Text

Proposed Changes to Tech Specs 3.7 & 4.1,upgrading Table 3.7-1, Reactor Trip Instrument Operating Conditions & Modifying Surveillance Requirements for Reactor Trip Sys Interlocks & Reactor Protection Sys,Respectively
	ML18150A109
Person / Time
Site:	Surry
Issue date:	05/22/1987
From:	; VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	;
Shared Package
ML18150A108	List: ML18150A107; ML18150A109;
References
	GL-83-28, GL-85-09, GL-85-9, NUDOCS 8706020046
	Download: ML18150A109 (28)
	v • d • e

(

e ATTACHMENT 1 PROPOSED TECHNICAL SPECIFICATION CHANGES FOR 8706020046 870522 PDR ADOCIA. 05000280 p

PDR SURRY UNITS 1 AND 2

e e

TS 3.7-1 3.7 INSTRUMENTATION SYSTEMS Operational Safety Instrumentation Applicability Applies to reactor and safety features instrumentation systems.

Objectives To provide for automatic initiation of the Engineered Safety Features in the event that principal process variable limits are

exceeded, and to delineate the conditions of the plant instrumentation and safety circuits necessary to ensure reactor safety.

Specification A.

For on-line testing or in the event of a

subsystem instrumentation channel failure, plant operation at rated power shall be permitted to continue in accordance with TS Tables 3.7-1 through 3.7-3.

B.l The reactor trip system instrumentation channels shall be operable as specified in TS Table 3.7-1.

B.2 In the event the number of channels of a particular subsystem in service falls below the limits given in the column entitled Minimum Operable Channels or Minimum Degree of Redundancy cannot be achieved, operation shall be limited according to the requirement shown in Column 4 of TS Tables 3.7-2 and 3.7-3.

j

c.

e TS 3.7-2 In the event of subsystem instrumentation channel failure permitted by Specification 3.7.B2, Tables 3.7-2 and 3.7-3 need not be observed during the short period of time an operable subsystem channel is tested where the failed channel must be blocked to prevent unnecessary reactor trip.

D.

The Engineered Safety Features initiation instrumentation setting limits shall be as stated in TS Table 3.7-4.

E.

The radioactive liquid and gaseous effluent monitoring instrumentation channels shown in Table 3.7-S(a) and Table 3.7-S(b) shall be operable with their alarm/trip setpoints set to ensure that the limits of Specifications 3.11.A.1 and 3.11.B.1 are not exceeded.

The alarm trip setpoints of these channels shall be determined and adjusted in accordance with the Offsite Dose Calculation Manual (ODCM).

1.

With a radioactive liquid or gaseous effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above specification, without delay suspend the release of radioactive liquid or gaseous effluents monitored by the affected channel and declare the channel inoperable or change the setpoint so it is acceptably conservative.

2.

With less than the minimum number of radioactive liquid or gaseous effluent monitoring instrumentation channels operable, take the action shown in Table 3.7-S(a) or Table 3.7-S(b).

Exert best efforts to return the instruments to operable status within explain in the next 30 days

and, if unsuccessful, Semiannual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.

1

TABLE 3.7-1 REACTOR TRIP INSTRUMENT OPERATING CONDITIONS FUNCTIONAL UNIT TOTAL NUMBER OF CHANNELS MINIMUM OPERABLE CHANNELS CHANNELS TO TRIP PERMISSIBLE BYPASS CONDITIONS OPERATOR ACTION

1. Manual

2. Nuclear Flux Power Range

3. Nuclear Flux Intermediate Range

4. Nuclear Flux Source Range A. Below P Note A B. Shutdown - Note B

5. Overtemperature LlT

6. Overpower LlT

7. Low Pressurizer Pressure

8. Hi Pressurizer Pressure 2

4 2

2 2

3 3

2 3

2 2

1 2

2 2

2 1

1 0

2 2

Low trip setting at P-10 P-10 P-6 P-7 Note A -

With the reactor trip breakers closed and the control rod drive system capable of rod withdrawal Note B -

With the reactor trip breakers open 1

2 3

4 5

6 6

7 7

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TABLE 3.7:...1 REACTOR TRIP INSTRUMENT OPERATING CONDITIONS MINIMUM TOTAL NUMBER OPERABLE CHANNELS PERMISSIBLE FUNCTIONAL UNIT OF CHANNELS CHANNELS TO TRIP BYPASS CONDITIONS OPERATOR ACTION

9. Pressurizer-Hi Water Level 3

2 2

P-7 6

10. Low Flow 3/loop 2/loop in 2/loop in P-8 6 e each aper-any aper-ating loop ating loop 2/loop in P-7 any 2 aper-ating loops

11. Turbine Trip A. Stop valve closure 4

1 4

P-7 11 B. Low fluid oil pressure 3

2 2

P-7 6

12. Lo-Lo Steam Generator 3/loop 2/loop in 2/loop in 7

Water Level each aper-any aper-ating loop ating loops

13. Underfrequency 4KV Bus 3-1/bus 2

2 P-7 6

14. Undervoltage 4KV Bus 3-1/bus 2

2 P-7 7

15. Safety Injection Input 2

2 1

8 A From ESF t-3

16. Reactor Coolant Pump Cf.l Breaker Position l,.).

I/breaker I/breaker 1

P-8 9

--.J I

per aper-2 P-7 10 I-'

I-'

ating loop

TABLE 3.7-1 REACTOR TRIP INSTRUMENT OPERATING CONDITIONS MINIMUM TOTAL NUMBER OPERABLE CHANNELS PERMISSIBLE FUNCTIONAL UNIT OF CHANNELS CHANNELS TO TRIP BYPASS CONDITIONS OPERATOR ACTION

17. Low steam generator water 2/loop-level I/loop-level I/loop level with steam/feedwater and and 2/loop level coin-flow mismatch 2/loop-flow flow mis-cident with mismatch match or I/loop-2 loop/level flow mis-and I/loop-match in flow mis-same loop match

18. A. Reactor Trip Breakers 2

2 1

B. Reactor Trip 2

1 1

Bypass Breakers - Note C

19. Automatic Trip Logic 2

2 1

Note C - With the Reactor Trip Breaker open for surveillance testing in accordance with Specification Table 4.1-1 (item 30)

PERMISSIBLE BYPASS CONDITIONS - P-6, P-7, P-8 and P-10 are defined in TS Table 4.1-A 7

8 11

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ACTION 1.

ACTION 2.A.

e TS 3.7-13 TABLE 3.7-1 (Continued)

TABLE NOTATION ACTION STATEMENTS With the number of channels OPERABLE, one less than required by the minimum OPERABLE Channels 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 SHUTDOWN 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 OPERABLE channels OPERABLE Channels, POWER OPERATION may following conditions are satisfied:

equal to the Minimum proceed provided the

1.

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 .

2.

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 redundant channel(s) per Specification 4.1.

3.

Either, THERMAL POWER is restricted to $75% of RATED POWER and the Power Range, Neutron Flux trip setpoint is reduced to

$85%

of RATED POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or, the QUADRANT POWER TILT is monitored at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

e TS 3.7-13a TABLE 3.7-1 (Continued)

4.

The QUADRANT POWER TILT shall be determined to be within the limit when above 75 percent of RATED 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 at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2.B.

With the number of operable channels one less than required by the Minimum Operable channels requirement, be in Hot Shutdown within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 3.

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

a.

Below 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 P-6, (Block of Source Range Reactor Trip) setpoint, but below 10%

of RATED

POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10% of RATED POWER.

c.

Above 10% of RATED POWER, POWER OPERATION may continue.

ACTION 4.

ACTION 5.

ACTION 6.A.

e TS 3.7-13b TABLE 3.7-1 (Continued)

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

a.

Below 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 P-6, operation may continue.

With the number of channels OPERABLE one less than required by the Minimum OPERABLE Channels requirement, verify compliance with the SHUTDOWN MARGIN requirements 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.

With the number of OPERABLE Channels Operable Channels, POWER OPERATION may following conditions are satisfied:

equal to the Minimum proceed provided the

1.

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 .

2.

The Minimum OPERABLE Channels 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 4.1.

6.B.

With the number of OPERABLE Channels one less than required by the Minimum Operable Channels requirement, be in Hot Shutdown within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 7.A.

7.B.

ACTION 8.A.

ACTION 8.B.

e TS 3. 7-13c TABLE 3.7-1 (Continued)

With the number of OPERABLE Channels equal to the Minimum Operable

Channels, POWER OPERATION may proceed until performance of the next required CHANNEL FUNCTIONAL TEST provided the inoperable channel is placed in the tripped condition within 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With the number of OPERABLE Channels one less than required by the Minimum Operable Channels requirement, be in Hot Shutdown within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With the number of OPERABLE Channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and 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.

However, one channel may be bypassed for up to 2

hours for surveillance testing per Specification 4.1. provided the other channel is OPERABLE.

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 8.A.

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 9.

ACTION 10.

ACTION 11.

ACTION 12.

TS 3.7-13d TABLE 3.7-1 (Continued)

With one channel inoperable, restore the inoperable channel to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or reduce THERMAL POWER to below the P-8, (Block of Low Reactor Coolant Pump Flow and Reactor Coolant Pump Breaker Position) setpoint, within the next 2

hours.

Operation below P-8 may continue pursuant to ACTION 10.

With less than the Minimum Number of Channels

OPERABLE, operation may continue provided 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 .

With the number of OPERABLE channels one less than Minimum Channels OPERABLE requirement, be in at least HOT SHUTDOWN within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-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 4.1.

provided the other channel is OPERABLE.

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 .

e TS 4.4-1 4.1 OPERATIONAL SAFETY REVIEW Applicability Applies to items directly related to safety limits and limiting conditions for operation.

Objective To specify the minimum frequency and type of surveillance to be applied to unit equipment and conditions.

Specification A.I Calibration, testing, and checking of instrumentation channels shall be performed as detailed in Table 4.1-1 and 4.1-2.

A.2 The logic for the reactor trip system interlocks listed in Table 4.1-A shall be demonstrated operable prior to each reactor startup unless performed during the preceeding 92 days.

The interlock function shall be demonstrated operable at each refueling by channel calibration testing of each channel affected by interlock operation.

B.

Equipment tests shall be conducted as detailed below and in Table 4.l-2A.

1.

Each Pressurizer PORV shall be demonstrated operable:

a.

At least once per 31 days by performance of a channel functional test, excluding valve operation, and

b.

At least once per 18 months by performance of a

channel calibration.

2.

Each Pressurizer PORV block valve shall be demonstrated operable at least once per 92 days by operating the valve through one complete cycle of full travel.

J

TABLE 4.1-1 MINIMUM FREQUENCIES FOR CHECK, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS Channel Description

1.

Nuclear Power Range

2.

Nuclear Intermediate Range (below P-10 setpoint)

3.

Nuclear Source Range (below P-6 setpoint)

4.

Reactor Coolant Temperature

5.

Reactor Coolant Flow

6.

Pressurizer Water Level

7.

Pressurizer Pressure (High & Low)

8.

4 kV Voltage and Frequency

9.

Analog Rod Position Check s

S

S

S s

s s

N.A.

S(l,2)

(4)

Calibrate D(l)

Q(3)

R(4)

R(2)

R R

Test M(2)

P(l)

M(l)

M(2)

M M

M(3)

Remarks

1)

Against a heat balance standard

2)

Signal at ~T; bistable action (permissive, rod stop, trip)

3)

Upper and lower chambers for symmetric offset by means of the movable incore detector system

4)

Neutron detectors may be excluded from Channel Calibration

1)

Log level; bistable action (permissive, rod stop, trip)

2)

Neutron detectors may be excluded from Channel Calibration

1)

Bistable action (alarm, trip)

2)

Neutron detectors may be excluded from Channel Calibration

1)

Overtemperature ~T

2)

Overpower ~T

1)

With step counters

2)

Each six inches of rod motion when data logger is out of service

3)

Rod bottom bistable action

4)

N.A. when reactor is in cold shutdown

TABLE 4.1-1 (Continued)

MINIMUM FREQUENCIES FOR CHECK, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS Channel Description

10.

Rod Position Bank Counters Check S (1, 2)

11.

Steam Generator Level S

12.

Charging Flow N.A.

13.

Residual Heat Removal Pump Flow N.A.

14.

Boric Acid Tank Level

D

15.

Refueling Water Storage Tank Level S

16.

Volume Control Tank Level

17.

Reactor Containment Pressure-CLS

18.

Boric Acid Control

19.

Containment Sump Level

20.

Accumulator Level and Pressure

21.

Containment Pressure-Vacuum Pump System

22.

Steam Line Pressure N.A.

D N.A.

N.A~

s s

s Calibrate N.A.

R R

Test N.A.

M N.A.

N.A.

M N.A.

M(l)

N.A.

M Remarks

1)

Each six inches of rod motion when data logger is out of service

2)

With analog rod position

1)

Isolation valve signal and spray signal.

J e

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TABLE 4.1-1 (Continued)

MINIMUM FREQUENCIES FOR CHECK, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS Channel Description

23.

Turbine First Stage Pressure

24.

Emergency Plan Radiation Instr.

25.

Environmental Radiation Monitors

26.

Logic Channel Testing

27.

Turbine Overspeed Protection Trip Channel (Electrical)

28.

Turbine Trip A.

Stop valve closure B.

Low fluid oil pressure

29.

Seismic Instrumentation

30.

Reactor Trip Breaker

31.

Reactor Coolant Pressure (Low)

32.

Auxiliary Feedwater

a.

Steam Generator Water Level Low-Low

b.

RCP Undervoltage

c. s.I.

d.

Station Blackout

e.

Main Feedwater Pump Trip Check s

M

M N.A.

N.A.

M N.A.

N.A.

s s

(All N.A.

N.A.

Calibrate R

R N.A.

N.A.

R N.A.

N.A.

R N.A.

R R

R Safety Injection R

N.A.

Test M

M N.A.

M R

p p

M M

N.A.

M M

surveillance N.A.

R Remarks TLD Dosimeters Setpoint verification is not applicable The test shall independently verify operability of the undervoltage and shunt trip attachments requirements) e H

Cf.l

.p..

I-'

I CXl

33.

34.

Loss

a.

b.

TABLE 4.1-1 (Continued)

MINIMUM FREQUENCIES FOR CHECK, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS Channel Description Check Calibrate Test of Power 4.16 KV Emergency Bus Under-N.A *.

R M

voltage (Loss of Voltage) 4.16 KV Emergency Bus Under-N.A.

R M

voltage (Degraded Voltage)

Control Room Chlorine Detectors Remarks

35.

Manual Reactor Trip s

N.A.

R N.A.

M R

The test shall independently verify the operability of the undervoltage and shunt trip attachments for the manual reactor trip function.

The test shall also verify the operability of the bypass breaker trip circuit.

36.

Reactor Trip Bypass Breaker

37.

38.

39.

Safety Injection Input from ESF Reactor Coolant Position Trip Steam/Feedwater S/G Water Level S

- Each shift D

- Daily Pump Breaker Flow and low N.A.- Not Applicable N.A.

N.A.

s Q

- Every 90 effective full power days

See Specification 4.1.D N.A.

N.A.

R M

p R

M(l), R(2)

R R

M

- Monthly (1) Local manual undervoltage trip prior to placing breaker in service.

(2)

Automatic shunt trip.

- Prior to each startup if not done within the previous week

- Each Refueling Shutdown e

DESIGNATION P-6 P-10 P-7 P-8 TABLE 4.1-A REACTOR TRIP SYSTEM INTERLOCKS CONDITION 1 of 2 Intermediate range above setpoint (increasing power level) 2 of 2 Intermediate range below setpoint (decreasing power level) 2 of 4 Power range above set-point (increasing power level) 3 of 4 Power range below set-point (decreasing power level) 2 of 4 Power range above setpoint or 1 of 2 Turbine Impulse chamber above setpoint (Power level increasing) 3 of 4 Power range below setpoint and 2 of 2 Turbine Impulse chamber pressure below setpoint (Power level decreasing) 2 of 4 Power range above setpoint (Power level increasing) 3 of 4 Power range below setpoint (Power level decreasing)

FUNCTION Allows manual block of source range reactor trip Automatically defeats the block of source range reactor trip Allows manual block of power range (low setpoint) and intermediate range reactor trips and inter-mediate range rod stop.

Automatically blocks source range reactor trip.

Automatically defeats the block of power range (low setpoint) and intermediate range reactor trips and intermediate range rod stop.

Input to P-7.

Allows reactor trip on:

Low flow or reactor coolant pump breakers open in more than one loop, Undervoltage (RCP busses), Underfrequency (RCP busses), Turbine Trip, Pressurizer low pressure, and Pressurizer high level.

Prevents reactor trip on:

Low flow or reactor coolant pump breakers open in more than one loop, Undervoltage (RCP busses), Underfrequency (RCP busses), Turbine Trip, Pressurizer low pressure, and Pressurizer high level Permit reactor trip on low flow or reactor coolant pump breaker open in a single loop.

Blocks reactor trip on low flow or reactor coolant pump breaker open in a single loop.

e J-3 Cf.l

TABLE 4.1-lA RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION CHANNEL DESCRIPTION

1.

GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE (a) Liquid Radwaste Effluent Line

2.

GROSS BETA OR GAMMA RADIOACTIVITY MONITORS PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC TERMINATION OF RELEASE (a) Circulating Water Discharge Line (b) Component Cooling Service Water System Effluent Line

3.

FLOW RATE MEASUREMENT DEVICES (a) Liquid Radwaste Effluent Line D - Daily M - Monthly R - Each Refueling Shutdown Q - Quarterly PR - Prior to each release N.A. - Not Applicable CHANNEL SOURCE CHECK CHECK D

PR D

M D

N.A.

SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL FUNCTIONAL CALIBRATION TEST R

Q R

N.A.

e 1-3 Cf.l I

00

(')

TABLE 4.1-lB RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL DESCRIPTION CHECK

1.

PROCESS VENT SYSTEM (a) Noble Gas Activity Monitor Providing Alarm and Automatic Termination of Release D

(b) Iodine Sampler w

(c) Particulate Sampler w

(d) Process Vent Flow Rate Monitor D

(e) Sampler Flow Rate Measuring Device D

2.

WASTE GAS HOLDUP SYSTEM EXPLOSIVE GAS MONITORING SYSTEM (a) Hydrogen Monitor D

(b) Oxygen Monitor D

3.

CONDENSER AIR EJECTOR SYSTEM (a) Gross Activity Monitor D

(b) Air Ejector Flow Rate Measuring Device D

4.

VENTILATION VENT SYSTEM (a) Noble Gas Activity Monitor D

(b) Iodine Sampler W

(c) Particulate Sampler W

(d) Ventilation Vent Flow Rate Monitor D

(e) Sampler Flow Rate Measuring Device D

(1) - The channel calibration shall include the use of standard

1.

one volume percent hydrogen, balance nitrogen, and

2.

four volume percent hydrogen, balance nitrogen.

CHECK CALIBRATION M*

R N.A N.A N.A.

N.A N.A.

R N.A.

SA N.A.

Q(l)

N.A.

Q(2)

M R

N.A.

R M

R N.A.

N.A.

R N.A.

SA gas samples containing a nominal:

(2) - The channel calibration shall include the use of standard gas samples containing a nominal:

1.

one volume percent oxygen, balance nitrogen, and

2.

four volume percent oxygen, balance nitrogen.

D - Daily W - Weekly M - Monthly R - Each Refueling Shutdown SA - Semi-annually NA - Not Applicable Q - Quarterly Monthly and prior to each Waste Gas Decay Tank Release TEST Q

N.A.

M M

Q N.A.

N.A.

TS 4.1-9 DELETE

ATTACHMENT 2 DISCUSSION OF PROPOSED CHANGES FOR SURRY UNITS 1 AND 2

DISCUSSION OF PROPOSED CHANGES Changes to Section 3.7 (Instrumentation System Limiting Condition of Operation) and 4.1 (Operational Safety Review) of Surry's Technical Specifications are being proposed as part of our response to Generic Letter 83-28, Item 4.5.3.

This item addressed the surveillance intervals for on-line functional testing of the reactor trip system instrumentation.

These changes are consistent with WCAP-10271 "Evaluation of Surveillance Frequencies and Out of Service Times for Reactor Protection Instrumentation System" and the associated NRG Safety Evaluation dated July 24,

1985, both documents concluded that the estimated change in reactor protection system unavailability is very small as is the estimated reduction in core damage frequency coming from inadvertent trips and that the change in core damage frequency and risk is insignificant.

A description of the changes to each section is provided below.

Section 3.7 The Specification statement for the reactor trip system instrumentation channels (T.S.

Table 3.7-1) has been separated from the Specification statements for the engineered safeguards action instrumentation (T.S. Table 3.7-2) and the instrument operating conditions for isolation functions (T.S.

Table 3.7-3).

This has been done because T.S. Table 3.7-1 is being revised and reformatted, and Specifications 3.7.B will no longer be applicable to T.S.

Table 3.7-1.

Similarly, Specification 3.7.C is being changed to be consistent with the changes proposed to Specification 3.7.B.

The number of minimum operable channels for 1 of 2 logic Reactor Protection System (RPS) channels has been changed from one (1) to two (2).

The channels affected include Manual Reactor Trip, Nuclear Intermediate and Source

Ranges, and Low Steam Generator Water Level coincident with Steam/Feedwater Flow Mismatch.

These changes are proposed in order to be consistent with the design philosophy presented in the Surry Power Station Updated Final Safety Analysis Report (UFSAR-Reference 1).

These changes represent more stringent operability requirements for these instrumentation channels.

Notes have been added for the Nuclear Flux Source Range Reactor Trip in order to clarify the operability requirement for that instrumentation channel.

The notes are consistent with the guidance provided in Section 3/4.3.1 of NUREG-0452, Revision 4, "Standard Technical Specifications for Westinghouse Pressurized Water Reactors" (W-STS-Reference 2), for the Nuclear Flux Source Range Reactor Trip.

The minimum operable channels requirement for the low flow reactor trip and low low steam generator water level reactor trip has been modified in order to clarify the requirement in accordance with the guidance provided in Section 3/4.3.1 of Reference 2

for the low flow reactor trip and low low steam generator water level reactor trip.

The Turbine Trip function has been subdivided into its component parts.

This change is proposed to provide clarification for the Turbine Trip/RPS interface and is being made in accordance with the guidance provided in the Model Technical Specifications enclosed with the letter from Mr. H. R. Denton (NRG) to Mr. L. D. Butterfield (Westinghouse Owners Group) dated July 24, 1985 (Reference 3), for the turbine trip.

The Safety Injection Function has been changed from referencing the initiating signals for Safety Injection to specifying the logic that makes up the Safety Injection/RPS interface.

This change is proposed to provide clarification of the actual interface between the Safety Injection actuation system and the RPS and is being made in accordance with the guidance provided in Section 3/4.3.1 of Reference 2 for the safety injection input from ESF reactor trip.

A Limiting Condition of Operation for Reactor Coolant Pump Breaker Position has been added to T.S. Table 3.7-1.

The position of the Reactor Coolant Pump Breakers is an input into the RPS as described in Section 7.2 of Reference 1 and should be included with the other reactor trip instruments.

The operability requirements for this trip channel have been established in accordance with the guidance provided in Section 3/4.3.1 of Reference 2,

for the Reactor Coolant Pump Breaker Position Trip.

The control rod misalignment monitor is being deleted from T.S. Table 3.7-1 because it is not part of the reactor trip instrumentation and does not provide a

reactor trip signal.

The T.S. have requirements on flux tilt and rod misalignment, the control rod misalignment monitor is not assumed to operate to mitigate the consequences of *any accident.

T.S.

Table 3.7-1, Reactor Trip Instrument Operating Conditions, has been reformatted to be consistent with the format used in Section 3/4.3.1 of Reference

2.

More specifically for each functional unit, column entries are being added for the "Total Number of Channels",

"Channels to Trip",

and specific "Operation Action".

The column for "Degree of Redundancy" is being deleted from the table.

The "Action Statements" have been modeled after the Action Statements of Section 3/4.3.1 of the North Anna Power Station Unit 2 Technical Specifications (Reference 4).

The times for testing and maintenance that are in the Action Statements have been established in accordance with the guidance in those provisions of WCAP-10271, "Evaluation of Surveillance Frequencies and out of Service Times for the Reactor Protection Instrumentation System" (Reference 5), and Supplement 1 (Reference 6),

portions of which have been approved in the NRC's SER issued in February, 1985 (Reference 7).

The changes to the operability requirements for the reactor trip breakers, the reactor trip bypass breakers and the automatic trip logic provide clarification for operation and testing of the reactor trip breakers, reactor trip bypass breakers and automatic trip logic.

More specific guidance is being provided in the Action Statements for these items that address the time allowed for testing and maintenance.

Currently, there are no restrictions on time for maintenance and testing of reactor trip breakers, reactor bypass breakers and the automatic trip logic included in the Surry Technical Specifications.

Therefore, the proposed Limiting Conditions for Operation and Action Statements are more conservative than existing Technical Specification requirements.

These changes are modeled after the NRC Staff guidance provided in Generic Letter 85-09 "Technical Specification for Generic Letter 83-28, item 4.3" and the Westinghouse WCAP-10271, "Evaluation of Surveillance Frequencies and out of Service Times for Reactor Protection System Instrumentation."

The only exception to the guidance provided in Generic Letter 85-09 is the allowed outage time for the Automatic Trip Logic (Functional Unit 19, p 3.7-12) where we are requesting 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months rather than 6

hours for maintenance.

Surry's Reactor Protection Logic is a Westinghouse, "7100 Process Instrumentation System", which utilizes relays to establish the logic matrix.

Normally defective relays are found during performance testing (surveillance) of the logic matrix.

Prior to commencing any troubleshooting or maintenance, a pre-job briefing is required as part of our Quality Maintenance Team process, which normally takes about one half hour.

Maintenance on a

defective relay requires approximately two (2) to three (3) hours of circuit evaluation and cable tracing in a very confined space within the instrument cabinets to identify the leads to lift or jumper.

To prevent the tripping of other relays in the logic matrix, while removing the effected relay from the series circuit, jumpers (temporary modifications) are required.

The jumpers require a

written safety evaluation (50.59 determination) and approval by the Station Nuclear Safety and Operations Committee prior to installation.

Westinghouse Technical Bulletins require testing of new relays/coils prior to installation.

The bench testing prior to installation includes overtravel measurement on contact makeup and coil pickup and dropout check which take about 1

hour.

Along with the time requirements for trouble shooting and safety evaluation the actual replacement of the defective relay is very tedious and requires extreme caution, to prevent unnecessary plant trips.

In addition, Quality Control verifies the bench testing as well as the actual installation and post maintenance testing of the new relay.

Previous plant experience with relay replacement during plant operations has shown that a

minimum of 6-8 hours is required to safely replace a defective reactor protection system logic matrix relay.

Section 4.1 A Specification is being added that defines the surveillance requirements for the logic for the reactor trip system interlocks and the interlock function.

This Specification has been modeled after the requirements of Section 3/4.3.1 of Reference 4 for the logic for the reactor trip system interlocks and the interlock function.

Additionally, T.S.

Table 4.1-A is being added to this section of the Technical Specifications.

This table describes the reactor trip system interlocks and is based on information contained in Section 7.2 of Reference 1.

The surveillance requirements listed in T.S. Table 4.1-1, Minimum Frequencies for

Check, Calibrations, and Test of Instrument Channels, for the Nuclear Power Range, Nuclear Intermediate Range, Nuclear Source Range, 4

KV Voltage and Frequency, Reactor Coolant Temperature and Turbine Trip are being modified in accordance with the guidance provided in Section 3/4.3.1 of Reference 3,

for these instrument channels.

This includes the addition of a note that allows exclusion of the neutron detectors from channel calibration of the Nuclear Power, Intermediate and Source Ranges.

Surveillance requirements have been added to T.S.

Table 4.1-1 for the following channels, Safety Injection Input from

ESF, Reactor Coolant Pump Breaker Position Trip, and Steam/Feedwater Flow Mismatch Coincident with Low Steam Generator Water Level.

The surveillance requirements for these channels have been established based on the guidance provided in Section 3/4.3.1 of Reference 2, for these channels.

The surveillance requirements for the Boron Injection Tank Level have been deleted from T.S.

Table 4.1-1 reflecting a recently proposed change to the Technical Specifications that was contained in the letter from Mr.

W. L. Stewart (Virginia Electric and Power Company) to Mr. H. R. Denton (NRC), dated July 12, 1985, Serial No.85-342 (Reference 10),

The surveillance requirements for the reactor trip breaker, the manual reactor trip, and the reactor bypass breaker are modeled after Generic Letter 85-09 and provide for independent testing of the shunt and undervoltage trip devices.

In addition, a Limiting Condition for Operation has been added to allow reactor operation for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months with either the undervoltage or shunt trip device of the Reactor Trip Breaker inoperable.

The entries for weekly, semiannually, every two weeks, and after each startup are being deleted from the listing of defined abbreviations in T.S.

Table 4.1-1 because they are no longer used in the table.

The pages for T.S.

Table 4.1-l(a), Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements, and T.S.

Table 4.1-l(b),

Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements, have been renumbered so that the page numbering in this section will run consecutively.

Page T.S. 4.1-9 is being deleted, because all of the information on that page duplicates information that is contained in other portions of T.S.

Table 4.1-1.

50.92 Significant Hazards Review Pursuant to 10 CFR 50.92, we have reviewed the proposed Technical Specification changes and have concluded that this change does not involve a

significant hazards consideration.

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 significantly 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.

These proposed changes do not alter the manner in which protection is afforded nor the manner in which limiting criteria are established.

(2)

The possibility for an accident or malfunction of a different type than any evaluation previously in the safety analysis report is not being created by these proposed changes because these proposed changes do not involve any alterations to the physical plant which introduce any new or unique operational modes or accident precursors; and (3)

The margin of safety as defined in the basis for any Technical Specification is not significantly reduced by these proposed changes because the operability and performance of the reactor trip system and instrumentation is not being significantly affected by these proposed changes.

The Commission has provided examples of changes that constitute no significant hazards consideration in Federal Register, Volume 48, page 14870.

Example (i) is a

purely administrative change to Technical Specifications; for

example, a change to achieve consistency throughout the Technical Specifications, correction of an

error, or a

change in nomenclature.

Example (ii) is a

change that constitutes an additional limitation, restriction, or control not presently included in the Technical Specifications; for

example, a

more stringent surveillance requirement.

Example (vi) is a change which either may result in some increase to the probability or consequences of a previously analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in the Standard Review Plan; for example, a change resulting from the application of a

small refinement of a previously used calculational model or design method.

Example (vii) is a change to make a license conform to changes in the regulations, where the license change results in very minor changes to facility operations clearly in keeping with the regulations.

The proposed changes that reformat T.S. Table 3.7-1 and renumber the pages that contain T.S.

Table 4.1-l(a) and T.S.

Table 4.1-l(b) are similar to example (i) in that they do not modify any technical requirement and will foster document consistency.

The proposed changes to the operability requirements for the manual reactor trip, the nuclear flux intermediate range reactor trip, the nuclear source range reactor

trip, and the low steam generator water level with steam/feedwater flow mismatch reactor trip, and the addition of the requirements for the reactor coolant pump breaker position trip on T.S. Table 3.7-1 are similar to example (ii) in that they constitute either new or additional requirements.

The proposed changes that address the surveillance requirements of the logic for the reactor trip system interlocks and the reactor trip system interlock function in Section 4.1, and the surveillance requirements for the safety injection input from

ESF, the reactor coolant pump breaker position trip, and the low steam generator water level with steam/feedwater flow mismatch reactor trip on T.S. Table 4.1-1 are similar to example (ii) in that they constitute additional requirements.

The proposed changes in the Action Statements of T.S. Table 3.7-1 that address allowed outage time for testing and maintenance and the proposed change to the prior to startup note at the end of T.S. Table 4.1-1 are similar to example (vi) in that the estimated change in reactor protection system availability is very small, and it has been concluded, by WCAP-10271, "Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System" and Supplement 1, portions of which have been approved in the NRC's SER issued in February

1985, that the change in core damage frequency and risk is insignificant as a result of these changes.

The proposed changes to the operability requirements for the turbine trip, the reactor trip breakers, the reactor trip bypass breaker, the auto trip logic and the Action Statements in T.S. Table 3.7-1 are similar to example (vii) in that they are based on previously issued NRC guidance.

The proposed changes to the surveillance requirements for the nuclear power range trip, the nuclear intermediate range trip, the nuclear source range trip, the 4 KV

e voltage and frequency trips, reactor coolant temperature and the turbine trip listed in T.S. Table 4.1-1 are similar to example (vii) in that they reflect previously issued NRC guidance, including NUREG 0452, Revision 4,

"Standard Technical Specifications,"

and Generic Letter 85-09, "Technical Specifications for Generic Letter 83-28, Item 4.3."

r References

1.

Surry Power Station Updated Revision 3, June, 1985.

Final

Safety Analysis

Report,

2.

NUREG-0452 (Revision 4),

"Standard Technical Specifications for Westinghouse Pressurized Water Reactors", Fall 1981.

3.

Letter from Mr.

H.

R.

Denton (NRC) to Mr.

L.

D.

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

4.

North Anna Power Station Unit 2 Technical Specifications, through Amendment No. 49.

5.

WCAP-10271, "Evaluation of Surveillance Frequencies and out of Service Times for the Reactor Protection Instrumentation System",

January, 1983.

6.

WCAP-10271, Supplement 1, "Evaluation of Surveillance Frequencies and out of Service Times for the Reactor Protection Instrumentation System", July, 1983.

7.

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 System", February 21, 1985.

8.

9.

10.

Letter from Mr. w.

Company) to Mr.

H.

85-229A).

Letter from Mr. w.

Company) to Mr.

H.

85-211C).

Letter from Mr.

W.

Company) to Mr.

H.85-342).

L.

Stewart (Virginia Electric and Power R. Denton (NRC), September 9, 1985 (Serial No.

L.

Stewart (Virginia Electric and Power R. Denton (NRC), October 31, 1985 (Serial No.

L.

Stewart (Virginia Electric and Power R.

Denton (NRC),

July 12, 1985 (Serial No.

ML18150A109

Text

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