ML20090E307

AO 73-11-20 on 731120,11 Control Rods Failed to Insert to Position 00 Following Scram.Caused by Damaged Stop Piston Seals Resulting in Excessive Leakage.Rods Will Be Overhauled

ML20090E307
Person / Time
Site:	Nine Mile Point
Issue date:	11/28/1973
From:	Schneider R NIAGARA MOHAWK POWER CORP.
To:	Skovholt D US ATOMIC ENERGY COMMISSION (AEC)
Shared Package
ML20090E309	List: ML20090E307 ML20090E313
References
8604, AO-73-11-20, AO-731128, NUDOCS 8303140597
Download: ML20090E307 (2)

Text

e ye I l

NIAGARA MOHAWK POWER CORPORATION NIAGARA MOHAWK 300 CRjt BOULCvARD wCST SYRACUSE. N. Y '32 02 November 28, 1973 AOR 73-11-20 _.

e Ib.3 Mr. Donald J. Skovholt '

}fI D Assistant Director for Reactor Operations Division of Reactor Licensing

/

h.(L

(,\

2a United States Atomic Energy Commission ggCS ggar b , dl Washington, D. C. 20545 9A ..jJ{ f Q ~" y'-lW8 gn Re: Provisional Operating License: DPR-17 g Docket No.: 50-220 Q~ # g h a ) g

Dear Mr. Skovholt:

On November 20, 1973 at approximately 0230' hours at Nine Mile Point Nuclear Station, Unit #1, eleven (11) control rod drives failed to insert to notch position 00 following a Reactor Scram. Persuant to Technical Specification 1.13 d.,

Failure of one or more components of an engineered safety feature or Station system that causes or threatens to cause the feature or system to be incapable of performing its intended function, this is an abnormal occurrence and was reported by telephone and photocopy to the Director of Regulatory Operations, Region Office I on November 20, 1973.

Prior to the occurrence a routine startup was in progress with reactor power being increased, using reactor recirculation flow, following a scheduled shutdown for Operator Demonstration criticals. Reactor power had reached the 77%

level, water level was norma 1r reactor pressure was 988 psig, the generator was on line with all systems operating normally for a routine start-up.

A sudden ir. crease in recirculation flow rate occurred at 77% power causing an increase in p'asitive reactivity (the result of voids being " swept" more rapidly from the core i.e. better moderation) resulting in a high flux reactor -

scram and anticipatory turbine generator trip. The control room operator immediately decreased recirculation. flow when it became obvious that recirculation flow was on an upward ramp and neutron instrumentation began steadily increasing,

~

j' however his proper action could not prevent the high flux flow bias reactor scram.

The recirculation flow ramp -increase will be the subject of an unusual; event Ictter before December 20, 1973. All systems operated nor.nally for; this transient except

~

the control rod drive systen. The operator noted that upon resetting of the-

~

scram eleven (11) control rods indicated ~ position 02 instead of 00. The operator immediately inserted the control rods to 00 using the reactor' manual control

, system with normal' control rod drive prgssure.

e ,A-o .

.8604~

8303140597 731120 s$ if#4 PDR ADOCK 05000220 s .

PDR V/t 4 F

u b COPY SENT REGION - Y_

/

c - .

. ~

1<'..

<- a

. ] ._ , .

p Mr. Donald - J. Sk:J 1t: .. M (Novcaber'28, 1973 --

- U.'3. Atbnic Energy _ Commission The apparent cause. of this occurrence has been ' identified as damaged stop piston seals causing' excessive leakage past these seals. . This was.

determined from stall leakage flow tests and scram testing of each of the' affected control rods prior to unirt power generation on the.. succeeding start-up.

Excessive flow past the stop piston-seals instead of thru the buffer holes (the number of those holes decreases as the control rod is inserted) will result

'in slowing .down of the control rod over the last 5% of its travel.

Throughout tne preceding transient' and during the ' period of time that the eleven (11) control rods were at position 02 no hazard was presented to the general public. The reactor was highly suberitical' and could' not have been made critical using th'ese eleven control rods at position 02.

To correct this situation, the control' rods affseted will be overhauled during the Spring 1974 refueling outage as well as any others which indicate the high stall . leakage flows.

In the past replacement of control rod drives during refueling outages has been determined in part by high stall leakage flows. The deterioration of the stop piston seals of the General Electric' control rod drive mechanism appears to increase with increasing' drive pressure required to operate the control rod. Orders have been issued to operations to maintain the drive -

pressure within normal limits during rod withdrawals. However those control rods with existing deteriorated seals will have tolbe moved'using higher than normal pressure.

Very truly yours,

' s ij ,

j

, '.- ~ '

udolph y Schnei er Vice President - Electric Operations - ,

RRS:cm cc: Mr. James O'Reilly -i-RO:I-b +

s _

'.~ -

a

h. .

s k J' 4 i

V 1

,e ., ,

X e . -

i' ,..

s _ < . _ ,

y r- , / A ., <

~ ~

- . . -_- 1 ~, ., 2 0 U . - -

- - - 1 . -. .