ML20126C652

Summary of 800304 Meeting W/Util,B&W Licensees Re Facility 800226 Nonnuclear Instrumentation Power Supply Failure Resulting in 43,000 Gallons of Coolant Escaping Primary Coolant Sys Into Reactor Bldg.Supporting Documentation Encl

ML20126C652
Person / Time
Site:	Crystal River
Issue date:	03/13/1980
From:	Reid R Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
TAC-12961, NUDOCS 8004080002
Download: ML20126C652 (57)

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NUCLEAR REGULATORY COMMISSION 1

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March 13,1930 Docket No 50-302 LICENSEE:

Florida Power Corporation (FPC)

FACILITY:

Crystal River Unit 3

SUMMARY

OF MARCH 4, 1980 MEETING REGARDING CRYSTAL RIVER UNIT 3 INCIDENT OF FEBRUARY 26, 1980

@_rpose The NRC staff requested FPC to make a presentation, to us and the other Bs operating plant licensees, on the February 26, 1980 Crystal River 3 (C.-3) event. This event was initiated by a non-nuclear instrumentation power supply failure and resulted in 43,000 gallons af coolant escaping the pri-mary coolant s."c' am into the reactor buildir g.

The event did not result in any inc' radioactivity outside tne reactor b"P 'ing.

B&W (Lynchburg.

-~ invited to attend.

Summary Mr. George Moore, Asst. Y.P. of FPC, in his introductory remarks presented these plans for restart of CR-3.

1.

Remove water from Reactor Building.

2.

Clean up Reactor Building.

3.

Complete NUREG-0578, Lessons Learned Items.

4.

Restart Unit.

5.

Shutdown in September 1980 for Reload.

Bill Kemper of FPC presented the agenda item of Event Sequence (9 page handout enclosed).

He stated that during 20 minutes of event until the PORV block valve was closed, the plant was operated using the small break LOCA procedure.

The Shift Technical Adviser (NUREG-0578 requircment) was in the control room in one to two minutes.

Other STAS reached the site very quickly.

No Safeguards nor Class 1 systems were lost during the event.

Forty-three thousand gallons of water were pumped through the PORV and one code safety valve (RCV-8) during the 25 minutes of HPI operation before HPI was throttled.

The HPI auto initiated three minutes and 20 seconds after initiation of event, due to low RCS pressure (1500 psig).

Containment temperature reached I

128 F (110 F is normal); pressure reached about 4 psig.

Highest radiation level reached was 60 R/hr due to release of non-condensable gases in pressurizer and coolant to containment atmosphere.

8004080002

." e l The event was initiated at 14:23:21 due to a short on a positive 24v bus in the Non-Nuclear Instrumentation "X" supply.

The short was cleared at 14:44:12, 21 minutes later, and NNI power was restored.

At this point in time it was verified that natural circulation was in progress and the plant subcooling margin was 131 F.

The lowest subcooling observed during the event was 8 F.

Code Safety Valve RCV-8 opened 100 psi low.

Ken Vogel of FPC presented Agenda Item entitled " Description of Equipment Defi-ciencies".

FPC concludes technicians working on the NNI Y bus did not cause the event because NNI Y is buffered from NNI X.

FPC, in a test, has found a short and duplicated failure.

FPC observed loss of positive bus on annunciator.

FPC is actively looking for physical evidence of the fault as last piece of evidence to establish cause beyond doubt.

Paul McKee of FPC discussed the Aaenda Item "Recent History (at CR-3) Examination".

FPC didn't find any precursors to this failure, immediately preceding the failure.

There were plant oscillations in the ICS (lasting 2 to 6 seconds) during three events from November 1979 thru February 1980.

However, the ICS did not parti-cipate in the February 26,1980 event.

FPC has isolated the component which caused these oscillations.

No operator action was needed durino these oscill-ations.

FPC stated that there was no connection between the module failure and the February 26,1980 event.

Ken Vogel of FPC discussed " Planned Corrective Action" item of the Agenda.

Prior to startup (about 3 weeks)

- Thorough testing of the NNI system to determine cause of failure

- Modify PORV so that NNI failure closes valve

- Modify pressurizer spray valve so that valve (doen't open) on NNI failure

- Provide positive indication of position of all three relief or safety valves

- Establish procedural control of NNI selector switches

- Train all operators in response to NNI failures

- Move 120v ICS "X" power to vital bus

- Initiate more extensive program for events recorder systen l

- Provide operator with redundant indication of main plant parameters At next refueling (Sep_tember 1980)

- Install indication lights on all panels to know if power is on

- Quick access to fuses is being designed into cabinets

- Modify EFW pump circuit to start pumps on any low steam generator level sianal i

n

-3,

Long Term

- Investigate upgrade of NNI capabilities - total loss of NNI

- Remote shutdown panel is being designed

- Provide backup AC sources to inverters with automatic transfer The other attending B&W licensees, SMUD (Rancho Seco), Duke (0conee),

Arkansas Power and Light Company (ANO-1), TECo (Davis-Besse) and Met Ed (TMI 1/2) presented their reevious similar power failures and dis-cussed their plants in reference to the CR-3 event. The plants have diverse NNI/ICS designs. Mr. Denton suggested that all licensees should determine the impact of an NNI failure by running tests.

FPC has per-formed such tests at CR-3.

However, licensees at the meeting did not commit to such tests.

All licensees agreed to update their response to IE Bulletin 79-27 based on their study of the CR-3 event and its applicability to their plants.

l Mr. Eisenhut concluded the meeting by asking the licensees to submit by C.0.B., March 12, 1980, their proposals to upgrade their NNI systems based on CR-3.

He asked them to ad cess the feasibility of running NNI tests 1

and to provide their schedules to complete IE Bulletin 79-27 upgrades.

He told all the attending licensees that NRC would send out a request for this information.

~

Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors

Enclosures:

1.

FPC Sequence of Events 2.

Description of Equipment Deficiencies & Block Diagrams 4

3.

Summary of Operable Instruments following NNI Power Failure 4.

Graphical Display of the Transient & Recorder Strips 5.

Attendance List,

cc w/ enclosures:

See last page

Enc 16sure 1 Rev. 5 Page 1 SEQUENCE (AS OF 2300 3/1/80) 26 February Transiene CR-3 EVENT SYNOPSIS At 14:23 on February 26, 1980 Crystal River -3 Nuclear Station experienced a reactor trip from approximately 100% full power. A synopsis of key events and parameters was obtained from the plant computer's post-trip review and plant alarm summary, the sequence of events monitor, control room strip charts, and the Shift Supervisor's log.

i The reactorewas operating at approximately 100 % full power with Integrated Control System (ICS) in automatic. No tests were in progress and minor main-tenance was being performed in the Non-Nuclear Instrumentation (NNI) cabinet "Y".

Time Event Cause / Comments 14:23:00 The following is a summary of plant conditions prior to the trip Flux 98.6%

RC Pressure 2157 psig PZR level 202 inches HU tank level 71 inches T

'"A" 599'F.

i "B" 600 *F.

C "A" 557'F.

T B" 556*F.

Rb Flow "A" 73 I 106 lbs/hr RC Flow "B" 73 I 106 lbs/hr Letdown Flow 48 gym OTSG "A" Iv1 (OP) 67%

OTSG "B" Iv1 (OP) 65%

OTSG "A" FRLV 242 inches OTSG "B" FRLV 254 inches OTSG "A" Pressure 911 psig OTSG "B" pressure 909 psig Main Steam Pressure 894 psig Main Steam Temp. 589 F.

Condenser Vacuum 1.76 Generated MR 834 DFT level 12.7 ft.

6 Feed. Flow "A" 5 I 10 lbs/hr Feed Flow "B" 5 I 106 lbs/hr Feed Pressure "A" 970 psig Feed Pressure "B" 968 psig 14:23:21

+24 Volt Bus Failure CNNI Cause still unknown. Apparently, power loss: "I" supply) the positive 24 VDC bus shorted dragging the bus voltage down to a

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Rev. 5 Page 2 Time Event Cause/ Comments low voltage trip condition. There

-I is a built-in k to second delay at which time all power supplies will trip. There was no trip indication on negative (-) voltage.

This event was missed by the annunciator. Following the NNI power failure, much of the control room indication was lost. Of the instrum-

-entation that remained operable transient. conditions made their indic-cation questionable to the operators.

16:23:21 PORV and Spray Open When the positive 24 VDC supply was i

lost due to the sequence discussed alove i.

the signal monitors in NNI changed st ate causing PORV/ Spray valves to open. Ile PORV circuitry is designed to seal ir l'

upon actuation and did so. The result ant loss of the negative 24 VDC halted spray valve motor operator and prevented PORV seal in from clearing on low l

pressure. It is postulated that the FORV l

opened fully and the spray valve stro<ed t:

for approximately second. The 40% osen indication on spray valve did not l

actuate, therefore, the spray valve dLd i

not exceed 40% open.

i i

16:23:21 Raduction in Feedwater As a result of the "I" power supply l

failure many primary plant control l

signals responded erroneously. Tcold failed to 570*F (normal indication war, 557'F) producing several spurious alarus.

Tave failed to 570*F (decreased). The resultant Tave error modified the reactor demand such that control rods were withdrawn'f to increse Tave and reactor power. The power increase was terminated at 103% by the ICS and a " Reactor Demand High Limit" alarm was received. Thot failed to g70* F (low) and RC flow f ailed to 40 X 10. Ibs/hr in each loop (low).

Both these failures created a BTU alarm and limit on feedwater which reduced feedwater flow to both OTSG's to i

essentially zero. Turbine Header Pressure failed to 900 psig (high) which caused l

the turbine valves to open slightly to

Rev. 5 Page 3 Time Event Cause/ Comments regulate header pressure thus increasing generated megawatts. These combined failures resulted in a loss of heat sink to the reactor initiating an excessively high RC pressure condition.

14:23:35 Reactor Trip / Turbine Rx trip caused by high RCS pressure at 2300 psi Trip Turbine was tripped by the reactor.

14:24:02 Ei Pressure Inj.

This was a computer printout and indicates Req. (Flag)

<50' subcooling.* See attached graph of RC Pressure / Temp. vs. Time. This graph is based on Post Trip data and actual incore thermo-couple data. From the reactor trip point (14:23 to 14:33, core exit temperature data was l

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obtained by extrapolation and calculated data.

This is supported by.tvo alarm data points plotted at 18' and 21* of subcooling during this period from the computer. It is important to note that lowest level of subcooling was 88F for a very short period of time.

• NOTE: This computer program was initiated as a result of the TMI incident.

14:24:02 Loss of Both Suspect condensate pump tripped due to high Condensate Pumps DFT level. This is verified by ???? printed by computer, indicatidg the level instrument was over ranged as well as g low flow indication in the gland steam condenser as also' indicated by computer.

14:25:50 PORV Isolated At this time a high RC Drain Tank level alarm was received. This was resultant from the PORV remaining open and was positive indication that the PORV was open. At this time, the

. operator closed the PORY block valve due to RCS pressure decreasing and high RCDT level.

14:26:41 HPI Auto Initiation HPI initiated automatically due to low RCS pressure of 1500 peig. The low pressure condition was resultant from the PORV remaining full open while the plant was tripped. Full HPI was initiated with 3 pumps resulting in approximately 1100 gpm flow to the RCS. At j

this time, all remaining non-essential R.B.

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Rsv. 5 Pags 4 Time Event Cause/ Comments were closed per TMI Lessons-L' arned Guidelines.

e 14:26:54 RC Pumps Shutdown Operator turned RC pumps off as required by the applicable emergency procedure and B & W small break guidelines.

14:27:20 RB Pressure Increasing This is first indication that RCDT rupture disc had Yuptured.

RB pressure increase data was obtained from Post Trip Review and Strip Chart indication.

14:31:32 R3 Pressure High Tids alarm was initiated by 2 psig in RB.

This l

1e attributed to steam release from RCDT. Code safeties had not opened at this time based upcc tail pipe temperatures recorded at 14:32:03 (Computcr).

14:31:49 OSTG "A" Rupture Katrix This occurred due to <600 psig in OTSG "A.

Actuation The low pressure was caused by OTSG "A" boiling dry which was resultant from the BTU limit and failed OTSG level transmitter. This resulted in the closure of all feedwater and steam block valves which service OTSG "A".

14:31:59 Main Feedwater Pump 1A Caused by suction valve shutting due to Tripped matrix actuation in previous step.

14:32+14:41 ES A/B Bypass Manually bypassed and HPI balanced between all i

4 nozzles (Total flow approximately 1100 gpm

-small break operating guidelines).

14:32:35 Started Steam Driven Started by operator to ensure feedwater was Emergency Feedwater Pump available to feed OTSG's.

14:33 Core Exit Temp. Verified The core exit incore thermocouples indicated the highest core outlet temperature value was 560*F. RCS pressure was 2353 psig atthis time, therefore, the subcooling margin at this time was 100*F.

Minimum subcooling margin for the 0

entire transient was 8 F.

It is postulated that some localized boiling occurred in the core at this point as indicated by the self powered neutron detectors.

14:33+14:44 Started hbtor Driven Emer-Same discussion as " Started Steam Driven Emer gency Feedwater Pump gency Feedwater Pump."

l 14:33:30 RC Pressure Righ (2395 psig) At this point, pressurizer is solid and code safety lifts (RCV-8).

This is the highest RCS pressure as recorded on Post Trip Review.

Apparently, RCV-8 lifted early due to seat

~

Rev. 5 Page 5 Time Event Cause/ Comments laakage prior to the transient and RCV-9 did not lift.

14:34:23 RB Dome Ei Rad Level RMG-19 alarmed at this point. Highest level indicated during course of incident was 50 R/hr. High radiation levels in RB caused by release of non-condensable gases in the press-urizer and coolant.

14:35:33 Attempted NNI Repower With-This resulted in spikes observed 6n de-ener-out Success gized strip charts.

14:36:50 Computer Overload caused by overload of buffer. Resulting in no further computer data until buffer catches up with printout.

14:38:15 FWV-34 Closed This valve was closed to prevent overfesd:sg OTSG "B" beyond 100%' indicated Operating Range.

Ii 14:44:12 NNI Power Restored Success-NNI was restored by removing the X-NNI Power

/

fully Supply Monitor Module.

This allowed the breakers to be reclosed. At this time, it was

.g observed that the "A" OTSG was dry, the press-urizer was solid (Indicated off scale high),

RC outlet temperature indicated 556*F (Loop A 4

& B average), and RC average temperature indi-L cated 532*F (Loop A & B).

The highest core exi thermocouple temperature at this time was 531*F RSC pressure was 2400 psig (saturation temp. at I

this pressure is 662*F.).

This data verified natural circulation was in progress and the plant subcooling margin was 131*F. (based on core exit thermocouples).

r 16:44:31 RB Isolation and Cooling Actuation At this time, RB pressure increased to 4 psig i

and initiated RB Isolation.

The operator verified all immediate actions occurred properl:

for HPI, LPI, and RB Isolation and Cooling.

Th-increasing RB pressure was resultant from RCV-8 passing HPI at this time.

i 3

l 16:46:10 Bypassed HPI, LPI and RB These "ES" systems were bypassed at this time Isolation and Cooling to again balance HPI flow and restore cooling water to essential auxiliary equipment (i.e.,

RCP's, letdown coolers, CRDM's etc.).

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Rav. 5 Page 6 Time Event Cause/ Comments 14:51:57 Rupture Matrix Actuation on The actuation was resultant from a deg-OTSG-B radation of OTSG-B pressure. Cold emer-gency feed was being injected into the OTSG at this time. This matrix actuation isolated all feedwater and steam block valves to the B-0TSG and tripped the "B" main FW pump. Both Emergency FW pumps were already in operation at this time.

B-0TSG 1evel at this time was 70% (Operation Range).

14:52 EP1 Throttled and RCS At this time, the maximum core exit thermo-Pressure Reduced to 2300 couple temperature was 515'F, RCS pressure psig was 2390 psig.

Therefore, the subcooling margin was 147*F. Natural circulation was in effect as verified previously. All con-ditions had been satisfied to throttle HPI.

Therefore, flow was throttled down to approx-imately 250 gpm to reduce RCS pressure to 2300 psig in order to attempt to reduce the flow rate through RCV-8 and into the RB.

14:53 Reestablished Letdown At this time, the operator was attempting to establish RCS pressure control via normal RC makeup and letdown.

14:56 Opened MU Pump Recire.

This was done to acsure the MU pumps would valves have minimum flow at all times to prevent possible pump damage.

14:56:43 Bypassed the A-0TSG Rupture Feedvater was slowly admitted Matrix and Reestablished to the A-0TSG which tras dry up to this point.

Faed to the A-0TSG Feedwater was admitted through the Auxiliary FW header via the EFW bypass valves.

The feedrate was very slow in order to minimize thermal shock to the OTSG and resultant depres-surization of the RCS. RCS pressure control was very unstable at this time. It is postulcted that some localized' boiling.occured in core at this point'as indicated by.self neutron detectors.

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Rev. 5 Page 7 Time Event Cause/ Comments 14:57:09

~ Bypassed the B-0TSG This was done to regain FW control of the Rupture Matrix B-0TSG. Level was still high in this OTSG (approximately 65% Operating Range). "herefore, feed was not necessary at this time. She Main Steam Isolation valves were open in preparation for bypass valve operation (when neceseary).

14:57:15 Established RC Pump This was done in preparation for a RCP start Seal Return (when necessary) and to minimize pump seal degradation.

15:00:09 keestablished Level This verified feedwater was being admitted to In A-0TSG the OTSG and made it available for core cooling via natural circulation. Feed to this a

6 generator was continued with the intent of proceeding to 95% on the Operating Range.

15:00:09 77'F Subcooled "A" Loop This value was based upon "A" RCS loop parameters at this time. The "A" loop was being cooled down at this time by the A-0TSG fill and the operator was attempting to equalize loop temperatures.

15:15 23*F Delta-T/ Manned the At this time, loop temperatures were nearing Technical Support Center equalization. This delta-T was calculated from loop A & B T 's and core exit thermo-c couples.

15:17 Declared Class "B" Emergency This was done based on the fact there was a loss of coolant through RCV-8 in the containment and HPI had been initiated. All non-essential CR# 3. personnel were-directed to tvacuate and 's c'ontact off-site agencies'be-

'gan.

Survey ~ team was'sent to Auxiliary Building 15:19 Opened Emergency FW Block At' this point the A-0TSG level was increasing and the decision was made to commence fillinn to B-0TSG the B-0TSG simultaneously. The intent wao to go 95% on both OTSG's without exceeding RCS cooldown limits (10Cf'F/hr) while maintaining RCS pressure control.

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Rev. 5 Page 8 Time Event Cause/Consments 15:26 Lo Level Alarm in Bodium This was recultant from the tank supply valve Rydroxide Tank opening when the 4 psig RB isclation and cool-ing signal actuated. The sodium hydroside was released to both LPI trains.

• Sodium Hydro'xide was admitted to the RCS via.HPI from.the BWST.

(Approtihately 2. ppa injected into the RCS.)

15:50 TerminatedhtPI At this time, all conditions had been satis-fied (per small break operating guidelines )

to terminate HPI. RCS pressure control had been established using normal makeup and letdown. HP1 was terministed and essentially all releases to the RB were discontinued.

16:00 Commenced Pressurizer At this time, RCS pressure and temperature Raatup were well under control. Natural circulation was functioning as designed (approximately 23*F delta-T). RCS temperature was being maintained at approximately 450'.

RCS pressure was approx-imately 2300 psig. The decision was made at this point to commence pressurizer heatup in preparation to re-establish a steam space in the pressurizer.

16:07 Survey Team Report The Raergency Survey Team reported no radiation survey results taken offsite were above back-ground.

16:08 104 Shutdown Steam Drive Emergency FW Pump The motor driven Emergency FW pump was running, therefore, the steam driven pump was not needed The plant remained in this condition for app-roximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, while heating up the press-uriser to saturation temperature for 1800 psig.

16:15 Press Release Media was notified of plant status.

18:05 Established Steam Space Pressurizar At this point, pressurizer temperature was approximately 620*F. Pressurizer level was brought back on scale by increasing letdown.

From this point pressurizer level was reduced to normal operating leval and normal pressure a

was established via pressure heaters.

State and Feder'l Agencies notified.

18:30 Terminated Class B Emergency a

l

Rev 5 Fage 9 Time Rvent Cause/Commments 21:07 Forced Flow Initiated The decision was made to re-establish forced in RCS flow cooling in the ES at this time.

B&W and NRC were consulted. ECP-1B and ID were started. At this point, RCS paramators were stabilized and maintained at RC pressure-2000 psig, RCS temperature-420*F. Pressurizer level-235 inches. The plant was considered in a normal configuration.

4

PLANNED CORRECTIVE ACTION AT CR-3 Immediate Thorough testing of NNI system to determine cause of failure 1

Modify PORV so that NNI failure closes valve Modify pressurizer spray valve so that valve doesn't open on NNI failure Provide positive indication of all three relief or safety valves Establish procedural control of NNI Selector switches Train all operators in response to NNI failures Move 120v ICS "X" power to vital bus.

Initiate more extensive program for events recorder system

- Provide operator with redundant indication of main plant parameters At Next Refueling ' September 1980)

(

Install indication lights on all panels to know if power on panel

- Quick access to fuses is being designed into cabinets Modify EFW pump circuit to start pumps on any low steam generator level signal Long Term Investigate upgrade of NNI capabilities - total loss of NNI Remote shutdown is being designed Provide backup AC sources to inverters with automatic transfer.

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=. _ __

Information reauested by COB, March 12 and March 17. 1980.

1.

Summarize power upset events on NNI/ICS Ihat have previously occurred at your plant.

2.

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

3.

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

4.

A:idress information available to the operator following various 4

Nh4/ICS power upset events, including a discussion of:

- how the operator determines which information is reliable

- what information is needed to bring the plant to cold shutdown 5.

Address the feasibility of performing a test to verify reliable information that remains following.various NNI/ICS power upsets.

6.

Address each CR-3 propo;ed corrective action in terms of applicability to your plant.

7.

Expand your review under IE Bulletin 79-27 to include the implications of the CR-3 event.

Inform us of your schedule for completion of this expanded review as discussed on March 4, 1980.

a In addition to the above, Florida Power Corporation should address:

1.

Sequence of events for the CR-3 trip.

2.

Proposed corrective actions at CR-3.

3.

Discuss the impact, whether it be beneficial or detrimental, of NRC Short Term Lessons Learned and Bulletins and Orders requirements.

)

I 1

ll

Enclosure E DESCRIPTION OF EQUIPMENT DEFICIENCIES The following equipment deficiencies were noted during an in depth review of the Crystal River Unit 3 transient on February 26, 1980.

1.

Loss of the Non Nuclear Instrumentation (NNI) System (X) DC power supplies.

2.

Power operated relief valve (PORV) failed open.

3.

Pressurizer spray valve failed open.

4.

Loss of indication of major plant parameters.

5.

Integrated Control System (ICS) response to NNI (X) failure.

A.

Auto initiation of Emergency Feedwater.

B.

Reactor trip on loss of feedwater.

6.

ICS loss of 120 v AC for 2 seconds.

7.

Events recorder degradation.

8.

Computer alarm buffer overflow.

9.

Systems that performed adequately.

A.

Reactor Protective System.

B.

Engineered Safeguards Activation.

C.

Control Rod Drive System.

D.

Steam Rupture Matrix.

E.

Secondary Plant Equipment.

F.

Radiation Monitoring System.

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6. OTSG "A" Level

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17. "B"-FW Flow

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SUBJECT 3ill carainnun u.

MARCH 4, 1980 MEETING RE: CR-3 INCIDENT OF FEBRUARY 26, 1980 LIST OF ATTENDEES NRC SMUD DPC M. Fairtile John Mattimoe W. A. Coley T. M. Novak Pierre Oubre F. C. Hayworth R. L. Gill V. A. Moore R. A. Knoerr

]

D. G. Eisenhut T. C. McNeekia J. J. Buzy R. W. Reid NUS C_0RP J. A. 01shinski J. T. Beard E. R. Schwardt EEI/INP0 G. C. Lainas J, Martore L. E. Mills C. Nelson T. Telford GPU R. W. Pack C. Long D. Garner R. R. Lentz R. Martin C. W. Smyth ASSOCIATED PRESS F. Jape M. D. Hunt Stan Benjamin J. S. Creswell Jack Guttmann BECHTEL Ell Igne Guy Vissing V. R. Marathe FPC D. C. Dilanni H. L. Ornstein C. Michaelson Bert Simpson D. Tondi TEC0 James Stanfield

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T. Dunning R. C. Wilson F. R. Miller W. F. Pasedag L. C. Stalter FLORIDA PUBLIC_ SERV. COMM.

]

S. Israel Monte Conner Jack V. McLean J. A. Murphy B. L. Siegel B&W S. Newberry J. Mazetis R. E. Ham R. H. Vollmer E. R. Kane GAI G. W. Rivenbark Don Hallman W. P. Gammill Bill Weaver C. H. Bittinq J. P. Joyce John A. Castanes Dale F. Thatcher J. D. Carlton J

J. Zudans K. Schroeder J. Lombardo D. H. Roy GILBERT ASSOC.

P. F. Collins L. M. Fauut J. Heltemes J. C. Deddens C. E. 0 bold Frank Orr R. W. Winks N. H. Wagner CPCo M. J. Salerno J. A. Pastor R. B. O'Donnell i

1

' Enclosure 5 NRC ORNL' Glenn Kelly J. L. Anderson Sammy Diab R. M. Satterfield Vic Benaroya B. A. Wilson VEPC0 C. Graves E. Jordan W. B. Rodill Harold Denton*

PORTLAND GE p

Mary L. Armstrong David G. Mardis Basil A. rker Phillip A. Almond NUCLEONICS WEEK Steve Wynkoop MET ED V. P. Orlandi BAILEY CONTROLS C0 J. L. Marsh 5

• Part time l

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MEETING

SUMMARY

DTSTRIBUT10N LTST-IN ADDlT10N TO THOSE LISTED IN ENCLOSURE 5 Mr. J. J. Mattimoe-SMUD Mr. R. C. Arnold-MET ED Mr. Lowell E. Roe-TECo Mr. William Cavanaugh, Ill-AP&L Mr. James H. Taylor-B&W Mr. J. A. Hancock-FPC Mr. William 0. Parker, Jr.-DPC Docket File VNoonan NRC PDR PCheck L PDR TERA GKnighton NSIC OELD l

ORB #4 Rdg IE (3) l NRR Rdg RIngram ECase RFraley (ACRS-16)

WRussell Program Suppret Branch BGrimes ORB #4 Mtg. Summary File TJCarter ASchwencer DZiemann TIppolito LShao JRMiller JHeltemes-AE00 i

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