05000369/LER-1982-058, Forwards LER 82-058/03L-0.Detailed Event Analysis Encl

Forwards LER 82-058/03L-0.Detailed Event Analysis Encl

ML20062M187
Person / Time
Site:	McGuire
Issue date:	08/06/1982
From:	Tucker H DUKE POWER CO.
To:	James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
Shared Package
ML20062M190	List: 05000369/LER-1982-058, Forwards LER 82-058/03L-0.Detailed Event Analysis Encl
References
NUDOCS 8208200022
Download: ML20062M187 (3)

LER-2082-058, Forwards LER 82-058/03L-0.Detailed Event Analysis Encl

Event date:
Report date:
3692082058R00 - NRC Website

text

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o DUKE Powru COMPANY STEAM PHODUCTION DEPT.

GENERAL OFFICES TELEPHONE: AREA 704

p. o. mox asses 422 SOUTH CHURCH STREET 373 4oss CIIAHLO1TE, N. C. 28242 August 6, 1982 Mr. Ja.ss 1. 8 O'Reilly, Regional Administrator U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, Suite 3100 Atlanta, Georgia 30303 Re: McGuire Nuclear Station Unit 1 Docket No. 50-369

Dear ?t . O'Reilly:

Please find attached Reportable Occurrence Report R0-369/82-58. This report concerns T.S.3.1.1.4, "The Reactor Coolant System lowest operating loop tem-perature (Tavg) shall be greater than or equal to 551oF." This incident was considered to be of no significance with respect to the health and safety of the public.

V ry truly yours, f Nw I. B. Tucker, Vice President Nuc1 car Production PBN/jfw Attachment l

l cc: Director Records Center Office of Management and Program Analysis Institute of Nuclear Power Operations U. S. Nuclear Regulatory Commission 1820 Water Place Washington, D. C. 20555 Atlanta, Georgia 30339 Mr. P. R. Bemis Sr. Resident Inspector McGuire Nuclear Station l

kk}CIALOOPY 8208200022 820806 PDR ADOCK 05000369 /

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DUKE POWER COMPANY MCGUIRE NUCLEAR STATION REPORTABLE OCCURRENCE REPORT NO. 82-58 REPORT DATE: August 6, 1982 FACILITY: McGuire Unit 1, Cornelius, NC IDENTIFICATION: The Unit 1 Reactor Coolant (NC) System Average Temperature (T"#")

Dropped Below 5510F With The Reactor Critical DESCRIPTION: During a unit shutdown on June 24, 1982, T ave dropped below 5510F for seven minutes with the reactor power between 10-5 and 10-7 amperes (less than 1% power). The plant was being shutdown following the successful completion of the Loss of Offsite Power Test and in preparation for a long outage. Some unusual conditions existed because of the test setup and because of preparations to place the steam generators (S/G's) in a laid up condition. Steam demand was too high to allow a reactor shutdown without an excessive cooldown rate. Heat loads included leaking main steam (SM) to moisture separator reheater (MSR) control valves, leaking steam drains, high blowdown and leaking feedwater regulating valves. These heat loads coupled with the fact that the operating auxiliary electric boiler was having conductivity problems (which lowered output) meant that rcactor power of about 1% was necessary to maintain Tave at 5570F. Since reactor pcwer was balanned by steam dezand without the aid of steam dump, the automatic steam bypass to condenser valves were closed. The control operator, not realizing that no autonatic decrease in nteam loads would compensate for reactor shutdown, directed a nuclear equipment operator GiEJ) in training to step-in the control rods to obtain a power level of 10-8 amperes (which is below the point of adding heat).

As reactor power decreased, Tave and pressurizer level began to drop rapidly. The control operators began to decrease letdown in an attempt to preserve pressurizer level. Once the operators in charge of the control rods realized that a rapid cooldown was in progress, they withdrew the rods to stabilize the plant. " Balance of plant" operators began isolating unnecessary steam drains, MSR's and S/G blow-down. During this time, they also swapped feedwater alignment from the main S/G nottles to the auxiliary (upper) S/G nozzles. During the transient, Pressurizer level dropped to 17%, letdown was automatically isolated, and the pressurizer heaters locked out. Tave dropped to a minimum of 5460F and then rose to 5600F where it stablized. Once the excess steam loads were isolated or swapped to auxiliary steam (AS) and the problems with the electric boiler were corrected, the shutdown was completed without further incident.

The operators attempted to shutdown the reactor before steam loads had been swapped to AS and before feedwater had been alighed to the upper nozzles. Specifically, the con-trol operator attempted to shutdown the reactor without a corresponding drop in steam loads. Thus this incident was the result of Personnel Error.

EVALUATION: The main contributing factor to the excessive steam loads apparently is the feedwater alignment.to the S/G's. Aligning the feedwater to the upper nozzles (auxiliary nozzles) increases the back pressure on the feedwater regule. ting valses and decreases the leakage (auxiliary nozzles are located at a higher elevation than the main nozzles). Realignment normally takes place when the turbine load drops below 244 MRe, but this point was passed while the plan was running back during the test.

b Report No. 82-58 Page 2 Once the plant was stabilized at the lower power level, onerators failed to correct the alignment before shutting down the reactor. The day shift operators began shutting down the plant immediately following the turbine overspeed test and subsequent trip.

They turned over control of the plant to the evening shift during the shutdown. Day shift operators felt that the feedwater alignment was adequately covered during the turnover. The evening shift control operator may have been aware of the fecdwater alignment but did not have sufficient time to change it before the incident began.

The alignment was corrected during the recovery effort.

A number of factors could account for excessive steam loads such as those experienced in this incident. A stuck open MSR regulating valve or leaking steau drains alona er in combination would have had the same result. The control operator, shutting down the reactor from low power levels, must analyze the steam load situation before removing reactor heat. He should insure that SM-bypass-to-the-condenser valves are open and the controls in automatic before driving in the control rods.

In this incident the control operator and NEO had full control of the shutdown because the turbine had already been shutdown, and the steam load variations were well within the capability of control rod movement. They also had total control of the speed of t!.a shutdown. Had they driven the control rods in more slowly, they could have watched the response of T ave and reacted sooner to the drcp in temperature.

SAFETY ANALYSIS: Operator reaction to the low NC temperature and pressurizer level were prompt end correct. Automatic isolation of NC letdown also helped limit the drop in pressurizer level. The cooldown and shrinkage of NC water volume were never outside the capability of the charging system so the reactor was not endangered.

The health and safety of the public were act affected by this incident.

CORRECTIVE ACTION: A note will be added to the controlling Procedure For Unit Shutdown cautioning the operators to insure that sufficient excess steam loads exist, as indicated by open SM to condenser bypass valves, to allow a reduction in reactor power below 1%.

This incident will be covered with all operating personnel through crew meetings.

They will be instructed to stop major plant evolution during turnovers, whenever time is available, to insure that the. in-coming shift is fully prepared to operate the plant. They will be specifically reminded to predict the effects of their actions on plant systems to insure that the end results will be satisfactory.