ML20215N757

Proposed Tech Specs,Revising Reactor Trip Sys Instrumentation to Include Independent Testing of Reactor Trip Breaker Undervoltage & Shunt Trip Attachments During Power Operation

ML20215N757
Person / Time
Site:	Beaver Valley
Issue date:	10/27/1986
From:	DUQUESNE LIGHT CO.
To:
Shared Package
ML20215N749	List: ML20215N748 ML20215N754 ML20215N757
References
NUDOCS 8611070238
Download: ML20215N757 (12)

Text

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TABLE 3.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION m

MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE m

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION h

&M

18. Turbine Trip (Above P-9)

N A.

Auto Stop Oil Pressure 3

2 2

1 7

B.

Turbine Stop Valve Closure 4

4 4

1 8

i 5

19. Safety Injection Input y

from ESF 2

1 2

1,2 1

20. Reactor Coolant Pump Breaker Position Trip (Above P-7) 1/ breaker 2

1/ breaker-1 11.

per oper-ating loop mx

21. Reactor Trip Breakers 2

1 2

1,2 1,40 0R 2

l' 2

3*,4*,5*

39,40 8*

gw

22. Automatic Trip Logic 2

1 2

1,' 2 1

1 2

1 2

3*,4*,5*

39 5g

23. Reactor Trip System s<

Interlocks E

A.

Intermediate Range 2

1 1

2 3

Neutron Flux, P-6 B.

Power Range 4

2 3

1-12 Neutron Flux, P-8 C.

Power Range 4

2 3

1 12 Neutron Flux, P-9 D.

Power Range 4

2 3

1 12 Neutron Flux, P-10 E.

Turbine Impulse 2

1 1

1 12 Chamber Pressure, P-13 8611070238 861027 PDR-ADOCK 05000334 P

PDR t

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.

(1) Trip function may be manually bypassed in this Mode above P-10.

(2) Trip function may be manually bypassed in this Mode above P-6.

ACTION STATEMENTS ACTION 1 -

With the number of OPERABLE Channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per specification 4.3.1.1, provided the other channel is OPERABLE.

ACTION 2 -

With the number of OPERABLE channels one less than the Total Number of Channels and with the THERMAL POWER level:

a.

Less than or equal to 5% of RATED THERMAL POWER, place the inoperable channel in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after increasing THERMAL POWER above 5% of RATED THERMAL POWER; otherwise, reduce thermal power to less than 5% RATED THERMAL POWER within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b.

Above 5%

of RATED THERMAL POWER, operation may continue provided all of the following conditions are satisfied:

1.

The inoperable channel is placed in the tripped condition within'1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.

The Minimum Channels OPERABLE requirement is met;

however, one additional channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.3.1.1.

3.

Either, THERMAL POWER is restricted to 1 75%

of RATED THERMAL and the Power Range,' Neutron 4

Flux Trip setpoint is reduced to 1 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; or, the QUANDRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per Specification l

4.2.4.c.

ACTION 3 -

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

BEAVER VALLEY - UNIT 1 3/4 3-5 PROPOSED WORDING

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

hours or be in HOT STANDBY within the next 6

hours; however, one channel associated with an operating loop may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />'for r

surveillance testing per Specification 4.3.1.1.

ACTION 10 Not applicable.

i ACTION 11 With less than the Minimum Number of Channels OPERABLE, j

operation may continue provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION 12 With the number of channels OPERABLE one less than-l required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and/or open the reactor trip breakers.

4 i

ACTION 39 With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or open the reactor trip breakers within the next hour.

i ACTION.40 With one of the diverse trip features (Undervoltage or i

shunt trip

-attachment) inoperable,

' restore it to OPERABLE ' status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or declare the breaker inoperable and apply ACTION 1

or ACTION 39 as applicable.

Neither breaker shall be bypassed while one i

of the diverse trip features is inoperable except for the time required for performing maintenance to restore the breaker to OPERABLE status.

i I

i l

BEAVER VALLEY - UNIT 1 3/4 3-7 (next page is 3/4 3-9) l PROPOSED WORDING l

TABLE 4 3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS tn Channel Modes in which j

Channel Channel Functional Surveillance m

Functional Unit Check Calibration Test Required

.E 1.

Manual Reactor Trip N.A.

N.A.

S/U(1), R(10)

N.A.

l E

2.

Power Range, Neutron Flux I

a.

High Setpoint S

D(2), M(3)

M 1,'2 E

and Q(6)

U b.

Low Setpoint S

N.A.

S/U(1) 2 w

3.

Power Range,-Neutron Flux, N.A.

~R M

1, 2 High Positive Rate 4.

Power Range, Neutron Flux, N.A.

R M

1, 2 High Negative Rate m

$w 5.

Intermediate Range, S

N.A.

S/U(1),

1, 2, 3*

g}

Neutron Flux M(7) 4*, 5*

un@T 6.

Source Range, Neutron Flux N.A.

N.A.

S/U(1),

2, 3*,

4*

g[

(Below P-10)

M(8) and 5*

o 7.

Overtemperature AT S

R M

1, 2 5

8.

Overpower AT S

R M

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

M 1, 2 12.

Loss of Flow-Single Loop S

R M

1 13.

Loss of Flow - Two Loops S

R N.A.

1 14.

Steam / Generator Water S

R M

1, 2 Level-Low-Low

TABLE 4 3-1, (CONTINUED)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS m

Channel' Modes in which Channel channel Functional Surveillance h

Functional Unit Check Calibration Test Required W

< 15.

Steam Feedwater Flow Mis-S R

M 1, 2 match and Low Steam Gen-g erator Water Level 16.

Undervoltage-Reactor N.A.

R M

1 Coolant Pumps cz y 17.

Underfrequency-Reactor N.A R

M 1

Coolant Pumps 18.

Turbine Trip a.

Auto Stop Oil Pressure N.A.

N.A.

S/U(1) 1, 2 b.

Turbine Stop Valve N.A N.A.

S/U(1) 1, 2 mg Closure mN

$ ^ 19.

Safety Injection Input from N.A N.A.

M(4) 1, 2 yy ESF 20.

Reactor Coolant Pump N.A N.A.

R N.A g

Breaker Position Trip 21.

Reactor Trip Breaker N.A.

N.A.

M(5, 11) 1, 2,

3*,

and S/U(1) 4*, 5*

22.

Automatic Trip Logic N.A.

N.A.

M(5) 1, 2, 3*,

4*, 5*

23.

Reactor Trip System Interlocks A.

P-6 N. A. -

N.A.

M(9) 1, 2 B.

P-8 N.A.

N.A.

M(9) 1 C.

P-9 N.A.

N.A.

M(9) 1 D.

P-10 N.A.

N.A.

M(9) 1 E.

P-13 N.A.

R M(9) 1 24.

Reactor Trip Bypass N.A.

N.A.

M(12), R(13),

1, 2, 3*

Breakers S/U(1) 4*,

5*

TABLE 4.3-1 (CONTINUED)

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

If not performed in previous 7 days.

(1)

(2)

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

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

THERMAL POWER.

Recalibrate if absolute difference > 3 percent.

Manual ESF functional input check every 18 months.

(4)

Each train tested every other month.

(5)

(6)

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(7)

Below P-10.

(8)

Below P-6.

Required only when below Interlock Trip Setpoint.

(9)

(10)

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).

(11)

The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers.

( 12 ) -

Local manual shunt trip. prior to placing breaker in service.

(13)

Automatic undervoltage trip.

BEAVER VALLEY - UNIT 1 3/4 3-13 PROPOSED WORDING

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 inititated when the parameter monitored by each channel or combination thereof exceeds its setpoint, 2) the specified coincidence logic is maintained, 3) sufficient redundancy is maintained to permit a

channel to be out of service for testing or maintenance, and 4) 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 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.

The surveillance requirements for the Manual Trip

Function, Reactor Trip Breakers and Reactor Trip Bypass Breakers are provided to reduce the possibility of an Anticipated Transient Without Scram (ATWS) event by ensuring OPERABILITY of the diverse trip features

(

Reference:

Generic Letter 85-09).

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.

The Engineered Safety Feature Actuation System interlocks perform the following functions:

Actuates turbine

trip, closes main P-4 Reactor tripped c".feedwater valves on T below

setpoint, prevents the opening ofthemainfeekw$tervalveswhichwereclosedbya safety injection or high steam generator water level

signal, allows safety injection block so that components can be reset or tripped.

Reactor not tripped - prevents manual block of safety injection.

BEAVER VALLEY - UNIT 1 B 3/4 3-1 PROPOSED WORDING

_~ -.-

3/4.3 INSTRUMENTATION 4

BASES

~

1 3/4.3.1 AND 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESF)

INSTRUMENTATION P-11 Above the setpoint P-ll automatically reinstates safety injection actuation on Low pressurizer

pressure, automatically blocks steamline isolation on high steam pressure

rate, enables safety injection and steamline isolation on (Loop Stop Valve Open) with low steamline

pressure, and enables auto actuation of the pressurizer PORVs.

Below the setpoint P-11 allows the manual block of safety injection actuation on low pressurizer pressure, allows manual block of safety injection and steamline isolation 4x1 (Loop Stop Valve Open) with Low steamline pressure and enabling steamline isolation on high steam pressure rate, automatically disables auto actuation of the pressurizer PORV's unless the Reactor Vessel Over Pressure Protection System is in service.

P-12 Above the setpoint P-12 automatically reinstates an arming signal to the steam dump system.

Below the setpoint P-12 blocks steam dump and allows manual bypass of the steam dump block to cooldown condenser dump valves.

E J

l 1

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BEAVER VALLEY - UNIT 1 B 3/4 3-la PROPOSED WORDING l

I ATTACHMENT B

Proposed Technical. Specification Change No. 129 No Significant Hazards Consideration s

[

Description of amendment request:

Change Request No. 129 would revise 'the Reactor Trip System Instrumentation to include independent i

testing of the reactor trip breaker undervoltage and shunt trip attachmenta during power operation, reactor trip bypass breakers and.

l testing the control room manual switch contacts during each refueling outage.

The following changes reflect the guidance provided by Generic Letter 85-09:

1 1.

Page 3/4 3-4, Table 3.3-1 Functional Unit 21, Reactor Trip

~

Breakers, the Applicable Modes have been revised by separating Modes 1,

2 from 3*,

4*,

5* since the Action statements have been revised to provide for the diverse trip-features.

Action 1, 40 applys during Modes 1, 2 and Action 39, 40 applies during Modes l

3*,

4*,

5*..

Functional Unit 22, Automatic. Trip Logic, the Applicable Modes have been revised by separating Modes 1, 2 from l

3*,

4*,

5*

since the Action statements have been revised to provide for the actions required in Modes 3*,

4*,

5*.

Action 1 applies during Modes.1, 2 and~ Action 39 applies during Modes 3*,.

l 4* and 5*.

2.

Page 3/4 3-5,

-Action statement 1 has been revised to allow the applicable channel to be bypassed for up to 2

hours.for surveillance testing per Specification 4.3.1.1 provided the other channel is operable.

In addition, Action statement 2.b.3 has i-been revised to reference monitoring the Quadrant Power Tilt Ratio (QPTR) in accordance with Specification 4.2.4.c.

3.

Page 3/4 3-7, Action 39 has been added to address the actions required in Modes 3*,

4*,

5*.

Action 40 has been added to address the diverse trip features.

The page number has been revised to 3/4 3-7 with a note (next page is 3/4 3-9).

Action 39 is provided to give direction for an inoperable channel in Modes i

3*,

4*,

5*..

This reflects the STS since the current action statement provides no action for an inoperable breaker in these modes.

Action 40 is intended to take credit for both trip

- features available to trip the breaker.

With one trip feature inoperable, the reliability of the breaker would be reduced, i

however, the test performed to verify that a trip-feature is inoperable (see Table 4.3-1, Functional Unit 21)- would also verify that the diverse feature is operable.

Since the diverse feature was' just-tested, this would-provide a high degree of confidence that the operable features would be able to trip the-i breaker.

Action 40 also limits bypassing with an inoperable trip l

feature.

The automatic trip of a bypass breaker is only tested at refuelings, as such, the confidence in tripping' automatically is lower than if tested every other month.

Therefore, bypassing a

breaker with an inoperable trip feature or bypassing the opposite train would further reduce system reliability and should

]

be minimized.

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Attcchment B PCg3 2 3

4.

Page 3/4 3-11, Table 4.3-1 Functional Unit 1 Manual Reactor Trip, the Channel Functional Test requirement has been' revised to include R(10).

Verification of the ' manual trip

switches, contacts and wiring was added to the Channel Functional Test on an 18-month frequency.

The methodology for testing the manual trip function for startup uses a system approach and verifies a trip of the breakers from the trip switch without verifying that both the undervoltage and shunt trip attachment operates.

The proposed test provides assurance that each trip switch is capable of activating both trip attachments on all breakers.

Since the intent of this test is to verify operable contacts and wiring, it is not necessary to physically trip the breakers but only verify voltage er ; continuity at the terminal blocks in the reactor trip switchgear.

5.

Page 3/4 3-12, Table 4.3-1 Functional Unit 21 Reactor Trip

Breaker, the Channel Functional Test requirement has been changed to include Note 11.

The Modes in which Surveillance is Required has been revised to 1, 2, 3*,

4*,

5* for Functional Units 21 and 22.

Verification of the operability of the shunt and undervoltage trip attachments was added to the monthly Channel Functional Test.

The modification to the reactor trip breakers, incorporated in response to Generic Letter 83-28, causes both the undervoltage and shunt trip devices to receive the automatic trip signal, these functions are verified during surveillance testing.

Functional Unit 24 Reactor Trip Bypass Breakers has been added, the Channel Functi,onal Test requirement is M(12), R(13), S/U(1),

and the Modes in'which Surveillance is Required is 1, 2, 3*,

4*,

5*.

Verification of the bypass breaker operability is provided by the addition of the bypass breaker surveillance requirements.

Prior to bypassing a

reactor trip breaker the bypass breaker shunt trip attachment shall be tested by the local trip switch.

This test along with the manual trip switch check during refueling provides some assurance that a manual trip signal will open the bypass breaker.

Since the undervoltage trip attachment cannot be tested during normal operation, operability will be verified on an 18-month frequency.

This test is intended to verify that an automatic trip signal willt trip the bypass breaker.

A test of the manual trip function performed prior to startup will verify that the breaker trips (without considering independent undervoltage or shunt trip) on a manual trip signal.

6.

Page 3/4 3-13, Notation,' Notes 10, 11, 12 and 13 have been added to address applicability to the above revisions.

~

Attcchment B i

POg3 3 7.

Page B

3/4 3-la, Bases Section 3/4.3.1 has been revised to describe the basis for the reactor trip and bypass breaker testing requirements.

The Bases'have been revised to provide a future reference and describe why these modifications. were required (i.e.,

to reduce the probability of Anticipated Transients Without scram (ATWS) events).

1 Basis for no significant hazards determination: ' Based on the criteria for determining whether a significant hazards consideration exists as setforth in 10CFR50.92(c), plant operation in accordance with the proposed amendment would not:

i 1)

Involve a

significant increase in the probability of occurrence or the consequence of an accident previously evaluated because:

The intent of the proposed changes is to decrease the possibility of an ATWS event by additional testing to provide adequate

. verification of reactor. trip and bypass breaker operability.

UFSAR Section 7.2.2.2.1 discusses single failures of the Solid State Protection System logic train and notes that since it is 1/2' coincidence at the system level, a single failure will not prevent a

protective action.

The proposed change will require testing the manual trip switch contacts and wiring.

To perform this test voltage or continuity measurements will be performed on the. trip attachments at the switchgear terminal blocks.

IE Information. Notice 85-18 and Westinghouse Technical Bulletin NSID-TB-85-16 note that-shorting across the Undervoltage (UV) coil could cause the UV output board.in the protection system to fail in a -manner that would prevent an automatic reactor trip.

The modification to add the automatic shunt trip to the reactor j

trip breakers also adds test points with in-line resistors to the l

switchgear terminal blocks cross the undervoltage coil.

The

.in-line resistors would limit current such that shorting across the test points should not cause a

failure of the UV board.

However, test points are not being added to the bypass' breakers at. this time,.thus during the manual trip switch test the possibility exists for the UV boards to fail on both logic trains.

Reviews of this potential problem have resulted in recommendations to check.the protection system after performance of the manual trip switch test to verify UV board operability prior to startup.

For ATWS events, studies have shown that the limiting event is a loss of normal-feedwater

(

Reference:

NSAC/91 and WCAP-10858).

This transient is evaluated in UFSAR Section 14.1.8 and assumes a reactor trip. occurs.

The proposed change incorporates additional testing to ensure the reactor trip breakers are operable and this enforces the validity of the assumption that a reactor trip will occur.

Therefore, it is concluded that these proposed changes do not involve a

significant increase in the probability or consequences of a previously evaluated accident.

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Attcchment B P 2 9 3 4' i

Action 2.b.3 has been revised to include reference to Specification 4.2.4.c for monitoring the QPTR.

This will reflect the Standard Technical Specifications and. eliminate potential confusion when interpreting the action requirements to ensure the QPTR is monitored using the incore moveable detectors.

This is i

an administrative change and does not affect the UFSAR accident 4

analysis.

l (2). Create the possibility of a

new or different kind of accident

~

from any accident previously evaluated because:

These changes involve testing methods'for the reactor trip and bypass breakers that are performed'during shutdown or with the breaker bypassed, therefore, challenges to. the system during testing will be minimized.

A concern has been identified, such that, the manual trip switch test could cause failure of the UV boards in both solid state protection systems.

However, as discussed above, reviews have been performed to address this problem with recommendations to verify UV board operability prior to startup.

j Another concern has been identified, such that, the additional testing may cause increased wear. on the breakers resulting in failures.

Life cycle' testing of the-breakers has been performed for.

.the Westinghouse owner's Group.by Westinghouse and is 1.

documented in WCAP-10852.

These. tests cycled breakers through 2500 trip operations with no significant performance degradation, therefore, a. trip attachment replacement interval of 1250 cycles has-been recommended.

An estimate of breaker cycles was provided to the NRC by letter dated June 17, 1985 and determined that 1250 breaker cycles would be achieved in about 11 years.

Therefore, it has been determined that these changes will not create the possibility of a

new or different kind of accident from those i

described in the safety analysis report.

l (3) Involve a

significant reduction in the margin of safety because:

l The additional testing will increase overall confidence that the reactor trip and bypass breakers are capable of tripping the reactor.

With any test there is concern for test induced l

failures, however, those concerns have been addressed above and shown to be safe therefore the margin of safety will not be reduced.

Conclusion-The proposed changes will aid in verifying the operability of the reactor trip and bypass breakers and thus to reduce the possibility of an ATWS -event.

Therefore, based on the above, it is proposed to characterize the change as involving no significant hazards consideration.

_ -_-