05000333/LER-1979-063-03, /03L-0:on 790906,during Normal Startup,Control Rod Worth Minimizer Was Intentionally Bypassed.Caused by Revision in Control Rod Withdrawal Sequence

/03L-0:on 790906,during Normal Startup,Control Rod Worth Minimizer Was Intentionally Bypassed.Caused by Revision in Control Rod Withdrawal Sequence

ML19209B277
Person / Time
Site:	FitzPatrick
Issue date:	10/03/1979
From:	W. Verne Childs POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML19209B272	List: 05000333/LER-1979-063, Forwards LER 79-063/03L-0 05000333/LER-1979-063-03, /03L-0:on 790906,during Normal Startup,Control Rod Worth Minimizer Was Intentionally Bypassed.Caused by Revision in Control Rod Withdrawal Sequence
References
LER-79-063-03L, LER-79-63-3L, NUDOCS 7910090432
Download: ML19209B277 (2)

LER-1979-063, /03L-0:on 790906,during Normal Startup,Control Rod Worth Minimizer Was Intentionally Bypassed.Caused by Revision in Control Rod Withdrawal Sequence

Event date:
Report date:
3331979063R03 - NRC Website

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NAME OF PF4EPARER PHONE:

POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 ATTACHMENT TO'LER 79-063/03L-0 Page 1

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During normal startup the Control Rod Worth Minimizer (RWM) was intentionally bypassed.

Prior to the initiation of control rod withdrawal, the RWM was verified operable utilizing Operations Surveillance Test F-ST-20A titled " Rod Worth Minimizer Functional Test" in accordance with the requirements of Technical Specifications Appendix A, Paragraph 4.3.B.3.b.

Subsequent to the initiation of the reactor startup, the cont rol rod wi thdrawal sequence was changed in that the control rod listings and positions for certain control rod groups was revised and " loaded" in-to the plants process computer and RWM.

As a resul t of these changes, the opera-bility demonstration provided by the successful completion of F-ST-20A was no longer valid.

As a result of this action, proper control rod wi thdrawal was verified by a second licensed reactor operator or a member of the Reactor Analyst staff as allowed by Technical Specifications Appendix A, Paragraph 3.3.B.3.b until reactor power was above 20% of rated. Thus the event did not represent any significant hazard to the public health and safety.

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