ML19256F148

Responds to NRC 790926 Ltr Re Violations Noted in IE Insp Repts 50-266/79-13 & 50-301/79-15.Corrective Actions: Will Request Tech Spec Change for Section 15.3.10 to Remove Ambiguity

ML19256F148
Person / Time
Site:	Point Beach
Issue date:	10/16/1979
From:	Fay C WISCONSIN ELECTRIC POWER CO.
To:	James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
Shared Package
ML19256F145	List: IR 05000266/1979013 ML19256F148
References
TAC-12663, TAC-12664, NUDOCS 7911210541
Download: ML19256F148 (3)

Text

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WISCONSIN Electnc mia coursur 231 W. MICHIGAN, P.O. BOX 2046. MILWAUKEE, WI 53201 October 16, 1979 Mr. James G. Keppler, Dimctor Office of Inspection and Enforcement Region III U. S. NUCLEAR REGULATORY COMMISSION 799 Roosevelt Road Glen Ellyn, Illinois 60137 Der Mr. Keppler:

DOCKET N05. 50-266 AND 50-301 REPLY TO NOTICE OF NONCOMPLIANCE POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 Your letter of Septenber 26, 1979, which forwarded IE Inspection Report No. 50-266/79-13; 50-301/79-15, included a notice of violation. This notice states that, contrary to the Point Beach Nuclear Plant Technical Specifications, Item 15.3.10C, on August 10,1979 from 0310 hours0.00359 days <br />0.0861 hours <br />5.125661e-4 weeks <br />1.17955e-4 months <br /> to 0452 hours0.00523 days <br />0.126 hours <br />7.473545e-4 weeks <br />1.71986e-4 months <br />, the reactor was operated with all rods inoperable. Pursuant to Section 2.201 of the Commission's Regulations, we are providing this response to that notice of violation.

We believe the declaration that the reactor was operated with all rods inoperable is misleading. At 0310, a " rod urgent failure" alarm was annunciated.

There was, at that time, no signal requiring rod motion and no load changes were planned. The reactor operator attempted to move the control rods and no bank responded. He attempted to reset the alam with the dedicated reset switch in the control room, but was not successful. At that point, the Duty Shiff Super-visor notifed the Duty and Call Superintendent and the Instrument and Control (I&C) Engineer. An I&C Supervisor and a Technician were called to the site by the I&C Engineer. The I&C personnel investigated the cause for the " rod urgent failure" alam at the rod control circuitry cabinet and had indications that the alarm had been annunciated by the stationary gripper coil circuitry. In preparation for further trouble shooting of these circuits, the local reset switch was pushed and the alarm cleared. Subsequent investigation by the Plant Staff revealed no apparent cause or mason for this spurious alam. There has been, to date, no recurrence of this situation.

In consideration of the corrective action and the circumstance involved, it is obvious that the inability to step the control rods was the result of an electrical control circuit spurious alam. Had power to the control rod holding coils been interrupted by a reactor protection system signal, the control rods would have shut down the reactor as designed. The primary function of the 1Hi !69 pf W 7911210 91./.

Mr. James G. Keppler, Dimctor October 16, 1979 control rods is to teminate the critical state of the reactor when a reactor protection signal is sensed. Therefore, it is misleading to state that the reactor was operated with all control rods inoperable, since this primary control rod function would have been satisfied had it been called upon.

When the Technical Specifications were written, it was recognized that if a control rod could be stepped in using the rod control system, the control rod would, under any conceivable condition, drop into the core upon demand.

Because it is unreasonable to verify control rod operability by periodically dropping a rod into the core, the Technical Specifications (Item 15.3.10.C.1.b) were written to pemit control rod stepping as an alternative method of verifying operability. This is supported by the basis for this Specification which states, "From operating experience to date, a control rod which steps 'in' properly will drop when a trip signal occurs . . .". Unfortunately, the Point Beach Technical Specifications do not differentiate between the inability to move rods due to mechanical interference from that. caused by control rod electrical problems.

Hence, by the inspector's verbatim interpretation of the specification, the control rods had to be assumed to be inoperable.

This type of electrical failure is not uncommon. At other nuclear facilities, the Nuclear Regulatory Commission has approved Technical Specifica-tions which pemit continued power operation for limited periods of time during the investigation and trouble shooting of multiple " inoperable" control rods.

It had been our interpretation of the Point Beach Technical Specifications that a similar period of power operation during trouble shooting of control rod control system failures was pemitted. This interpretation is based on Specification 15.3.10.C.2 which states, "No more than one inoperable control rod shall be pemitted during sustained power operation". We believe that one hour and 42 minutes of power operation while this specific problem was investigated was not

" sustained" operation.

Nevertheless, to clarify the Technical Specification interpretation and to provide a positive response to your notice of noncompliance, we shall request a Technical Specification change for Section 15.3.10 which removes the antiguity from the Specification. An initial draft of the proposed wording we intend to submit with this change request is attached. This position will be reviewed by both the Manager's Supervisory Staff and the Offsite Review Committee and submitted as a fomal license amendment mquest in both Point Beach Nuclear Plant Dockets in about 60 days. Until this license amendment is submitted and approved by the NRC, it is requested that this letter and attachment be accepted as the basis for interpretation of the existing Technical Specifications regarding operability of control rods.

Very truly yours ,

( 'NJ {

C. W. Fay, Di ctor Nuclear Power Department Attachment 15// ?/0

C. Inoperable Rod Cluster Control Assembly (RCCA)

1. A control RCCA shall be considered inoperable if the following occurs:

a. The RCCA does not drop upon removal of stationary gripper coil voltage,

b. The RCCS does not step in properly when the proper voltage sequences are applied to the control rod drive mechanism coils. It shall then be assumed inoperable until it has been tested to verify that the RCCA would drop.

c. If an RCCA does not step in upon demand up to six hours is allowed to determine that the problem with stepping is an electrical problem. If the problem cannot be resolved within six hours, the RCCA shall be assumed inoperable until it has been verified that it will step in or would drop upon demand.

,' d. If more than or,a RCCA does not step in, apparently due to electri-cal problems, the situation shall be rectified or clearly defined that it is an electrical problem and the RCCAs are capable of dropping upon demand or an orderly shutdown shall conmence within six hours.

e. The RCCA is shown by the rod position indicator channel to be misaligned by more than 15 inches. It shall be assumed inoper-able until it has been tested to verify that it does step in properly or that it does drop.

2. No more than one inoperable RCCA shall be permitted during sustained power operation.

3. When it has been determined that a RCCA does not drop on removal of stationary gripper coil voltage, the shutdown margin shall be increased by boration as necessary to compensate for withdrawn worth of the inoperable RCCA.

13./7 ?71