Text

Proposed Tech Spec 3/4.3.1, Reactor Trip Sys Instrumentation, Modifying Surveillance Frequencies & out-of-service & Test Times for Selected Functional Units
	ML20210F676
Person / Time
Site:	North Anna
Issue date:	01/30/1987
From:	; VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	;
Shared Package
ML20210F657	List: ML20210F653; ML20210F676; ML20210F682;
References
	NUDOCS 8702110135
	Download: ML20210F676 (29)
	v • d • e

-

ATTACHMENT 1 Proposed Technical Specifications Changes for North Anna Unit I P

... -. ~

z TABLE 3.3-1

@d

[

REACTOR TRIP SYSTEM INSTRUMENTATION 5>

MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE y

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES

' ACTION 1.

Manual Reactor Trip 2

1 2

1, 2:and

12 2.

Power Range, Neutron Flux 4

2 3

1, 2 2#

3.

Power Range, Neutron Flux 4

2 3

1, 2 2#

High Positive Rate 4.

Power Range, Neutron Flux 4

2 3

1, 2 2i High Negative Rate 5.

Intermediate Range, Neutron Flux 2

1 2

1, 2 and

3 w

N 6.

Source Range, Neutron Flux w

,8 A.

Startup 2

1 2

2##

4 3

B.

Shutdewn 2

1 2

'3*, 4* and 5*

15 C.

Shutdown-2 0

1 3, 4 and.5 5

7.

Overtemperature AT Three Loop Operation 3

2 2

1, 2 6#'

Two Loop Operation 3

1**

2 1, 2 9.

5 TABLE 3.3-1 (Continued)

Ei*

REACTOR TRIP SYSTEM INSTRUMENTATION E

MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICAELE-c:3 FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION a

8.

Overpower AT 6f~l Three Loop Operation 3

2 2

1, 2 -

Two Loop Operation 3

1**

2 1, 2 9

9.

Pressurizer Pressure-Low 3

2 2

1, 2 -

7f 10.

Pressurizer Pressure-High 3

2 2

1, 2 7f-11.

Pressurizer Water Level-High 3

2 2

1, 2 6f 12.

Loss of Flow - Single Loop 3/ loop 2/ loop in 2/ loop in 1

6f (Above P-8) any oper-each oper-w is ating loop ating loop u

13. Loss of Flow - Two Loops 3/ loop 2/ loop in 2/ loop in 1

6f (Above P-7 and below P-8) two oper-each oper-ating loops ating loop 14.

Steam Generator Water 3/ loop 2/ loop in 2/ loop in 1, 2 7f-Level--Low-Low any oper-each oper-ating loops ating loop-15.

Steam /Feedwater Flow 2/ loop-level 1/ loop-level 1/ loop 1, 2 7f Mismatch and Low Steam and coincident level and Generator Water Level 2/ loop-flow with

_2/ loop-flow mismatch 1/ loop-flow mismatch or mismatch in 2/ loop-level same loop and 1/ loop-flow mismatch

TABLE 3.3-1 (Continued) g

=g REACTOR TRIP SYSTEM INSTRUMENTATION 5>

MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTIONo G

16. Undervoltage-Reactor Coolant Pump Busses 3-1/ bus 2

2 1

7#

17.

Underfrequency-Reactor Coolant 3-1/ bus 2

2 1

6fl Pump Busses 18.

Turbine Trip A.

Low Auto Stop Oil Pressure 3

2 2

1 6#

B.

Turbine Step Valve Closure 4

4 4

1 16f 19.

Safety Injection Input from ESF 2

1 2

1, 2 1

R*

20. Reactor Coolant Pump Breaker y

Position Trip A.

Above P-8 1/ breaker

'l 1/ breaker 1

10 B.

Above P-7 1/ breaker 2

1/ breaker 1

11-per oper-ating loop 21.

A.

Reactor Trip Breakers 2

1 2

1, 2.

1, 14 2

1 2

3*,4*,5*

15 B.

Reactor Trip Bypass Breakers 2

1 2

13 22.

Automatic Trip Logic 2

1 2

1, 2 1

2 1

2 3*,4*,5*

15-

- 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.

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

- - -With-the Reactor Trip Breaker open for surveillance testing in accordance with Specification-Table 4.3-1 (item 21A).

9The_ provisions of Specification 3.0.4 are not applicable.

High voltage to detector may be de-energized above P-6.

ACTION STATEMENTS ACTION 1 -

.With the number of channels OPERABLE one less than required by the Ifinimum Channels OPERABLE requirement, be in 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 up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1.1 provided the other channel is operable.

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:

a.

The inoperable channel is placed in the tripped condition l

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months l

for surveillance testing of the redundant channel (s) per Specification 4.3.1.1.1.

c.

Either, THERMAL POWER is restricted to s 75% of RATED THERMAL and the Power Range, Neutron Flux trip setpoint is reduced to s 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 once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

d.

The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75 percent of RATED THERMAL POWER with one Power Range Channel inoperable by using the moveable incore detectors to confirm that the normalized symmetric power distribution, obtained from 2 sets of 4 symmetric thimble locations or a full-core flux map, is consistent with the indicated QUADRANT POWER TILT RATIO at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 3 -

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

NORTH ANNA - UNIT 1 3/4 3-5

TABLE 3.3-1 (Continued) a.

Below P-6, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint.

b.

Above P-6 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.

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 P-6, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint.

b.

Above P-6, operation.may continue.

ACTION 5 -

With the number of channels OPERABLE one less than required by the Minimum Channelo 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 inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and b.

The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4.3.1.1.1.

ACTION 7 -

With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and POWER OPERATION may proceed until performance of the next required CHANNEL FUNCTIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 8 -

Not applicable l

l l

NORTH ANNA - UNIT 1 3/4 3-6

F TABLE 3.3-1 (Continued)

ACTION 9 -

With a channel associated with an operating loop inoperable, restore the inoperable channel to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 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 ; however, one channel associated with an operating loop may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1.1.

ACTION 10 -

With one channel inoperable, restore the inoperable channel to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to below P-8 within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Operation below P-8 may continue pursuant to ACTION 11.

. ACTION 11 -

With less than the Minimum Number of Channels OPERABLE, operation may continue provided the inoperable channel is placed in the tripped 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, 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 13 -

With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within (1) hour or terminate testing of the Reactor Trip Breaker and open the Reactor Trip Bypass Breaker.

ACTION 14 -

With one of the diverse trip features (undervoltage or shunt trip device) inoperable, restore it to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the breaker inoperable and apply Action 1.

The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time required for performing maintenance to restore the breaker to OPERABLE status.

ACTION 15 -

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 open the reactor trip breakers within the next hour.

ACTION 16 -

With the number of OPERABLE channels less than the Total Number of Channels, Operation may continue provided the inoperable channels are placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

t NORTH ANNA - UNIT 1 3/4 3-7

TABLE 4.3-1 z

O REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENT 3 E

R CHANNEL

_ MODES IN WHICH g

CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED e

5 H

1.

Manual Reactor Trip E.A.

N.A.

R(8) 1,2 and

2.

Power Range, Neutron Flux a.

High Setpoint S

D(2), M(3) and Q(6)

Q(7) 1, 2 b.

Low Setpoint S

R(6)

S/U(1) 1***,

2 3.

Power Range, Neutron Flux.

N.A.

R(6)

Q(7) 1, 2 High Positive Rate 4.

Power Range, Neutron Flux, N.A.

R(6)

Q(7) 1, 2

.High Negative Rate 5.

Intermediate Range, S

R(6)

S/U(1) 1***, 2 and

wy Neutron Flux u

,L 6.

Source Range, Neutron Flux N.A.

R(6)

Q(7), S/U(1) 2**, 3, 4 and 5 w

7.

Overtemperature AT S

R(6)

Q(7) 1, 2 8.

Overpower AT S

R(6)

Q(7) 1, 2 9.

Pressurizer Pressure--Low S

R M

1, 2 10.

Pressurizer Pressure--High S

R M

1, 2 11.

Pressurizer Water Level--High S

R Q(7) 1, 2 12.

Loss of Flow - Single Loop S

R Q(7) 1

TABLE 4.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS E!

g CHANNEL MODES IN WHICH y

CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED c

13. Loss of Flow - Two Loops S

R N.A.

1 e

14.

Steam Generator Water Level-S R

M 1, 2

~

Low-Low 15.

Steam /Feedwater Flow Mismatch and S

R M

1, 2 Low Steam Generator Water Level

16. Undervoltage - Reactor Coolant N.A.

R N.A.

1 Pump 3usses 17.

Underfrequency - Reactor Coolant N.A.

R Q(7) 1 R.

Pump Busses s

y 18.

Turbine Trip C

A.

Low Auto Stop Oil Pressure N.A.

N.A.

S/U(1) 1, 2 B.

Turbine Stop Valve Closure N.A.

N.A.

S/U(1) 1, 2 19.

Safety Injection Input from ESF N.A.

N.A.

M(4) 1, 2 20.

Reactor Coolant Pump Breaker N.A.

N.A.

R N.A.

Position Trip 21.

A. Reactor Trip Breaker N.A.

N.A.

M(5), (9) and (11) 1, 2 and

B. Reactor Trip Bypass Breaker N.A.

N.A.

M(5), (9) and R(10) 1, 2 and

l 22.

Automatic Trip Logic N.A.

N.A.

M(5) 1, 2 and *

{

l

(

TABLE 4.3-1 (Continued)

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

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

Below the P-10 (Low Setpoint Power Range Neutron Flux Interlock) setpoint.

(1) -

If not performed in the previous 31 days.

(2) -

Heat balance only, above 15% of RATED THERMAL POWER.

(3) -

Compare incore to excore axial offset above 15% of RATED THERMAL POWER. Adjust channel if absolute difference 2 3 percent.

(4) -

Manual ESF functional input check every 18 months.

Each train or logic channel shall be tested at least every 62 days on (5) a STAGGERED TEST BASIS.

1 (6)

Neutron detectors may be excirded from CHANNEL CALIBRATION.

Each channel shall be tested at least every 92 days on a STAGGERED (7)

TEST BASIS.

The CHANNEL FUNCTIONAL TEST shall independently verify the l

(8)

OPERABILITY of the undervoltage and shunt trip circuits for the manual reactor trip function.

The test shall also verify the operability of the Bypass Breaker Trip circuit (s).

Local manual shunt trip prior to placing the bypass breaker into l

(9) service.

Automatic undervoltage trip.

l (10)

(11) -

The CHANNEL FUNCTIONAL TEST shall independently verify the l

OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers.

NORTH ANNA - UNIT 1 3/4 3-14

3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 PROTECTIVE AND ENGINEEPED SAFETY FEATURES (ESF)

INSTRUMENTATION The OPERABILITY of the protective and ESF instrumentation systems and interlocks ensure that 1) the associated ESF action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof exceeds its setpoint, 2) the specified coincidence logic and sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance consistent with maintaining an appropriate level of reliability of the Reactor Protection and Engineered Safety Features instrumentation, and 3) sufficient system functional capability is available for protective and ESF purposes from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundance and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions.

The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses.

The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards.

The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.

Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with WCAP-10271, " Evaluation of Surveillance Frequencies and Out of Service Times for Reactor Protection Instrumentation System",

and supplements to that report.

Surveillance intervals and out of service times were determined based on maintaining an appropriate level of reliability of the Reactor Protection System and Engineered Safety Features instrumentation.

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.

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 NORTH ANNA - UNIT 1 B 3/4 3-1

ATTACHMENT 2 Proposed Technical Specifications Changes For North Anna Unit 2 L

c-

z TABLE 3.3-1 E!

REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM 8

TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 3

1.

Manual Reactor Trip 2

1 2

1, 2 and

12 y

2.

Power Range, Neutron Flux 4

2 3

1, 2 2

3.

Power Range, Neutron Flux 4

2 3

1, 2 2'

High Positive Rate 4

Power Range, Neutron, Flux, 4

2 3

1, 2 2

High Negative Rate 5.

Intermediate Range, Neutron Flux 2

1 2

1, 2 and

3 6.

Source Range, Neutron Flux gg R

A.

Startup 2

1 2

2 4

B.

Shutdown 2

1 2

3*, 4* and 5*

15 C.

Shutdown 2

0 1

3, 4, and 5 5

7.

Overtemperature AT g

Three Loop Operation 3

2 2

1, 2 6

l Two Loop Operation 3

1**

2 1, 2 9

5 TABLE 3.3-1 xd REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE i

c FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 3*

3.

Over power AT Three Loop Operation 3

2 2

1, 2.

6 Two Loop Operation 3

1**

2 1, 2 9

i 9.

Pressurizer Pressure-Low 3

2 2

1, 2 7

10.

Pressurizer Pressure-High 3

2 2

1, 2 7

11.

Pressurizer Water Level--High 3

2 2

1, 2 6*

12.

Loss of Flow - Single Loop 3/ loop 2/ loop in 2/ loop in 1.

6 (Above P-8) any oper-each oper-ating loop ating-loop 6'

13.

Loss of Flow - Two Loops 3/ loop 2/ loop in 2/ loop 1

(Above P-7 and below P-8) two oper-each oper-ating loops ating loop 14.

Steam Generator Water 3/ loop 2/ loop in 2/ loop in 1, 2 7'

Level-Low-Low any oper-

'each oper-ating loops ating loop 15.

Steam /Feedwater Flow 2/ loop-level 1/ loop-level 1/ loop level 1, 2 7'

Mismatch and Low Steam and coincident and Generator Water Level 2/ loop-flow with 2/ loop-flow mismatch 1/ loop-flow mismatch or mismatch in 2/ loop-level same loop and 1/ loop-flow e

mismatch F

i

~

1 TABLE 3.3-1 (Continued)

E g

REACTOR TRIP SYSTEM' INSTRUMENTATION MINIMUM s

TOTAL NO.'

CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS

'TO TRIP OPERABLE MODES ACTION i

j2i 16.

Undervoltage-Reactor Coolant 3-1/ bus 2

2 1

~7 H

Pump Busses 4:::,

n f

j

~

Underfrequency-Reactor Coolant 3-1/ bus

-2 2

1 6

47.

Pump Busses

!$. Turbine Trip p

-fj, A.

Low Auto Stop Oil Pressure 3

2 2

1 6

B.

Turbine Stop Valve Closure 4

4 4

1 16, 3..

19.

Safety Injection Input 4

{

from ESF 2

1 2

1, 2 1-w i

N 20.

Reactor Coolant Pump Breaker j

ws Position Trip

~ / breaker 1

10 1

A.

Above P-8 1/ breaker 1

s, B.

Above P-7

~1/ breaker 2

1/ breaker 1-11 per oper-j ating loop I

i l

21.

A. Reactor Trip Breakers 2

1 2

1, 2

.1.14 5

i 2

1 2

3*,4*,5*

15

)

l B. Reactor Trip Bypass Breakers 2

1 2

13 1

22.

Automatic Trip Logic 2

1 2

1, 2 1

l 2

1 2

3*,4*,5*

15 l

i i

1 I

i i

i

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.

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

- - With the Reactor Trip Breaker open for surveillance testing in accordance with Specification Table 4.3-1 (item 21A).

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 With the number of channels OPERABLE one less than required by ACTION 1 the Minimum Channels OPERABLE requirement, be in 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 up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1.1 provided the other channel is OPERABLE.

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

a.

The inc4erable channel is placed in the tripped condition win M f hours.

l b.

The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for l

surveillance testing of the redundant channel (s) per Specification 4.3.1.1.1.

c.

Either, THERMAL POWER is restricted to s 75% of RATED THERMAL POWER and the Power Range, Neutron Flux trip setpoint is reduced to s 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 once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

d.

The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75 percent of RATED THERMAL POWER with one Power Range Channel inoperable by using the movable incore detectors to confirm that the normalized symmetric power distribution, obtained from 2 sets of 4 symmetric thimble locations or a full-core flux map, is consistent with the indicated QUADRANT POWER TILT RATIO at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

1 NORTH ANNA - UNIT 2 3/4 3-5

TABLE 3.3-1 (Continued)

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

a.

Below the P-6, (Block of Source Range keactor 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 continued.

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 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 OP'ERATION may proceed provided the following conditions are satisfied:

a.

The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and b.

The Minimum channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of the other channels per Specification 4.3.1.1.1.

ACTION 7

- With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and POWER OPERATION may proceed until performance of the next required CHANNEL FUNCTIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 8

- Not applicable.

NORTH ANNA - UNIT 2 3/4 3-6

TABLE 3.3-1 (Continued)

With a channel associated with an operating loop inoperable, ACTION 9 restore the inoperable channel to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 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 ; however, one channel associated with an operating loop may by bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1.1.

With one channel inoperable, restore the inoperable channel to ACTION 10 OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to below P-8, (Block of Low Reactor Coolant Pump Flow and Reactor Coolant Pump Breaker Position) setpoint, within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Operation below the P-8, (Block of Low Reactor Coolant Pump Flow and Reactor Coolant Pump Breaker Position) setpoint, may continue pursuant to ACTION 11.

ACTION 11 - With less than the Minimum Number of Channels OPERABLE, operation may continue provided the inoperable channel is placed in the tripped condition within I hour.

With the number of channels OPERABLE one less than required by ACTION 12 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.

With the number of channels OPERABLE one less than required by ACTION 13 the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within (1) hour or terminate testing of the Reactor Trip Breaker and open the Reactor Trip Bypass Breaker.

ACTION 14 - With one of the diverse trip features (undervoltage or shunt trip device) inoperable, restore it to OPERABLE atatus within 48 hoers or declare the breaker inoperable and apply Action 1.

The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time required for performing maintenance to restore the breaker to OPERABLE status.

I With the number of channels OPERABLE one less than required by ACTION 15 the Minimum Channels OPERABLE requirement restore the inoperable l

channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the reactor j

trip breakers within the next hour.

ACTION 16 -

With the number of OPERABLE channels less than the Total Number l

Channels, operation may continue provided the inoperable I

channels are placed in the trip condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l NORTH ANNA

'INIT 2 3/4 3-7

[

.l TABLE 6.3-1 j

g 4

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH g

CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED E

1.

Manual Reactor Trip N.A.

N.A.

R(8) 1,2 and

w 2.

Power Range, Neutron Flux a.

High Setpoint S

D(2), M(3) and Q(6)

Q(7) 1, 2

+

b.

Low Setpoint S

R(6)

S/U(1) 1***,

2 3.

Power Range, Neutron Flux, N.A.

R(6)

Q(7) 1, 2 High Positive Rate 4.

Power Range, Neutron Flux, N.A.

R(6)

Q(7) 1, 2 High Negative Rate 5.

Intermediate Range, S

R(6)

S/U(1) 1***, 2 and

g Neutron Flux 4-6.

Source Range, Neutron Flux S

R(6)

Q(7), S/U(1) 2**,

3, 4, 5 and u

7.

Overtemperature AT S

R(6)

Q(7')

1, 2 8.

Overpower AT S

R(6)

Q(7) 1, 2 3

9.

Pressurizer Pressure-Low S

R M

1, 2 10.

Pressurizer Pressure--High S

R M

1, 2 11.

Pressurizer Water Level--High S

R Q(7) 1, 2 12.

Loss of Flow - Single Loop S

R Q(7) 1

TABLE 4.3-1 (continued)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS D

CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION

_, TEST REQUIRED e

3 13.

Loss of Flow - Two Loops S

R N.A.

I 4

14.

Steam Generator Water Level-S R

M 1, 2 Low-Low 15.

Steam /Feedwater Flow Mismatch and S

R M

1, 2 Low Steam Generator Water Level 16.

Undervoltage - Reactor Coolant N.A.

R M

1 Pump Busses 17.

Underfrequency - Reactor Coolant N.A.

R Q(7) 1 Pump Busses M*

18.

Turbine Trip Y

A.

Low Auto Stop Oil Pressure N.A.

N.A.

S/U(1)

N.A.

O B.

Turbine Stop Valve Closure N.A.

' N.A.

S/U(1)

N.A.

19.

Safety Injection Input from ESF N.A.

N.A.

M(4) 1, 2 20.

Reactor Coolant Pump Breaker N.A.

N.A.

R 1

Position Trip 21.

A. Reactor Trip Breaker N.A.

N.A.

M(5),(9) & (11) 1, 2 an4

i B. Reactor Trip Bypass Breaker N.A.

N.A.

M(5),(9) & R(10) 1, 2 and

4 1

J l

22.

Automatic Trip Logic N.A.

N.A.

M(5) 1, 2 and

i I

1

TABLE 4.3-1 (Continued)

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

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

- - - Below P-10 (Low Setpoint Power Range Fautron Flux Interlock) Setpoint.

If not performed in previous 31 days.

(1)

(2) - Heat balance only, above 15% of RATED THERMAL P0KER.

Adjust channel if absolute difference > 2 percent.

Compare incore to excore axial flux difference above 15% of RATED (3)

THERMAL POWER. Recalibrate if the absolute difference 2 3 percent.

(4) - Manual ESF functional input check every 18 months.

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

- - Neutron detectors may be excluded from CHANNEL CALIBRATION.

(6)

Each channel shall be tested at least every 92 days on a STAGGERED (7)

TEST BASIS.

.(8) - The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip circuits for the Manual Reactor Trip Function.

The test shall also verify the OPERABILITY of the Bypass Breaker trip circuit (s).

(9)

Local manual shunt trip prior to placing the bypass breaker into service.

l (10) - Automatic undervoltage trip.

(11) - The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt crip attachments of the Reactor Trip l

Breakers.

l NORTH ANNA - UNIT 2 3/4 3-14 l

m

3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESF)

INSTRUMENTATION The OPERABILITY of the protective and ESF instrumentation systems and interlocks ensure that 1) the associated ESF action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint, 2) the specified coincidence logic and sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance consistent with maintaining and appropriate level of reliability of the Reactor Protection and Engineared Safety Features instrumentation, and 3) sufficient system functional capability is available for protective and ESF purposes from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions.

The integrated operation of each of these systems is consistent with the as-sumptions used in the accident analyses.

The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards.

The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.

Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with WCAP-10271, " Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System",

and supplements to that report.

Surveillance intervals and out of service times were determined based on maintaining an appropriate level of reliability of the Reactor Protection System and Engineered Safety Features instrumentation.

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, overlap-ping 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.

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 individu-al channals and 2) the alarm or automatic action is initiated when the ra-diation level trip setpoint is exceeded, i

NORTH ANNA - UNIT 2 B 3/4 3-1

i ATTACHMENT 3 Discussion Of Proposed Changes i

l l

Discussion Of Proposed Changes Background' In response to growing concerns of the impact of current testing and maintenance requirements on plant. operation, particularly as related to instrumentation ~ systems, the Westinghouse Owners Group (WOG) initiated a program' to develop 'a justification to be used to revise generic and plant c

specific instrumentation-technical specifications..

Significant time and effort on. the part of ~ the operating staff was devoted to performing, reviewing, documenting and tracking the various surveillance activities, which in many instances seemed unwarranted based on the high reliability of the equipment.

Significant benefits for operating plants appeared to be achievable through revision of instrumentation test and maintenance requirements.

On February 3,

1983 the Westinghouse Owners Group submitted WCAP-10271,

" Evaluation of Surveillance Frequencies and Out of Service Times for the

- Reactor Protection Instrumentation System" (Reference 1) to the NRC as the first step in gaining approval of the instrumentation program.

WCAP-10271 documents the justification to be used to justify revisions - to technical specifications.

The technical specification revisions evaluated were increased test and maintenance times, less frequent surveillance, and testing in bypass.

~

On October 4, 1983 the WOG submitted Supplement 1 to WCAP-10271 (Reference 2) to the NRC.

Supplement I demonstrates the applicability of the justification contained-in WCAP-10271 to reactor protection systems for two, three and four loop plants with either relay or solid state logic.

Additionally this supplement extends the evaluation to topics not addressed in the original WCAP-such as the interdependence (or lack there of) of surveillance intervals and hardware failure rates.

1 In February 1985 tne NRC issued the Safety Evaluation Report (Reference 3) for i

WCAP-10271 and Supplement 1.

The SER approved quarterly testing, a 6-hour t.

outage time, increased test-time and testing in bypass for analog channels.

Specifically, the reactor trip system instrumentation functional units that are affected by the proposed changes in this submittal are the power range

-neutron flux, the power range neutron flux high positive rate, the power range S

neutron flux high negative rate, the intermediate range neutron flux, the source ' range neutron flux, the overtemperature A T, the overpower AT, the pressurizer water level-high, the loss of flow-single loop, the loss of flow-two loops, the under-frequency-reactor coolant pump busses, and the turbine - trip.

The proposed changes are to increase surveillance intervals from once per month to once per quarter, to increase the time during which an i

inoperable channel may be maintained in an untripped condition from one hour to six hours, and to increase the time an inoperable channel may be bypassed l'

to. a) low tecting of another channel in the same function from two hours to four hours.

4 i

The proposed changes are based on the results and conclusions presented in references 1 and 2 and approved generically by reference 3.

These changes 3

have been made in accordance with the guidance provided in references 4 and 5.

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

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Item by Item Discussion of the Changes 1.

Unit I and Unit 2 Technical Specifications Table 3.3-1.

A.

For the power range neutron flux, the power range neutron flux high positive rate, and the power range neutron flux high negative rate functional units the allowed maintenance time has been increased from one hour to six hours, and the time allowed for testing with an inoperable channel being bypassed has been increased from two hours to four hours.

B.

For the overtemperature A T (three loop operation), the overpower A T (three loop operation), the pressurizer water level-high, the loss of flow-single

loop, the loss of flow two

loops, the under-frequency-reactor coolant pump busses, and the turbine trip -

low auto stop oil pressure functional units the allowed maintenance time has been increased from one hour to six hours. Also, specific provisions have been added that allow for testing with an inoperable channel being bypassed for up to four hours.

C.

For the turbine trip - turbine stop valve closure functional unit the allowed maintenance time has been increased from one hour to six hours.

2.

Unit 1 and Unit 2 Technical Specifications Table 4.3-1.

A.

For the power range neutron flux functional unit separate surveillance requirements have been established for the high setpoint and the low setpoint.

For the high setpoint the surveillance interval for the channel fur.ctional test has been changed from monthly to quarterly on a staggered test basis.

For the low setpoint the surveillance int'ervals have been established as each shift for the channel

check, each 18 months for the channel calibration, and prior to startup if not performed in the previous thirty one days for the channel functional test.

For the low setpoint, surveillance is required in Mode 1 below P-10, and Mode 2 B.

For the power range neutron flux high positive rate, the power range neutron flux high negative rate, the overtemperature A T, the overpower A T, the pressurizer water level-high, the loss of flow-single loop, and the under-frequency-reactor coolant pump busses functional units the surveillance interval for the channel functional test has been changed from monthly to quarterly on a staggered test basis.

C.

For the intermediate range neutron flux functional unit the surveillance interval for the channel functional test has been changed from prior to startup if not performed in the previous seven days to prior to startup if not performed in the previous thirty one days.

Additionally, no surveillance of any type is required to be performed in Mode 1 above the P-10 (low setpoint power range neutron flux interlock) setpoint.

q D.

For the source range neutron flux functional unit the surveillance interval for the channel functional test has been changed. The monthly interval has been changed to quarterly on a staggered test basis; and the prior to startup if not performed in the previous seven days surveillance interval has been changed to prior to startup if not performed in the previous thirty one days.

Additionally, no surveillance of any type is required to be performed in Mode 2 above the P-6 (block of the source range reactor trip) setpoint.

E.

For Unit 1 only, the note numbers for noten 8, 9, and 10 on technical specifications pages 3/4 3-13 and 3/4 3-14 have been changed in order to be consistent with the Unit 2 Technical Specifications.

NRC Stipulations For Acceptance Of The Proposed Changes In reference 3 the NRC identified five stipulations that must be addressed by utilities seeking to implement the technical specification changes approved generically as a result of their review of WCAP-10271 and Supplement 1.

These items are dispositioned below.

1.

Staggered Test Plan - reference 3, page 7 state: "Accordingly, the staff's acceptance of less frequent surveillance is contingent on the implementation of the staggered test plan."

Disposition: A note has been added to Technical Specification Table 4.3-1.

The note is associated with the reactor trip system instrumentation functional units that are required to have channel functional tests performed quarterly.

The note states "Each channel shall be tested at least every 92 days on a STAGGERED TEST Basis."

2.

Identification of Common Cause Failures - reference 3, page 8 states

"..., the staff's acceptance of less frequent surveillance is contingent on implementation of procedures to identify common cause failures and to test the other channels which may be affected by the common cause.

.... licensees should confirm that this practice is formalized in plant procedures.

First, the procedure should require an evaluation of any RTS channel test failure to determine if that failure could be a common cause failure.

Second, the procedure should require testing of the additional channels in that function if the failure is determined to be a plausible common cause failure."

Disposition:

Prior to implementing the proposed changes to the technical specifications, the plant surveillance procedures for the power range neutron flux, the power range neutron flux high positive rate, the power range neutron flux high negative rate, the source range neutron flux, the overtemperature AT, the overpower AT, the pressurizer water level-high, the loss of flow-single loop, and the underfrequency-reactor coolant pump busses functional units will be modified. The procedures will require an evaluation of a test failure to determine if that failure could be a common cause failure.

The procedure will also require testing of the additional channels in that function if the failure is determined to be a plausible common cause failure.

E i

3. Testing In The Bypassed Mode - reference 3, page 8 states "...., licensees choosing the option to perform routine channel testing in the bypass mode should ensure that the plant design allows testing in bypass without lifting leads or installing temporary jumpers. The staff's acceptance of this option is contingent on confirmation of this capability."

Disposition:

The plant design currently would not allow testing in bypass without lifting leads or installing temporary jumpers.

Therefore, we are not pursuing this option at this time.

4.

Engineered Safety Features Actuation System - reference 3, page 8 states "Some RTS functional units also provide input to safety related systems such as the Engineered Safety Feature Actuation System (ESFAS).

In order.

to avoid confusion in plant technical specifications regarding such dual function channels, the staff concludes that either 1) the channels should not be changed in the RTS tables until the ESFAS review is finished or 2) cautionary notes in the RTS tables should refer to the more stringent ESFAS requirements."

Disposition: The pressurizer pressure-low, the pressurizer pressure-high, the steam generator water level-low low, the steam /feedwater flow mismatch and low steam generator water level, and the undervoltage-reactor coolant pump busses functional units are dual function units. We have chosen option 1; no changes are being proposed for these functional units at this time.

5.

Setpoint Determination reference 3,

page 9 states "...., the staff's acceptance is contingent on confirmation that the instrument setpoint methodology includes sufficient adjustments to offset the drift anticipated as a result of less frequent surveillance."

Disposition:

Enclosure A to this attachment documents our findings in this area and. confirms that the instrument setpoint methodology includes sufficient adjustments to offset the drift anticipated as a result of less frequent surveillance.

In summary we have met the requirements of the NRC stipulations.

50.59 Safety Review Pursuant to 10 CFR 50.59, we have reviewed the proposed Technical Specification changes and have concluded that no unreviewed safety question exists.

Specifically:

(1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased by these proposed changes because the estimated change in reactor protection system unavailability is very small as is the estimated reduction in core damage frequency coming from inadvertent trips.

Overall, it has been concluded that the change in core damage frequency and risk is insignificant.

r (2) -The possibility for an accident or malfunction of a different type than any evaluated previously.in the safety analysis report is not being created by these proposed changes because these proposed changes do not change the manner in which the Reactor Protection System provides protection to the plant..

(3) -The margin of safety as defined in the basis for any technical specification is not reduced because these proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operations are established and because the operability and performance of the reactor trip system instrumentation is not being significantly affected by these proposed changes.

50.91 Significant Hazards Review In accordance with the provisions of 10 CFR 50.91 we are providing an analysis about the issue of no significant hazards consideration, using the standards of 10 CFR 50.92 and the guidance provided in Generic Letter 86-03.

We have concluded

-that these proposed changes involve no significant hazards considerations. Specifically operation of the facility in accordance with the proposed changes would not:

(1)

Involve a significant increase in the probability or consequences

~

of an accident previously evaluated because overall it has been concluded that the associated change in core damage frequency and risk is insignificant.

(2) Create the possibility of a new or different kind of accident from any accident previously evaluated because these proposed changes do not involve hardware changes and do not change the manner in which the Reactor Protection System provides plant protection.

(3)

Involve a significant reduction in the margin of safety because the operability and performance of the reactor trip system instrumentation is not being significantly affected by these proposed changes.

References 1.

WCAP-10271, " Evaluation of Surveillance Frequencies And Out Of Service Times For The Reactor Protection Instrumentation System," January,1983.

2.

WCAP-10271, Supplement 1, " Evaluation Of Surveillance Frequencies And Out Of Service Times For The Reactor Protection Instrumentation System," July, 1983.

3.

Letter from Mr. C. O. Thomas (NRC) to Mr.

J.

Sheppard (Westinghouse Owners Group), " Acceptance For Referencing Of Licensing Topical Report WCAP-10271, Evaluation Of Surveillance Frequencies And Out Of Service Times For The Reactor Protection Instrumentation Systems," February 21, 1985.

4.

Letter from Mr. L. D. Butterfield (Westinghouse Owners Group) to Mr. H. R.

Denton(NRC),May 16,~1985.

5.

Letter from Mr. H. R. Denton (NRC) to Mr. L. D.

Butterfield (Westinghouse Owners Group), July 24, 1985.

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ML20210F676

Text

Response

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