ML19290F027

Submits Responses to Items 1-5 in NRC 800306 Ltr Re Facility 800226 Event.Provides Description of Event,Summary of Previous Power Upset Events & Plant Susceptibility to Event Occurrences.Believes Continued Operation Acceptable

ML19290F027
Person / Time
Site:	Arkansas Nuclear
Issue date:	03/12/1980
From:	Steel C ARKANSAS POWER & LIGHT CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
1-030-05, 1-30-5, NUDOCS 8003170478
Download: ML19290F027 (8)

Text

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ARKANE AS POWER & LIGHT COMPANY POST OFFICE BOX 551 UTTLE ROCK. ARKANSAS 72203 (501)371-4000 March 12,1980 1-030-05 Mr. H. R. Denton, Director Office of Nuclear Reactor Regulation U. S.11uclear Regulatory Commission Washington, D. C. 20555

Subject:

Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. DPR-51 Evaluation of Crystal River -

Unit 3 Event on February 26, T0 (File: 1510)

Gentlemen:

In accordance with your letter of March 6,1980, Arkansas Power & Light Company herein provides written responses under 10 CFR 50.54 (f) to Items 1 through 5 of Enclosure 3 to your letter. The responses pertain to Arkansas Nuclear One - Unit 1 (ANO-1). Written responses for Items 6 and 7 will be provided no later than March 17, 1980.

Item 1 Summarize power upset events on UNI /ICS that have previously occurred at your plant.

Response

Including the startup testing period prior to commercial operation, the following three (3) events were a result of power upsets to the ANO-1 NNI/ICS:

Date Description 12/6/74 While operating at approximately 80% FP, water leakage into the "B" Main Chiller load controller resulted in an electrical fault, causing a loss of NNI-Y power and a loss of condenser vacuum pumps induced by the under-voltage spike. The r ector tripped due to high RCS pressure. The leak was repaired and the 4160 VAC pro-tective relay system was checked prior to resumption of power operation.

MEMBEA M. COLE SOUTH UT!UTIES SYSTEM

1-030-05 Mr. H. R. Denton March 12, 1980 Date Description 7/8/76 While operating at approximately 94% FP, a NNI power supply was inadvertently shorted during maintenance, causing the breaker to trip and resulting in a loss of indication from some ICS components. The reactor tripped on High RCS Pressure caused by an automatic increase in p ;r. Revised administrative controls were used to prevent the recurrence of this type of event.

11/18/77 While operating at approximately 99% FP, the inadver-tent shorting of a ICS power supply did not cause a reactor trip but caused a runback to 60% FP. Power escalation commenced upon resetting of the tripped breaker.

Item 2 Specifically review the Crystal River event, and address your plant's susceptability to it in general.

Response

Preliminary data indicates that the stimuli for the Crystal River event may have been an inadvertent short circuit between the positive twenty-four volt bus and the power supply common bus in NNI Channel X. The power monitor performed its function properly. During the interval be-tween detection of the low voltage condition and breaker opening, the instrumentation modules in NNI Channel X would have been operating with unbalanced supply voltages. This could have resulted in momentary trip conditions of some signal monitors in NNI-X. These trip conditions would have cleared once power was removed. Data indicates that the pressurizer pilot operated relief valve and the pressurizer spray valve were signaled to open by. the momentary trip states. The problem created by the momentary open commands could have been aggravated by latching features in the pres-surizer pilot operated relief valve controller and the " fail as is" char-acteristics of the pressurizer spray valve and the fact that the NNI system does not transmit close commands when de-energized.

Tha AN0 Non-Nuclear Instrumentation system is the same vintage and w3s supplied by the same vendor as the Crystal River NNI. It is therefore susceptible to the fault stimulus that appears to have precipitated the Crystal River event. The response of the NNI system to the fault would be the same. The pressurizer pilot operated relief valve and pressurizer spray valve are similar to the extent that they could be expected to behave as did the Crystal River valves. The emergency feedwater design at ANO is different than the Crystal River design to the extent that the failure of NNI Channel X would not have precluded initiation of one train of emergency

1-030-05 Mr. H. R. Denton March 12, 1980 feedwater by flNI Channel Y due to its normal initiating parameters which are:

1. Loss of all reactor coolant pumps,

2. Loss of both main feedwater pumps, or

3. Low level in steam generator B.

The foregoing answer is based on the following understanding of the Crystal River event and lillI system characteristics. The Crystal River f4NI is divid-ed into two channels (X and Y). TillI-X contains power supplies which power legative and positive twenty-four volt D.C. buses. This arrangement is duplicated in lifil-!. In both f1NI-X and Y, the bus voltages are monitored by a power monitor module. When the power monitor module detects low voltage on either bus, it trips breakers which remove primary input power from the power supplies in that channel. Time delay from detection of a low voltage condition to removal of primary power is approximately one-half second.

Item 3 Set forth the information presented by your representative (s) in the meeting on March 4, 1980.

Response

Arkansas Power & Light Company was informed of the Crystal River event on February 26, 1980. Since that time, we have ' evaluated the available.

information describing the sequenc.e of events and causes of the incident.

On the af ternoon of March 3,1980, members of the Florida Power Corpora-tion staff presented a sequence of events (as of 2300 on 3/1/80) and proposed causes of the event to representatives of the B&W Owners Group.

Based on the brief time available to evaluate accurate information (less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />), we felt we were unable to provide an adequate or useful comparison between the fillI/ICS power supplies of At;0-1 and Crystal River.

Based upon the above, only general comments were made. All pertinent information requested in the meeting is now addressed or will be provided by March 17, 1980.

Item 4 Address information available to the operator following various flNI/ICS power upset events, including a discussion of:

- how the operator determines what information is reliable, and

- what information is needed to bring the plant to cold shutdown.

1-030-05 Mr. H. R. Denton March 12, 1980

Response

isent Afi0-1 procedures for dealing with loss of fiNI/ICS power are included

o. OP. 1203.12, Annunciator Corrective Action; OP. 1202.02, Blackout; and our 14SS Data Point Log.

Our Annunciator Corrective Action procedure simply addresses restoration of lost power. Our Blackout procedure directs the operators to select inverter-powered indications and refers to the use of certain remote, direct-reading indications. The NSS Data Point Log lists all selectable indicators, provides guidance as to which transmitter is " preferred" by virtue of its power source, and indicates the source of power for each indication. A power supply failure can be identified via local descriptive drop panel alarms, repeated through a common alarm on the Main Control Room Annunicator, and thus the operator has an indication of reliable instrumentation. Simulator training, operator li-cense training and requalilication training include fiNI/ICS failure training.

This training is relied upon for the operator's ability to diagnose power and component failures.

We are currently drafting new procedures based on our findings from a very close review of our UNI /ICS (and certain other) preferred indicatcrs cir-cuitry which will detail specific information as to which indicators are accurate (and which are not) under several postulated power losses to the NNI/ICS systems. Alternate indications, such as ccmputer points, RPS and ES panel instruments, and remote local indications will also be referenced.

This interim procedure, based on the presently installed system, will be in effect by March 31, 1980.

Attachment 1 is a listing of the i;NI information needed to bring the plant to cold shutdown which is available to the operator following four loss of power events including: loss of 24VDC-X power; loss of 118VAC-X power; loss of 24VDC-Y power; loss of 118VAC-Y' power.

Item 5 Address the feasibility of performing a test to verify remaining information following various Nfil/ICS power upsets.

Response

A power upset test could be performed to verify the operability of all NNI/

ICS parameter indications in the control room during loss of various power buses internal and external to the UNI. This test would be performed during cold shutdown conditions with appropriate precautions to ensure the operation and monitoring of the decay heat system.

The scope of the proposed test covers disabling the NNI internal and external power buses in each NNI channel and testing the effect on the control room indications on loss of each bus. The validity of each indication would be determined by testing each individual instrument string with simulated inputs to verify operation. The results of this test would verify the validity of control room indications during power upsets, verify the coincidence of installed equipment and its documentation, and provide input to and veri-fication of operator procedural controls used during loss of NNI power events.

1-030-05 Mr. H. R. Denton March 12, 1980 Such a test would involve several days of operation with reduced instru-mentation; with a non-redundant instrument arrangement such that an error in the test performance could result in loss of all instrumentation. The consequences of such a test must be weighed against the verification ob-tained by the test before a final decision is made to proceed.

Based upon of review of the causes of the Crystal River event and the ney-ligible effect of the event on the health and safety of the public, we contend the continued operation of ANO-1 to be acceptable. This conclusion was drawn from an apparent lack of a common mode failure to cause a power supply failure. Specifically, the ANO-1 saturation meters are currently installed and operable, adtr.iaistrative coni.rols exist to limit work activ-ities inside the NNI/ICS cabinets to cold shutdown conditions with experi-enced personnel concurrent with the knowledge of the Shif t Supervisor, and the absence of module failures at operating plants.

Very truly yours,

$ m L L3 Dcz9\

Charles L. Steel CS:DGM:skm Attachment

STATE OF ARVANSAS )

) SS COUNTY OF PULASKI )

C harles L. Steel, being duly sworn, states that he is Vice President / Assistant to President for Arkansas Power and Light Company; that he is authorized on the part of said Company to sign and file with the Nuclear Regulatory Commission this Supplementary Information; that he has reviewed or caused to have reviewed all of the statments contained in such information, and that all such statements made and matters set forth therein are true and correct to the best of his knowl edg e, information and belief.

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.h Charles L. Steel SUBSCRIBED AND SWORN TO before me, a Notary Public in and for the County and State above named, this /.'"i day of fYpu/[, , 1980.

/ a:en . tilnv M otary N Public My Commission Expires:

8- /6- [2

ATTACHMENT 1 Power Loss Power Loss Power Loss Power Loss Parameter 24VDC-X 24VDC-Y 118VAC-X 118VAC-Y Pressurizer Level C C C C Pressurizer Level Temp Comp -

SRX -

SRX Pressurizer Water Temp SIY C,SIXl - C,SIXl RC Outlet Temp (Th )

Loop A WR C C C C RC Outlet Temp (Th )

Loop B WR C C C C RC Inlet Temp (Tc )

Loop A WR SIY C,$1XI -

C,SIX l RC Inlet Temp (Tc )

Loop B WR SIY C,SIX l - C,SIXl OTSG A Startup Level - C,SIX1 ,3 -

C,SIXl 0TSG B Startup Level C SIX3 C SIX OTSG A Operating Level SRY C,SRX2 - C,SRX2 OTSG B Operating Level C SRX C SRX OTSG A Pressure SIY C,SIX I -

C,SIX 1 OTSG B Pressure C,SIY 1 SIX C SIX DH Removal Flow A -

I -

I DH Removal Flow B I -

I -

viakeup Tank Level -

C,SRX 2 -

C,SRX2 BWST Level SIY SIX -

SIX RCS Wide Range Press. R4 R4 -

R4 DH Cooler Temp Loop A DSCHG I ,C -

I,C -

DH Cooler Temp Loop B DSCHG -

I,C -

I,C Saturation Meters Ch 2 R Ch 1 R Ch 2 R Ch 1 R

ATTACHMEllT 1 (continued)

Symbols:

(-) -

Indicates none available C -

Computer I -

Indicator R -

Recorder Prefix S - Selectable between X or Y Suffix X or Y -

Denotes signal origin or dependency liotes :

1. Either indication to operator and input to other function og; computer input, but not both.

2. Either recorded and input to other functions or computer input, but not both.

3. Either X or Y transmitter can be selected for indication.

4. Transmitter is from Engineered Safeguards.