Manual Initiation Automatic Actuation Logic N.A.

R[([/ 1,2,3,4

(! [{ b. N.A. M%

1 s o i

c. Containment Pressure--High-High S R Q 1,2,3 i i:s i S

' d '

1 r , .

' t,

a I

=

E TABLE 4.3-2 (Continued)

E s

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS

[E CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

3. CONTAINMENT ISOLATION -

a. Phase "A" Isolation

' g

1) Manual N.A. N.A.

)(% 1, 2, 3, 4

2) From Safety Injection N. A. N.A. M 1, 2, 3, 4 Automatic Actuation Logic

b. Phase "B" Isolation p w o 2 1) Manual N.A. N.A. 1, 2, 3, 4 w

a 2h. Automatic Actuation /

N.A. N.A. 1,2,3,4 m

Logic M%

3) Containment Pressure-- S R Q 1,2,3 High-High

c. Containment Ventilation Isolation g I 7

1) Manual N.A. N.A. 1, 2, 3, 4 p.-

[ E- 2) Automatic Isolation Logic N.A.

I h r-N.A. M([/ 1,2,3,4 f.J

! 3) Containment Gas Monitor S R H 1, 2, 3, 4 m .. Radioactivity-High f E!

L)

. B

,i, kk

M TABLE 4.3-2 (Continued) '

o S ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION E SURVEILLANCE REQUIREMENTS C

CHANNEL MODES IN WHICH

[ CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE

~

CHECK CALIBRATION TEST REQUIRED FUNCTIONAL UNIT M 1, 2, 3, 4

4) Containment Purge Air S R Exhaust Monitor Radio-activity-High .

R M 1, 2, 3, 4

5) Containment Particulate S Activity-High

4. STEAM LINE ISOLATION

a. Manual N.A. N.A.

([ 1, 2, 3 1,2,3 9

b. Automatic Actuation Logic N.A. N.A.

M([/

R Q 1, 2, 3 M c. Containment Pressure-- S High-High S R Q 1,2,3

d. Steam Flow in Two Steam Lines--High Coincident with T,yg-- Low-Low or Steam Line Pressure--Low ,

;( -

{ ! ?' 5. TURBINE TRIP AND FEEDWATER j .: ISOLATION

! ll S R Q 1,2,3 F -

a. Steam Generator Water

[c g : ; Level--High-High U

U 6. AUXILIARY FEEDWATER 1,2,3 a.

b.

Manual Automatic Actuation Logic N.A.

N.A.

R[([/

M(p 1, 2, 3 l*

O O O J

TABLE 4.'3-2 (Continued) .

l

[ $ ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION

,' ' Q SURVEILLANCE REQUIREMENTS

! E

' ' CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE

- E

q FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

8. ENGINEERED SAFETY FEATURE

ACTUATION SYSTEM INTERLOCKS

a. Pressurizer Pressure, N.A. R( N.A. 1, 2, 3 P-11

b. T,yg, P-12 N.A. R( N.A. 1, 2, 3

c. Steam Generator Level, P-14 N.A. R([ N.A. 1, 2

$ 9. AUTOMATIC SWITCHOVER TO

, CONTAINMENT SUMP i

y a. RSWT Level - Low S R H 1,2,3,4 8

COINCIDENT WITH

. Containment Sump Level - High S R M 1,2,3,4 AND Safety Injection (See 1 above for all Safety Injection Surveillance Requirements) b I

l

%l * .

t S i:.

e l

i.5.

TABLE 4.3-2 (Contfdaed) ..

TABLE NOTATION DForrt (U B n" I be tested at ame+ cas - 5 ~

i during shut own, .-"r - -

- an: manual safeguards actuatinn e M ' g T M "."' M--._.' easLange per '31 am -

[/J(/) Each train or logic channel shall be tested at least every 62 dayt; on a STAGGERED TEST BASIS.

Dtti.ert UT~ iim L.5- 7 JL'f_.0TIONAllE_ST shall include - u. 'r = .. U.er oy applying either a lar"L M __u G - -r ._ anr_ogriate side of the

- g m;rrse

) The total interlock function shall be, demonstrated OPERABLE during CHANNEL (2 CALIBRATION testing of each channel affected by intgrlock operation.

f P

l l

uAGAP *enn m n n 6 V 4U UC.

l SEQUOYAH - UNIT 1 3/4 3-38 .'_:; n.:.T.c . . t ." . M l

[ ..

. .._..__ 1,.....tdre4

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

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

Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or

2) utilizing replacement sensors with certified response times.

ZNSERT~ & Q<xr PAH;.C) 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 radiation levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation (

level trip setpoint is exceeded.

(

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

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

9 aH Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in 1

recalibrati6n of the excore neutron flux detection system, and full incore

, flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT l POWER TILT RATIO when one Power Range Channel is inoperable.

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 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 SEQUOYAH - UNIT 1 8 3/4 3-2 i

... a - --~ -c^

. = _ - . . _ . . - . . - - - .

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

0 t

UNIT 2 PROPOSED CHANGES i

.m. .

. _ - - .,rr., -- -

t

! TABLE 3.3-1 (Continued)

p REACTOR TRIP SYSTEM INSTRUMENTATION i Eo y

MINIMUM y TOTAL NO. CHAtlNELS CHANNELS FUNCTIONAL UllIT APPLICABLE OF CilANNELS TO TRIP OPERABLE MODES ACTION

c-x 19. Safety Injection Input r H from ESF 2 1 2 1, 2 j - m 12 j 20. Reactor Trip Breakers 2 2 4

l 1 1, 2, and

12,15

21. Automatic Trip Logic 2 i 1 2 1, 2, and

12

! 22. Reactor Trip System Interlocks j A. Intermediate Range if Neutron Flux, P-6 2 1 2 2, and*

Ba

] ,

B. Power Range Neutron Flux, P-7 4 2 3' w 1 8b

}1 1 w C. Power Range Neutron Flux, P-8 4 2 3

E 1 8c j D. Power Range Neutron j Flux, P-10 4 2 3 3

1, 2 8d E. Turbine Impulse Chamber j Pressure, P-13 2 1 2 1 8b F. Pr.wcr Range Neutron

{f i

Flux, P-9 4 2 3 1 8e 4 t-G. Reactor Trip, P-4 2 1 2 1, 2, and *

\t.

14 ull

}

7

6 m

J -

TABLE 3.3-1 (Continued)

TABLE NOTATION With the reactor trip system breakers in the closed position, the control rod drive system capable of rod withdrawal, and fuel in the reactor vessel.

an The channel (s) associated with the protective functions derived from the out of service Reactor Coolant Loop shall be placed in the tripped condition.

1. The provisions of Specification 3.0.4 are not applicable.

High voltage Range Reactor toTrip) detector may be de energized above the P-6 (Block of Source setpoint.

ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than required b the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STAN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the reactor trip breakers.

ACTION 2 - With the number of OPERABLE channels one less than th Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

(~ ' a.

The inoperable channel is placed in the tripped condition hour.S.

within /(a b.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to f hours for surveillance testing per Specification 4.3.1.1. .

c.

Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER and the Power Range, Neutron Flux trip setpoint is reduced to less than or e 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; theor, qual to QUADRANT 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . POWER TILT RATIO is monitored at least once per d.

The QUADRANT POWER TILT RATIO, as indicated by the remaining three detectors, is verified consistent with the normalized symmetric power distribution obtained by using the movable incore detectors in the four pairs of symmetric thimble locations at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when THERMAL POWER is greater than 75% of RATED THERMAL POWER.

{ A- . . ' .a 2 3/13/31 SEQUOYAH - UNIT 2 3/4 3-5 t

. _. _; - a

_ TABLE 3.3-1 (Continued)

ACTION 3 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL POWER level: -

a.

Below the P-6 (Block of Source Range Reactor Trip) setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint.

b.

Above the P-6 (Block of Source Range Reactor Trip) setpoint, but below 5% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 5% of RATED THERMAL POWER.

c. Above 5% of RATED THERMAL POWER, POWER OPERATION may continue.

d. Above 10% of RATED THERMAL POWER, the provisions of Specification 3.0.3 are not applicable.

ACTION 4 - With the number of OPERABLE channels one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

a.

Below the P-6 (Block of Source Range Reactor Trip) setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint. {'

b. Above the P-6 (Block of Source Range Reactor Trip) setpoint, operation may continue.

ACTION 5 - With the number of OPERABLE channels one less than required by the Minimum Channels OPERABLE requirement, verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the.following conditions are satisfied:

a. The inope,rable channel is placed in the tripped condition within / hour.5,

b. The Minimum Channels OPERABLE requirement is met; however, one additionaltesting for surveillance channel may be bypassed per Specification 4.3.1.1.1. for up to t h (s ACTION 7 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed until performance of the next required CHANNEL FUNCTIONAL TEST .

provided the inoperable channel is placed in the tripped condition within / hourJ. D_

4 SEQUOYAH - UNIT 2 3/4 3-6 l- A" 2 2 t .':".

~ ' - - - # i

TABLE 3.3-1 (Continued) f- ACTION 8 - With less than the Minimum Number of Channels OPERABLE, declare

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

a. Source Range Reactor Trip.

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

Undervoltage '

Underfrequency Pressurizer Lcw Pressure Pressurizer High level

c. Reactor Trip Low Reactor Coolant Loop F, low (1 loop)

d. Reactor Trip Intermediate Range Low Power Range Source Range

(, e. Reactor Trip Turbine Trip pugrg ACTION 9 - channel associated with an operating loop inoperabl resto inoperable channel to OPERABLE status with' urs or be in R channel assocTs e NDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; howe e ith an operating loop may ssed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for su 1 nce testing per Sp ' ation 4.3.1.1.1.

ACTION 10 - With one channel inoperab tor inoperable channel to OPERABLE status within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months r uce THERMAL POWER to below the P-8 (Block of Low Reac o t mp Flow) setpoint, I within the next 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . eration b ow e P-8 (Block of Low Reactor Coolant Pu ,1 ) setpoint, may h t e pursuant to ACTION 11.

/

ACTION 11 - With 1 55 an the Minimum Number of Channels OPERABL , tion m 3n inue provided the inoperable channel is placed in '

I rXped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

l l ACTION 12 - With the number of OPERABLE channels one less than required by the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for i

up top / hour for surveillance testing per Specification 4.3.1.1.1

( provided the other channel is OPERABLE.

' c d;;..: 2

.,, in,

/p a si v &

SEQUOYAH - UNIT 2 3/4 3-7 L

. . . _. . - . . - - M

i TABLE 3.3-1 (Continued) ['

i' .

ACTION 13 - With the number of OPERABLE channels one less than the Total Number of Channels and with the THERMAL POWER level above the P-7 (enable reactor trips) setpoint place the inoperable channel in the tripped condition within /4 ours h operation y may continue until performance of the next required CHANNEL FUNCTIONAL TEST.

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

peg I.5 - W ors + ooK c/= rH4~ Di vd/aS E Ytt/P pggur2CS

, ( Ls N O</C Vo ' rA4 % CA S H+w r YK-s,* A m +c.H p.cd t-)

liv ot'dA'o96 L*C , /2s[s ro /2 < 11~ TD DNAb80~(i GrArtsS w m+m 4 6 poats, orz Dicci Mg rac Alc arta (Al oth[sl Adi-G s4WO RYALj kOI70A lE , T~/-/G Wstand sanu- xor /3s seyssssgo wi+,uc ONG o t= 'rHE Di vga se rR *P f~ dart <scs is IN oPGAA4 Lot Gxever FOR uf 70 4 How25 I'0A ;

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SEQUOYAH - UNIT 2 3/4 3-8 ,, _

.I

f. -

v I: -

m g TABLE 4.3-1 i

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS

. i CHANNEL MODES FOR WHICH

[ g CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS t' q FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

1. Manual Reactor Trip N.A. N.A. S/U(1) Ago R(9) 1, 2, and

y

{ 2. Power Range, Neutron Flux 5 D(2), M(3)

Q 1, 2 i

! and Q(6) ,

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

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

1, 2, and

w Neutron Flux

6. Source Range, Neutron Flux S(7)

, R(6) M and S/U(1) 2, 3, 4, 5, and *

7. Overtemperature AT S R M 1, 2

8. Overpower AT S R M -

1, 2

. 9. Pressurizer Pressure--Low S R

. - Q 1, 2 s 4

i; i

, [p 10. Pressurizer Pressure--High S R Q 1, 2 i i j[.E 11. Pressurizer Water Level--High S R Q 1, 2 I f((

.o

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

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

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

i

INSTRUMENTATION lL TABLE 4.3-1 (Continued)

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

(1) -

If not performed in previous 7 days.

(2) -

Heat balance only, above 15% of RATED THERMAL POWER. Adjust channel if absolute difference greater than 2 percent.

(3) -

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

(4) -

Manual ESF functional input check every 18 months.

(5) -

Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST 8 ASIS. THC ras 7- 2+ns a '^'04^t~Denr v' Vea g,y, age,rv oc rac t.woexvos-r-sac Ano Auro"m s'nuar sfy (6) -

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(7) -

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

(8) -

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

(9) - 7?M CHAsvHEL fu M t Yled T;ES T S H A L.E.

o

/N OTAENrt.y VE)2W1 TUC OtMAs mst frf of nu wicMAv'ol RAG <

l}N o s H H N r rY2tP Cisteutr$ fort rg ffdNugt.

pcac.ron. pt.is" f u N e-rioiv.

l SEQUOYAH - UNIT 2 3/4 3-13

, . _ . . _ _ . _ -- J=i

~:

-l]. -

p-

. A O ,

h .

1

> us g TABLE 4.3-2 kE

,' g ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION

' 2 SURVEILLANCE REQUIRENENTS L

g , CHANNEL MDDES FOR WHICH I: CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS

[, q , FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

1. SAFETY INJECTION AND FEE 0 WATER ISOLATION

a. Manual Initiation N.A. N.A ([ 1, 2, 3, 4 k}l- b. Automatic Actuation Logic N.A. N.A.

M([l 1, 2, 3, 4

'T

\ c. Containment Pressure-High S R Q 1, 2, 3

! y d. Pressurizer Pressure--Low S R Q 1,2,3 n * '

y e. Differential Pressure S R Q 1, 2, 3

y Between Steam Lines--High r f. Steam Flow in Two Steam S R 1,2,3 Q

s Lines--High Coincident with t

T avq --Low-Low or Steam Line

.. Pressure--Low .

t :

2. CONTAINHENT SPeY

f ,

a. Manual Initiation N.A. N.A

(/) 1, 2, 3, 4 i -

! t:-

,' g b. Automatic Actuation Logic N.A. N.A. M(/)/ 1, 2, 3, 4

, . . r.

iat I g c. Containment Pressure--High-hlgh S R Q 1,2,3 3 c3 o

t" (%

. ~

i- s Q.

y .

.j M TABLE 4.3-2 (Continued) 4 C T." S ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION

j. $ SURVEILLANCE REQUIREMENTS CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS FUNCTIONAL UNIT CHECK CALIBRATION TEST

" REQUIRED z

$ 3. CONTAINMENT ISOLATION 3 a. Phase "A" Isolation

1) Manual N.A. N.A. (d f 1, 2, 3, 4

[ 2) From Safety Injection N.A. N.A.

M(/)/ 1, 2, 3, 4 Automatic Actuation Logic

b. Phase "B" Isolation

w u

1) Manual N.A. N.A. }{

[( /) '

1, 2, 3, 4 f h 2) Automatic Actuation Logic H.A. N.A. M(/)/ 1, 3, 3, 4

3) Containment Pressure-- S R Q 1,2,3 High-High

c. Containment Ventilation Isolation

1) Manual N.A. N.A. /(/) 1, 2, 3, 4

2) Automatic Isolation Logic

, , i f N.A. N.A. M(/)/ 1, 2, 3, 4 1 ; e

~

IF 3) Containment Gas Monitor S R H 1,2,3,4

Radioactivity-High

.- = -

I h ' 'c$

3" i

e

g TABLE 4.3-2 (Continued) 0

.S ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION

$ SURVEILLANCE REQUIREMENTS e

e CHANNEL 2 MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS FUNCTIONAL UNIT

] CHECK CALIBRATION TEST REQUIRED

4) Containment Purge Air S R M 1,2,3,4 Exhaust Monitor Radio-activity-High ,

5) Containment Particulate S R M 1,2,3,4 Activity-High

4. STEAM LINE ISOLATION R+

a. Manual N.A. N.A. (% 1,2,3 Y b. Automatic Actuation Logic N.A. N.A. M%, 1,2,3

c. Containment Pressure-- S R Q 1,2,3 High-High

d. Steam Flow in Two Steam S R 1,2,3 Q

Lines--High Coincident with T,yg -- Low-Low or Steam Line Pressure--Low

a i

!I i M S. TURBINE TRIP AND FEEDWATER j ISOLATION y - *A r "-

a. Steam-Generator Water S R 1,2,3 Q

-! FG Level--High-High

x. .

3

I 6. AUXILIARY FEEDWATER

] x g

! A'N- % ual N.A. N.A.

)((/) 1, 2, 3

b. Automatic Ach;= tion Logic N.A. N.A.

H(/) 1, 2, 3 E

TABLE 4.3-2 (Continued) m

@ ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION S SURVEILLANCE REQUIREMENTS E

' CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE .

E TEST REQUIRED p FUNCTIONAL UNIT CHECK CALIBRATION ro

8. ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INTERLOCKS

a. Pressurizer Pressure, N.A. R(QL) N.A. 1, 2, 3 P-11

b. T,yg, P-12 N.A. R(M) N.A. 1, 2, 3

c. Steam Generator N.A. N.A. 1, 2 R (f)(4)

Level, P-14 I

i

$ 9. AUTOMATIC SWITCHOVER T0 CONTAINMENT SUMP w

a. RSWT Level - Low S R M 1,2,3,4 COINCIDENT WITH Containment Sump Level - High S R M 1,2,3,4 l AND j Safety Injection (See 1 above for all Safety Injection Surveillance Requirements)

' l-2E[p -

'I c il l I :

I

O TABLE 4.3-2 (Continued)

TABLE NOTATION M rK (1) Manua a hes shall be tested at 1 months

~ -

during shutdown. A . c' a ed with manual safeguards actuation NEL at least once per

(/)% Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST BASIS. -

Tvcccc DD:E:- applying either a L TEST shall include ex h G iEF by riu appropriate side of the (2)(/) The total interlock function shall be demonstrated OPERABLE during CHANNEL CALIBRATION testing of each channel affected by interlock operation.

L' SEQUOYAH - UNIT 2 3/4 3-39

-~

_ L___--- -"A

t INSTRUMENTATION BASES Ci REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM IN5TRUMENTATION (Continueo)

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

Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or

2) utilizin replacement sensors with certified response times.

.ZNSEAT (mr- Foss.)

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

{

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

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

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

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

April 1974. -

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

. rl~ '

INSERT B 4

U-2, SEQU3YAH NUCLEAR PLANT, PAGE B 3/4 3-2 Action 15 of Table 3.3-1, Reactor Trip System Instrumentation, allows the breaker to be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for the purpose of performing maintenance. The a hours is based on a Westinghouse analysis performed in i WCAP 10271, Supplement 1 which determines bypass breaker availability.

1 4

I i

l I

4 l --a- -- 4

t ENCLOSURE 2 PROPOSED TECHNICAL SPECIFICATION CHANGES SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 (TVA SQN TS 67)

Justification to add requirements to LCO 3.3.1.1 (unit 1) and LCO 3.3.1 (unit 2) technical specifications for operability of the new shunt trip attachment and also incorporate relief of survelliance and outage times as allowed by Westinghouse Standard Technical Specifications for applicable reactor protection system instrumentation and engineered safety features manual actuation circuitry.

l I

l 4

t d

I

=. - . . . , sa , ,e s

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

Description of Chango Add requirement to LC0 3.3.1.1 (unit 1) and LCO 3.3.1 (unit 2) technical specifications for operability of the new shunt trip attachment to the reactor trip breakers. Further additions are for surveillance of these attachments and consequential inoperability of the trip breakers. The surveillance and outage times also apply to the bypass breakers and inoperable channel conditions.

Additionally, this change will extend the surveillance interval for the functional testing of the engineered safety features (ESF) manual actuation channels from monthly to once per cycle (during refueling).

This change will also delete action statements 9, 10, and 11 of Table 3.3-1, Reactor Trip System Instrumentation, and the original Note 3 of Table 4.3-2, ESF Actuation Instrumentation Surveillance Requirements. Finally, clarification for action 15 of Table 3.3-1 is being added to section 3/4.3.1 and 3/4.3.2 of the bases. (See Enclosure 1)

Reason for Change Item 4.3 of Generic Letter 83-28, " Required Actions Based on Generic i .

Impilcations of Salem Anticipated Transient Without Scram (ATWS) Events,"

l established NRC requirements (Generic Letter 85-09) for automatic actuation of the shunt trip attachment for Westinghouse plants. NRC concluded that technical specification changes should be proposed to explicitly require independont testing of the undervoltage (UV) and shunt trip attachments during 1

power operation and independent testing of the control room manual switch contacts during each refueling outage. These tests are necessary to ensure

! reliable reactor trip breaker (RTB) operation.

_n

s .

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

Also, times currently allowed by technical specifications for testing and maintenance of reactor protection system (RPS) instrumentation are insufficient to accomplish the necessary work. It is proposed that allowable test times specified for RPS instrumentation be extended from 2 to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and that maintenance times be extended to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . .For analog channels, the i first 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months of maintenance would be accomplished with the channel in an contripped condition; thereafter, the channel must be tripped. The proposed test and maintenance times have been identified by WCAP-10271, " Evaluation of Surveillance Frequencies and Out-of-Service Times for the Reactor Protection Instrumentation System," as being more representative of actual conditions.

. NRC has also required adding surveillance testing on the bypass breakers.

Based on current Westinghouse Owners Group (WOG) calculations of the reactor trip system unavailability, there is an insignificant reliability improvement from including periodic surveillance tests of the bypass breakers in the technical specifications. Given the minimal impact of bypass breaker testing, it is recommended that SQN be allowed to omit this requirement as proposed in Generic Letter 85-09.

SQN technical specifications presently require monthly functional testing of the ESF manual actuation channels. Industry experience has demonstrated that monthly functional testing of manual actuation circuitry was not optimal. The requirement forces testing in operational modes which are not

l

.l

' 0 1

l i ..

2 conducive to the testing. This requires unusual system configurations which ,

I are difficult to control and can easily lead to spurious ESF actuations. SQN is requesting that this functional testing only be required during each refueling outage.

Note 3, Table 4.3-2, and action 9, 10, and 11. Table 3.3.-1 are being deleted because they do not apply to any item in the tables. Review of the original technical specifications show no applicability for note 3 and .

i actions 10 and 11, and action 9 applies only to 31 1oop operations. In order 4

to avoid undue confusion, SQN is requesting this unnecessary information be, g ;

. o.1 i .. i .a. . r u.m t. i . .i n a

. e .

tn v n , . . .. ,, s.. .

w. n ,..r,u.a, .o a nu,scu . c :,; . . .w . . , .. _

. . i Justification for Change

, , , ca.:.c. : n i . ;. - . . . . i .. . . . . i. u , .ne eu 5:.s; t.: v . ,.

j The auto shunt trip feature will upgrade the RPS to automatically trip the i .

RTD in a similar manner that is acc.omplished by the UV output circuit.

a ' "

j

'Tha UV Butput circuit'is designe'd so that in the absence of a reactor trip signE1'froim the universal logic boirds or"s' witched' inputs Darlington Pair l 9.. ,

j Transistors in the UV output circuit will conduct' current and energize the UV trip coil of the associated RTB.

A' reactor Efip input will result in the l turning off of the pair transistors, thereby interrupting the current flow to l

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

Similarly, with the addition of the automatic shunt trip feature, a reactor

trip input will close the contacts on the shunt relay in the shunt trip coil i

l circuit. This energizes the shunt trip coil and trips the same RTB by a '

i

, diverse mechanism.

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

shunt and UV trip features of the RTBs as part of a single sequential test 4

i .

1 t

-= - . . _ _ . . _ _ _ -

4 3

1 procedure. Therefore, the survel'11ance test whleh identifies a failure of one i diverse trip feature also confirms the operability of the other trip feature.

As a consequence, there is a high degreo of confidence that the operable trip feature will be capable of initiating a reactor trip in the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Accordingly, an additional action statement will be included in the technical specifications for the RTBs to permit continued plant operation for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months with one of the diverse. trip features inoperable before further action needs to be taken. If unable to restore operability within the 48-hour timo requirement, the breaker must be declared inoperable and the plant taken to hot standby within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . During this 48-hour time period, th9 breaker can' only be bypassett for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for the purpose of performing maintenance.

This surveillance testing and additional action statement meets the request by NRC as proposed by item 4.3 of Generic Letter 83-28.

The surveillance testing for the automatic shunt trip attachment.will take an additional 10 minutes above the 40 minutes it presently takes to complete the test. This'40-minute time period does not allow for any delays that may be encountered during testing. Should the logic test fall to meet its acceptance criteria, the stringent one-hour does not allow sufficient time to evaluate the problem due to the complex electronics involved with the SSPS.

There have been three potential reportable occurrences (PRO) written as a result of exceeding the one-hour time limit. SQN is requesting this time I interval for survoillance testing of the RTBs be extended to two hours,

, provided the other breaker and associated channel is operable. Current Westinghouse Standard Technical Specifications (STS) allow two houro for this J .

f.e s t i n g ,

f

- . _ . _ . . . . . - - - _ _ _ . - - , . , . . - _ _ . - , . - -. - -_ ?- _... . l_ ._-. L . --- - - , . _ .

Weatinghouse STS will also incorporate extended testing and maintenance times for RPS instrumentation as proposed by WCAP-10271 and approved by NRC.

! These changes will allow:

I 1

1.

An inoperable channel to be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for survelliance testing provided that the minimum operable channel he cr* .1 - requirement is met (SQN technical specifications actions 2 and 6)...

r, c s c* tc - - -' - c~n. . - .. m- x i

r.~.2. Six hours before an inoperable channel must be placed in the tripped ho. . - condition when the minimum number of channel requirements is violated

""' (SQN technical specifications actions 2, 6, 7, and 13). .. ,

n , ,

Uc The evaluations of the impact of implementing the proposed technical specifications revisions that have been presented in WCAP-10271 and supplement 1 provide a comprehensive analysis of the RPS. The reliability of the RPS remlins high. Plant safety is maintained or improved; plant availability is improved; and the burden to the plant caused by technical

~

specifications compliance is significantly reduced.

The genoric letter also required periodic independent testing of the i

control room manual switch contacts during oach refueling outage. The procedures for testing the breakers will be revised to require voltage measurements in the RTB cabinets to verify operability of the reactor trip switch contacts and wiring to the circuit breakers. Precautions will be taken to ensure that the " block auto shunt trip" switch will be used to preclude sensing the appilcation of power to the shunt trip coil via the automatic shunt trip feature. Additionally, with the breaker in a tripped condition, voltage will be measured across the conbination of the shunt trip coil and

- - , , -.-,,--r w- - - , - - . - - , - - - . , , - - . - - - - _ - - - , , , , _ _ - -. -

, - - . - - --r-. +-..w. --- - _ - - - _ - , , w--,p: w -

I series "a" auxiliary contact, due to the presence of the breaker closed position status light located in parallel with the normally open manual i reactor trip switch contacts. This indicating light will be removed to avoid l

ambiguous voltage measurements.

Finally, the generic letter required periodic surveillance testing of the bypass breakers. SQN plant surveillance instructions will require that the bypass breakers be tested prior to being..put.in. service. However, this is not, 4

j a technical specification requirement. WOG has calculated the impact of the I bypass breaker failure probability on,the reactor trip system failure .. c probability and concludes that the bypass breaker contribution is '-

' cd!

insignificant. .These calculations are based on the trip breaker fault tree model presented ~1n supplement 1 to WCAP-10271.

In WOG Letter'No.> 0G-106s which' transmitted the; WOG response to NRC

! questions on WCAP-10271, a typical Westinghouse pressurized water reactor (PWR) trip unavailability is estimated to be 1.5 E -5. No credit was taken I

for operati'on of the bypass breaker in the evaluation from which these calculations were derived. ..The impact on the reactor trip system t

! unavailability, including the unavailability of the reactor trip bypass breakers, was calculated with the following results.

i l

1. The bypass breakers are placed in service only when one train of the 1 RPS is in test or for the purpose of performing maintenance to restore the breaker to an operable status (up to four hours). The only i

circumstance in which the bypass breaker could affect RPS l

unavailability is the cutset when one train is in test or during i

l maintenance. (A signal is generated in the operable redundant train i

j and the main breaker fails to open.)

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2. The unavailability of the RPS attributable to failure of a main trip breaker with the opposite train in test or during maintenance is 3.7 E -7 or 2.5 percent of the total RPS unavailability (i.e., 1.5 E -5).

This cutset constitutes the only configuration in which the bypass breaker can affect RPS unavailabilty.

3. Taking credit for the bypass breaker would reduce the probability

m. i . . n . ; , . . . . . . . a.- 6. u . . .m ...u. .. , , ,, , , , , n,,,

value of this cutset to a r:-h: : . e c. t - .

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no (3.7 E -7)(3.5 E -4) = 1.3 Ea -10 where 3.5 E-4 is the unavailability of the bypass breaker assuming

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. bimonthly testing

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where 3.5 E -3 is the unavailability of the bypass breaker assuming testing on an 18-month interval.

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Based on the above, there is an insignificant reliability improvement from I

including periodic surveillance tests of the bypass breakers in the technical in:. . .. . . . .n .

1

, specifications. Testing the bypass breakers on a 2-month or 18-month test

' interval will result in an E -9 or E -10 level contribution to the RPS 9

unavailability of approximately E -5. Alternatively, the RPS unavailability increase that occurs by increasing the bypass failure probability from O percent to 100 percent is only 2.5 percent at the RPS level.

Given the minimal impact of bypass breaker testing, it is recommended that l

the proposed requirement in Generic Letter 85-09 to test the bypass breakers prior to the main breaker periodic surveillance test be deleted.

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Westinghouse STS revision 5 was revised to allow functional testing of the ESF manual actuation channels to be performed at each refueling outage when unusual system configuration can be more safely accommodated. These revisions were made to preclude spurious ESF actuations that could develop from the unusual configuration.

The ESF manual actuation circuitry consists of W-2 handswitches, terminal blocks, and associated control cables to the relays. These components reside in the seismic category I. Control Building. The ambient environment is mild and well controlled which should preclude any unusual climate induced degradation. The SQN maintenance history. records were reviewed, and there was no indication of time dependent reliability reductions. The license event report (LER) files were reviewed also, and they provided independent verification that the manual. actuation circuit is reliable.

This change proposal would bring SQN technical specifications into i conformance with the currently approved revision of the Westinghouse STSs for ESF manual, actuation circuitry functional test requirements. Setting the testing interval at least once per 18 months allows the manual channel ;

! functional test to be performed during refueling outages. At cold shutdown '

I i this test can be more safely and easily accomplished. !

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{ . . . . . . . _ . . . - - YD

, e ENCLOSURE 3 TVA SQN TS 67 PROPOSED TECHNICAL SPECIFICATION CHANCES SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 Determination of no significant hazards considerations to add requirements to LCO 3.3.1.1 (unit 1) and LCO 3.3.1 (unit 2) technical specifications for operability of the now shunt trip attachment and also incorporate relief of surveillance and outage times as allowed by Westinghouse Standard Technical Specifications for applicable reactor protection system instrumentation and engineered safety features manual actuation circuitry.

Sir.nificant Hazards Considerations

1. Is the probability of an occurrence or the consequences of an accident previously evaluated in the safety analysis report significantly increased?

No. The RPS is fundamental to plant safety because all transient and accident analysis are predicted on the basis that the RPS operate to terminate the energy released by the fission process. If the reactor fails to shut down following a transient or accident, an ATWS will exist. This change is intended to further minimize the possibility that

, an ATWS will occur. The additional testing requirements will not preclude any present testing procedures. The RTus and UV attachment will be subject to the same testing and perform as originally intended. The additional testing and maintenance times and relief of functional testing of ESF manual actuation channels do not alter the manner in which

!' protection is afforded. W' JAP-10271, supplement 1 evaluates the impact of increasing surveillance intervals on hardware failure rates based on various failure mechanisms and the nature of the periodic testing being performed. The evaluation ahows the not effect of the reduced testing is expected to be a decrease in f ailures, hence, an increase in equipment reliability.

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

No. The intent of the proposed change is to increase the reliability of the RpS. The auto shunt trip attachment will meet all electrical and phyoical separation requiremonta as defined by IEEE 279-1971 and will be s41smically and environmentally quallfled. Surveillance testing and maintenance activities are still performed in a manner to ensure equipment reliability. Also, plant surveillance instructions will be changed to require that the bypass breakers be physically tested before being put in service. The propored changes do not result in a change in the manner in which the RPS and EOF provido plant protection; therefore, this change will not create the possibility of occurrence of a new or different type accident.

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3. Is the margin of safety significant reduced?

No. The proposed changes are expected to ircrease the overail margin of safety. The independent testing of the UV and shunt, trip httachment ensures that each RTB can be automatically tripped'by two diverse mechanisms. The reliability of manual trip from the control room is retained by the periodic testing of the switch contacts, and the bypass breakers are ensured to be operable before service. The increased maintenance times are expected to improve overall safety with higher quality repairs leading to improved equiprent reliability. By increasing ~

surveillance times, plant safety;is maintained or' improved along with plant availability. The proposed changes do not significantly alter the manner in which safety limits, limiting safety system setpoints, or limiting conditions for operation are determined.

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