ML20129F628
| ML20129F628 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 11/28/1990 |
| From: | Hayes B NRC OFFICE OF INVESTIGATIONS (OI) |
| To: | Carr NRC COMMISSION (OCM) |
| Shared Package | |
| ML20129F106 | List:
|
| References | |
| FOIA-94-208 NUDOCS 9610070007 | |
| Download: ML20129F628 (200) | |
Text
_ _
LIMik :STRIBUTION -- NOT FOR PUBLICLOSURE D 1
~
November 28, 1990 MEMORANDUM FOR: Chairman Carr THRU: Hug L. Thompson, Jr. /s/s Depu Executive Director for clear Materials Safety, Safeguards, and Operations Support FROM: Ben 8. Hayes, Director Office of Investigations DM for BH
SUBJECT:
INVESTIGATION NOTIFICATION The following information is submitted in compliance with the Commission directive that you be informed whenever the Office of Investigations (OI) opens a new investigation:
Vogtle Elec. Gen. Plant Case No. 2-90-020 Opened: 11/06/90 On November 1, 1990, the Regional Administrator, Region II, requested investigative assistance after two former employees of Georgia Power Company petitioned the NRC asserting that the licensee knowingly provided false statements intending to mislead the NkC with false assurances about the reliability of the diesel generator, and that SONOPCO knew the diesel generator had actually continued to experience an excessive number of trips, failures, and problems similar in nature to the failure which led to the March 20, 1990, station blackout. ECD 08/91.
If you desire more detafis on this or any other OI investigation, please let me know.
cc: Commissioner Rogers Comissioner Curtiss -
Commissioner Remick J. Taylor, EDO H. Thompson, Jr. , DEDS S. Chilk, SECY D. Williams, OIG W. Parler, OGC S. Ebneter, RA:RI!
Distribution:
s/f(2-90-020) c/f B. Barber, 01:HQ P. Thompson, DI:RII EDO r/f
.RI f b !:RII h:D :D JVorst.g DMurphy rtu , ayes l !/7 90/11/ 90/11/;f g' 0/11 /11/f
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LIMITED DISTRIBUTION -- NOT FOR PUBLIC DISCLOSURE WITHOUT OI APPROVAL 9610070007 960827 ,
PDR FOIA p- COLAPIN94-208 PDR ,
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- ,s.c,......._. .. . - , o.i l On 3-20-90, Unit I was in a refueling outage and Unit 2 was operating at 100%
power. At 0820 CST, the driver of a fuel truck in the switchyard backed into a
- support for the phase "C" insulator for the Unit 1 Reserve Auxiliary Transformer
! (RAT) 1A. The insulator and line fell causing a phase to ground fault. Both Unit 1 RAT 1A and Unit 2 RAT 2B High Side and Low Side breakers tripped, causing
- a loss of offsite power condition (LOSP). Unit 1 Diesel Generator (DG) 1A and Unit 2 DG28 started, but DGIA tripped, causing a loss of residual heat removal
! service for maintenance. A Site Area Emergency (SAE) was declared and the site Emergency Plan was implemented. The Reactor Coolant System heated up to 136 degrevst fMarf0"degrwest twfew-h06F wer emergenerstsete6 a& 986&CET and
- RHR was restored. The initial notifications were not made within the required '
15 minutes due to the loss of power to the Emergency Notification Network (ENN).
At Ogl5 CST, the SAE was downgraded to an Alert after onsite power was restored.
l
! The direct cause of this series of events was a cognitive personnel error. The l truck driver failed to use proper backing procedures and hit a support, causing
! the phase to ground fault and LOSP. The most probable cause of the DGIA trip was the intermittent actuation cf the DG jacket water temperature switches.
- Cbrrective actions includd strengthening policies for control of vehteter, l extensive testing of the DG, replacement of suspect DG temperature switches, and improvements in the ENN system.
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'. REQUIREMENT FOR REPORT This event is reportable per:
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i unplanned Engineered Safety Featurea) 10 CFR 50.73 (a)(2)(iv), because an i ActuetforSyrtersequencer startea, a(nd b.)~ Technical SESF!
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this repor,t serves as a summaryencyofevent. .
the Site Are Additionally, i
UNIT STATUS AT TIME OF EVENT Unit I was in Mode 6 (Refueling been shut down since 2-23-90 for) a 45 day scheduled The reactor had refat;
- reactor core reload had been completed, thethe ueling outage.. The initial tens reactor vessel head studs was complete, andofawaiting the outage te Coolant System (RCSpermission from the control ng. Reactor room to be i Residual Heat Remova)l level was(RHR being maintained at mid-loop with t i
temperature 90 degrees F. of the RCS pump inwas)being service for decay maintained heat removal.atThe approxi
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- Due to the refueling outage maintenance rogress some activities in p
{ equipment configurations. was out of service and several systems were
- Transformer The Train 8 Diesel a required 36 month maintenance inspection. Generator (DG18)
The Train B Reserve Auxiliary i
i Train Train B Class A RAT 1A through (RAT 18) had its alternate supply been breaker removed The A
i was from thebeing powered switchyard. from Generator All Steam the Unit Auxiliary Transformers (ll .
removed, but only S UAT by backfeeding t
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Maintenance personne/G's I and rimary 4 had their p(S/G) nortle dans manways secured.
i on S/G's 2 and 3.
j ,, RCS level was being maintained at repairs and the S/G manway restorations. ,
i manwar was removed to provide an RCS vent path,In addition, the pressu ;
3 C.
i DESCRIPTION OF EVENT i On March 20, 1990, escort the plant entered operating staff the protected area in a rfuel ver with truck.at a security approxim i
i belonging to a service gro,up used to perfom various did,oot..neet fuel. driver checked the weldingservices. machine The that was a and found that it ii j, backing out1A.of the area when he hit aphaseC" insulator for RAT support holding i fault, and the transformer breakers tripped.The insulator and line
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' At 0820 CST, both Unit 1 RAT 1A and the Unit 2 RAT 28 High Side and Low Side Breakers trippaid tausing a loss of offsterposer condition (LOSP) trthe < -
Unit 1 Train A Class IE 4160 volt bus IAA02, the Unit 2 Train B Class IE bus 28A03, and the 480 volt busses supplied by 1AA02 and 2BA03. The Unit 1 Train B Class IE 4160 volt bus IBA03 also lost power since RAT 1A was feeding both Trains of Class IE 4160 volt busses. The loss of power caused the associated ESF Actuation System Sequencers to send a start signal to one Unit I and one Unit 2 Diesel Generator. DGIA and DG2B started and sequenced the loads to their respective busses. Further description of the Unit 2 response to this event is provided in LER 50-425/1990-002.
1 One minute and twenty seconds after DGIA started and sequenced the loads.to the Class IE bus, the engine tripped. This again caused an undervoltage (UV) condition to class IE bus IAA02. The UV signal is a maintained signal at the sequencer. However, since DGIA was coasting ewn from the trip, the shutdown logic did not allow the DG fuel racks or starting air solenoids to open and start the engine. This properly caused the engine starting logic to lock up, a condition that existed until the UV signa was reset. For i this reason, DGIA did not automatically re-start after it tripped.
l 1
After the trip, operators were dispatched to the engine control panel to i investigate the cause of the trip. According to the operator, several annunciators were lit. The operator briefly reviewed several instrument {
l read-outs and detected no immediate problem. In order to restore emergency i
- power, the operator reset the annunciators without delaying to evaluate or l 4
record the annunciators that were present. During this time, a Shift ;
! Supervisor (SS) and a Plant Equipment Operator (PE0) went to the sequencer !
i panel to determine if any problems were present on the IA sequencer. The SS l pushed the UV reset button, then reset the sequencer by deenergizing and '
) energizing the power supply to the sequencer. This caused the DG air start l solanaAAtmanatsize for another.Lancands.which caused the engine to. start. I This happened 19 minutes after the DG tripped the first time. The engine i i started and the sequencer sequenced the available loads as designed. After l
- 1 minute and 10 seconds, the breaker and the engine tripped a second time. l l It did not automatically re-start due to the starting logic being blocked as i described above. By this time operators, a maintenance foreman and the ;
i diesel generator vendor represe,ntative were in the DG room. The initial '
report was that the jacket water pressure trip was the cause of the trip.
i This report was discounted because the maintenance foreman and vendor
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representative observed that the jacket water pressure at the gauge was abeet 12-13.PSIG. .The trip setpoint is L.PS K and the alare setpejnt is 8 '
) PSIG. Also, the control room observed a lube oil sensor malfunction alars.
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Fifteen minutes after the second DGIA trip, DGIA was started from the engine control panel using the emergency start breakglass button. The engine
! started and loads were manua ly loaded. When the DG is started in the emergency mode, all the trips except four are bypassed. However, all alarms
- will be annunciated. During the emergency run, no trip alams were noticed bX the personnel either at the contro room or at the engine control panel.
! The only alams noted by the controT' room operator assigned forSG' operation -
- were lube oil pressure sensor malfunction and fuel oil evel high/ low alarm, neither of which would have tripped the diesel.
9 j
At 1040 CST, RAT IB was energized to supply power to 4160 volt bus 1BA03.
- DGIA supplied power to 4160 volt bus IAA02 until 1157 CST, at which time bus
] 1AA02 was tied to RAT IB.
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A Site Area Emergency was declared at 0840 CST, due to a loss of all offsite j and onsite AC power for more than 15 minutes. The Emergency Director signed j the notification form used to inform offsite government agencies of the
! emergency at 0848 CST. The shift clerk attempted to initiate offsite notification utilizing the primary ENN in the control room but found it.
l inoperable due to loss of power. The shift clerk then went to the back-up
! ENN and initiated notification after roll call on this system at 0857 CST.
l Due to the loss of power, which rendered the primary Emergency Notification
- Network (ENN) inoperable, and some mis-communication, the initial I notification was not received by all agencies until og35 CST.
i The Emergency Director instructed personnel to complete various tasks for l
restoring containment and RCS integrity. All work was accomplished and maintenance personnel exited containment by 1050 CST.
.i The SAE was downgraded to an Alert Emergency at 0915 CST after restoration l
of core cooling and one train of electrical power. By 1200 CST, plant
- conditions had stabilized with both trains of electrical power being
! supplied from an offsite source (RAT 18). After discussions with the NRC i
and local government agencies, the energency was teminated at 1247 CST and j all agencies were notified by 1256 CST.
D. CAUSE OF EVENT
! Direct Cause:
! 1. The direct cause of the loss of offsite Class IE AC power was the fuel i truck hitting a pole supporting a 230kV line for RAT 1A. This was a
! cognitive personnel error on the part of the truck driver. There were
' no unusual characteristics of the work location that directly
- contributed to this personnel error.
l 2. The direct cause of the loss of onsite Class IE AC power was the failure of the operable DG, DGIA, to start and load the LOSP loads on bus IAA02.
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- 3. The direct cause of the failure of the primary ENN system in the l control room was the loss of electrical power to Unit 1. The pH eary l t.NN in the control room is powered from Unit I Class IE AC xnver.
j - Tharafora .whan. Unit.1. lost Class IE AC electrical power, tie primary :
i ENN in the control room did not work.
2 :
j Root Cause:
- 1. The truck driver met all current site training and qualification l -
requirements, including holding a Class 2 Georgia driver's license.
! However, site safety rules, which require a flagman for backing j vehicles when viewing is impaired, were violated.
! 2. The roct cause for the failure of DGIA has not been conclusively
- l determined. There is no record of the trips that were annunciated '
after the first trip because the annunciators were reset before the
! condition was fully evaluated. Therefore, the cause of the first trip j can only be postulated, but it was most likely the same as that which
- caused the second trip. The second trip occurred at the end of the l' timed sequence of the group 2 block logic. This logic allows the DG to
! achieve operating conditions before the trips become active. The block
- logic timed out and the trip occurred at about 70 seconds. The j annunciators observed at the second trip included jacket water high
!, temperature along with other trips. In conducting an investigation, l the trip conditions that were observed on the second DG trip on 3-20-90 could be duplicated by venting 2 out of 3 jacket water temperature sensors, simulating a tripped condition. The simulation duplicated l i
- both the annunciators and the 70 sec. trip time. The most likely cause
! of the DG trips was intermittent actuation of the jacket water temperature switches.
l l Following the 3-20-90 event, all three jacket water temperature
! switches, which all have a design setpoint of 2000F, were bench tested.
Strttetr*TS4MWwarfoun&Mee>+,setpotat .eL1,SLt-==e E, wtLich
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was approximately 6 degrees below its previous setting. Switch
! TS-19111 was found to have a setpoint of 199 degrees F, which was approximately the same as the original setting. Switch TS-19:12 was found to have a setpoi..'. of 186 degrees F, w!.;ch was approximately 17 l degrees F below the previous setting and was re-adjusted. Switch
! TS-19112 also had a small leak which was judged to be acceptable to support diagnostic engine tests and was reinstalled. The switches were recalibrated with the manufacturer's assistance to ensure a consistent j ,,.
calibration technique.
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. During the subsequent test run of the DG on 3-30-90, one of the '
! switches (TS-19111) tripped and would not reset. This appeared to be :
4 an intermittent failure because it subsequently mechanically reset. :
j This switch and the leaking switch (TS-19112) were replaced with new !
sw.itchas...All_ subsequent testing was conducted with no additional *'
- problems. ;
1 . !
l A test of the jacket water system temperature transient during engine !
a starts was conducted. The purpose of this test was to determine the !
l actual jacket water temperature at the switch locations with the engine j i in a normal standby lineup, and then followed by a series of starts !
- without air rolling the engine to replicate the starts of 3-20-90. The i
- test showed that jacket water temperature at the switch location 4 decreased from a standby temperature of 163 degrees F to approximately i 156 degrees F and remained steady.
! Numerous sensor calibrations (including jacket water temperatures),
l special pneumatic leak testing, and multiple engine starts and runs 1 were performed under various conditions. After the 3-20-90 event, the l control systems of both engines were subjected to a comprehensive test i program. Additionally, the jacket water high temperature switches were i sent to an independent laboratory, which found the switches set at i temperatures ranging from 162 degrees F to 195 degrees F rather than the 200 degree F setting that was required. The calibration technique j was changed and switches were re-calibrated and installed on DGIB on ;
! 5-23-90. However, another failure occurred on DGIB (See Technical
! Specification Special Report 1-90-4.). These switches were also sent
! to the independent laboratory, which found the settings to be from 164
! degrees F to 169 degrees F. Subsequent to this testing, the onsite ca ibration procedure was again revised to provide a technique that is consistent with the actual operating conditions that the switches
- experience. Switches were calibrated using this new technique, l installed and found to operate within the expected parameters. Since the event of 3-20-90 through 6-7-90, DGIB had received 12 valid tests wttir thr one4rtierva menttonodi a% am600.E had. encatund .1 A valid _ _
tests with no failures.
Based on the above facts, it is concluded that the jacket water high i temperature switchas were the most probable cause of both trips on
, 3-20-90.
! The investigation and testing following the 3-20-90 event revealed that pressure sensors in the diesel generator lube oil system had not been replaced in accordance with a 10 CFR 21 notification from the manuf5cturer datet 5-12-88. The 10 CFR 21 notification was sensusing -
i relative to the requirements for their replacement. It was
- subsequently revised in an addendum dated 6-8-90. The pressure trip l ans som anon enam
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! instructions. GPC does not believe that these sensors contributed to j the diesel generator trip on 3-20-90.
I E. ANALYSIS OF EVENT The loss of offsite power to Class IE bus IBA03 and the failure of DGIA to start and operate successfully, coupled with DGlB and RAT IB being out of !
service for maintenance, resulted in Unit I being without AC power to both l
Class IE busses. With both Class IE busses deenergized, the RHR System j could not perform its required safety function. Based on a noted rate of ,
j rise in the RCS temperature of 46 degrees F in 36 minutues, the RCS water l
- would not have been expected to begin boiling untti approximately I hour and 1 l 36 minutes after the beginning of the event. Using more conservative j
! assumptions and methods, but the same actual time of the event, the , ,
j calculated worst case time to boiling was found to be approximately I hour i
and 11 minutes, and time to core uncovering was found to be approximately 11 4 hours and 5 minutes. This assumed no gravity feed from the RWST.
Restoration of RHR and closure of the containment equipment hatch were
, completed well within the estimated I hour and 36 minutes for the projected onset of boiling in the RCS. A review of information obtained from the Process and Effluent Radiation Monitoring System (PERMS) and grab sample
! analysis indicated all normal values. As a result of this event, no
! increase in radioactive releases to either the containment or the l environment occurred.
)
l Additional systems were either available or could have been made available l
to ensure the continued safe operation of the plant:
3
service approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> into the event.
27 Offetter pouemoveMobhr 4mes>4c'M *W tha.,gsparator step-up transfomers which were being used to "back-feed" the Unit
! and supply the non-lE busses. Provided
' Auxiliary that the phase to ground fau was cleared, Class IE busses IAA02 and Transfomers (UAT)lt j
IBA03 could have been powered by feeding through non-lE bus INA01.
i
! 3. The Refueling Water Storage Tank could have been used to manually l
establish gravity feed to the RCS to maintain a supply of cooling water l
to the reactor.
l Consequently, neither plant safety nor the health and safety of tha public .
- was adversely affected by this event.
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- F. CORRECTIVE ACTIONS i
j 1. A management policy on control and operation of vehicles has been established.
- 2. Jeeporary barricades have been erected with signs which direct i authorization for control of switchyard traffic to the SS.
- 3. The Loss of Offsite Power (LOSP) diesel start and trip logic has been modified on both Unit I and Unit 2 so that an automatic ' emergency" l i start will occur upon LOSP. Therefore, non-essential diesel engine 4 trips are blocked upon LOSP. Additionally, high jacket water i temperature has been deleted as a trip signal in the emergency start f
mode, l l
i
! 4. The DGIA test frequency was increased to three times per week until '
- 4-20-90 when the test frequency was changed to once every 7 days in i t
accordance with Technical Specification Table 4.8-1. This frequency ;
1 will be continued untti 7 consecutive valid tests are completed with no !
- more than one valid failure in the last 20 valid tests. Up to and including the two valid failures of the 3-20-90 event, there were a total of four valid failures in alid tests of DGIA.
? 4
- 5. The jacket water temperature sw1;c es for each DG were replaced or i re-calibrated using a more appropriate technique prior to their installation.
1 i 6. A back-up ENN system powered from the AT&T system, which previously
! existed and was operational for South Carolina agencies, has been
{ extended to include Georgia local 7,ad st:ste agencies. Instructions have
- been given to Emergency Directors and Communicators concerning use of l the emergency communication syst,tes.
i l j G. ADDITIONAL INFORMATION 1
- 1. Failed Components:
Jacket Water High Temperature Switches manufactured by California
! Controls Company.
! Model #A-3500-W3 i
l
- 2. Previous Similar Events:
l i None
- M M:.:
194 m 8 y of h k. ~
anse.-anonnam , ,.
ane ,.am men. . . u - --aseuuueer --
__ m g 8 W9RGE M UCENSEE EVENT CEPORT (LER) J!2.",*, ".,'"a"m*u'S%".'a"e*e"6m"d' *"4 M l
TEXT CONTINUATION M M T !S!"*,".aE'EI. Te I
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VEGP - UNIT 1 o islololel 4l2l4 9l 0 -
0l0 5 -
01 1 Of 9 or 0l9 i i wr . mme w wim E Energy Industry Identif5satton System Code: ., ,j 1
. Reactor Coolant System - A8 '
1 Residual Heat Removal System - BP l l Diesal Generator Lube 011 System - LA Diesel Generator Starting Air System - LC Diesel Generator Cooling Water System - LB <
Diesel Generator Power Supply System - EK l Safety Injection System - 80 13.8 kV Power System - EA 4160 volt non-lE power system - EA ,
4160 volt Class IE power system - EB Chemical and Volume Control System - CB !
Containment Building - NH i 480 volt Class IE Power System - ED Engineered Safety Features Actuation System - JE Radiation Monitoring System - IL N-r
- w o-$ l 5 c .: , ;. .
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~ - - - - - - _ - . . . _ , -
.u .x 'ge 2.4 sg ..tpisgtOPCO-VOGTLE TEL NO:1-205-877-7985 ..EZii P22 I
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southem CompanyServices
! July 31, 1990
, .. . ~,. ,
3 Vaatle Electr' c Generatina p ant - Un i 1 =d i . a. ;
l Fire Event safe Shutdown Eva'untion of Diesel Generator Trips - + e - v'"
- File:~REA'VG'0047 security Code:
NC Log: 5G-9510 ~ .
{, ..c . . . ..
Y
! Mr. C. C. Miller -
l Vogt er of E ManafeProjec1inoering w Nuclear Operations .$
M' ~
- k. V. '
Georgia Power Company post Office Sox.Itg5 A .
\' s 81rmingham, Alabama 35201 l .
Dear Mr. Miller:
1 We have reviewed the potential emergency diesel generator ground fault trip j described in our letter $G-9471 for requirements to notify the NRC per 10 CFR
$0.72. As described in FSAR Parapraph 9.5.1.1.3 the !tre protection program i provides assurance that a fire wi . , . .
shutdown systems'with or without offsite power. A fire in; fire ireest '
1 not cause the loss of function;ef saf j
1/2 CS-LC-B was analysed and protective asasures were included ta: '
en 4
~
as described in FSAR Appendix 9A. Thus s' fin inothis even with' sin' 4"L within the design bases of the plant.
l i i , f!M@9f.p/ne.Jes..M.g, M 3 . g., i gl%
Due to.a fire in these areas, A train cables may be damaged and: A train .
is assumed to be lost, however the B train cables are protected effects of the fire and no damage to B train equipment and. cable $ts s regulfed : the'T . .
for safe shutdown would occur. Thus while an unentlysed ifes1t shich?;r W;. .
could separate the diesel generator from the 5 train sa 'related tes1st tn .; A ..1
. occur. t w equipment required to achieve and atletata; shutdeum $r ,
'g ndamaged'end the plant configuratien~would .be s.ie(14ffteht;stgey k f , ,,, ,
., . ,,n .
_ . ....., m, w ,
We"believe that the corrective actions to tselate the greind' fault and *a. . . . ..
re estabitsk power to the g train are straight forward and.readily - ~
accomplished within the time frees previously analyzed fer e statten '
blackout. Thus adeguate time is availeble to provide sewer to the safety related equipment required to shut down the the design bases of the plant is maintained plant and the casehtlity to meet p In conclusion, were this to occur, it could represent an unanalysed condition, but would not be a significant compresise of plant safety, and ->
- k. -
' l therefore is not reportable per the requirements of 10 CFR 50.72. I Jw%%3 K Q3 T
,.., e m ; . 1 N-wjx*g.,.K4 4
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n294 P03 s l TEL NO: A-205-07h*7985 4 1 FUL-31 '90 34:9A ID 50tOPCO-00GTLE File: REA VG 0047 Mr. C. C. M111er Log : $G 9510 July 31, 1990 i Y*9* I :=:
If you have any questions, contact Andy Wehrenberg at extension 6768.
l Very truly yours.
{ A W. C. Ramsey, Jr. 1 I
WCRJr/cm !
XCI Becital Power Corocration I. Stetrzyk l
- cre1a 7 :r r - anv P. D. Aussten N0lM5 l
Eaut.S ;; r-nv tarwicas. Inc.
C. R. Myer !
R. E. Petrick J. A. Wehrenberg ,
Document File l l
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O . fNT OF THE 18 0/G. NEUTRAL GkCud OvERCURRE#L ELAY.Wuc s d e.n .
A FIRE IN ZONE 80, CAUSING DAMAGE TO THE FEE 0ER CABLF TO INB10000VLOJ .w;2.J, 4.ih .
j dm. i RESULT IN THE D/G TRIPPING BEFORE 1hB10 TRIPS. IF A LOSO.QCCOR$ ,. c.y .,/ g" ,.7., d ;
- J n. ON UNIT T.MyERIFLTHERE IS NO F1RE. ALAPM IN ZONE 80. IF M JRE Mf'lALAM:!N'ZONET 80 EXIST 5',
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j i DUE TO THE SETPOINT OF THE UNIT 1 & UNIT 2 'D" D/G NEUTRAL GROUhD OVERCURRENT . <
! RELAY. A FIRE IN 70NE 80 CAUSING DAMAGE TO THE FEEDER CA8tE TO N810 COULO RESULT IN THE j "B" D/G ON THE AFFECTED UNlT TRlFPING BEFORE NB10 TRIPS.
IF A LOSP OCCURS ON EITHER UNIT, VERIFY THE:f 15 N^ TIRE ALARM IN THE AFFECTED 3 UNIT'S ZONE 80. IF A FIRE ALARM IN ZONE 80 F7ISTS, IPHEDIATELY TRIP NB10
~
ON THE AFFICTED UNIT. THIS PROBLEM EXISTF ON BOTH UNIT'S "B" D/G AND WILL , ,
BE CORRECTED IN THE NEAR FUTURE BY DELETING THE NEUTRAL GROUND DVERCURRENT ~ ., . . i D/G TRIP. ' * ' " ' ~
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Performance PayyBlan -
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Survey 1990 l . .
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Je,y : 74;> is a .d~faJ >~~r i
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3 l l An.>weo doC5plcdio's SS. SG . 58 n on rvadwy # s., si, 54j l_
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- VARTLE PPP NEVEY ffEMEK i
pi M T V0ETLE i
1 v Sif SlestARY TAR:
None of the overall dimensions wcra pos . ,c. *we.ar, all either 4
i improved er remained the seau eince 1989
.y M11TIVE FIEIM51 i
THE 1990 PAY PLAN (PPP):
1 1. Has been explain 6d to me in scch a way ttat wt<arstand it.
( !
! WITH RERAM TO 80AL SETTING IN THE PPPs
- 4. I understand the corporate goals used in the PPP.
- 21. Accomplishing my individual goals contributes to the achievement of eer organtaational goals.
J JN Senn 4Ls
- 61. Increasing competition represents a tajor threat to our company.
- 64. I would like to continue working under stee type of performance
' pay plan.
v i MSATIVE PImiMit i
s
- THE1990PERF0mANCEPAYPLAN(PPP)
- 1. Is att having positive effects on the cuegany's perfornance.
! 3. Is having a positive effect on my work habits and perforsance.
- 4. offer enough money in the incentive portion to make it
- 5. Will Bat appropriately compensate me for my contributions, s l i Vogthr het t good payout. Unr <fft tive peopte- feet' tttfr way?
Old the large numbers of people rated in the top two blocks have en affect on this? They might have believed they were going to receive a 15-205 incentive Increase. In n ted performance ratings a l apparently hurt.
1 i i
I v
- Donetes items of htsh interest 20 Jul 199G
. 4 . ,
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4
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- WCATLE PPP EnftY FEIBMCK l
! i v
WITN RtAAM To eML stTTING IN TM PPP
! 7. I don't see the connection between to '.'P.1ctial vals and the 1 corporate goals.
3
- 8. Itr manager has nat asked for my input in understanding the results i
ed last year's survey.
i g. ItyPPPfacentivepaymentwasnatbased:reyjobperformance.
i
- 11. The PPP goals of my organization are more difficult than the goals
- of other organizations.
- 13. It is agg possible to achieve our PPP organizational goals at the top leve J. (5' Tevel 'of performance) j 14. Werk groups inside my organization do D21 cooperate in accomplishin our o
- 24. fly individua goals 1anizational goals. eave out isportant aspects of my job l 1 difficult to asasure.
l 25. As a result of having individsal goals. my co workers are less
- willing to help each other.
- 26. To reach our 1989 organizational goals, people made decisions
- which will hurt my organization in the long run.
- 27. Accomplishing my individual goals is beyond my control.
l
- 28. As circumstances change, I will 281 be able to change my i individual goals to make them more appropriate.
- HV 1990 IISIV10uAL 80ALS 00 Eli REQUIRE Mt To
- 29. Se more flexible in deaWng with my customers, supervisor, and l v co-verkers. '
- 34. Take intelligent risks.
j l
NY llAllAGER/90PERVISOR IS EU, EFFECTIVE AT:
1 35. Treating all employees in sur group f airl.r ud consistently.
- P- 4trating a strans belief that ghgngg jging_on in the comoan,y
@ Apg enemanary, j ~. ---_ . Z :; " - -dine people based on nark' -?
~
- . Encesreging and environgs communications.
- 4. .;,; , z --:; ,..i and tonest n..s.4 = h:" ora av job.
- 41. Melping me, in a positive way, to understand how t e PPP works.
- 42. Setting sy work group to function as a team.
1 43. C- 7 2--the 6 W espee6s of men.
- 44. Cennunicating about organizational goals and my role in helping to l achieve them.
- 45. Giving me the authority to make decisions ! need to make.
- 46. Taking the right action with poor perfomers. l v
- Genotes itene of high interest 20-Jul-1990 5
I t
1 8 168 ??? R E E 1 1.E n nin In seenAL:
- 47. I am 301 paid fairly compared itt ut% Moyees ir my company who do staller work.
- 48. Employees are nat rewarded for icquirtr ; .w skills.
Are people told in perforinarcs feedbad z suiten ibet they have sees developmental needs and they doJ1J inrme in certain areas.
Then, either management doesn't provide deseloreental opportunities or reward the acquisition d new skin s?
- Writ *= informatian an the PPP nas out; I,Qd a to understand tne an. w
- eyees are afraid to voice ac opintor uragement does 31 wan h-
- 52. There are nE ways for employees to foIm7[p~articipate in selvfee prTo ens for the company.
- . people in sy orsanization do nel trust end other,
- 5. Unnecessary change is occurring althin oar company.
-t is not n"-ertine.the changtaoing..gh ficers of sty company are ngi aware f the problems at p
@ {he E7. h [ gehg ggggggtjens for taeravament..l_as lancred.
@ There ys m ,,___ up and down consnunicat1ons throughout the
' St. Ities'aggiventhetrainingIneededtomakethePPPwork.
- 60. I was sat giving training in how to set effective goals.
St. N4negers d'd agt use last year's survey data to improve the
= at'en of the neeen1Zation. -
~
ce an opinliin" thit'eT'anageM6Tdoesn't want Eo7
[* h I'an afraid to vo' -- . - . . ,
- 66. I de att understand how y base salary 'r:rease was determined.
ElIIM CfMEE1r N
people liked being judged on merit and the money.
ISRATIVE:
They didn't like the lack of fairness and favoritism of the allocations. They felt the goal quality was poor, too trivial, toe easy to achieve, and too sub.jective. .
v i
1
' I
- Ossetes items of high interest 20 Jul 1990 6 l
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Deficiency Card I-90-003 -
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On January 5,1990, th- # . .M t o < e v ' e w t he re port ald t.i ty,ofdL ...J'i;,.l1 Deficiency Card 1-90-00.1. a. . . ,
violation of a licence c m , . J c.f r ov irncheo 1' y "'e ea xrctentiaL..iimit ter oower, of 3411 MW, The board V - .
Vil :. c. cision that the iterr wasnotrepbr. table.
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OPERATION OUTSIDE THE DESIGN BASIS FIRE PLUS LOSP DEFEATS SAFE SHUTDOWN DESIGN CAPABILITY On 7-19-90 two def1clency cards (DC's 1-90-299 and 2-90-080) were written by a plant electrical engineering supervisor.
The DC's identified that 1E electrical train cable separation problems ex1sted in the 1NB10 switchgear room for Unit 1 and in the 2NB10 switchgear room for Unit 2. The DC's stated that a fire in Zone 80 could cause Diesel Generator 18 or 2B +^ trip thru the neutral over current relay and Safety Related A train 480 V and 120 V circuits would be .
lost due to fire damage because cables affecting both A '
train and B train were located in the same fire zone.The train A cables include " Safe Shutdown" cables.
This condition (of having a design where a single fire event ;
plus loss of offs 1te power would result in the loss of Safe ;
Shutdown functions) is a condit1on outside the design basis 1 of the plant. It v1olates the general des 1gn criteria of i 10CFR50 Append 1x A criterlon 3, 17,21 and the requirments of l
Branch Technical Pos1 tion CMEB 9.5.1 of NUREG 800.In addition the operability of effected equipment so designed is also in question. Operation outside the safety design basis essentially places the plants 1n " motherhood".
To the operator the observed condition would be extremely confusing.In the event of a fire in zone 80, for example, the safe shutdown procedures direct the operator to use train B.The operator would find the B train diesel tripped with a ground fault that coulc not be cleared.A train would be unavailable due to fire damage.
Fire areas were evaluated in the Vogtle FSAR Fire Hazzards Review and are described in the Vogtle FSAR sections 9A.2.29,9A.1.40,9A.1.50,and 9A.1.47 (see chart below )These sections are incorrect.
They fail to consider the design dificiency described above and provide guidance for " safe shutdown" to use a train that would be faulted, tripped,and unavailable. ,
Despite the above conditions being reported to the Shift Supervisor thru the DC's the Shift Supervisor did not make the required prompt not1fication to the NRC within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> !
under 10CFR50. 72(b)( 1 )( 11 )( A )or(B) .This condition would also I' De considered reportable under10CFR50.72( b)(2 )( iii )( A ) which requires a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report.
Operations did in1tlato a " Standing Order" (1-90-11) on 7-19-90 that describes the problem as being only on Unit 1.The standing order is incomplete and does not address all the ;
fire zones (only Zone 80) where the electrical separation j design is deficient ( see chart i, j It relies on operator action and may not mitigate the def1ctency.
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On 7-30-90 a secono stancing order was :ssued correcting tne first one. stating tnat notn uni- 1 anc unit 2 were affected.It still only accresses Fire Zone 80.
By 7-27-30 the Vogtle NSAC supervisor had determined that the event was a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> reportaoie and the DC's had been elevateo to SONOPCO licensing ano discussions up.to and including the Vice Pres 1 cent had been held cut actions to place the Plants in a safe condition and to notify the NRC with1n 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> were not initiated.
On the morning of 7-30-90 the Vogtle General. Manager was d1scussing the reportab111ty of tnis event with the Technical Support Manager and the Ooerations Manager.He had been told that SONCPCO engineers thought it was outside the design basis and th1s angered him.He stated that "Whatever eng1neer thinks 1t 1s outs 1de the des 1gn bas 1s should wr1te a DC and sign h1s name on it and show exactly what criteria 1s violated.'.This seemed very strange since a DC had already been written 11 days Defore ident1fing the def1clent condition. It was the General Manager's and h1s operations staff's job to promptly evaluate the existing DC and make appropriate reportability calls,not to atempt to reset tne time clock with another DC or make intimicating statements about tne eng1neers.
By 7-30-90 the SONOPCO General Manager Plant Support,the Engineering and Licensing Manager, and the Licensing Manager had all concluded that it was a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> reportable but still no report was made .It was said that Southern design engineering was still looking at 1t.
On 7-31-90 the SOPNOPCO Eng1neering and Licensing Manager informed the Plant that SONOPCO and SCS had decided that this condition was not outside the design basis and was not reportable as a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />,a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and not even a LER,1t was merly a deficiency that should be corrected.
FSAR CHAPTER 9A FSAR FIRE EVENT SAFE SHUTDCWN EVALUATICN ZGNE AFFECTEC' TRAIN' says to use 1-80 8 8 train 1-60 B A or B train 2-80 'O O train 2-83 B A or 8 train
- -74 a- A me a tra'n
Since tne :ord1*.1on was initia11< discoverea there has been tne opoortunity to actuall. '14 the ceficiency.The fix is simple clifting 2 leads 1 ar.d can De acne at power.In addition Unit I had tripped off line on 7-23-90 to 7-26-90 ano the work could De done as cart of tne forced outage.
Management has not pushed,to lift tne leads and the fact that the unsafe condition was not corrected immed1ately is apalling but also confusing.It would appear that the deemphasis is des 1gned to support the decision to not report.
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e massea==a 30 mammypnamanat E 2.2sd PETITrest, amoffoAf 111.2
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I. petitianers' A11aantiana.
The petitioners assert that GPC, through the soNOPCC Project, submitted known, false statements to the NRC intended to mislead the NRC about the reliability of the Vogtle emergency diesel generators. As basis, the petitioners allege that on April 10. 1L990 Mr. M
" diesel air quality"ysbaugh informed statements made inthe General Manager a Confirmation that the of Action response letter were falso and that on April 19, 1990 Georgia Power submitted a Licensee Event Report ("LER*) after Mr. Mosbaugh advised the senior vice President, a corporate officer located in Birmingham, that the information contained in the LER wa's incorrect. The petitioners further allege that the Company " intentionally delayed" revising the LER until after a June 8, 1990 presentation to the Commissioner, drafted multiple transmittal letters for the revised LER which contained " false explanations" in an attempt to " cover up" errors in the original LER, and retaliated against Mr. Mosbaugh for identifying the alleged false information submitted to the NRC.
II. cPC Rannonsa to Petitioners 8 A11aantions.
A. Diesel Generator Pneumatic System Air Quality.
The petitioners' allegation that the cocirmation.of Action
( response letter, dated April 9,1990 (the ' . CAR"), was false and the implication that Mr. Mosbaugh informed the General Manager of' the inaccurate statements in the letter prior to its transmittal First, the COAR was dated April 9, to thg 1990. NRC are without merit.Mr. Mosbaugh's memorandum to the Vogtle Gene which addresses " air quality" (Exhibit 1) is dated April 10, i
j I Diesel air quality refers to the dryness of the air in the !
l pneumatic control system of the diesel engines.
t 2The COAR states, in part, the following:
I In addition, the following actions have been or are being l implemented to ensure a high state of diesel reliability . . . . GPC has reviewed air quality of the D/G air system including dew point control and has concluded that air quality is satisfactory. Initial reports of higher l
' than expected dew points were later attributed to faulty l instrumentation. This was co.. firmed by internal inspection i
' of one air receiver on April 6, 1990, the periodic replacement of the control air filters last done in March,
' 1990 which showed no indication of corrosion [,) and daily air receiver blowdowns with no significant water discharge.
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i se 1990. Thus, Mr. Mosbaugh's alleged notification to the General 6
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Manager that the " diesel air quality statements made in the letter were false" would have occurred after the letter was transmitted to the NRC. Second, the memorandum of Mr. Mosbaugh i
j (Exhibit 1) does not mention the COAR nor state that the air quality at that time was not satisfactory. Rather, the J memorandum identifies three types of historic problems.
l Third, the basis of Mr. Mosbaugh's memo is believed to be a
] document, consisting of five pages (Exhibit 2), also dated This memorandum also discusses historic
{ April 10, 1990. maintenance of the diesel air dryers and suggests that dew point j
measurement practices needed to be investigated to ensure reliable results. The memorandum does not, however, conclude i
i that the then current air quality was deficient.
I Fourth, the COAR acknowledges initial concern associated j
with air quality (i.e., " initial reports of higher than expected dew points") and deficient measurement ofMr. dewMosbaugh, point (i.e.,
it i
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" attributed to faulty instrumentation").
- appears, was focused on historic air quality issues based on maintenance history, and was unaware of GPC efforts relative to l better instrumentation and measurement. These efforts included l obtaining instrumentation from another plant.
)
Fifth, the COAR lists some of the activities which form the basis for the conclusion that the air quality.was satisfactory
! ( (April 6, 1990 internal inspection, replacement of control air
?
' filters and daily air receiver blowdowns). In addition, Mr. Mosbaugh apparently was unaware of other technical l considerations, including the views of knowledgeable engineers j
that the air quality of the pneumatic system was satisfactory.
t Finally, the NRC Staff, thought to include Mr. Pete Taylor, l reviewed the issue of the possible contribution of moisture in i
the diesel engine's pneumatic control system to the March 20, 1990 event. The NRC Staff, then, is more aware of the i
verification of adequate diesel engine air quality based on personal review than Mr. Mosbaugh, who bases his conclusion on dated information.
j l
B. Diesel Generator Start Information. )
Petitioners' allegation that the April 9, 1990 COAR and a Licensee Event Report ("LER") (90-006) dated April 19, 1990 j
contained known falso statements intended to mislead the NRC about the reliability of the VEGP diesel generators is without l merit.
i The C0AR states, in part, that the "A" Unit 1 diesel "B" diesel generator generator had been started 18 times, and the i
had buen started 19 times and that no failures or problems had l
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' occurred during any of these starts. The LER refers to both
! [ diesel generators as "having been started at least 18 times each and no failures or problems have occurred during any of these starts." As can be confirmed in statements in the custody of the l 8 l
i NRC's Operational Safety Inspection team which reviewed this i matter in August, 1990, unit control logs and shift supervisor logs were the source of the data used in developing the numbers i
"18" and "19" found in the COAR and the original LER. The numbers originally were included in a transparency developed by Vogtle plant personnel; this transparency was included in handouts at an April 9, 1990 meeting wita the NRC in Atlanta, The COAR, written the same day as the meeting with NRC i Georgia.
representatives in Atlanta, adopted the "18" and "19" numbers.
l The LER, written later, also was predicated upon the "18" and 4
"19" start count.
statements in the custody of the Operational l
Safety Inspection team confirm that both documents basically used the information developed for the April 9 transparency.
In addition, successive draft revisions of the LER have been reviewed by GPC. A vereion of the LER prepared by the site, dated April 17, 1990, identified "several starts" rather than An l
i specifying a number of starts (Exhibit 3, p. 6 of draft LER).
attendee at the Plant Review Board stated that a specific number should be used and the next draft version of the LER stated that i'
the number was "more than 20 times" (Exhibit 4, p. 6). The "more than 20 times" phraseology was provided to corporate I representatives in Birmingham (Exhibit 5). These representatives, with knowledge of the 18 and 19 numbers used in the trknaparency on April 9, questioned the "more than 20 times i each" language provided by the site. More specifically, Mr. Hairston, the Senior Vice President, requested the corporate LER coordinator to " verify >20 starts." Retained copies of the I
LER drafts confirm other efforts by corporate representatives to verify other information and assure the accuracy of the LER
, (Exhibit 6).
4 Additional diesel generator starts had occurred subsequent i to April 9, 1990 (the date of the GPC meeting in Atlanta with NRC representatives), and the final April 19th LER wording stated
! that egchCPC diesel wasengine aware had beeninspectors that NRC started "athad least 18 times followed the each."
Company's efforts to troubleshoot and test the operability of the diesel generators and believed that the NRC had all relevant information on the diesel generators' operability and reliability. Nevertheless, 'either before or concurrent with the 3 The wording was reviewed by corporate and site representatives in a telephone conference call late on April 19, 1990. Although Mr. Hairston was not a participant in that call, he had every reason to believe the final draft LER presented to l
him after the call was accurate and complete.
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transmittal of the LER to the NRC, the Senior Vice President
[ instructed Mr. C. Ken McCcy, the Vice President for Vogtle, to l
- e call the NRC's Mr. Ken Brockman and to discuss the fact that the j number of starts indicated inA the phone LER calldiffered from Mr. from McCoy the number on was placed the April 9 transparency.
i to Mr. Brockman on April 19,' in the afternoon.
Mr. Mosbaugh and employees who reported to him controlled
- e the development of the original LER. To the extent Mr. Mosbaugh J actually had concerns about the substance of this document, he had direct and immediate ability to change the information contained in it. His own actions relative to the LER establish Indeed, as reflected in the PRB comment review sheet 1
! this fact.
for its meeting No. 90-59, held on April 18, 1990 (Exhibit 4),
1 Mr. Mosbaugh directed three changes to the draftMe, LER, two of therefore, which he directed as " mandatory" word changes.
l had an opportunity to require any other correction. similarly, on April 19, 1990 in a telephone conversation between the site
]'
representatives and Corporate office representatives, he had the opportunity to suggest corrective language but, apparently, failed to do so. Not until April 30, 1990 does it appear that i
j Mr. Mosbaugh articulated for the benefit of his management that the diesel engine start count data contained in the LER was i
inaccurate. At that time, he was. assigned, in writing, to l-correct the NRC documentation (Exhibit 7). He, therefore, was j
tasked with correcting the inaccuracy which his Technical Support group had created by supplying "more than 20 times" wording to j
( the Corporate office.
In September or October, 1990, in the presence of Mr. Brian Bonser, an NRC Resident Inspector, Mr. John Aufdenkampe (the l former Technical Support Manager under Mr. Mosbaugh responsible
- 8 for LER development in April,1990), stated his opinion that the LER used the numbers in the transparencies developed for the April 9, 1990 meeting with the NRC in Atlanta and that his group l
i had merely added additional starts from and after April 9 to reach the conclusion that "more than 20" successful starts had occurred during the relevant time frame. '
In addition to directing changes in documents as required, l nn April 30, 1990 the General Manager also verbally notified the l
NRC Resident Inspector of the erroneous data, as he testified to l l
- the operational Safety Inspection team. Further, Mr. Nairston called Mr. Stewart Ebneter, the NRC Regional Administrator, on May 14 and May 24, 1990. He believes that in the longer call on 2
May 24 he informed Mr. Ebnc'.r that the count c' successful starts in the LER was in error. He further recalls that he i ,
i conveyed the then-current " correct" numbers at that time to Mr. l l
l Ebneter and informed him that revisions to the LER would be i forthcoming. Mr. McCoy recalls calling the NRC's Mr. Ken l l
Brockman about the same time and informing him of the error, telephone billing reports reflect several telephone calls from
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4 Mr. McCoy to Mr. Brockman on May 24. The petitioners' allegation ,
[
i that an intentional delay in revising the LER until after a
) June 8, 1990 NRC meeting is founded, then, on the false premise j that the revised LER was the mechanism by which the NRC first I j learned of the inaccuracy of the LER. Such was not the case.
i 1 On or about May 9, 1990, Mr. Mosbaugh provided a revised '
i draft of the LER language which addressed diesel generator i starts. The revised language proposed by Mr. Mosbaugh (Exhibit i 8) conveys the same substantive message as the language in the i
April 9th COAR and the original April 19th LER. All three state l
that each engine had been started successfully, and none
- indicated failures or problems indicative of an unreliable diesel j engine. Mr. Mosbaugh's proposed revision, in pertinent part, j states " including the under-voltage test each engine has been.
- successfully started eleven times with no start failures." If,
{ as he now alleges, Mr. Mosbaugh truly had concerns related to the
- original LER, his inaction on April 18 (at the PRB), in the
! April 19 telephone conference, and his April 30 assignment from j his General Manager to provide revised LER language provided him
- with numerous opportunities to direct revision or to revise the i alleged "falso statements." This he failed to do.
i 4 The allegation that GPC officers would attempt to mislead l
the NRC with incorrect information is, in a word, absurd. As l Appendix B to NUREG-1410 indicates, from March 26, 1990 through 1
f April 17, 1990 numerous interviews and meetings were held
- \ concerning the event at Vogtle, including transcribed diesel i generator meetings between the NRC and GPC. The Incident i Investigation Team (IIT) reviewed voluminous plant records, including records associated with historic diesel generator
- operations and maintenance. Numerous informal discussions concerning diesel operability also occurred, including l discussions concerning operability of the diesel generators
, between the General Manager and the NRC's Mr. Allen Chaffee.
- Extensive telephone discussions were also held between NRC and
! GPC after the March 20 event, including 25 calls to IIT
[ representatives in Bethesda between April 6 and May 11, 1990 i concerning the diesel generators' sensors. Many of these calls
- lasted for over an hour and typically involved several IIT team i members. Given the widespread and extensive discussions between i GPC and NRC representatives at various functional levels, the suggestion that GPC officers or upper level managers, who were aware of these efforts, would knowingly provide false information is ludicrous. The converse is the truth; their staffs were j tasked to verify the information provided from the site group which was under Mr. Mosbaugh's direction and control. And, when
- it became apparent that information provided to the NRC was l inaccurate, various GPC representatives informed the NRC of the
- fact. ;
i l
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Finally, the Petition f ails to point out 1990 asthat Mr. Mosbaugh a consequence of the
[ was removed from the PR8 on May 11, I
permanent Assistant General Manager - Support reassuming his position after completion of SRO training, not as a consequence j
of identifying an inace' racy in the LER or COAR.
l C. The Revised LER.
i
! t A revision to the original LER to correct the diesel generator start count was contemplated as early as April 30, 9
1990, as reflected by the General Manager's meno of that date i
(Exhibit 7). Due to the several sources of inaccuracy, as j identified in GPC's August 30, 1990 letter to the NRC, a 1
j consensus on the " correct" count was not reached for some time.
In addition, examination and testing of diesel engine sensors was being pursued (representatives of the IIT readily can verify the extensive, almost daily discussions with GPC representatives i
i concerning these efforts). A draft of Revision 1 of LER 1-90-6 1990 was approved by the PRB in PRB meeting Mo. 90-66 on May 8, Mr. Mosbaugh, and by the Vogtle General Manager on May 14, 1990.
i as Acting PRB Chairman, signed the approval sheet on this l
revision draft on or about May 8. Exhibit 9 reflects this approval, as well as the Technical Support Manager's May 4th i
l approval of this draft revision, which was telecopied to the Corporate Office on May 14, 1990.
A comparison of this site-approved draft revision and the f( prior draft prepared just a few days before (Exhibit 10) reveals changes in diesel generator starts. As stated previously, on May 24 the Senior Vice President called Mr. Stewart Ebneter, the Region II Administrator. The Senior Vice President recalls that he supplied the Regional Administrator the then-current " correct" numbers which were "14" and "15." This recollection He also is confirmed recalls 1990 draft of Revision 1.
by the May 14, informing the Regional Administrator that two revisions to the LER were then contemplated, one to correct the diesel generator start data and the other to document the results of the sensor test program. 9 The draft revised LER was further modified over time.
Exhibit 11 is a June 11, 1990 corporate edition of the Exhibit 12,revision a site draft which reflects "15" and "14" starts.
version of the revision updated to include starts through June 11, shows "14" and "11." The Senior Vice President noted about this time that the diesel generator start count data was differoat than previous data. Irritated at the data variation and withvut a satisfactory explanation of why the data was different, the Senior Vice President tasked the Safety Analysis and Engineering Review ("SAER") quality assurance group at the plant with the verification of the " correct" numbers for the LER revision.
6
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j As of June 11, 1990 the then-current draft revision of the l j [ LER identified "14 valid tests of DG1 (sic) with no valid i failures" and "11 valid tests of DG1B with one valid failure" (Exhibit 12). On June 14 the Senior Vice President called the Regional Administrator again. The Senior Vice President informed i
the Regional Administrator that the " count" data had changed once l again and was different than the information previously provided to the Regional Administrator on May 24. He also informed hia j l that the SAER group had been assigned to conduct an audit on the :
f numbers. The conversation reflected upper management's l l
commitment to obtain and supply accurate information. The Senior l l
Vice President also instructed that the NRC's Mr. Brockman or Mr. Hardt be contacted and informed about the change in " count" i data; Mr. William Shipman, the General Manager - Supporf did so, l either on June 14 or June 15, based upon telephone billing ;
- reports. <
4 l
By June 15, 1990 information related to the testing of the l jacket water temperature sensors had been sufficiently developed j for inclusion in the revised LER (Exhibit 13). )
1 I i By June 23, 1990, the Manager of Technical Support, the PRB,
{
and the General Manager at the plant had approved a draft
' Revision 1 to LER 1-90-6 (Exhibit 14). Concurrently, the SAER I group was conducting its comprehensive review of available diesel j
generator start data. As of June 28, 1990, the SAER group had reviewed diesel generator start data available and prepared a i ( number of spread sheets comparing various data sources. These
! spread sheets eventually were attached to the group's report (Exhibit 15). Again, this report was developed at the request of 4
the Senior Vice President, who instructed that a copy of the report be provided to the Resident Inspector at Vogtle.
As demonstrated by the foregoing, the delay in submittal of l
a revised LER was principally a function of assuring an accurate document and providing additional information concerning the temperature sensor testing. Numerous informal notifications to l
i j the NRC preceded the formal revision as well as the IIT i
presentation to the Commission. The various draft' revisions j unequivocally demonstrate an on-going evolution of LER draft 4
revisions and the significant variation which led, appropriately, to the Senior Vice President's request for independent verification by the SAER group prior to submission of the revised LER.
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i f D. Transmittal Letter for Revision 1 of LDt 1-90-006.
[ r i
l Mr. Mairston instructed his staff to prepare a transmittal letter to the NRC for the revised LER which explained the j
differences in the count numbers between the original and revised LIms. The transmittal letter informs even the most casual reader j
that the revision was necessary "to correct the LER" and borrows i
i heavily from the SAER report (Exhibit 15). The third sentence of the transmittal letter comes from the "results" section of the SAER report, page 3. The revision's shift to " valid diesel generator tests in accordance with Reg. Guide 1.10s rather than i the number of successful starts since the event" is stated
! clearly. One key phrase is "since the event," which connotes to the knowledgeable reader a shift in the time frame for the counts from (1) a'fter the March 20 event until April 19 (the date of the l
original LER) to (2) after completion of the test program (as j
i defined in the June 29 letter) through April 19 (*10" and "12" set forth in the transmittal letter) or through June 7, 1990 i
(*12" and "16" set forth in LER 90-006-1, p. 6 of 9).
The petitioners ascribe nefarious intent to the fact that J
various drafts of the June 29, 1990 transmittal letter were prepared. The fact that several transmittal letters were prepared merely evidences the difficulty inherent in dealing with the subject matter (i.e., " tests,"Further, " valid failures," the drafts"were validjust j tests" and " successful starts").
j thatt preliminary documents which were subject to further verification and approval. None of them were the document i ( forwarded to the NRC. Nonetheless, a review of the drafts in their full text demonstrates the on-going effort of GPC to improve the accuracy of the transmittal letter in spite of the petitioners' selective paraphrasing of their contents on page 12 4
of their Petition under " Explanation contained in Draft."
First, the June 28, 1990 draft of 0751 AM Central Time (Exhibit 16) states that "only valid failures were considered in 4
I the conclusion that no problems or failures occurred" and that i
the number of tests was determined by counting testa regardless i of whether or not the test constituted a " valid test" under regulatory definition. These would have been inaccurate l statements of fact since, as established by interviews of the osI in August, 1990, " tests" and " valid fallures" were not counted by l involved personnel.
4 1
The June 28, 1990 0855 draft (Exhibit 17) appears identical i
to the earlier draft of 0751 except that the revised draft appropriately deletes reference to " valid failures" and changes l the key word " tests" to " starts": "the number of starts was These i
determined by counting Diesel Generator starts...." l l
modifications increase the accuracy of the draft by correctly identifying, using a lay ters, the things that were, indeed, ;
counted.
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[ The June 29, 1990 0755 (Exhibit 18) draft and 1142 draft
! (Exhibit 19) of the same date are each longer than the preceding ;
l draft, accurately describe the substance of the April 9, 1990 l
! letter and focus on the wording " subsequent to the test program" '
in the original LER. In both instances, the draft transmittal i
letter explains that if the report had stated " subsequent to the event," rather than " subsequent to the test program," the LER :
4 would have been consistent with the April 9 COAR and the "18" and '
l "19" numbess included in the transparencies provided by GPC to the NRC on April 9. This is a correct statement of fact. l I
?
The 1142 "Thedraft statement (Exhitit 19) in made includes the LER the andadditional in the April 9
' sentence:
letter did'not consider troubleshooting problems associated with j
the restarting of Diesel Generator 13, which was This,out also, of service is'a i
for maintenance at the time correct statement -- made with hindsight -- because the SAER of the event."
report identified " successful" starts associated with non-valid tests where post-maintenance problems were identified (e.g., fuel
! j priming) and these problems were not counted.
i i
With each iteration additional information was added to the !
prior draft to provide a more complete explanation of the " count" i
in the original LER and April 9 letter. This is indicative of
! the Company's attempt to assure accurate and useful information to the NRC Staff -- a revision to the. original LER, standing l f alone, would have resulted in a " correct" count and would have satistied notification requirements but would not have explained
(
i i why the revision was appropriate.
The June 29, 1990 1311 draft is essentially the same as the l transmittal letter forwarded to the NRC, with one exception. The i
word " discrepancy" in the last sentence of the first paragraph
! (Exhibit This final20) was more wording modified to "connotes clearly difference" a contrastin the final between version.
the
" count" in the transmittal letter and the " count" in the original j
m.
The final LER Rev.1 transmittal letter, then, draws on statements and conclusions made by the SAER group in its report of June 29, 1990. This makes sense, since the Senior Vice President had consissioned this effort by the group and would i And, as can be observed by execute the transmittal letter.
! reviewing prior drafts and comparing them to the final version of i
' the transmittal letter, the final version not only reconciles the original LER and the April 9 letter with the facts known as of j June 29, 1990, but also identifies the causes of the error (e.g.,
recordkeeping practices and the lack of definition of the time l
frame of the " count" due to the vagaries associated with the original " test program" wording).
9
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I The August 30, 1990 letter (Attachment 11 to GPC's '
f september 28, 1990 response to the Petition) from GPC to the NRC further expounded upon the differences in the " counts." The l
attachments to the August 30, 1990 letter contain tables which I list the starts using more extensive information than used as the '
basis for the April 9 transparency and letter and designates starts considered " successes" under a definition which is spelled The letter also ;
, out in the text of the August 30 letter. acknowledges error on the part o compiled the " counts" in his review of operations logs.
j In light of the revised LER, the information supplied to the osI, the independent review of diesel generator start data 1990 letter submitte ,
conducted py the OSI, the August 30, J to the Petition, and the further information provided in this :
response, the lack of merit of the allegation that GPC attempted J to mislead the NRC has been demonstrated exhaustively.
Relevant and controlling facts, including interviews conducted by the osI, the text of draft documents provided in this response, and the itsformal notification of the " count" error in the original LER were either unknown or not provided by the petitioners.
On these bases, sach standing alone, the allegation is demonstrated to be without merit.
E. Request for NRC Review of Diesel Generator Performance.
The petitioners, on page 12 of the Petition, request a such
( review of the performance records of the diesel generators.
a review, according to the Petition, will show unreliability based upon (1) the initiation of three different design changes, (2) additional " failures" after the original LER was submitted to the NRC, and (3) the unreliability of the ccaponents, apparently the temperature sensors, which are alleged The review to have requested bybeen known to the petitioners be unreliable for years.
has been fulfilled. First, GPC understands that several NRC representatives diesel engines.
have reviewed the performance records of theMr Hunt are believed to have professional opinions as to the reliability of the diesel engines, second, while GPC cannot i
divine the "three" specific design changes referred to by the petitioners, the NRC Staff is intimately familiar with the performance of and design change associated with the temperature l sensors of the energency diesel generators (see, for example, NRC staff comments filed January 11, 1991 in ASLBP No.90-617 OLA). Third, revised LER 90-06 and other special Reports to the NRC subsequent to the original LER have formally notified the NRC of additional problems, including " invalid" and " valid"
- failures." Fourth, the petitioners' allegation that the reliability of "the components" was "known toA be unreliablefact documented fer is years" is supported by no articulated fact.that the ARC has examine )
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.-- ._. -. . . = .. . . - - . . . . - .- ,
i i I
! l enforcement action resulting from the March 20, 1990 event, ;
/ whether information available to GPC should have been identified as precursors of that event, including the failures of the temperature sensors (sna Confirmation of Meeting letter dated l i
! August 22,1990 (at page 2) and October 1,1990 Enforcement Conference Summary letter from Mr. Luis A. Reyes (NRC) to ;
J
,.' Mr. W. G. Hairston, III (GPC)) . GPC's knowledge of the
! components' historic reliability, therefore, has already been l
considered by the NRC Staff. Further review is simply not ,
4 appropriate on the basis of a bald, conclusory allegation. l 1 F. Alleged Retaliation. ;
1 l I The retaliation alleged by Mr. Mosbaugh is the subject .
matter of ongoing Department of Labor proceedings, as explained i in the Enclosure to this submittal. By letter dated January 10, 1991, GPC provided the NRC with an explanation of the basis for
< the employment action taken with regard to Mr. Nosbaugh.
The Petition is not the apprepriate vehicle for resolution ,
4 of Mr, Mosbaugh's private cause of action, if any, and the l l
requested relief is inappropriate for this employment-related !
1 matter.
, III. ggnplusion.
Based on the foregoing, the company concludes that the petitioners' allegations are without merit.
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Ceer7s y
. , Pos ONce Bos 1295 Sarmngham. Ambaw 35201 Tewohene 205 877 7122
- c. x. mecoy GeorgiaPbwer Vce Prescent. Ne> ear Vogte Pro,ect the sotshem etcrre syste-December 10, 1991 ELV-03293 Docket Nos. 50-424.
50-425 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 )
Attn Thomas E. Nurley, Director Office of Nuclear Reactor Regulation subject: Additional Information Regarding Amended 10 CFR $ 2.206 Petition Filed by l Mr. Marvin Hobby and Mr. Allen Mosbauch Gentlemen:
By letter dated October 3, 1991, Georgia Power Company
("GPC") responded to your August 22, 1991 request for additional information. The request for additional information sought information associated with several allegations submitted by the petitioners in a July 8, 1991 supplement to their original petition.
Subsequent to GPC's October 3, 1991 response, we have l identified information which is relevant to the preparation of a l Director's Decision in this matter. The NRC may be aware of most of the information provided in this letter. Nevertheless, the purpose of this letter is to assure that you have the benefit of all material and relevant facts and circumstances during your deliberations.
I. Recommended Decision and Order in Marvin Hobby v. Georaia Power Company.
By Order dated November 8, 1991 in the U.S. Department of Labor case No. 90-ERA-30, (the " Order") the Honorable Joel R.
Williams recommended to the Secretary of Labor that the complaint of Mr. Narvin Hobby be dismissed with nreiudice. GPC previously had called the NRC's attention .o this proceeding, as explained in the Company's response, dated September 28, 1990 and April 1, 1991, in this matter.
______e __ _ _ _ - _ _ .._ _ _ _._. _ _ _ _ _ _ _ _ _ _ _ _ .
l l GeorgiaPower A l
[ A. Alleged False Testimony of Mr. R. Patrick Mcdonald. <
l l
The petitioners have repeatedly alleged that Mr. R. Patrick Mcdonald, the Executive Vice President of GPC responsible for i j nuclear operations, submitted perjured testimony during Energy l Reorganization Act proceedings before the Department of Labor. ;
Foremost is the claim that Mr. Mcdonald's testimony regarding the selection of SONOPCo Project staff was falso and that a thorough NRC investigation would demonstrate that the staffing was made in j a two-day meeting at the 270 Peachtree Street Building in i Atlarta, Georgia (see, e.g., pp. 10-15 of the petitioners' ;
l July 8, 1991 " Amendments" to the original petition).
l Judge Williams' Order addresses Mr. Hobby's assertions as l l follows: ,
l The meeting in preparation for the Fuchko and Yunker trial occurred six days after the meno establishing the NOCA l
. (Nuclear Operations Contract Administration group) was issued. I find that Complainant's (Hobby's) testimony, in l regard to his having been told by anybody involved in the ,
i proceeding that he would have to change any testimor.y that j he would give in that matter to conform to that of
' Mr. Mcdonald, to be totally unbelievable. I fail to see where Respondent's (Georgia Power Company's) attorneys would even consider having the Complainant testify about the sopoPCO selection process as he was not involved in the same and any testimony he would have given relating thereto would have been nothing more than hearsay. The Complainant is unable to identify the attorney who purportedly approached his with such an incredible request. The two partner attorneys, who conducted the two sessions which the Complainant attended, have denied making such a statement and I consider them to be credible witnesses. There were two other associate attorneys present in the meeting, but the Complainant made no attempt to subpoena them to the hearing. Although he allegedly relayed the purported conversation to Mr. McHenry the next day, Mr. McHenry was not examined at the hearing in regard thereto and I decline to credit his affidavit, prepared with the Complainant's assistance 1 1/2 years after the purported event.
Order at p. 40-41.
B. Alleged Unlawful Management Direction of the Licensee.
The petitioners also alleged that Mr. Mcdonald received his management directiott from Mr. Joseph Farley and, as a result, Georgia Power Comparty has improperly transferred control over its 2-e
l j GeorgiaPower A i
nuclear operating licenses. (see, principally, section III.1 of 4
the september 11, 1990 original petition.) After Judge Williams
- outlined the origin of this concern of Mr. Hobby (p. 21), the
! Judge then found as follows:
- , I recognize that in addition to the memorandum, the e
Complainant did mention a concern, as to Mr. Mcdonald's receiving his management direction from Mr. Farley instead of Mr. [Dahlberg] to Mr. Evans and perhaps others.
Mr. Evans did acknowledge the Complainant's having mentioned
. such concern 'in passing.' Depending upon the tone of such conversation, Mr. Evans could have taken the concern as the Complainant's personal one. Nevertheless, the time frame i for the oral complaints is not established in the record, i Mr. Smith (of Oglethorpe Power) laid the matter to rest in May 1989 upon receipt of the organization chart and 3 Mr. Williams' meno (of May 15, 1989). Although the l Complainant continued to be concerned about the reporting
- relationship in June 1989, when he corresponded with Admiral i Wilkinson, there is no evidence of record to establish that
- he continued to raise the subject with anyone beyend that
- time. Perhaps he had become as convinced as I as that Mr. Mcdonald did, in fact, take his management direction from Mr. Dahlberg in regard to the two nuclear plants owned,
- in part, by Georgia Power. Certainly, any doubts in his
! mind concerning the same should have been dispelled by the
. August 1989 meeting in reference to the Public Service l Commission case. The evidence referable to what transpired in this meeting clearly established that Mr. Dahlberg exercised control over Mr. Mcdonald regarding Georgia j Power's nuclear operations.
l Order at 42.
] C. Alleged Retaliation for Raising Concerns. >
Finally, the Order is relevant to the petitioners' non-4 specific allegation that Georgia Power retaliates against
- managers who raise regulatory concerns (Item III.9(d) of the j
- original petition). 2 I find that the decision to eliminate (Mr. Hobby's) position of manager of NOCA was in no way related to the 1 Complainant's participation in the January 2, 1989 meeting ,
! [in which he allegedly raised the concern about the accuracy )
i of Mr. Mcdonald's testimony regarding the selection of I employees for the SONOPCO Project) or the concern raised in l i his April 27, 1989 memorandum as to from whom Mr. Mcdonald j receives his management direction for operation of the i I
1 j e I
. Georgia Power d i
J Georgia Power nuclear plants. I find that, instead, the decision to eliminate the position was fully justified as a l
measure to operate the Respondent's nuclear program more economically and efficiently.
]
f, order at 44.
II. Vostle Snacial Team Inspection Report Nos. 50-424. 425/90-l
' 19 Sunnlement 1. Nov==har 1. 1991.
Sections 2.1, 2.2, 2.4, 2.5, and 2.7 of the above-cited Inspection Report address, in whole or in part, allegations in l
Sections III.8, 6(e)(iii), 5(a), 3 and 9(d) of the original 4 petition of September 11, 1990, respectively. Each provides a factual basis for concluding that allegations of wrongdoing are i
l unsubstantiated, although the alleged events and technical 4
deficiencies may have occurred. It should be noted that the NRC's inspection efforts which form the basis of the Inspection Report were initiated over a month prior to the September 11, j 199C submission of the original petition.
]
GPC is aware of other NRC Inspection Reports on this docket
' which are relevant to aspects of the Petition. First, Inspection
- Report 91-20, dated September 12, 1991, at page 4 of the report i " Details," addresses the allegation contained in Section III.6(c)
- of the original petition. Second, Inspection Report 91-14, dated i
July 19, 1991, in particular Sections 2.b, c, f, and 3.c provide factual bases demonstrating the falsity of the general allegation l contained in Section III.6 of the Petition that GPC " subverts"
! the requirements of Technical Specifications. Third, the two allegations in Section III.6(e)(i) and (ii) were addressed by the j NRC well before the submission of the original petition.
Inspection Report 90-10 dated June 14, 1990 sets forth factual j conclusions relative to the two events which are the subject j natter of LERs 1-90-004 and 2-90-001 and which were previously identified as non-cited violations. See, also, Inspection Report 90-19, Supplement 1, page 2, first paragraph, last sentence.
l III. Petitioner's Possession of Draft 759 1-90-004.
1 l
The petitioners alleged that GPC personnel purposely violated Technical Specifications in order to keep the Vogtle
- Electric Generating Plant operating or to hasten the restart of a i unit (Section III.6(e)(ii) of the original petition). GPC has
- recontly obtained, in conjunction with discovery in
- Mr. Mosbaugh's Department of Labor proceeding, a draft (Enclosure i 1) of Licensee Event Report 1-90-004, dated March 7, 1990.
Notations on the draft, in what appears to be Mr. Mosbaugh's handwriting, indicate that orior to submitting the original
! -4 4 - , - -
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Georgia Power d petition Mr. Mosbaugh knew: 1) operations Department management 1
~
identified the noncompliance of Technical specifications on a morning status conference call, and 2) the reason for the i noncompliance of Technical Specifications was that information concerning equipment out-of-service had been placed on the back of a Limiting Condition of Operation status sheet. Thus, it appears Mr. Mosbaugh possessed information prior to September l 1990 which was contrary to the statements made in his
- september 11, 1990 petition. .
1 IV. Particinants in an Acril 19. 1990 conference call Reaardina j fMR 1-90-006.
GPC's October 3, 1991 supplemental response sets forth the basis of GPC's April 1, 1991 statement regarding the Senior Vice President's lack of participation in a telephone conference call late on April 19, 1990 which finalized LER 1-90-006. In late October, 1991 in conjunction with discovery in Mr. Mosbaugh's l Department of Labor proceeding, GPC obtained cassette audio tapes i j which were surreptitiously made by Mr. Mosbaugh during, approximately, the February-September, 1990 time frame. One of i those tapes of April 22, 1990 discussions (identified as Tape No.
! 71) indicates that Mr. Hairston was R21 a participant during the j April 11, 1990 telephone conference call when language concerning emergency diesel generator start counts was finalized in the LER.
l The following is a transcript of a portion of this tape which contains a discussion between Mr. Mosbaugh (" ALM") and another participant ("P") on the April 19th conference call.
ALM: I think there is a high probability that there is a problem with their statement (in LER 90-06 concerning diesel generator start information).
P: What George told me over the phone--
l ALM: George who?
P: George Bockhold--
ALM: When?
P: Before we issued the LER.
ALM: Yeah.
P: We had a big conversation on those numbers with George [Bockhold), uh, [ George) Emirston--
ALM: Yeah.
d
I Cleorgia P(nter d$k F --or mot Emirston, [3111] shipman.
ALM: They were all on there.
~
P --and what George [Bockhold) said is they had explained to the Region that they had had--they used--
Tom got those numbers from what we presented to the NRC 4 and then he just added the additional starts after I
- that---
s - ALM That's right.
- P: But--and we questioned that with George (Bockhold) and what George (Bockhold) said is, 'Yes, we did have 1 failures; the Region was aware we had failures, but we '
were in the troubleshooting mode and once we cleared the troubleshooting mode then we had that many successful starts.'
- ALM That's not true--
i - P: That's what I was told.
ALM:
You can interpret-- !
i
- ALM Yeah, but that was a presumption on George's )
- part-- l 1
P: No. George did not presume anything. He made that
- as a statement of fact and all that information was presented to him before he made the NRC presentation
[on April 9, 1990) and that's the way he made the presentation. ,
- ALM: OK, well that's--
1 P George was aware of the fact that [ inaudible)-- 1 ALM: You've got to establish, you know, you have 4
weasel words in this thing, you've got to establish a i criteria, OK, between X and X, how many successful i starts do I think I had? What'. my criteria? The words in there say ' failures and problems.' What's a problem?
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GeorgiaPower A 1 l h
P: Well, I think probably the more appropriate way would have been to word it to say, 'We have had--we {
have had eighteen consecutive starts without a trip j j from this date going back.'
l
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ALM: Whatever--the words in there are weasely,
, OK,-- )
P: Well, they weren't-- l l -
ALM: --they say ' failures and probleas' and they say j 'since the 20th.'
P: They weren't intended to be weasely. From my {
r standpoint, they weren't intended to be weasely. )
1 ALM: You can read those words a couple of different l ways. All I'm saying is that somebody, you know, we
- need to decide what w; missed--
1 i -
P: You'll probably want to mention that to George j [Bockhold)--
ALM: We need to decide what we missed, then we need to
' review the data and see if what we meant is true or not, but I have yet to be able to figure out, among the i various ways of interpreting it, I find a flaw with j aach method of interpreting the words.
i As can be observed from the highlighted portion of this excerpt, the participant indicated that Mr. Shipman and not Mr. Hairston participated in conversations which finalized the This is consistent with the collective recollection of LER.
5 participants during the August, 1990 special inspection, as reflected in documents enclosed in GPC's October 3, 1991 j supplemental response. Moreover, the conversation indicates that '
1 Mr. Mosbaugh, as of April 27, 1990, had not reached a personal conclusion of " material false statement" relative to the LER.
! GPC's prior responses include an April 30, 1990 meno to
- Mr. Bockhold from Mr. Mosbaugh which indicates to us that Mr. Mosbaugh's conclusion regarding incorrect, material j information in the LER crystallized only after the original LER
- had been 1:orwarded to the NRC.
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Georgia Power d i
j I hope this information will be helpful to resolve these matters in an expeditious manner. If I can be of any further assistance, please do not hesitate to contact me.
Very truly yours, l
C. K. McCoy l
Enclosure cc: Georaia Power Company '
Mr. A. W. Dahlberg Mr. W. G. Hairston, III 1 Mr. W. B. Shipman l l Mr. P. D. Rushton l Mr. J. T. Beckham Mr. M. Shelbani' NORMS U. S. Nuclear Reculatory 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 l
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- i ENCLOSURE l
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1 l L E A l 60'i - -
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! Mgr. Engr. Supt. Date Mgr. MP/chen. Date Im/ M7/f/ Date Mgr. Maint.
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Date pr.Tggh. Sept.
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Mgr. Ops.
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1 Mgr. Admin. Date Date Aff / Lh /
- Mgr. Trng. Data Mgr. Out./ Plan Date LN / A /
l Asst. General Mgr. - Support Data Asst. General Mgr. - Operations Date l l i I i
j ru Mts. no.
Please review, sign where appropriate and return to ial!Ihe ImMk by .
)
If there are any questions, please call N!ff*P S ed at extensten 331@.
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RL90770
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.ER 1-90-004
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" FAILURE TO COMPLY WITH TECHNICAL SPECIFICATION 3.0.4 OCCURS ON i4 ENTRY INTO MODE 8' i EVENT DATE: 3-1-90 i ASSTRACT:
t On 5-1-90. at 0133 CST, a failure to cocely with Technical Spec 181 cation (T.S.) 3.0.4 occurred when Unit 1 entered Mode s i
(Refueling) f rom Mode 5 (Hot Shutdown). Prior to entering Mode 4
i 6, a Limiting Condition for Operation (LCO) had been Tnitiated for Source Range Channel 1N31 to ellow performance of a 18 sonth channel calibration. Although this LCO remained in effect, the
]
j Shift Superintendent signed of f on th's applicable procedure to
- indicate he had reviewed the LCO Sook for impact en entering Mode 5 and that approval was granted to change status free Mode 5 to Superintendent Moos 5. After entry into Mode 6 the Shift accognized that T.S. 3.9.2 reauires two Source Range Monitors to j
ope.able in Mode 6 and that a failure to comply with T.S.
be -
3.0.4 had occurred. No immediate action was required since the f
. action reauirements of T.S. 3.9.2 were satisfied.
cause for this event is considered as cognitive
- The root The Shift personnel error by the Shift Superintendent.
- - be Superintendent has been counseled and a copy of this LER will a
placed in the Operations Required Reading Sock.
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I 1 E90772
9 A. REOU!REMENT FOR REeOR*
l report is reevired per 10CFR50.73(a)(2)(1) because of a This failure to comply with Technical Specification (T.S.) 3.0.4
- S. UNIT STATUS AT TIME OF EVENT Unit 1 had shutdown to commence its second refueling outage.
, This event occurred when Unit i entered Mode 6 (Refueling) from Mode'5 (Cold Shutdown). Reactor coolant temperature and pressure l' Fahrenheit and 0 psig were approximately 110 degrees respectively. Additionally, the Reactor Coolant System was i
i drained to midloop and nozzle dans had been installed.
, C. DESCRIPTION OF EVENT ,
On 2-20-90, a Limiting Condition for Operation (LCO) was entere'd i
allow performance of 18 month surveillance 24695-1
- Naclose to Channel Instrumentation Syster (NIS) Source Range Channel 1N31
]
Calibration". Entry of the LCO f or Source Range Channel 1N31 was
! appropriately recorded in the LCO Sock and in the Unit 1 Shift Supervisor Log.
On 3-1-90, procedure 12007-C
- Refueling Entry (Mode 5 to Mode 6)*
I was being performed in preparation for entry into Mode 6. Items by the Shift (4) and (5) of step 4.3.1.c wereStepcompleted 4.3.1.c reads:
- REVIEW Superintendent and initialed off.
(1) Jumper and
! the following for impact on entering Mode 6: (3) Equipment L1fted Wire Log, (2) Temporary Modification Log, (4) LCO Book. (5) Outstanding Work Orders.'
- At Clearance Log, l
0014 CST, the Shift Superintendent signed off on procedure 12007-Mode 6.
C to indicate soproval to change status f rom Mode 5 to At 0133 CST. Mode 6 was antered when Reactor Vessel Head detensionTng commenced.
Several hours later, the Shif t Superintendent was reviewing the l LCO Sock in preparation for shif t turnover and recognized that a failure to comply with T.S. 3.0.4 had occurred on the entry into l Mode 6. At the time of the w. ode change, the LCO for Source Range still in l Channel 1N31 was atill in effect and the channel was T.S.
!
- test
- for performance of surveillance 24695-1 3.9.2 i eseu1res two Source Range Neutron Flux Monitors to be operable in Mode 6. Therefore, the reauirements of T.S. 3.0.4, which state in oart " Entry into en OPERATIONAL MODE or other soecified condition Limiting shall not be made unless the conditions for t he provisions Condition for Operation are met without reliance on contained in the ACTION recuirements', had not been fully complied with. No im=ediate corrective action was required due I to this discovery since the action reautrenants of T.S. 3.9.2 l
! l were satisfied.
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RL90772 )
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- D. CAUSE OF EVENT The root cause for this event is considered as cognitive personnel error on the part of the Shift Superintendent. In reviewing the LCO Sook and signing off on procedure 12007-C, the Shift Superintendent should have recognized Source Range Channel IN31 as being a modo change restraint. There were no unusual characteristics of the work location that contributed to the occurrence of this event.
- b~
E. ANALY3!S OF EVENT
- Qn$%eht* '
The action requirements of T.S. 3.9.2 state that with. one Source l Range Neutron Flux Monitor inoperable or not operating, to j 1mmediately suspend all operations invo,1ving COPE ALTERATIONS er positive reactivity changes. These action requirements were complied with. By 1120 CST on 3-1-90, surveillance 24 895-1 had been completed and the LCO for Source Range Channel 1N31 was exited at that time. Since the action reouirements of T.S. 3.9.2
- were complied with, it is concluded that there was no adverse effect on plant safety or on~ the health and saf ety of the public.
) F. CORRECTIVE ACTIONS
- 1. The involved Shift Superintendent has been counseled regarding his f ailure to recognize IN31 as a modo change l aestraint.
- 2. A copy of this LER will be placed in the Operations l Recu1*ed Reading Sock to reemphasire the need to be aware of mode change restraints.
G. ADDITIONAL INFORMATION i 1. Failed Comenment Identification
- usa..
1 l 2. Previous Similar Events i
A failure to fully comply with T.S. 3.0.4 previously occurred for Unit 1 on 10-28-07 (reference LER 424/87-051),
- when the Unit changed status from Mode 4 (Hot Shutdown) to
! Mode 3 (Hot Standby) with certain recuired equipment having not been verified as ope'r able prior to completing the mode i c hange. However, the root cause for these two events differ slightly in that the earlier event resulted from a failure to implement "Information Only LCO's".
- 3. Energy Industry Identification System Codes Incore/Excore Monitoring System - 10 i
RL90773
P.1 Jtt 15 '93 02:32PM OGC/f%ILROOM O
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. AL 15 '93 02:32PM OGC/m8LPOOM ]
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. N r.i UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION '50 dL 14 A:: M6 Commissioners: * [,{, .
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Ivan Salin, Chairman ;
Kenneth C. Rogers Forrest J. Remick E. Gail de Planque .
)
M I4M
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In the Matter of )
' ) Docket Nos. 50-321 GEORGIA POWER COMPANY, 50-366
) 50-424 at 11& ) 50-425
)
(Hatch Nuclear plant, Units 1 & 2; Vogtle Electric )
(10 C.F.R. 5 2.206)
Generating Plant, Units )
)
1 & 2) )
v=vnesunum nun nunna ;
CLI-93 -15 staff's partial i The Nuclear Regulatory Commission (NRC) j 37 NRC ___ (Apr. 26, decision under 10 C.F.R. $ 2.206, 0D-93-08, in 1993), is pending before the Commission for possibis review l For the reasons stated in accordance with 10 C.F.R. $ 2.206(c).
this order, the Commission is vacating the staff's partial l
decision and remanding the matters decided therein to the staff for further consideration.
The staff's partial decision responds to a petition filed by and Allen L. Mosbaugh and Marvin B. Hobby in Sept.mber 1990, further supplemented in October 1990 and July 1991, which asked for initiation of proceedings and other enforcement action The petitioners based their against Georgia Power Company (GPC).
. p.3
. JUL 15 '93 02:33PM OGC41A8LRO0t1 l.
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petition on various allegations of falso statements, willful In DD violations of NRC requirements, and other misconduct.
08, the staff denied the petition with respect to certain of the petitioners' allegations which the staff believed were capable of final resolution. However, the staff declined to reach a determination with respect to allegations of unlawful discrimination against Messrs. Hobby and Mosbaugh which are related to pending proceedings before the United states Department of Labor and to other allegations of wrongdoing which are still under consideration by the NRC staff.
In addition to his filing of the section 2.206 petition, Mr.
Mosbaugh has been admitted as an intervenor in a proceeding on the transfer of operating authority over the Vogtle Electric 1
Generating Plant from GPC to Southern Nuclear Operating Company Among the bases for his admitted (Southern Nuclear).'
consolidated contention in the adjudicatory proceeding are the allegations also contained in the section 2.206 petition that GPC and Southern Nuclear had consummated an unlawful da facto transfer of control to Southern Nuclear of the operating licenses for the vogtle and Hatch facilities, and that GPC's executive 1991, vice president in a meeting with NRC staff on January 11, made material falso statements about the formation of Southern Nuclear. The staff denied the petition on the merits with I LBP-93-5, 37 NRC 96 (1993) (appeal pending before the Commission).
Mr. Hobby also petitioned to intervene in the He transfer proceeding, but was denied standing to intervene.
has not appealed the denial of his intervention.
P.4 AA.13 '93 02:34PM OGC/MAZLRCOM a
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i- 133 slip op. at 5-17 and ,
respect to these matters in DD-93-08. ,
t 53.
The Commission has generally discouraged use of section >
2.206 procedures as an avenue for deciding matters that are under I
l Thus, the commission '
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consideration in a pending adjudication.
j ordinarily would expect the staff to deny a section 2.206 idered in ,
!, -petition that raises the same issues that are being cons dency of the
.a pending adjudication on the basis of t he pen identical matters in a proceeding involving the same licensee or facility This general rule is not intended to bar petitioners l
from seeking immediate enforcement action from the staff in
- f circumstances in which the presiding officer in a proceeding is l Moreover, we recognize here not empowered to grant such relief.
' that Mr. Mosbaugh has not invoked section 2.206 to avoid a pending adjudication and that his section 2.206 petition seeks relief with respect to issues and facilities that are not before l However, the Licensing Board in the pending transfer proceeding.
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in view of the overlap and similarity of some issues between the section 2.206 petition and the transfer proceeding (particularly those addressed.in sections II.A. and II.B. of DD-93-08), the l
' l staff's final determ!.ation of the common issues should take into 1
3 .
Nuclear Coro.. (Three Mile Island I 333 caneral Pub. Utila.
Nuclear Station, Units 1 & 2; Oyster Creek Nuclear Generating CLI-85-4, 21 NRC 561, 563-65 (1985); Pacific Gas &
-Station), CLI-Elec. Co. . (Disblo Canyon Nucisar Power Plant, Units 1 & 2) ,
81-6, 13 NRC 443 (1981).
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. JUL 39 '93 02 34PM OGC/t1AZLR00M P.5 4
account the Licensing Board's findings and the outcome of the transfer proceeding.3 Apart from the commenality of some issues decided in DD 08 with pending issues in the adjudicatory proceeding, a common thread runs throughout the allegations raised in the section 2.206 petition. The issues raised in the petition generally e
concern the integrity of GPC or Southern Nuclear officers and t corporate organization responsible for operation of the Hatch ano Vogtle plants. Under the particular circumstances of this case, rather than address the issues in the section 2.206 petition in a 1 l
piecemeal fashion, the staff should reach a determination of.all )
inaues in an integrated manner after consideration of the remaining matters raised in the section 2.206 petition and the outcome of the transfer proceeding.
We therefore vacate DD-93-08 and remand to the staff those portions of the section 2.206 petition decided therein for the staff's further evaluation and final decision in conjunction with the staff's resolution of the other remaining matters in the petition and in light of the outcome of the transfer proceeding.
In taking these actions, we intimate no view on the soundness of the staff's analysis of the issues in DD-93-08. We also do not l
3 We recognize that GPC has appealed the Licensing Board's admission of Mr. Mosbaugh as a party and of the consolidated contention. We expect to render a decision on the appeal in the near future. Nonetheless, at least pending a further Commission order on appeal, Mr. Mosbaugh is entitled to party status and his contention is deemed admitted.
P.6
, JUL 15 '93 02:35PM OGC414ILR00M 5
b r the staff from taking prompt enforcement action at any time during its ongoing review of the matters raised in the petition.
It is so ORDERED.'
For the Commission
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SAMUEL . CHILK Secretary of he Commission q
Dated a Rockville, Maryland, this/ ~ day of July 1993.
d
' Commissioner da Planque did not participate in the Commission's consideration of this order.
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SOUTHERN COMPANY SERVICES l NUCLEAR PLANT SUPPORT V0GTLE 4
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FARAMETERS USED IN EVALUATING THE RADIOLOGICAL . >
CONSEQUENCES oF A LIQUID RADWANTE TANK FAILURE .
Source data I
! l Core power level (MWt) 3565 Defective fuel (%) 1 Atmospheric dispersion factors i
see table 15A-2.
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Iodine gas activity (percent 1 1
of tank contents)
Tank contents subject to release ..
- activity released to the See table 15.7.3-1.
environment j Muslide o to 2 h (C1)
Kr-47 i Kr-84 5.5E+2 i
Kr-89 1.6E+3 6.6E+1 Xe-133 i
Xe-135 1.1E+5 j Xe-138 3.1E+3 i
2.7t+2 4
1- 13 1 1-132 1.2t+0 I-133 1.2E+0 I-135 1.st+0 4
1.0E+0
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Radioactive contents 4
)
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Kr-SS 1.56 x los
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W cale Xac1901, man 2 a _ . P--- T af /d ARE Sffet te Bana Anm1wata eram E. # Pd dman #1U43 l Alternate Calculation of Offsite Releases From ARB Pipe Crack Purposea This alternate calculatten is performed to verify the conclusion in J. A. Wehrenberg's calculation that the offsite effect of the failure of a pipe in the ARS is within the limite of 10CFRec. L Scope ' This esiculation addresses only the offsite releases from the primary calculation. The release rates are the same as those in the primary except as noted in the assumptions. Conclusions This alternate calculation clearly demonstrates that a s
/M MN failure of a pipe in the ARE with resultant spray out the building will not create a condition which would cause the dN p plant's limitsannual average release concentrations value is 43M to ofexceed the j Q of 10CFReo. The computed the 10CFR3C limit. 01ven the different assumptions this agrees well with h ad ' the value of 85M in Mr. Wehrenberg's calculation. While slightly higher than the previous FSAR value of 35%, the computed release is well within the regulatory limit.
Assumpttensa j The saan basic assumptions apply as stated in the principat
- calculation except that 100M of the water is assumed to be released instead of the SOM assumed by Mr. Wehrenberg. The ,
l effect of this assumption is to double the releases. {- Methods i The alternate calc uses PSAR Table 11.E.5-2 as a starting , point. That table is a listing of the expected releases of radioactivity per year from the plant for each isotope of MQ - AS concetti Jhe total annual release _ of .1-131, was divided, by g th's prejecte fe'engeniration to ob_tain a design besle-j* J114tig3r The activity released from %e ARE pipe crack in
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b p[(My, ed thirty minutes is then computed for each of the major isotopes by multiplying its concentration (in uC1/ml) by hoF 63 SPM times thirty minutes times the factor to convert gal )6M 3 4 to al. That value is added to the FSAR release value to That r.tesse is in.n multipli.d by bd d, obtain . total r. lease.sho w tution facer; to obtain a prosseted annual !
- p- conc.ntrati.n. Th. n.w av.ra.. cone.ntrations are eu the total displayed at the Bottom of the page.
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Vogtle electric Generating Plant Prepared., .r ARN offsite Dose Analysis Verifle _ date .f.d Alternate Calculation for Att 1 .date__ . Ca l c . X4C1901.8503 Page Jf_ of 1[ NOTE: The following is an alternate calculation for the assessment of the impact of a pipe failure in the ARE on the plant's complian'ce with 10CFRt0. This calculation is based upon the dilution factor derived froc. FSAR Table 11.2.3-8. l ! h Estimate of Effect of Postulated ARB Radweste Line Failure on 10CFR30 Compliance ! Dilution Factor = 2.5E*13 Spray Mate = 63 GPM Isotope 1-131 1-134 1-133 I-134 Cs-134 Cs-136 FSAR TEL i 11.E.3-2 Release tuC1) ! i 1.10E+05 5.50E+03 S.SOE+04 E.00E+01 E.10E+04 8.60E+03
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11.E.3=E l Conc (uC1/ml) 4.46E-09 E.23E-10 3.45E-09 8.11E-13 8.51E-10 1.0SE-10 l Spray Event Release Conc. (uct/ge) 8.80E-01 E.90E-01 4.20E-01 E.30E-01 3.07E-08 3.87E-CE t Total ! Melease (uct) R.00E+06 8.00Evo6 3.00E+06 1.65E+06 E.19E+05 E.77t+05 Combined Release Total Release (uC1) 2.11E+06 E.01E+06 3.09E+06 1.65E+06 E.40E+05 8.79E+05 Release Cons. (uCi/ge) 8.97E-05 0.14E-08 1.ESE-07 6.67E-08 9.75E-09 1.13E-08 Maximum Perm. Oone. (uCL/ml) 3.00E-07 B.00E-06 1.00E-06 E.00E-05 9.00E-06 0.00E-05 Maximum Perm. , conc . (ratio) E.368-01 1.0tE-08 1.25E-01 3.34E-03 1.00E-03 1.26E-04 Sum of.MPC ration = 0.486 g. f gh
WOGTLE ELECTWec GranamATggeg ptany a.es menesa.as rowsm oewessom CALCULATION TITLE SHEET JOB NO. 9510
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ALTERNATE RADWASTE BUILDING LI?uIC PACWASTE SYSTEM FAILURE OFFSITE DCSC ANALYSIS
?
SCS calculation X4C1901S53 was performed under REA VG-9057 l on 3-4-90 to calculate a groundwater and airborne release j from a failure in the ARB.It concluded that the release was ; only about 33% that previously calculated in the FSAR for a ! failure of the Recycle Holdup Tank.This calculation d1d not : evaluate a pathway by which 11 auld leaves the ARB as a surface water release. On 6-27-90 an additional calculation was performed to ! evaluate the surface water pathway.It wrongly concludes that l Regulatory limits are met.Severa,1 major errors are made in ; the calculation as follows: i The calculation assumes that the radioactive surface water flows west and north as this area 1s used to ; compute the drainage area for rainfall dilution.In fact a s1mple review of storm dra1n drawings f AX2D45S001 and AX2D45S002 shows that the water flows l east then southeast. The calculation assumes that the water 1s discharged j into the Savanna r1ver at the plant discharge point ; as it assumes "near field dilution" factors of 10 and i references the dilution effects of cooling tower ! blowdown and dilution flow.In fact the radloactive I water woulo leave Georgia Power property in the small stream that flows under the low spot on New R1ver Road (Rt 56 spur) Just before tne intersection with the Visitor access road near the Training Center (see drawing CX2D45V002). As a result of these fundamental errors the calculations performed are vold.As prev 1ously addressed part 21 definitions of" significant safety hazzard" include radioactivity concentrations in unrestricted areas in excess of 500 times the part 20 limits.Previously it was not stated that yearly average part 20 limits would be violated but considering the actual pathway leaving GpC property, near field dilution credit is inappropriate and yearly averages may well be exceeded, f a D 1 e
4 i NRC REACTIVE IRSPECTION 8-06-90 4 NRC: Vanoenburgh, Chris A. - Team Leader
! Wilcox. John D. -
Assistant Team Leader l Hunamu11er. Neal - INTR Operator License Branch (sim. cert.) 4 Branch Morris - Senior Resident Matts Bar Thomas. McKenzie Region II Inspector (Q.A.) Taylor. Pete - Region II Inspector (Reactor Sefety) Gamer. Larry Senior Resident Pabinson e3 ass 7a 0 ,1 1 Brockman. Ken - Section Chief Anactor Projects / h Mattheus. Dave - Project Director i i i Reyes. Luis- - Director Division Reactor Projects
- j Robinson. Sheila - Secretary l CM TACTS OPERATIONS -
Smartzwelder Jim Extension 3618 i J Beeper 044 i MAINTENANCE - Handfinger. Harvey Extension 4278 Beeper 310 i H.P./CEMISTRY - Seepe. Mike Extensten 3380 l Beeper 252
- Q.A. - Frederick, George Extensten 3228 Beeper 170 l
- TECMICAL S!!*POR? -
Williams. Sus Extensten 4279 Beeper 019 l l Aufdenkampe John Extension 3600
- Beeper 101 i SECURITY -
Danneuiller. Ted Extension 3o37 ! Beeper 444 i i ADMINISTRATION - Quick Brent Extensten 3114 i Beeper ! ENGINEERING SUPPORT - Ealick, John Extensten 3545 Beeper 102 i Horton. Mike Extension 3121 ! Beeper 107 TRAINING - Broun, Bob Extensten 3923 i OUTAGES, AER PLAIELING - Beasley. Bernie Extension 4209 _ Beeper 194
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l 'v%kenz:e I hoano n Rene h < Zit.sneeL- N&clRee;oa E ! v i / Cow 4%ru e /d G C zah'* yz s/9 tv ' c //?E l b n n**- A-e ru. c+s 7. ne ex a #s ar A VC S N T {L.Jt N F b7 N R. f._ //?,. T i r o v m t w sets a ra scrAt+ nc/M i Nur X hmvua An i L . ora m hvit , l J L V A<o ,s L/Ew. % A < 4479 Te A % s s a. f / d k I J.8 Be os/.u.i de 64%sl//omi-' dzov iL L UA a 'L l A C.. b [ n % -te % r .,,a. 9,n wi fcam G 2. Fireu aick suryt - saa 73ze ? c,-/c. j E.M. humi, e cp r Qaupe hem g,q gg3, .yr,se m / g . Sn, frmeerain Msr Lar Aswin 3/i 3 GPC. s JELaertafLDEE - $>5. Me,e . %t8 C r'.S / G W i kU[$ N $ & $'L5C.0L b 't ? Y &fY (~2d Bl u GH 3ist GPc-f0'h h W kbih .:13 Y SA. . l Lee Gew Mu, Gowr (wee.s ,1-ll,1r reu, m / se O.-D 3L9V
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e AUG-87 '90 13:31 ID:50POPCO-UDGTLE TEL to:1-295-977-7985 c343 Pe6 E
'. Dee*9 e *omer Cempe% >
j* 233 PecmorW A.enue e Ar41a Georga 30300 Tetonnone dos $26 3196 j .'
. t . Wa6n0 Adoress 4
40leverness corner Parmeer 8s0 Cwles Som 1795 S.rmingham. Mooema 36301 weenon.20s ses sset "8"'* 7' IO en. ,,.a m,c ,,m. i n. a. nennen. m sener ve. P=.orv Nucnear Operstone ELV-01995 0535 ] Docket No. 50-424 i 50-425 i U. S. Nuclear Regulatory Commission i ATTN: Document Control Desk l Washington, D. C. 20555 l 1 Gentlemen: ]. ! V0GTLE ELECTRIC GENERATING PLANT SPECIAL REPORT VALID DIESEL GENERATDR FAILURES i In accordance with the requirements of the Vogtle Electric Generating Plant
- Technical Specifications, Sections 4.8.1.1.3 and 6.8.2, Georgia Power Company
- hereby submits the enclosed Special Report concerning three valid diesel 4
generator failures.
- Sincerely, i
- W.J'. ,d 1 Y _- .
1 W. G. Nairston, !!! WGH,111/NJS/gm
Enclosure:
Special Report 1-90-05 a xc: Genreia pauer c - =nv
- Mr. C. K. McCoy Mr. G. tockhold, Jr.
j Mr. P. D. Rushton i Mr. R. M. Odes NORMS U. 1. W mlaar eaaulatary Cnem1ision Mr. $. 1). Ebneter, Regions' Administrator I
- Mr. T. A. Reed, Licensing Project Manager, NRR l Mr. B. R. Sonser, Senior Resident Inspector, Vogtle <
l/g' / i ;
, f .
M o o r e n d ____ 4 3 P j
. . - . - _ _ - - . ~~ - - - - - . -
- M 5-97 '98 13
- 31 ID SONOPCO-V0GTLE TEL N0:1-205-e77-7895 a343 Pe7 e i r '. * '
i
= -
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' i 1
i ! i ENCLOSURE 1 V0GTLE ELECTRIC GENERATING ptANT - UNITS I AND 2
- TECHNICAL SPECIFICATION SPECIAL REPORT l-90-05 i VALID DIfsft efNfaATok FAILUAfs i A.- REQUIREMENT FOR REPORT
! l'
- This report is required in accordance with Technical Specification (TS) 4.8.1.1.3.
j This specification requires that all diesel generator (DG) j failures, valid Report pursuant ortonon-valid, TS 6.8.2. be reported to the Coastssion in a Special B. DESCRIPTION OF EVENT j . On 7-!!-90, Diesel Generator (DG) 2A was being tested during a routine surveillance per procedure 14980-2, " Diesel Generator Operability Test". - The right air start bank was isolated to allow testing of the left air start bank. The engine start button was pushed by the control roon 1 i operator and the engine began to roll with starting air. ; I i the local operator in the diesel room, the engine rolled twice andAccording to stopped. The DG was declared inoperable and the T1 action statement was
- initiated. The DG was unavailable for emergency operation for a period l
of 67 hours and 49 minutes. i During the review of this event, it was determined that similar events ; had occurred on 4-12-90 and 7-5-90. These previous sistlar events had i
- not been recognized as failures and therefore had not been reported as such. These events are described as follows
- ;
i l on 4-12-90, operators conducted a TS surveillance test of DG 1A per procedure 14980-2. The manual start button was pushed, but no start ; i occurred. Operators decided that the $vshbutton had not been depressed ' long enough and made another attempt wtich resulted in a successful start. On 7-5-90, a steilar incident occurred on OG 18, and a successful start again resulted en the second attempt. Neither DG was considered to ! be unavailable for emergency operation as a result of these two events. ' C. CAU$t 0F EVENT An investigation into the 7-11-90 event by utility and vendor personnel i
- found that the starting air valve pistons could stick in their cap i
assemblies due to inadequate manufacturing tolerances. This condition was apparently the result of the initial manufacturing process which left insufficient clearances between some of the pistons and caps. A failure to start would occur only after the engine had been shut down from a
- previous run and the engine stopped with a particular alignment of faulty air start valves and crankshaft position.
i . !, AUG-07 '90 13:32 ID:SONOPCO-v0GTLE i TEL PC:1-295-977-7985 c343 Pee 5 e *
- i
}=
i i ENCLO5URE l 2 V0GTL'E ELECTRIC GENERATING PLANT - UNITS 1 AND t j TECHNICAL SPECIFICATION SPECIAL REPORT l-90-05 VALID DIftft efNfeATOR FAILURft
- i 1
On a non emergency manual start with the air start pliot valves malfunctioning, the initial burst of air was not adequate to start the engine. The ourst of air was adequate to change the altenment of the ; j crankshaft with respect to the faulty air start pilot valves so that any I ! subsequent attempt to start the engine could be successful. This problem l is now believed to have been the cause of the OG failures on 1-24-90 and l-25-90, which were reported to the Commission on 2-19-90 as Special i Report 2-90-02. On 7-19-90, the manufacturer of the valves submitted a j 20 CFR 21 report to the Commission as a result of the above findings. The failure of the DG operators to recognize the initial start attempts 3 of 4-12-90 and 7-5-90 as DG failures is partially attributed to l limitations of the simulator computer. The simulator requires operators i to hold the DG manual start push >utton in order to have the proper i control signals annunciate, creating the misconception that the l pushbutton must remain depressed for a given period of time in order for
- a DG start to occur.
{ D. CORRECTIVE ACTIONS i
- 1. The sixteen starting air valves on each of the four DC's were tested l and polished where necessary to provide adequate clearance between the pistons and caps.
- 2. The appropriate maintenance procedures will be revised by the next j
refueling outages to require testin0 of the starting air valves to
- demonstrate freedom of movement following 06 overhaul.
i ! 3. During shift briefings, operators were advised that the DG should l start when the sensa pusibutton is depressed, any failure to manually 1 i start is a reportable event, and such information should be relayed to j the appropriate personnel so that a report can be initiated. t i
- 4. Operator trainint will be enhanced during the next training cycle to i
advise personnel that a DG start should occur without having to
- continue depressing the manual start p.shbutton.
j pl.TheDG18andIAtestfrequencyiscurrentlyonceper7daysin accordance with T5 Table 4.8-1. This frequency will be continued until 7 consecutive valid tests are completed with no more than one valid failure in the last 20 valid tests and/or no more than 4 valid failures in the last 100 valid tests. Up to and including the 7-5-90 L i valid failure, there have been a total of 4 valid fa11eres in 79 valid tests of DG18. Up to and including the 7-11-90 valid failure, there j , have been a total of 5 valid failures in 43 valid tests of DG2A. I _ , __
interoffice Correspondence GeorgiaPoner A J l August 3, 1990 Memo To: George Sockhold, Jr. ) General Manager Nuclear Plant-Vogtle l 1 j
Subject:
Vogtle Electric Generating Plant - Units 1 & 2 QA Audit of Surveillance Program / Technical Specification Compliance OP09-90/31 4 File: X78G17-P-OP09 ) Log No: VSAER-90-186 l 4 Audit Scone: The purpose of this audit was to verify comp 11ance with, and the { effectiveness of, the Plant Vogtle Quality Assurance Program as applied to j Surveillance Program / Technical Specification Compliance. The scope of this audit included a review of Technical Specification surveillances covering power distribution limits, electrical power systems, limiting safety system i settings, reactivity control systems, and instrumentation. Su ry of Problent Found: l i o Procedure 88075-C inadequately implemented Technical Specification surveillance 4.2.5.3 requirements that instrumentation used for the ' precision heat balance calculation must be calibrated within seven days prior to the surveillance. Therefore, the Unit I surveillance was improperly performed prior to operation above 75% rated thermal power following the Unit I second refueling outage. (AFR #432) o Some corrective actions for the event of March 20 1990, in which loss of offsite power lead to a site area emergency, were, not implemented in an
- effective and timely manner. (AFR #433)
- Evaluation
- With the exception of the Unit 1 precision heat balance i calculation, surveillance tasks reviewed had been performed within the specified time periods, had received required reviews and approvals, and had
- mat.approprlata ******=== c.r.1tarit. Tha Tarhaira1 Spac.ifh unssaM1anca which was improperly performed was caused by procedural inadequacy. However, i
the results of this audit, based on sample size, do not indicate a j programmatic weakness concerning the technical adequacy of implementing procedures for surveillances. ! During this audit, ladication of inadequate corrective action was also ! identified. The diesel generator 18 (7/5/90) and 2A (4/12/90) failures are examples that effective and timely corrective actions from the event of March 20, 1990 were not taken. These examples indicate that additional management ! nttentiear map be . required end that other corrective actions from the event . 4 reviewed for adequacy of implementation. i
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QA Audit of Sur* e111ance Program / Technical Spectffcat1on Compifance
, OP09-90/31 Page 2 Action: In accordance with ANSI N45.2.12. you are requested to respond to the '
attached audit report no later than September 2, 1990.
/ h4l G. R. Frederick
. Supervisor - SAER I
CMB/GRF/btp Attachment xc: R. P. Mcdonald J. H. McGowan W. G. Hairston, III H. D. Drinkard ; C. K. McCoy Q. A. File O. M. Fraser J. E. Swartzwelder H. M. Handfinger M. W. Norton W. F. Kitchens T. V. Greene M. J. Aj1unt G. A. McCarley ' W. E. Mundy NORMS J. G. Aufdenkampe Jr. (Org. AFR's #432 and #433)
)
l l I l f f 4 E e 6
g Plant Vogtle - Units 1 & 2 ; ACTIVITY QA Audit of Surveillance Program / Technical Specification Compilance AUDIT NO. i OP09-90/31 DATES AUDITED June 28 through July 12, 1990 i AUDITOR C. M. Burke. Senior QA Fleid Representative (Audit Team Leader) PRE-AUDIT POST-AUDIT CONTACTS CONFERENCE EDil CONFERENCE { G. R. Frederick x x x T. V. Greene x d J. A. Rodgers x x x , M. E. Mundy x i J. R. Petro x
, H. M. Handfinger x x 4 J. C. Milliams x J. E. Swartzwelder x x R. L. LeGrand x D. R. Lee x
. J. S. Bowden x x l' J. G. Aufdenkampe x N. F. Kitchens x M. L. Hobbs x x T. D. Gentry x
- 0. D. Hayes x
- S. A. Bradley x x M. L. Burmeister x
, C. L. Coursey x M. S. Briney x T. L. Mendt x D. R. Christiansen x i C. A. Griffin x 1 9. Sw hMtaher s J. M. Grandy x , T. G. Lamb x M. E. McGrath x J. E. Bowles x C. H. Milliams x M. C. Henry x T. L. Morris x R. K. Pope x G. A. Ovellette x
- 3. M. Sowsevat . -
x .
1 OA T.. ;_Audit,o,f
... Surveillance. Program
#.a_ c. / . . .
. OP09-90/31 Page 2 R. M. Odom x
; A. Rickman x I R. P. Farrow (Nisco) x
! R. M. Smith (Nisco) x
! T. L. Hillis x i J. J. Godbee x l J. Redding x REEEBENCES l Vogtle Electric Generating Plant (VEGP) Units I and 2 Technical Specif1. cations I PROCEDURE REVISION DESCRIPTION j 11885-C 13 Diesel Generator Operating Log 13145-1 22 Diesel Generators i 14230-1 4 AC Source Verification :
Power Range Reactor Trip Interlocks 18 Month ACOT i 14925-1 3 j [ Analog Channel Operational Test) , 14940-1 9 Estimated Critical Condition Calculation l 1 14940-2 3 Estimated Critical Condition Calculation I l 14980-1 20 Diesel Generator Operability Test i 24525-1 7 Pressurizer Pressure Protection Channel I 1P-455 Analog Channel Operational Test and Channel Calibration 24553-1 3 Turbine Trip - Reactor Trip Hydraulic Pressure IP-6161 Analog Channel Operational Test and Channel Calibration 24700-1 22 Nuclear Instrumentation System Power Range Channel
- IN41 Channel Calibration j 24782-1 9 Reactor Coolant Flow Loop 1 Protection Channel I IF-414 Analog Channel Operational Test and Channel i
Calibration i 24783-1 8 Reactor Coolant Flow Loop 2 Protection Channel I
- 1F-424 Analog Channel Operational Test and Channel i Calibration -
- 24784-1 8 Reactor Coolant Flow Loop 3 Protection Channel I
! 1F-434 Analog Channel Operational Test and Channel i Calibration . 24810-1 13 Delta T/T AVG Loop 1 Protection Channel I IT-411
- Analog Channel Operational Test and Channel Calibration
! 88003-C 0 Shutdown Margin by Minimum Bank Height ! 88013-C 1 Overall Core Reactivity Balance F. 28044 1F Ases;tes Cae&eek Spetem>FleeMosewemose t 88023-C 3 One Point Incore/Excore Detector Calibration 88025-2 0 Determination of Movable Incore Detector Operating i Voltages
- 88075-C 1 Precision Heat Balance 4
l PURPOSE / SCOPE The purpose of this audit was to verify compliance with, and the effectiveness l of, the Plant Vogtle Quality Assurance Program as applied to Surveillance o vv. .-. mimkal Specification Compliance. The scope of this auditatnetodet-a review of power distribution limits, electrical power systems, limiting safety system settings, reactivity control systems, and instrumentation. i
i OA Audit of S rveillance Program / Technfcal Spectffest'forr Comp Mince
-' j 4 .
j . .0P09-90/31 Page 3 i EVALUATION I Observation of Technical Specification surveillances being performed noted compliance with procedures and that personnel were knowledgeable of the surveillance requirements. With the exception of the Unit 1 precision heat i balance calculation, surveillance tasks reviewed had been performed within the i specified time periods, had received required reviews and approvals, and had met appropriate acceptance criteria. The Technical Specification surveillance j which was improperly performed was caused by procedural inadequacy. However, the r.esul.ts of thin audit, based on sample size, do not indicate a i j programmatic weakness concerning the technical adequacy of implementing ; procedures for surveillances. The diesel generator 18 and 2A failures are examples of inadequate corrective actions from the event of March 20, 1990. These examples indicated that attention to detail is required in the area of corrective actions and that other corrective actions from the event should be reviewed for adequacy and implementation. l } AUDIT DETAILS l 7 Power Distribution Limits A. Requirement VEGP Units 1 and 2 Technical Specifications, section 4.2.5.3 states that after each fuel loading, the Reactor Coolant System (RCS) flow . rate shall be determined by precision heat balance prior to The RCS flow rate l operation above 751 rated thermal power (RTP). i shall also be determined by precision heat balance at least once j per 18 months. Within seven days prior to performing the precision l heat balance flow measurement, the instrumentation used for i l ! performing the precision heat balance shall be calibrated. B. Results 7 ! This surveillance requirement is satisfied by performance of procedures 88014-C and 88075-C. Review of completed documentation for tasks 88014-102, 88014-103, and 88075-101 noted that the ' surveillances had been performed following the Unit i second
, .T..L, oeggyer. However', .14a eP v.m..e.m 8085P5.@ mest ' .hat -
- the procedure addressed calibration of special test instrumentation i but did not address plant instrumentation. Discussions with the Reactor Engineering Supervisor and review of completed 1-TE-15201, procedure 88075-C noted that plant instruments 1-TE-15200, 1-TE-15202, and 1-TE-15203 were used for feedwater temperature input for the precision heat balance calculation.
Discussions with the Instruments and Controls Superintendent noted j that the instruments are calibrated on a six months basis and were calibrated on 1/23/90. Additional review noted that Maintenance Work Order Op40) 1-90-02215 calibrated the instruments on 4/28/90. However, the precision heat balance calculation and the RCS flow measurement were performed on 4/23/90 at 2330 CDT. Also, 757. RTP
1 . OA Audit ef Surveillance Pr ram / Tectmtraf 5,..;G& Ctaqr - OP09-90/31 Page 4 was achieved et approximately 0130 CDT on 4/24/90. Therefore, surveillance 4.2.5.3 was incorrectly performed because procedure 88075-C did not require calibration of plant instrumentation used for feedwater temperature input for the precision heat balance calculation. This discrepancy v111 be included in Audit Findina Renert OP09-90/31 #432 After the post-audit conference, Deficiency Card #1-90-293 was initiated by Reactor Engineering personnel to facilitate the reportability review of this ites. C. Requirement . VEGP Units 1 and 2 Technical Specifications, section 4.2.5.2 requires the RCS flow rate indicators to be subjected to channel calibration at each fuel loading and at least once per 18 months. D. Results Review of completed documents for surveillance tasks 24782-101, 24783-101, 24784-101, and 24785-101 noted that the 18 months surveillances were performed with acceptable results and within the required time limits during the Unit I second refueling outage. No problems were noted. II. Electrical Power Systems A. Requirement VEGP Units 1 &2 Technical Specifications, section 4.8.1.1.2.a requires each diesel generator to be demonstrated operable in
- accordance with the frequency specified in Table 4.8-1 on a
- staggered test basis by verifying seven specific items.
i 8. Results The auditor observed performance of surveillance procedure 14980-1 , for the Unit 1 Train 8 diesel generator operability test. This test is normally performed monthly, but is being performed weekly ! as required by Table 4.8-1 due to the number of recent test failures. In the Unit 1 Train 8 Diesel Generator Building, the ! avdHer veMW& the& tb fo44ew6ey c d.;; be6ay oso&were tts . . current revision: 11885-C. 13145-1, and 14980-1. Additionally, the auditor observed performance of the cylinder moisture check, an independent verification of the cylinder moisture check, the diesel i generator air start compressor test, and completion of the Diesel Generator Operating Log. In the Unit 1 Control Room, the auditor observed loading of the diesel generator and the Diesel Generator Fuel 011 Transfer System test. Additionally, the auditor verified that the stop watches, i required by step 4.2.a. were currently calibrated. Upon completion - of the test, the auditor verified that the acceptance criteria of
~
section 6.0 had been met. 4 J
l
. l 04 A::d1t cf Surveillance Program /
Tecturtert ^ ^rT^ n'nm Congritancv- -
., OP09 90/31 Page 5 During performance of the test, the diesel generator failed to start on the first attempt. Discussions with an Operations Shift Superintendent and a Shift . Supervisor noted that on the first attempt a Reactor Operator trainee had not held the diesel generator start push button for sufficient time to start the diesel. The failure was attributed to operating error and was not considered to be a valid test failure as allowed by Table 1 of procedure 14980-1. Alignments were reverified by Operations personnel, and on the second attempt the diesel generator started.
No problems were noted with the diesel generator operability test. NOTE: Because of a similar problem identified with the Unit 2 Train A diesel generator on 7/11/90 (VEGP Event Report
#2-90-005), this audit detail was brought to the attention of plant management during the write-up of this report on 7/17/90. Immediste corrective action was initiated by the General Manager Nuclear - Plant Vogtle and a special test 0 440 1-90-03340) was conducted on 7/18/90 for the Unit 1 Train B diesel generator. This test, witnessed by the auditor, identified that ten of sixteen air pilot valves on the cylinders failed when subjected to 100 psig air pressure.
Additional review by the auditor noted that the 7/5/90 18 diesel generator failure and associated pertinent alarms and indications were not documented in the Unit 1 Shift Supervisor Log, the Unit 1 Control Log, or in the
- technical specification surveillance package for task
- #14980-102-25185. This event should have been identified l as a failure to start and should have been logged i accordingly. Also, a 4/12/90 2A diesel generator fatture i - for the same reasons as the IB failure - was not
! critiqued until after the similar failure on 7/11/90 for !
- diesel generator 2A. These discrepancies indicate that:
) o The requirement to conduct an event critique for each diesel failure until plant management decides critiques are no longer required (commitment 18758), and I ! o The requirement to revise or develop a procedura/ policy outlining guidelines for logging 4 pertinent alarms and indications to assist in ! evaluation of equipment or system malfunctions j (commitment 18756) i were not implemented in an effective and timely manner i
- for the event of 3/20/90, in which loss of offsite power l
1ead to a site area emergency. Audit Findina Reoort OP09-90/31. #433 will be issued to track -corrective actions for this item.
~
OA Audit of Surveillance Program /
. TechnicaT Specificaticn Compliance OP09-90/31 Page 6 C. Requirement VEGP Units I and 2 Technical Speci fications, section 4.8.1.2 requires the following AC electrical power sources to be demonstrated operable during modes 5 and 6 by the performance of each of the requirements of Specifications 4.8.1.1.1 4.8.1.1.2 (except for specification 4.8.1.1.2a.5), and 4.8.1.1.3:
- 1. One circuit between the offsite transe~.ssion network and the Quita Class IE Distribution System, and .
- 2. One diesel generator with:
o a day tank containing at least 650 gallons of fuel, o a fuel storage system containing at least 68.000 gallons of fuel, and o a fuel transfer pump. , D. Results ! Specification 4.8.1.1.1 is met by the weekly performance' of
- procedure 14230-1. Specification 4.8.1.1.3 is satisfied by reports j as applicable. Fcr Spectfication 4.8.1.1.2. the auditor verified
! performance of surveillance task 14980-103 for the Unit 1 Train A . diesel generator. In Document Control, the auditor reviewed l completed procedure 14230-1 data sheets for the period of 3/13/90 . through 6/20/90 and noted that acceptance criteria had been met. l Also, the surveillances were performed within the required time ' limit (weekly since 4/9/90) and had been properly approved. No problems were noted.
- III. Limiting Safety System Settings
! A. Requirement VEGP Units 1 and 2 Technical Specifications, section 2.2.1 requires i the Reactor Trip System Instrumentation and Interlock Setpoints to ce" set comrtstvert wttfr effir t'rtir myw6 vnfuer smnrrr 'rr Table 2.2-1.
- 8. Results
, The auditor verified by calculation that the setpoint and allowable values for eleven of twenty-eight functional units (overtemperature
- delta T. overpower delta T, power range neutron flux, etc.) from Table 2.2-1 were accurately incorporated into Instruments and Controls (I&C) procedures 24700-1, 24525-1, and 24553-1 for channel '
- calibration and analog channel operational tests (ACOT's).
Additionally. the auditor verified that setpoint and allowable values for four functional units were properly incorporated into i Operations procedure 14925-1 for the eighteen months ACOT of the t
- power range reactor trip interlocks. No problems were noted. ,
i
OA Audit of Surveillance Progras/ Tecantes ? w. 1 r n..u ver t"ompifancy :
,' OP09-90/31 Page 7 ,
i IV. Reactivity Control Systems 1 A. Requirement I VEGP Units 1 and 2 Technical Specifications, section 4.1.1.1.1.c requires the shutdown margin to be determined to be greater than or equal to 1.3% delta k/k with K,ff less than 1, within 4 hours . prior to achieving reactor criticality by verifying that the ' predicted critical control rod position is within the limits of Specification 3.1.3.6. , B. Results The auditor determined the dates of the Units 1 and 2 reactor trips and reviewed completed 14940-1 ano 14940-2 procedures in Document Control. The review noted that an estimated critical condition ' calculation utilizing either an estimated critical boron l concentration (ECC) calculation or an estimated critical position l (ECP) calculation had been performed prior to the reactors achieving criticality after the trips. No problems were noted. C. Requirement VEGP Units 1 and 2 Technical Specifications, section 4.1.1.1.1.d requires the shutdown margin to be determined to be greater than or equal to 1.3% delta k/k prior to initial operation above 5% rated thermal power (RTP) after each fuel loading, by consideration of l six factors, with the control banks at the maximum insertion limit of Specification 3.1.3.6. D. Results Review of completed documents for surveillance task 88003-101 noted l that the surveillance had been performed prior to operation above ! 51 RTP after the second refueling of Unit 1. No problems were I noted. I E. Requirement
- '/EEF N b ant. 3 Testup6eek L M ..e h , 6 M .1.1.2 requires the overall core reactivity balance to be compared to predicted values to demonstrate agreement within i 11 delta k/k at least once per 31 effective full power days (EFPD). The l predicted reactivity values shall be adjusted (normalized) to !
correspond to the actual core conditions prior to exceeding a fuel ; burnup of 60 EFP0 after each fuel loading. F. Results i The auditor reviewed completed documentation for surveillance tasks i
#88013-201 performed on Unit 2 since January 1,1990. The review verified that the seven calculations had been performed at least once per 31 EFP0 and that appropriate acceptance criteria had been met. No problems were noted.
j '. 04 Audit of Surveillance Program / Tecfmfcar SpectfTeaffen Cbspiraner i . OP09-90/31 Page 8 l Concerning normalization of predicted reactivity values after fuel 4 loading, the auditor verified that procedure 88013-C had been 1 performed satisfactority for Unit 1 prior to 60 EFPD after the j second refueling. No problems were noted. V. Instrumentation j A. Require ent VEGP Units 1 and 2 Technical Specifications, section 4.3.1.1
- requires each Reactor Trip System instrumentation channel and i interlock and the automatic trip logic to be demonstrated operable i
j by the performance of the Reactor Trip System instrumentation 4 surveillance requirements specified in Table 4.3-1. I B. Results
- The auditor reviewed completed documentation for twelve
- surveillance tasks and noted that the surveillances for the
! applicable functional items in Table 4.3-i had been performed i within the required time limits and with acceptable results. No
- prohless were noted.
? ! Additionally, the auditor observed performance of surveillance task l
- 24810-102 for the 31 days Analog Channel Operational Test
! surveillance requirement for Table 4.3-1 item 7 (overtemperature i delta T) and item 8 (overpower delta T). Test equipment was ) i currently calibrated and documents used were noted to be current. j Quality Control hold points were noted to be honored and performed. Also, the surveillance was performed in a professional manner by all involved personnel. No problems were noted. J i Also, the auditor accompanied Reactor Engineering personnel to the Unit 2 Control Room to observe performance of surveillance tasks j 88023-201 and 88023-202. The auditor observed the Reactor Engineer obtain clearance 2-90-10046 and perform procedure 88025-2 for ! Nestinghouse incore instrumentation detectors A. B, C. D. & E as a i prerequisite to performance of procedure 88023-C. However, the surveillance was postponed after a Reactor Engineer noted that the z reh um neb at e conytte power dueWP m tM sevud l ) days earlier. For the activities observed, no problems were noted. l ! OPEN ITDt1 From previous audits: None were reviewed. J l From this audit: I
- OP09-90/31 Plant instrumentation used for feedwater temperature 43Z input in the Uhtt 1 precision heat balance caleviatierr-was not calibrated within seven days prior to the i surveillance as required by Technical Specification i
, 4.2.5.3.
l QA Audit cf Surveillance Program / Tectattar Sperffftaffen CBapitsucy OP09-90/31
. Page 9 !
I I i I OP09-90/31 Some corrective actions for the event of March 20, 433 1990, in which loss of offsite power lead to a site area emergency, were not implemented in an effective i and timely manner. POST-AUDIT CDNFERENCE A post-audit conference was held on July 12, 1990. The audit results were i presented, discussed, acknowledged, and agreed upon by those attending the setting-C. M. Burke
/h b G. R. Frederick Senior QA Field Representative Supervisor - SAER l
l I l
i f,' ORTGINAI. Audit Finding Report Trend Code: P05D Safety Audit and Engineering RevisW Responsible Party - J. G. Aufdenkampe MR No, Trocar No, OP09-90/31 #432 C n.e nem.o erw.u. nnen N.. VSAER-90-186
} Company /oreenization i Georgia Power Company QA Audit of Surveillance ; Proj.c t/Ac tMey i Plant Vogtle - Units 1 & 2. Program / Technical Specification Compliance I
Auditor (s) C. H. Burke, Senior QA Field Representative (ATL) June 28-July 12. 1990 signatur.: [ d l[ u oste: 8-3-9o j REFERENCE / REQUIREMENTS VEGP Units 1 and 2 Technical Specification surveillance requirement 4.2.5.3, "After each fuel loading, the RCS [ Reactor Coolant System] flow rate shall be determined by precision heat balance prior to operation above 757. RATED THERMAL POWER. The RCS flow
- rate shall also be determined by precision heat balance at least once per 18 months.
Hithin 7 days prior to performing the precision heat balance flow measurement, the instrumentation used for performing the precision heat balance shall be calibrated. The provisions of 4.0.4 are not applicable for performing the precision heat balance flow j measurement." 4 flHDlHG Review of completed Technical Specification surveillance procedure 88075-C, " Precision Heat Balance," and discussions with plant personnel noted that plant instruments 1-TE-15200,1-TE-15201,1-TE-15202, and 1-TE-15203 used for feedwater temperature input . for the Unit I precision heat balance were not calibrated within seven days prior to i performance of the surveillance. Therefore, procedure 88075-C inadequately implements I i Technical Specification 4.2.5.3 requirements and the surveillance was improperly performed.
- RECOM4 ENDED ITEMS TO BE CONSIDERED IN CDRRECTIVE ACTION I
- 1. Investigate the problem and any potential similar conditions.
- 2. State actions taken to resolve the specific and similar problems.
- 3. Determine root cause.
- 4. State actions taken to prevent recurrence.
CDRRECTIVE ACTION / REVIEW VERIFICATION 4 l 1 4 Sinnature: Date:
,g /
)i ORTGINAL I* Audit Finding Report Trend Code: P168
) Safety Audit and Ensincerina Review R. on.ibi. p.ag : G 80ckhold. Jr.
ON9-90/31#433 C n= en w % ,y ne ne me. VS M M 186
; compenv/oreenlannen j Georgia Power Company QA Audit of Surveillance l Plant Vogtle - Units 1 & 2. Program / Technical Specification Compliance j Austarte) --
j C. M. Burke, Senior QA Field Representative (ATL) June 28-July 12, 1990
- .. w .- 6/ad A && _- .. B-3-90 RU 3 dNCE/RE(1]IREMENTS
- 1. Title 10 Code of Federal Regulations (CFR) 50, Appendix 8. Criterion XVI/" Measures shall be established to assure that conditions adverse to quality, such as failures, i malfunctions, deficiencies, deviations, defective material and equipment, and i nonconformances are promptly identified and corrected. In the case of significant
! conditions adverse to quality, the measures shall assure that the cause of the i condition is determined and corrective action taken to preclude repetition." ]
- 2. Open Item Tracking System #18756 indicated that a procedure / policy would be revised
! ::r d:veloped outlining guidelines for logging pertinent alares and indications to assist in evaluation of equipment or system malfunctions. It also indciated that this data would be emphasized in training and by appropriate management attention. i 3. Open Item Tracking System #18758 indicated that the trend program would be reviewed i for adequacy of coverage of diesel generator component failures and that event j critiques would be required for diesel generator failures. I FINDING i Effective and timely corrective actions for the event of 3/20/90, in which loss of offsito power lead to a site area emergency, were not adequately implemented. The following are examples: o Finding #1 (corresponds to requirements #1 & #2) On 7/5/90, the Unit 1 Train 8 diesel failed to start. However, this failure was not ' documented in the Unit 1 Shift Supervisor Log, the Unit 1 Control Log, or in the associated Technical Specification surveillance package for surveillance task j #14980-102-25185. i j o Finding #2 (corresponds to requirements #1 & #3) The 4/12/90 2A 01esel Generator failure was not critiqued untti after the 7/11/90 ] failure of diesel generator 2A. The IB diesel generator failure of 7/5/90 was not critiqued untti after management was made aware of the failure on 7/17/90. REC 304 ENDED ITEMS TO BE CONSIDERED IN CDRRECTIVE ACTION
- 1. Investigate the problem and any potential similar conditions.
1 2. State actions taken to resolve the specific and similar problems.
- 3. Qatsretas rest <a us ,
- 4. State actions taken to prevent recurrence.
i mine p.,.i.i_2.c.
-- . _ __ _. . - - - - - _ - - L_ ____ _ _ _ _ - .___ . -
RIGINAL dudit rimang neport conen.
* % 9-90/31 #433 VSAER-90-186 MAfCTIVE ACTION /REVIDi VERIFICATION 51cnature: Date I
i i I4# 1 Page e of
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- w. o n..eca. n:
s.,.... e ,. - rec <a cm .- CLV-018til 0233 Ducket Nos. 50-424 50-425 U. S. Nuclear Regulatory Commission Washington, D. C. 20555 ATTN: James M. Taylor Executive Director for Operations Gentica.en. V0GTLE ELECTRIC CENERATING PLANT CO W NTS ON NUREG-1410 Your letter of June S,1990 transmitted a copy of the NRC Incident Investigating
- Team (!!T) report (NvREG-1410) and stated that any comments should be provided by July 9, 1990. We believe that the !!T performed a thorough investigation of the event and '. hat the report in general represents an accurate d'!scription of the circumstan
- es surrounding the $tte Area Emergency on March 20, 1990. We have reviewed the document and specific comnents are attached.
Sincerely, A! ). U : N-~2ii~~ W. G. Hairston, Ill WGH,lil/HWM/gm Att achttent xc: Georcia power Cor.cany Mr. C. K. McCoy Mr. G. Bockhold, Jr. Mr. R. M. Odom Mr. P. D. Rushton NORMS , k. S. Nuclear Reaulatory Commission Mr. S. D. Ebneter, Regional Administrator Mr. T. A. Reed, Licensing Project Manager, NRR Mr. B. R. Bonser, Senior Resident inspector, Vogtle
-f000/G vC/E l W
y
- j. .L-10 *:-0 c~t;E :::!Ct G C., .cG ,E E. .;:; ,Og-r . ;3e .:: g _; :c ,
g i e enc:.05URE i GmEf4T.LQft NUttEG-1410
- 1. P l-5 suta v.htt tf.e risk uf core damage is greater from m'id-loop 5 optrat u. t.rai d'.ie to a fuel handling accidert; no basSs to support this assunptien is provided.
- 2. D. 2-3 states erecceously that " spec.at containment building penetrations had been constructe: in anticipation of the need to quickly close the contcinment bu11 ding curing an outage." Iney iere in fact constructed to permit SG ludge lancing with the equipment hatch closed.
- 3. P. 3-11 (3.2.5) needs to ret.utjnize that an emergency Start of the diesel generators at VEGP has teen changed to include starting on loss of offsite povier.
l 4 P. 5-22 (5.5.2.2) states that maintena' ice personnel stivald te trsined in I nic-loop ecerations. This has not been determined to be appropriate for l reQu1reo maintenance training through systematic analysis of job tasks. !
- 5. P. 7-5 anc Appendix K (p. k-8) (Also, on page 7-7 fur the first refueling outage) !ist several *nonconservative conditions" that existed in the plant, with the wri1 cation that, in retrospect, these conditions should not have existed. GPC contends that sufficient risk analysis dces not exist to show
'hJt the Ilit is r.on-conserve:tve.
- 6. P.10-3 (Cond usion 10.4.1) an: J-28 state that foreign material in the DG lar et v.a'..!* terperature switches is considered the most likely cause of the I M te : 5 :cring tne event. GPC concure that internal ccntamination is the o:t likely cause of one switch tripping. The second switch, although it had it.terral foreign material, was also set low as a rt:sult of inade4Jate
- alibrition procedures. SubseQaent additional testing has shown that these su tches are temperature sensitive, requiring that consistent calibration
'9thniQves tie used to achieve the desired setting (reference 1.ER alai!9 W-005, revision 1). Therefore, GPC has con:1uded that the primary
-s o of tne intermittent actuation of the jacket water temperature switches
.. 4nacecote calibration in conjunction with foreign material.
1 F. ' 8 (Ccnclusion 10.7.3) states that Vogtle had a high number of failures c f *. : sensors compared to tile rest of the industry. From Appendix 1 (p.1-8) of tu report wnich itemizes these failures, a large percentage of the pret @ s were calibration setpoint out of specification during construction accept:nce testirg. Out of calibration in this application is not typically count?c as a failure by either Vogtle or other plants in accordance with Ml;5 reporting criteria; hence the NRC conclusion that Vogtle has a higher i n Woer of switch failures is not based on comparable data.
- 8. ' P. K-2, states erroneeusly that future SG eddy current testing will be cenducted on only 20 percent of the tubes. VEGP currently plans to follow the EPRI guiceltnes which require a minimum of 20 percent of the tubes to be testad as a base' scope, plus an augmented sample of additional tubes for any
.ctIve da. mage mechanisne.
t
1600 8-1 6-90 u LER 1-90-004 1 og Original NRC Concern To detemine if Technical Specification 3.0.4. was violated. when Unit 1 entered Mode 6 from Mode 5 while Source Range IN31 ms out of service for an 18 monta calibration, solely for the purpose of progress on the ; critical path schedule; if the shift was subsequently congratulated for making that progress; and if the Shift Superintendent demonstrated a willingness to violate Technical Specifications for the sake of schedule. Found to be unsubstantiated NRC Concern The inadequacy of the root cause detemination and corrective actions cf LER 1-90-004, in that, human factor problems involving the LCO sheet may have contributed to the Shift Superintendent's failure to note the LCO mode change restriction.
/^/AD6ID (M 7F C044 cn t/6 A8cT704 NRC Doctanentation -/ar Mf g M CM Technical Specification 3.9.2 Deficiency Card 1-90-0050 LER 1-90-004 12007-C. Refueline Entry (Mode 5 to Mode 6)
! Unit 1 Control Log ! Unit 1 Shift Supervisor Log (2/28/90 and 3/1/90) ! LCO Status Sheet 1-90-152 LCO Log (10008-C. P.8 of 11. dated 2/28/90) 14000-1 Operations Shift and Daily Surveillance Loos, dated 2/28/90 , IR2 Outage Schedule (actual vs. schedule) Turnover Checklist (11870-C. dated 2/28/90 Completed Procedures, dated 2/28/90 - 12007-C. 14000-1,11871-C and 11872-C VEGP Position VEGP's position is that human factor problems with the LCO sheet was not a significant contributing causal factor in this event. However, due to a number of htsnan factor concerns noted during the IR2 refueling outage. VEGP has revised procedure 10008-C twice, to enhance' usability and htsnan factoring. Furthemore. VEGP will review Procedum 10008-C to detemine if further enhancements are warranted. ;
, v[7 l
I
l c 1600 8 90 LER 1-90-004 4 VEGP Doceentation l LER 1-90-004 ' 12007-C. Refueline Entry (Mode 5 to Mode 6) Unit 1 Control Log, 2/27/90 to 3/2/90 i Unit 1 Shift Supervisor Log, 2/27/90 to 3/2/90 l Unit SS Relief Checklists, 2/28/90 and 3/1/90 Support SS Relief Checklists 2/28/90 and 3/1/90 i Operations Supervisor Reitef Checklist, 2/28/90 and 3/1/90 l R0 Relief Checklists 2/28/90 and 3/1/90 BOP Relief Checklists, 2/28/90 and 3/1/90 14000-1, Operations Shift and Daily Surveillance Loos, 2/28/90 : 1R2 Outage Schedule, 2/23/90 thru 3/3/90 ' 10008-C. Recording Limitino conditions for Operation, Rev.12 l 1 4 h l
. l 08/16/90 UNIT 2 LER 90-001 1600 PAGE 1 0F 2 MISSED SURVEILLANCE ON
" CONTAINMENT INTEGRITY VERIFICATION VALVES OUTSIDE CONTAINMENT" NRC CONCERN: Required Tech Spec actions may have been delayed by initiating an investigation. Was management pressure a contributing factor?
l Concern: Potential concealment of correct Tech Spec LCO ! entry time to prevent a forced shutdown and immediate notification of the NRC. Finding: The correct T.S. LCO entry was not concealed. I l Concern: Cause for confusion over the Surveillance Task Sheet. Finding: The cause for the confusion was an inconsistent use of equipment identification numbers on these sheets. Although corrective actions have been taken to eliminate this inconsistency, potential for confusion still exists.
- Concern: Extent of emphasis on keeping the plant in
! operation and limiting NRC notifications. i Finding: There was no indication of unreasonable emphasis on keeping the plant in operation or limiting NRC ; notifications. ' NRC DOCUMENTATION: D.C. 2-90-022 Surveillance 14475-201 Jan 3,1990, Feb 1,1990, Feb 28, 1990 Unit !! LER 90-001 Control Room Logs from Feb 27 and Feb 28, 1990
- NRC Inspection Report 90-10
l . 08/16/90 ! , 1600 l UNIT 2 LER 90-001 l PAGE 2 0F 2 VEGP POSITION: Management does not apply pressure to delay action statement entry. Investigations are only for the purpose of determining if a problem exists. Timely resolution of problems is required. Suspected problems are promptly reviewed to detemine if a problem exists. LER 90-001 gives details of event, the cause and corrective actions. The SS acted to instruct the personnel to complete i all valves on the procedure and concurrently notified Surveillance Tracking. Surveillance Tracking went to Document Control and concluded we had a missed , Surveillance. Surveillance Tracking then contacted i the SS and notified him of the discovery and I initiated D.C. 2-90-022 in accordance with Procedure I 00150-C. The personnel involved acted properly. j l l VEGP DOCUMENTATION: Same as NRC Documentation l Task 14475-201 Verification Sheets l i
1600 8-16-90 RNR PtMP 18 VIBRATION 8llC
~
g Original NRC Concern A non-conservative decision was made concerning the operability of the IB RHR ptsnp in order to avoid substantial impact to the outage critical path schedule j Found to be unsubstantiated 4 NRC Concern - A Deficiency Card was not generated in a timely fashion concerning the IB RHR pump cooler leak and elevated vibration levels. , j i CJL A ricn (: d au Jil G 720 Wfl ' Tli OC NRC Doctanentation Unknown e l VEGP Position VEGP concurs that a Deficiency Carti was not generated in a timely fashion.
- Since the occurrence of this event, VEGP management has taken positive action to improve the effectiveness of the Deficicacy Ca rd Program. These improvements include:
- 1. Revision of Reactor Trip Review Procedure,10006-C to specifically require a sign-off indicating a Deficiency Card has been written.
- 2. Address by General Manager to the PRB stressing the necessity for timely Deficiency Card generation, and memo to all managers from the Technical Support Manager stressing the requirements for timely submittal of Deficiency Cards.
In addition, the Deficiency Card Program has received incitased management attention and oversight to ensure Deficiency Cards are generated in a timely ; fashion. This will ensure that operability and reportability deteminations 1 and appropriate engineering ev4luations art perfonned. YEGP Doctsnentation Letter. Manager. Technical Support to Department Managers dated 6/22/90, i i
1600 8-1 6-90
' ~
IMPROPER TCP PROCESSING i j NRC Concern - TCP 18028-C-7-90-1 was "back-dated" to avoid violating section 6.7.3.c of Technical Specifications. NRC Doctanentation 4 TCP 18028-C-7-90-1 DC 1-90-282 DC 1- W 283 PR8 Minutes 90-81 and W82 4 Procedurt 00052-C i. YEGP Position 4 A violation of Section 6.7.3.c of the VEGP Technical Specifications i occurred. However, the cover sheet of TCP 18028-C-7-W1 was not dated l 6-12-90 to avoid this violation.
, TCP 18028-C-7-90-1 (written against Rev. 7 of the permanent plant procedure). Loss of Instrianent Air, was approved by the Operations Manager
- on 5 90. On 6-8-90 the PR8 tabled this TCP to allow the Operations
- Depa rtment to determine if additional instructions for Modes 3 4, 5, i and 6 should be added to the revision to strengthen the AOP. Revision
- 8 of the permanent procedum was prepared by Operations and approved by
- the PR8 on 6-12-90. This revision addrissed both TCP 18028-C-7-W1 and additional instructions for a Loss of Instrtsnent Air in Modes 3, 4, 5 i and 6. The Acting Operations Manager understood that the TCP would not i be used in the field once Rev. 8 was issued. Upon approval of Rev. 8 of the permanent procedure by the PRB. verbal instructions were given by the Acting Operations Manager to the procedure coordinator to void TCP 18028-C+ 7-90-1. The TCP was next in the procedure coortlinator's
, possession on 6-15-90. On that date the acting Operations Manager signed l the TCP cover sheet and dated it 6-12-90 to reflect his understanding,
- based on discussions with the procedure coordinator and his verbal instructions of 6-12-90, that the TCP was voided on 6-12-90.
The Acting Operations Manager assumed that the approval of Rev. 8 , of the permanent procedure (which he asstaned occurred on 6-12-90) resulted , in the voiding of the TCP, and that his vertal instruction to the Operations i staff was adequate to close-out required paperwork. This was an error 1 and resulted in a failure to comply with Procedurt 00052-C. Section 4.6.2 in a timely manner. GPC notes that the minutes of PR8 meeting 90-82 indicate the TCP "was voided" on 6-12-90 which reflected the understanding of the Acting Operations Manager.
1600 8-16-90 u IFROPER TCP PROCESSING Ckll On 6-18-90, the Operations Manager instructed the Acting Operations Manager to write a DC on the inaccurate dating of the TCP close-out sheet and a failure to meet the 14 day period set under Procedure 00052-C Section 3.2.4. This instruction was a result of nonnal Technical Support Group review and verification close-out of TCPs. On 6-22-90 two DCs were written by Technical Support. On the same date the (former) Acting Operations Manager, in preparing a DC on the inaccurate dating of the TCP cover sheet, detennined that the pennanent procedurt. Rev. 8, was not issued until 6-13-90, that the TCP was pulled from the Control Room on 6-13-90 and that the 14 day limit under Procedure 00052-C had been exceeded. i
8-16-90 1600 SEQUENCER IN0PERASILITY QC NRC Concern Inadequate information exists for shif t personnel to detemine which Technical Specification to apply if the seqancer is inope rable. In addition, previous verbal guidance was inadequate. NRC Docunentation ~
- 1. Sequencer related work orders
- 2. Previous sequencer LCO sheets
- 3. Control room narrative logs
- 4. Sequencer related surveillances VEGP Position The NRC position is accurate in that no Tech. Spec. interpretation exists. Previous guidance connected sequencer inoperability to diesel generator inoperability. Recent information has demonstrated that sequencer inoperability should also be tied to " actuation logic and actuation relays",
as found in the instrumentation specifications. VEGP will further review and evaluate this issue to ensure an adequate interpretation exists for the shift personnel. VEGP Docunentation As above
t i f 8/16/90 1600 i t ALTERNATE RADMASTE SUILDING l NRC Concern Concerned that the FAVA system was installed without performing adequate i engineering and safety evaluation (50.59), because the fabrication and quality ; of the system did not meet the RG-1.143 and ASME code requirements. Concluded that the FAVA system was originally installed without an adequate safety evaluation. As a result of a VEGP QA finding in early 1989 involving a breakdown in the procurement and failure to meet FSAR connitments, the system l was removed from service. Subsequently the system was returned to service 1 following two SEs (dated 11/89 and 2/90) which adequately addressed the use i of PVC piping with respect to radiation ciegradation and pipe rupture. Although these SEs did not address the effects of a breat in the hoses (which could
- result in wall spray down or leakage), the use of hoses and effects of hose breaks (i.e., airborne activity and puddling) were addressed in SER Supplements 3 and 4. Although these SEs did not address high temperature effects our interview indicated that these effects were considered in performance of the SE.
i' Concluded that the SE perforsned on 6/90 at the request of RII to evaluate !' the effects of a FAVA system wall spray down and wall leakage to an unrestricted area have been adequately addressed for the use of the FAVA system, because the FAVA System has a protective cover. However, the June 90 safety evaluation 4 inadequately addresses the potential effects of wall spray down from any other source in the ARB due to erroneous assumptions concerning the release path and the dilution volumes. This is a potentially unreviewed safety question concerning the use of the alternate radwaste building. And as such will be j- followed as an unresolved item pending further review and evaluation. i (Unresolved item concerning unreviewed safety question). J ' NRC Documentation I t- IEC 80-18 l SSER 3 Section 11.4 ! l SSER 4 Section 11.4 1
- SSER 8 Section 11.4
- VEGP Position i
The safety evaluation for the FAYA microfiltration system was adequate for use of'the system. The calculation performed to evaluate the " spray accident" in the AR8 was flawed due to erroneous assumptions regarding the release path. ' F These flawed assumptions do not affect the 50.59 evaluation made for the FAVA i- unit. The analysis of the " spray accident" in the AR8 should not have been included as a revision to the safety evaluation in the FAVA unit. Doing so
- confused these two separate issues and was not appropriate.
- . . .- - . - - - . . - - - . - _ -. , ~ . _ - -
~
DRAFT =. During the plant if censing process, details of the construction of the ARB were provided to the NRC. NRC personnel visually inspected the as-built condition of the ARB and associated solid waste processing steams and interface connections that tie other support systems to the equipment in the ARB. Flexible hoses and couplings were in use in the facility at the time of this inspection. The NRC found the facility acceptable for use based on our submittal and their visual inspection with one exception which was subsequently corrected and had to do with exhaust air filtration. This infomation is well documented in the SER Supplements 3. 4. and 8. In June,1990. NRC requested YEGP to perfom an analysis of the " spray accident" in the ARB. This analysis was performed but was based on erroneous assumptions as noted above. VEGP can reperfom an analysis, but since we know of no one i who has performed any similar analysis, a methodology will have to be developed j first. VEGP Documentation Safety Evaluation dated 2/26/90 Other documentation same as NRC i i 4 J < V
1 s . Dater August 16, 1990 T Time: 1600 Snubber Reduction and Use of the Anoranciate Technical Snecification NRC Concern Voluntary entry into a LC0 action statement is acceptable for the purpose of surveillance testing but is discouraged for modification work. (See NRC internal correspondence Murley to Martin dated May 18,1990). NRC has determined that l applying the action staterrent associated with Specification 3.7.8 and then applying the action statement of the applicable system is a correct interpretation of ! Technical Specification requirements. Specification 3.7.8 is intended for broken ' snubbers or functional testing and not for other purposes. With respect to snubber reduction you must have a valid safety evaluation which considers the ramifications of performing the modification at power. If the modification renders the system J inoperable during the installation, as detemined by analysis, then the applicable ! system action statement must be applied. I l Entering a LC0 action statement should represent a net safety benefit and be ! warranted by operational necessity, not just by convenience. The practice should j not be abused by repeated entry into and exit from the LCO. Implementation of l i j snubber reduction during power operation is non-conservative.
\
RG_pocumentation LCO's j Safety Evaluations for DCP's 88-VIN 0ll4, 89-VIN 0047 Copies of MW0's for NSCW "A" Train Letter from Murley to Martin, dated 5/18/90 , Letter from W. C. Ramsey to C. C. Miller, dated 8/15/90 ' . Letter from Denton to Norelius, dated 5/27/90 t ' j VEGP Position l The position associated with voluntary entry into a LCO action statement is [ discussed in a separate position paper. 4 VEGP agrees with the NRC position that applying the action statement associated with specification 3.7.8 then applying the associated system action statement is the ! appropriate way to implement Technical Specification requirements. In addition, VEGP agrees that when performing a modification a Safety Evaluation must be i performed. If the evaluation determines the system is rendered inoperable during the installation process then the actia statement associated with the system must
- be followed and 3.7.8 cannot be applied. L Since snubber reduction increases system reliability by eliminating potential failure modes, implementation of snubber reduction during power operation is conservative. '
r VERP Documentation None r
Date: August 16, 1990 Time 1600 LCD Action Statement Entry to Implement Desian Chance N.tc concern Entering a LCO action statement should represent a net safety benefit and be warranted by operational necessity, not just by convenience. The practice should not be abused by repeated entry into and exit from the LCO. NRC Documentation l i l Letter from Murley to Martin, dated 5/18/90 t l VEGP Position i Voluntary entry into a LCO action statement for the purpose of implementing i a design change is acceptable provided the activity is accomplished within 1 the provisions of the Technical Specification and proper consideration has l been given to the impact on plant safety. This position is supported by NRC Standard Technical Specification interpretation which actually endorses j voluntary entry into an action statement condition on the basis that the NRC "has structured the Technical Specifications to permit the licensee to i exercise judgement within the latitude permitted by the Action Statement , language in the Technical Specification". VEGP implements design changes on safety related systems for the purpose of
- improving system reliability and thereby enhancing plant safety. VEGP maintains that voluntary entry into a LCO action statement to implement a design change is acceptable and desirable in specific cases. VEGP considers this consistent with industry practice.
VEGP Documentation I STS, Section 3.0 Voluntary Entry Into Action Statements ! Issue Date 1/1/82 l l I r
4 CONTAINMENT 8/16/90 i CONCERNS: HYDROGEN MONITORS i I, II. 1 I. ANALYZER OPERATION FOLLOP OWING SI 0 1 OPENING H2S MONITOR NRCQUESTION AT POWER 3 VALVE f, d Are the followingesvalv ,P
/t !9 '
O \ 1 HV-2792A f i HV-27928 considered containment is l I HV-27938 o 1 HV-27918 ation valves 7 \ s a NRC CONCERN
\
NRC feels they are bas i ed on: , VEG FSAR Table 16.3-4FSAR Table 6.2.4-1 Yes.Pthe POSITION i the A-Train Containmabove ar identified valves ' 3 NRCQUESTION If the they are containment i ent H2 Monitor. e containment isolatio n valves for operation of NRC CONCERP these valves 7 solation valves does Tech
- Tech Spec 3.6NRC feels Tech Spec pec 3.6.3 3 apply 6 to
- We say they ar.3 applies to 3 applies for the followin interpretation
- Operations .. to 3 6 3 eProcedu containmen ( ont isolati isolationg reasons:
containment 1-18-90 valves.
- Maintenance Procedure 13130-2 valves in page).
Tech Spec
- FSAR 6.2. 4. 2. 3 re 24932-2 step 3.2 "Pr4 " Caution" statement VEGP POSITION Tech Spec 3 erequisites or.
these valves.6.3 applies to Initial t
- Conditions" if cne of threnderto perform thehannel c calicontainment valve inoper b isolation excessively, ese va a le bration valve .
and therefore T operability. ; the associated etc. then ,lves con) became Tech Spec 3.6 of tai 3 inoperable the H2 analyzer (e.ach Spec 3.6. does notOpening Tcch Spec 3.6.3pret nment inter penetration.would
- 1. would not close entered.
valvas apply to thatleaked va,lv g' e and iProcedure 13as described above.ation (1-18-90) appli 130-2 confirms the way es to hydrogen monitor we want oto perate thes e valves. \ O
1* r !. 8/16/90
. 1600 Procedure 24932-2 was only recently revised to include the reference to the LCO condition. The LCO condition was with reference to breaching
- . the piping boundary outside containment. l
)
i FSAR 6.2.4.2.3 states these essential lines are normally closed and remain I i closed during power operation. The configuration of these ~'?ves are normally closed during power operation. Opening these val n t to perform { l i calibration does not conflict with the system description in the FSAR. ' CALIBRATION PROCEDURE 24551-2 1
^
NRC QUESTION Do we feel step 2.0 still valid, precautions / limitations "may be performed in any plant mode"? VEGP POSITION ; Yes, precaution 2.9 is valid. The procedure for H2 monitor calibration is required to be performed every 92 days on a staggered test basis (i.e. one channel must be tested approximately every 46 days), per Tech Spec 4.6.4.1. Thus Tech Specs recognize this surveillance as one that can be and should be performed at power. NRC QUESTION Do we feel it is necessary to open the isolation valves to perform calibration? VEGP POSITION Yes, for the following reasons: s
- 1) By establishing a flow path to and from containment we are verifying an open flow path exists.
- 2) Verification that pump will operate in the normal flow path configuration is confirmed.
- 3) ALARA concerns associated with positioning the vent valves.
- 4) Risk associated with vent valve manipulations.
NRC QUESTION Was operations involved in review and approval of the procedure? VEGP POSITION No NRC QUESTION Was the 50.59 safety evaluation performed adequately? VEGP POSITION Yes
8/16/90 1600
' NRC OUESTION on Unit I the next day?When this issue was pointed out 8/7/90 on Unit VEGP POSITION When the on infonnation issue was brought up with the Operations Manager, ha bega the issue.
and did not review the nextAtday's the time he felt our procedures were correct activities. When shift personnel were made aware of the NRC concern the test was terminated imediately. NRC QUESTION What Tech Spec requires LLRT testing of the system? 24910 24930 24931 24932 24933 VEGP POSITION Procedures 24930, 24931, 24932, 24933 Spec 4.6.1.2.d for components deffned in FSAR Table 6.2.4-1. satisfy the requ Procedures 249106.7.4.A. Spec and 24932 satisfy the requirements of leakage assessment of Tech NRC QUESTION , Is leak rate testing perfonned on these containment isolation valves added to overall containment leak rate? VEGP POSITION
- Leak C rate for these isolation valves is added to the total type B and leakage. It is not added to type A results.
NRC OUESTION Evaluate applicabf1ity of Tech Spec 6.7.4.A to this system VEGP POSITION i The piping outside containment is covered under the leakage assessment
- program as addressed in Tech Spec 6.7.4.A.
NRC QUESTION Do we feel we violated Tech Spec 3.6.3 on the following two occasions? 4 Unit 2 0411 8/6/90 to 0122 8/7/90 21 hrs. 11 min. Unit 1 2053 8/7/90 8/8/90 18 hrs. 47 min. VEGP POSITION No the containment isolation valves were not inoperable on these dates. Further, in the past calibrations nave been scheduled in accordance with Tech. Spec. requirements (approximately every 90 days).
8/16/90 1600 II. ANALYZER OPERATION FOLLOWING SI NRC QUESTION Are the analyzers placed in service 30 minutes after a safety injection? Requirement: NUREG 0737 IIF1 Attachment 6 TMI requirement Provide station position relative to NUREG 0737 also provide proof of implementation.
)
NRC has looked the following places: 19000-C 19251-C Loss of primary & secondary coolant Reference SER 6-4 VEGP POSITION l Procedure 19010-C (Loss of Primary or Secondary Coolant) step 12 currently ' addresses obtaining containment H2 samples fo11cwing an SI. VEGP intends to enhance this procedure relative to placing the H2 analyzers in service for this purpose. l l l 1 l l
Date: Iugust 16, 1990 Timet 1600 i Precision Heat Balance NRC Concern
/
- The corrective actions associated with LER 90 0l$ (Failure to Calibrate Computer Points Prior to Precision Heat Balance Flow Measurements) are technically correct, the decision not to re-perform the surveillance test was non-conservative.
J NRC Documentatlon i LER 90-015-00 1 DC cards, RCN's Reactor Engineering Calculations 4 Completed test procedures i , VEGP Position The decision not to reperform the surveillance was conservative, based on Engineering evaluation of available data. Additionally, all associated data was reanalyzed assuming potential calibration errors. This reanalysis
- verified the initial engineering analysis.
When Unit I final feedwater temperature instrumentation was determined to be out of calibration, examination of the data indicated sufficient margin to address the out of calibration condition. Reanalysis of the data considering the out of calibration condition confirmed the conclusion. 4 For Unit 2 the calibration of final feedwater temperature was never i suspect. Again, examination of calibration data taken after the test indicated sufficient margin to address potential miscalibration problems. Reanalysis of the data also confirmed this conclusion. l VEGP Documentation i None 4
8/16/90 1600 4) 7 g b f PER$GREL ACC0tNTABILITY NRC Concern Holding shift 'upervision accountable for the number of reactor s trips, LER's and ESFAS actuations has a potential negative influence on plant safety because personnel might not be open about reporting these types of plant problems. NRC Documentation Typical Shift Superintendent Accountabilities VEGP Position These accountabilities enhance reactor safety because they focus personnel attention on safety and compliance issues. Reporting problems is required to achieve good SALP ratings and is part of shift supervision accountabilities. VEGP Documentation Typical Shift Superintendent Accountabilities l 1990 Organizational Goals Performance Appraisal Forms l l l 1 4
1600 8-1 6-90 1 l (/ 3.0.3 1 HR. ACTIONS C*4 p .g b)I
\ l t .
_NRC Concern f Inadequate doctanentation exists to demonstrate all actions taken i during the firct hour af ter entry into Technical Specification 3.0.3. ( l NRC Documentation
- Control Room Narrative Logs j VEGP Position Docisnentation of all actions taker, during the first hour after entry into Technical Specification 3.0.3 does not exist as stated by the NRC.
Howeve r, this infonnation (doctanentation) is not procedurally requi red , nor is it a regulatory compliance issue. Appropriate actions have been taker in the past 3.0.3 entries to meet the time table cf the action statement. VEGP Documentation 4 As Above l l 1 I e 1 i I
Date: August 16. 1990 i
- Time: 1600 E5FAS REPORTABILITY IS$llE /p',
(
~ p NRC Concerns In regard to reportability of ESF actuation NRC has developed a position that '
are for "If, any reason caused (except expected responses to testing) the ESF components to operate then an ESF actuation did occur". formulated in response, to GPC Corporate internal memo (JuneThis 11,1987)position which was provides guidance concerning ESF actuation reportability. i i NRC Documentation
- 1. Internal NRC memo dated July 12, 1990 from Charles E. Rossi !
to Gus C. Lainas. I
- 2. Internal GPC memo dated June 11, 1987 from R. Baker to L. T. Gucwa.
t VEGP Position The June 11, 1987 letter was written by a member of the Corporate Nuclear Safety and Licensing Department to his Supervisor. The information contained in the attachmen*, to this letter was intended to be used as guidance when determining ESF actuation reportability. This inforination was never adopted by the Vo guidance. gtle Electric Generating Plant (VEGP) to be used for reportability The Vogtle practice has been to report the ESF actuation regardless of "what caused the actuation" or "how the actuation occurred". Based on discussions with individuals who review deffetencies for reportability and a review of past deficiency evaluations identified no instances where the position described in the June 11, 1987 letter was utilized in ESFAS reportability determination at VEGP. i ' VEGP Documentation
- 1. W. F. Kitchens memo to OSOS dated June 9, 1987
- RE
- ESF Actuation l
- 2. Sort of DC's by keyword "ESFAS"
- 3. List of ESFAS LER's for Vogtle
1600 8-16-90 TECMIICAL SPECIFICATION INTERPRETATI0RS 4 9 v _NRC Concern 1. formal enough.The control of the generation, approval and distribution is not 2. is not aThe level of review and aps:roval of Tech. Spec. interpretations high enough level (i.e. PR8 review and requi red). concurrence should be In fact. Tech. Spec. 6.4.1.6 a or d ray apply. 3.
- " Info Copy".If NRC guidance is used, author of guidance should be tent a l
MCDocumentation I None VEGP Position Since VEGP Tech. Spec. interpretations are not designed to modify t the intent or breadth of the Technical Specification but merely to clarify thenot is specification re for the on-shift operations crews, a formal process Manager, quired and T.S. 6.4.1.6 a and d are not applicable. The Operations ! being the a Senior Reactor ser.ior member Operators Licenseof plant management required to maintain is the appropriate approval authority for the Technical Specification interpretations generated. As required by subject matter, input from other i ard Compliance, is utilfred in the development of the interpretations. sections In order to provide the NRC with information on VEGP positions on Technical copies of our Specifications, the Pesident Inspectors will be provided with interpretations. VEGP' Documentation None
8-16-90 1600 i 0VERTIME7
, 0 1
4 NRC Concern No provision exists in our procedures that would prevent operators from working an excessive amount cf overtime during an extended period i l of time, f.e. monthly or yearly. Our restriction of no more than 72 hours worked in a sever day period would not prevent excessive evertime en a montnly or yearly basis. NRC Doctanentttion i None VEGP Pasition Operations department personnel use established procedures and guidelines, based on existing regulatory guidance, that limit the i possibility of this situation occurring and ensure cousliance with the regulations governing overtime. J Procedurt OC005-C gives the overtime guidelines and mquires the j department head to evaluate and approve the consistent use of overtime.
, The GMNP, or designee is also requimd to review excess overtime assigned
- to individuals each month to ensuru proper authorization per Figure 1 i
of this procedure. Also, he reviews the overtime to ensure that assignment of excess overtime does not become routine. l For operators working under the union contract, additional guidance 1 is provided for overtime assignment and equalization in the Memorandtsn of Agreement, paragraph 49. Based on review of overtime recortis. LER's. Reactor Trips, and ESFAS , Actuations, no conclusions can be drawn that indicate excessive cvertime has caused operator fatigue or an increased frequer.cy of 4erator errors. ) i Howe,ver. VEGP intends to review this item for potential enhLicements. VEGP Doctanentation Week at a Glance Paragraph 40. Memorandtsn of Agreement j 4 Precedure OC005-C ! i
1 600 8-1 6-90 NON-LICENSED OPERATOR TRAINING i NRC Concern F 1. The PE0 training program does not include under-instruction watches
!- for building qualification.
] 2. The training program may not be adequate to train and evaluate the ability to make routine rounds; some operators may not have completed j their actual rounds task properly. l NRC Docismentation In NRC interviews with new building operators some said they did 3 not actually perform their rounds in training. J l } VEGP Position 1 l 1. Qualification is based on the successful completion of required ! knowledges and skills, which are arrived at throu 4 of a systematic approach to training (SAT) process.gh The the analysis phase INP0 accredited 1 I i program does not rely on any arbitrary number of under-instruction shifts for qualification. Houever, due to.. requests from Plant Equipment Operators surveyed, VEGP will re-evaluate the addition of under-instruction watches j to the building operator qualification checklists.
- 2. The routine conduct of rounds is an identified task with associated supporting knowledges in the PE0 training and qualification program.
1 A coms mhensive instructional unit provides sufficient information and
- j. guidance to assurv that operators have the ability to perform routine
{ rounds. A specific weakness has been identified in the taplementation
- j. of the evaluation process for the task conducting rounds. This weakness j will be handled through the SAT feedback process.
I VEGp Documentation i (- Procedure 11956-C, " Auxiliary Building Operator Training Qualification j
. Checkli st" Qualification Signoff Criteria Cluster $1 - NLO Administrative Duties VEGP Instructional Unit NL.IU-51401-001-C, Conduct Auxiliary Building
- Rounds
- VEGP Instructional Unit NL-IU-51401-0C2-C, Conduct Control Buildin 2
Management Observation Report (MORE) - TQ.3, "On-The-Job-Training"g Round i i i i
. ~ , -, .-
1600 8-16-90 OPERATOR ROUNDS GENERAL INSPECTION 9 , NRC Concern There are differences in the depth of general inspections performed by operators during their rounds. j NRC Documenta' tion NRC observation of rounds by new Auxiliary Building Operators. VEGP Position . The general inspection is intended to identify acy type cf abnonnal condition which may develop. The procedural guidelines are accordingly , very broad. It is not our intent to. detail every possible check which the operator could make in our procedums. The guidance in Operations
, Procedure 10001-C "Logkeeping", describes the overall areas of inspection I required of the Plant Equipment Operators.
We expect there will be diffemnces in the focus of different operators based on their personal experience and shift supervision instructions. , This diversity is a plus to incmase the breadth of the general inspection. I For the rest of 1990, VEGP will increase the ntauber of supervisors and managers doing Management Obsenation Reports on operator rounds, 4 during both day and night shifts. These observations will be reviewed to establish a baseline performance standard and any needed corrective actions will be implemented in procedures, training and practice. I i VEGP Doctamentation Procedure 10001-C. Logkeeping
; Training Cluster 51-NLO Administrative Duties 1
j l i i i 4
\- !k i 8/16/90 16:00 ELECTRICAL SEPARATION ZONE 80 NRC Concern l [f Upon an LOSP a postulated fire in Zone 80 would render Train A inoperable.
- ' and may trip the Train 8 Diesel Generator output breaker. VEGP should insure that no equipment required to cope with this condition would be damaged by the fire while the diesel generator output breaker is being reclosed.
4 NRC Documentation j DC 1-90-299 and 2-90-080 VEGP Position i 1. The design requirement for a fire in this area is to be able to shut down the plant using Train "B" equipment (FSAR 9A.1.40.L.la). j This scenario does not affect our ability to safely shut down the plant, is not a condition outside our design basis, and is therefore not a reportable conditian. A. The postulated fire scenario would not damage the Train "B" safety related equipment necessary for safe shutdown of the
, plant. The D/G would continue to run and an annunciator would l indicate D/G trouble.
B. The operator would be required to observe a loss of power on the "8" Train safety related bus, recognize there was a fire i in the room where the attached non-safety related bus is located, i separate the nonsafety related bus from the safety related bus, and reclose the D/G output breaker. There is adequate time
, for the operator to take these actions.
4 l 2. As a conservative measure, the feature which could cause the D/G , output breaker to open in this scenario is being eliminated so ar. , not to rely on operator action to reclose the breaker.
- 3. VEGP intends to provide additional information to support these conclusions.
VEGP Documentation
; FSAR 9A.I.40.L.la Letter #56-9471 Letter #5G-9510 , DC 1-90-299 DC 2-90-080
AUGUST 16. 1990 AREAS OF CONCERNS NRC VEGP CONTACT CORPORATE CON'I.CT
- D/G Records Starts / Failures Pete Taylor G. Frederick 3/1/90 S R Monitor inop Mode Change Neal Hunemuller JES/D. Carter
- Hissed Sury. Cont. Isol. Heal Hunemull JES/S. Swanson
- March 15 RHR Train B Ron Aiello JES/J. Gasser P. D. Rushton
- Temp. Change Notice to A0P Robert Carrol JES/J. Cash 18028-C-7-90-1
- ESFA Sequencer Out of Service Robert Carrol JES/Horton J. A. Bailey Alternate Radwaste Building Ron Aiello Ron LeGrand/JES P. D. Rushton
- Snubber Reduction Larry Garner Gus Williams Ward /Stringfellow LC0 Action Statement Larry Garner Gus Williams Ward /Stringfellow Cont. Integrity Hydrogen Ponitor Morris Branch Dean Gustafson Ward /Stringfellow Valve Opened Precisicn Heat Balance Morris Branch Gus Williams B. Florian
- Personnel Accountability C. VanDenburgh JES/GB Methodology for Reporting
- Tech. Spec. 3.0.3 Philosophy J. D. Wilcox J. E. Swartzwelder J. Stringfellow
- ESFAS Reportability J. D. Wilcox R. M. Odom J. A. Bailey
- Tech. Specs. Interpretation Morris Branch J. E. Swartzwelder J. Stringfellow
- Overtime / Training & Qualification Larry Garner J. E. Swartzwelder
- Electrical Separation Zone 80 Larry Garner M. Horton P. D. Rushton T. S. 3.4.7.3 CCW J. D. Wilcox J. E. Swartzwelder j j M% [h
,/ '
~
!' s,,
. DIESEL STARTS AND FAILURE REPORTING 8/17/90 Time: 12:00 Page 1 of 2 1
NRC Concern
- 1. The NRC is concerned about the incorrect number of diesel starts reported in LER 1-90-06 and the number of starts presented to the NRC on April 9, 1990 and in the confirmation response letter of April 9,1990. The major ;
issue remaining is to try and determine through personal interviews, how l the number of 19 for diesel 18 was arrived at in the April 9 letter to the , NRC. The revised response to LER 90-06 did not clarify the number of l starts reported to the NP.C April 9. j i
- 2. The inspector noted that documentation provided by Operations to support ;
diesel trending (14980-C and 13145-C data sheets) does not contain an i adequate description of what happens dering the start attempt. The plar.t : is not interpreting Reg Guide 1.108 properly with regard to reporting l valid and non-valid failures. There may be valid and non-valid failures that were not reported. The NRC does not consider the current status of reporting diesel failures to be in compliance with commitments made to the NRC in Violation 50-424/87-57. nxC Documentation The NRC has reviewed the diesel start log and s'9 porting documentation (14980-C and 13145-C data sheets). The NRC current 1,y blieves some problems identified on 14980's and 13145's should be classified as non-valid failures and reported to the NRC. The NRC has requested and received written analysis to explain the disposition of the following 18 diesel starts: #'s 123, 124, 132, 133, 134, 136, 160, 161, 162, 164, 165, and 190. LER 1-90-06, revision 1: 0A Audit Report OP26-90/33: QA Audit Report OP09-90/31; and Special Report 1-90-05, dated August 7, 1990; GPC confirmatory action letter dated April 9, 1990. VEGP Position
- 1. The error made in the number of diesel starts reported to the NRC on April 9, 1990, and in LER 1-90-06 is attributed to two factors:
- a. The testing as described in LER 90-06, revision 0, was in the
" context of" and "in reference to" the diesel control systems. The first two sentences of the 5th paragraph explain actions taken with regard to sensor calibrations and control system testing. In this context, the test orocram correlates to testing discussed with the NRC on April 9,1990, and reported in the April 9,1990, confirmatory letter. The LER 90-06 comment of " subsequent to the test program" was not intended to exclude successful diesel starts before declaring the diesel operable. As a result, diesel starts after testing of the control systems, but before a declaration of operability were counted. The transmittal letter for LER 90-06, revision 1, describes the confusion and attempts to clarify the concern by redefining the types of starts and the point of counting. .
/dAmig E/r*
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DIESEL STARTS AND FAILURE REPORTING 5 Page 2 of 2
- b. LER 90-06, revision 1, was intended to clarify any inadvertent
" misleading" of the NRC on successful operation of the diesel control systems. When Vogtle Management was aware of the problem in LER 90-06, revision 0, management . notified the NRC Residents. Also at the corporate office on 6/11/90, W. Shipman contacted Kon Brockman and on about 6/11/90, H. G. Hairston, III, contacted Stu Ebneter of NRC Region II. The revised LER was submitted on 6/29/90.
- 2. After a thorough review of Reg Guide 1.108 Engineering Support (Nike Horton) agreed that all diesel start eroblems have not been reported as failures. GPC's response to NRC Violation 424/87-57 committed to report such equipment problems as failures; however, due to internal administrative problems, the commitment was not implemented. Engineering Support intends to review diesel start records for any unreported failures.
VEGP Documentation o LER 1-90-06, revision 1; QA Audit Report OP26-90/33; QA Audit Report OP09-90/31; and Special Report 1-90-05, dated August 7, 1990; GPC confirmstory action letter dated April 9, 1990. o 18 diesel start analysis available 8/15/90 and Reg Guide 1.108 position i from Engineering Support. I I l
i o
~ 8/17/90
' Response to NRC Question Concerning Time: 12:00 Diesel Starts Reported on April 9,1990 and in LER 90-06, Revisions 0 and 1 4
Ouestion #1
. 1. Who prepared the slide for the 4/9/90 presentation?
j Answer: G. Bockhold, Jr., J. P. Cash, and K. Burr working as a group. l 2. Who approved use of the slide? l Answer: G. Bockhold, Jr. Question #2
; 1. Who prepared the confirmatory lette- of April 9, 19907 i j Answer: C. K. McCoy, J. A. Bailey, H. G. Hairston, III as a group. ; 2. Who approved the letter?
Answer: .H. G. Hairston, III j Ouestion #3 (with regard to LER 90-06, revision 0, dated 4/19/90)
- 1. Who prepared the LER?
j Answer: Several draft revisions of the LER were prepared by Tom Webb and others of the NSAC group of the Vogtle Site Technical Support. i These drafts were reviewed and commented on by the Plant Review l Board. The final revision of LER 90-06, revision 0 was prepared by a phonecon between site management and corporate management. l Those participating are believed to be G. Bockhold, Jr., A. L.
- Mosbaugh, J. G. Aufdenkampe, H. Shipman.
- 2. Who reviewed the LER7
- Answer: All revisions of the LER were reviewed by the PR8 and the
! General Manager-Plant Vogtle.
- 3. Who approved the LER7
) Answer: The LER was approved by W. G. Hairston, III Ouestion #4
- 1. Who prepared the cover letter for LER 90-06, revision I?
L Answer: The cover letter was prepared by H. H. Majors of the corporate , staff. This letter was prepared under the guidance of W. G. i Hairston.
- 2. What was the purpose (intent) in the wording of the cover letter with
! regard to the number of diesel starts? ! Answer: The cover letter was intended to document discussions with NRC Region II to clarify the starts documented in LER 90-06, revision O. By picking a well defined point to specify
" subsequent to the. test program" it was possible to identify a substantial number of successful diesel starts. This was intended to remove any edditional ambiguity.
i , Duestion #5 l 1. Who in corporate added the words " subsequent to the test program" in LER L 90-06, revision 0? i Answer: Corporate Licensing personnel in conjunction with the phone i conversation described above made editorial changes as directed. Those present during the phone conversation are thought to be W. Shipman, G. Bockhold, Jr. , A. L. Mosbaugh, l J. G. Aufdenkampe, and J. Stringfellow. 7
- i 2 'd OSIC1 06/c2/B0 v1NU"11W-2*D30 WOEd
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4 e j,i UNITso sTAfts i NucLEAn nsoVLAfonY COMMisslON i i E neo60es II tot maastTTA svattT,hm, ATLA 8tTA,StonelA 30833 D$hn*Nos. 50-424. 50-425
! I,1 cense Nos. NPF-68. NPF-81 '
i Georgia Power Company i ATTN: Mr. W. G. Hairston. III i Senior Vice president - Nuclear Operations i P. D. Box 12g5 Sirmingham AL 35201 i, , Gentlemen: I i
$UBJECT:
! CONFIRMATION OF MEETING - ENFORCEMENT CONFERENCE 1 (NUREG-1410. NRC INSPECTION REPORT N05. 50-424/90-50-425/9016) i This confirms a telephone conversation between Mr. C. K. McCoy of your sta
- and Mr. K. Brockman of my staff on August 10. 1990, i Conference september to be conducted at the NRC Region !! Office at 10:00 a.m. oncon 5.1990.
This meeting was requested to discuss numerous items 1 identified by the Incident Investigation Team (!!T) which was chartered in response to the Site Area Emergency event of March i'
- 20. 1990. The team's
- Heat 1990.* Removal System Ituring Mid-Loop operations a I'
The first issue to be discussed concerns the site's fa timely manner. I what was the sequence of events which precluded the rOf particula , being provided through alternate coessunications means.equired information from l and 50-425/90-16.concerning this issue is provided in NRC 50-424/90-16 Inspection Report Mo The second issue to be discussed concerns the inability of the site perso to establish containment integrity within the time limits re Closing Containment Hatch." procedures 12006 C. " Shutdown 0pening andto Cold Shu Your response to Generic Letter 88-17 made 4 numekus commitments concer. ling analyzing the conditions and time req necessary to ensure an appropriate establishment of containment integrity and for developing the procedures necessary to ensure the closing of hatch before the plant could enter reduced inventory conditions. the eoufpment You saould be closure. how you verified whether the time limit could b hatch was not able to be closed in the required time on March 20, 1990. Seth event should be addressed.the worst case environmental conditions and
' f/, '
, i s SCYDt't*9F.% #
cd istc1 os/c2/so WANE 71w-2*D3G wogs a t
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Georgia Power Company
]
2 - g g g gg i The third issue concerns the failure of the emergency diesel generator to provide AC power as intended. The plant has ex
, grevious indicators of diesel generator problems,perienced a long litany of ! ahake down" the system after maintenance outage:. includingacommonneedto These shake. downs" were i
often in response to, what now seems to have been, failures within the
, pneumatic control system. You should address why ;/our root cause analysis J
program did not investigate the totality of continuing diesel problems which had been experienced at the plant and what has been done to improve the root ] cause analysis program at the Vogtle site. i Should you have any questions regarding these arrangements, we will be pleased
- to discuss them.
i j Sincerely. ) i Luis A. Reyes Director Division of Reactor Projects 3
Enclosure:
Proposed Meeting Agenda ' 1 ! cc w/ enc 1: R. P. Mcdonald Executive Vice President. Nuclear , Operations Georgia Power Company P. O. Sox 1295 1 Straingham,AL 35201 ! C. K. McCoy ! Vice President-Nuclear Georgia Power Company 4
- P. O. 1295 1
! 81mingham, AL 35201 ! 8. Sockhold, Jr. i General Manager, Nuclear Operations ! Seorgia Power Company P. O. 1600 ! Naynesboro GA 30830 J. A. Sailey Manager-Licensing
- Georgia Power Company P. O. Sox 1295 Simingham, AL - 35201 l
, (ce w/ enc 1 cont'd - See page 3) j
1 y *d 2EtC1 06/C2/B0 01N0"IlW-2*D30 WOGd 1 Georgia Power Company 3 ! i AUG 12 990 1 cc w/ enc 1: Cont'd 4 4 Ernest L. Blake, Esquire
' Shaw, Pittman, Potts and .
Trowbridge l
- 2300 N 5treet, NW
' 1 Washington, D. C. 20037 i J. E. Joiner, Esquire 1
1 Troutman, Sanders Lockeman, and 1 Ashmore l i ' 1400 Candler Butiding 127 Peachtree street, NE :
- Atlanta, GA 30303 \
1 1 D. Kirkland, III, Counsel l Office of the Consumer's i Utility Council i Suite 225, 32 Peachtree Street, NE Atlanta, GA 30302 i Office of Planning and Budget Room 6158 l 270 Washington Street, SW ! Atlanta, GA 30334 4 Office of the County Comissioner Surke Count Waynesboro,y GA Comission 30830 i Lonice Barrett, Comissioner Department of Natural Resources . 205 lutler street, 3E, Suite 1252 Atlanta, GA 30334 Thomas Hill, Manager i Radioactive Mater'als Program 4 Department of Natural Resources 878 Peachtree St., NE., Room 600
- Atlanta, GA 3030g !
Attorney General Law Department 132 Judicial Butidin Atlanta, GA 30334 g
- State of Georgfa t
i
, E 'd 10101 E 'd 2EICI 06/C2/80 WANW11e-Z'D3u woes 1 . . .
i (NCLOSURE PROPOSED MEETING AGENDA Georgia Power Company Enforcement Conference 10:00 a.m. September 5, 1990 1 A. Opening Remarks
$. D. Ebneter W. G. Hairston, !!!
- 3. General Discussion C. K. McCoy i 1. Circumstances surrounding the failure to E. Bockhold make a $1te Area Emergency notification in a timely manner
- 2. Inadequate root cause analysis program G. 80ckhold
- 3. Circumstances surrounding the inability 6. Bockhold i
to establish containment integrity in a timely manner. C. Closing Remarks
- W. E. Hairston !!!
S. D. Ebneter 1 4 l i j i Y Ca W s. . '
j
- ' '~
- -: /i .: '~4 ~~~
..:~.. -. : .-: 1-3 i:5 -
- 4 ::: e Weroffice Correspondence G c O'j. ? R: v.c d i
DATF: August 18, 1990 RC. Vaatle ~1e-tet: hr.trat h.o Ping Audit Ftr. ann; 0:'13 90/29 8430 Log: ELV-02051 Security Code: NC FRCM- C. C. Miller TU: G. R. frederi<.r. > I'rocedure revisions resultino from the findings in this aud't re" port will be coroieted by Dece.v.ber 21, 1990. We welcome tne opport#.ity to di: cuss any details in addition to those given in the aacit report.
/ //s. -
C. C. Miller C CM,*ci r ; t ac: Geercia Power Co san. P. D. Rusnton , M. J. Ajitmi ; J. A. Rod;ers NOV.S
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GEORGIA POWER COMPANY ' Inverness Building 40 4 P.O. Box 1295 Birmingham, Alabama 35242 TELECOPY COVER SHEET SONOFCO-V0GTLE - 4TE FLOOR 4 Teleceptors (205)-877-7885
- Verf17 : (205)-877-7897 1
August 29, 1990
- caEs NUMBER OF PAGES
(Escluding Cover Page) 4 ) RECIPIENT: Pisase notify us if you have problems receiving this talesopy. 4 I Tos F10M: George Bockhold ! Lewis Ward 7802 EITEN5103: 3118 EITENSION: Admin B1dg. LOCATION: Inverness LOCATION: Plant Vogtle i I HLIW1E l, Should this docusant be returned to you after it has been senef
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SUMMARY
OF OIE5EL GENERATOR PRESSURESENSOR(CALCON)PR08LENSATV0GTLE ) l A. Statoment of Problem: NUREG-1410 lists Calcon pressure switch prob' ens, which far exceeded industry experience. This precursor indicator to the 3/20/90 event had not been adequately addressed.
- 1. Pressure Switch Out-of-calibration nroblems t
. 12 DG pressure switches have been out-of-calibration and havs required readjustment over the lifetime history of Vogtle from i l
NUREG-1410.
. This covers a population of 24 sensors each of 4 engines), ,
over a lifetime of about 92 sensor-years! (6 on l
. Typical out-of-calibration values range from to psig.
khisdatadoesnotrevealanyunexpectedresultsbasedonthedevice type (hydraulic-pneumatic), setpoint drift amount, or frequency. Thus, a formal root cause evaluation would not be wacranted.
- 2. pressure Switch (Defective) oroblems:
. 5 pressure switches were replaced (prior to 3/26/90) due to being termed " defective" by the technician.
. For 4 of these sensors, no reason for the failure was determined.
. After the 5th " failure" (P-3 relay removed from DG 1A on 3/25/90 as part of post-event troubleshooting), the switch was sent to Cooper Industries for failure root cause determination. The sensor was determined to be operating properly.
. Subsequent to the 3/20/90 event, all 3 low lube oil pressure switches were removed from the 1A DG to investigate a sensor malfunction alarm that had occurred during the event. The switches were tested by Cooper Industries on May 30 and 31, 1990 with the following results:
A. The 'B" and "C" switches were operating properly and were set i correctly. B. The "A" switch was stuck in the tripped condition. This condition was stated by Cooper to be the same as reported by Cooper's 10CFR21 report addendum of May 12, 1988. The Part 21 states " Devices that are already installed and operating after several hours between tests have demonstrated their reliabtitty.
.IMO Delaval recommends that all devices not installed, or that are installed but have not operated for several hours between tests, be returned to IMO Delaval for remachining, inspection and testing."
.1.
AJG-29 '90 1::19 :"): SCNCPC-VCGTLE TE_ 'O 21-225-57~'-7995 :716 "?? I _ C. In 1988 VEGP reviewed this Part 21 and returned all spare sensors for rework. Installed sensors were not returned since they were believed to have met the reliability conditions stated in the Part 21. D. Following the May 31, 1990 testing results from Cooper, modified switches were ordered to replace all installed Calcon pressure switches. These switches (6 per engine) were received, calibrated and installed; with the last OG completed on June 15, 1990. E. Cooper was requested by Georgia Power Company to clarify the original Part 21 notification. On June 8, 1990, Addendum 3 was issued stating:
"Our recommendation of May 12, 1988 clay have been confusing and in light of this failure after 9 years, it is appropriate to restate our recommendation. Cooper Industries recommends that all pressure sensor devices, Cooper P/N F-573-156, be modified or replaced by devices identified as Calcon P/N B4400B."
- 9. Root cause
. A design deficiency existed in the Calcon pressure switches.
Replacement switches had not been installed due to mis-interpretation on the May 1988 Part 21 notice. C. Corrective Action
. All Calcon pressure switches have been replaced with the improved model within the last 6 months.
. The Part 21 was reissued for clarification.
D. Sionificance of Problem
. None in fact. Multiple simultaneous failures were unlikely, but could have rendered one or more OG inoperable.
E.
Conclusion:
. Georgia Power Company took prompt actions to identify and correct the ooserved switch failure in 1990.
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SUMMARY
0F DIESEL GENERATOR TERPERATURE SENSOR (CALCON) PROBLENS AT V0GTLE , ] A. Statement of Prob 1m; NUREG-1410 lists Calcon temperature switch ' problems, which far exceeded industry experience. This precursor indicator <
! to the 3/20/90 event had not been adequately addressed. -
7 1 I . Following the 3/20/90 event, all 3 Jacket Water Temperature Switches - were removed from the lA DG for testing. ; 4 i
- . A test program to determine the as-found condition and failure mechanism :
- of these switches was developed. The purposes of the test were
- 1. Determine the reliability and potential failure mechanisms of two )
i new sensors.
- 2. Determine the cause of failure of the installed switches.
ram was conducted at Wyle Laboratories in Huntsville, This*testpro$/23/90to5/4/90. 4 Alabama from Copies of the report were furnished to
- the NRC and all TDI Owners through the owners Group. Pertinent
!- conclusions from this test program include:
- 1. Insufficient temperature stabilization period prior to calibration -
! The sensor exhibits a setpoint shift as the sensor body and internal J components change temperature. Note: Failure to recognize this
- phenomena and properly compensate for it during switch calibration i was a contributing cause of one switch trip on DG 1A on 3/20/90, and l subsequently on DG IB on 5/22/90 during switch replacement. Note
- This phenomena is undesirable in a sensor that is designed to sense i temperature.
- 2. Contaminants on the temperature sensor (tip) - Direct immersion of the sensor tip in a calibration bath can result in residue buildup
, that can affect the setpoint. Although this is not a standard practice, isolated cases may have occurred during previous , calibrations, which could have contributed to the numbers on the ! , chart. :
- 3. Water bath heatup -rate - A slow, controlled bath heatup rate is necessary to allow the sensor temperature and bath temperature to be l approximately the same, while avoiding excessive sensor heatup.
- Previous calibration procedures did not recognize this affect. This ,
. affect could have produced some of the previous setpoint adjustments j i that contributed to the numbers on the chart.
- 4. Thermowell setscrew tightness - This produced a 20 setpoint shift.
Although not large, in relation to the tolerance band of 140, it ; could have been a factor in making previous setpoint adjustments ; that contributed to the numbers on the chart. i i
. _ _ r - _ . = - . - _ -
AUG-29 '90 14: 19 ID:50HCPCC-VGGTLE TEL NO 1-205-977-7995 #716 P25 4
- 5. Spacer-tube tightness - The s)acer tube can self-loosen when not locked with thread-sealant, whic1 produces a setpoint shift of about 800 per turn. This could have been a direct contributor to some of the past failures or setpoint shifts. The vendor has stated that all new switches are supplied with sealant; however, this deficiency '
has continued to be observed.
- 6. Internal contaminants - Several switches had internal contaminants in the poppet valve area consisting of thread sealant material and metal slivers apparently from the inlet air port threads.
Subsequent examination of new switches at the plant site revealed similar contamination, concluding that the manufacturer or vendor can be introducing contaminants during calibration. Internal contamination was the direct cause of one switch failing on DG 1A on 3/20/90, and was a contributor to the second failure. , . . Upon completion of the above Wyle testing program, a calibration procedure was written specifically for these switches to include: ! 1. Q,isassembly of the switch, internal cleaning, and provisions to prevent re-contamination. d
- 2. Inspection and application of thread-sealant to the spacer tube.
- 3. Requirements for the sensor to be calibrated in a thermowell.
1
- 4. Temperature stabilization prior to calibration.
l . The new procedure was used to calibrate 3 new Jacket water temperature sensors, which were installed on DG 1B on 5/22/90. During the i subsequent maintenance start of the engine, these sensors tripped. They were then removed and carried to Wyle Laboratories where as-found setpoint testing showed all 3 to trip between 1600 - 1660F, j . Subsequent evaluation of the differences between Wyle and plant test techniques led to the following conclusions: 4
- 1. The plant bath had internal flow blockage that did not permit a uniform temperature or heatup rate in the bath. This produced several degrees difference at the reference vs. test specimen
- locations.
- 2. The soak temperatore requirement was poorly worded. Although the minimum soak times from the Wyle report had been observed, the sensors had been soaked too iong (up to 6 hours) at near the setpoint (2000F). This resulted in further adjustments, longer soak, and more adjustment; producing switches that were far out of adjustment when subsequently returned to their normal ambient
- operating condition. 1 4 . Upon correction of the above bath and procedural deficiencies, consistent settings were obtained at the site on the same switches that had been reset at Wyle Labs. _4
- 716 P06 0 4JG-19 ' c0 118 20 :D:50tOPCO '.CG7LE EL +C:1-205-5'7-7E95 1
. All jacket water temperature switches on all DG's were removed, cleaned and recalibrated per the latest procedure in early June 1990. l l
. A copy of the latest procedure, with lessons learned, was provided to j the TOI owners Group in June 1990. i B. Root cause Internal contamination can cause a properly calibrated switch to trip.
Cs11bratton of the switches was inadequate to ensure the desired setpoints. C. Corrective Action
. A calibration procedure that cleans and properly controls the calibration requirements has been written and implemented. j
. Reliability of the basic switch component was established at an indepsndent laboratory.
Switches have been defeated in the emergency start mode. l
. Currently evaluating replacement of pneumatic sensors with the vendor.
D. Sionificance of Problem f 1
. Trip of DG 1A on 3/20/90, and trip of DG IB on 5/22/90.
E. Conclusion !
. Georgia Power Company took prompt action to identify the cause of ;
temperature switch malfunctions, following the 3/20/90 event. ! l
. Prior to the 3/ao/So event, in retrospect, the previous ca11Dration )
drift and failure history should have been recognized as a problem to be l resolved. However, discussions with the switch manufacturer did not ' produce any clues that the switch has to be inspected, cleaned and i calibrated under the conditions that Georgia Power subsequently l developed. In fact, the 10CFR21, Addendum 3 June 8, 1990 update to the l NRC states: "While no specific component failures have occurred, the , setting and verification of same, is procedure sensitive." Note that 1 this was Afler VEGP had resolved the issue internally. ! l
i
/G-19 ' E0 14:20 !D:5CtOPCC-VCGTLE ~EL NC 1-205-677-7855 =~'15 F07 j-l 4
)- i DIESEL GENERATOR AIR START VALVE FAILURE AND ROOT CAUSE EVALUATION ! l I i A. History of Problem- - i i . 1/24/90 - DG 2A rolled but failed to start during routine surveillance ! testing. One air receiver had been isolated to test independent l starting on the other bank. The operator noticed an air Irak on one of 1 the air start valves solenoids, unisolated the second bank, and ! successfully started the engine. l . 1/25/90 - (a) DG ZA start was attempted to satisfy Tech Spec action l requirements. The engine slow rolled but did not start. A second start l attempt was successful. j i (b) In order to isolate the cause of the start failure, troubleshooting l i was conducted including: replace air start distributor filters, ; check,ing operation of the governor, fuel racks, fuel system and cy1 W er 4 air start valves. No apparent problems were found, l
- (c) The engine was started 3 more times with no problems. l I . 4/12/90_ - DG 2A rolled but did not start during normal surveillance. l l_ Operators decided that the start pushbutton had not been depressed long I
- enough and reran the test. The engine successfully started, which
! seemed to confirm the operator's decision. The first attempt was not l considered to have been valid, and was not reported. i 1 l . 7/5/90 - A similar event occurred on DG 18 during surveillance testing.
- Again, the operator attributes this to short pushbutton action, and did I not log or report the start attempt or failure. I l
i . 7/11/90 - OG 2A again slow rolled and failed to start. The system engineer had developed a test plan for troubleshooting the start F pushbutton and seal-in circuit root cause of the previous observed l problems on 2A. This investigation led to the following conclusion:
- (a) The seal-in circuit for the start pushbutton seals in any attempted start.. Operator belief that the start pushbutton had to be i depressed for a certain minimum time i period ste med from a simulator phenomena, and does not exist in the plant.
J (b) Discussions with Cooper Industries resulted in parallel investigations in the air start solenoid valves, air start i distributors, air line routing, and individual cylinder air start valves. l (c) Several individual cylinder air start valves were determined to be
; sticking- in the open position. If the engine crankshaft posttien i was such that a particular combination of stuck valves also L occurred, then certain cylinders would be opposing each other and
4 l AUG-29 '90 14:21 "O:50rOPCC-VCGTL- - NO:1-205-577-?E85 c716 F09 9 - l the engine would not start. This likelihood would be increased f half the cylinders were isolated, as in do,ie during the surveillance test. (d) During disassembly of the cylinde- air start valves for troubleshooting, some of the piston:, had to be removed with air pressure or excessive p.ysical force, indicating that the pistons were binding in the valve ups. (e) Detailed measurements of the pistons and caps were made for vendor review. These concluded that several caps were machined with the 4 bore slightly oval-shaped and tapered by several mils, this distortion was subsequently observed in several new caps in the - warehouse. . (f) Vendor direction to correct the binding was to polish each piston to provide at least I mil clearance with its matched cylinder cap, followed by an engine run to heat the assemblies to nor:nal engine temperatures, followed by a pop test of each valve to ensure that
..it was cycling freely.
(g) This corrective action was taken on all 4 DG's. 7-13-90 DG 2A completed . 7-18-90 DG 18 completed 4 7-21-90 DG 1A completed 7-23-90 DG 2B completed. (h) Based on this problem, Cooper Industries issued a Part 21 report to the NRC on July , 1990. Long term corrective actions have not been finalized bu Finclude areas such as: (1) Factory QC check of cap dimensions. (2) Possible material changes in the piston or tap. (3) Expanded testing during routine overhauls. (i) An Event critique Team reviewed the July 11, 1990 start failure. This team concluded that the April 12, 1990 start failure should have been critiqued, but was not. Additional conclusions from the review process include: (1) Clarification of plant vs. simulator pushbutton delay times for licensed operators.
-(2) Instructions to operators to consider and log all attempted DG starts.
B. Root Cause of Events
. Manufacturing defect in individual air start valves, resulting in occasional binding in the open position.
Afs-19 ' 90 1 :: 21 ~ 2: 50 CPcC-.0GTLE TEL eO 1-205-877-7893 2716 P09 I C. corrective Action
. Increase clearance in valves to prevent binding, and verify.
, Notify vendor for Part 21 issue.
. Long-term correction not yet complete? by vendor.
D. 51ont ficance of Problem
. DG would not start if a particular crankshaft alignment and stuck air valve combination existed. This condition was ex during surveillance testing when half of the cylinders were isolated from '
starting air. 1 l
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1 ELV-02059 ' j 0579 j
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Docket No. 50-43 b 1 U. S. Nuclear Regulatory Comnission - ! Region II - 1 i 101 Marietta Street, N. W. ! ! Atlanta, GA 30323 ~l ATTN: Mr. S. D. Ebneter .
- Dear Mr. Ebneter
- V0GTLE ELECTRIC GENERATING PLANT !
- CLARIFICATION OF RESPONSE TO CONFIRMATION {
j OF ACTION LETTER J ! By letter dated April 9, 1990 (ELV-01516), Georgia Power Company (GPC) responded L to a Confirmation of Action Letter dated March 23, 1990.- In that letter and in
- our meeting notes, GPC reported 1 hat Diesel Generator (DG) 1A had been started 18 times and DG 1B had been started 19 times with no failures or problems
- between March 20 and April 9, 1990. Similar information was reported in-l Revision 0 of Licensee Event Report (LER) 50-424/1990-006 dated Apr.il 19, 1990 i (ELV-01545). As reported in our telephone calls to the NRC, we subsequently discovered that this information was in error.
i In Revision I to LER 50-424/1990-006 dated June 29, 1990 (ELV-01729), GPC l- attempted to clarify the correct number of DG starts occurring in this time !~ period by using regulatory guide terminology (i.e., valid vs. successful
- starts). This revised LER accurately reports the number of valid DG starts
- during the period of March 21 through June 7, 1990. However, during a recent NRC inspection it was pointed out that the revised LER did not adequately clarify th4 numbers in the April 9th letter.
The confusion.in the April 9th letter and the original LER appear to be the r result of two factors. First, there was confusion in the distinction between.a . successful start and a valid test. For the purpose of this letter, a start was considered successful when the DG was started and either ran or was intention-l ally shut down due to testing in progress, as identified on the attached tables. Our-use of the term " successful" was never intended to imply a " valid successful test" in the context of Regulatory Guide 1.108. Many start attempts were made !. to test the DG's IA and 18 using applicable operating procedures. These F procedures and data sheets do not contain criteria for determining if a start is
. successful which resulted in determinations of success which were inconsistent, ,
. with the above definition. Second, an error was made by the individual who i
- performed the count.of DG starts for the NRC April 9th letter.
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U.'S. Nuclear Regulatory Commission : ELV-02059 Page 2 The purpose of this letter is to correct the figures related to the number of DG l starts reported in the April Sth letter. Attached are tables 1 and 2 which > summarize the DG starts for the period indicated. For DG 1A, there was a total ; of 31 start attempts and 29 of these attempts were considered successful-after the two failures associated with the March 20 ever.t. For DG IB there was a-total of 29 start attempts and 21 of these attempts were considered successful. i Further,- for DG 18 there were 12 successful sequential starts. _ 1 f Sincerely
. . [0A W. G. Hairston, 11 i WGH,III/NJS/gm Attachments _
xc: Gegraia Power Comoany Mr. C. K. McCoy Mr. G. Bockhold, Jr. Mr. R. M.-Odom Mr. P. D. Rushton NORMS U. S. Nuclear Reaulatory Commission Document Control Desk Mr. T. A. Reed, Licensing Project Manager, NRR Mr. B. R. Bonser, Senior Resident inspector, Vogtle l
', s, ,
TABLE 1 DIESEL GENERATOR 1A START RUN UNPLANNED L DAl[ SUCCESS Tif,E TRIP DISCUSSION 139 03-20-90 No 1 min Yes Failure to maintain load. 140 03-20-90 No 1 min Yes failure to maintain load. 141 01.20-90 Yes 4 1/2hr No Manual start, load maintained. 142 03-20-90 Yes 45 min No Normal reserve auxiliary transformer swap method. 143 03-20-90 Ye s -" 5 min No Observation run. 144 03-20-90 Yes 20 min No __ Observation run. 145 03-23-90 Yes 60 min No Observation run. 146 03-23-90 Yes O min No Started wrong diesel generator. 147 03-29-90 Yes 50 min No UV test start #1. 148 03-30-90 Yes 2 hr Yes* Bubble test #1, high temperature jacket water sensor vented. 149 03-30-90 Yes 6 min No Trip simulation test. 150 03-30-90 Yes 6 min No Trip simulation test. 151 03-30-90 Yes 3 min No Trip simulation test. 152 03-30-90 Yes 6 min No Trip simulation test. 153 03-30-90 Yes 4 min No Orifice modification functional test. 154 03-30-90 Yes 10 min No Orifice modification
- functional test.
155 03-31-90 Yes 2 min No Orifice modification functional test. 156 03-31-90 Yes 3 min No y~ I Orifice modification functional test. 157 03-31-90 Yes 10 min No Bubble test #2 158 03-31-90 Yes I min No Sensor trip timing test. 159 03-31-90 Yes 1 min No Sensor trip timing test. 160 03-31-90 Yes 2 min No Sensor trip timing test. 161 03-31-90 Yes 1 min No Sensor trip timing test. 162 .03-31-90 Yes 75 min No Sensor trip timing test 163 03-31-90 Yes 27 min No UV test start #2. 164 04-01-90 Yes 1 1/2 hr No Normal surveillance test. 165 04-06-90 Yes 1 min No Jacket water temperature test. 166 04-06-90 Yes 1 min No Jacket water temperature test. 167 04-06-90 Yes 10 min No Jacket water temperature test.
I
. l 1
I I TABLE I (CONTINUED) DIESEL GENERATOR 1A START RUN UNPLANNED I L DAlf SUCCESS IIE TRIP DISCUSSION __ 168 04-06-90 Yes 2 1/2 hr No LOSP trip modification ! functional test. : 169 04-09-90 Yes 1 3/4 hr No Normal surveillance test. )
- Unit CPipped during bubble testing due to one sensor venting and another sensing line being disconnected _for testing. This is further described in ;
NUREG-1410. ' WW N i l I
1 D-TABLE 2-DIESEL GENERATOR 18 START RUN UNPLANNED NL.__ DAll SUCCESS llE TRIP DISCUSSION 120 03-21-90 No O min No Post-maintenance run, prime fuel lines. I 121 03-21-90. No 0 min No Post-maintenance run, prime fuel ' lines. 122 03-21-90 No 15 min No Post-maintenance run, adjust
-' governor. l 123 03-21-90 No 2 min No Post-maintenance run, fuel oil '
__ delta pressure high. 124 .03-21-90 No 4 min No Functional test run, fuel oil delta pressure high. 125 03-22-90 Yes 6 min No Functional test for maintenance. 126 03-22-90 Yes 1 min No Functional test for maintenance. 127. 03-22-90 Yes 15 min No Post-maintenance overspeed test. 128' 03-22-90 Yes 3 min No Post-maintenance overspeed test. 1 129 03-22-90 Yes 5 min No Post-maintenance overspeed test. 130 03-22-90 Yes 5 min No Voltage clamp circuit adjust. 131 03-22-90 Yes 2 min No Voltage clamp c4rcuit adjust. 132 03-22-90 No 1 1/2 hr Yes Post-maintenance load test, high temperature lube oil trip. 133 03-23-90 Yes 7 hr No Post-maintenance load test -- 134 03-23-90 No 3 min Yes Post-maintenance load test, low pressure jacket water trip.
-- 135 03-23-90 Yes 4 1/2 hr No Post-maintenance load test.
136 03-24-90 No* 33 min No Post-maintenance load test, high ! temperature jacket water alarm. 137 03-27-90 Yes 1 1/2 hr No Bubble test. 138 03-27-90 Yes 42 min No Trip simulation test. 139 03-27-90 Yes 3 min No Trip simulation test. 140 03-27-90 Yes 2 min No Trip simulation test. 141 03-27-90 Yes 6 min No Trip simulation test. 142 03-27-90 Yes 57 min No UV test. 143 03-28-90 Yes 1 3/4 hr No Normal surveillance 144 03-28-90 Yes 4 min No Low pressure lube oil modification functional test. 145 03-28-90 Yes 4 min No Low pressure lube oil modification functional test. 146 04-04-90 Yes 1 1/4 hr No Post-maintenance load test. 147 04-05-90 l' Yes 5 min No LOSP trip modification functional test.
.148 04-05-90 Yes 2 hr No Normal surveillance. j
- High temperature jacket water trip alarm was. received and the engine kept !
running. i d
- - - _ _ = . -. . . . . . _ -, , - .- - -. . - _ , _ ,
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KO H N. KO H N & CO LAPINTD. P.C. ATTORNEYS AT 4.Aw a e e r6o. ion Avtwut. == wasenwetose. oc 3000 t 8303e 334 4663 esi...t. O s D a w" O' COWe#8th geg g= m a O m a...
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'[3 September 11, 1990 SEP 11 8 $
^
0 19 0 0 tl$ W W to Hon. Kenneth M. Carr, Chairman / United States fluclear Regulatory Commission .N ((t h Washington, D.C. 20555 4
Dear Mr. Chairman:
Please find enclosed the original and one copy of Petitioners Marvin B. Hobby's and Allen L. Mosbaugh's request for proceedings and 11.. position of civil penalties. l
, The exhibits identified in the petition will be forwarded 1 under seperate cover.
4 Petitioners stand ready to assist the Commission in any I way they can. Please do not hesitate to contact me in this l regard. ) 1 very truly yours, j
/~ /M Michael D. Kohn Counsel to Marvin B. Hobby and Allen L. Mosbaugh cc: A.W. Dahlberg (petition and cover letter) 1 1
I l
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\ UNITED STATES OP AMERICA BEFoRE THE NUCLEAR .. RICULATORY COMMISSION , - - l I
)
L ' In the Matter of, l
) Docket Nos. _
GEORGIA POWER CO., l . REQUEST FOR PROCEEDING 8 AND IMPOSITION OF CIVIL PENALTIES FOR IMPROPERLY TRANSFERRING CONTROL OF 4 2 GEORGIA POWER COMPANY'S LICENSES To THE SONOPCO PROJECT AND FOR THS UNSAFE AND IMPROPER OPERATION
' or CECRGIA POWER COMPANY LICENSED FACILITIES To: NennethCarr, Chairman
- U.S. Nuclear Regulatory Commission .
I. Relief Souaht Petitioners Marvin B. Hobby and Allen L. Nosbaugh, hereby request that the Commission, 111 122D13, pursuant to l institute licensing proceedings to determine 1 l 10 cFA 2 104 (C) whether coorgia Power Company ("GPC'): f Illegally transferred control of its licenses 1) to SONOPCO Project (hereinafter 80N0PCC) and the Southern I Company system in violation of 10 C.F.R. 50.80(C); ! 11) Can reasonably assure that s0N0PCC or oth?.T entities operating CPC's licensed nuclear facilities are j complying with HRC regulations; l
^
111) can reasonably assure that SONOPCo's or other t antities' operation of GPC's licensed nuclear I : facilities is not endangering the health and safety of l 5 the publier l iv) Has the character, competence, fundamental trustworthiness and commitment to safscy to operate a . nuclear facility. i 4 4 8 88 .e,GW me og ,, ,
~
e The need for immediate and swift action by the Commission, in light of the seriousness of the allegations
- contained herein, cannot be overstated.
II. Backaround of Petitioner a) Marvin B. Hubby. Petitioner Marvin 8. Hobby, has devoted the last 22 years to promoting the safe and reliable use of nuclear power. In 1940, the Institute of Nuclear Power operations ("IMPO") was created and Admiral Dennis Wilkinson was named as its , y i first President. In April of 1940, Admiral Wilkinson 3 selected Mr. Hobby to serve as INPo's Communications 4 He later became Assistant to the President. In j Manager. 1983 Mr. Hobby became INPO's corporate Secretary as well as 4 4 continuing in his role as Assistant to the President. In 1984, Mr. Hobby accepted a position with the Nuclear l Utilities Management and Resources Cor.mittee ("NUMARC") and
- served as the Project Manager, Congressional Education.
4 . In June of 1985, Mr. Hobby accepted an offer of ! employment f rom GPC's then President Mr. J.H. Miller, Jr. , to serve as his assistant. In 1986, in addition to being f-l Assistant to the President, Mr. Hobby was also named as Assistant to GPc's senior Executive Vice President. In 1987 Mr. Hobby was named GPC's Manager of Nuclear Support. The i following year he served as cPc's Manager of Nuclear Support services. In December of 1988 he was then named as GPC's General Manager of Nuclear Operations Contract
- Administration and Assistant to the Senior Vice President.
2
I Hr. Hobby served as General Manager of Nuclear Operations contract Administration until April of J990, at l ' which time he was forced from the company after attempting to bring to GPC management's attention that it had l improperly transfered control of its nuclear licenses to In this regard, on April 80N0PC0 and the southern company. 27, 1989, Mr. Hobby wrote a highly confidential memorandus (co-signed by GPC's then Senior Vica President George F. A redacted copy of this memo is attached as Exhibit Head). l A. This memo alerted GPC management to the fact that it ) i l appeared that GPC was violating its licenses by improperly l on that tran,sferring control of its nuclear facilities, 1989, a GPC vice day, and the following day, April 28, president, Fred R. Williams, instructed Mr. Hobby to destroy Mr Hobby, concerned all copies of his April 27, 1999 memo. on about his potential liability, sought outside advice, June 8, 1989, Mr. Hobby wrote to Admiral Wilkinson to explain the concern he had regarding his perception that CPC improperly transferred control of its nuclear facilities to the southern company and 50NOPCO as well as CPC's A reaction to his raising the concern to management. redacted copy of Mr. Hobby's letter to Admiral Wilkinson is "S". Before Hr. Hobby could get attached hereto as Exhibit GPC to resolve his concern over the improper transfer of GPC's license to Southern Company and SON 0PCo, he was for Inaenuch as Mr. Hobby was unable to resolve his _ from GPC. concern internally, as he f aithfully tried to do, he is now forced to petition the Commission directly. l l 3 l l l
~
i . J . b) Allen 1. Mosbaugh. _, - . . . Petitioner Allen L. Mosbaugh has devoted the last 14 f, and testing of l years to the safe start up, operations, Mr. Mosbaugh has been 1 commercial nuclear power reactors. l trained and certified as a Senior Reactor Operat':or (8RO) and Shift Technical Advisor (STA), and holds undergraduate and graduate degrees in Nuclear Engineering. Between 1986 and May of 1990, Mr. Mosbaugh served as
.i
- He GPC's Vogtle project Assistant Plant support Manager.
was in charge of a staff of over 400 people in the areas of i technical support, engineering support, security, j administration, training, and quality concerns. Mr. Mosbaugh, until May of 1990, was the Vice-Chairman of the l Plant Review Board ("PRB"), one of four Plant Duty Managers, l I as well as a vogtle project Emergency Director. Nr. l Mosbaugh was removed from the PRB by Plant General Manager f ! George Sockhold after he attempted to resolve safety issues I with the PR8. f Between 1984 and 1986, Mr. Mosbaugh served as GPc's { Pre-operational superintendent and superintendent of 2ngineering services for the vogtle project. In this 4 [ capacity he was responsible for start up and pre-operational testing of Unit 1. He assembled and managed 150 engineers ' and additional support and clerical staff. By April of 1990, Mr. Mosbaugh came to the conclusion l that the highest levels of the SONOPCC project, including R. l _ i Patrick Mcdonald and Joseph M. Earley, were in control of Mr. Mosbaugh the operation of GPC's Vogtle project. 1 concluded that SONOPc0 was needlessly andangering the i i 4 f
I
- 1 l l -.
i public's health and safety byi" refusing to adhere to 1 technical specifications in the interest of schedulet carelessly diersgarding for reactor criticality safetyt operating radioactive waste systems as to be in gross violation of NRC requiremental refusing to report adverse events and conditions to the NRC as required by regulations; submitting falso information to the NRC; repettadiy allowing the Vogtle project to enter Technical Specification 3.0.3
" motherhood" conditions without notifying the NRC or correcting the ejverse condition within the required time 4
spant adopting a policy of intentionally "taking" Licensee Event Reports ("LER's") to keep the Vogtle plant on line and i l i on schedule during planned shut downs and rewarding managers for engaging in non-conservative and questionable compliance practices. III. Facts Petitioners submit the following information in support of their request for proceedings.
- 1. Illegal Transfer of Licenses to SON 0pco.
GPc improperly transferred control of its nuclear operating licenses to The southern Company and to 80HopC0 without first obtaining permission from the Commission to do so pursuant to 10 CFR 50.80 ("No license...shall be transferred, assigned, on in any manner disposed of, either l voluntarily.or individually, directly or indirectly, through - transfer of control of the license to any person, unless the commission shall gise its consent in writing"). Evidence 5
. that GPC transferred control of its licenses to SONOPCO was ,
r obtained as a result of Petiffo'ners' witnessing the day-to- ~~ ! day operation of GPC's nuclear facilities both at the site ! (by Mr. Mosbaugh) and at GPC's corporate offices (by Mr. ! Hobby). . Although Mr. Parley asserts under oath that he is not a l corporate officer of GPC (rather he is an officer of southern Company services and The southern Company), he is ! the 80NOPC0 Chief Executive of ficer ("CE0") . Indeed, GPC's ! Senior Executive Vice President through May of 1990, H. Grady Baker, Jr., acknowledges that Mr. Farley is 8t. M /C0's i de facto CEOs "The appropriate oversight of 80NOPC0 m ists, in that the Chief Operating officer, Pat Mcdonald and tje 3 CEO or not the CEO because its not a corporation == but Farley and Mcdonald are officers of Georgia power Company, , reporting to the president, Bill Dahlberg." {g3, Baker f Deposition Transcript at pp. 16-17 (excerpt attached as Exhibit "C"). As the above quotation also demonstrates, Mr. l Beker was also led to believe that Mr. Farley was an officer 2 of GPC reporting to GPC's President. M. Yet Mr. Farley denies that he is an officer of OPC. See Farley Deposition et p. 10 (excerpt attached as Exhibit D). A thorough review of SONOPCO's operation will demonstrate that SONOPCO's CEO is Mr. Farley, not Mr. Dahlberg. The actual chain of command is General Plant Manager George Rockhold to 80N0PC0 Vice President McCoy; McCoy to SONOPCO's Senior Vice President, George Mairston, , Hairston to SONOPCO's Executive Vice President and Chief operations officer, R. Patrick Mcdonald Mcdonald to 6
4 .
~. l soNOPCO's Chief Executive Of fice'r, Mr. Parley.
l In an April 27, 1989 memo, Mr. Hobby advised GPC Vice 5 President Fred Williams in writing that in the course of j l attempting to perform his function as General Manager, ' 4 Nuclear Contract Administration, he observed that Messrs. \ l Farley and Mcdonald -- not Mr. Dahlberg -- were in control of and were operating GPC's nuclear facilities. kr. Hobby , was instructed to destroy all copies of the memorandum, Thus, since April of 1989, GPC was advised in a confidential ! f memorandum that in the opinion of its contract f l 4 1 I ! Administration Group, GPC had 111ogs14y transferred control of its nuclear licenses to SONOPC0; and soNoPCO's CEO, Mr. Farley, and Chief operating officer, Mr. Mcdonald, were in 4 control of GPC's nuclear licenses in violation of NRC s i regulations. l- 2. GPC misled _the co'nmission about the chain of co==end from t1o_Voatle croiect's Plant i Manaaer to its CEO. I GPC's Executive Vice President of Nuclear operations, l R. Patrick Mcdonald, knowingly made f also statesents to the l NRC commissioners in the presence of GPC's President, A.W. l Dahlberg; Vice President for Nuclear Generation, C. Een McCoys and General Manager of the vogtle Project, George l Bockhold. Yet no one from GPC attending the meeting with the Commission corrected the falso statement made by Mr. Mcdonald to the Commission. i on March 30, 1989, during the course of a transcribed - } proceeding held before the Commission, Mr. Mcdonald was asked by then Commissioner (now Chairman) Carr to state the l i
- 7 l l
" hierarchy between the CEO anif~the plant manager
- of the Vogtle project so Mr. Carr could evaluate his " management concern" he had that the plant manager, Mr. Bockhold, being "a long way from the CEO." Mr. Mcdonald misled Consissioner l
l Carr when he eliminated one entire level of management i l The transcript of between the plant manager and the cEo. f the proceeding demonstrates that Mr. Mcdonald stated that i the reporting chain was General Manager, George sockhold, to Vice President for Nuclear Generation, Ken McCoys who l reported to Executive Vice President of Nuclear Operations, l R. Patrick Mcdonald, who in turn reported directly the the CEO, GPC's President, A. William Dahlberg. A copy of the relevant transcript pages is attached hereto as Exhibit E. . In reality, Mr. McCoy did not report to Mr. Mcdonald but rather to SONCPCo's senior Vice President, Mr. George Hairston, who then reported to Mr. Mcdonald. While it may be conceivable that Mr. Mcdonald may have suffered from a that Messrs. Dahlberg, lapse of memory, it is inconceivable McCoy and Bockhold suffered the same lapse of memory at the exact same time. Messrs. Dahlberg, McCoy and Bockhold should have known that Mr. Mcdonald's statements were false and should have brought this to the immediate attention of the Commission and otherwise corrected the record h3J,g,g,the Commission acted on the Vogtle full power license request. Although GPC eventually corrected a portion of the falso statement, it was not corrected until after the full power , license was granted. Moreover, the correction did not address the fact that Mr. Mcdonald was and continued to e l
- - - ~ - - - - - - -- .- __
reports to soNoPco's (3 facto CEO, Mr. Joseph Farley, rather than to GPC's President. gge 1, infra. - -- 1 l 3. SONOPCo intentionally mis l,ed_the NRC about i the conditici of the Voat;,e Plant after ! a site Aree Emergency in Orser to_nasten the restart of the reactor. , As the Commission is weil aware, a near disaster befell l the nuolear industry when, in March of 1990, the Vogtle nuclear station had a total loss of electrical power while ! the Reactor coolant System was at "midloop" and without containment integrity. Following the station blackout, SONOPCO submitted a Confirmation of Action Response letter (COAR) and a follow-up Licensee Event Report (LER), No. 90-i 006. Both the C0AR and LER contained known falso statements a ! intended to mislead the NRC with falso assurances about the ' reliability of the diesel generator whose failure resulted I in the site Area Emergency. The NRC was advised in the C0AR i j and LER that the back-up diesel generator that failed to start and caused the blackout had been returned to a safe operating and reliable condition. 20NopC0 alleged in the C0AR and LER that the diesel was reliable because it had been successfully started multiple times without suffering a failure, trip or problems. But, as SONoPCC knew, the diesel generator had actually continued to experience an excessive number of trips, failures and problems similar in nature to the failure which led to the March 20, 1990 station blackout. Indeed, on April 10, 1990, Mr. Mosbaugh wrote a memo to the Vogtle General Manager, Mr. Beckhold, and informed him that the diesel air quality statements made in _ the CoAR were false. On April 19, 1990 Mr. Mosbaugh had 9
informed 80NOPCO's Senior Vice President, Mr. George Hairston, that the diesel had suffered trips and' failures. Nonetheless, later that same day $cNOPCO's Senior Vice President, Georg,e Hairston, signed LER 9o-ood-00, after he was advised that the information stated therein contained falso information. On April 30, 1990 Mr. Mosbaugh submitted a me'morandum to Mr Bookhold stating diesel start data contained in the LER was incorrect. On May 10, 1990, Mr. Mosbaugh, acting as the chairman of the Vogtle Plant Review Board (8PRB"), issued an action item to Mr. Beckhold requiring the resolution of the incorrect stataments contained in the COAR. The following day, May 11, 1990, Mr. Mosbaugh was removed from the PRB. An independent investigation into this matter will demonstrate thats a) The statements made in the COA and LER were used by tne NRC to make decisions "significant to the regulatory process." b) The LER wording is falso because it overstated the reliability of the diesel and did not count I numerous failures and problems if the diesel when CPC attempted to start it up. i c) Concern over the accuracy of the data contained in the LER was raised by Mr. Mosbaugh before the LER i was submitted. d) soNoPC0 personnel recognized that the C0AR statements I were falso before submitting the LER; Petitioner Mosbaugh submitted detailed factual ' e) 10
- I)I I !Ql ]p ;],{ "J")j y , .g j .ppf
I i . l information to CPC af ter the LER was ll* ._. . .. submitted which conclusively demonstrated that the [ LER contained false information; f) 80NoPCO intentionally delayed revising the LER
! until after critical meetings with the NRC and 1
Comruission were held on June 8, 1990 (ITT presentation to commissioners);
- g) After the IPP Presentation to the NRC Commissioners, SONOPC0 further delayed correcting j the LER even though QA had already substantiated j the inaccuracies contained in the LER.
h) SON 0PC0 proceeded with actions to submit the revised LER only after Petitioner Mosbaugh continued to ask "why the revised LER has yet to l be submitted" to the NRC (Mr. Mosbaugh raised this j question in a meeting attended by the NRC and the General Manager); i) on June 28, 1990 and June 29, 1990, 30NOPC0 drafted j at least six cover letters to be submitted with the revised LER. Each of these six cover letters states false explanations and were concocted after the fact without regard to the truth. The multiple drafts of the cover letters demonstrate that SONOPCO does not intend to
- advise the NRC why the errors in the LER were actually made, To wit, the different j explanations stated in five of the cover letters
- (
i are as follows: l l 11 : i , 1 4
i
; DATE/ TIME EXPEANATION CONTAINED IN DRAFT i 6-28-90 07851 All tests of diesel Were counted but only valid failures were considered in the conclusion that j no problems or failures occurred I 6-28-90 08:55 All tests were counted regardless i of whether they were valid or not.
j-l 6-29-90 07:55 This draft asserts that the LER i really meant to read ' subsequent to the event" but was inadvertently i worded " subsequent to the test I ~ progras." ; ! 6-29-90 11:42 This draft states that the LER did ! not consider failures and problems associated with the troubleshooting i and restarting the diesel. ] 6-29-90 13:11 This draft states that the error occurred due to poor record keen ng i practices and no definition of the end of the test progrant l
- 3) A review of the performance records of the diesel generator will demonstrate that it was unreliable l and that the statements provided to the NRC were false and intentionally misleading. Indeed, the i
! diesel generator was so unreliable after the site l - ) j area emergency that GPC was eventually forced to l initiate three different design changes to remove ( ' 1 l . or modify numerous unreliable components from the l design and control logic when the diesel ! experienced additional failures after the LER was submitted. l k) The unreliability of hte components that caused the diesel to fail to perform its intended safety ! function when actually called upon to work was .
- known to be unreliable for years and remained i 12
~
uncorrected by the71censeet - ~ ~' i
- 1) SONOPCO retallat d against Allen Mosbaugh by 1
i removing him from the PRR after submitting manorendum to George Bockhold demonstrating that
- his personal presentation to the NRC contained incorrect information and that the LER and C0AR l 1etters contained falso information.
Thus, the diesel was not as reliable as the C0AR and LER conveyed to the NRC. As such, SONOPC0 provided false ! and misleading information to the NRC about the actual reliability of the diesel generator and the actual failure rate of the generator. Thereafter, a cover-up of the ! reliability of the diesel followed and Mr. Mosbaugh's i attempts to correct the falso statements contained in the LER and C0AR resulted in his removal form the PRB. ! By misleading the NRC about the safe operating ! condition of the diesel generator, 80NOPCO demonstated a i complete lack of concern for the safe operation of the Vogtle facility.
- 4. OPC's Execut:,ve Vi_ce President submitted
! periured tecnimony durino the Course of a
- proceedine_9. grime c_ed__under section 210 of the i Enerav Reorcan sa ion Act.
i l i GPC's Executive Vice President, R. Patrick Mcdonald, ! knowingly submitted falso testimony during proceedings comunenced pursuant to Section 210 of the Energy
- Reorganisation Act. In the ruchko & Yunker v. Georgia Power Company Section 210 proceedings, Mr. Mcdonald stated _
under oath at the hearing and in a previous deposition that j the staff of the newly formed SONOPCO project was chosen , 13
e j
"from the top down" (i.e. management picked the.Vice j j Presidents, the Vice Presidents picked the General Managers, ,
t then the General Managers picked their own Managers, then t i the Managers picked their Supervisors, etc.) . , 133 Deposition transcript at pp. 50, 42, and January 4, 1989 l Mearing Testimony at p. 429 (both of which are attached i
- hereto as Exhibit F).
a
- Moreover, GPC's counsel was advised that NY. Mcdonald's
i testimony was falso prior to that counsel's calling Mr. ; Mcdonald as a witness at the Yunker/Fuchko hearings on - l January 4, 1989. In this regard, on January 2, 1989 (two j
- f. days before Mr. Mcdonald was to testify), a meeting of GPC's l witnesses was scheduled by GPC's counsel, the Troutman, i sanders, Lockerman and Ashmore law firm. As Mr. Hobby was one of the witnessas GPC planned to call at the hearing, he l
- was asked to attend and did, in fact, attend this meeting.
j At this meeting, Mr. Mcdonald addressed the group and stated 2 how the 80N0PC0 staff was selected. Mr. Mcdonald stated that the " top down" approach was used to staff the SON 0PCO l l project. Near the end of the meeting, Mr. Hobby informed ! GPC's counsel that Mr. Mcdonald's statement regarding how 1 i the SONOPC0 staff was chosen "from the top down" was f41se. Mr. Hobby also advised GPC's counsel that other statements 1 j made by Mr. Mcdonald were also false. CPC's counsel j l responded to Mr. Hobby's statement that he believed Mr.
; Mcdonald's statements to be falso by advising Mr. Hobby that
) he would have to change his testimony. When Mr. Nobby was , , f instructed to change his testimony, he refused to do so and ] 1 ( : 0 1 14
- s .
I / I
1 k'.
; advised CPC counsel that he.would not cooperate in their attempt to submit perjurious testimony during the course of f the I g Egg and Fuchko proceedinge.
l The following day, on January 3, 1989, Mr. Hobby j advised Mr. Thomas McHenry of the falso statements Mr. j Mcdonald made at the January 2, 1989 neeting. Mr. McNonry ) confirmed that if the statement about the " top down" approach of filling SONOPCO staff positions was made by Mr. I Mcdonald, then Mr. Mcdonald would not be telling the truth. Mr. McHenry advised Mr. Hobby that he had 1st hand knowledge j that the assertion was false. Mr. Hobby further advised Mr. McHenry that he was instructed to change his testimony to coincide with Mr. Mcdonald's but he refused to do so. ggg Affidavit of Thomas McHenry, attached as Exhibit "0". ! Indeed, the false " top down* statement was made by Mr. Mcdonald under oath on December 23, 1988 and January 4, l j 1989, during the course of the section 210 proceedings. ggg Exhibit "M". More troubling is the fact that grigg to l allowing Mr. Mcdonald to take the witness stand at the ! Yunker/ruchko proceed',ngs, GPC's counsel had arranged, l during confidential settlement discussions, that Mr. Yunker's and Fuchko's counsel would not subject Mr. Mcdonald to vigorous cross-examination when he testified thereby
~
assuring that 1) his perjurous testimony would not be
- challenged and, 2) it could be used in subsequent I
proceedings before the NRC. The falso testimony Mr. Mcdonald gave prior to and , during the Yunker and Fuchko hearing was of a critical nature. Messrs. Fuchko and Yunker alleged that they were 15 1 i __ _ _ __ _ _ I
. l 1
~~ ~~
r prohibited from transferring ou't of GPC's nuclear security department and into soNOPCO's organization because they had l
' raised valid safety concerns about GPc's improper handling i
of safeguards materials. In an attempt to demonstrate that i Messrs. Yunker and Fuchko were not improperly kept out of SONOPCC, GPC alleged, through the testimony of Mr. McDonsid, 1 that son 0pco positions could not be filled until a manager ]
) of the security department was chosen because of the " top i
)
down" method routinely employed to fill all positions at l ) 80N0PCO. ggg December 23, 1988 Deposition Transcript of R. Patrick Mcdonald at p. 50, 42 (attached as Exhibit I). l ( ! 5. S0N0pCO routinely threatens the safe operation i 3 or arc's nuclear facilities by allowine them to enter "notnernood.= f 80NOPC0 repeatedly allowed the Vogtle Plant to violate 4 j . Technical specification 3.0.3 and enter " motherhood" without correcting the situation or notifying the NRC. Technical specification 3.o.3 is the "last eschelon of defense" in assuring that sufficient redundancy and margins of safety are maintained for safe plant operation. To wit, under l Technical s'pecification 3.0.3, a plant shut down and NRC ! notification are' required within one (1) hour. The following are some examples where Plant Vogtle entered l
" motherhood" without management appropriately correcting the s
situation or notifying the NRC a) Both Unit l's and Unit 2's A and 5 train safety related load sequencers have been inoperable due to . failure, downpowering and other conditions rendering them inoperable on numerous occasions. 14 1 l
Licensed operators-were not knowledgeable that the - loss of this sequencer resulted in the plant entering " motherhood" and as such the NRC was never notified of the condition pursuant to Technical specification 3.0.3.
- 6. SONCPCO routinely endancers the public's safety l
by len_orino technical soecifications. SONOPCO has made a conscious decision to endanger the public's safety by subverting technical specifications so as to be able to keep the Vogtle Plant operating and/or to speed up the restart of the Vogtle facility. Illustrative examples of the willful and repeated technical I specification violations as follows: a) The licensee willfully and knowingly violated Vogtle Unit 1 Technical spscifications by opening
" Dilution valves" required locked closed by technical specifications. The valves were opened while the Reactor Coolant System was at "mid=
loop,d thus placing the plant in an unanalyzed ' I condition and risking an uncontrolled " dilution
- acoident" and "inadvertant reactor criticality."
The valves were opened to speed the outage so the l l plant could be placed back on line according to i
- the outage schedule. Breaching technical specifications to stay on schedule is undoubtedly due, in part, to son 0PCo's philosophy --
l attributed to Messrs. Farley, Mcdonald, Hairston _ and three soNOPC0 Vice Presidents (it was not 17
attributed to Mr. Dahlberg which further . . , . demonstrates that Mr. Farley and not Mr. Mcdonald controls GPc's operating licenses) -- that outages must be scheduled assuming thatt
. . .everything goes right. Everything falls into place right. That you do not contigency or extra time in there.put .." any (quotation verbatum from vice President McCoy).
The pressure to keep on schedule will necessarily result in managers intentionally breaching technical specifications and "taking" LERs in order to remain on schedule. b) On February 26, 1990, the NRC found that the same dilution valves identified in 6(a) above were again unsecured while at "mid loop" in viciation of Technical specifications. Vogtle senior management wilifully violated Technical specifications again by not "immediately" securing the valves as required by technical specifications because management was too busy due to the outage echedule. Instead, they argued for five hours that locking the valvues was not required and that apapeIIIAgwouldbesufflaient. c) on January 20, 1989 procedural errors made by two shifts of licensed operators miscalculated the shutdown margin for Vogtle Unit 1 which was shutdown at the time. The RCS boron concentration was dangerously low at 1396 ppm and xenon was decaying rapidly (aero percent shutdown margin was 1420 ppa boron). There were no plans to alter RCS 18
boron concentrations.- By pure luck, a-reactor engineer came to the control room and felt uneasy I with the low boren concentration. He recalculated the shutdown margin revealing an error of 3.64. Immediate horation was ordered to averting an inadvertent criticality. Senior plant management clearly realized the gravity of the event by responding in a private meeting that this event could have causer not only a shut down of Unit i but also that it could have interfered with the licensing of Unit 2. Moreover, after these events were brought to Mr. Sockhold's attention, no deficiency was writen, no critique ever conducted, no reivew to assure Technical spooitications were not violated was conducted and no report to the NRC was made. d) on March 22, 1990, GPC employees were told to keep planned shutdowns on schedule by "taking" LER's (i.e. create an LER situation in order to keep the plant running). This practice is indicitive of the philosophy employed by SON 0PCC's CEO, Mr. Farley, and coo, Mr. Mcdonald. 133 4(a) above for the " philosophy" being employed on site. e) Other lesser examples include:
- 1) The Licensee knowingly concealed a technical violation which if recognized would have resulted in a safety-related shut down ,
of Vogtle's Unit 1. This technical violation 19 l . l L .-- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ - _ _ _ _ _ _ _ _ - .
. l concerned the.. failure to properly. test ,
approximately 39 containment isolation valves in violation of technical specification surveillance 4.6.1.1.a. Because the ! surveillance tosts had not been' performed, the valves were to be considered inoperable. [ The licensee had one hour to assure the t valves were in a " safe" condition. The surveillance tests were not completed for two i hours, thereby requiring the shut down of Unit 1 for one hour. Instead, SONCPCC subverted the safety procedure by performing the surveillance tests without initiating a Limiting Condition of Operation, (Leo). Had the LCO been initiated, 80NOPC0 would also have been required to submit to the NRC a Notification of Unusual Event, (NUE), causing further embarrassment since Unit i had to l report a NUE for the same reasons on February 23, 1990. 333 D.C. 2-90-0022;
- 11) SON 0PCC knowingly concealed another l
l technical violation on March 1, 1990 when a ! mode 5 to mode 6 change occurred even though i required equipment wce not operable. The
- _ failure to comply with the technical l specification translated to a 12 hour schedule unhar.cament at a critical juncture; _
l iii) On March 5, 1990, 80NOPC0 knavirgly concealed another technical violation when "B i
- 20 k
i 4 _. . . . . _ . . _ . - _ .__
t train" PHR pump. vibrations resulted in the
- cracking of a Nscw water cooling line. With j the pump vibrating severely and with a failed
! cooling line, the pump should have been .. declared inoperable. At the same time the "A train" RHR pump was drained due to outage- ) related work. Under Technical specification j 3.9.4.1, both trains were not operational. A q I40 and action statement for this condition i ehould have been entered. Had an LCO and j action statement been entered, certain { l
" actions or operations" would'have had to be j suspended. Instead, the pump was not declared inoperable and the LCO was not
, entered. 7. , 20NOPCO repeatally concealed safeguards i problems from tTe NRC. l a) $0NOPCC personnel (including a SOHCPCO Vice l President and SONOPCO Ceneral Manager, and a
- Southern company services Manager) knowingly and repeatedly hid safeguards problems from the NRC and willfully refused to comply with mandatory reporting requirements. Moreover the SONOPCO Vice Presidow nade falsa statements to the NRC during an Enforcement Conference about the status of safeguards materials in Birmingham, Alabama. The false and misleading information presented at the Enforcement conference and other information -
withheld from the NRC are highly significant to 21
i* . O l the regulatory proc _e.ss and were relied upon as a i basis for NRC decisions, which had the NRC had the i benefit of complete, factual information, the Nr.C j would hava, most probably, increased the civil I penalties tros the minimum $50,000 into the j hundreds of thousands of dollars (i.e. 100% increase in the base penalty due to past f l performance, 1004 increase in the base due to I
- multiple events, 50% increase for failure to l report, a 50% increase for no prompt corrective i-action, and an increase due to avillfulness").
- b) On July 23, 1990, Plant and 80NopCO senior
! management prevented the site amourity Manager 4 l from making a Red Phone notification within one
- hour ao required by 10 C.F.A. 73.71. The manager was so prevented hoping to delay or defuse NRC knowledge of programatio problems within s0NCPCC l
(and its design agencies which include southern company services) with safeguard documents.
- 8. SONOPc0 has endangered the nublic's hea?,th _
and safemt hv~operatine rad l,oactive wasi :e systems and fact: . ties Anown~to me : ,n eroes vio ,ation of NRC reauirements. In early 1984, Plant Vogtle's radioactive waste filter system was installed and operated at Plant Vogtle in gross violation of Regulatory Guide 1.143. Although the system was shutdown by Quality Assurance due to programatio breakdowns in procurment and design, in February of 1990, _ SONOPCO approved its resumption eventhough the violations 22 e
- 1 l -
observed by quality Assuranca.had not been remedied. when the Plant Review Board attempted to consider whether a the
- system shold be resumed, vogtle's General Manager, George j Bockhold, intimidated members of the PRS. The end result
- was that the system was returned to service eventhough 10 i C.F.R. 5C.$9 safety evaluations and accident analysis are j inadequate and/or incorrect. As such, a spray of
radioactive let' cage from rubber hoses and plastic pipe used i in this makeshift system can flow unrestricted into storm
- drains which would result in its discharge into Beaver
-i- creak.
- 9. soNoPCO manecement rfMtinelv rasks the safe 4
operation or arc s utoloar rac;.11 ties throu i ign-conservitive ana questionable managemen preo<:icas. l SoNOPCC encourages non-conservative and questionable compliance practices by: a) Praising managers for taking risks.
>} Not taking any adverse action against managers or employees who engage in non-conservative and questionable compliance practices.
c) Nefusing to oritically investigate events or practices resulting in Lens, d) Retaliating against managers who make their regulatory concerns known to GPC and/or 80NoPC0 management. WHEREFORE, Petitioners request that this Honorable commission, aue sponte, institute proceedings to considert 23
'i a) whether GPC illegally transfered control of its i licenses to soNOPCC and the Southern Company System in violation of 10 C.F.R. 50.80(C);
I b) whether there is reasonable assurance.that 50N0PCO's or other entities' cperating GPC's licensed nuclear facilities are oosplying with NRC regulationes ! c) whether thers is reasonable assurance that 80NOPCO or other entities operation of GPC's licensed ! nuclear facilities is endangering the health and safety of the publict and d) whether the licensee has the character, competence, fundamental trustworthiness and commitment to safety to operate a nuclear foo111ty. Respectfully subeitted, ' Mionaal D. Kohn Stephen M. Xohn KONN, KOHN AND COLAPINTO, P.C. 517 Florida Ave., N.W. Washington, D.C. 20001 (202) 234 4663 Counsel to Marvin 5. Hobby and Allen L. Mosbaugh Dated: September 11, 1990 - 65/NRCOP . 24
l. I i FACSIMEE TRANSMITTAL l U.S. XRC - RII ATLA 3TA, GA ! l [_'Q,' $7EV2 YCHN CHECK APPROPRIATE BOX: INPO WEB PHWPS 9/952-6728 492-7142 492-8110 W5fE R N WOODWONT PAYROLL 492-0260 492-7056 492-4371 1RMENG CTR. 2 RE 615/855-6543 346-5324 388-5691 m w Res0ev sre 728-8210 463-3804 (SEE k W any. WMH/A/C7zW DC , OmCE/LOCAMN FAX NO.: #2M42-4/ff VENRCADON: p N0.OFPAGES M + 1RANSWTTM. SEET f ! FROM: . O FAX: 841-4479 / VERIF. - 5510 #l i'O w risjo l
9 DECLARATION OF LARRY L. ROBINSON I, Larry L. Robinson, do hereby declare that the following is true and correct, under penalty of perjury, to the dist of my ability.
- 1. My name is Larry L. Robinson. I en employed as an Investigator with the Of flee of Investigations, United States Nuclear Regulatory Consission. My duties include the conduct of investigations of licensees, applicants, their contractors or vendors, including the investigation of all allegations of wrongdoing by other than NRC esployees and contractors.
- 2. I aske these statements based upon my own personal knowledge, or upon knowledge obtained by se during the course of my employsent, and is relied upon by se in the performance of my official duties.
- 3. The Office of Investigations (01), Region II (Rl!), NRC, currently has two pending investigations regarding allegations of 1.stentional wrongdoing on the part of Georgia Power Company (6PC) Managers at the Vogtle Electric 6enerating Plant (VESP). These investigations basically involve allegations of ;
deliberate Violations of Technical Specifications, and Material Falsa l Statements. If these allegations are substantiated, they could constitute i violations of NRC regulations enacted to protect the public hedith and safety. In addition, a recent Special Inspection, conducted by MRC at VESP during the period August 6-17, 1990, addressed additional related allegations of wrongdoing by EPC Management at VESP that will, in all likelihood, be referred to 01 in the near future.
- 4. On September 12, 1990 Stephen Kohn, of the Law Offices of Kohn, Kohn, and
! Colapinto, telephoned se and advised se that their client, Allen L. Mosbaugh, a SPC employee at VESP, was in possession of audio tape recordings that he, ! Mos5augh, had made of conversations with VE6P Managers th.at may be pertinent to the ongoing NRC investigations /Special Inspectic-. Kohn advised as that Mosbaugh had been officia!!y ordered to turn these tapes over to the Law Offices of Troutaan, Sanders, Lockersan, and Ashmore, representatives of SPC in a Department of Labor (DOLI Case, No. 90-ERA-59, initiated by Mosbaugh. Kohn stated that his understanding was that Mosbaugh was going to have to turn over these tapes on Sept. 13, 1990. S. Allen L. Mosbaugh had been interviewed by se on February 8, 1990, during i the course of my investigation of one of thu aforementioned allegaticas.
- 6. On September 12, 1990, I telephoned Mosbaugh, and he verified that he did aske such tape recordings, that he was in possession of them, that he had been ordered by a DOL Administrative Law Judge to turn them over to the Troutman, Sanders Law Fire. Mosbaugh told se that, in his opinion, some of these tapes show evidence of intentional wrongdoing on the part of 6PC Management at VE8P, and SPC Management at the of fices of SONOPCO Project, Birminghas Alabama, in connection with the allegations in the ongoing O! investigations and the Special Inspection.
- 7. Also on September 12, 1990, in response to my sessage, Hichael Kohn, also with the Law Fire of Kohn, Kohn, and Colapinto, telephoned se at ey residence and advised as that his client, Mosbaugh, per an order from DDL Administrative ;
Law Judge Bernard J. 811 day, Jr., was required to turn over the tapes to the !
3 GPC attorneys by sidnight, September 13, 1990. Michael Kohn said that he had not been able to personally review all the tapes, but that it was his understanding f ree conversations with his client that there was evidence of wrongdoing on the part of GPC Management, pertaining to the ongoing investigation /Special Inspection issues, contained in the conversations on the tapes. Kohn stated that his client would be willing to turn the tapes over to NRC f or review f or evidentiary purposes. Kohn stated that he would prefer to have the NRC subpoena the tapes. i
- 8. The Office of Investigations has reasonable cause to believe that these tapes contain direct evidence of intentional violations of regulatory requirements by SPC personnel that pertains to ongoing NRC investigations / inspections.
- 9. O! has reasonable cause to believe that the review of these tapes by 6PC personnel, or their representativer, prior to the completion of the af oresentioned investigations, would severely compromise the integrity of
- these investigations.
t Further, declarent sayeth naught. Dated this 13th day of September, 1990 at Atlanta, Georgia. 4 Larry . Robinson l l i 4 d i
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se l l September 14, 1990 i BIBLIOGRAPHY
- 1. ;
Memorandum from J.
Subject:
Investigation of March 20, Taylor to the Commission da Demand at Unit 1 Nuclear Power Plant Involving loss of offsite Pow 19 er on 2. Documents Team (AIT) Collected and Provided by Augmented ectionInsp 2-1 2-2 Instruction (Outdoor 230kv) Manual For S & C Circuit Switches 2-3 Overcurrent RelaysInstruction Manual For Westingho 2-4 Differential Protective RelaysInstructioner Manua
& AOS HighInstruction Manual For Brown BoveriAOK ,
Type AOT 2-5 Voltage Current Transformer .
*2-6(124) Differential Voltage RelayInstruction Manual For I i
$2-7 it.-
Nuclear S P3 ant Maiatansnce c rhv..1 'es-une. Wr4k Orcer
*.c .
289004'S ; n:. t i) Y i n- t_A .' 0.t. (Maa n 0..e nine i <- ' J.uc 2)> :.tol D raings 2X3D-AA-A01A (Main One Line - 2-9 Common Units 1Bechtel Drawings AX3D-AA-A01A (Main one 2-10 & 2) - Bechtel Drawings AX3DLO60 ' 2-11 Arrangement) (Switchyard General Bechtel Drawings IX3DH7Al 2-12 Plan) (Low Voltage Switchyard Offsite Source No. 1 C.T. Company Services v /GPC 2-13 Sheet 1) & P.T. Connecting - 2-14 -Electrical System Generator Tripping)Bec ram Southern Company f,rvices/GPC Drawing AX3D BA L57D (500kV PCB Nv. - 2-15 ~ 161520 Close & Trip No. 1) Southern L50T Company 6ervices/GPC Drawing AX9D AA - - 2-16 Line) (500kV PCBr (o. 161520/161620/161660 Single L55A (230kv Offsite Source No. & P.T. 1 C.T. Southe 2-1~ Connecting - Sheet 1) - Southern Company Serviceci ?C Drm ;rn ."'u a l ) l)I de ' L55B -(230kV Offsite Connecting - Sheet 1) Source No. 1 ', .
- r. T .
2-1B , L55C (230 offsite Source No. Southern Company S Relaying) 1 Diff. & Backup 'f s
- Refer to document number in parenth eses g;M 1
2-19 Southern Company Services /GPC Drawing AX3D-AA- ' L50B (230kv Single Line For PCBs 161760/161860/161960) 2-20 Bechtel Drawings 2X3D-AA-B04A (Three Line Diagram 2-21
- Unit 2 AC Generator)
Bechtel Drawings 2X3D-AA-B02A (One Line - Relays & Meters For RATS) 2-22 Bechtel Drawings 2X3D-AA-B01A (Relays & Meters - Generator, Main, & UAT) 2-23 Bechtel Drawings 1X3D-BH-B55B (Res. Aux. XFMR ; INXRB CKT Switcher) ' 2-24 Southern Company Services /GPC Drawing AX3D-AAL50A (500kv & 230kv Sunstation Single Line Index Drawing) 2-25 2-26 Nuclear Plant Maintenance Work Order No. 18906364 Southern Company Services /GPC Drawing AX3D-BA-L52R 2-27 Southern Company Services /GPC Drawing AX3D-BA-L52N Elementary 2-28 Southern Company Services /GPC Drawing AX3D-BA-L52P Diagrams For 230kv PCB 2-29 Southern Company Services /GPC Drawing AX3D-BA-L52Q No. 161860 4-20 Southern Company Ser"icas/GPr nr= wing AX1n-RA-L52S 2-31 Southern Company Se.rviccc/GPC Drawing AX3D-BA-L52H 2-32 Southern Company Services /GPC Drawing AX3D-CA-L72K 23 Bechtel Drawing 1X3D-AA-D03B (4160 V Switchgear 1BA03)
*2-34 2-35 (2-40) Bechtel Drawing 1X3D-AA-D03A (4160 V Switchgear)
Bechtel Drawing 1X3D-AA-D02B (4160 V Switchgear 1AA02) 2-36 Bechtel Drawing 1X3D-AA-D02A (4160 V Switchgear 1BA02) 2-37 Bechtel Drawing 1X3D-AA-D01A (4160 V Switchgear 1NA01) 2-38 Bechtel Drawing 1X3D-AA-D04A (4160 V Switchgear 1NA04) 2-39 Bechtel Drawing 1X3D-AA-D02A (4160 V Switchgear ANA02) 2-40 Bechtel Drawing 1X3D-AA-D03A (4160 V Switchgear i ANA03) 2-41 Bechtel Drawing AX3D-BA-D02C (4160 V Breaker ANA0203) 2-42 Bechtel Drawing AX3D-BA-D03C (4160 V Breaker gea c1.* 7 :
* ~>
3echter Drawing AX3D-BA-D03B (4160 V Breaker ANA0301) 2-44 Bechtel Drawing 1X3D-BA-D01J (4160 V Breaker 1NA0111) oRefer to document number in parentheses 2
;-45 Bechtel Drawing 1X3D-BA-D04D (4160 V Breaker 1NA0412) 2-46 Bechtel Drawing 1X3D-BA-D02C (4160 V Swgr 1AA02
;-47 INCM Brkr INXRB)
Bechtel Drawing 1X3D-BA-D02B (4160 V Swgr 1AA02
- -48 INCM Brkr INXRA)
Bechtel Drawing 1X3D-BA-002D (4160 INCM Brkr 152-
-49 1AA0219 EDG)
Bechtel Drawing 1X3D-BA-D01C (4160 Swgr INA01 INCM
-50 Brkr INXRA) 2-51 GPC Procedure No. 13145-1 " Diesel Generators" GPC
& CAL"Procedure No. 24614-1 " Train B Sequencer ACOT 2-52 GPC Procedure No.
Control Panel Functional Test"27563-C " Generator And Engine 2-53 2-54 Delaval Drawing 09-500-76021 Sh : 2-55 Delaval Drawing 09-500-76021 Sh 2 Delaval Drawing 09-500-76021 Sh 3 Diesel Generator 2-56 Delaval Drawing 09-500-76021 Sh 4 2-57 Engine Control Delaval Drawing 09-500-76021 Sh 5 Panel Schematics 2-58 Delaval Drawing 09-500-76021 Sh 6 2-59 Delaval Drawing 09-500-76021 Sh ' 2-60 Delaval Drawing 09-500-76021 Sh 8 2-51 Delaval Drawing 09-695-76071 "l'ngi r: e Fr.eu.na t i.; Schematic" 2-62 Prints 1X3D-AA-A01A, Rev. 16 1X3D-AA-F27A, Rev. 13 1X3D-AA-F28A, Rev. 14 2-63 1X3DDG020, Rev. 15 w/FCR 10 Mile EP2 Map 2-64 2-65 NOUE ED Checklist Of 3/23/90 2-66 EOF Personnel For 3 Security Emergency Response
/20/90 Organization for 3/20/90 2-67 2-68 TSC Personnel 3/20/90 2-69 CR Personnel 3/20/90 2-70 OSC Personnel 3/20/90 fraining Records EOF 2-71 Communicator Package Consisting Of:
Course Completion & Attendance Records Training Student Handout Dated 6/27/89 Lab / Performance Exercise Guides Lesson Plan RE LP 07001-02 Training Student Handout Dated 4/10/88 Leu.en ??a c6 TT 27001-n1 2-72 Contral Room Lavouts Procedure 10003-C 3
... i r i 2-73 Previous Inspection Report: 50-424,425 89-21 & 89-25 8/25-27/89 Exercise i 50-424/425 88-38 & 88-42 8/15-19/89 ERF Appraisal 2-74 -QA Open Item OQA-87-292 2-75 ' Safety Standard Handbook . i' 2-76 Emergency Director Log 2 TSC Log 2-78 Emergency Notification Messages: 81 ' 42
#3
#4 !
#5 46
#7 28 39 !
2-79 EOF Manager Log 2-80 Met Info 2-81 , Dictated From GEMA To VEGP (FAX 3/22/90) Notification Information 2-82 A= sorted News Dispatcher & Newspaper Articles 2-83 Procedures: 91001-C, Rev. 7 91002-0, Rev. 15 1 91102-C, Rev. 6 91401-C, Rev. 5 91704-C, Rev. 8 2-84 GEK Training Student Handout 2-85 Handbook For General Employee Badge Training 2-86 3/2/90 Test " Data Sheet 4" Of 91204-C & Data Sheet 7 2-87 3/20/90 Notes " Emergency Response" - From Region II 2-88 3/21/90 Notes 3 Pgs. Notes 2-89 8 Hr Report Dated 3/20-90 2-90 3/22 Results Sheet On Accountability & TSC Procedure Verification; 3/22 Restriction On Offsite Interaction - Phone Call; 3/22 BUENN Notes ) l 2-91 2-92 3/22 Interview - Herb Whitener RII I 3/23 Notes CR Procedure Verification i 2-93 RII Plant Note On Capt. T. Nash ! 2-94 Interview Pauline Jenkins - Communication 2-95 Interview Theresa Jones - Communication WE 2-96 Interview Jimmy Cash - Operations Supt.
.- NE 2-97 Interview William Burraister - Plant Duty Manager
?-49 Results Of Selected Sampling Ct E:upivyde 2..
Security' Bldg. l 3
2-99 E Plan Section H-3 2-100 Emergency Facility" " Activities And Staffing of Interview John Hopkins - Shift Supt. Unit 1 2-101 (Interview Incomplete) Interview Capt. Log Form William Johnson Accountability & 2-102 Interview Lt. William Stewart Accountability 2-103 Historical Classification of Emergency & Phone Communication 2-104 Notes - Bob Trojanowski - RII GAO 2-105 E Plan, Section B, "Onsite Emergency Organization" 2-106 ResultsENN Backup of CR Walkdown of Procedure File As Found Description 2-107 Met Bldg. Inspection Notes 2-108 2-109 Listing Of Onshift People For Inspection Drug And Alcohol Screening Report On Donnie 2-110 Willhite Inventory(Fuel Truck Of Fuel TruckDriver) 2-111 Procedure 00656-C, Rev. 0 Traffic And Parking Control, 2-112 2-113 Procedure 90015-C, Vehicle Access, Rev. 8 Procedure Control, Rev.00653-C, 8 Protected Area Entry / Exit 2-114 Procedure 70010-c, Rav. 2 Traffic And Farking Control,
* ?-113 2-116(2-112 Procedure
) Preceriur e 90015-C, Vehicle Access To The PA 00260-C, Control, Rev. 5 Hazardous Substance And Waste 2-117 Procedure Rev. 5 00261-C, Fuel Oil Handling And Safety ,
2-118 The Alvin W. Safety StandardsVogtle Electric Generating Plant: 2-119 Figure 16-1 "Offsite AC" (FSAR)
*2-120(307)GPC - VEGP - Unit 1 Control Log pp 5283-84 2-121 (3/20/90)
GPC - VEGP - Unit 2 Control Log pp 2601-03 i 2-122 (3/20/90) 2-123 GPC - VEGP - Unit 1 SS Log pp 2-124 GPC - VEGP - Unit 2 SS Log pp 5413-14 (3/20/90) VEGP - Unit 2 Trip Report 2875-77 (3/20/90) I
*2-125(33) Proteus Alarm Printout - Unit 2
*2-126(32) 2-127 Proteus DC Alarm Printout - Unit 1 (3/20/90) 1-90-0097 (3/20/90) 2-128 Control Room Layouts (1A D/G Sequencer Prob) 2-129 Procedure 13145-1, Rev.
j 2-130 O " Diesel Generators" i Licensed Operator Electrical Power Training Materials For Loss of 1-121 Licensed Operator Training Materials For Loce O' 3 2-132 Pacidual Heat Removal I Plan ImplementingLicensed Operator Training Materials For
- Refer to document number in parentheses i
5
;-133 Outside Operator Generator Operation Training Material For Diesel 2-134 Procedure3/8/90 Shutdown No. 12006-C Unit Cooldown to Cold 2-135 Procedure Draining No. 13005-1 Reactor Coolant System 2-23-90 2-136 Procedure 3-26-90 No. 13011-1 Residual Heat Removal System 2-137 Procedure No.
Refueling VEGP Entry 00052-C Procedure No. 12007-C 3-26-90 2-138 Procedure No. 12007-C Refueling Entry 3/8/90 2-139 Procedure No. 12000-C Refueling Entry 3/8/90 2-140 Procedure No. 18019-C Abnormal Operating Procedure 2-141 Loss of Residual Procedure No. Heat Removal 3/8/90 19100-C Emergency Operating 2-142 ProcedureNo. Procedure ECA-0.0 Loss of ALL AC Power 3/16/90 23985-1 RCS Temporary Water Level System Emergency Operating Procedure ECA - O>=0.0 2-143 Loss of All AC Power 5/12/89 Procedure No. 54840-1 Installation and removal , instructions for the RCS Temp. Level Indication l Tygon Remova) Tube and the Defeat of the Residual Heat 2-144 Suction valve aute closuie intetloCr; ! 0950076021 Control Panelpp. 9 of 9 Delaval Drawing Engir.c l 2-145 0968876021 pp. 1 of 3 Devalve Drawing ingine and l 2-146 Skid Ele. Schem I 1X4DB111 System No. Rev. 1201 17 P&I Diagram Reactor Coolant l 2-147 1X4DB112 System No. Rev. 1201 24 P&I Diagram Reactor Coolant i 2-148 1X4DB114 Rev. System No. 120825 P&I Diagram Reactor Coolant 2-149 1X4DB115 Rev. 23 P&I Diagram Chemical & Valve No. 1208 2-150 1X4DB116-1 No. 1208 Rev. 23 P&I Diagram Chemical & Valve 2-151 1X4DB116-2 Rev. 16 P&I Diagram Chemical & Valve No. 1208 2-152 1X4DB117 Rev. 1208 19 P&I Diagram Chemical & Valve No. 2-153 1X4DB118 Rev. 20 P&I Diagram Chemical & Valve No. 1208 2-154 1X4DB119 Rev. System No. 120421 P&I Diagram Safety Injection 2-155 1X4DB120 Rev. Systar w. '.224 15 P&I Diagram Safety Injection 2-156 1X4DB121 Pev. 75 System No. 1204 P&I D'.agram Safety Injection 2-157 1X4DB122 Rev. System No.120527 P&I Diagram Residual Heat Removal 6
i
- -158 Loss of All AC/4160V !E Power Lesson Plan
- 2. 3riefing Book document for IIT Leader 2-1 Ltr to W. G. Harston, III fm S. D. Ebneter dtd 2,23/90.. )
Sub]ect: Confirmation of Action Letter < 3-2 PNO-II-90-16,
Subject:
Site Area Emergency at Vogtle Unit 1 Loss Of Offsite Power (3/20/90) , 3-3 PNO-II-90-16A,
Subject:
Site Area Emergency at Vogtle ; Unit 1 Loss of Offsite Power (3/21/90) 3-4 PNO-II-90-16B,
Subject:
Augmented Inspection Team is ; Dispatched to Vogtle Unit 1 (3/22/90) , 3-5 EVENT NUMBER: 18024 (3/20/90)' , l 2-6 DRAFT: NRC Staff Dispatches Augmented Inspection Team i to Vogtle Nuclear Power Plant
. I 3-7 Meeting
Purpose:
NRC ENTRANCE (3/22/90) 3-6 Status Summary 1: (J/20/90) 3 Status Summaty 2: (3/20/90) 3-10 Article: Alert Declared at Vogtle after Truck Hits Tower (3/21/90) (The Augusta Chronicle) 3-11 Article: Outage at Vogtle Means Moment of Fear of Some (3/21/90) (The Augusta Chronicle) 3-12 Questions and Responses ) 3-13 Electrical Distribution Schematic Charts 3-14 VR-1 Update 1500 3/22, Site Area Emergency (3/20/90) 3-15 VR-2 Update 1500 3/23, Site Area Emergency (3/20/90) ) 3-16 Attention - Quarantine List and Licensee Restrictions 3-17 Chart - D/G Testing Unit 1 (3/24/90)
,_ 46 3-18 lst Draft of GPC's Event Critique (uneventful) . Event l
Title:
Loss of Offsite & Onsite AC Power (3/20/90) 1 3-19 Power Level / Mode and L.:gtr- bl. Equip!.unt/.'.br.TJr.a l i System Alignment-(Info) 1 7
I i 3-2 0' Interview ::ctes f(3/20/90) :
- -21 Interview List (3/23/90) l l
3-22 Miscellaneous Notes Kendall l Status of AIT Charter Item Assigned to Rick 3-23 (Itens Nos. 5& 6) (3/24/90) j (3/25/90) notes l w/ Paul Kochery (10: 30 a.m. )
#$ '3-24 Discussion :
fren Ric?. Kendall 3-25 Offsite Commendation 5a 3-26 Onsite Notification 5b 1 Auxiliary Building Radwaste Unit 3-27 GPC, Vogtle Plant,
' Operator Log (3/20/90)
Unit 1 Outside , 3-33 GPU, Vcgtle Electric Generating Plant,. Area Operating-Log (3/20/90) Unit 1 > i 3-34 Prelininary Thermocouple Reading Charts for 3-35 Interview Schedcle (0/?1/90) 3/24/90) 3-36 Interview Ot;.31ule - AIT l Item 1 (3/24/90) 3-37 Warren Lyon - Status Report Notes for Charter Item 3, ; 3-38 Stntus Regarding AIT Charter Item No. Description (3/24/90) (Testa) 3-39 Status Regarding Summer (#4) (Trager) ' 3-40 Vogtle AIT Chart Item 7 Switchyard l 4. Evaluation of Potential Explosion in the Vogt e 4-1 Event Evaluation #1 (3/24/90)
' 4-2 Event' Evaluation #2 (4/10/90) 4-3 Event Evaluation #2 (4/12/90) 3/26/90 Entrance Presentation by Georgia Power Company
. 3.
Subject:
IIT arrive at Vogtle Site
- 6. PNO-IIT-90-02 3/26/90 S- ,
b
1 I
- j Agreements signed by (3) industry representatives on Waiver of Compensation, conflict of Interest and release of
- nvestigation Information for industry participating in IIT i l
- 8. Sulletin Board Notice u- 3gt 9. Letter to Document control desk,USNRC from W. G. Hairston III, Sr. V.P., Nuclear Operation, Georgia Power Co., dated 11/30/89. Subject Vogtle Electric Generating Plant Hardware Modifications pursuant to generic Letter 88-17
- 10. Letter to document control desk, USNRC from W. G. Hairston III, GPC, dated 2/2/89.
Subject:
Plant Vogtle - Units 1 and 2, NRC' docket 50-424, 50-425, Operating License NPF-68, Construction Permit CPPR-109 Response to Generic Letter 88-17
- 11. Letter to document control desk, USNRC W. G. Hairston, III,GPC, dated 12/29/88:
Subject:
same as #10
- 12. Sequence of Events Chronology of Site Area Emergency 3/20/90 (Received 3/27/90) from license 13 Letter to W. G. Hairston, III, CPC, from A. 3. Scrdt, MRC , 1 dated 7/31/89.
Subject:
Notice of Violation (Inspection j Report Nos. 50-424/89-19 and 50-425/89-23) s/NRC Inspection j Manual Temporary Instruction 2515/101 1
- 14. Local newspaper coverage - March 27, 1990 ,
- 15. Interoffice Memo from G. Bockhold, Jr., Plant Manager, to Vogtle Site Personnel dated 3/27/90.
Subject:
Vehicles in Perimeter Area
- 16. Entrance Meeting with Licensee and Personnel Statements (3/26/90) 16-1 Entrance Meeting Notes 16-2 J. Hopkins - SS 16-3 R. B. Snyder - SS 16-4 P. Vannier - RO 16-5 K. Jones - CRO
- 17. Order to quarantine
- 18. Letter to C. C. Miller, Mgr. of Engineering, Vogtle, from W.
C. T.0 7s e'r . Tr., dated ?/16/90.
Subject:
Vogtle, Units 1 & 1, final Rerponse to Request for Engir.eering Assistance. Attacnment: Loss of becay Heat removal Analytical Studies for Vogtle 1 & 2, A response to GL 88-17 9
i 1 4
)
5 19, Training' Student Handout'No. GE-HO-88002-00-001-C Continuing ( Training--RHR Mid-Loop Oper" )
;3. Training "esson. Plans: 'I 20-A Continuing Training--RHR Mid-Loop Oper. No. GE- ,
~P-88002-00-C 20-B Emergency Diesel Generator Auxiliaries Fuel Oil System No. .
NL-LP-11202-01-C' I 20-C Emergency Diesel Generator General' Overview No. , NL-LP-11201-00-C 20-D- Emergency Diesel Generator Auxiliaries No. NL-LP- ! 11203-02-C l
- 21. EOP No. 19100-C, Revision 4, ECA-0.0 Loss of All AC Power 13427-2, 122, 4160V AC 1E Electrical Distribution, Procedure No. ,
Revision 5 18031- l 23. uoss of Class 1E Electrical System, AO Procedure No. j C, Rev3sion o i l
' 4.
Eoron In]ectior. Flow Path Verification - Snuccown, Procedure ! No. 14406-1, Revision 3 !
.25.
Generator and Engine Control Panel Functional Test Procedure ) No. 27563-C, Revision 2 ) j "A-TRAIN UNDERVOLTAGE TEST" Expiration
- 26. T-ENG-90-11, Rev. 1, l
. Data: 4/8/90 f B-Train Undervoltage 1
- 27. Temporary Procedure No. T-ENG-90-12, I
)
Test f
- 28. Temporary Procedure No. T-ENG-90-13, Sequencer Operability Check
- 29. Temporary Procedure No. T-ENG-90-14, Unit One Train B DCP
'88-VIN 0070 Sequencer Functional Test
- 30. T-ENG-90-15, Unit one Train A, DCP 88-VIN 0070 Sequencer Junctional Test ,
' ' 1. Maintenance Work Order (MWO) 19001576, 3/28/90 (D/G 1A)
*(;;;
*32. Proteus Alarm Printout..(M?)
- Refer to-document number in parentheses 10 i
J
t 23. Proteus Alarm Printout (U2) 34. List of= Quarantined Equipment (Revised 3/29/90 Rev. :)
- 35. Personnel Interviews 35-1 K. Pope - SS 35-2 W. Burmeister - Unit Superintendent 35-3 N. Dewbre, P. Jenkins, 25-4 F. Thompson - EGS; R.Moye T. Jones
- ESS
- Shift Clerks 35-5. G. Bockhold - PM ;
35-6 D. Vineyard - SS 35-7 35-8 W. F. Kitchens - Ass't. PM D. Hines, D. Daughhetee, E. Pickett, J. 35-9 Stanley J. P. Cash - OS 35-10 T. C. Eckert - Oper. Dept. !
'3 6.
List Of quarantined equipment (Revised 3/29/90 Rev. 3) l
- 27. Personnel interviews
*37-1 W. Johnson, W.
Stewart - Security 37-2 M. Lackey - Outace Planning Mgr. ) 37-3 M. Lackey, R. Barlow, J. l 37
- J. Roberts - EP D'Amico - Schedulinc .
37-b H. Handfingar - Maintenance Mgr. 37-6 G. 6tenenoorg, I. Kochery
- 38. PNO-IIT-90-02A i
39. Training Lesson Plan No. NL-11204-OOC Emergency Diesel Generator-Engine Control and Protection 5/11/89 40. Procedure No.6/30/89 Transformers, 13415-1,Rev.6, Reserve Auxiliary 4
- 41. Training Lesson Plan No. LO-LP-28201-09-C Sequence Operation, 7/26/89 ;
- 42. EOP No. 19101-C, ECA-0.1 Loss of all AC Power Recovery without SI Required 43.
- Training Lesson Plan No. NL-11205-01C, Emergency Diesel Generator Control and Protection, 8/29/89
. 44. Training Lesson Plan No. LO-LP-11102-05-C Ft.orgency Diesel Generator Auxiliaries A !"
St?rt System, 12/8/89
- 45. '
Training Lesson Plan No. LO-LP-11001-06-C,
-Emergency Diesel Generator Introduction and Overview, 12/11/89
- SAFEGUARDS DOCUMENT NOT TO BE RELEASED 11
,s. .
4 6 '. Personnel Interviews 46-1 E. Dannemiller, D. Huyck - Security 46-2 J. D. Jiles - Safety Specialist 46-3 R. Berry - Security 46-4 S. Chestnut -Training 46-5 K. Stokes - Sr. Plant Eng.
,J. Frocedure No. 13426-C, 4160V AC Common
-than IE Electrical Distribution System, 1/26/90
- 48. . Training Lesson Plan No. LO-LP-11103-06-C Emergency Diesel Generator Auxiliaries:
Combustion air and exhaust, 2/28/90 4 9. . Training Lesson Plan No. LO-LP-11104-C, Emergency Diesel Generator Auxiliaries Lube Oil and Crank Case Ventilation, 12/8/89 n
- 50. Training Lesson Plan No. LO-LP-11105-08-C, Emergency Diesel Generator Auxiliaries Jacket l water cooling system, 12/8/89 :
;1. Training Lesson Plan No. LO-LP-11101-07-C Diesal Generator Auxiliaries ruei 011 System l 14/e/69
- 52. Letter to J. P. Kane, GPC, from h.C. Reasey (unsigned and undated).
Subject:
Response to REA VG-9010, Loss of decay heal removal
- 53. Emergency Response Facilities Input List, Revision v7.05, 12/11/86
- 54. Procedure No. 14406-1, Revision 3, Boron Injection Flow Path Verification-Shutdown, 2/6/89
- 55. VEGP Standing Order No. 1-90-05, Emergency Boration Flow Path, 3/1/90
- 56. Memorandum for A. Chaffee from S. Ebneter, undated (received 3/30/90).
Subject:
Designated Regional Point of Contact
- 57. Procedure No. 17038-1, Rev.7, Annunciator Response Procedures for ALB 38 on EAB Panel, 3/11/90 ;
I 5 8. - Procedure No. 13c11-1, Rev.18, RHR System, 3/11/90 l
- 59. *lehicle Access Request, 3/20/90
- 60. .AOP) Frocedure No. 18019-C,Rev.7, Loss of RHR,3/16/90 12
- 61. PRB Comment Review Sheet (PRB-90-44) for Temporary Procedure No. T-ENG-90-14
- 52. 3PC VEGP Handbook for General Employee Badge Trainingc GE-HO-00101-001-C, Rev. 5, 10/23/89 w/ record of training dates
'for D. Willhite
- 63. Event Report No. 1-90-003, Additional Support Items .
- 64. SPDS Checklist i i
- 65. Items on Fuel Truck 3/2-30/90
- 66. Photographs ,
66-1 Roll 1 - (3/24/90) t 66-1-1 Voltmeters 66-1-2 Auto XFML No. 2 5 66-1-3 Auto XFMRs Nos. 1&2 l 66-1-4 Electrical Panel ' 66-1-5 Annunciators and Electrical Panel ! 66-1-6 Electrical Panel and Annunciator j
.66-1-7,8 Letdown /Chg. Flow and Bit Press 66-1-9 Accumulator Pressure TanXs 1 and 2 00-1-10 Accumulator pressure,' Tanks 3 an.d 4 66-1-11 PROTEUS Computer Display 66-1-12 RWST Reset Switches (2) and RHR Suct Vent Line TRN-B 66-1-13 RHR to HL, RHR Suct Vent Line TRN-A, RWST Reset 66-1-14 RHR X Train A & B Outlet and Bypass 66-1-15 RHR Pump Pressure Trains A & B 66-1-16 Incore TC 66-1-17 Operator Aid for mid-loop on RCS Loop 1 Hot Leg NR Level, and RCS Loop 4 Not Leg WK Level 66-1-18 SI Pump Disch Trains A & B 1
66-2 Roll 2 - (3/24/90) 66-2-1,2 Plant Safety Monitoring System 66-2-3 Control Rod Position 66-2-4 RCS Flow Trip Alarms 66-2-5 PRZR Pressurizer, PRZK Spray, PRZK LVL 66-2-6 RCS Press, RCS HL Temp 66-2-7 .RCS CL Temp OP delta T, OT Delta T, Delta T 66-2-8 Chg Fl ow ~. RCF Ienos 1-4, Delta T 66-2-9 Press, LTDN Flow, RCS Luop 4, RLs lemp Loop , a, ue1La T 66-2-10 RCS Loop 4, RCS Temp Loop 2, Delta T
'66-2-11 Safety A, RCS Loop l'HL Press, RCS Temp Loop 1, Delta T ,
13 :
i RCS Loop 4 (Delta T, OP Delta T, OT Delta T, 56-2-12 & RC Flow Loop 4) T-AVG 66-2-13 RCS Loop 4, & RCS Flow Loop 4 66-2-14 RCS Loop 2, & RCS RCS Flow Loop 2 Flow Loop 1 RCS Loop 1, Delta T 66-2-15 Loop 1 HL Press, RCS Temp Loop 1, 66-2-16 RCS PRZR Relief Temp, RCS Loop 1 66-2-17 66-2-18 In Core TC 66-2-19-23 Electrical Panel and Annunciators
- Truck 66-3 Roll 3 - (3/25/90) 66-3-1,2 Rear View Blind Spot Assessment (180 to 195 feet) 66-3-3-6 66-3-7 Left View 66-3-8 Fuel Can Closeup of event-related damaged area 1 66-3-9 Closeup of event-related damaged area 2 66-3-10 66-3-11,12 Closeup of event-related damaged area 1 Closeup of event-related damaged area 2 66-3-13 Fire Extinguisher 66-3-14 Fire Extinguisher 66-3-15 66-3-16,17 Front View (Driver's Door Open) ;
66-3-1R Inside Cab inside) l Front Windshield (fenm i ht 46-3-19
'6 ' 20-23 Blind Spot Assessment through laft and r g l
l mirrors l 66-3-24 Lett side view I Lee, J. Aufdenkampe, W. F. l
- 67. Personnel Interview of G.
Kitchen, W. Burmeister & D. West 68. Diesel Generator Failure Analysis l D. DeLoach, J. Jackson, and S. i
- 69. Personnel Interview: Whitman !
I
- 70. Photographs l 70-1 Roll 4 (3/24/90) l 70-1-1 Operator Aid - RCS Loop 1 HL NR Level, and l RCS Loop 4 HL WK Level and Containment 70-1-2 SI Sump Disch Trains A & B, Press 70-1-3 Letdown /Chg. Flow, BIT Press 70-1-4,5 Proteus Computer: In Core TC 70-1-6 Accumulator Level RHR. Pump Pressure iO-1~/ - Tne p/3r.,90) 72-2 Roll 5 - Photos provide by Licen=ct ,
Welder 70-2-1-3 Side View 70-2-1-4,5 Front View 14 1
-- _1
- 71. ' Meeting Attendance Record
~1-1 I T Entrance (3/26/90) l 71-2 Diesel Generator (3/28/90) ,
71-3 Diesel Generator (3/30/90) l 1
- 72. Access Record History, 3/20/90 73.- Badge l Record - 3/30/90 l
74.- Evaluations of Initial Plant Conditions - J. D'Amico
- 75. Procedure No. PTDB-1 Tab 8.0, Rev. 2 - Pictorial Aids: RCS Elevations and Mid Loop Level Instrumentation (3/2/90) ]
l
- 76. . Procedure No. 18015-C, Rev. 5, Loss of Instrument Air j l
- 77. Procedure No. 12000-C, Rev. 16, Refueling Recovery 3/8/90 l 1
- 78. Procedure No. 12006-C, Rev. 15, Unit Cooldown to Cold Shutdown 3/8/90
~9. Procedure No. 12007-C, Rev. 14, Refueling Entry (Mode 5 to Mode 4). 3/8/90
- 80. Procedure No. 13 005 -2 . Rev. 10, Reactor Coolar.t System Draining, 2/23/90
- 81. Procedure No. 17006-1, Rev. 11, Annunciator Response Procedure for ALB 06 on Panel 1A2 on MCB, 3/6/90
- 82. Bechtel Drawings 1X3D-AA-G02 C, Rev. 6, Vital Inster. Distr. Pnis. I AX3 D- AA- A01A , Rev. 13, Main One Line 1 & 2 )
AX3D-AA-F19A, Rev. 10, 480V Motor Control Center I 1X3D-AA-G05B, Rev. 14, 120V AC Non-Class IE key Instr. AX3D-AA-HO1A, Rev. 9, TSC 125V DC/120V AC Non-Class IE Distr. Panels AX3D-AA-A01A, Rev. 16, Main One Line Unit 1 AX3D-AA-G02B, Rev. Main One Line Non-Class IE 125V DC & 120V Ess. AC System
- 83. MWO 18906587, SG#4, Unit 1, 12/23/89 8 4 '. Procedure No. .25270-C, REv.6, SG Nozzle Dam Checkout,
!nstallation and Removal 12/16/88 15
- 35. Precedure No. 18004-C, Rev. 6, AOP RCS Leakage, 12/9/90 36.
'AOP) Procedure No. 18006-C, Rev. 2, 3/4/88 Fuel Handling Event, 87 (AOP) Procedure Cooling No. 18020-C, Rev.3, Loss of Component Water, 2/14/90 38.
(AOP) Procedure No. 18021-C,Rev. 4, Cooling Water System, 3/16/90 Loss of Nuclear Service
- 39. Procedure No. 18034-1,Rev. 1, Power, 9/12/89 Loss of Class IE 125V DC 90.
(AOP) Procedure No. 18028-C, Rev. 3/16/90 7, Loss of Instrument Air, 91. (AOP) Procedure Shutdown Panels, No. 18038-1, Rev. 10, Operation from Remote 8/29/89 92. Radiation Monitors Status 93. Relief Request 7 and 11
- n]ection System No. 1204(with accompanying documents) Safety
- 94. Piuceaura No. 00350 C, (12/27/89) Rev.19, Work F.esucst Progra:.
95-1. NWO #1887~ 146 3/31/90 95-2. 4WO #1906.339 3/31/90 96. Section 6.0, Administrative Controls (Vogtle 1 & 2) 97 Information on Critical Safety Function Status Frees on SPDS (on EXF computer)
- 98. Training Records: A. L. Blalock, J. D. Williams, W. P. Stephens, W.F.M. Redivannz, Watkins, W. Hennessy, l J. W. Covington, R. LeGrand, D. Haile, G. J. Durrence 99.
Information Ops on Maintenance Personnel Training on Mid-Loop 100. i Procedure No. 00400-C, Pav. 11, Plant Desi,a Control, 2/24/90 l l 101. Request for "ngiran i.g " .4. w 'o System 500 kV - reced' Ire Mc. 004 00-C) of I 16
4 l 102. Relaying Data Sheet - Gen. No. 2 Main Bk. Primary 103. Sequence of Events W/ source of information 104. 2.0 Hardware Configuration 105. Vendor 9/9/86 Document Status Sheet-Manual Change No. 76021-9, 106. Security Department Report No. 3941-30 ~ 107. Joint News Release - 3/20/90 108. MWO (continued) No. 19001576 109. T-ENG-90-11, Rev. MWO's 1, A-Train under voltage Test w/related 4 110. lA Diesel Generator Reference Material 111. 3/20/90 Logs: Unit 1 & 2 Shift Supervisor; 1 & 2 Control ' 112. Memo to G. Bockhold from G. R.
Subject:
Fredrick, dated 9/27/B9. VEGP-1 and 2 QA Audit finding Report 350 i
)
411 Release #1 and # 2 - 3/20/90 1 114. OSC Log 115. MWO 18906328, 3/27/90 116. Attandance Roster - Diesel Generator Mtg. (3/31/90) 117 Personn21 Interviews: 117-1 K. Ex1y 117-2 P. Humphrey 117-3 J. Williams, D. Gustafson, G. McCarley 118. Operational Journals, 3/20, w/ Situation Chart 119. MWO TimingNo. 19001684, of Two (3/31/90) Trips During LOSPDiesel Generator - To Verify 120. . ersonnel Interviewsi 120-1 G. 320-2 R. Schnieder (EP) e .d 9. Thr at* '*RF-EPC) Dorman - Training 17
r
; I 120-3 M. Cagle - Maintenance 4 120-4 S. Driver - Training 120-5 D.
120-6 E. Willhite - Truck Driver J. Kozinsky - SS 120 7 S. Young - PFS
*21. Design Change Request No.
i Closure - Train A & B Sequencer Panels 88-VIN 0070 w/ Design Change Package 122. Operation and Maintenance Instructions (SFSS) w/ drawings i j
) 123. Vogtle Operating Records - VPO401GP l i
124. MWO 19001684 - 3/31/90 MWO 19001576 - 3/31/90 3 MWO 28900466 - 3/31/90 125. PROPRIETARY DOCUMENTS: . W FIELD SERVICE PROCEDURES
- 125-1
- MRS 2.2.2 GPC-1, Post Activity Sign-Off for Area Cleanliness 125-2 {
i 125-3 MRS 2.2.2 GPC-21 Rev. O, Nozzle Dam Hydrotest 425-4 MRS 2.2.2 GPC-72 Rav n, Nor:le Dam 125 5 MRS 2.2.2 GPC-23 Rev. O, Nozzle nam Manual HR3 2.2.2 GPC-24 Rev. O, Nozr.le Dam, Leak Decacticn Manual 126. Procedure No. 29536-C, Cutage Management Program (3/4/88) 127. Procedure No. 27505-C, Rev. 2, i Opening and Closing Containment Equipment Hatch (7/18/88) I 128. Procedure No. 10013-C, Rev. 6, Writing EOPs from the Westinghouse Emergency Response Guidelines (9/22/88) 129. Procedure No. 29537-C, Rev. 1, f Outage Scheduling (4/11/89) 130. Procedure No. 01000-C, Rev. (6/9/89) 1, Management of Outages 4 , 131. Procedure No. 10011-C, Rev. 13, Operations Procedure Preparation and Review Guidelines 4 (7/5/89( 4 132. Procedure (9/26/89) No. 10018-C, Rev. 11, Annunciator Control
'52.
-oceO:rc MO. 1.2000-{, Rev. 16, Conduct of Operations (3.3f9n) -
*NOT TO BE RELEASED 18
l l 1 134. Procedure No. 11011-1, Rev. 7, RHR Removal System Alignment
'4/1B/89) 135. Temporary Change to Procedure Form (TCP) No. 13001-1-12 i 1, RCS Filling and Venting (Expiration Date: 4/6/90) l
)
136. Procedure No. 14230-1, Rev. 4, AC Source Verification l (7/27/88) ' i 137. Procedure No. 91403-C, Rev. 4, Site Evacuation (12/6/88) 138. (EOP) Procedure No. 19101-C, Rev, 8, ECA-0.1 Loss of All AC l Power Recovery Without SI Required (7/26/89) l 139. Procedure No. 19111-C, Rev. 8, ECA-1.1 Loss of Emergency Coolant Recirculation (7/26/89) ! 140. Procedure No. 13001-1, Rev. 12, Reactor Coolant System Filling and Venting (10/10/89) 141. System Block Diagrams - SFSS 142. BeChtel Drawings: ! 1X4DB113, RTD BY-PASS REACTOR CO3L7PT SYSTEM NO. 1291 1X4DB100, P& ID'S AND FLOW DIAGRAM IIGEND AX3AE03-9-10, ELECTRICAL SCHEMATIC ZX3AE03-10-10, ELECTRICAL SCHEMATIC 143. Unit 1 D/G Trip Sensor History w/ Licensee's Draf t Analysis of 1A Diesel Shutdown 144. PNO-IIT-90-02B - 4/2/90 145. Transcript: Briefing Meeting - 3/28/90 146. Personnel Interviews 1 - J. Acree - SS 2-J. Aufdenkampe - Tech Support 3 - F. Pope - EO 4 - S. Owyoung 5-J. Ealick 6 - M. Cagle 7 - M. Lackey (3/30/90)
'! .e t i."q !!T and Licensee Personnel re Diesel
}$ 147. Transcii f t:
Malfunction (1/10/90) 19 l l
,,; ;h148. Transcript: Meeting w/ Event Critique Team - 3/31/90 149. Personnel Interview - D. Hines - PE Sfg150. Transcript: . Discussion re Results of Testing on A-Diesel Performed on_3/30/90 :
151.' Memo for A. Chaffee from G. Zech, dated 4/3/90. i Lessons Learned: Vogtle Site Area Emergency
Subject:
,,jp152. Memo from M. S. Briney to G.
l Calcon Temperature Bockhold, switches Jr., dated 4/3/90.
Subject:
}'
153. PROPRIETARY DOCUMENTS-(INPO)* Operating Experience Close Out Packages for the following: 'I' SOER 85-01 ' SER 17-88 SER 73-83 SER 42-84 OEMR 272 i SOER 85-04 O&MR 365 SOER 88-03 SER 5-80 I SER 26-89 SER 36-88 l SIR 2-87 I SER 35-86 ! SER 31-86 ! SER 23-86 SER 17-86 ' 154. Calcon Switch Information ' 155. Quarantined Equipment List, Rev. 4 - 4/2/90 156. Meeting 4/2/90 Attendance Record - Diesel Generator and IIT Exit - 157. Procedure Test (2/5/90)14980-1, Rev. 18 - Diesel Generator Operability 158. Letter'to Director, I&E, NRC, from B. C. Guntrum, Mgr., QA, IMO.
Subject:
' Notification of a Potential Defect in a component of a DSR or DCP.*, ."tendF', D/c .' t '2 ? /A R )
*NOT TO BE RELEASED 20
~59. 7HS Video Tape Recording: RHR Mid-Loop OPS 160. Photegraphs 160-1 Roll 6 - (3/31/90) i i'
160-1-1,2 TSC 160-1-3-7 D/G 1A CK Control Panel 160-1-8,9 D/G 1A CK Annunciator Panel 4 160-1-10 PSMS Codes 1 160-1-11 Plasma Display and Keyboard of PSMS 160-1-12,13 Proteus Computer 3
?
4 160-1-14-16 NSCW, CCW 160-1-17,18 NSCW: Return Temp, Basin Lvl. & Hdr. Press. 160-1-19 NSCW A Flow: Supply and Return , 160-1-20,21 Misc. Status Lights for NSCW ! 160-1-22,23 CCW PMP-1, 3 : 160-1-24,25 Rx MW Wtr. to CCW ' f 160-1-26,27 CCW Train A Surge Tank 160-1-28,29 CCW Cnat. Spray, & SI Panels 160-1-30,31 RHR & CVCS, & Accum. Panels ' 160-1-32-33 RW9T Ieve) 360-1-34 Page 1 of D/G Procedure (13145-1) a 160-2 Roll 7 - (3/33/90) , 160-2-1-5 Jacket Water Pressure Sensor 160-2-6-9 Pressure Sensor 160-2-10,11 160-2-12-14 Lube 011 Pressure Sensors (3) Lube Oil Temperature Sensor ' 160-2-15-25 Jacket Water Temperature Sensors (3) 160-2-26-31 Turbo Lube Oil Sonsor 160-2-32-34 Emergency Break Glass Start 160-2-35 Break Glass Hammer vs Flashlight Size 161. List of Vogtle 1 Event Headquarters Operations Center Participants 162. History of Quarantined Items 163. Quarantine Sign 164. Unit 1 Second Refueling Outage Cha.rts: ! 164-1 Electrical System Work lC ; < ~3-1 Target vs. Actual Critical Pe*3 21
^
I l l l c
,_ _151- 1, First Drain to Mid-loop
~~(3p 164-4'
~
IR-2 Target vs. Actual Critical Path i 7 .05 D/G Activities /TRN B l'o4-6 D/G Activities /TRN A 165. Unit 1 First Refueling Outage Charts: I 165-1 \ Mid-loop Work and Support 165-2 Electrical System Work 165-3 D/G Activities /TRN B 165-4 D/G Activities /TRN A ; 166. Photographs ; 166-1 Roll w8 - (3/31/90) i 166-1-1 D/G Control Panel 166-1-2 166-1-3,4 D/G Control & Annunciator Panels 166-1-5 D/G Annunciator Panel SI Signal 166-1-6,7 Emergency Start 166-1-8 Procedure Sec. 4.4.3.1 166-1-9 Procedure Sec. 4.4.3 166-1-10 Jacket Water Lv. In. 166-1-11,12 Jacket Wther In 166-1-13,14 Generator Bearing Tamp. i 166-1-15-17 Day Tank Level > 166-1-18 D/G Control Panel 166-1-19-21 Local Remote Control Switch 166-1-22-25 166-1-26 O/G Back Panel (Inside) 166-1-27 Gaitronics Phone 166-1-28,29 Commercial Phone [ 166-1-30 Sound-Powered Phone P. 1 of Procedure 13145-1 166-1-31 Lube Oil Press 166-2 Roll #9 - (3/30/90) 166-2-1-7 Sensor Points on Blackboard. in Large 166-2-8-10 Conference Rm (LCR) 166-2-11 Electrical Configuration of Vogtle Plant 166-2-12-19 Blackboard Note to IIT 166-2-20-34 Sensor Points on Blackboard in LCR Sensor Points on Blackboard in LCR showing steps to de-eaergize Sequencer , on Panel l 4 t z 22
166-3 Roll 410 - (3/26/90) 166-3-1,2 i 166-3-3,4 New Restricted Area Sign and Rope D/G Panel 166-3-5 D/G Annunciator Panel 167. Procedure No. Recording 00053-C D/G 1A - Temperature Monitoring and (4/4/90) i 168. Meeting Transcripts: u-. 3k 168-1 Telecon Personnel. between IIT, Licensee Personnel, and RII ' 3gt 168-2 D/G Meeting (4/2/90) I 169. ! Technical Specifications - Vogtle Unit 1 - Sections 1-5 170. 1985 Sensor Maintenance Switches Work Orders (MWOs) - History, Unit 1, D/G-171. e 1936 MWOs - History, Unit 1,'D/G-Sensor Switches 1 172. 1987 MWOs - History, Unit 1, , D/G-Sensor Switches t 2 '; 3 . 1988 MWos - History, Unit 1, D/G-Senant Switches
;,;. 1989 MWCs - History, Unit 1, D/G-Sensor Switches a
175. 1990 MWOs - History, Unit 1, D/G-Sensor Switches 176. 1988 MWOs - History, Unit 2, D/G-Sensor Switches
- 177. ERF Computer Points 178.
D/G Temperature Switch Calibration Data Received from Licensee - 4/6/90 t 179. PROPRIETARY DOCUMENTS - WESTINGHOUSE
- 179-1 Core Map - Instrument Locations 179-2 Reference Loading Pattern 4 180. D/G (1A/1B) Start Logs 181. Ltr to R. Newton, WOG, fm A. Thadani, NRC, dtd 12/11/89.
Labject: Filled, Loss of RHRS WCAP-11916, July Cooling 1988, While the RCS is Partially Activities and Other Related WOG
=NOT TO BE RELEASED ;
23
*B2.-II Bulletins I
182-1 No. 80-12 (5/9/80) l 182-2 No. 86.01 (5/23/86) 183. Information Notices 193-1 No. 80-20 183-2 No. 80-41
*83-3 'No. 81-09 183-4 No. 21-10 183-5 No. 86-39 183-6 No. 86-74 183-7 No. 86-101 183-8 No. 87-01 183-9 No. 87-23 183-10 -No. 87-59 1 183-11 No. 89-41
*B3-12 No. 89-67 184. Memorandum for T. Murley, NRC, and E. Beckjord, NRC, fm E.
Jordan, NRC, dtd 5/18/87.
Subject:
Loss of Decay Heat Removal Function at Pressurized Water Reactors with Partially Drained RCSs 185. NURE6/CR-5915 ; BNL-NI1RFG-5?l21, Improved Reliability of Residual Heat hemoval capability in PWRs as Related to Resolution of GI-99 l 186. Case Study Report AEOD/C503, Decay Heat Removal Problems at U.S. PWRs, 12/85 187. Documentation of Special Test of RHR Lineup to be Used with Fath to Loops 1&2 only 188. MWOs on S/G Manway and Nozzle Dams 188-1 No. 18906589 188-2 No. 18906581 188-3 No. 18906579 188-4 No. 18906580 l 188-5 No. 18906582 188-u No. 18906588 188-7 No. 18906590 188-8 No. 18906587 4 99 O, aerator Aids - Local Control Stations 24
e l l h 190. l Ltr to USNRC GPC, dtd 8/15/89. Document Control Desk fm W. G. Hairston, III, I
SUBJECT:
VEGP LER 50-425/1989-023 I i 191 1 Selected Licensed Operator RHR Training Materini l 192. Selected Licensed operator Annunciator Training Material 191. . Selected PEO, D/G Training Material: Lessons Plans & . Response to Annunciators l i 194. Procedure No. 13431-1, Rev. 4, Distribution System (4/20/90) 120V AC IE Vital Instrument ! 195. Presentation to Region II NRC on Vogtle Site Area Emergency, March 20, 1990 (4/9/90 - by Licensee) j t 196. Generic Letters i i I 196-1 ) No. 87-12, Loss of RHR While the RCS is Partially ' rilled (7/9/87) - 196-2 No. 88-17, Loss of Decay Heat Pemoval (10/17/88) 197. Memo for J. P. Stohr, NRC, fm A. T. Boland, NRC, dtd 4/3/90, suc]cct. Critique cf RII's Response to the Vogtle Incident 198. . Appendix A to NRC TI 2575/103, Supplemental Information, Containment Closure and Reaccor Coolant System Level Instrumentation (2/26/90) 199. Letter to R. C. Jones, NRC, from L. A. Walsh, WOG, dated 4/6/90;
Subject:
WOG Transmittal of Information Copy for Loss of RHR While and Background Documentation Operating at Mid-loop Conditions Guideline
,, p 200. Transcript: Telephone Conversation w/IIT, Licensee, R-II (4/5/90) 201. NRC Personnel Interviews: 4/5/90 201-1 A. Thadani, NRR 201-2 W. Lyon, NRR-l 202. . Unit 1A Train D/G - Air Peceiver Dew Point Measurements ,,, jg ;03. : Transcript: Telephone Conference w/IIT, Licensee, R-II tA/F ,a, r .
) 1
- 25-
i 204. Personnel Interviews: 4/6/90 204-1 L. Walsh, Seabrook (telecon) 204-2 D. Marksberry, NRC/AEOD 204-3 J. MacKinnon, NRC/AEOD 204-4 P. Ray, NRC/AEOD 204-5 D. Ross, NRC/AEOD "' jj' 205. Transcript: Telephone Conference w/IIT, Licensee, R-II (4/7/90) _. b- 206. Transcript: Telephone Conference w/IIT, Licensee (4/9/90) 263. NRC Personnsl~ Interviews: 4/9/90
~
207-1 W. Hodges, R-I 207-2 G. Zech, AEOD 208. 2E 2nformation Notices 208-1 No. 83-56 208-2 No. 84-42 209 Prelicinary D/G Instrument Test Outline (4/11/ vu)
; 10. Foilures of Calcon Temperature & Pressure Sensors at Vogtle 1&2 211. NRC Information Notice No. 88-36
- gfhI 212. Transcript: Telephone Conference w/IIT, Licensee, R-II (4/10/90) 213. NRC Personnel Interviews: 4/10/90 213-1 R. Jones, DST l l 213-2 R. Eckenrode, HFAB 213-3 D. Tondi, ESB 214. Graph: Times To Core Uncovery Following Loss of RRR - During Mid-Loop Operations (4 loop W Plants Including Vogtle) - H. Ornstein (3/29/90) 215. Memorandum for M. W. Hodges, NRR, From J. E. Rosenthal, AEOD, dated 6/29/87;
Subject:
Additional Information on Loss of DHR at PWRs with Partially Drained RCSs l 26 .
e 1 l 216. Memorandum for K. Kniel, RES, From J. E. Rosenthal, AEOD, dated 8/28/87;
Subject:
BNLs 8/18/87 Presentation of the Results of Draft Report " Improved Reliability of RHR , Capaoility in PWRs as Related to Resolution of GI-99" ! 217. Memorandum 2rd T. E. Murley, NRR, from E. S. Beckjord, RES, dated 4/4/88;
Subject:
GI-99 Risk Assessment Results
-- ;g fk 218. Event Report #1-90-003, Loss of Off-Site and On-Site AC Power (1E); Date of Event: 3/20/90 219. Data Sheets - MWO 19000092 220. Data Sheets - MWO 19000093 221. Commitments and Commitment Change Reports for D/G Preventive Maintenance: Report Nos. 9023, 9034 through 9038, 9052, 9082, 9086, and 13764 222. Data Sheets - MWO 19000094 1 223. Data Shee".s - MWO 19000095 224. Letter to Director, IE/NRP, f"nn E. R. Slock, CA Engincer, IMO, dated 5/12/83;
Subject:
Supp).ement Letter of 4/29/88 Additional Information tc l 225. Operation and Maintenance Manual, Appendix VII, Alarms and Safety Shutdowns (pp. 8-8, 8-8A) 226. Procedure No. 29101-C, Rev. 1, Emergency Lighting Surveillance (FSAR Fire Protection Surveillance), 12/5/89 227. Procedure No. 00414-C, Rev. 7, Operating Experience Program, 11/29/89 228. Procedure No. 11885-C, key. 10, D/G Operating Log, 5/11/89 228-1 Data Sheets, DG1A - 3/20/90 228-2 Data Sheet, DG1B - 3/23/90 229. Videocassettes - GPC, Vogtle, March 1990 D/G Tests 230. Bechtel Drawings: 230-1 Lube Oil Piping Schematic 1X4AKO1-27-11 220-2 Starting Air Piping Schtm.' tic '.X'AKn1 79 1? 27
- .... - - . - . . . - ~ -. - - . - . - - . - . - ~ . - - .. - - . - - i e [ L 230-3 230-4 Jacket Water Piping Schematic 1X4AK01-26-11 i 230-5 '4160V INCM.BRKR 152-1BA0319 (DG 18) 1X3D-BA-D03D 4160V INCM.BRKR 152-1AA0219 (DG 1A) 1X3D-BA-D02D ,,
'231. Operating Experience Close-Out Packages for the following:
IN 88-36 IN 86-101 IN 87-23 IN 80-20 : IN 84-42 t IN 89-64 !
'GL 87-12 IN 89-67 i
!E Bulletin No. 80-12 232. PROPRIETARY DOCUMENTS (INPO):*
i SER 60-83 , SER 79-64 ! ga j (- 233.. Transcript: Telephone' Conference w/IIT, Licensee (4/11/90) ia*. NRC Personnel.Interviaws: 4/11/00 234-1 S. Snan'(man, NRR 234-2 234-3 C. McCrauxen. NRR {' N. Ornstein, AEOD 235. Maintenance Work Request (MWO 19001563 malfunction (communications failure) wa)s recognizedindicating when 236. Procedure 1,1-3KJ,03,Rev.1)- to Test Annunciator Including First-Out (VEGP-237. VEGP Plant Review Board Meeting Minutes Which Include Discussion of Plant Condition on 3/20/90 238. Note to A. Chaffee, IIT,-from R. Jones, RSB, dated 4/11/90;
Subject:
Documents Requested by IIT 239. MWO No. 19000909 (Data Concentrator) 2/22/90 240. Drawings: 240-1 Field Change Requests; Reference Drawing Numbers: i 2X4ANO1-05* R/8 - Eng. Cont. Pnl. Scha. 2X4A401-498 P/8 - Elec. Scha. { 2A4AK01-361 R/8: ! 2X4AK01-369 R/8 1
*NOT TO'BE RELEASED I I
28 i
'l
240-2 VEGP Procedure No. 50009-C, As Built Notice No. 89-V1MO70A001 (1/18/90) 240-3 Bechtel Drawings: 1X4AK01-42-11 Engine Control Panel Installation 1X4AKO1-52-9 Engine Control Panel Schematic 1X4AK01-357-9 A.C. Schematic 3 Y.42.K01-3 58-8 Control Schematic 1X4AK01-443-4 Engine Pneumatic Schematic 1X4AK01-458-7 Instrument Identification Schedule 2X4AK01-367-8 Engine Cont si Panel Schematics (2) 2X4AK01-428-5 Engine Pneumatic Schematic 2X4AK01-459-6 Instrument Identification Schedule 241. Receiver 3/30/90 Air Pressures observed by Plant Personnel on ! 242. History (Unit 1 - From First RHR During Second Refueling Outage to Initiation of 3/20/90 Event) Defining " Windows" of Operational Configurations Potentially Affecting the NSSS Containment, and Supporting Systems , 243. Licensing Document Change Request Pkg. No. FS 88-99-Rev. 1 2e4. Description Fault Recorderof Electronic Data Reading EkT, Protcus and ,, g 245. Transcript: (4/12/90) Telephone Conference w/IIT, Licensee, R-II 246. NRC Personnel Interviews: 4/12/90 246-1 K. Brockman, R-II 246-2 S. Ebneter, R-II
,-- 246-3 F. Rosa, NRR 247.
Outage Scheduling Information 248. RAT Scheduling Information 249. PRT, RCS, RWST, SG, TS Information 250. Letter to C. K. McCoy, GPC, fror J. L. Tain, Westinghouse, dated 3/16/90, w/RHR B Pump Vibration Data 251. MWO No. 1o0014 S, 1/40/90 (D/0 ini 29
;' : i t
252. MWO No. 19001433, 3/20/90 (D/G 1A) : 253. MWO No. 19001537, 2/25/90 (D/G 1B) 254. MWO No. 19001576, 3/28/90 (D/G 1A) . 255. MWO No. 19001684, 3/31/90 (D/G) ' 256. !
~
D/G Function Checkout Data 9P 257. Transcript ; V/IIT, (unmonitored tape recording): Teleconference Licensee, R-II (4/3/90) !
., 9hg258. Transcript:
(4/13/90) Telephone Conference w/IIT, Licensee, R-II ! i l 259.' Transcript: . (4/16/90) Telephone Conference w/IIT, Licensee, R-II
,, 260.
GPC Personnel Interview: (4/17/90) R. Odom, G. McCarley, M. Sheibani !'
*'HS Audiocassette:
i Ior D/G (4/18/90) Calibration Tests en Temperature Sensot 2 34 . i Meeting Attendanca Records:
-(4/17,'D0) ; IIT Exit (4/18/90) IIT Entrance Briefing t
263. MWOa: 263 .t No. 18906592 - Containment Equipment Hatch
\
i 263-2 (12/23/89) ; No. 18906593 - j Personnel Hatch (12/23/90) 264. Calcon Temperature Switch Response Test on 4/4/90 265. Note 4/13/90;to D.
Subject:
Gustafson and K. Burr from R. Jones (GPC), Piping - 1A Diesel Vibration Readings on Temperature Switch 266. Drawings: As Built Notices No. 00517 ! No. 00121 I No. 00122 l 30
_ _ _ _ _ - _ _ . _ . _ _ ~ . - . . _ _ _ ._...._.. _. _ _ ._ . _ _ __ ...._ _ _ . _ . ___ - - _ ..-.. _ _
.. i t
i i 267. Bechtel. Drawings: One Line Diagrams i 1X3D-AA-E17A,'Rev. 6 - 480V Switchgear 18816 1X3D-AA-E07A, Rev. 7 - 480V Switchgear 1B807 1X3D-AA-E06A, Rev. 6 - 480V Switchgear 1BB06 ! 1X3D-AA-E10A, Rev. 9 - 480V Switchgear 1NB10 1X3D-AA-D03B, Rev. 9 - 4160V Switchgear 1BA03 l 1X3D-AA-D03A, Rev. 8 - 4160V Switchgear 1BA03 ! 1X3D-AA-E01A, Rev. 11 - 480V Switchgear 1NB01 j IX3D-AA-E04A, Rev. 6 - 480V Switchgear 1AB04 ; 1X3D-AA-E05A, Rev. 8 - 4807 Switchgear 1AB05 ; 1X3D-AA-D02B, Rev. 6 - 4160V Switchgear 1AA02 l 1X3D-AA-D02A,.Rev. 8 - 4160V Switchgear 1AA02 i 1X3D-AA-E16A, Rev. 5 - 480V Switchgear 1AB15 , 1X3D-AA-A01A, Rev. 16 - Unit 1 l 268. Deficiency Tracking System : 269. l l Containment Penetration Control Package (Procedures and l Surveillance Task Sheets) ' 270. Letter T. to All Holders E. Murley, NRC/NRR, of RO and 11/1/88r detad SRO Licences for PWRs from
Subject:
Operator Diligence While in Shutdown Conditions 271. Letter Lo H. P. Allen, Southern California Edison Co., T. E. Murley, NRC/NRR, dated 12/2/88; from-Decay Heat Removal
Subject:
Loss of 272. Calibration Data Sheets for Pressure Switch 273. Annunciator Response Procedures for ALB-09 273-1 Panel-1C1 on MCB 273-2 Panel 2C1 on MCB 274. ERF Computer Points (continued) 275. Training Student Handout - RHR System 276. Annunciator Response Procedures for ALB 36 on EAB Panel 277. Conoseal Status and Description at Time of Site Area Emergency on 3/20/90 (Note: _ontains proprietary data *) Drawing No. 1X6AB02-288-1 l 4;2. Procedurn No. 00000-C, Rev. 5 - Hazardous Subsua..-c. a-a Wacea Centrol (10/24/C;;
*NOT To BE RELEASED i 31 '
I l i 279. ! Containment Building Penetrations Verification - Refueling i 279-1 Procedure No. 14210-1, Rev. 4 279-2 Procedure No. 14210-2, Rev. 2 ; 280. Bechtel Drawings: Containment Building Piping Areas l 231. Westinghouse Drawing: PROPRIETARY
- Safeguard Actuation Systes -
282. Drawings: ! Electrical, control Systems, Mechanical, Nuclear Pressurizar Press Control i Steam Dump Control ) Rod Controls Steam Generator Trip Signals
.283. Security Vehicle Log -3/20/90 284. NRC Personnel Interview: J. Calvo, NRR 285. MW0s:
235-1 230-2 No. 19001482 - D/G 1B 285-3 No. 19C015e4 - D/G B l Mo. 19001677 - D/G A 286. Dsficiency Cards: 186-1 286-2 No. 1-88-3016 - D/G A 286-3 No. 1-88-3083 - D/G B 286-4 No. 1-88-3453 - D/G 1A No. 1-90-0182 - D/G 1A 287. Information Notice No. 90-25: Loss of Vital AC Power with Subsequent RCS Neat-Up 288. Interoffice Ajiuni, GPC,Memorandum to C. K. McCoy, GPC, from M. J. dated 10/16/89;
Subject:
Vogtle's Operations Experience Program Assessment 289. INPO Database KEYWORD; DIESEL 290. Vogtle Status of NRC and Reviews as of 4/18/90 INPO Operating E..perience Doicument 291. Bechtal D"Muinos R=lsted *e Lighting
*NOT To BE RELEASED i
- 32
i ( 292. Bechtel One-Line Diagrams: i i ' Non-Class 1E Distr. Panels 480V Motor Control Center ! 293. Bechtel Drawings: 1K4-1208-486-01, Reactor Head Vent . 1K4-1201-064-02, RCS 2X6AB02-66-4, Reactor Vessel General Arrangements 294. VHS Audiocassette Tape: D/G Local, control Room, Sequencer, TSC Local; Counter Nos.: 0000-363 Procedural Steps to Reset Sequencer (3/29/90) 0364-969 DG1A Sensor ** (3/31/90) - Lube Oil High Temperature Switch, Lub 011' Low Pressure Switches, Jacket Water Temperature Switches, Turbo Oil Pressure Switch, Jacket Water Pressure Sensor and Governor 0970-1004 Back of DG1A Control Station 1005-1380 Tested Emergency Lighting j 1381-1507 Scanned DG1A Control Panel '
-1508 Headset and Extension Cord
-1530 Normal and Emergency Jacks for Cord 1550-1614 Overvi.w vf DG1A noom 1615-1S68 I". side of DG1A Control Panel Showing sensor lines 1Gow-1.wi0 Front of DG1A control Panel Showing Emergency Start Button Using Break Glass Instrument (or Unscrew Glass) 1811-2207 Walkthrough of Emergency Start Procedure, SOP No.
13145-1, Section 4.4.3 (2043) Operator Aid Needed: Trips on Front Panel 2208-2837 First out Walkthrough, Noting Annunciator Trips Provided (2289) Operator Aid: Stand for Procedures (2423) Required by ARP to Log First Out and Report to Control Room (2544) Tested Annunciators Using Annunciator Response Control Buttons (2740) Must Report All Major Functions and Malfunctions (DG Trips, First out According to Admin. 10000 Procedure; also 13145, 14980); If Tripped, Logged and Reported to Engineering 1 1 33 l
o 0 2838-2925 Front of DG1A Panel; Magnetized Procedure Book; Ties on First out Discussed (Two Valid Trips Simultaneously); Hight Not Work one Reactor Trip is Eliminated or 2926-2993 Communication Phones: Gaitronics, Commercial Headphones, Sound Powered, and Extension Cords (25 feet) 2994-3311 Technical Support Center 3312-3430 DG1A Annunciator and Centrol Panels (3329) Sticker on C-9 for MWO and Another i Sticker (3400) First out on Reactor Trip Shown (3429) First out Response Control Buttons for Reactor System 3431-3500 Safety Parameter Display System (SPDS)
-3505 PROTEUS
~3517 ERF 3526-3804 Walkthrough of Control Boards (3535) Plant Status Monitoring System (RVLIS., ]
Incores, etc.) on Plasma Ecreen j (1621) Relevant Indications
+
(3636) Reactor Vessel Level and RVLIS, Scannad (3680) NCSW System (3703) CCW system (3732) RHR Systcm (3800) Tower Plaffarm in Control Room 3805-3844 Jacket Fct6c Temperature Sensor in I/C Shop 3845-4179(3855)Unit 1 Control Room (4/2/90) Description of Response Control Buttons in Control Room for DG1A: ACK, RESET , TEST, Including Contrast of What Happens (4130) at DGIA Local Control Station Reactor Trip; First out Described and Shown 416F " 4. 7 Turbine Trip (on Back Panel); First out Explairled, Including Response Button Controls; Electrical 3 Hydraulic Control Cabinet Shown i (4290) i First Hit Reset and Electrical Malfunction Explained 4440-4934 DG1A Control Station (4625) How Clears to clear Itself F6 (Switch Not in Auto)J 4680-4878 Summary of What Operator Has to Do to Clear , Audible Response Condition; Alarm Goes Solid (If ; in series, Silence, Acknowledge, Reset) (4878) How Operator Knows Condition is Cleared 493* END w- --, - . - - , -
P 295. Spare Temperature Sensor Calibration Data 296. Test Procedure No. 17133 297. Security Access Printout for M. Lackey and S. Chessnut (3/20/90) 098. Letter to C. K. McCoy, GPC, from J. L. Tain, Westinghouse, dated 4/20/90;
Subject:
Vogtle, Unit 1, Reactor Vessel Head, Thermocouple AssembliesVent Rate Through 299. Calcon Temperature and Pressure Switch Data (3 Volumes) 300. Letters to NRC Document Control Desk from W. G. Hairston, III, GPC, dated 4/19/90; Subjects: 300-1 VEGP LER, Loss of Offsite Power Leads to Site Area Emergency 300-2 VEGP LER, Unit 2 Reactor Trip from Unit 1 Reserve Auxiliary Transformer Feeder Line Fault 301. Security Logs for Entering and Exiting Control Room, Containment, D/G Room (3/20/90) 302. PROPRIETARY DOCUMFNTS (TNPO)* Operating Experience Closeout Package: for the Following: IN-81-09 SER-5-83 ' SER-38-83 SER-76-84 SER-36-87 SER-5-89 SER-74-81 SER-78-81 SER-87-8 OEMR-295 SER-15-87 SER-26-89 303. Wyle Laboratories Test Results - Temperature Sensors (submitted 4/26/90) 304. Response to Questions Level, Gravity Feed Regarding Containment LCO Logging, RCS
*NOT TO BE RET.FASED .
l 35 !
,t ,*
305. Event Data Collection System - Equipment Locations Sketch 306. Wyle Laboratories Test Results - Temperature Sensors ' (submitted 4/27/90) 307. Unit 5287) 1 Control Log for 3/19/90 through 3/21/90 (pp. 5281- i , 308. VEGP (3/2/90Plant Review through Board Meeting Minutes Prior to the Event 3/19/90) 309. Wyle Laboratories Test Results - Temperature Sensors (submitted 4/30/90) 310. MWO for Head Removal / Replacement 311. Core-Exit Thermocouple Channel Calibration 312. Steam Generator Drawing 313. RHR Pump B Modification Data 314. Wyle Laboratories Test Results - Temperature Sensors Icith.mittad S/1/90) i 315. Test Precedure for As-Received Testing und Calibration of Sevne by WyleCalcon Model A3500-W3 Laboratories for GPC Temperature Sensors (4/30/wo; - l 316. VEGP Procedure No. 93240-C, Rev. BT, Reactor Vessel Assembly / Disassembly Instructions 317. MWO No. 18905286 - #036 Check Valve 318. Letter 5/2/90; dated to GPC,
Subject:
Attn. K. S. Burr, from Wyle Laboratories , Calcon Model A3500-W3 Temperature SensorsReliability Evaluation Test 319. Information Notices 319-1 No. 82-20 319-2 No. 86-09 319-3 No. 83-17 319-4 No. 83-51 i 319-5 No. 85-28 I 319-6 No. 85-73 319-7 No. 85 01 319-8 No. 84-69 319-9 No. 84-e9, supplement 1 l
.6
319-10 No. 86-73 319-11 No. 88-75 319-12 No. 89-87 320. IE Bulletin 77-01, Pneumatic Time Delay Relay Setpoint Drift 321. Trip IE Circular Lock-out77-16, Emergency Diesel Generator Electrical Features 222. MWO for Accumulator Isolation valve 323. Note to W. Lyon, IIT, from J. F. D'Amico, GPC, dated 5/2/90; Re: Information on Seal Table 324. Pressurizer Detail Drawing ! 325. MWO for Charging Line Check Valve 036 and 035 326. Wyle Laboratories Test Results - Temperature Sensors (submitted S/4/90) . i 327. Bechtel Drawings of Accumulator Isolation Valve 328. Clearanew bheet for Reactor Head Vent 320 Tnstruction State & Operating Anan*ncistor SystemsManual Series X12 & X16 Models Solid 330. PROPRIETARY DOCUMENTS (INPO)
- 1 i
1 Operating Experience Closeout Packages for the Following: SOER 83-01 SOER 81-10 SOER 83-06 O&MR 97 O&MR 110 O&MR 334 SER 82-78 SER 82-079 SER 09-85 i SER 25-85 i SOER 86-003 SER 89-028 SOER 85-001 SER 84-72 (also includes 82-008 and 84-007) SED 88-031 "Nuf TO BE RELEASED 37
. .= ,
331. Operating Experience Closecut Packages for the Following: IN 77-01 IN 80-41 IN 82-20 IN 83-17 IN 83-51 IN 84-69 (and Supplement 1) IN 85-28 IN 85-73 ' IN 85-91 IN 86-70 IN 86-73 IN 88-75 IN 89-87 332. Wyle Laboratories Test Results - Temperature Sensors (submitted 5/7/90) 333. Operating Experience Closecut Package for IMO Report ! 334. Deficiency Card No. 1900125 - S/G Manway 335. Bechtel Drawings: Swing Check Valve 3-C88; Motor np. Gate Valve Med. 10001GM39FNM010 336. Draft " Corrective Actions for site Area Emergency" and Unit 1 Status Report from 3-18 to 4-1-90 (submitted by licensee) 337. Famorandum 5/11/90;
Subject:
to H. Wyckoff, EPRI, from J. O'Brien, EPRI, dated Extreme Environmental conditionsNuclear Plant Worker Capabilities Under 4
-p4 338. Letter to NRC Document Control Desk, from W.
III, GPC, dated 5/14/90:
Subject:
G. Hairston, Vogtle Electric Generating Plant Corrective Actions for Site Area Emergency 339. Cooper Industries Test Results - Pressure Sensors and Shutdown Logic Board (submitted 5/15/90) 340. Motor 1ABF Control Center Load Lists for 1NBS, 1NBI, and 1ABA to 341. Wyle Laboratories Test Report (Preliminary),
" Reliability W3 Evaluation Temperature Testing of Ten Calcon Model A3500-dated 5/12/90, Sensors" qc 342. List of P0rsonnel On Site on March 20, 1990 38
o i s' - l 343. Procedure No. 00150-C, Rev. 10, Deficiency Control (5/10/90) 344. Photographs - Switchyard Where Incident occurred (Affected Pole, Fuel Truck, Fallen Insulator) 345. D/G 1B Troubleshooting Plan; Procedure 22981-C, Calcon t Pneumatic 5/24/90) Temperature Sensor Calibration (telecopy received i 346. Unit one "B" D/G Sensor Testing Sequence of Events - Preliminary (telecopy reewived 5/24/90) 347. Unit One "B" 5/29/90) D/G Sequence of Events (telecopy received 348. Telephone Conversations Center on March 20, 1990 at NRC Headquarters Operations 4 348-1 348-2 Unmonitored Transcript Memorandum from E. Weiss, AEOD to C. Siegel, IIT, dated May 31, 1990;
Subject:
Transcript of t Telephone Conversatiens Relevant to the Vogtle Event ! I 149. ! GPC Interoffica Enrrespon.dencs from J. U. Swart:velddr to Department Heads dated 1/13/89; Re: Deficiency Contral Eid QA Audit Unit 1 and Surveillance Finding Trend Report for Vogtle . 350. Test Report No. 17133-1, Ten CalconforModel Laboratories A3500-W3 GPC (May 12, Temperature Sensors," by Wy 1990) 4 351. Letter to M. W. Hodges, NRC, from R. A. Newton, HOG, dated November 21, 1988, with enclosure, WCAP-11916 352. NRC Inspection Manual, Temporary Instruction of Decay Heat Removal" (GL 88-17) 2515/101 , " Loss 353. l NRC Inspe.ction Manual, Temporary Instruction I of Decay (Long Heat Term) Removal" (GL 88-17) " Programmed2515/103 Review" Enhancements
" Loss i l
l 4 39
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nFrICE or THE SECRETARY
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CORRESPONDENCE CONTROL TICKET PAPER HUMBER: CRC-90-1000 14GGING DATE: Sep 17 90 i ACTION OFFICE: EDO AUTHOR: Michael D. Kohn ATTILIATION: DC (DISTRICT OF COLUMBIA)
- LETTER DATE: Sep 11 90 FILE CODE:
SUMECT: Request for proceedings and imposition of civil penalties ACTION: Appropriate DISTRIBUTION: Records, Chilk SPECIAL MANDLING: 2.206 PETITION
- NOTES
- TREAT AS A PETITION UNDER 20 CFR 2.206 AND FOR ACKNOWLEDGEMENT PER D&SB THRU MR. CHILX AND OGC DATE DUE:
! SIGNATURE: i DATE SIGNED: AFFILIATION: i ) [ . i (I) I e
OFFICE OF THE SECRETARY CORRESPONDENCE CONTROL TICKET , PAPER NUMBER: CRC-90-1000 IDGGING DATE: Sep 17 90 1 ACTION OFFICE: EDO AUTHOR: Michael D. Kohn AFFILIATION: DC (DISTRICT OF COLUMBIA) l l LETTER DATE: Sep 11 90 FILE CODE:
SUBJECT:
Request for proceedings and imposition of civil ! penalties ACTION: Appropriate DISTRIBUTION: Records, Chilk i SPECIAL HANDLING: 2.206 PETITION NOTES:. TREAT AS A PETITION UNDER 10 CFR 2.206 AND FOR ACKNOWLEDGEMENT PER D&SB THRU MR. CHILK AND OGC DATE DUE:
; SIGNATURE: . DATE SIGNED:
AFFILIATION: EDO --- 005836 , fe. llig3 . I
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I BULL & ASSOCIATES ! l
. Certified Shorthand Court Reposters Cosaputerised R4 pordag l since 1966 OctoberLl', 1990 ]
i J 1 Jesse P.'Schaudies, Jr., Esq. : 127 Peach.treo' Street, N.E. l
'400 The , Candler Building .)
Atlanta,;GA. 30303 ; RE: -UllITED STATES DEPARTMENT OF-LABOR -f l MOSBAUGH VS. GA.' POWER CO.' i C. A. FILE NO. 90-ER.*.-58 , 1
' DEPOSITION OF: ROBERT PATRICK MC DONALD TAKEN ON 9/17/90 1 I
Enclosed is your copy of the' deposition, which was stenographically reported in the above-captioned matter. ; d Also, enclosed is the jurat page from the original deposition. It' is requested'that the deponent read the deposition for accuracy, making'certain the court reporter correctly reported the testimony. If thereLare corrections to typing or spelling, et cetera, these should be noted on a separate sheet of. paper, indicating the page
- and line-numbers on which...these appear. , , ,,
'In addition to the foregoing, it is requested that the personally
~
> subscribed jurat page, together with.the errata sheet, if any, be returned to the office of Bull & Associates, within the next
- thirty-day period. We shall, upon receipt, include same with the j
' original deposition.
i
'Your. assistance and cooperation in.this matter is appreciated.
~
* \
Sincerely,/ i ULL & . Enclosure-Copy of deposition g/-
-a IAJ S.
< Original jurat page f
, /p ay1 l '
CC: Michael D. Kohn, Es * !
.. Administrative istant l
. hd
. O 4651 Ro.wein' Road, N.E. FAX (404) 848-3616 0 125 Haberahaan Drive .
. Suite F504. . , WATS 1 (800) 447-8855 Suite D E l
- Atlanta, Georgia 80848 " * . Fayettevitic, Georgia 30214 . [. .j
.-(404)256-8886 0 7 For Retura Informauon (404) 460-9774 ^ ( .i
A}a; UNtTED STATES
/psafspq',, NUCLEAR REGULATORY cOMMisslON
, , , 3* '** REGION 88 e
# 101 MARIETTA STREET,N.W.
' E ATLANT A, GEORGI A 30323 i k **"* 007 01 1990 l Docket Nos. 50-424, 50-425 License Nos. NPF-68, NPF-81 EA 90-129 ,,5a,1 s s 7of,,, Georgia Power Company ATTN: Mr. W. G. Hairston, III / kg 8
- - Senior Vice President - @ 6 Nuclear Operations v P. O. Box 1295 O //'
i Birmingham, AL 35201 k #'
&l Gentlemen:
SUBJECT:
ENFORCEMENT CONFERENCE SUK. J (NUREG-1410 AND NRC INSPECTibi. httuRT N05. 50-424/90-16 AND50-425/90-16) l' This letter refers to the Enforcement Conference held at our request on September 5,1990, concerning activities authorized for your Vogtle facility. 1 This meeting was requested to discuss numerous items identified by the Incident
- Investigation Team which was chartered in response to the Site Area Emergency j event of March 20, 1990.
1 i The circumstances, root causes, corrective actions, and safety significance of i three areas of concern were discussed at this meeting. These three issues
- included the site'. failure to make the required emergency notifications to .
state and local government agencies in a timely manner, the inability of site I ! personnel to establish containment integrity within the required time limits, and the failure of the emergency diesel generator to provide AC power as intended (inadequate root cause analysis program). Your presentation was effective in clarifying the reasons for the apparent violations and in delineating your corrective actions. This meeting was also l beneficial in the fostering of open connunications. A list of attendees and a : copy of your handout are enclosed. We are continuing our review of these issues to determine the appropriate enforcement action. ; In accordance with Section 2.790 of the NRC's ' Rules of Practice," Part 2 j Title 10. Code of Federal Regulations, a copy of this letter and its enclosures l will be placed in the NRC Public Document Room. Sincerely, , l l Ik Luis A. Reyes, Di r ' Division of Reacto Projects
Enclosures:
(Seepage 2)
)
b f/ WN{D Rc2T--- fh
Georgia Power Company 2 OCT e j g
Enclosures:
- 1. List of Attendees
- 2. Conference Handout 4 cc w/encls:
R. P. Mcdonald Executive Vice President-Nuclear
' Operations Georgia Power Company P. O. Box 1295 Birmingham, AL 35201 C. K. McCoy Vice President-Nuclear Georgia Power Company P. O. 1295 Birmingham, AL 35201 G. Bockhold, Jr.
- General Manager, Nuclear Operations Georgia Power Conipany P. O. 1600
. Waynesboro, GA 30830 ! J. A. Bailey Manager-Licensing Georgia Power Company . P. O. Box 1295 l Birmingham, AL 35201
- D. Kirkland, III, Counsel Office of the Consumer's Utility Council Suite 225, 32 Peachtree Street NE 2 Atlanta, GA 30302
' Office of Planning and Sviget Room 6158 l 270 Washington Street, $W Atlanta, GA 30334 Office of the County Connissioner Burke C~Jnty Connission Waynesboro, GA 30830 Lonice Barrett, Connissioner ,
l Department of Natural Resources 205 Butler Street, SE, Suite 1252 l Atlanta, GA 30334 4 l (cc w/encls cont'd - see page 3) 4
-* Georgia Power Company 3 ' ' 007 o f luso cc w/ enc 15: (Continued) '
Thomas Hill, Manager Radioactive Materials Program Department of Natural Resources
- 878 Peachtree St., NE., Room 600 i Atlanta, GA 30309 Attorney General '
Law Department 132 Judicial Building
- Atlanta, GA 30334 1 State of Georgia b l l
i i a a I 4 4 i l l r ..y -, . .-. -
4 i ENCLOSURE I , LIST OF ATTENDEES U.S. Nuclear Regulatory Commission S. D. Ebneter, Regional Administrator, Region II (RII)
-J. L. Milhoan, Deputy Regional Administrator, RII L. A. Reyes, Director, Division of Reactor Projects (DRP), RII A. F. Gibson, Director, Division of Reactor Safety. RII J. P. Stohr Director, Division of Radiation Safety and Safeguards (DRSS), RII i G. R. Jenkins, Director, Enforcement and Investigation Coordination Staff (EICS)
A. R. Herdt, Chief, Reactor Projects Branch 3, DRP, RII K. E. Brockman, Chief Reactor Projects Section 38. DRP, RI! W.andH.Radiological Rankin, Chief. Emergency Protection (Preparedness Section, Emergency Prepa EPRP) Branch,DRSS,RII l
- E. D. Testa, Senior Radiation Specialist. EPRP, DRSS, RI!
' B. R. Bonser, Senior Resident Inspector, Vogtle, DRP, RII ' R. D. Starkey, Resident Inspector, Vogtle, DRP, RII i P. A. Balmain, Resident Inspector, Vogtle, DRP, RII W. H. Miller, Jr., Project Engineer, DRP, RI! R. W. Borchardt, Regional Coordinator, Office of the Executive Director of Operations, NRC T. A. Reed, Project Manager, McGuire, Office of Nuclear Reactor Regulation , (NRR) D. Hood, Project M. nager, Vogtle, NRR B. Urye, Jr., Senior Enforcement Coordinator EICS, R!I A. D. McQueen, Office of Enforcement Georgia Power Company l R. P. Mcdonald, Executive Vice President W. G. Hairston, III, Senior Vice President - Nuclear Operations C. K. McCoy, Vice President - Vogtle Project G. Bockhold, General Manager - Vogtle P. D. Rushton, Manager, Engineering and Licensing - J. A. Bailey, Manager, Licensing L. A. Ward, Manager, Maintenance and Support ' K. R. Holmes, Manager, Trainirfg and Emergency Preparedness
- . . - - . - . - . . - - . . . - - - - . - _ _ . . _ = . . . _ . . . - -
ENCLO50RE"2 AGENDA i ! OPENING . . . . . . . . . . . . . . . . . . R . P . MCD O NA LD t 4 INTRODUCTION . . . . . . . . . . . . . . . . C. K. MCC0Y ; J EMERGENCY NOTIFICATION . . . . . . . . . KEN HOLMES : l A. 1. STATEMENT OF WNAT OCCURRED I (TIME LINE), EQUIPMENT i 2. ROOT CAUSE i 3. CORRECTIVE ACTIDH
- 4. SIGNIFICANCE l
! B. EQUIPMENT HATCH CLOSURE . . . . . . . . PAUL RUSHTON l 1. DISCUSSION OF HATCH CLOSURE I 2. ANALYSIS OF LOSS OF RHR l 3. ACTIONS FOR UPCOMING OUTAGE l; AND FUTURE CONSIDERATION
)
I
- 4. SIGNIFICANCE OF MARCH 20 EVENT C. D/G FAILURE . . . . . . . . . . . . . . LEWIS WARD y.
J ArtJCPCC
- 1. STATEMENT 0F FACTS 2
- 2. ROOT CAUSE
- 3. CORRECTIVE ACTION i
- 4. SIGNIFICANCE l I
l l
l l HISTORY OF DG PNEUMATIC CONTROLS PROBLEMS l l 5 l .. IT HAS SEEN STATED THAT YOGTLE HAS LONG HISTORY OF i DG PR0sLEMS.
. NUREG-1410 LISTS 69 CALCON SENSOR FAILURES.
i
. PRESENTATION WILL COVER 3 AREAS:
. CALCON TEMPERATURE SWITCHES l
l . CALCON PRESSURE SWITCHES IN ADDITION, WE HAVE SEEN REQUESTED TO ADDRESS ( RECENT PROBLEMS WITH: i COOPER STARTING AIR ADMISSION VALVES f 4 I FOLLOWING A PRESENTATION OF THE HISTORY AND l TECHNICAL FACTS RELATED TO THESE 3 SEPARATE l PROBLEMS, OUR CONCLUSION IS THAT:
. NO CLEAR PRECURSORS TO THE 3/20/90 EVENT j ARE EVIDENT FROM A REVIEW OF THE PREVIOUS I SENSOR WORK HISTORY.
i l . GPC DID 00 na0 ROUGH TESTING AND IDENTIFIED j 3 GENERIC COMPONENT DEFICIENCIES WITH TDI SUPPLIED COMPONENTS. I GPC DID NOTIFY OTHER, OWNERS OF THESE PROBLEMS i . AND SOLUTIONS. I 1 l j i
- l. . .
CALCON TEMPERATURE SENSOR MODEL A3500-W3 Y TOP PLATE 1 ml I TOP PLATE GASKET l O 0-RING CALIBRATION R]NG DISK SHAFT e POPPET DISK i e POPPET - AIR INLET . r POPPET SPRING i BASE GA5KET
]
ys l\\ b SPRING RETAINER A ! SENSOR TENSION SPRINC BASE SENSOR DISK 5 EXPAND AS SPACER TUBE TEMPERATURE INCREASES. THEREBY PULLING THE POPPET DISK DOVH ONTO THE POPPET. AS THE POPPET 15 PUSHED ' DOWNVARD. THE AIR ESCAPES THROUGH
- THE YENT.
AS THE TEMPERATURE DECREASES. THE SENSOR DISKS CONTRACT. TE SENSOR DISKS i SENSOR TENSION SPRING PUSHES
, LFWARD AGAINST THE SPRING RETAINER.
WHICH PUSHES TE POPPET DISK UPWARD 0FF TE POPPET. THE AIR StPPLY AND POPPET SPRING PUSH i TE POPPET BACK INTO THE POPPET HOLE.
SUMARY OF DIESEL GENERATOR TEMPERATURE SENSOR PROBLEMS AT V0GTLE , UNIT 1 UNIT 2 UNIT 1 1R1 UNIT 1 CYCLE 2 POST EVENT CONSTRUCTION CONSTRUCTION (9/88-10/88) PRE- OUTAGE ISTRUNENT INCTIdM PROBLEM (8/85-12/86) (1/88-12/88) OUTAGE PRE-EVENT 3/20-3/25/90
. W. Temp. Out-of-cal. 8 7 4 4 2
. W. Temp. Defective 1 6 1
. O. Temp. Out-of-cal. 1 1 1 rg. Temp. Defective l l 10 1
A lb. Snsrs. Defective 3 . i
l f* - # i' . i 4 i JACKET WATER SWITCH PROGRAM RESULTS
~
. INSUFFICIENT TEMPERATURE STABILIZATION PERIOD :
l PRIOR TO CALIBRATION. ! CONTAMINANTS ON THE TEMPERATURE SENSOR (TIP ~
. l l
. CALIBRATION BATH HEATUP RATE.
. THERM 0WELL SETSCREW TIGNTNESS. !
. 1 i
. $ PACER-TUBE TIGHTNESS. !
I d
. INTERNAL CONTAMINANTS.
- 4 l
i J i l i 1 I . l 4 1 i l
l - I JACKET WATER TEMPERATURE SWITCH PROBLENS l I i Roor CAUSE [ . j -
. INTERNAL CONTAMINATION CAN CAUSE SWITCH TO j CONTINUOUSLY YENT.
- . CALIBRATION OF THE SWITCHES WAS INADEQUATE.
i 1 J CORRECTIVE ACTION , . CALIBRATION PROCEDURE TNAT CLEANS AND PROPERLY f CONTROLS THE CALIBRATION REQUIREMENTS. i
. RELIABILITY OF THE BASIC SWITCH COMPONENT WAS j ESTABLISHED.
! . SWITCNES HAVE BEEN DEFEATED IN THE EMERGENCY START MODE. ! . CURRENTLY EVALUATING REPLACEMENT. i SIGNIFICANCE OF PaostrM J ). . INADE0'JATE CALIBRATION PROCEDURES AND TECHNIQUES i RESULTED IN REDUCED RELIABILITY OF THE EMERGENCY ! DIESEL GENERATORS. EONCLUSION l l . BEGINNING WITH PROBLEMS ENCOUNTERED DURING THE l FIRST UNIT 1 REFUELING IN 1988, GPC HAS TAKEN ! A SERIES OF ACTIONS TO RESOLVE DG TEMPERATURE ! SWITCN MALFUNCTIONS. 1
I 1 [ .- - EMt8 r ) 1. & = 9&M ATR18Aala I fB FBIA MF628. l l . samatcnn mms esses . - I EtB SW CEWREG O*333*L. , j, elaCTTTE 4 MI/4379 i i 4 j b"" 4390 i SENS840 NtAS 4847 4395 ett ese g as . eence 430 3 A 84M1 . . , 4 gg3 ggayy - \ 4373 I A 4383 ; BAD 1 4370 ' ' 4384 l passaggE FEATE 1 A l 4300 i ' 4378 349 [ _ - an. EW
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= = = tant ==mmeg 3 3439 l Arts
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! ,84400 l PRESSURE SENSOR SUB ASSEMlllES j"i'i i _ _ .- . ___ _
=
SUMARY OF DIESEL GENERATOR PRES 5URE SENSON PROBLEMS AT V0GTLE 1 l UNIT 1 UNIT 3. UNIT 1 1R1 UNIT 1 1R2 POST EVENT N5TRUMENT CONSTRUCTION CONSTRUCTION (9/88-10/88) PRE- OUTAGE UNCTION PROBLEM (8/85-12/86) (1/88-12/88) OUTAGE PRE-EVENT 3/20-3/25/90
. O. Pres. Out-of-cal. 4 2
. O. Pres. Defective 1 --
1
. O. Pres. Out-of-cal. 1 1
. O. Pres. Defective 1
. P. (P-3) Out-of-cal. 1
. P. (P-3) Defective 1 . 1
. W. Pres. Out-of-cal. 1 1 2
DIESEL GENERATOR CALCON PRESSURE SWITCHES { 10CFR PART 21 (MAY 1988) i ! ~ j
" DEVICES THAT ARE ALREADY INSTALLED AND OPERATING !
i j AFTER SEVERAL HOURS BETWEEN TESTS MAVE DEMONSTRATED THEIR RELIABILITY. IMO DELAVAL RECOMMENDS THAT ALL DEVICES NOT INSTALLED, OR THAT ARE INSTALLED BUT HAVE i i NOT OPERATED FOR SEVERAL HOURS BETWEEN TESTS, BE l ! RETURNED TO IMO DELAVAL FOR REMACHINING, INSPECTION l AND TESTING." i 10CFR PART 21 (JUNr 8, 1990. Anormouw 3) l l i "OuR RECO#e4ENDATION Or NAY 12, 1988 MAY HAVE BEEN ! CONFUSING AND IN LIGHT OF THIS FAILURE AFTER 9 YEARS, i IT IS APPROPRIATE TO RESTATE OUR RECODG4ENDATION. COOPER INDUSTRIES RECOP94 ENDS THAT ALL PRESSURE l DEVICES, COOPER P/N F-573-156, BE MODIFIED OR REPLACED BY DEVICES IDENTIFIED AS CALCON P/N 844008." ' I! i i ! l l l v -
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4 i
- PRESSURE SWITCH FAILURE ANALYSIS RESULTS 1
1 l
. Roor CAUSE i
l . . MIS-INTERPRETATION OF PART 21 (NAY 1988). j CONNECTIVE ACTION l . ALL CALCON PRESSURE SWITCHES HAVE BEEN REPLACED. ? i l
- SrGNIFICANCE or Pn0BLEM i
. STICKING PRESSURE SWITCHES COULD HAVE RESULTED l IN REDUCED RELIABILITY OF THE EHERGENCY DIESEL l GENERATORS.
i l CONCLUSION i AS A RESULT OF THE GPC ROOT CAUSE EVALUATION OF i A PRESSURE SWITCH PROBLEM IN 1990, THE VENDOR ! CLARIFIED AND REISSUED THE 10CFR21 NOTICE. 1 l 1
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i i -InnicATION OF Arn START PRonLEMS i 1/24/90 - DG 2A ROLLED SUT FAILED TO S ROUTINE SURVEILLANCE TESTING. 1
. i
{ 1/25/90 - DG 2A START WAS ATTEMPTED. THE ENGINE SLOW ROLLED SUT DID NOT START. ! I 4/12/90 - DG 2A ROLLEo sUT FAILED TO STA ROUTINE SURVEILLANCE TESTING. ) t 7/5/90 > A SIMILAR EVENT OCCURRED ON D SURVEILLANCE TESTING. l i 7/11/90 - DG 2A AGAIN SLOW ROLLED AND FA START. i ! i ! I I \ i t s l
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- STARTING AIR ADMISSION VALVE l .
RESULTS 0F INVESTIGATION I SEVERAL INDIVIDUAL CYLINDER AIR START VALVES WERE DETERMINED TO BE STICKING IN THE OPEN POSITION. SEVERAL VALVE CAPS WERE MACHINED WITH THE BORE i SLIGHTLY OVAL-SKAPED AND TAPERED BY SEVERAL MILS. SOLUTION WAS TO POLISH EACH PISTON TO PROVIDE l APPROXIMATELY 2 - 3 MIL CLEARANCE WITH ITS MATCHED i CYLINDER CAP. COOPER INDUSTRIES ISSUED A PART 21 REPORT TO THE NRC ON JULY 19, 1990. l e
i j.. t i Roor CAUSE OF EVENTS
. MANUFACTURING DEFECT IN INDIVIDUAL AIR START VALVES I ,
WAS DISCOVERED AND RESOLVED BY GEORGIA POWER l CoHPANY. . l - i -l CORRECTrYE AcTr0N i . INCREASE CLEARANCE IN VALVES TO PREVENT BINDING. 1 i
! . NOTIFY VENDOR FOR PART 21 ISSUE.
! . LONG-TERM CORRECTION NOT YET DETERMINED BY VENDOR. j SraNrrrCANCE OF PnOBLEM
- . DG STARTING RELIABILITY WAS ADVERSELY AFFECTED BY THIS MANUFACTURING DEFECT.
t l l - l 4 t i i 4 1
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Roor CAUSE OF EVENTS ,
. MANUFACTURING DEFECT IN INDIVIDUAL AIR START VALVES WAS DISCOVERED AND RESOLVED BY GEORGIA P0wER COMPANY.
CORRECTIVE ACTION 1
. INCREASE CLEARANCE IN VALVES TO PREVENT BINDING.
. NOTIFY VENDOR FOR PART 21 ISSUE.
. LONG-TERM CORRECTION NOT YET DETERMINED BY VENDOR.
j SIGNIFICANCE oF PROBLEM
. DG STARTING RELIABILITY WAS ADVERSELY AFFECTED BY
! THIS MANUFACTURING DEFECT. I i j l 4
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[ y g UNITED $TATES U.S. !:f C NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20ssa n"\ Z hf'ff, _ OhRill
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DOC DT: 09/11/90 FINAL REPLY: Michael D. Kohn Cnunrel to Marvin B. Hobby and A11cn L. Mosbau9h TO: Chairman Carr FOR CIGNATURE OF: ** GRN ** CRC NO: 90-1000 DESC: ROUTING: 2.206 - REQUEST FOR PROCEEDINGS & IMPOSITION OF Taylor CIVIL PENALTIES FOR IMPROPERLY TRANSFERRING Sniezek CONTROL OF GEORGIA POWER CO.'S LICENSES TO THE ' on SONOPCO PROJ. & FOR THE UNSAFE & IMPROPER a OPERATION OF GEORGIA POWER CO. LICENSED FACILITIES M 1ey, NRR 1eberman, OE DATE: 09/17/90 ASSIGNED TO: CONTACT: - OGC Scinto SPECIAL INSTRUCTIONS OR REMARKS: A NRR TO COORDINATE WITH OE. [- 81 h I / 'l 7 s' /
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UNITED STATES [p Eth 'o NUCLEAR REGULATORY COMMISSION . p'" *
, REGION ll y y 101 MARIETTA STREET N.W. # # ATLANTA GEORGl A 30323 NOV 011990 % . . . . . /'
d LIMITED DilTRT5UTlDN -NOT-FOR4UBLIreDISC05ttRE U Request No. RII-90-12 T0: James Y. Vorse, Director Office of Investigations ; Region II Field Office l FROM: Stewart D. Ebneter : Regional Administrator REQUEST FOR INVESTIGATION Georgia Power Compar.y 50-424, 50-425 Licensee Docket Nos. Vogtle Electric Generating Plant RII-90-A-0092 Facility Allegation No. t A. Request What is the matter that is being requested for investigation? (Be as i specific as possible regarding the underlying incident.) 1 l By letter dated September 11, 1990, the law firm of Kohn, Kohn, and ] Colapinto, 517 Florida Avenue, N.W. , Washington, D.C., petitioned the Nuclear Regulatory Commission on behalf of Messrs. Marvin B. Hobby and Allen L. Mosbaugh (petitioners), former employees of Georgia Power Company.
...for proceedings and imposition of civil penalties." A petition entitled " REQUEST FOR PROCEEDINGS AND IMPOSITION OF CIVIL PENALTIES FOR IMPROPERLY TRANSFERRING CONTROL OF GEORGIA POWER COMPANY'S LICENSES TO THE SON 0PC0 PROJECT AND FOR THE UNSAFE AND IMPROPER OPERATION OF GEORGIA POWER COMPANY LICENSED FACILITIES" was attached to the above referenced letter. This document was received in the Office of the Secretary of the Commission on September 11, 1990, and instructions were provided to process the petition under 10 CFR 2.206.
Contained within the petition Jer Part 3, Section *, " SON 0PC0 inten-tionally mislead the NRC about the condition of the Vogtle Plant after a Site Area Emergency in order to hasten the restart of the reactor," the petitioners assert that the licensee knowingly provided " false statements intended to mislead the NRC with false assurances about the reliability of l the diesel generator whose failure resulted in the Site Area Emergency [on l March 20, 1990]." The petitioners further alleged that SON 0PCO. knew "the l I[Y F0 BLIC DISCLOSURE WITHOUT THE APPROVAL 9%oS29 w& ? 4PP j
1 i
, , l QM1TEfTDIiSTl18tfTTO T RE-410V 011990 l 2
diesel generator had actually continued to experience an excessive number j of trips, failures and problems similar in nature to the failure which l' lead to the March 20, 1990 station blackout." The petitioners provided additional information to support their allegation in the petition. ; B. Purpose of Investigation
- 1. What is the basis for the belief that the violation of a regulatory requirement is more likely to have been intentional or to have resulted from careless disregard or reckless indifference than from error or oversight? (Be as specific as possible.)
There have been several inspections related specifically to the diesel generator start issue performed by Region II inspectors. Those inspections, although not specifically focusing on possible intentional actions by the licensee to mislead the NRC with regard to diesel generator starts, did develop enough information to indicate that there may, in fact, have been a " counting problem" with respect to enumerating the number of starts and defining what actually constituted a valid start for counting purposes. The entire issue was clouded with differing views and opinions among the licensee's staff, enough so that an accurate assessment of the issue could not j be established. The staff is concerned that because of the lack of ! clarity regarding this matter that there may be sufficient confusion, I intentional or otherwise, involved in the issue to conceal possible efforts on the part of the licensee to obtain the most favorable position with respect to the NRC. The information provided by the i petitioners adds to the confusion surrounding the matter and cannot ! be discounted because one of the petitioners claims to have first-hand knowledge and actual involvement in the issue.
- 2. What are the potential regulatory requirements that may have been l violated?
10 CFR 50.9 1
- 3. If no violation is suspected, what is the specific regulatory I concern?
N/A I
- 4. Why is an investigatisn needed for regulatory action and what is the regulatory impact of this matter, if true?
Due to the claims and counter-claims of the various parties involved ! in this matter, a formal investigation is the most expeditious means of resolution. If the allegations that the licensee failed to provide accurate and complete information regarding the diesel generator to the NRC are substantiated, legitimate concerns as to the j I KTFORPUSL4C-015CLOSUREWITHOUTTHEAPPROVAL s j OF THETIRECTOR rQI y ' .
i m h0R6i(DIspw$QE NOV 011990 f 3 licensee's ability and determination to operate the facility in a manner consistent with ensuring the public health and safety would be ' paramount and require significant evaluation and action by the staff. Any substantive action would require a formal investigative basis. ;
- 5. In addition to the accuracy of information issues involved with the diesel generator, there are other similar issues relating to the question of the licensee providing accurate information to the NRC .
during a special team inspection conducted at the facility. I During the period August 6-17, 1990, a special team inspection was conducted at the Vogtle Electric Generating Plant (VEGP) to determine ; if the licensee operates the facility in accordance with approved procedures and within the requirements and intent of the facility's operating license. The inspection was predicated upon recent activities at VEGP which raised concerns within the NRC as to the , licensee's ability and determination to operate the facility in a . safe and conservative manner. An aggregation of facts and circum-stances associated with operational events and unsubstantiated allegations were viewed as a possible indicator of a non-conservative attitude on the part of the licensee's operating staff which warranted immediate initiation of the special team inspection. The results of this special inspection were detailed in Draft No. 7 of NRC Inspection Report Nos. 50-424/90-19 and 50-425/90-19. The final report, which has not yet been issued, will not include any of the details for items referred to 01 or for items included in the 10 CFR 2.206 petition. In addition to a review of specific technical issues made during the special inspection, the inspection team concluded that during the inspection inaccurate information was received on several occasions from responsible managers and operators on topics well within the scope of their specific responsibility. In five instances the initial information supplied by cognizant licensee representatives was clearly incorrect or inadequately researched. The inspection team concluded that in each of these examples, that licensee officials provided inaccurate, unsworn, oral statements concerning information which concerned topics well within their specific responsibilities and expertise. In the first two cases listed below, the inaccurate information was considered significant to the outcome i , of the inspection prc;ess. Specifically- l l
- a. a Unit 1 surveillance procedure, the unit shift super- !
During(USS) visor stated, and the operations manager later confirmed, , that the containment isolation valves for the hydrogen monitor i system were allowed to be opened without entering the LCO action l l NOT--F0E UBl.IC DISCLOSURE WITHOUT THE APPROVAL y '0FTHE DlRECTOR, Ol' V
D(MITED DI5TRI @ SURE NOV 011990 requirements for Technical Specification (TS.) 3.6.3 because the valves received an automatic isolation signal. The inspection identified that these containment isolation valves were remotely- i operated, manual valves without automatic isolation signals. j
- b. The operations manager stated that, after Unit I refueling i outage IR2, the modifications to the snubbers were done in i conjunction with preplanned system outages which were required t for other preventive or corrective maintenance or testing. The inspection identified that few of the snubber modifications were done jointly with pre-planned system outages.
- c. The general manager stated that VEGP complied with the corporate position regard ESF actuation reportability. The inspection identified that VEGP did not follow this policy and, in fact, complied with the requirements of 10 CFR 50.72 in that all ESF ,
actuations were reported.
- d. The operations manager stated that the shift superintendents (SS) reported directly to the operations manager and that he personally prepared their performance appraisals. The inspection identified that the SSs reported to the unit superintendent (US), and that the US personally prepared the performance appraisals of the SSs.
- e. The US indicated that there were no Operations Department actions which were anticipated or required within the first three hours of entering the action statement of TS 3.0.3. The inspection identified that the VEGP management policy and stated practice required preparations for a power reduction, including informing the load dispatcher within the first hour.
C. Requester's Priority
- 1. Is the priority of the investigation high, normal, or low?
High
- 2. What example fram Appendix 0517, Part III, does this incident most closely fit, if any?
4.a.(3) ! 1
/
NO
'LEOR_PUBLIC RE WITHOUT THE APPRQVAC L OF DIS l
THE DI T P
i i o s l AfMITEtFDISTb[BUT}' 11Rf f r n Q C p NOV 011990 ; 5
- 3. What is the estimated date when the results of the investigation are !
needed? ) December 20, 1990
- 4. What is the basis for the date and the impact of not meeting this date? (For example, is there an immediate safety issue that must be addressed or are the results necessary to resolve any ongoing reg-ulatory istue and if so, what actions are dependent on the out- come of the investigation?)
The date when the results of the investigation are needed is based on the urgent need to have multiple issues involving VEGP resolved on a timely basis. There cre currently two other investigations pending that involve activity related to VEGP, as well as the special team inspection. The issue involving the diesel generator is included in the 10 CFR 2.206 petition which is currently pending before the NRC. The outcome of this investigation will directly impact the outcome of the special inspection and the staff's response to the 2.206 petition. D. Actions by Staff i i
- 1. What actions have been taken by the staff (e.g., inspections Notice '
of Violation, Enforcement Conferences, Confirmatory Actions Letters, etc.)? The special team inspection has been co;apleted and its report is currently pending. The staff has begun preparation of the response to the 2.206 petition. The results of this investigation will he incorporated into the response.
- 2. Actions to be taken if investigation is closed without a report (based on currently available information).
Additional inspection will have to be performed to resolve the issues through the inspection process. E. Contact
- 1. Staff members: K. Brockman, Region II, Ext. 16299 A. Herdt, Region II, Ext. 15583 B. Uryc, Region II, Ext. 14192
- 2. Alleger identification with address and telephone number if not confidential . (Indicate if any confidential sources are involved and who may be contacted for the identifying details.)
See petition for identifying data pertaining to the petitioners. J0LFORJilBLIC DISCLOSURE WITJH UTQI NOF-THE IIIRECTOR, O!
s tiMR N TRIB FjR40BMR OV 0 1 19g0 F. Other Relevant Information Additional details regarding the special team inspection are provided in the draft special inspection report. , f.;; Stewart D. Ebneter .
Enclosure:
Ltr M. Kohn to.K. Carr, 9/11/90, i w/ encl as indicated cc: J. Taylor, EDO J. Sniezek, DEDR H. Thompson, DEDS B. Hayes, 01 , J. Lieberman, OE
! J. Partlow, NRR J. Goldberg, 0GC l
4 I
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LIMFTEO18tfT UBLIC DJ'"LOSURE 01 RESPONSE T0/ STATUS OF RE0 VEST FOR INVESTIGATION November 6, 1990 T0: Stewart D. Ebneter - Regional Administrator FROM: James Y. Vorse Director Office of Investigations Field Office, Region II , REQUEST NO.: RII-90-12 DATE OF REQUEST: 11/01/90 Licensee / Vendor / Applicant: Georgia Power Company Facility or Site Location: Vogtle Electric Genating Plant 50-424, 50-425 License No.:
. Docket No.: N/A CASE INITIATED: [X] CASE N0.: 2-90-020 Date Opened: 11/06/90 Type of Case: (I)
Assist (A) Inquiry (Q) l ECD: 08/91 Investigation ( ) Case Priority: H H - High N - Normal L - Low CASE NOT INITIATED: [ ] REASON: i COMMENTS: i Distribution: ' (1 / cc: D. Murphy, 01:HQ J. Weddle 01:HQ s/f 2-90-020 A J. Liebeman, OE c/ J' , 01:RII ) 93
-LRobinson OI:RII)
JVorse 90/11/ & 90/11/ $ <\s 4 m
l INVESTIGATIVE PLAN Date November 6, 1990 i Case Agent L. Robinson Case No. 2-90-020 Controlling Office / Requester RA R2 view of request & applicable regulaticns by F0D //-O/- 10 ! Coordination with staff Co:rdination with Regional Counsel /0GC /d RIquest for additional information from staff l l Submission of OI response / status of request for investigation ;
. (Case.0pening Paper) 11/06/90 :
Interview of alleger: Confiden 1t anted YES[) N0( _) CS No. d7-90-/5 Interview of appropriate sta'ff memb[r(s) I Submission of initial monthly investigation status report (ISR) with ECD 11/06/90 Initial discussion with RA regarding ECD and priority of investigation
/ Z -Qs. 90 f case file and discussion with :ase agent: j
/2//2 Rsviewo/f8 3 26 41
/ViA !/4/ 4 DM. A'I i 2lf/2 f/ff, A MMI
'2/hz / 9 / L 35" Al i
I Discussion of case progress with RA: P /yf/ /-g.f2. b./2.fa l t 'l ~ O C -9 O '4=t RI 4- A 4/ 3 16-42. 2-/0 f z i 1 - o ~7 -4 i f I"! 4 i la m-f/ 3.'//.#2. ' k-I\~4i L t1 4I /1 11: 11 t/.14.f2 3* L 4l 7-I 4 / z) .ar:(/ A b/F(2
/ J of monthly ISR:
Submission it @ [40
#1 -T -40 rd Tllen I .
,) h Date field work completed: Iyif j jr !
Date draft report reev'd by F0D: ; Date final report signed by F0D and/or forwarded to HQ for review: b 1 i _m_ _ _ _ _ . . _ . . _ _
OIMIS DATA ENTRY SHEET CASE NUMBER 2-90-020 01 DOCKET NUMBER 50-424, 50-425 03 FACILITY- Vogtle Elec. Gen. Plant 04 TYPE OF CASE 05 DATEOPENED(omit /) 11/06/90 06 ESTIMATED COMPLETION DSTE 08/91 07 REQUESTED BY RA 08 STATUS 10 AltrGATION (2 lines max) ALLtGED FALSE STATEMENTS REGARDING OPERABILITY OF DIESEL GENERATOR 123456789012345678901234567890123456789012345678901234567890 11 COMMENTS / REMARKS (10 lines max) 123456789012345678901234567890123456789012345678901234567890 12 CASE AGENT Robinson j 13 CATEGORY WR l 15 STATUTE DATE ! 16 CLOSED DATE , l 18 ISSUED DATE l 19 DOJ REFERRAL 20 00J ACTION 22 RELATED CASES 23 FOIA NUMBER j 26 PRIORITY H A ! 27 SOURCE OF ALLEGATION 28' VENDOR CODE h n af, s}}