ML20078C439

Amend 174 to License DPR-51,re Changes to TS Concerning RPS to Permit Plant to Operate Indefinitely W/One RPS Channel in Bypass

ML20078C439
Person / Time
Site:	Arkansas Nuclear
Issue date:	10/24/1994
From:	Beckner W Office of Nuclear Reactor Regulation
To:
Shared Package
ML20078C441	List: ML20078C439 ML20078C450
References
NUDOCS 9410310252
Download: ML20078C439 (8)

Text

.

Yp ets

,A u

O t

[

r E

UNITED STATES 3;g]ik,f NUCLEAR REGULATORY COMMISSION

?

WASHINGTON, D.C. 2055M001 ENTERGY OPERATIONS INC.

DOCKET NO. 50-313 ARKANSAS NUCLEAR ONE. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 174 License No. DPR-51 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Entergy Operations, Inc. (the licensee) dated January 13, 1994, complies with the standards and requirements 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 conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance:

(i) that the activities authorized 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 Commission's regulations; D.

The issuance of this license 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 Commission's regulations and all applicable requirements have been satisfied.

9410310252 941024 PDR ADOCK 05000313 P

PDR

. 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No. DPR-51 is hereby amended to read is follows:

2.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.174 are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

The license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION QPLO f.yb-William D. Beckner, Director Project Directorate IV-1 Division cf Reactor Projects - III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: October 24, 1994

l ATTACHMENT TO LICENSE AMENDMENT NO.174 FACIU TY OPERATING LICENSE NO OPR-51 DOCKET NO. 50-313 i

Replace the following pages of the Appendix "A" Technical Specifications with i

the attached pages.

The revised pages are identified by Amendment number and contain vertical lines indicating the area of change REMOVE PAGES INSERT PAGES 42 42 43 43 1

43a 43a 43b 43b 45e 45e l

l

3.5 INSTRUMENTATION SYSTEMS 3.5.1 Operational Safety Instrumentation Appl i cabili ty Applies to unit instrumentation and control systems.

Objectives To delineate the conditions of the unit instrumentation and safety circuits necessary to assure reactor safety.

Specifications 3.5.1.1 Startup and operation are not permitted unless the requirements of Table 3.5.1-1, columns 3 and 4 are met.

3.5.1.2 In the event the number of protection channels operable falls below the limit given under Table 3.5.1-1. Columns 3 and 4 operation shall be limited as specified in Column 5.

3.5.1.3 for on-line testing or in the event of a protection instrument or channel failure, a key operated channel bypass switch associated with each reactor protection channel may be used to lock the l

j channel trip relay in the untrippeo state as indicated by a light.

Only one channel shall be locked in the untripped state or contain inoperable functions in the untripped state at any one time.

In the event more than one protection channel contains inoperable functions in the untripped state, or a protection channel or function becomes inoperable concurrent with another protection channel locked in the untripped state, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> implement the actions required by Table 3.5.1-1 Note 6.

Only one channel bypass key shall be accessible f ar use in the control room.

While operating with an inoperable functien unbypassed in the untripped state, the remaining RPS key opetated channel bypass switches shall be tagged to prevent their op9 ration.

3.5.1.4 The key operated shutdown bypass switch associated with each reactor protection channel shall not be used during reactor ower operation except during channel testing.

3.5.1.5 During startup when the intermediate range instruments come on scale, the overlap between the intermediate range and the source range instrumentation shall not be less than one decade.

If the overlap is less than one decade, the flux level shall be maintained in the source range until the one decade overlap is achieved.

3.5.1.6 In the event that one of the trip devices in either of the sources supplying power to the control rod drive mechanisms fails in the untripped state, the power supplied to the rod drive mechanisms through the failed trip device shall be manually removed within 30 minutes following detection.

The condition will be corrected and the remaining trip devices shall be tested within eight hours following detection.

If the condition is not corrected and the remaining trip devices are not tested within the eight-hour period, the reactor shall be placed in the hot shutdown condition within an additional four hours.

Amendment No. 440,174 42

~

l Bases Every reasonable effort will be made to maintain all safety instrumentation in operation.

A startup is not permitted unless the requirements of Teble 3.5.1-1, Columns 3 and 4, are met.

Operation at rated power is permitted as long as the systems have at least the redundancy requirements of Column 4 (Table 3.5.1 1).

This is in agreement w'*h redundancy and single failure criteria of IEEE 279 as described in FSAR, Section 7.

There are four reactor protection channels.

Normal trip logic is two out of four.

Required trip logic for the power range instrumentation channels is two out of three. Minimum trip logic on other instrumentation channels is one out of two.

The four reactor protection channels were provided with key operated bypass switches to allow on-line testing or maintenance on only one channel at a time during power operation.

Each channel is provided with alarm and lights to indicate when that channel is bypassed.

There will be one reactor protection system channel bypass switch key permitteo in the control room.

Upon the discovery of inoperable functions in any one reactor protection channel, the effect of the failure on the reactor protection system and other interconnected systems is evaluated.

The affected reactor protection channel may be placed in channel bypass, remain in operation in a degraded condition, or placed in the tripped condition as determined by operating conditions and management judgment.

This action allows placing the plant in the safest condition possible considering the extent of the failure, plant conditions, and guidance from plant management.

Should the failure in the reactor protection channel prohibit the proper operation of another system, l

the appropriate actions for the effected system are implemented.

Administrative controls are established to preclude placing a reactor protection channel in channel bypass when any other reactor protection channel contains an inoperable function in operation in the untripped state.

Each reactor protection channel key operated shutdown bypass switch is provided with alarm and lights to indicate when the shutdown bypass switch is being used.

The source range and intermediate range nuclear flux instrumentation scales overlap by one decade.

This decade overlap will be achieved at 10-10 amps on the intermediate range scale.

The ESAS employs three independent and identical analog channels, which supply trip signals to two independent, identical digital subsystems.

In order to actuate the safeguards systems, two out of three analog channels must trip.

This will cause both digital subsystems to trip.

Tripping of either digital subsystem will actuate all safeguards systems associated with that digital subsystem.

Because only one digital subsystem is necessary to actuate the safeguards systems and these systems are capable of tripping even when they are being tested, a single f ailure in a digital subsystem cannot prevent protective action.

Amendment No. 174 43

Removal of a module required for protection from a RPS channel will cause that channel to trip, unless that channel has been bypassed, so that only one channel of the other three must trip to cause a reactor trip.

Thus, sufficient redundancy has been built into the system to cover this situation.

Removal of a module required for protective action, from an analog ESAS channel will cause that channel to trip, so that only one of the other two must trip to actuate the safeguards systems.

Removal of a module required for protective action from a digital ESAS subsystem will not cause thot subsystem to trip.

The fact that a module has been removed will be continuously annunciated to the operator.

The redundant digital subsystem is still sufficient to indicate complete ESAS action.

The testing schemes of the RPS, the ESAS, and the EFIC enables complete system testing while the reactor is operating.

Each channel is capable of being tested independently so that operation of individual channels may be evaluated.

The EFIC is designed to allow testing during power operation.

One channel may be placed in key locked " maintenance bypass" prior to testing.

This will bypass only one channel of EFW initiate logic.

An interlock feature prevents bypassing more than one channel at a time.

In addition, since the EFIC receives signals from the NI/RPS. the maintenance bypass from the NI/RPS is interlocked with the EFIC.

If one channel of the NI/RPS is in maintenance bypass, only the corresponding channel of EFIC may be bypassed.

Prior to placing a channel of EFIC in maintenance bypass, any NI/RPS channel containing inoperable functions in the untripped state is evaluated for its effect on EFIC.

Only the EFIC channel corresponding to the NI/RPS channel containing the inoperable function may be placed in maintenance bypass unless it can be shown that the failure in the NI/RPS channel has no effect on EFIC actuation, actions are taken to ensure EFIC actuation when required, or the appropriate actions of Table 3.5.1-1 are implemented.

The EFIC can be tested from its input terminals to the actuated device controllers.

A test of the EFIC trip logic will actuate one of two relays in the controllers.

Activation of both relays is required in order to actuate the controllers.

The two relays are tested individually to prevent automatic actuation of the component.

The EFIC trip logic is two (one-out of-two).

Reactor trips on loss of all main feedwater and on turbine trips will sense the start of a loss of OTSG heat sink and actuate earlier than other trip signals.

This early actuation will provide a lower peak RC pressure during the initial over pressurization following a loss of feedwater or turbine trip event. The LOFW trip may be bypassed up to 10% to allow sufficient margin for bringing the MFW pumps into use at approximately 7%.

The Turbine Trip may be bypassed up to 45% based on BAW-1893, " Basis for Raising Arming Threshold for Anticipatory Reactor Trip on Turbine Trip,"

October 1985 and the NRC Safety Evaluation Report for BAW-1893 issued from Mr. D. M. Crutchfield to Mr. J. H. Taylor via letter dated April 25, 1986.

The Automatic Closure and Isolation System (ACI) is designed to close the Decay Heat Removal System (DHRS) return line isolation valves when the Reactor Coolant System (RCS) pressure exceeds a selected fraction of the DHRS design pressure or when core flooding system isolation valves are opened.

The ACI is designed to permit manual operation of the DHRS return line isolation valves when permissive conditions exist.

In addition, the ACI is designed to disallow manual operation of the valves when permissive conditions do not exist.

Amendment No. M,60,41 M.M4,4M,174 43a

I Power is normally supplied to the control rod drive mechanisms from two separate parallel 480 volt sources.

Redundant trip devices are employed in each of these sources.

If any one of these trip devices fails in the untripped state, on-line repairs to the failed device, when practical, will be made and the remaining trip devices will be tested.

Four hours is ample time to test the remaining trip devices and, in many cases, make on-line repairs.

The Degraded Voltage Monitoring relay settings are based on the short term starting voltage protection as well as long term running voltage protection.

The 4.16 KV undervoltage relay setpoints are based on the allowable starting voltage plus maximum system voltage drops to the motor terminals, which allows approximately /81 of motor rated voltace at the motor terminals.

The 460V undervoltage relay setpoint is based on long term motor voltage requirements plus the maximum feeder voltage drop allowance resulting in a 92% setting of motor rated voltage.

The OPERABILITY of the accident monitoring instrumentation ensures that sLfficient information is available on selected plant parameters to monitor and j

assess these variables during and following an accident.

This capability is 1

consistent with the recommendation of Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During 1

and Following an Accident," December 1975 and NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations."

l The OPEkABILITY of the chicrine detection system ensures that sufficient capability is available to promptly detect and initiate protective action in the

)

event of an accidental chlorine release.

This capability is required to protect control room personnel and is consistent with the recommendations of Regulatory Guide 1.95. " Protection of Nuclear Power Plant Control Room Operators against an Accidental Chlorine Release," February 1975.

The subcooled margin monitors (SMM), and core-exit thermocouples (CET). Reactor Vessel Level Monitoring System (RVLMS) and Hot leg Level Measurement System (HLLMS) are a result of the Inadequate Core Cooling (ICC) instrumentation j

required by Item II.F.2 NUREG-0737. The function of the ICC instrumentation is to increase the ability of the plant operators to diagnose the approach to and recovery from ICC.

Additionally, they aid in tracking reactor coolant inventory.

These instruments are included in the Technical Specifications at j

the request of NRC Generic Letter 83-37 and are not required by the accident analysis, nor to bring the plant to cold shutdown conditions.

The Reactor Vessel Level Monitor is provided as a means of indicating level in the reactor 1

vessel during accident conditions. The channel operability of the RVLMS is defined as a minimum of three sensors in the upper plenum region and two sensors in the dome region operable. When Reactor Coolant Pumps are running, all except the dome sensors are interlocked to read " invalid" due to flow induced variables that may offset the sensor outputs.

The channel operability of the HLLMS is defined as a minimum of one wide range and any two of the narrow range transmitters in the same channel operable.

If the equipment is inaccessible due to health and industrial safety concerns (for example, high radiation area, low oxygen content of the containment atmosphere) or due to physical location of the i

fault (for example, probe f ailure in the reactor vessel), then operation may continue until the next scheduled refueling outage and a report filed.

Amendment No. M.M,M, m m, W.174 43b

TABLE 3.5.1-1 (Cont'd)

Notes:

1. Initiate i shutdown using normal operating instructions and place the reactor in the hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if the requirements of Columns 3 and 4 are not met.

2. When 2 of 4 power range instrument channels are greater than 10% rated power, hot shutdown is not required.

3. When 1 of 2 intermediate range instrument channels is greater than 10-10 amps, hot shutdown is not required.

4. For channel testing, calibration, or maintenance, the minimum number of operable channels may be two and a degree of redundancy of one for a maximum of four hours, after which Note 1 applies,

5. If the requirements of Columns 3 or 4 cannot be met within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, place the reactor in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

6. The minimum number of operable channels may be reduced to 2, provided that the system is reduced to 1 out of 2 coincidence by tripping the remaining channel.

Otherwise, the actions required by Column 5 l

shall apply.

7. These channen initiate control rod withdrawal inhibits not reactor trips at -10% rated power.

Above 10% rated power, those inhibits are bypassed.

8. If any one component of a digital subsystem is inoperable, the entire digital subsystem is considered inoperable.

Hence, the associated safety features are inoperable and Specification 3.3 applies.

9. The mininium number of operable channels may be reduced to one and the minimum degree of redundancy to zero for a maximum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, after which Note I applies.

10. With the number of operable channels less than required, either restore the inoperable channel to operable status within 30 days, or be in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

11. With the number of operable channels less than required, !solate the electromatic relief valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, otherwise Note 9 applies.

Amendment No. 50.60,91,174 45e

-