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Amend 94 to License DPR-36,revising Tech Specs Re Operability & Surveillance for New Noble Gas Effluent monitors,high-range Monitors & Water Level Monitors in Containment
	ML20197D418
Person / Time
Site:	Maine Yankee
Issue date:	03/26/1987
From:	Thadani A; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20197D405	List: ML20197D402; ML20197D418; ML20197D424;
References
	NUDOCS 8704080317
	Download: ML20197D418 (13)
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'o,, UNITED STATES l 8 o ' NUCLEAR REGULATORY COMMISSION

.h $ WASHINGTON, D. C. 20066 kg*****,of l.

FAINE YANKEE ATOMIC POWER COMPANY L DOCKET NO. 50-309 MAINE YANKEE ATOMIC POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE-4 Amendment No. 94 License No. DPR-36

1. The Nuclear Regulatory Comission (the Comission) has found that:

A. The applications for amendment by Maine Yankee Atomic Power

. Company (the licensee) dated October 7, 1982 and March 4, 1985,.

supplemented April 26, 1985, comply with the standards and require-ments of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in confonnity with the applications the provisions of the Act, and the rules and regulations of the Comission; C. There is reasonable assurance (i) that the activities aui.horized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such. activities will be conducted in compliance with the Comission's regulations; D. -The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and ,

E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

T B704080317 FI70326 PDR ADOCK 05000309 P PDR _

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2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and-paragraph 2.B(6)(b) of Facility Operating License No. DPR-36 is hereby amended to read as follows: ;

(b) -Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 94 , are hereby-incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION h[A h StT Ashok C Yh E ni, Director PWR Pr ject Directorate #8 Division of PWR Licensing-B

Attachment:

Changes to the Technical Specifications s Date of Issuance: March 26, 1987

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l ATTACHMENT TO LICENSE AMENDMENT N0. 94 FACILITY OPERATING LICENSE NO. DPR-36 DOCKET NO. 50-309 Revise Appendix A as follows:

Remove Pages Insert Pages 3.9-1 3.9-1 through through 3.9-6 3.9-7

-4.1-10 4.1-10 4.1-12 - 4.1-12 4.1-13 4.1 .

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3.9 OPERATIONAL SAFETY INSTRUMENTATION, CONTR0L' SYSTEMS, AND ACCIDENT MONITORING SYSTEMS t Applicability:

Applies to plant instrumentation system.

Obj ective_:

To specify the conditions of the plant instrumentation and control systems necessary to ensure reactor safety.

Specification:

A. Reactor Protection System The minimum' number of operable reactor protection system functional unit channels shall be as shown in Table 3.9-1 when the reactor is at hot standby (Condition 6) or higher. A channel in the tripped condition is considered operable.

Remedial Action:

With less than the minimum number of channels in a functional unit operable, restore the minimum number to operable within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

This remedial action does not apply to the manual trip functional unit.

Exception:

1. One channel in each functional unit may be made inoperable for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for the purpose of surveillance trouble shooting or maintenance.

2. One channel in each functional unit may be made inoperable after the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months period while actively

performing required surveil-lance or actively

performing maintenance, trouble shooting, or surveillance activities associated with restoring the channel to operable status.

B. Engineered Safeguards Features Actuation System The minimum number of operable channels per engineered safeguards system actuation subsystem shall be as shown in Table 3.9-2 whenever automatic initiation of engineered safeguards systems is required to be operable. A channel in the tripped condition is considered operable.

Short delays such as encountered during shift turnovers, meal breaks or brief rest breaks not to exceed 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months shall not require returning an inoperable channel to the tripped condition.

3.9-1 Amendment No. 52,65,8J, 94

3.9 OPERATIONAL SAFETY INSTRUMENTATION, CONTROL SYSTEMS, AND ACCIDENT MONITORING SYSTEMS (Continued)

B. Engineered-Safeguards Features Actuation System (Continued)-

Remedial Action:

With'less than the minimum number of channels in a subsystem operable, restore the minimum number to operable within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . This remedial action does not apply to manual actuation devices.

Exception:

1. .One subsystem may be removed from service for maintenance or on-line testing for a period not to exceed 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

2. One channel in each subsystem unit may be made inoperable after the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months period only while actively

performing required surveillance or actively

performing maintenance, trouble shooting, or surveillance activities associated with restoring the channel to operable status.

C. Accident Monitoring System -

Wheneser the reactor is at power, the minimum Accident Monitoring Instrumentation listed in Table 3.9-3 shall be operable.

Remedial Action:

In the event the number of operable accident monitoring instrumenta-tion channels falls below the Minimum Channels Operable requirements in Table 3.9-3, either restore the inoperable channel (s) to operable status within 7 days or submit a report to the NRC describing plans to restore the channel (s) to operable status.

Exception:

An instrument may be removed from service for maintenance or on-line testing for a period not to exceed 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Basis:

The reactor protection system is designed to rapidly shut down the reactor automatically'in the event that selected nuclear steam supply system conditions deviate from predetermined ranges. The system acts to prevent violation of safety limits, and maintains operation within bounds assumed in safety analyses.

The system is designed for high reliability. The design incorporates sufficient redundancy to assure that surveillance testing, trouble shooting.

Short delays such as encountered during shift turnovers, meal breaks or brief rest breaks not to exceed 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months shall not require returning an inoperable channel to the tripped condition.

3-9.2 Amendnent No. 52,8J,65,8J. 94

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3.9 : OPERATIONAL SAFETY INSTRUMENTATION, CONTROL' SYSTEMS, AND ACCIDENT MONITORING SYSTEMS (Continued)

Basis (Continued)

H.

l= and maintenance activities can be accomodated with margin. - This speci-

. fication establishes the limiting conditions -for operation necessary to maintain adequate system reliability under anticipated operating conditions.

! Each-automatic trip function usually-operates in a two-out-of-four coincidence mode. - One' trip channel at a time within a functional unit can be bypassed in which case a trip signal generated by that channel will not

contribute to .the coincidence necessary to produce a reactor _ trip. In i this case, a coincidence trip of two of the three remaining channels will produce a reactor trip. Bypass of a channel is limited to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to

. restrict the- time when the degree of-redundancy is reduced to one.

l An inoperable channel, i.e., one that cannot automatically develop a trip

. signal as_ required, may be restored to operability by placing it in the

- trip mode. In this case, only one additional trip of the three remaining channels in the functional unit is required to produce a reactor trip.

This restores the degree of redundancy to two. An inoperable channel made

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operable by placing it in 'the trip mode may be subsequently rendered inop-erable by bypassing for maintenance, trouble shooting, or surveillance when such activities are best perfomed under this condition. A clarifica-tion is added to the exception which allows the channel to remain bypassed

, for short periods such as shift turnovers, meal and rest breaks.

) This flexibility is permitted to allow a malfunctioning channel to be ,

2 repaired and restored to service without necessitating a shutdown and to avoid long periods of operation of a repairable channel in the trip mode.

Although no credit is taken for the high rate-of-change-of-power channel

! -in the Maine Yankee accident analysis, operability of this channel at low p power levels provides backup assurance against excessive power rate increases. Temperature feedback effects protect against excessive power i rate increases at higher power levels.

Redundant sensors and logic are. provided for the initiation of all engineered safeguards systems. In both the containment isolation and containment spray

, systems, two identical ~ subsystems are used in each system. In the safety

. injection actutation systems diverse sensors are used for the initiation of

! two identical subsystems. Each of these three engineered safeguards systems j may be operated, as shown in Table 3.9-2 and the associated exceptions, without jeopardizing safeguards initiation. One subsystem may be removed r from service for a limited time for purposes of maintenance or testing

! because it is highly unlikely that a failure of the operable subsystem would i occur concurrent with an accident requiring engineered safety features

, actuation. The second exception allows bypassing beyond the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months limit i for testing, trouble shooting or maintenance that must be performed or is

best performed under bypass conditions. A clarification is added to the 4

l 3.9-3 Amendment No. 52,$J,65,$J 94 1

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3.9 OPERATIONAL SAFETY JNSTRUMENTATION, CONTROL SYSTEMS, AND ACCIDENT MONITORING SYSTEMS (Continued)

Basis (Continued) exception which allows the channel to remain bypassed for short periods to accommodate matters such as shift turnovers, meal breaks, or brief rest breaks. This flexibility is permitted to allow a malfunctioning sensor to be repaired and restored to service without necessitating a shutdown and to avoid long periods of operation of a repairable channel in the trip mode.

The safety injection actuation system is initiated by two out of four pressure sensor channels. When-three channels are operable, the degree of redundancy, as defined in the definitions section, is one. This degree of redundancy is also provided when two channels are operable with a third sensor placed in a configuration which simulates the tripped condition.

The minimum number of operable channels for the accident monitoring instru-mentation is given in Table 3.9-3. The accident monitoring instrumentation is used to evaluate and aid in mitigating the consequences of an accident.

. 3.9-4 Amendment No. $J.94

1 e.

TABLE 3.9-1 INSTRUMENTATION OPERATING' REQUIREMENTS FOR REACTOR PROTECTIVE SYSTEM Minimum Operable No. Functional Unit Channels (a) Bypass Conditions 1 Manual.(tripbuttons) 1 set None 2 High Rate-of-Change Power 4(c) Below 10-4% and Above 10% of Rated Power (b) 3 High Power Level 4(c) None 4 Thermal Margin / Low 4(c) Below 10% of Rated Power (b) 5 High Pressurizer Pressure 4(c) None 6 Low Reactor Coolant Flow 4(c) Below 2% of Rated Power (b) 7 Low Steam Generator Water 4(c) None Level 8 Low Steam Generator Pressure 4(c) 100 psi Above the Trip Setpoint 9 High Containment Pressure 4(c) None 10 Axial Flux Offset 4(c) Below'15% of RatedPower(b)

(a) The minimum degree of redundancy is one, except for manual trip which has a minimum degree of redundancy of zero.

(b) As indicated on Nuclear Instrumentation Channels.

(c) A channel in tripped position is considered operable.

3.9-5 Amendment No. 19,55, 94

v.

TABLE 3.9-2 Instrumentation Operating Requirements for Engineered Safeguards Systems .

' Minimum Operable Channel- Bypass Initiation No. Functional Unit Per Subsystem Conditions Set Points 1 Safety Injection:

A. Manual 1 High Containment

Less than B. 3(a)-

Pressure 5 psig Low Pressurizer

Greater C. 3(a)

Pressure than 1585 psig 2 ' Containment Spray:

A. Manual 1 B. Containment High

Less than I 3/ set (a)(b)

Pressure 20 psig 3 Containment Isolation:

A.

Manual 1 B. Containment High

Less than 3/ set (a)(b) l Pressure 5 psig (a) A channel which is placed in a configuration which simulates tripped is considered operable.

_(b) Each subsystem is initiated by two out of three pressure sensor channels.

The minimum degree of redundancy in'each subsystem is one.

Reactor coolant pressure less than 1685 psig.

3.9-6 Amendnent No. $J,65, 94

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o TABLE'3.9-3 Accident Monitoring Instrumentation Instrument Minimum Channels Operable

1. Containment High Range Radiation Area Monitor 2 2.. Primary Vent Stack High Range Noble Gases Effluent Monitor 1

3. Pressurizer Water Level 1

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4. Auxiliary Feedwater Flow Rate 1 per Steam Generator

5. Reactor Coolant System Subcooling Margin Monitor 1

6. PORV Position Indicator (Acoustic Flow Sensor) 1/ valve

7. Safe'ty Valve Position Indicator (Acoustic Flow Sensor) 1

8. Containment Water Level Monitor 1

9. Containment Hydrogen Monitor 1

10. Containment Pressure Monitor. 1 3.9-7 Amendment No. 57.65.SJ 94 p

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Table 4.1-3 Minimum Frequencies for Checks, Calibrations and -

Testing of Miscellaneous Instrumentation and Controls

, Surveillance Channel Description Function Frequency Surveillance Method

1. Reed Switch Rod Position Check S Compare rod position indication from-

Indication System the Reed Switch Position Indication System with that of the Pulse Count -

ing Position Indication System when-ever the reactor is critical and the

plant computer is available.

l 2. Pulse Counting Rod Position Check S Compare rod position indication from

+

Indication System the Pulse Counting Position Indication

, System with that of the Reed Switch Position Indication System whenever L

o the reactor is critical and the plant computer is available.

3. Area, Process and Effluent a. Test D(6) a. Internal test signals used to Monitors verify instrument operation.

b. Calibrate R b. Exposure to known external radia-tion source.

4. Emergency Plant Radiation a. Check D(3) a. Internal radiation source provides g Containment High Range continuous signal verifying g instrument operations

[ b. Calibrate R b. Exposure to known radiation source.

? -

, 5. Environmental Monitors a. Check M a. Operational check.

! E b. Calibrate A b. Verify airflow indicator.

. .E

', g 6. Pressurizer Level a. Check S(3) a. Comparison of independent level readings.

b. Calibrate R b. Known. differential pressure applied to sensor.

Table 4.1-3 (Continued)

Surveillance Channel Description Function Frequency Surveillance Method

13. Safety Valve Position a. Calibrate R Apply known frequency to sensor.

Indicator (Acoustic

< Flow Sensor)

14. PORV Actuation Circuit a. Check M(3)(4) a. Bistable trip test.

b._ Calibrate R b. Apply known pressure to sensors.

15. Pressurizer Power Operated a. Calibrate R Apply known pressure to the pressure

Relief - Variable Pressure sensor. Manual actuation of each Setpoint PORY to verify solenoid operation.

m

b. Test R(5)

h 16. Purge Valve Isolation on i

High Radiation Signal

a. High Containment a. Calibrate R Exposure to an external radiation Gaseous Activity source.

b. High Primary Vent Stack b. Calibrate R Exposure to an external radiation Gaseous Activity source.

'l g 17. Containment Hydrogen a. Check W a. Verify expected values on readout.

g Monitors

b. Calibrate SA b. Apply gasses containing known con-

$a centration of H2 with balance N2 .

[ 18. Primary Vent Stack High a. Check D(6) a. Expose detector to internal check o Range Noble Gas Monitor source.

g b. Calibrate R b. Apply known radiation source to y detector.

w h 19. Containment Water Level a. Check M a. Compare readings of'the two

y channels for consistency.

w

b. Calibrate R b. Adjust indicator and verify 2 readouts based on signals.

Table 4.1-3 (Continued)

Surveillance Channel Description Function Frequency Surveillance Method

20. Containment Pressure a. Check M. a. Compare readings of the.two channels for consistency. -

b. Calibrate R b. Apply known pressure to sensors.

(1) Not required unless the reactor is in the power operating condition.

(2) Not required during plant startup and shutdown periods.

(3) Not required when plant is in the cold shutdown, refueling shutdown or refueling operations mode, f,

(4) Must be performed within 30 days prior-to attaining a power operating condition.

(5) Must be performed prior to cooldown below MPT plus margin.

(6) When required to be operable.

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ML20197D418

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