ML20236C762

Forwards Listing of Changes,Tests & Experiments Completed During Month of Feb 1989

ML20236C762
Person / Time
Site:	Quad Cities
Issue date:	03/01/1989
From:	Robey R COMMONWEALTH EDISON CO.
To:	Office of Nuclear Reactor Regulation
References
RAR-89-11, NUDOCS 8903220305
Download: ML20236C762 (13)

Text

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Commonwealth Edison C,

' Ouad Cities Nuclear Power Station s

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RAR-89-11 March 1, 1989 Director of Nuclear Reactor Regulations U. S. Nuclear Regulatory Commission Mail Station P1-137 Washington, D. C.

20555 En.les-d pl. ease find a listing of those changes, tests, and experiments uring the month of February, 1989, for Quad-Cities Station complet=* v Units 1 and 2, DPR-29 and DPR-30.

A summary of the safety evaluation is being reported in compliance with 10 CFR 50.59.

Thirty-nine copies are provided for your use.

Respectfully, COMMONWEALTH EDISON COMPANY QUAD-CITIES NUCLEAR POWER STATION k

R. A. Rooey Services Superintendent RAR/vmk/eb Enclosure cc:

R. Stohls T. Watts /J. Galligan I

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SPECIAL TEST 1-124 l

t Special Test No. 1-124 was completed on February 8, 1989. The purpose of this test was to verify response of control rod drive E-9 to Reactor Manual Control System and execution of proper inward and outward Rod Notch motion.

1.

The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report is not increased because rod manipulations were conducted in accordance with existing approved rod manual control procedures.

In the event of a target position overshoot, the subject control rod would be fully inserted to the O

full-in position. A qualified Nuclear Engineer would direct l

recovery of the control rod if overshoot occurs.

I 2.

The probability for an accident or malfunction of a different type than any'previously evaluated in the Final Safety Analysis Report is not created because the possible range of rod movement has been analyzed in developing B.P.W.S. and control rod manual control procedures.

3.

The margin of safety, as defined in the basis for any Technical Specification, is not reduced because the range of control rod movement required in the special test is within the control rod manipulations allowed by Technical Specifications. No condition which renders control rods inoperable is specified or required to conduct the test.

Performance of the test is not expected to result in control rod inoperability.

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Modification M-4-1-86-12A a

Description.

The purpose of this modification is to satisfy commitments.for_ Regulatory Guide 1.97.

The modification provides redundant loops for measurement and display of reactor water level and pressure'. New reactor water level'indicat1on is' located on a new three pen recorder on the 901-5 panel and a new reactor water level indicator on the 901-3 panel. Reactor water level indication uses existing level transmitters. The new reactor pressure indication is located on the 901-5 and 3, panels and fed by new pressure transmitters'on the 5'and 6 racks.

Evaluation The safety of the system is increased because indication loops are increased in number providing a more conservative monitoring system.

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L Modification M-4-1-86-12B-

- Description-N The. purpose of this modification is'to satisfy commitments'for Regulatory Guide 1.97.. To meet these commitiuents drywell1 pressure instrumentation bas calf.brated to measure pressure from -10"'Hg to 70 psig. The top' scale on' recorder 8740-12 and scales on indicators 2540-009A&B were changed to reflect the new range. Recalibration of PT 2541-12A and B, PT 1625, and. pressure

- switches 2540-16A&B,-and 17A&B were also calibrated to reflect'the new range.

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'N Evaluation

- The safety aspects of the system will increase because the instrumentation

' is now capable of indicating vacuum.

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- Modification M-4-1 12C r

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The' purpose of.this modification is to satisfy commitments for Regulatory r

l Guide 1.97 and Human Factors. A new torus water. temperature recorder and j

signal' isolators were added to.the 901-4 panel.

In order to make room for the new recorder, relocation of existing RCIC instrumentation (square root

, converter,.1340-10, and power supply, 1340-12) from.the 901-4 panel'to the i

s 901-19 panel was necessary, i

Evaluation-t e

The safety of the system will not change because of this modificatiori.

The torus water temperature is redundant instrumentation of instruments' located on the back control room panels.

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I Modification M4-1-86-12D Description The purpose of the modification was to assure that Quad Cities Station Primary Containment Radiation Monitoring equipment is consistant with actual plant parameters. This modification was performed to fulfill the commitments made to the NRC in order to be in compliance with Regulatory. Guide 1.97, dated July 31, 1985.

This included verifying the calibration of existing radiation level monitors (1-2418A,B,C,D and 1-2419A,B,C,D) located on the 901-55 and 56 i

I panel and the torus radiation monitoring on Recorder 1-2420A and B (Pen 2) reflected a range of 1 to 106 R/hr.i This modification also changed the scale fordrgwellradiationrecorders (1-2420A&B) (Pen 1) from 1 x 106 R/hr to 1 x 10 R/hr and installed proper name plates.

Evaluation This modification did not change the existing system operation. The modi-fication simply revised scales and verified calibration of Drywell and Torus Radiation monitors and recorders to be in compliance with the installed equipment capabilities and to provide Human Factors enhancements to the present displays.

Therefore, the safety aspects of the system does not change.

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Modification M-4-1-86-12E Description The purpose of the modification was to satisfy commitments for Regulatory Guide 1.97.

To meet these commitments existing thermocouple and thermocouple extension wires were replaced with environmentally qualified thermocouple and thermocouple extension wires. The replacement was performed on Drywell Atmosphere Temperature thermocouple (1-5741-42A&B, 1-5741-73A&D, 1-5741-44B&C, 1-261-38A&B) and the RHR heat exchanger outlet temperature and cooling water to ESF components temperature thermocouple (1-1047A&B, 1-1052A&B).

Evaluation The safety aspects of the system will increase because non-EQ equipment is being replaced with EQ equipment.

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4, Modification M-4-1-86-12F, Description

.. Recalibrates existing reactor water level. instrumentation, (the 1-263-106A&B,.

'l 901-3 panel, 1-640-27 recorder, 901-5 panel,' and reactor water'1evel transmitters,.

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1-263-73A&B) from -243" - +57" to -340" - +60".

The change is to. provide environmentally qualified' instrumentation for reactor water level that will.

tj monitor a minimum'of -290" for commitments to Regulatory Guide 1.97.

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Evaluation I

The increased span will allow for the operators to monitor reactor-water level: during emergency conditions 'and reduce the chances of off scale ' indications.

The 2/3 core covered interlocks, provided by the wide range level instrumentation,

. remains unchanged.

Therefore,-the safety of the system is increased because of l

the increased span.

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Procedure Change QAP 300-2 l

.This. procedure' change' includes the actions required when a' surveillance-J cannotLbe completed due to plant status and a statement'has been added to i

prohibit placing ECCS pumps in pull-to-lock until' water level'is above the-top ofcfactive-fuel.

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1.. : The probability o'f uan occurrence or the conseque'nce of an accident, 3

or malfunction of equipment important to safety as previously evaluated in.the Final Safety Analysis l Report is not increased because the j

-changes to this procedure 1do not affect'any previously evaluated accidents or malfunction.

2.

TheLpossibility for an accident or malfunction of a different type l

than any previously evaluated in'the Final Safety Analysis Report is not created because these changes provide guidance lon what to

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do when a surveillance cannot be performed and further guidance on'

. placing equipment in pull-to-lock.

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3.

The margin of safety, as defined in the basis for any Technical Speci-

-fication is not reduced because the changes made are not defined in any Technical Specification. basis.

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A Procedure Change QAP 1500-2 h

This procedure change clarifies the responsibilities of the E.Q. Coordinator.

1.

The probability of an occurrence'or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report is not increased by defining the E.Q. Coordinator's responsibilities in regards to work requests.

2.

The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report is not created because the procedure is simply defining the responsi-bilities of the E.Q. Coordinator.

3.

The margin of safety, as defined in the basis for any Technical Specification is not reduced because the function of the E.Q. Coordinator is not delineated in any Technical Specification basis.

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Procedure Change Q0A 010.

This procedure has been changed..go require taking auxiliary equipment L room keys when-evacuating the control' room.. Instead of opening a breaker, afuses are pulled.to initiate an isolation.. The! pumps started for the turbine.

are identified. The requirement to open the. condenser vacuum-breaker is' removed.

A~means'for verifying the Group I is provided.

1.

The probability of an occurrence or the consequence of an accident,:

o or malfunction of~ equipment important.to safety'as previously evaluated in the Final Safety Analysis Report is not increased because the changes listed above are enhancements to the procedure and do not affect any previous FSAR evaluations.

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'2.. The. possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report

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is not created because the changes listed above.mainly provide a means to perform the steps of the procedure. -These clarifications and'the other changes do not change or create any conditions that require new evaluations.

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The margin of safety, as defined in the basis for any Technical Specification, is not' reduced because the evacuation of the' control-room is not addressed in any Technical. Specification basis.

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Procedure Change QoS 300-1, 12, 4, 6 This procedure revision changes, deletes, or inserts precautions for EGC operation while performing the procedure.

i 1.

The probability of an occurrence or the consequence of an accident, or malfunction of-equipment important to safety as previously evaluated in the Final Safety Analysis Report is not increased because the affect of EGC operation on the affected procedures has been evaluated.

Where restrictive precautions were added to the procedure to limit EGC operations.

2.

The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report is not created because EGC operation is independent of the Reactor Manual Control system. These changes do not impact or create the possibility for an accident or malfunction, j

3.

The margin of safety, as defined in the basis for any Technical Specification, is not reduced because it will not be altered.

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Procedure Change QOS 1300-1

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lThis RCIC, procedure was changed to account /forflockwiringaof the trip l

'and. throttle handwheel..

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.The probability of an occurrence or_the consequence of an accident,

,orfmalfunction of: equipment important to safety as:previouslyfevaluated i

in_the' Final Safety Analysis Report'is'not increased because;the RCIC system operation was unchanged by this' procedure revisioni LFrom thel. control room the operators.will not alter-the operation of the system automatically or_ manually.

2..

The possibility for an accideat or malfunction of a different type:

than anyfpreviously evaluated in the. Final' Safety Analysis Report.

o is not created'because the system. operation is unchanged. Turbine'-

i reset will require removal'of.a'lockwire'to' accomplish the, task.

.This is easily.done.and will not hamper execution of the procedure.

3.

The margin of safety, as defined in the basis for any:-Technical

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Specification, is not reduced _because'the Technical Specificat!L' y

ibasis is-unchanged.

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