ELV-03114, Forwards Responses to Sections I-IV of Additional Suppl to 910708 Suppl to 900911 Petition of M Hobby & Mosbaugh

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Forwards Responses to Sections I-IV of Additional Suppl to 910708 Suppl to 900911 Petition of M Hobby & Mosbaugh
ML20129H771
Person / Time
Site: Vogtle  
Issue date: 10/03/1991
From: Mcdonald R
GEORGIA POWER CO.
To: Murley T
Office of Nuclear Reactor Regulation
Shared Package
ML082401288 List: ... further results
References
FOIA-95-211 2.206, ELV-03114, ELV-3114, NUDOCS 9611040036
Download: ML20129H771 (80)
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Text

{{#Wiki_filter:' 3YP IA u h-h.- Aft'ritt Geor0:230308 Telephone 404 526 3848, u summ Mashng Addressi (/- 40 inverness Center Parkway Post Office Box 1295 Barrningnam. Alabama 35201.- Telephone 205 868 5540 the southem electrc system R. P. Mcdonald ' Executive Vice President

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Nuclear Operations ELV-03114 1124 Docket Nos. 50-424 50-425 U.- S; Nuclear' Regulatory Commission [ Washington, DC 20555 ATTN: Thomas E. Murley, Director Office of Nuclear Reactor Regulation Gentlemen: VOGTLE ELECTRIC GENERATING PLANT REGARDING PETITION OF M. B. HOBBY AND A. L. MOSBAUGH By letter dated August 22, 1991, the NRC requested Georgia Power Company ("GPC" or the " Company") to provide a response to each of the-allegations contained in a July 8, 1991 supplement to the . September-11, 1990 petition of Messrs. Marvin Hobby and Allen Mosbaugh (the " Additional' Supplement"). Enclosed herewith the Company provides the requested responses to Sections I, II, III and IV of the Additional Supplement (Attachments I, II, III and' IV, respectively). Mr. R. P. Mc' Donald states that he is an Executive Vice President of-Georgia Power Company and is authorized to execute this oath on behalf of Georgia Power Company and that,.to the best of his F ' knowledge and belief, the facts set forth'in this letter are true. l, GEORGIA POWER COMPANY [ i / By: ) f (M R. P. Mcdonald yd Sworn.to and subscribed'before me this v day of October, 1991. S.g & _ Notary ( Public

WMSSION DFIRESJANUARY12,1993 l

9611040036 960827 PDR-FOIA KOHN95-211 PDR

a ...-x., I ~ i ,l; Georgia Power h U. S. Nuclear Regulatory Commission ELV-03114 - Page 2 3 - xc: Georaia Power Connany Mr. A. W. Dahlberg Mr. W. G. Hairston, III Mr. C. K. McCoy Mr. W. B. Shipman 'Mr. P. D. Rushton [ Mr. J. T. Beckham Mr. M. Sheibani. NORMS U. S. Nuclear Reaulatory Commission Mr. S. D. Ebneter, Regional Administrator Mr. D. S. Hood, Licensing Project Manager, NRR Mr. B. R. Bonser, Senior Resident Inspector, Vogtle Document Control Desk i f l '..l

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ae sciso 16/62/01 J. ~i o s ' A Wii soia - ~'e e l l* j RHR EVENT CRITIQUE SEQUENCE OF EVENTS }' p. y s t;. s j /[ 1830 APO IN THE CONTAINMENT CALLED TO MAN PHONES FOR TYGON TUBE WATCH FOR DRAINDOWN. APO AWARE OF NEW PERMANANT 4 SITE GLASS AND INTERPRETED THE TYGON TUBE TO BE NEW PERMANANT GAGE GLASS. l 1833 CAVITY PUMPDOWN BEGUN. CAVITY LEVEL AT 210' 4". 3 i 1844 NITE SHIFT USS RELIEVES. CAVITY LEVEL AT 208'. l 1930 CONTAINMENT APO REPORTS THAT THE LEVEL IS NOT ON SCALE l YET (STILL LOOKING AT NEW LVL SYSTEM). ONCOMNING PE0 RELIEVING ON. STATION NOTICES THAT VALVES M1H, D1H, AND DIL SHUT, GD, GV OPEN AND CAPPED. DRAINING WAS STOPPED. 1 PERFORNED FILL AND VENT OF NEN GAGE 1 LG 10401. CONTROL i ROOM REPORTED THAT LEVEL SHOULD BE COMMING ON SCALE SOON. i AFTER FILL AND VENT LEVEL ROSE AND STEADIED OUT AT 206' 5", WHICH WAS CONSISTENT. WITH THE CONTROL ROOM l i INDICATIONS. DRAINING THEN RESUMED. 2200 LEVEL AT 194 ' BY 1 I4 10401 AND TYGON TUBE. DRAINING WAS STOPPED. RCS LO LEVEL ANNUNCIATOR WAS RECEIVED. THE I REACTOR OPERATOR CHECKED 1-L1-957 (TEMPORARY WIDE RANGE IN CR), IT READ 100%. HE TAPPED THE GAGE AND IT READ l APPROXIMATELY 60%. DRAINING WAS STOPPED BY CICSING 1-HV-8715A. THE CONTAINMENT LEVEL WATCH WAS ASKED TO CHECK THE CAVITY LEVEL AND REPORTED IT AT THE, VESSEL FLANGE i WHICH IS APPROXIMATELY 194 '. CONTROL ROOM CONTACTED I&C ABOUT A SUSPECTED 1L987 AND AFTER DISCUSSIONS BELIEVED l 1L957 WAS NOT READING CORRECTLY. THE IAC FORMAN INDICATED l THAT THE INSTRUMENT REFERENCE LEGS MIGHT NEED TO BE FILLED. THE USS AND RO BELIEVED THAT THEY HAD 3 RELIABLE i INDICATIONS (TYGON, NEW LEVEL GAGE GLASS, AND VISUAL) AND l CONTINUED THE DRAINDOWN, SIANLY TO 192 8 l 2235 INDICATIONS OF REDUCED RHR B (TRAIN PROVIDING. COOLING) DISCHARGE PRESSURE AND FIDW. IL 987 INDICATED ABOUT 304 (APPROX 188'3" - TOP OF HOT LEG) AND 1L980 INDICATED ABOUT 60% (187 ' 6" -JUST ABOVE MID LOOP) RO INFORMED USS AND SS. B RHR PUMP WAS PUT ON MINIFLOW AND THE "A" RHR j PUMP USED TO RAISE LEVEL. No ABNORMAL INDICATIONS WERE NOTED ON THE "A" RHR PUMP, POSSIBLY SINCE ITS FLOW HAD I SEEN ABOUT 300 - 400 GPM AS CONTRASTED TO THE B TRAIN I PUMP WHOSE FLOW PRIOR.TO THE EVENT HAD BEEN APPROXIMATELY 7 600 GPM. WHILE ON MINIFLOW THE "B" RHR PUMP AMPS AND DISCHARGE PRESSURE INDICATIONS WERE NORMAL, INDICATING 4 THAT THE PUMP WASN'T AIR BOUND. THE OPERATIONS STAFF i CARRIED OUT THE AOP 18019 ACTIONS AltD RESTORED LEVEL. 2316 AFTTt th RD BEEN RESTORED "B" TRAIN RHR COOLING MODE J. lle i .,,. t e /

E 'd e.cabo so<ou v. WAS REESTABLISHED WITH NO OBSERVABLE OPERATING PROBLENS. i IEC WAS ASKED TO CHECK THE FILL ON 1L950 AND iL957 AND ADCIITIONAL OPERATORS WERE SENT TO CHECK FOR LEVEL SYS , LINEUP PROBLENS. ED, K05INSKY AND OTHERS FOUND NEN LEVEL GAGE ISOLkTION CIASED AND 0015 (TCP ISOLATION TO THE NEN GAGE GLASS) " MIL" THE VALVE WAS IOCATED IN A HICH RADIATION RED TAGGED. ALSO REPORTED A HEPA VENTILATION UNIT CONNECTED TO j AREA. THE REMOVED PRESSURISER SAFETY VALVE. THE ELEPHANT TRUN BETNEEN THE VALVE FLANGE AND THE HEFA WAS COLIAPSED. i 4 e i i i 1 i i i j 4

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PRoCEDUAE NJ. REVISION ' AQ E NO. -~ e PTDB-1 TAB 8.0 4 3 of 5 TAB 8.2 i -848' HID LOOP LEVEL 'oe*! f- - --EpYNir gSjRUMENTATION s.e.. a l =to3' PRtska!!tt l - - aca LM - hl f ;g -301' c -300' / -is.' = i / j y 1BLE igs-TM 1. -197' V jo i -1se' t 194'., 194'.'- -= - 100X 4 = DRA!N DOWN h he[ 3 #3 - tok 1 slow? Ast #1 gga* Lt l - 70x l-M it a's.. t RCP 9tAL9 - - -190' - SOM f - a= l fp, Jf!;-& y.pe'-3.,i =Jex_.. 3 h ista - 20x

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j u -.., uy... ugg"' NOTE: e This Operator Aid is to be used only when in mid-loop configuration and temporary level instrumentation is installed. e LI-957 and LI-950 may read higher than actual RCS level i if opening between RC8 and PRZR is blocked and RCS is pressurised. e Lancer Tap.(Jpper Tap l LI-950 'E LW FRZR LT-0459 Upper Tap LI-937 HL 4* PRZR LT-0459 Upper Tap /FI8/fo i

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Fieenwa u. newesen rsee n-VEGF 23985-1 3 24 of 40 Pressurtaer Level Transattler l c / j h"j, Q -)(gt ILT 469 1 (2 LNsRMATICN g ONLY 8 L, } O Q f Poly Tubing ~ l { Poly Tubing gg,,,j,,, Poly Tubing Tubing Vent t Vent Port Port 1LI.1310 ILE-1320 8 i ( RIVLIS AIVLIS l Isolater Stainless Tubing Stainless Tub,- Isolater Y A .;gr. ~ Wide Ran 1 Narrow Ranged "g JM7 j ILT-11320 l [lLT-11310 i i Y NP P LP HP I pPoly Poly ? j ubing Tuting h i i dHP 1 DP LP LP i Valve Valve Monifold Mantfold 4 HP LP L HP Poly Poly d Tv6 tag futing < ~ - n.. 1 NOTE Poly tubing to be as short and straight as possibia. FIGURE 2

J 8 i VEGP l October 26, 1991 Loss of RHR Incident A/E Technical Questions 8. Determine the ability of the system to provide the minimum flow rate necessary to maintain 140 F core temperature during the event, i assuming the following conditions: (list assumptions) o Upon receipt of indication of abnormal operations, Train B is placed in mini-flow, and Train A is aligned in shut-down cooling mode with suction from the hotleg o Train A is aligned with suction from the RWST o Provide discussion of system hydraulic effects due the present of the reactor internals installed. The flowrate to maintain the RCS water at 140 F at the time of the incident using Train A in the shutdown cooling mode with suction from the hot leg would have to have been 400 gpm (this assumes a CCW flowrate of $500 gpm to the RHR heat exchanger at 80 F). The ERF data i showed indications of Train "A" pump performance degradation at the time the Train "A" heat exchanger discharge isolation valve was closed (bypass valve was already closed). The flowrate of Train "A" at this 4 time was approximately 553 gpm. Therefore, based on the ERF data, it l is inconclusive if the Train "A" pump would have flowed the required 400 gpm in the shutdown cooling mode. i The flowrate to maintain the RCS water at 140 F at the time of the incident using 88 F water flowing from the RWST would have had to have been 400 gpm. At this flowrate the reactor internals would not have a effect on the system hydraulics. This conclusion has been confirmed by Westinghouse. 1 t 'i Naf s M(' l

a... VEGP October 26, 1991 Loss of RNR Incident A/E Technical Questions 9. Determine if Train "A" RHR pump could have maintained 3000 gpm during the event (list assumptions). Based on the ERF data the pump performance data for the Train A pump showed indications of vortexing occurring at the time the operator identified the problem with the Train B pump. At this time Train A had a flowrate of approximately 690 gpm. Therefore, the Train A RHR pump could not have maintained a flowrate of 3000 gpm using Train A in the shutdown cooling mode with suction from the hotleg after the operator identified the problem with the Train B RHR pump. i l i 1 i - N. t 024040 l

l IDES OP IICIW BEAT REICVAL ANIG2TIchL RIDEIZES for 6 ELECI5 TIC GDGIRATDG PIANT IM235 OtB AND DIO A REERMEE TO GENERIC mfMR 88-17 for GREGIA 70iGIR CEMDANY aulapCD PR:0ECT-90GIZE l W4 8002B5W CGEPANY SEWICES, INC. NOCEEhR FIANT SUPPCRMcGZZE

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Propd En95Werf9 MafSper = Vogee SouthemCompany5ervices February 16, 1990 Voetle Electrit C:reratine Plant - Units 1 and 2 No. VG-9011 Final Response to Request for Engineering Assistance File: X780111 Log: SG-8817 Security Code: NC Mr. C. C. Miller Manager of Engineering Vogtle Project - Nuclear Operations Georgia Power Company Post Office Box 1295 Birmingham, Alabama 35201

Dear Mr. Miller:

addresses the specific NRC concerns idThe attached report is the P 1 Number 88-17 and subsequent responses.entified in Generic Letter Also, this report verifies plant The results from the RCS venting analysis were discusse contact at Westinghouse for concurrence prior to the issuance of this report. This document coepletes activities concerning REA VG-9011 any questions, please call David Dotson at extension 6850. If you have Very truly yours ./A ) W. C. Ramsey, Jr. WCRJr/DRD/sm Attachment G. Rockhold, Jr. (w/att.) 1 xc: A. E. Cardona $C. R. Nyer~Jw(/att.)) w/att. M. W. Horton ~ R. E. Patrick

3. Pietrzyk (w(wj att.)

P. D. Rushton /att.) \\ NORNS DocumentFileK6[att.) Project File D _ _ - _ _. - - - ~ ~ ~ ~ ~

-. _ _. _. _ _ _ ~ i. D i l [~" ,2.2.2 (Eth?Z2Y F3.DIf 2NVIrfItEY am E 21s mim'*4m detseminas h$ differerst systems, to the ECB.naurates far gravity news from the NET, D e systamm analyzed include the him1 and voltan control systua rics:smi charging Saw path netich tes menticmed in the NChP, injection systa n%, and the resi&ml heat runwal system cold lag irdecticut

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{ are listed below. -+4-used in this r=1~1*4~s 5 1.m 1,.sinom a ,m... m s ar. .d.s ,1.h..a1 s. I

2. Ptaps are andaled as a zetacar and an elbow.
3. WaldMP-1sta asume 4=*i-*4*iw pressure drops for gravity Saw conditions and therneceu are not =rda1=4-
4. me Es water Inval is at 287 ft-c in.

S. 2s Merr is full for each ks preneure atmditim. Full was defined as a level just atme the miniaman inval allowed to seat tacenismi fa-4*4*-=. De 1 9 4=1 water Inval far the mert was demmin=9 by the low alars =*=L'it; of the tank. She water 1sumi in tbs vuesel is 187'-0". the pipe tram the All of tank to the entry point in the Es was modeled for each of the systaus in tktit 1. De system ftr 13 nit 1 with the hiM nowrata was modelad fortanit 2 analysis. [ the cazurant 5b etstain pips 4 ' imaastric drawings were used for detusmining the langth of -i,

pipe, raaber of fittings, and elevations.. the h itti ---*im, ardiftm9 for use with agaivalent lanyths, was used to estamnine the news.

me.-P,.enz evaw + w fumwn2/2, at i D e variables are as folleus: Q = total flow l Pata =ah9h8Pic PEWEEurt hAEd P = M 5 preneure head rag 62 - =1Mim dih-c= D = pipe diamatar - a = pipe area I l f = grichien factor \\ N N 4 l

T l ~. J 2000 5 b.N'N D 1500 f. b E ,N 2 E (\\.. f-y 1000 w.N' I v> r l' l j SCO E.. 5 O O to 20 30 40 7 l PSIG l RHR1 ~ * - - RHR2 ) ) ~ M 2.3 MAVZTY ped 3t for M IMZIS 1 atd 2 I l 19 TOTAL P.05 l

4 ' VEGP october 26, 1991 Loss of RHR Incident A/E Technical Questions 3. Determine boiling & core uncovery times. Specify calculations based on 88 F starting temperature, taking all conservatisms out. Summarize assumptions on an attachment. j The time to boiling will be 82 minutes ano the time to core uncovery will be 543 minutes (this assumes an initial water temperature of 88 F and a water level elevation of 187'-0"). 'h. -N.t a J

/ i' VEGP October 26, 1991 Loss of RHR Incident A/E Technical Questions What was the impact of a negative pressure in the pressurizer on the 4 1. mid-loop and drain down level instrumentation. ) (Requested 10/28/91)? SCS Is the temperature rise indicated on the RHR heat exchanger inlet i 2a. temperature trends reasonable. (Requested 10/28/91)? SCS 3' Requested additional information in regard to the temperature issue - i 2b. Determine boiling & core uncovery times. i (Requested 10/30/91) SCS Determine initial RCS elevation assuming 10,000 gallons of water was added back into system. Westinghouse to develop a curve (gallons of 3. I water vs.RCS elevation) - This will provide more comprehensive data. l (Requested 10/30/91) Westinghouse j i Detemine if all temporary level indication has common mode failure i 4. for vent path system - provide justification if adequate or alternate design solution if act. l (Requested 10/31/91) SCS Estimate for how long_ the RHR pump B was operating without mini-flow i 5a. (complete loss of flow through pump) since discharge pressure indicated near zero for those times when the pump was still running. (Requested 10/31/91) SCS 5b. How long can RHR pump operate without flow? b (Requested 10/31/91) Westinghouse s 6. Detemine lowest flow rate necessary to maintain core cooling j (assuming 10/26/91, Unit 1 plant conditions) f i (Requested 10/31/91) SCS I.g

  1. 3053 t

1 i i i I t, TOTAL P.02 I

_-_ ~. m i-iggi es 34 FREN SNC-UDCITLE BG4.!C 1D 3RD FL SERVI E P.01 VESP October 26, 1991 Loss of Rift incident A/E Technical Questions s 1. What was the impact of a negative pressure in the pressurizer on the mid-loop and drain down level instrumentation? (asked 10/28/91) Initial Rennenne on october ts. fool [ Visual check of level in RPV indicated l' below flange (193'-11"). At this time, ILI-957 was reading 60%, though it should have been reading 2005. This 405 difference in s approximately 40' Water Column. pan corresponds to an error of The collapse of the elephant trunk on the safety valve indicated that this error was induced by vacuum in the pressurizer. Instruments ILT-11310 and ILT-11320 would not be damaged by pressure or vacuum of this magnitude. Undated Responsa If vacuum existed in the pressurizer, the maximum vacuum which would have existed can be detemined by the height of the water column drained out of the RCS. Assuming that the drain down occurred from elevation 210'-0" to elevation 188'-0", the pressure would be 22' water column vacuum or 5.2 psia. This negative pressure would be applied to the reference legs of the transmitters (ILT-11310 and ILT-11320), the pomanent sight glass (ILG-10401) and the temporary tygon hose. Per vendor literature the transmitters will remain within specified performance and suffer no damage down to 0.5 psia. The sight glass and temporary tygon hose may have collapsed at this level of vacuum. 1 l 4 ) 1 i i 1 % 4 f i 4 d i ' - - ^ - " -^^

M n-el-1991 M05 FROM 9C-MIm.E ENM.1C TO 3 D FL SERU!G P.02 (. VESP i, October 28, 1991 Loss of RHR Incident A/E Technical Questiens l !a. Is the temperature rise indicated on the RHR heat exchanger inlet temperature trends reasonable? (asked 10/28/91) 4 i Per calculation X4C1201304, the decay heat after refueling was 0.2022 times 10 to the 8th BTU /Hr. and the total heat capacity of the water l and metal in the mid-loop condition was 295.882 BTU /F. Therefore, the heat rise would have been approximately 0.58 F/ min. This was an average expected heat rise. However, the water in the vicinity of the core would have a higher heat rise. When flow was shut off, the water temperature in the core incriased. When RHR flow was re-established, the RHR temperature increased initially 1:acause of the increased temperature of the water in the RPV. As the cooler RWST water mixed with the RPV water the temperature trended downward. Therefore, per out initial review, we believe that a temperature of 1070F is reasonable. i i 1 1 r-m.- m

l. VEGP October 26, 1991 Loss of RHR Incident A/E Technical Questions 2b. Requested additional infonnation in regard to the temperature issue - They want to know boiling & core uncovery times (i.e. - same info they i j requested from Duke Power). Calculation X4C1201504 determined the time to boiling and core uncovery. The calculation assumed an initial water temperature of 1000F which will envelope the 10/26/91. incident. A copy of the applicable Section of the calculation is attached. j FnoM Taew.E ord Face llF or z o - tooe, sn.4 (,ay.y 91 tMN 2PTWL i S H ufpDW M.' U MG T~o BeiL '. 1 2 4-H R.4 1 Tomg To LIM CoV M L t otLG t (s.L1 Htf & ASSvHf3 MD er*Veb FragD ' pyu l., Log of AC-NO

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1 P W i-1991 08:37 FROM 9C 6 ENM.lC TD 3RD FL SERVIG P.11 VEGP October 26, 1991 Loss of RHR Incident A/E Technical Questions l 4. Determine if all temporary level indication has common mode failure for vent path system - provide justification if adequate or alterr. ate design solution if not. (asked 10/31/91) Answer was provided in letter log number SG-10756 (Attached). 1 e i h me

N3uH31-1991 08:37 FRO 1 SN~-MXITLE ENM.lc TD .RD FL SERVIG P.12 1, SouthemCompanyServices l nw soumem eecmc neem October 30, 1991 i l Voetle E1 metric Reneratina Plant - IJnits 1 and 2 j Midloop Instrumentation Design File: X780111/REA VG-9010 Log: SG-10756 Security Code: NC j Mr. C. C Miller Manager of Engineering i Vogtle Project - Nuclear Operations 1 Southern Nuclear Operating Company i Post Office Box 12g5 l Birmingham, Alabama 35201

Dear Mr. Miller:

l During the NRC entrance on the RHR incident,from the pressurizer a question was raised as j to why the level instrumentation was vented i instead of to atmosphere. The question was raised because the l' pressurizer was not vented sufficiently during the incident, resulting in inaccurate level indications. In reviewin the design basis of the level instrumentation, it was f determined hat the subject instruments were designed for use during reduced inventory and mid-loop operations. During this time the reactor head must be in place. It would be preferred to have the instrument vent reference connected directly to the reactor head. However, this is not acceptable due to the need to move the head i around. Therefore, the instrument is connected to the pressurizer. l which, during aid-loop operations, should be at the same pressure as the reactor. Venting to atmosphere during this mode would be unacceptable since the reactor could potentially become pressurized l causing an instrument vented to atmosphere to give incorrect readings. During refueling modes of operation when the reactor head is removed, it would be acceptable to vent the level instrumentation to i atmosphere. This would, however, require careful procedure review to i ensure that atmospheric vent valves were closed prior to putting the i head in ;6See for mid-loop operations. h ~mt I f l i i

.~ M).H31-1991 08 G FRD1 9C6 D&t.1C TO 3RD FL SERVIG P.13 i

7..

Mr. C. C. Miller File: X780111/REA VG-9010 October 30, 1991 Log: 3G-10756 Page 2 If you have any questions, please contact me on extension 5335. Sincerely, C. R. er Jr. Proj Design Manager - Vogtle CllMJr/ss ac: Ree stel P==r tornaratian R. :). Kiss

  • m ia H e t - any 100RMS Southern Nuclear Omaratina Connany P. D. Rushton L. A. Warti Southern P =aany Servicet Inc.

J. R. Singhan W. C. Ramsey File t 8, i j i 024031 i ~ e a e

PC)-el-1991 08:38 FRCM SNC-4J0GTLE BG4.!C TO 3RD FL SERV 1 P.15 VMP October 26, 1991 Loss of RHR Incident A/E Technical Questions Eb. How long can RHR pump operate without flow? (asked 10/31/91) Answer provided by Westinghouse. l i 1 d j i j i l j i ) 1 i 1 l } i i i TOTR. P.15

5 ll4td vj@ w ' 0 v4 Won. [$ . '(printcd:ct 10/31/91 17:20) .2 t I' To: TAIN.J.(1048:WES4029) Cc:' GREENWOOD.G.L-(1038:WTX3496). i Cc:.HACKMANN.E.K.(1038:WTX3581) j Cc: COLVIN.E.R (1048:WES2759) Cc: MAGEE.R.D (WST2160) j From:- ' WALKER.L.I (WST6274) Delivered: Thu 31-Oct-91 16:38 EST Sys 1049 (41)

Subject:

RHR PUMP-OPERATION WITH DRY SUCTION 4 ' Mail Id: IPM-1049-911031-149770371 i

      • MED-AEE-6409***

CC: File Room STC-701, 106: GAE/205/2

Th3 Vogtle RHR pump was inadvertently operated with an apparent suction F vortsx condition that resulted in loss of the pump discharge pressure.- The j loma of" discharge pressure indicates that the pump suction experienced either j' dry cr
2 phase conditions.

Operation of a centrifugal pump under these

conditions ~is not advisable due to the potential for mechanical seal damage,

, bsaring damage and waar ring contacting. i l'Tharaisnowaytodeterminehowlongapumpcanrunwiththeseadverse ! auction conditions before pump damage will occur. Westinghouse is aware of

csvaral events of this type that have occurred with RHR pumps that are very
  • similar in design.to the Vogtle RHR pumps.

At least one of these events .reculted in pump seizure,.while other events caused no damage to the pumps. 4 j Tho level of pump. damage incurred, if any, is likely dependent on the length , of cparating time and the amount of water that is available at the pump t, euction. i[ Wh:n a pump is operated under the adverse suction conditions, the level of , pump damage and the continued pump operab l ty is best determined by ii ! insp2ction and testing of the pump.. It is advisable to rotate the pump shaft

by hand and to vent the pump casing prior to operation.

During the substguent operation, visual observation for leakage can confirm that the f pump seal has not been damaged. The pump should be operated while monitoring

ths vibration levels, pump developed head and motor amparage.

If these puremste s meet the Vogtle surveillance test requirements, it will' confirm r , that the pump suffered no internal damage that will affect the pump hydraulic parformance or operability. g I Wsatinghouse recommends that observation and testing as described be used to ! asssen the operability of the'Vogtle RHR pump. i 1%3Vards',1 I J. G.. Dudiak, Engineer . Auxiliary Equipment Engineering 4 I

( FLOWRATE REQUIRED THRU [";ll!^,J.".i"' CORE TO MAINTAIN 195F F L 1300 O g FULL SPENT CORE W \\j R 1100 A T E 900 T \\ H 700 \\ U x ~ 3 0-C x O 2/3 SPE NT COR E N 4 j ~ R 300 g E C l (GPM)100 4, 0 100 200 300 400 500 600 700 800 ,e TIME AFTER SHUTDOWN (HOURS) FIGURE A

J Statement on Loss of RHR 26 October 1991 j 'During turnover from day shift, the off-going SS told me draining of the Rx cavity and vessel to the 192 ft elevation was in progress. The level was at 210 ft. During the Shift briefing, I assigned Chris Hutton to relieve the operator at the tygon tube.- I probably said the words "tygon tube" since that had always been the term used for the cavity /RPV level ~ indicator'in containment. At.about 1040pm, I received a call from the CR stating we .were having a problem with RHR and I immediately returned to the CR. When I arrived, I saw indications of cavitation in the B RHR train on the ERF computer (low press, low flow, and low current). The CR had raised level several inches from when they first saw the problem and the pump discharge press and current were normal when I arrived. To restore j cooling to the RCS, the operators started to increase flow in the B train and at approx. 1500 to 1800 gpa we again saw indications of the cavitation. RPV level was reported to be about 194 ft from the tube. I then instructed the CR to raise level to greater than the cavity floor elevation to remove any question in my mind as to the actual level. l My thinking was that we had erroneous level indication at the time of the cavitation and the only way to ensure adequate level was to restore level to an elevation where we would be assured we had good indication. I contacted the Duty Engineer who had Mike Chance contact me. Mike and I discussed several possible scenarios where we could have actually had level lower than indicated. Ed j Kozinsky was on-site for the planned head lift and he inspected the new tube for potential problems. Ed found the Je#,e upper isolation closed and hold tagged. Since this was the j i i. /T __ tube he operator was monitoring it seemed to explain the

OLJgef*} tube 1 eve We had the clearance released and'placed the new n service.

It now agreed with the old tube and actual i cavity level. Since we now felt we had two accurate leve indicators, we decreased the level to 192 ft for the Asa set. j j s j I _ha. ( /o/16 NQ-6h g4,E" %g k, v. O5C c~- l r b i / /47 I w er

6Jt:18Gl'illGy wsH W Card # / Qf QC"- ILMInst 1 O Unit 2 O Common g: (Additionit Sheets Attached? 2'fis --. No) Descriptiontf Deficiency bsbr abs /2600-C shte % l.6 w.ith a m+u A wur 1: AaS E45 n -h ex % 4 m224'C,s% % % s % O n k e/ \\ 'lW 2 h h i4s &wtaa e a C / M n ahk 'BLna, -L fri w a b, s w d a J. /MP '# was Es/4 Af/#4 2Pvd 5 LIh"'B' m >Ld 2/h % A (Md4 wic-um re dific-ftM ekd ANA 'Tb B /he-da+ g MPL Tag Number / I Location Of Deficiency 1 g g What is Affected By The Deficiency?, g I ( 7 / / l How Was The DeflCHinCy Discovered? &WM 0 Event Time gpg Date//[g[g/ Discovery Time 3pg/ Date/gg[ Discovered By? f[g [ Work # M Dept. / y 2: Shift Supervloor Review Name Of USS Reported To? f,,- Time gfd Date //E_ ~/ hg [gg M /g Plant Mode / Condition: is immediate Notification Rkuired? byes ho # e If Yes, D1 Hour, 04 Hour, or 024 Hou,r Reported: Date Time Tech. Spec, Required Action Taken? d ONo ON/A f, R List Applicable Tech. Spec. Section(s) gg E E Summartze Compensatory Action Taken: ht/ W l 4 a.s H fl /f LCO initiated: Oyes effo' # Type: Info LCO Fire WRT Initiated: Oyes d# d ~ re j 7_ pf Time oye Date Signature of usS 706666C 1230 i i 4~ me ..e ) (g O

10- 2 0-9l 3: NSAC Evolustion/ Review (Check Appropriate Box) Date Received: g A. No Defciency Card Required. Send Copy To Responsible Dept., Close Onginal )( Renma* Deficiency Rep *

  • W l uby LER / DLER tt I-91 W a

) / C. Deficiency, Not Reportable- $eL 10 df {1 ) E \\ s ai d Responsible Dept.: m it. sma /yf Ac) i.01 I NSAC Reviewerggg Date:g,yj,y/ NSAC Supervisor:h [/ Date:f[ ff f dCCd kuired Oyes ONo l 4: -

- ' : Dept. Review Explanation:

Disposition: 1 a 1 5 3 Cause Code (s): Event Code (s): Goch.Spt initials: Causing Dept (s): Department Manager: Date: 106686C 1238 1 i

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4-1 .o PAGE 1 OF 1 ATTACHMENT: ' DC #1-91-475 BLOCK 3: EXPLANATION OF REPORTABILITY The following facts are noted regarding the RHR Train "B" event which occurred on 10/26/91: e After observing indications of cavitation, Plant Operators responded appropriately per contingency actions provided in AOP 18019-C to reduce the flow of the 1B RHR pump, to stop the RCS draindown via the 1 A RHR pump, and to realign the 1 A RHR pump to raise RCS level. This action was effective in preventing the 1B RHR pump from becoming air bound and prevented air entrainment/ cavitation from occurring for the.1 A RHR pump. Therefore, procedural controls (and training) were effective in minimizing the operability impact on the 1 B RHR pump and in preserving the operability of the 1 A RHR pump. e-After the flow of the 18 RHR pump was reduced, operators closed 1HV-8716A and opened 1HV-8809A to stop the draindown via the 1A RHR pump. This action momentarily (for approximately 2 minutes) placed the RHR Train in the recirculation / shutdown cooling mode at approximately o cavitation was observed for the 1 A RHR pump at this flowrote. (After o his alignment for approximatelypinutes, the 1 A RHR pump was shutdown and the suction was realigned to the RWST to refill the RCS). e While shutdown core cooling flow provided by the 1B RHR pump was temporarily interrupted, it is noted that the 1B RHR pump did not become air bound and it was never necessary to completely shut the pump down. Based on these facts, it is concluded that plant procedures and operator training provided effective mechanisms to ensure that the RHR system remained capable of performing its shutdown cooling safety function and, therefore, no reportable condition is considered to exist. However, due to generic industry implications, a voluntary LER will be submitted. ['d$ d.k / / t'- V-9/ W. K. Smith 10/31/91 (D/ I/ ~ b l V

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/ ( "% UNITED STATES \\;. NUCLEAR REGULATORY COMMISSION i 1 Office of Governmental and Public Affairs \\...l..) Region 11 ici undetta street, N.W., Adanta, GA 30323 No.: 1I-91-70

Contact:

Ken Clark FOR IPMEDIATE RELEASE Telephone: 404-331-4503 (Tuesday, October 29,1991) NRC STAFF SENDS AUGMENTED lhSPECTION TEAM TO PLANT V0GTLE TO REVIEW PROBLEMS ASSOCIATED WITH HEAT REMOVAL SYSTEM The Nuclear Regulatory Comission staff has dispatched a special Augmented Inspection Team (AIT) to the Vogtle nuclear power plant, operated by Georgia 1 Power Company near Augusta, Georgia, to review events associated with a temporary degradation of the Unit I heat removal system while the reactor was shut dean for refueling. i 30 p.m. (EST) on Saturday, October 26,1991, Unit 1 was shut down for l ...ueling with the reactor head off. The reactor cavity had been filled with water during refueling and was being drained down to the top of the reactor vessel so that the head could be replaced. Water to cool the residual heat from the reactor core was being circulated by one of two residual heat removal system pumps. Tne other pump was being utilized to drain the reactor cavity. As draining continued, plant personnel observed indications of " cavitation" on the pump being used to recirculate the water, indicating that the pump was pumping air instead of water. Plant operators took imediate action to restore the pump suction and full flow; however, the reactor was without full decay heat flow for about 25 minutes, which allowed the core temperature to rise from 90 degrees (F) to about 107 degrees (F). NRC officials said initial information indicated that there was no release of 4 radioactivity to the environment and that the water level in the reactor vessel remained at a high enough level to maintain a safe shutdown margin with no threat to the fuel. The special NRC inspection team will review all circumstances associated with i the event, and a report of findings will be made available to the public when -the inspectionWtomplete. i lN j7 hg

1-Meeting Attendanc's Record GeorgiaPower A j u .ns Poroo NRC ENTRANCE MEETM C, om conooci.a e, Ocr. 29,199I i PttNE Name (Pnntj Title (Print) DepartmentiCompany HERBERT BEACHER. 5A ENGR. 3%9 /i38 GEosiisiA power / TECH.Jwpg(7* h/s 94W OPc f>tneMS z-tuAD d Eo b Ein w xJ SUP4 OPb TUOU'1 396 3 GPC. i 7 E.Saaerzarwte-. Ma. dar.c U?dA4 36/8 OIP/G L C uc u e. s r; a a s a a w eenv.soa-ss en. s228 S A E a. /Sov nicaa,at uen i

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i l October 29,1991 j AGENDA i i j e INTRODUCTION GPC/ BILL SHIPMAN NRO/ PIERCE SKINNER i e EVENT DESCRIPTION BARNIE BEASLEY i 4 e PRELIMINARY CAUSES BARNIE BEASLEY e ADDITIONAL OBSERVATIONS BARNIE BEASLEY i e DISCUSSION 4 i l"' i l ll' c

SEQUENCE OF EVENTS {r ot' '((,.. ' ' ' 10/26/91 1830 APO IN THE CONTAINMENT, ' CALLED TO ESTABLISH COMMUNICATIONS FOR TYGON TUBE WATCH FOR CAVITY DRAINDOWN. / 1833 CAVITY PUMPDOWN BEGUN. CAVITY LEVEL AT 210' 4" (23% COLD CALC. PZR LEVEL). (<.., 1848 NIGHT SHIFT USS RELIEVES. CAVITY LEVEL AT 208' ' ' T", " ' i 1930 ONCOMING PEO RELIEVING ON STATION NOTICES THAT VALVES

i. *'

WAS STOPPED. CAVITY LEVEL AT 205' 6" (APPR.15% , /" /HlH, DlH, AND DIL SHUT, GD, GV OPEN AND CAPPED. DRAINING ~;',,.,.,,,p 2200

jRCS LO LEVEL ANNUNCIATOR WAS RECEIVED. DRAINING WAS i

STOPPED BY CLOSING 8716 AND OPENING 8809A. AFTER VISUAL, j TYGON,AND 1LG10401 CHECKED SAT AT 194', BELIEVED 1Ll957 WAS IN ERROR. DRAINDOWN PROCEEDED SLOWLY. 2235 INDICATIONS OF REDUCED RHR B (TRAIN PROVIDING COOLING) DISCHARGE PRESSURE AND FLOW.1Ll957 INDICATED ABOUT 30% l (APPROX 188' 3" -TOP OF HOT LEG) AND 1Ll950 INDICATED ABOUT i 60% (187' 6" - JUST ABOVE MID LOOP) RO INFORMED USS AND SS. B RHR PUMP WAS PUT ON MINIFLOW AND THE "A" RHR PUMP WAS REALIGNED TO RAISE LEVEL. l f f 2245 SHIFT SUPERINTENDENT DETERMINED EVENT DID NOT REQUIRE EMERGENCY CLASSIFICATION. i l l 2316 AFTER LEVEL HAD BEEN RESTORED "B" TRAIN RHR COOLING MODE WAS REESTABLISHED. 2330 SHIFT SUPERINTENDENT CONTACTED OPERATIONS MANAGER TO DISCUSS EVENT. j 10 27 91 0015 DISCOVERED NEW SIGHT GLASS ISOLATION VALVE "HIL" (TOP a ISOLATION TO THE NEW SIGHT GLASS) CLOSED AND RED TAGGED. 4 l- ., tHEPA FOUND CONNECTED TO PZR SAFETY FLANGE (INTENDED RCS ) VENT PATH). DUCT HOSE WAS COLLAPSED. HEPA VENT TURNED OFF AND POSITIVE VENT PATH ESTABLISHED. l 0200 SHIFT SUPERINTENDENT CONTACTED OPERATIONS UNIT SUPERINTENDENT AND CONFIRMED NO REPORTABILITY INVOLVED. { i

=. b 'l ' ' SEQUENCE OF EVENTS 10/27/91 .y.i ' 0630 ASSISTANT GENERAL MANAGER INFORMED OF EVENT WHO THEN PROCEEDED TO THE SITE TO START EVENT INVESTIGATION AND ORGANIZE CRITIQUE TEAM. EVENT TEAM ESTABLISHED. - g* 't-{ 0830 u,.' i. -. ' ,.e \\. b t s C jf \\ j ~.I. t st.( I t

i i. PRELIMINARY ROOT CAUSES AND CORRECTIVE ACTIONS 1. PROCEDURAL CONTROLS FOR REDUCING RCS LEVEL OTHER THAN INITIAL I RCS DRAIN DOWN DID NOT CONTAIN SUFFICIENT INSTRUCTIONS TO REVERIFY SPECIAL LEVEL INSTRUMENT INSTALLATION, PROPER VENT PATH, OR DRAIN FLOW WITH UPPER INTERNALS INSTALLED. j j CORRECTIVE ACTION: 1 e TEMPORARY PROCEDURE CHANGES HAVE BEEN MADE TO INCLUDE APPROPRIATE ADDITIONAL CAUTIONS AND REQUIREMENTS TO REVERIFY LEVEL INSTRUMENT INSTALLATION AND VENT PATH AVAILABILITY j 2. OPERATOR AWARENESS OF PLANT CONrlGURATION DURING OUTAGE EVOLUTIONS NEEDS IMPROVEMENT. EXAMPLES INCLUDE: J-l- i OPERATOR KNOWLEDGE OF THE SIGHT GLASS MODIFICATION / CLEARANCE e y o INSTALLATION OF HEPA FILTER ON VENT CORRECTIVE ACTION: e STEPS WILL BE TAKEN TO ASSURE MORE COMPREHENSIVE AND METHODICAL APPROACH TO ASSURE OPERATOR KNOWLEDGE OF CHANGING PLANT CONFIGURATIONS DUE TO MAINTENANCE OR j MODIFICATIONS. [ e BRIEFINGS WILL ASSURE OPERATORS ARE AWARE SYSTEM CONFIGURATION, INCLUDING PROPER INSTRUMENTS TO MONITOR. l e TURNOVER REVIEWS TO ASSURE KNOWLEDGE OF CLEARANCES AND l MODIFICATIONS WILL BE REQUIRED. i e THE MODIFICATIONS PROGRAM WILL BE REVISED TO ALLOW EASIER j REVIEW OF MODIFICATIONS STATUS BY OPERATORS. I 3. INVESTIGATION OF ANOMALIES WERE NOT SUFFICIENTLY THOROUGH AT l THE TIME. i e CONTROL ROOM DID NOT CORRECT APO'S ACTION TO l MONITOR THE NEW SIGHT GLASS INSTEAD OF INTENDED i TYGON. !l" e' ATTRE TIME OF LOW LEVEL ALARM VISUAL OBSERVATION OF SIGHT GLASS, TYGON TUBE, AND VESSEL, CREATED ERRONEOUS UNDERSTANDING OF ACTUAL LEVEL. e LINEUP PROBLEMS WITH THE NEW GAGE GLASS WERE NOT FULLY INVESTIGATED. I !~

.=== -- -

I i I CORRECTIVE ACTIONS: j e CASE STUDY TRAINING WITH OPERATORS INVOLVED WILL BE CONDUCTED WITH ALL SHIFT CREWS FOCUSING ON l 1 PROBLEMS DISCUSSED IN THIS EVENT TO ENSURE LESSONS LEARNED AND FULLY COMMUNICATED. l 4. CREW BRIEFING IMPROVEMENTS ARE NEEDED, ESPECIALLY WHEN ] COMPLEX EVENTS CONTINUE INTO SUBSEQUENT SHIFTS. ) CORRECTIVE ACTION: 1 e INCREASED BRIEFINGS WILL BE REQUIRED BY PROCEDURE 10000-C AND 00053-C. ALL SHIFTS WILL CONDUCT BRIEFINGS PRIOR TO BEGINNING INFREQUENTLY USED OR COMPLEX TASKS. EMPHA818 WILL BE PLACED ON TASKS l THAT CONTINUE PAST ONE SHIFT. j 5. OPERATOR KNOWLEDGE OF WATER LEVEL BEHAVIOR DURING DRAINDOWN i WITH THE UPPER INTERNALS INSTALLED NEEDS UPGRADING. i l CORRECTIVE ACTION: j PROCEDURES HAVE BEEN MODIFIED TO LOWER DRAIN DOWN l FLOW TO APPROXIMATELY 100 GPM WHEN LESS THAN 194 FEET WITH THE INTERNALS INSTALLED. e OPERATOR TRAINING WILL BE CONDUCTED ON POSSIBLE l MISLEADING VISUAL INDICATIONS DURING DRAINDOWN j WITH INTERNALS INSTALLED AS WELL AS CONSEQUENCES j OF POSSIBLE VORTEXING WITH TOO RAPID A DRAIN RATE. i i i 4 h d%f i I i y,-'w w

== 1 1. PRELIMINARY OBSERVATIONS-GOOD 1. OPERATIONS STAFF REMAINED FOCUSED ON THE RCS LEVEL CHANGE EVOLUTION AND ENSURED A DEDICATED OPERATOR WAS ASSIGNED WITH NO COLLATERAL DUTIES. 2. THOUGH THE OPERATORS ALLOWED THE USE OF THE NEW GAGE GLASS BEFORE IT WAS RELEASED FOR USE, THERE WAS NEAR CONSTANT COMMUNICATIONS TRACKING THE LEVEL AND DRAINING PROCESS. THE OPERATORS REQUESTED CONFlRMATION WITH THE TYGON TUBE WHICH WAS THE AUTHORIZED LOCAL LEVEL INSTRUMENT. 3.. THE RHR FLOW AND DISCHARGE PROBLEMS WERE PROMPTLY DETECTED TO PREVENT PROBLEMS WITH THE PUMP. SENSITIVITY AND PROCEDURES TO MONITOR FOR CAVITATION PROBLEMS WAS EFFECTIVE BASED ON TRAINING AND KNOWLEDGE OF INDUSTRY EVENTS. 4. ACTIONS REQUIRED BY THE AOP WERE TAKEN PROMPTLY, AND ONE TRAIN OF RHR WAS ALWAYS AVAILABLE. 5. EPIP AND REPORTABILITY DISCUSSIONS OCCURRED PROMPTLY AND PROPER DETERMINATIONS MADE. //.' ([ A. A

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i ADDi?lONAL CONCLUSIONS I e "A" TRAIN RHR PUMP SHOWED NO SIGNS OF CAVITATION AND WAS } MAINTAINED OPERABLE. I j e "B" TRAIN RHR PUMP CAVITATION CEASED PROMPTLY AFTER FLOW l REDUCTION AND PUMP WAS NOT STOPPED. INSPECTING AND TESTING HAS SHOWN NO OPERABILITY PROBLEM AND THE PUMP IS OPERATING NORMALLY TODAY. e ESTIMATED PEAK RCS TEMPERATURE WAS APPROXIMATELY 107' F. B'l f j e LOWEST RCS LEVEL DURING DRAIN IS CURRENTLY INCONCLIJSIVE. i P 3 i s 4 -m+ t.- 5 a d- "r ,,6

I TRAIN A RHR l 1000-- b. soa = t PUMIS DISCHARGE TEMP \\ goo = r i N ~ 70' = RCSTO RWST PUMPDOWN RCS RH R SUCTION FILL FROIA Al.lGNMENT RCS RWST TO RCS FILL FROM 1000 -- RWST i GPM G 53 = PUMP FLOW I J l -5 0 5 10 15 20 25 30 45 2235 2240 2245 2250 2255 2300 2305 2330 ren a r

[ TRAIN B RHR llo "F l i i 100o t \\1 3 i /!! \\' i ) k) PUMt DISCHANGE TEMP 90* = J i i 300 = i 10e - t i 60 = l AMPS yteu 3800 GPM t I f 3800 GPM PUMP CURF,ENT / FLOW 3000 GPM 2600 GPM P i - r j _g 1800 GPM i 30 - 300 GPM I AMPS Il [ RllNIFLOW MINIFLOW ~ 1800 GPM i l, l \\ _ l l -5 0 5 10 15 20 25 30 55 2235 2240 2245 2250 2255 2300 2305 2330

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m .,.,..... ~.. -- -,- m ..~m.. .3 l Procedure 900-Rewsson Page h lfd' VEGP' 54840-1 4 23 of 23 TOP OF PRESSURIZERM X ><I- ' - 1201-HV TEMP VENT JU4-100 VALVE ' I HIL fgy d k E Dil c-TEMPORARY S PLASilC (IF REQ'D) HOSE i x- = S ~~ EG _ ASS / . = FV1 - 1AJ14 FV2 M ALL VALVE i NUMBERS ARE PRECEDED BY DlH GD 1-Lc-10401 ' I HIH d L TEMP DEMIN 1-1201-PI FILL U4-003 LOOP 1 INTERBEDI ATE H3 CONNECTION LEG M ]G )q ><}. m. .o..- . m + i a.- a

i , enoCEDURE ND. REWSION PAGE NO. } PTDB-1 TAB 8.0 4 3 of 5 1 2 TAB 5.2 i 848' = HID LOOP LEVEL ' 08 *! i IF -344 R ENTATION ~~ ~~~~~ a ~ =808' NiskalH$ i N = aca = ~ hyf;p- ^

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-801' -too' ,s,. i.<. TM \\([ N =197' = M -194' i 19 8 '.. -194'.'-

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u - 9..s OLAS$ #3 //3 gs [( 3/O NOTE: e This Operator Aid is to be used only when in mid-loop confesuration and temporary level instrumentation is installed. I e LI-957 and LI-950 may read higher than actual RCS level if opening between RCS and PRZR is blocked and RCS is pressurised. e Lower Tap,(Jpper_ Tap LI-950 HL 1" FRZR LT-0459 Upper Top c//h LI-957 HL 4* PRER LT-0459 / FIB /fo U)per Tap

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V0GTLE, UNIT 1 91-19 INADVERTENT REACTOR VESSEL DRAINDOWN OCTOBER 26, 1991 4 PROBLEM: A LOSS OF SHUTDOWN COOLING RESULTED WHEN REACTOR COOLANT WAS DRAINED TO THE POINT WHERE A MID LOOP ALARM ACTIVATED AND RHR PUMP "B" 4 CAVITATED. TEMPERATURE INCREASED APPROXIMATELY 20F DURING THE 20-25 MINUTES THAT SHUTDOWN COOLING WAS LOST. CAME: THE VENT PATH FOR LEVEL INDICATION BEING USED FOR THE DRAINDOWN WAS RESTRICTED CAUSING AN ERR 0NEOUS HIGH Li"/EL INDICATION. ALSO, AN ISOLATION VALVE TO A GAGE GLASS WAS FOUND TO BE CLOSED AND RED TAGGED. % Aay W 9 a e.. :-bo 6 :., e o e... - L ~ wb 2: k SAFETY SIGNIFICANCE: bM& ' # d "A " CN LOSS OF SHUTDOWN COOLING COULD RESULT IN UNC0VERED FUEL DUE TO B0ILING 4 ) AND A SUBSEQUENT RADIOACTIVE RELEASE. DISCUSSION: 4 0 V0GTLE UNIT 1 WAS 42 DAYS INTO A REFUELING OUTAGE. 0 ONE RHR PUMP ("B") WAS RUNNING TO REMOVE DECAY HEAT. 0 THE LICENSEE WAS PERFORMING A PLANNED PARTIAL DRAINDOWN OF THE REFUELING CAVITY. M u'u d'Ng 4 *4.t 0 THELICENSEEWASUSINGANRCSSITEGLASS,ATYGONTUYE,AND DIRECT OBSERVATION FOR LEVEL INDICATION. TWO PERMANENT LEVEL 4 INDICATORS WERE ALSO AVAILABLE (LI-957, L1-950), CONTACT: D. GAMBERONI, NRR/EAB AIT: YES

REFERENCES:

TELEPHONE CALL FROM REGION 11 SIGEVENT: TBD

V0GTLE, UNIT 1 91-19 i 4 0 THE FOLLOWING SEQUENCE OF EVENTS OCCURRED: i ~1833: CAVITY PUMPDOWN COMMENCED FROM A CAVITY LEVEL 0F 210'4". ) l l 1930: THE ONCOMING' SHIFT NOTICED THAT THE RCS SITEGLASS j WAS ISOLATED. DRAINING WAS STOPPED. THE RCS SITEGLASS WAS FILLED AND VENTED.y DRAINING WAS RESUMED. P h " # " # [ '". en.L L.~ d 2200: REACTOR CAVITY LEVEL WAS AT 194' BY RCS SITEGLASS i AND TYGON TUBE. DRAINING WAS STOPPED. -TN 19W-tEVEt--ANNUNG4* TOR-WAS-RECE4VED. LI-957 WAS READING 100%. AN OPERATOR TAPPED LI-957 AND IT j DECRE'ASED TO 60%. THE CONTAINMENT LEVEL WATCH c CHECKED THE CAVITY LEVEL AND REPORTED IT AT THE VESSEL FLANGE (APPROX. 194'). THE CONTROL ROOM CONTACTED I&C AND AFTER DISCUSSIONS THEY DECIDED i L1-957 WAS NOT READING CORRECTLY. THE CONTROL ROOM OPERATORS BELIEVED THEY HAD THREE RELIABLE j INDICATIONS OF LEVEL (RCS SITEGLASS, TYGON TUBE, 4 AND. VISUAL) S0 THEY CONTINUED THE DRAINDOWN. 2235: INDICATIONS OF REDUCED RHR "B" DISCHARGE PRESSURE AND FLOW WERE OBSERVED. LI-957 INDICATED ABOUT 30% AND LI-950 INDICATED ABOUT 60% (JUST AB0VE i MID LOOP). THE MID LOOP ALARM ACTIVATED. i i "B" RHR PUMP WAS PUT ON MINIFLOW AND THE "A" RHR --- PUMP WAS USED TO RAISE LEVEL. 2316: AFTER LEVEL HAD BEEN RESTORED, "B" TRAIN RHR COOLING MODE WAS REESTABLISHED WITH NO OBSERVABLE PROBLEMS. i 9 g g N W4 9 e 8e e s ,n..

~ 1 vi 'V0GTLE' UNIT'1 91-19

J 0015:

RCS SITEGLASS TOP ISOLATION VALVE ("HIL") WAS l FOUND TO BE CLOSED AND RED TAGGED. THE VALVE WAS LOCATED IN A HIGH RA'DIATION AREA, l A HEPA VENTILATION UNIT WAS FOUND CONNECTED TO THE REMOVED PRESSURIZER SAFETY VALVE WHICH WAS THE VENT PATH FOR THE RCS SITEGLASS AND THE i TYGON TUBE. THE ELEPHANT TRUNK BETWEEN THE i VALVE FLANGE AND THE HEPA WAS COLLAPSED. i INVESTIGATION REVEALED SHUTDOWN COOLING MAY HAVE BEEN LOST FOR 20-25 MINUTES AND TEMPERATURE INCREASED FROM 90F TO 107F OVER THAT PERIOD OF j.. TIME. 1 i SJMILAR EVENTS: 0 ON MARCH 8, 1991, OCONEE UNIT 3, AT 24 DAYS INTO_A REFUELING l l OUTAGE, LOST 14,250 GALLONS OF PRIMARY WATER DURING TESTING. THE LOW-PRESSURE INJECTION PUMP CAVITATED AND SHUTDOWN COOLING i WAS LOST FOR 18 MINUTES. THIS EVENT WAS REPORTED TO INDUSTRY VIA IN 91-22. O IN 90-55 DISCUSSES THREE EVENTS WHERE RCS INVENTORY WAS INADVERTENTLY REDUCED AS A RESULT OF DEFICIENCIES IN HUMAN PERFORMANCE WITH THE REACTOR IN A SHUTDOWN CONDITION. 0 ON MARCH 20, 1990, V0GTLE UNIT 1, WHILE IN MID-LOOP OPERATION, LOST SHUTDOWN COOLING FOR 36 MINUTES DUE TO A LOSS OF VITAL AC POWER. THIS EVENT WAS REPORTED TO INDUSTRY VIA IN 90-25, i IN 90-25 SUPPLEMENT 1 AND THE IIT REPORT, NUREG-1410, FOLL OWilP !~ 0 AN AIT W DISPATCHED.T0 THE SITE ON TUESDAY, OCTOBER 29, 1991, L TO GATHER INFORMATION PERTAINING TO THIS EVENT, 0 THIS EVENT WAS NOT REPORTED VIA A 10 CFR 50.72 REPORT. f , ~,

Vogtle 1 Briefing 91-19 Noctoumt No. ggVillON i

  • g PTDB-1 TAB 8.0 PAgg No.

4 4 of 5 TAB 8.2.1 HID LOOP LEVEL INSTRUMENTATION: LI-950 L1-957 s)! 5 ill P. i9l d: i l l ] e I h I! g i a i-j:j i i !3l i ' em -. 'Nt %q l Eg' jtJJld n/as, -. g, ,.u

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4 Q<_ October 29, 1991 u f d PRELIMINARY NOTIFICATION OF EVENT OR' UNUSUAL OCCURRENCE PNO-II-91-70 This ' preliminary notification constitutes EARLY notice of-events of POSSIBLE safety or .The information is as initia'"y received without veriff- . public interest significance. cation -or evaluation, and is basically all that is known by the Region 11 staff on this ~date. ~ Georgia. Power Company Licensee Emergency Classification: FACILITY: Vogtle Unit.1 Notification of Unusual Event . Docket No. 50-424 Alert Waynesboro, GA . Site Area Emergency General Emergency X.Not Applicable

SUBJECT:

DEGRADATION OF DECAY HEAT REMOVAL AT V0GTLE - AUGMENTED INSPECTION T DISPATCHED' 26, 1991, while Vogtle Unit I was in Mode 6 At approximately 10:30 p.m., on October (refueling) with the reactor head off, the licensee was draining down the reactor cavity to The 'B' Residual Heat just below the vessel flange to facilitate replacing the head. Removal (RHR) pump was in use for decay heat removal, while the 'A' RHR pump was in use for draining.the reactor cavity. As draining continued, the licensee observed indications.of -cavitation on the 'B' RHR pump. The 'B' RHR pump was placed on miniflow, and the ' A' RHR Decay heat removal was pump was realigned and used to raise reactor water level.The reactor was without decay heat remova reestablished after reactor level was restored. for 'approximately 25 minutes,. and core temperature rose from approximately 90F to 107F. There was no release to the environment and the reactor safe shutdown margin wer f maintained. Aa Augmented Inspection Team (AIT) consisting of a team leader and four inspectors has been t dispatched to-Vigtle. The team will arrive on-site and begin the inspection on October 29. Unit 2 is cv.ntly operating at full power. Th3 licensee has been responding to media inquiries; the NRC will issue a press release. The NRC recei" itial notification of this event by telephone from the licensee to the

resident ins, at approximately 12:00 p.m., (EST) on October 27, 1991.

t 'This information is current as of.11:30 a.m., on October 29, 1991. CONTACT: S -+t Sparks - 84r 619 UTETRTisui. ~ J0ne White-M' land Nat'l DCS((OriginalIE34)Transportatio [ Flint North Bank '"dj Regions MAIL T0: DOT O Chairman 5elin TRT Region I . sConn. Rogers OIGL Region II 6 -P em. Curtiss ~4E90. Region III FAX TO: Chairman Selin ~ Cunn. Remick - NRC Ops Ctr-Region.IV O!G 0GC Region V INP0 OCA' LICENSEE (Corporate) GPA/SLITP/PA. Nicholson Lane Phillips Bida and site EDO - -RES ACRS RI! Resident NRR1 PM, NRR

NMSS

. Street EWW i OE: L.. 'DR XKAP ,d i 15520:- 10/29/91 0 TO REGIONS AND HQ d-ocean * * * * * * * * * * * *mnrwww* * * * * ** * * * * * * * * * * * * * ** * * * * * * * * * * * * * * * * * * * * * * * ** * * * * * * * * * * * * * * * * * ** iu ^ O l .}}

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