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Docket No. 50-346 MEMORANDUM FOR: John F. Stolz, Chief, Operating Reactors Branch No. 4, Division of Licensing FROM:
Olan D. Parr, Chief, Auxiliary Systems Branch, Division of Systems Integration
SUBJECT:
DAVIS-BESSE STARTUP FEE 0 WATER /AFW PUMP - TAC NO. 55sB1 We have reviewed the licensee's 1,etter dated November 21, 1985, in which the conceptual design for adding a motor driven startup feedwater pump (SUFP) which will serve the dual purpose of start-up and as a full sized AFW pump, is provided. This.new pump will prnvide a diverse powered backup to the existing turbine driven AFW pumps. We are delighted that the licensee is adding this third, motor driven pump. We are concerned with one aspect of the design - the manual loading of the pump onto the diesel and manual alignment of the appropriate valves.
In order to use the SUFP in the "AFW" mode, five manual valves must be repositioned locally.
Since LER 84-3 identified a boil dry time of five minutes for the steam generator, we are concerned that unless these five valves are made remote manual valves which are operated from the control room they cannot be aligned before dryout of the steam generators. We also believe that the two manual valves, one to each steam generator, should be tied into the FOGG system so as to prevent possible pump damage due to pump run out. The suction of the pump should have the same automatic switchove" to the station service water (SSW) as the turbine driven AFW pumps to prevent cavitation of the SUFP.
As an alternative to making the valves remote manual, the SUFP should have all of the manual valves lined up for the "AFW" mode of SUFP operation when the plant is in operating modes 1, 2 or 3.
In this configuration, the SUFP will take suction frem the condensate storage tank and discharge to the AFW headers. The only manual operation necessary would be to load the SUFP onto a diesel, manually start the pump, and operate the motor control valve from the control rcom. We also suggest that the suction of the SUFP have automatic switchover to the SSW system, as previously discus:ed, for this configuration.
Contact:
J. Ridgely X29566 XA w.
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Since the'SUFP will have an "AFW" mode of operation, we request that the 4 @t,(
el licensee propose technical specifications for the operability of the SUFP system. We suggest an LCO and surveillance requirements on the A
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pump and valves to provide assurance of the system's availability.
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We are prepared to meet with the licensee to discuss these suggestions i
l if such a freeting would be appropriate.
This completes our work on TAC i
No. 55581.
i ba
. Olan D. Parr, Chief Auxiliary Systems Branch Division of Systems Integration ec:
R. Bernero L. Rubenstein J. Wilson J. Ridgely i
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o CONFfRMATORY ACTION LETTER
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June 10, 1985
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Docket No. 50-346 Toledo Edison company ATTN: Mr. Richard P. Crouse Vice President Nuclear Edison Pla:a 303 Madison Avenue Toleco, OH 43652 Gentlemen:
This letter confirms the conversations on June 10, 1985, between you and Charles E. Norelius of my staff related to the reactor trip at the Davis-Besse Nuclear Plant on June 9,1985.
While operating initially at 90% power, a reactor trip and other subsequent problems with the main and auxiliary feedwater systems resulted in a situation where no feedwater-flow was ava11able for 10-12 minutes causing low level in the steam generators.
With regarc to the matters discussed, we understand you have taken or will take tne following actions:
1.
Hold in abeyance any work in pregress (electrical or mechanical) and/or any work planned,on equipment that malfunctioned during the incident, such as disnantling or disturbing existing evidence, until the NRC Investigative Team has had an opportunity to evaluate this event and concur on your preposed corrective action (s).
8.
Review main steam isolation valve actuation, a.
Establish the cause(s) of the unexplained closure of ooth MSIVs.
b.
Cetermine and implement the corrective action (s) recessary to prevent recurrence.
c.
Perform additional testing to ensure the MSIVs operate as required.
3.
Review the auxiliary feed system actuation, a.
Establish the cause(s) of the inadvertent trip of both auxiliary feedwater pumps (AFP).
b.
Determine and implement the corrective action (s) necessary to prevent recurrence.
CONF 1RMATORY ACTION LETTER QQl4flJ SA
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CONFIRMATORY ACTION LETTER r
Toledo Edison Company 2
June 10, 1985 4.
Review the main feed system operation, a.
Establish the cause(s) of the inadvertent trip of the No.1 main feed pump (MFP).
b.
Detemine and implement corrective action (s) to prevent recurrence for both main feed pumps.
5.
Complete and submit to NRC Region III the results of your investigation of this event including your safety evaluation, an evaluation of thermal shcck consicerations for both Steam Generators (SG), the maximum S/G snell differential temperature, and possible general mechanical damage to any affectec system (s).
Also report your basis for event classification and the adequacy of information as originally reported to the NRC on June 9, 1985.
6.
Comolete the actions required by items 1-5 and obtain verbal concurrence frem the NRC Region III Regional Administrator or his designee prior to authorizing restart (Mode 2) of the reactor.
7.
Main ano Auxiiiary Feed System Testing.
Perform testing which can be performed below 5% pcwer, of the PFP's a.
to demonstrate all pumos operate as required.
b.
Perform testing of the AFP's to assure all pumps operate as requirec.
Appropriate test results, your conclusions, and a summary of c.
corrective actions taken will be orovided to the NRC resicent office upon ccepletion.
8.
Af ter all actions requirec above are completed obtain verbal concurrence from the NRC Region III Regional Administrator or his designee prior to exceeding 5% reactor power.
CONFIRMATORY ACTION LETTER
4 s'
CONFIRMATORY' ACTION LETTER RIII-CAL-85-06 Toledo Edison Company 3
June 10, 1985 Please let us know immediately if your understanding differs from that set forth above.
Sincerely.
Cric!9PI nIcned by J *..: G. L',. : r James G. Xeppler Regional Administrator cc w/ enclosure:
S. Quennot, Station Superintendent:
DM8/Cocu:nent Control Desk (RIOS)
Resicent Inspector, RIII Harold W. Kohn, Ohio EPA James W. Harris, State of Ohio R:bart H. Quillin, Ohio Department of Health i
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'In the matter of:
MEETING BETWEEN THE NRC STAFF AND TOLEDO-EDISON COMPANY CONCERNING AFW SYSTEMS A
Docket No.
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Lccation: Bethesda, Maryland 1 - 50 Date: Monday, June 17, 1985 Pages:
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.Gi 3IL*;'t & ASSOCIATES Court Reporters 108 ' e r..
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Suite 921 Washington, D.C.
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1 UNITED STATES OF AMERICA 2
NUCLEAR REGULATORY COMMISSION 3
4 MEETING BETWEEN THE NRC STAFF AND TOLEDO-5 EDISON COMPANY CONCERNING AFW SYSTEMS 6
7 Room P-110 8
9 Phillips Building 10 7920 Norfolk Avenue 11 Bethesda, Maryland 12 Monday, June 17, 198b 13 14 The meeting in the above-entitled matter convened, 15 pursuant to notice, at 10:30 o' clock, a.m.,
Mr. Harold Denton, 16 presiding.
17 NRC ATTENDEES:
18 H.
Denton NRR 19 B.
Sheron DSI 20 C.
Parr DSl 21 W.
Houston DSI 22 J.
Stolz NRR 23 F.
Rowsome DST 24 A.
DeAgazio NRN 25 D.
Eisenhut NRR
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2 1
NRC ATTENDEES (continued) 2 A.
Thadani Safety Technology 3
D.
Wessman Division of Licensing 4
A.
El-Bassioni PRAB 5
8.
Newlin Public Affairs 6
G.
Lainas NRR 7
F.
Miragtla NRR 8
9 LICENSEE ATTENDEES:
10 J.
Silberg, Esquire M
11 B.
Fink
. L
-\\
j 12 R.
Gradomski 13 B.
Peters 14 T.
Myers 15 S.
Jain 16 M.
O'Reilly 17
- O ALSC PRESENT.
19 L.
Connor Doc-Search Associates 20 R.
Borsum Babcock & W l I co:<
21 J.
Nurmi Engineering Planning & Mgmt 22 M
'?yan NucIeenies Woek 23 24 25 t
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'6 3
e p RO C EED I N GS 1
(10:30 a.m.)
2 3
MR. DENTON:
Good morning.
My name is Harold j
4 Denton.
I want to thank you for coming in on such short 5
notice.
6 The purpose of this meeting is to deve!cp a 7
chronology of actions that have been taken on the feedwater
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O system at Davis-Besse and the studies which have been done 9
leading to those actions.
10 I reali=a 1 didn't give you much lead time to 11 prepare for this meeting, and all this information was not yet fr 1
12 available.
But I have had the Staff go through our files on 13 what has transpired between us with the steamline accident, 14 and I wanted to be sure we had a complete chronology developed 15 on that point.
16 I think there have been some questions asked since 17 the incident at Davis-Besse as to what the status of 18 operaticna
.4 a r.d it.h e n 5t wcia l d be comp'oted tnd this sort of 19 thing, and I would like to get a clear understanding of what 20 you have done, what studies you've done, and we'll chip in 21 with our knowledge of the process where we can.
I l
OC Before we begin. I think it might be useful to go i
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scound the room and each person identify themselves and their 24 responsibilities for the record.
25 Why don't we start, Al, with you.
.]
'. D 4
I am Al DeAgn=lo, Davis-Besse 2
Project Manager in the Division of Licensing.
t, 3
MR. STOLZ:
I'm John Stolz, Branch Chief of the 3
4 branch to which Davis-Besse is assigned.
5 MR. ROWSOME:
I'm Frank Rowsome, Division of Safety
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6 Technology.
7 MR. EISENHUT:
Darrell Elsenhut, Deputy Director of 8
NRR.
t 9
MR. DENTON:
Harold Denton, Director of NRR.
(
I 10 MR. SHERON:
I'm Brian Sheron, Chief of Reactor 11 Systems Bra.ich, OSI.
i MR. PARR:
Olan Parr, Chief, Auxiliary Systems 12 7
13 Branch.
14 MR. HOUSTON:
Wayne Houston, Assistant Director for OSI 15 Reactor Safety and today Acting Division Director,
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16 MR. THADANI Ashok Thadani, Chief, Reliability and 17 Assessment Branch, 19 MR. WESSMAN:
I'm Dick Wessman, representing the 19 Branch Chief of the Division of Licensing.
20 MR. FINK:
Bob Fink, MPR Associates, consultant to 21 Toledo Edison.
22 MR. GRADOMSKI Rich Gradomski. Engineer.
23 MR. PETERS:
Bob Peters, Licensing Manager.
24 MR SILBERG:
Jay Silberg, Shaw, Pittman, Potts &
25 Trowbridge, Attorneys.
]
/*
5 1
MR. JAIN:
Sushil Jain, I'm with Toledo Edison.
2 MR. MYERS:
Director, Core Safety and Licensing, 3
Toledo Edison..
4 MS. BRYAN:
Margaret Bryan with " Nucleonics Week."
5 MS. NURMI:
Joy Nurmi with Engineering Alanning &
6 Management.
7 MR. NEWLIN:
Bob Newlin, public Affairs, NRC.
8 MS. O'REILLY:
Mary O'Reilly, Toledo Edison.
9 MR. DENTON:
We have two more seats for anyone wno 10 would like to move forward.
11 MR. BORSUM:
Babcock & Wilcox.
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12 MS. CONNOR:
Lynn Connor, Occ-Search Associates.
13 MR. MIRAGLIA:
Frank MiragIla, Deputy Director of 14 Licensing, NRR.
15 MR. LAINAS:
Gus Lainas, Assistant Director, 16 Operating Reactors, Division of Licensing.
17 MR. EL-BASSICNI Adel El-Bassioni l'm with ARAB.
1G MR. GENTON.
For the benetit o t' the people dho just 19 arrived, what we're doing today is developing a chronology of 20 actions that have been taken at Davis-Besse regarding t he-21 feedwater system and the auxiliary feedwater system since the 22 TMI accident We do not intend to get into any discussions of 2?
what happonad recently st Onvis Ga4*o.
but t%is !t to qo 5ack 24 to some of the earliest actions that were taken in the summer 25 of '79 and to bring those forward.
And I would like to go J
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1 forward right up until your plans for late this year regarding 2
completing the operating of the auxiliary feedwater. system, 3
including the larger pump associated with the electric drive 4
system for the auxiliary feedwater.
5 So at this point, Al, unless there are some other 6
preliminaries to get into, I propose that we let Teledo Edison 7
have the floor and give us your understanding of the actions O
that have transpired.
9 MR. DeAGAZIO:
I think we can probably do that.
10 MR. MYERS:
I am here on behalf of Dick Crouse, who 11 apologizes for not being able to come down here.
We have i
12 brought several of our key individuals from the engineering j
13 and licensing area, who have been involved with the auxiliary 14 feedwater system over the years, it was our intent to 15 exchange information today at your request from Saturday, a
16 much as we could, and if anything was left unvalidated or 17 preliminary, like much of our information is that we developed
- ?
yesterday and this -e-ning, we would v e - ' '/ *hst in time to 19 support any further use of that information that we've been 20 involved with that perhaps people have not been aware of over 21 the years.
22 MR. DENTON:
I think what might be useful would be 23 for us to give you a period of time to review the transcript 24 and then provide any additional clarifying material that you 25 think might be appropriate.
s 7
1 MR. MYERS:
We appreciate that.
2 The intention we have this morning was to provide a 3
short description of our auxiliary feedwater system, 4
recognizing several people may be new to it or have not been 5
involved recently in it to bring the system back into focus.
6 I think most of the individuals or many of you here have been 7
involved over the years since TMI with Davis-Besse, but I will 8
give you a short review of its operation and its basic design 9
and then go into both the procedural and the modification 10 activity that we've been involved with since TMI.
11 Again, much of that has been exchanged'with you, h.
j 12 Some of that has not specifically been Identified to you.
The 13 Intent today is to come up with a complete list.
Hopefully, 14 that would give you both of those packages.
~
15 As I said, we have some of our people here, but some 16 of our key people are also back supporting the fact-finding 17 team.
So if we need to validate any information or if there 19 are burning questions, we :aa probabty :t t i l l de
- hat ti: d a y by 19 phone.
But to at least give you an overview of what those 20 changes have been, we should be able to supply that here.
21 We are working again back at Toledo to actually 22 Identify actual installation and modification dates for you to 23 give you an accurate chronology of exactly when these 24 modifications were in place.
We can give you estimates today, 25 to the best of our knowledge and what we were able to come up
G 1
with yesterday.
But we will be validat6ng, and you should 2
recognize this is preliminary information, so we can append 3
the transcript, possibly, with any changes that we would find 4
in the interim to validate that information for you.
5 I would like to introduce now Mr. Jain, who will be 6
describing in general the system function.
I had expected a 7
smaller meeting, so unfortunately our overhead doesn't seem to S
focus large enough.
k 9
We have a system layout drawing that we will pass 10 around that will help in the discussion.
M 11 MR. DENTON:
Do you have sufficient copies?
If not, d
1 12 we would make some.
j 13 MR. MYERS:
I think we do, hopefully.
14 This is a simplified drawing and is not the actual, 15 exact detail that's in the field.
Any differences will be 16 pointed out as we go through, but I think it will offer a 17 basis for discussion.
P?
MR.
JAIN:
C4.! 3-80s ta r i.J P. t now has twc 19 independent, redundant trains of auxiliary feedwater systems.
20 The mode of power provided right now is a steam turbine, and 21 each train, which drives one full hundred percent capacity aux 22 fe=dwater pump, normally is takinq suction from the e.nadencate 23 storage tank.
The Seigmic 1 supplies and the safaty arAde 24 supplies the service water system, for which we have an 25 automatic switchover from the condensate to the service water
i 1
9 1
in the event of a LOCA condition from the condensate storage 2
tank.
3 On the discharge side, we have each aux feed pump 4
discharging to its respective steam generator with provisions 5
for crosstles, so that each aux feed pump can feed either 6
steam generator at a given time.
We have similar crossties on 7
the steam inlet side on the aux feed pump turbine.
Normally O
the No.
I steam generator would be providing steam to the 9
No.
1 aux feed pump turbine, and the No. 2 steam generator 10 would be doing that for the No. 2 aux feed pump turbine.
We 11 have similar crossovers where any steam generator can provide 12 steam to either aux feed pump turbine.
13 The initiating system for the aux feedwater system 14 is what we call a steam and feedwater rupture control system.
15 The SFRCS, as we call it, is actuated on four ditierent 16 parameters ranging from steam at low-level for 17 loss-of-feedwater condition, a loss at the coolant pumps to 13 premoto natwral circulation, arid t h.7 r: a steam:..na break, which 19 is a low steam pressure condition.
20 The alignment of valves on the discharge of the 21 pumps, as well as the steam inlet to the turbines, is 22 dependent upon which steam generator is good, meaninq which 23 has good pre 4sure.
If a 4 team qmnerator has a break, that is 04 considered to be a bad generator, and auxiliary feedwater is l
25 isolated from that generator, and so is the steam coming from i
s 10 1
that steam generator.
2 The system that we have right now is essentially all S
safety-grade components, both on the steam inlet as well as on 4
the discharge side, with all the motor-operated valves 5
supplied from either the diesel, in the case of loss of 6
offsite power, or from the DC powee 4upply from one train of 7
the aux feedwater system.
8 That is the basic overview of the system, very 9
briefly.
10 MR. DENTONr What is the function of the system 11 labeled " steam generator wet layup generation and m
)
12 recirculation pumps?"
13 MR. JAIN:
This drawing includes two systems.
The 14 basic intent here is to look at the aux feed pumps.
The 15 recirc pumps are used when you have flooded the steam 16 generator during a shutdown condition, and you have the steam 17 generators in a wet layup condition, and then you want to
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insert te naintain e.h em i s t r y,
La d y e J % gvtr 4t9am Tt tritnes 19 flooded up to the top nozzle.
20 MR. DENTON:
So that's merely a recirculation 21 system?
22 MR.
JAIN:
Correct That is merely relavant to what 23 we're talking about today.
24 MR. DENTON:
Let's stop and see if the Staff has any 25 questions about this system.
e.
11 1
MR. STOLZ:
Were you also going to discuss the 2
startup pump and its relationship to supporting the aux feed 3
system that you show here?
4 MR. MYERS:
Verbally describe that.
5 MR. EISENHUT Before you do, though, if we could 6
maybe on the aux feed system, wtth what you were just 7
describing, the normal mode, then, on any event such as a low 8
steam generator level would be that both aux feed pumps get a 9
signal to turn on?
10 MR. JAIN Correct.
11 MR. EISENHUT:
And they actually get a signal to 12 turn on, and are there any valves that have to open?
Where 13 would the normal water supply be coming from in such an 14 event?
Could you just describe the scenarlo of what would 15 have to happen to turn on the auxiliary feedwater?
16 MR. JAINs For a steam generator low-level 17 condition, there is one valve that has to open to supply ateam 18 to the aux feed pump turbine, and those are shown on the left 10 top of this drawing.
But for just a low-level condition, the 20 respective valve from the No.
I generator would open the 21 No.
1 aux feed pump turbine.
With respect to the valve for 00 the No. J generator, it would open the f4o. 2 aux feed pump P1 barh8ng 24 MR. EISENHUT!
Could you tell me which those are on 25 this diagram?
12 1
MR. JAIN:
100 and 107 is for the No. 2 aum feed 2
pump turbine.
3 On the discharge side, I will tell you what the 4
configuration is today, because we have changed it over the 5
last few years.
In the configuration today, there will be one 6
diacharge valve that would have to open on a low-level signal, 7
and that would be 3070 and 3072 for the other generator.
8 MR. EISENHUT:
So given an event of a low-level or 9
dry generator, for whatever reason, upon the loss of or 10 low-level upon the loss of the main feed system, those four 11 valves would open, and that would normally turn on the aux 12 feed system, and it would deliver flow to the steam generator?
13 MR. JAINt Correct.
The normal water supply in this 14 case would be the condensate storage tank.
15 16 17 18 19 20 21 22 23 24 25
s 13 0
1 MR. ROWSOME:
Would not HF-360 also have to open?
2 MR. JAIN:
This is an older drawing AF,300 and 3
368, which were also on the discharge of the pump, they used 4
to be speed control valves, it used to open on an epm of 5
2800.
We have since deleted that Interlock in order to 6
Improve the reliability.
It's one of the changes 1*ll be 7
talking about later as to what has transpired.
8 MR. ROWSOME7 okay.
They're normally open, then?
9 MR. JAIN:
- Yes, 10 MR. DENTON:
Could you start your description of the 11 changes, beginning about the time of the TM1 accident?
- 12 MR. MYERS:
Excuse me.
We had one other question 18 that John wanted, having to do with the current use of the 14 startup feed pump and how that would be utilized.
We can do 15 that before we get into othat.
16 MR. JAIN:
The startup feed pump, as it is 17 configured right now, delivers water to the main feedwater 18 nozzles into the steam generators, it takes suction or it 19 could also take suction f*om the condensate storage tank.
20 That's the normal alignment for the startup feed pump.
We 21 could go into detail as to what actions are required for it to 22 be put in service, given a complete loss of feedwater -- main 23 and auxillary feedwater.
24 MR. DENTONi What is its normal function?
1 25 MR. JAIN:
The startup feed pump is basically there
14 1
to support low-power operation and for feedwater cleanup for 2
the steam generator chemistry again.
3 MR. DENTON:
And the capacity of that system again?
4 MR. JAIN:
It's 300 gpm.
It's not enough to remove 5
decay heat by itself.
6 MR LAINAS:
Was that about half the capacity' 7
MR. JAIN:
Yes; e
MR. MYERS:
Any other questions as far as the basic k
9 operation currently?
10 MR. THACANI:
That startup pump, what is the mode of M
11 power for the pump and the associated valves?
k l
12 MR. JAIN:
It's supplied from a nonessential 4160 13 bus, but we have provisions for it to be backfit from an 14 essential 4160 bus.
15 MR. MYERS:
Those are also discussed.
That activity 16 is also discussed in the changes we've provided over the 17 years.
18 Anything elso' Mr. Jain will also continue, then, 19 Okay, we have 20 if you want to talk about the equipment and associated 21 procedural changes.
Again, scme of these the exact times
)
22 when they're implemented, we're in the process of validating.
23 We've kind of placed them in a general flow, early to late, 24 but we would jockey their positions, based on the actual times 25 we would hopefully get back to you today or tomorrow or in the
_j
4 15 1
transcript here.
2 MR, JAIN:
Originally the Davls-Besse aux.iliary 3
feedwater system motor-operated valves were all AC power 4
operated.
We made changes so that the No.
I train over the 5
auxiliary feedwater system has valves which are all DC power 6
operated, so one train of valves is entirely independent of AC 7
power.
This change, we think, was dor:0 sometime in 1980 in O
order to reduce the p'robability of human errors on manual 9
valve mispositioning in the flow path of both the aux 10 feedwater system as well as the startup feedwater system.
So 11 we have put padlocks on the local handwheels.
We have also 12 put padlocks on local handwheels, as well as the pushbutton 13 stations for vital motor-operated valves in the field.
So 14 nobody in the field could inadvertently misposition those 15 valves.
16 There is an administrative procedure in place which 17 is a control on the positioning of both the manual valves, as
'9 well as the motor-operated vafves 19 We have mado provisions for the startup feedwater 20 control valve and the feedwater block valve to be controlled 21 from the control room by the operator.
These are the two 22 valves that get isolated on a steam and feedwater rupture 23 centrol system, whicn is the initiating system for the 24 auxiliary feedwater system.
So for a loss-of-feedwater 25 condition, these valves will get isolated.
However, they will
16 1
have to be reopened to provide startup feedwater to the steam 2
generators for a complete loss-of-feedwater condition.
3 We have provided the capability, so that the SFRCS 4
signal to these salves can be blocked from the control room, 5
and these valves can be reopened in order to expeditiously 6
provide feedwater to the steam generator and the star
- tup feed 7
pump.
8 MR. EISENHUT:
Can you point out where they are on 9
this?
10 MR. JAIN:
I'm afraid this startup feed pump is not 11 on this drawing.
).
)y 12 MR. EISENHUT:
But the valves also, they're not 13 on here, obviously.
Okay, thank you.
14 MR. JAIN:
Okay.
After the TMI accident, we were to 15 demonstrate that the aux feed pump turbine could be driven on 16 a dry steam generator.
We actually conducted a test where we 17 ran the steam generators dry and started the aux feed pump l0 turbino w!!h a dry steam genv atcr Th e twa: u s.4 a u G o a a r* w '
19 The SFRCS --
20 MR. DENTCN:
Let's stop there.
Can you describe a 21 bit more about the test?
Is that because when it's dry, it's 22 not really dry?
23 MR.
JAIN:
It'4 drv in the sense that it has bottfed 24 up steam in it with all the isolation on the feedwater as well 25 as the steam side.
L
17 1
MR. DENTON:
When you ran the test, do you recall 2
what the pressure wasi 3
MR. JAIN:
I don't recall personally.
We would have 4
to look at the test results.
5 MR. EISENHUT But the reason you can the test was 6
the concern that the pumps mould not have any inlet flows 7
sufficient inlet flows so that they would overstrip and G
9 MR. JAIN:
The reason to do that test was to 10 demonstrate that you had enough power, if you will, in the 11 steam generator that could initially roll the turbine, and
)
12 once the turbine was rolled, you had enough water going into la the steam generator to start the cycle, the production of 14 steam, and then running the turbine again.
15 MR. EISENHUT:
But yet enough to actually start the 16 turbine flow?
17-MR. MYERS:
There was a question -- the question
'O we're tryInG to answer was,
.n att tha '
.:s and s v e,"/ thing, 19 ence the system was bottled up, would we lose the motor force 20 prior to getting water backflashing to raise pressure to 21 recover the motor force.
22 MR.
THADHN1*
Can you tell me two things 9 Nurber 23 one, how long did you run the test for9 24 Number two, is there any implication 6n terms of 25 controls?
Steam inlet valves, for example, and so on.
The
'~
l 18 the speed of the turbine, what happens?
1 speed of the pump 2
MR. JAIN:
I'm afraid I cannot answer that right 3
now.
I don't have firsthand knowledge of that 4
MM. MYERS:
The test engineer is not here.
We can 5
get that information in the detailed test.
As a matter of 6
fact, I believe that was even submitted.
We can dig those 7
details up.
e MM. DENTONr' Do you recall if this is one that we 9
asked you to perform, or you performed It on your own 10 initiative?
11 MM. MYER$t I'm pretty sure the details were
.I 12 provided.
I don't know whether it was one that was asked for 13 or was part of our program anyway.
I'd have to -- we changed 14 Project Managers and a lot of people since then.
I'm not sure 15 that we can dig that up,'too.
16 MR. DENTONt I think we've had a turnover of 17 personnel also in the intervening five or aim years, and 6t's 18 difficult for us to reconstruct all the details 19 Let me ask just about dryout.
When you use the term 20 "dryout," what do you really mean with regard to the water 21 level in the steam generator?
What does 8&W normally mean 22 when they talk about dryout conditions?
la that no liquid 23 remaining or 24 MM. JAIN:
Well, the procedure that the operators again, you have to realize what's 25 are told to use is i
19 1
happening to the level instrumentation, because the level 2
instrumentation doesn't read the actual level, it gives you 3
the indicated level, based on the conditions it was calibrated 4
to and so on.
5 The procedure they have been given is eight inchen I
I 6
in the steam generator indicated, eight Inches ledicated it 7
could be anything actual, depending on what the steam 9
generator pressure is, 9
MR. DENTON:
I was asking in connection with that 10 test you performed, whether there was water still in thn 11 bottom of the steam generator, or whether it was --
12 MR. JAIN:
1 think the assumption was that there was 13 no water, but we will have to dig further into that.
14 MR. EISENHUTi A correlated followup question, if 15 you could look at it at the same time.
The B&W eight inches 16 left in the bottom of the vessel, la that the definition of a 17
" dry generator?"
What's the instrument error?
'9 I t % ' nie alqwt inchn1 8s 9 V) 19 MR. MYERS:
It's a determination by the H700 l
l l
20 guidelines, which utilizes an either/or level, eight inches or 21 below, or pressure below, and i believe that's 900 pounde.
22 Whether there is water or not, you should assume that that 21 generator is dry.
24 MM, DENTONr Why don't you go ahead, then, with your l
25 chronology?
l
20 1
MR. JAIN The steam and feedwater rupture control 2
system, as well as the aux 616ary feedwater system, utilizes 3
several pressure switches, either for actuation or for 4
protection of the pump or the steamlines.
Initially, for a 5
few years after we started up, we had recurring problems with 6
the pressure switches.
They were falling at a very fast rate 7
because of corros ion prob l ems in the diaphragm.
We have since gone to a modified design for the O
1 9
pressure switches, and since then, the failure rates of these 10 pressure switches have gone down considerably, which 11 essentially improves the reliability, because it eliminates T
,/
12 the potential failure mechanism for valves going inadvedtently 13 closed or so on.
14 As part of the NUREG-0737 requirements, we installed 15 a control-grade flow Indicator in each train of the aumellary 16 feedwater system, and we also Installed a safety-grade flow 17 Indicator on e a s.h train.
These indicators have since been 10 toch spec'ed.
19 MR. MYCRGi Those were part of the 0777 20 requirements.
Spocifically the control-grade. I believe, was 21 part of the original startup after TMl, and 0737 required a 22 safety-grade upgrade "3
MR. STOLZ:
Ted, as you go along, can you indicate 24 the dates when these were implemented?
45 MR MYERSr We're going to try to do that wherever
l 21 l
I we can s however, the concern is that the best dates that we 2
have right now are off the top of a couple of engineers' 3
heads.
We will get those dates specifically down, so we can 4
have a chronology.
Where we have them, we'll provide them.
5 The control-grade was prior to restart after TMl, 6
and the safety grade was sometime in 1902 I think it was, 7
1942.
4 MM, JAIN:
We have also modified the control room l
9 annunciator in order for the operator to be better able to 10 diagnose what has caused an aux feedwater pump or train 11 Inoperability, given that the pump has been called upon to y,
12 actuate during an accident or a transient, 13 Using that revised annunciator window and the 14 computer printout, we can tell what enactly has gone wrong 15 with the aux feedwater train and take remedial action to l
1 16 correct it.
17 MM. LAINUst These are additional annunciators?
19 "A
JA'N Th i s we e sa 141'ti m4' 1 * *'ar c n '. o r', for 19 which we revised the logic, so it's for him to tell emactly 20 what went wrong, i
21 As i mentioned earlier while describing the system.
22 we deleted the speed switch interlock from the auxlilary 23 feedwater peamp discharge valves and left these valves normally I
i 24 open w6th the local handwheel and the pushbutton stations 25 locked to minimize the number of valves that had to open on
_ ~ - _ _ - - - _
22 1
the discharge side.
2 in terms of the analyses we have done after TM1, the 3
startup and makeup pump pORV analysis, which talked about 4
using the startup feed pump, if you lost both trains of 5
aumtilary feedwater, there was also a B4W reliability analysis 6
done as part of NUREG-0737 requirements, which was done on the 7
auxillary feedwater system, and then in December '91 we 8
submitted a detailed ORA analysis of the Davis-tesse aum 9
feedwater system and the several improvements that could be 10 made to it to improve the reliability overall.
11 MM. DENTON:
la that the one that was done by EOS7 12 MR. JAIN:
Correct.
13 MM. CENTON r Was there also a Bechtel study?
Was 14 there a Bechtel study of the sum feedwater system reliability.
15 in addition to the EDS report?
16 MM. JAIN:
I don't think there ever was a Bechtel 17 reliability.
There was a 84W aux feedwater reliability 18 analysis done and submitted to the NRC.
19 MR. OENTON I Just to be sure I understand, you have 20 mentioned so far three reports:
the original report done in 21 response to bulletins and orders, and then wnst was the second 22 report that you mentioned, the one done by 84W7 23 MN, JAIN:
Right.
24 MR. OENTON:
That was the reliabtlity analys6s of 25 the aumillary feedwater system.
And then the third report was
23 1
1 the reliability analysis of the auxiliary feedwater system l
2 also, but done by EDS?
3 MR. JAIN:
Correct.
To a different scope and to a 4
different detall.
l j
5 MR. EISENHUT Let's see, in fact, the last one that 6
EDS did was really the report submitted under NUREO-0737 of 7
the PRA study?
MR. JAIN:
No.
We never claimed that.
The B&W 8
O analysis was part of the NUREG-0737 requirements.
The EDS was 10 submitted basically to assess what we in the company could do 11 to improve the overall reliability of the aux feedwater PT ep 12 system, what options.
13 MR. EISENHUT Was it the study that concludes that 14 a third train of aux feedwater is not required?
l 15 MR. JAIN:
Correct.
l 16 MR. MYERS:
The conclusion was that there was equal 17 Improvement in reliabilities that could be mado in utilizing 19 an installed pump and other procedural activities as the third 19 pump, part of that third capacity full pump.
20 21 l
22 23 l
l 24
\\
l i
25 I
24 1
MR.
JAIN:
There have been some procedural changes excuse me.
I'm sorry.
We have also undertaken a 2
3 comprehensive government governor improvement program, 4
and Rick Gradomski is going to talk about that 5
MR. DENTON:
We could also probably use an overall 6
government improvement program, too.
7 CLaughter3 8
MR. GRADCMSKl I am Rick Oradomski from Toledo 9
Edison.
In speaking to the governors that are installed on 10 the Terry Turbine / Byron Jackson pump auxiliary f eedwa t er-pump 11 turbine system, the Woodward governors that were originally
')
12 installed on this system were Woodward pOpL pneumatic type
}
13 governors that were modified with the addition of a Bodie 14 motor attached to the hand control knob to accept electric 15 pulse signals from the steam generator level control 16 There are two major points to this program other 17 than the fact that we continue to try to adjust and make 13 repairs as necessary to the modified popL system.
19 MR. DENTON:
When did this system go into operation?
20 MR. GRADCMSKl*
This was September 1977 or July 1977 21 when we started it up.
22 MR. MYERS:
it was the original operational system.
23 MR, GRADCMSKI From roughly startup unt6i May of 44 e9dd, we curitinuou to uos A and resulvw pr ov i em. weth the JUGL 25 system until it locame apparent to us in roughly late 1901,
O 25 1
early 1982 that some serious work and evaluation had to be 2
done on this in order to improve the reliability of this 3
system in general.
4 in May of 1982 the Woodward Governor Company began 5
conducting an evaluation of the Woodward governors supplied by 6
the Terry Turbine Company and modified by the Terry Turbine 7
Company for Davis Besse, e
in approximately September of 1983, we had completed 9
a major modification to eliminate speed setting problems 10 associated with adaptation of the electric motor delve to the 11 manual hand control knob.
At that point in time there was a 12 1983 refueling outage, and we made changes to both the 13 installed governors and the spare governor that we had, so all 14 three of them were now modified as a result of the exhaustive 15 testing program that both Woodward and Toledo Edison had 16 conducted in order to solve a lot of the recurring mechanical 17 problems that we were seeing.
19 in October of 1983, we began a program for the to qualification and installation of a replacement governor.
The 20 fixes that we had instituted in September of 1983 were at best 21 considered short term.
We knew that at this point in time we 22 couldn't continue to operate with the system as it was.
We 23 felt confident that the changes we had made had drastically 24 improved the reliability of the governors and that had been 25 proved out in testing, that the recurring problems that we had
26 1
seen had, in fact, been solved.
2 Since that time, I don't recall any of the recurring 3
problems with either jamming on the h6gh and low speed stops 4
nor the problems of the slip clutch recurring since that point 5
in time.
6 MR. MYERS:
I think in raapense to nna of the ear t y 7
bulletins and orders, we went back through the failures and O
desc'ribed those that'had happened over the period of time, k
9 Many of those were associated with valve operation as well as 10 governor speed settings.
This is the main thrust, that the 9
11 failures had been in that area.
We didn't mention that.
I og 8
10 just wanted to bring that in.
13 So go ahead.
14 MR. ORADCMSKl All right 15 Again, in October 1983, we began a program for 16 qualification and installation to improve the reliability of 17 the aun feedwater pump speed control He a result of that, in 18 November 1904 during our lagt redoel' q ao'aa*
'w a 'attsl'ad a 19 Woodward model p00 governor on auxillary feedwater pump No.
2, 00 and we plan to install at the 1906 refueling outagu, depending 21 upon the continued evaluation of the performance of the new 00 model POO governor, we intend to 6nstall a new model DOO 23 governor on the No.
I auxillary feedwater pump-24 MR. DETER 3' l'm Hob poters.
I'm going to discusa 25 the procedure generation and modif6 cation that we went through
27 1
to support the use of the electric-driven startup feed pump.
2 prior to our restart from the TMl accident approximately May 3
1979, we generated a procedure that allowed the operators to 4
utilize the electric-driven startup feed pump to feed one 5
steam generator at a time.
Although we are not sure of the 6
emaet dates, early on we provided the capab6lity of power 6ng 7
the electric-driven startup feed pump from one emergency 8
diospl generator, and subsequently we modified the plant to 9
allow the startup feed pump to be powered from either 10 emergency diesel generator.
11 As Mr. Jaim mentioned earlier, in September of 1990, i
12 we modified the procedure again to reflect the plant j
13 modification that allowed the reset of the SFRCS trip on the 14 main feedwater block valves from the main control room.
This 15 allowed the control room operator to regain control of the 16 main feedwater block valve to utilize the startup feed pumu 17 feed the selected steam generator.
14 Then in July of 1901, we again mod 6 fled the 19 procedure to reflect the modification to the plant to allow 20 use of the startup feedwater valve I'm sorry. Rega6n 21 control of the startup feedwater valve from the control 22 rocm. That, again, would be to reset the steam rupture control 21 system trip from the control room.
.e And then f6nally in January of this year, January 25 1905, this procedure was incorporated into our abnormal 1.
9 28 1
transient operating guidel6nes, the symptom-based operating 2
procedures that we have utilized that were developed 44 4 3
result of the TM1 accident.
And that's it.
4 NM. NYERS That brings you up to current changes.
5 There are several additional changes that are being planned 6
that are down in the areas of lower contributors but are still 7
in our plans in addition to the mode of the startup feed pump.
- 8 Sushul, would you like to cover those?
g NM. DENTON I would like for you to be sure and to cover the status of the latest amendment.
I guess it was 11 Amendment 83 that requires Installation of a new startup 1
12 feedwater pump prior to starting cycle No.
6.
13 NM. JAIN6 As a result of the AMA study that we 14 submitted in 1981, we had identified there that one of the 15 most dominant contributdes to the aux feedwater unrollability 16 was the fal' lure of the motor-operated valves.
To that end wo 17 designed and engineered several changes so as to reduce the 10 number of valves that have to open on demand for the aux to feedwater system.
20 We are planning to leave the discharge valve on this 21 drawing 3670 and 72, leaving them normally open and locked 22 open, so that there wouldn't be any valve in the discharge of 23 the pump which will need to be opened to provide water to the 44 steam generator.
25 The other change that we are planning to make la the
'o 29 1
change to the logic of the steam and feedwater rupture control 2
system, as I had mentioned earlier.
On low level or loss of 3
feedwater condition, the respective steam generator provides 4
steam to the corresponding aux feed pump turbine through its 5
normal path, MS-106 or 107 in this path.
6 The logic is to be changed such thac when steam 7
generator can previde steam to both aux feed compartments 8
in other words, two valves will open, providing two paths, two 9
redundant paths for a given aux feed pump turbine, so if one to path or one valve falls to'open, the other path can still 11 provide steam to the aux feed pump turbine.
12 So essentially, each aux feed pump turbine has got r
18 two paths of steam for it to be run.
14 We are proceeding on designing changes to improve 15 the operation of the steam inlet valves, The steam inlet 16 valves have several interlocks.
One interlock la the one that 17 closes these valves if you had a break in the steam inlet l i ne 18 to the turbine itself, We have pecposed a tech spec change to 19 the NRC to delete that interlock so that the closure of the 20 valve is eliminated and thereby eliminating a potential 21 failure by that valve, 22 These valves are also interlocked with the decay 23 heat drop line valves, and the interlock has been there 24 because once you go to the decay heat mode, you trip all your 25 reactor coolant pumps, and tripping all four reactor coolant
~
l 30 1
pumps starts the auxiliary feedwater s y. t te..
st Davis Besse, so 2
the interlock was provided to prevent in*fvertent start when 3
you are going to the decay heat mode.
4 We are proceeding to delete that interlock and 5
taking manual action to de-energize the valve, which we 6
already have been doing for t%e last six or seven years anyway 7
to remove power from those valves so they don't come on open 8
to start the aux feed pump turbine.
9 These two modifications essentially reduce all the 10 control systems or the interlock failures that could 11 potentially contribute for these valves to fall to come up.
.I 12 As part of the control room de. sign review, we have 13 Identified several changes to the steam generator system, as 14 well as the SFRCS.
We are going to be providing a redundant 15 steam geneerator level and pressure indication in the control 16 room so the operator has better knowledge of the steam 17 generator status as far as what the level is and what the IA pre 1sure IS 51 tse ss its usability for now '9meus'7e Tys*em 19 We are also going to be relocating some of the SFRCS 20 manual trip switches to enhance human engineering.
21 The normal suction supply from the condensate 22 storage tank has a valve that is normally open to provide 23 suction to the aux feed pump turbine, and that goes closed on 24 a low suction pressure to transfer to service water, which is 25 a seismic suction for the pump.
31 1
One potential failure mechanism would be that the 2
valve in the suction from the CST could spuriously.go closed.
3 thereby robbing the aux feed pump of water for a few seconds 4
before it is called upon to actuate, and then it transfers to 5
We are proposing to delete or remove power 6
from that valve so that particular failure mechanism le again 7
eliminated.
'~
8 And finally, we are putting in a new startup feed 9
pump which is of a higher capacity, which has a capability of 10 feeding water into the steam generator, both in the main 11 feedwater nozzles as well as the aux feedwater nozzles.
We 12 were planning to implement that in the next refueling outage.
13 MR. DENTON:
What would the capacity of that pump 14 be?
15 MR. JAIN:
As I recall, 600 gpm, but we could 16 provide the exact number.
17 MR. DENTON:
So that would be an estimato of the IR eqia l va l en t e,apacity.
19 MR. JAIN:
It is equivalent to 100 percent capacity 20 decay heat removal aux feedwater pump.
21 MR. DENTON:
Have you had any studies done of 22 reliability of this system since that one that you mentioned 23 done by EDS?
24 MR, JAIN:
What we have is we have internall=ed most 25 of the risk assessment work, We have fault trees and other
'l
32 1
models made up for the system, but we are unable to support at 2
the present time any numbers in order to compare what the 3
numbers were when we submitted the EDS pRA study.
4 MR. DENTDN:
Could you summarize what the EDS study 5
stated?
6 MR. JAIN:
The premise of the EDS s*udy was to 7
evaluate what we should be doing as far as spending our money 8
on installing either a third aux feed pump or otherwise A
9 improving the existing aux feedwater system to include the 10 reliability, and the analysis-based configuration in that M
11 report was the one that was aimed at addressing each of the
. A 12 most dominant contributors to the unreliability and making 13 changes to the system to eliminate those dominant 14 contributors.
15 There was a third change analyzed in the report, 16 which analyzed a third aux feed pump, if you will, and the 17 analysis showed that you have a bigger improvement in 18 reliability for the analysis-based configuration for a lesser 19 expenditure, and you had lesser improvement in reliability for 20 the third aux feed pump with a much greater expenditure.
21 MR. MYERS:
I think the third the moving of the 22 startup feed pump was initiated and actually has several other 23 functions in addition to providing 100 percent auxiliary 24 feedwater backup to our concept it eliminates some of our 25 operational concerns about limited startup flow, it gives us
l+
33 1
improved operability in the normal function of the. system, and 2
also helps separate some of our fire hazards as under the l
3 Appendix R activities for control circuits and shutdown panels 4
and that sort of thing.
5 So there are quite a few issues that moving the 6
startup feed pump into a different area and increasing its 7
size actually addresses.
~
If your configuration on the auxiliary 8
MR. DENTON:
9 feed pump system is different than ot,her B&W plants, as i 10 recall, do you happen t o r emember wna t the decision process 11 was leading to that decision?
)
12 MR. MYERS:
Maybe there is some here that can.
13 help. In reconstructing -- and that's as good as I can do.
14 trying to put it together I think if we remember 15 timeframe-wise, Davis Besse was late in the licensing period 16 to where safety-grade auxiliary feedwater systems were being 17 discussed as requirements, and Davis Besse perhaps was the 18 first to actually enter design a raquired safety-grade system, 19 safety-grade initial operation and control of the B&W units, 20 the 177 B&W units.
21 At that time the -
~l'm not sure exactly how this f
22 particular configuration came out, but it was driven a lot on 23 this feedwater rupture control system, and the dedicated i
24 discharge you notice that is fairly unique pump to 25 generator rather than two pumps to header arrangement.
One I
~*
l l
34 1
pump to one generator, and if you need cross-connectors, you 2
dedicate cross-connect, so that original thought process in 3
the instrumentation and in the original system seems to have 4
been generated in those days of converting from a normal 5
control grade for older plant's aux feedwater system into a 6
fully safety-grade aux feedwater seismic qualified 7
control-started system.
8 MR. DENTON:- is that normally under the jurisdiction 9
of the vendor or the architect engineer?
10 MR. MYERS:
Architect engineer. The basic 11 requirements for steam generator cooling are provided by the C
12 vendor s however, the particular design is architect engineer 13 specific, and safety-grade systems being new, probably was 14 fairly unique, I would imagine, during that timeframe.
l*m 15 not sure if maybe someone here can help illuminate, who was in 16 a position at the time who could help.
This was the '74 17 through '76 timeframe.
18 MR. DENTON:
Well, I had heard at least I had a 19 memory, a recall that you had a study done, a Bechtel study 20 done, but you are saying you don't think Bechtel was ever 21 involved.
22 MR. MYERS:
In the original design?
23 MR. DENTON:
I'm not sure.
I thought there was l
24 Bechtel involvement.
25 MR. MYERS:
Absolutely.
I'm sure there was Bechtel t
35 1
in the original design of the system, and Consolidated 2
Controls was our vendor for the steam and feedwater rupture 3
control system. So in the original design, that definitely was 4
the case.
5 MR. JAIN:
The effort that you may be thinking about 6
would be maybe a study that we had 8echtel do trying to define 7
different options for a third aux feed pump, and they did a O
cos t-ana l ys i s as to which alternative was going to cost how D
much, and that's the one we were quoting when we met here four 10 or five years ago with the NRC would cost $6 million or $7 11 million.
That's the only involvement that I can think of in h
)
12 those timeframes.
13 MR. DENTON:
That's probably the one i remember.
14 MR. ROWSCME:
I can clarify the record a little 15 bit.
There was a system reliability analysis done at Bechtel 16 by me.
I was at Bechtel at the time.
17 CLaughter.]
18 19 20 21 22 23 24 25
36 in the 1974-75 period.
It was not at the request of 1
2 Toledo Edison.
It was submitted to the project design team 3
and i presume passed on to you all But it was not at your 4
request.
5 ft did not result, to my knowledge, in any design 6
changes.
7 MR. MYERS:
Do you recall in this timeframe of the O
or i gJ na l design, wher.e you were off project?
9 MR. ROWSOME:
I was off project.
It was supported 10 under overhead as an exercise in developing Bechtel's 11 capability in using fault tree analysis and system reliability 12 techniques, and it was not requested by the project.
13 MR. MYERS:
So you weren't really disowning the 14 actual design?
15 MR. ROWSOME:
No, it wasn't part of the design 16 effort.
17 MR. MYERS:
Good timeframe, though.
18 MR. DENTON:
Well, does that complete your 19 presentation?
20 MR. MYERS:
That completes our presentati6eT.
- But, 21 if there are any questions, again as I said, we are in the 22 process now, still now, trying to actually get dates for 23 chronologies here.
c4 Any discussion or questions con c er n i rig wither the 25 studies or
37 1
MR. DENTON:
What I wanted was to be sure that we 2
had as complete a record of the studies and the physical 3
changes to the plant that we could amass.
4 Are there other studies you have now underway in the 5
system, or is there anything more you are doing with it, or do 6
you see the completion of the effort lugt described as 7
providing a satisfactory feedwater system absent some further 8
probiems?
9 MR. MYERS:
Let me cover the philosophy which 10 wasn't, obviously, in the discussion of the individual items.
11 In our early review of the system it was felt that s
12 having the installed system being somewhat unique and somewh a t 13 inflexible to change, particularly without having a common 14 header on the discharge like many of the newer designs, would 15 have to allow options of dropping pumps onto the headce and 16 that sort of thing, that we work on the major contributors to 17 unreliability of the installed septic rate aux feedwater 18 system.
Recause, as we saw them, they wera quite identit3sbfe 19 and quite attackable.
The governor valve systems and 20 elimination of potential for new signals coming into isolate.
21 In addition we would provide procedurally a backup 22 to the condition of loss of all main feed and auxiliary feed, 23 and provide that to the operator, and also continually improve l
24 that for reliability.
So, just bringing an additional power l
25 supply capability to it, and then from either side of the j
i l
l
38 1
plant, as we saw the opportunity to actually move the startup 2
feed pump to not only increase the size, but also increase the 3
capability to deliver directly into the feedwater nozzles, 4
which is highly desirable on the B&W generic design.
We 5
certainly did not let that opportunity go by at all 6
The intention is to continue to improve the 7
as-installed system while we are providing an upgrade also in
~
8 the startup date of system supports.
1 9
MR. DENTON:
Will this new system you are describing 10 actually replace the previous system?
11 Cr,'will you keep it in addition to the system you m
. k 4
j 12 are talking about, talking about the upgraded pump and motor?
13 MR. MYERS:
The upgraded pump and motor will be 14 normally use to replace the startup feed pump.
It is in place
~
15 now.
16 MR. DENTON:
It will go in the same place in the 17 plant?
18 MR. MYERS:
Ch, no No We are moving 3t te an 19 entirely different location in the plant.
The size and piping 20 limitations are one of the reasons for moving it.
The 21 piping energy line break considerations, and also fire 22 considerations for moving it out of fire areas to give us 23 additional support so it will be moved into a different 24 portion of the plant, the turbine area where there is more l
l 25 room, a capability for a different pump activity and cross
i l
1 39 1
connects for water supply and discharge lines.
2 MR. EISENHUT:
When is Cycle 6 scheduled?
3 MR. MYERS:
It was scheduled for spring of 1986.
4 The long lead time components, as I believe Dick Crouse had 5
mentioned to you, they are already on order.
6 Again, we ara trying to get the detall* of when they 7
were put on order to provide to you.
But that design process
~ '
8 is well underway and" procurement is underway for that.
9 MR. EISENHUT:
How much time is programmed on your 10 schedule in Cycle 6 to actually do the installation, do you 11 know?
Is it a three month job, six month job 7 12 MR. MYERS:
I'll have to find that out.
Actually 1 13 can give you the philosophy also of the location.
It is 14 relatively free of activity currently.
I don't want to say it 15 is a dead spot in the plant where the square footage is doing 16 nothing.
But, a lot of the work was planned to do pre-outtage 17 so that we would not impact in working around the area with 19 more important pramps while we are acartting, but we 'vou l d be 19 able to do a lot of that construction independent of our 20 operation.
It would not hamper that.
21 I can certainly find out what the construction 22 module says now for actual outage time.
That is a dedicated 23 plot in our planning.
24 MR. DENTON:
Any other questions?
25 MR. THADANI Do you have procedures today,
40 1
recognizing the low capacity of the start-up feed pump, in the 2
event of extended loss of main and aux feed?
Can the 3
operators perform combinations of actions, the start-up feed 4
pump as well as the other mechanisms for removing decay heat?
5 Do you have procedures in place 6
First of a l l, is it feasible?
7 Second, if it is feasible, do you have procedures
" ~
8 today in place to be 'able to remove decay heat by a multiple 9
source of actions?
The start-up feed pumps, maybe open up the 10 PORU, try to get the pressure down s that kind of stuff?
11 MR. MYERS:
Yes.
The procedure that Bob peters o
12 mentioned earlier, the one that we originally developed, was 13 called an abnormal procedure, total loss of main and auxiliary 14 feedwater, and specifically did give instructions to utilize 15 and make up the high head capability of the makeup system 16 utilizing the pilot operated relief valve and the start-up 17 feed pump to provide full decay heat capability, and that's 18 osar procedural action and has been in place since that 19 procedure was initially approved.
20 Since then, this last cycle, we converted to the 21 ATOG, the symptom-oriented guidelines and incorporated that l
l 22 whole thing into our current program, current procedures i
23 program and that, along with, of course, the recovery of 24 auxiliary feedwater, is stressed in the procedure and actually 25 was the procedure we followed, the process we followed, last
c.
41 1
summer and it is a result within, I believe, two minutes of 2
when we lined up the feed pump and the auxiliary pumps were l
3 recovered also.
4 So, yes, that is proceduralized and is an integral 5
part of our ATOG program.
6 MR THADANI in reference to some analysis that was 7'
done by you and supported this procedure, I assume?
8 MR. MYERS:
We have done analysis to support it i
9 don't know what the ATOG references are.
We can find those 10 out.
They are probably in Volume 1 of the book, of the ATOG 11 guidelines for Davis-Besse.
We can dig out exactly what t
[
12 the actual emergency procedure in the plant probably does not 13 have that as "this references such and such a"
14 MR. THADANI.
No, I understand that, but the
~
15 procedures were developed on the basis of whatever analyses 16 were done.
17 CMr. Myers nodding.]
te MR. RONSCME:
Two technicsf questiens on whst 19 Mr. Jain told us.
I didn't quite follow what you said about 20 dropping the auto start on the trip of the reactor coolant 21 pumps.
How is that being implemented?
22 MR. JAIN:
These valves get an open signal from the 23 steam and feedwater rupture control system, when we have lost 24 all four reactor coolant pumps.
25 Now we are proposing to and we have been doing this
e s
42 1
ever since we started up, is when we go into the decay heat 2
mode, we take the power off of the steam inlet valve so the 3
interlock really never has been used, per se.
4 So we are deleting that interlock so that you are in 5
mode 1 operation, nothing can go wrong in the interlock to 6
cau<a the valve to go closed, you know, a spurious failure.
7 MR. ROWSOME:
Instead of manually depowering the 8
- valvbs, you are going to change the logic so that they are 9
automatically cut out of the logic when you are in mode 57 10 MR. JAIN:
We will be depowering it still, but we 11 don't need the interlock any more.
I guess we never needed 3.
~'
12 the interlock, but it was put in there for that purpose, so 13 that you don't start the aux feed pump turbine.
14 Maybe I am not very clear.
~
15 MR. ROWSOME:
I don't quite follow, but I am 16 reassured.
We will find out in due time.
17 You are saying that you would lock open the 19 discharge valves ?970 and ?8729 19 MR. JAIN:
Yes.
20 MR. ROWSOME:
I think that gives you trouble with 21 isolating a main steamline break, does it not?
Do you depend 22 on manual action to isolate the affected steam generator' 23 MR. JAIN:
This is a motor-operated valve and i
i 24 locking it open in the field, the hand wheel and the local 25 push button station doesn't disable the control from steam and
43 1
feedwater rupture control system.
2 MR. ROWSOME Oh, it's just the local control and 3
hand wheel All right.
4 MR. JAIN:
So most of your SFRCS actuations are on G
5 the low level.
6 MR. ROWSOME:
So it will still be normally closed, 7
it will just be that the --
" An' operator in the plant could not 8
MR. MYERS:
9 misposition the valve because it's locked.
He would have to 10 have there's a rigorous control for him to change the 11 position of that valve.
[.I 12 MR. ROWSOME:
But it remains normally closed?
13 MR. JAIN:
Open.
We are proposing it to be open.
14 MR. EISENHUT:
Open, and the logic would close it.
~
15 MR. ROWSOME:
"And you are counting on check valves 16 to constitute the pressure boundary?
17 MR. JAIN:
Right.
tG MR CENTCN.
Let me ssk about the main feed pumpJ.
19 Once the main steam line isolation valve closes, are you able 20 to bring them back to service again following a reactor trip?
21 What prohibits those from being used as " auxiliary" feedwater?
22 MR. MYERS:
The steam Supoly essentially would not 23 be available to run them because there are large steam needs.
24 I understand, like in Combustion Engineering plants, they talk 25 about condensate booster pumps and things like that.
We have
44 1
not pursued the evaluation of that type of additional backup 2
to date.
We felt much more reliable feed pump pressurewise 3
and availability and power supply-wise 4
MR. DENTON:
Well, then, a normal unplanned trip 5
what role does the main feed pump play?
6 MR. MYERS:
In just a normal plant trip without a 7
steam and feedwater rupture control signal?
MR. DENTON:
Whatever would normally happen, e
9 assuming no equipment failures.
10 MR. JAIN:
Steam and feedwater control system, the M
the integrated control system will take the 11 ICS will take
-)
. 4 12 plant to a steam generator low level condition where the main 13 feed pumps will be used to maintain level at 35 inches in the 14 steam generators and removing decay heat thereby.
15 Once the pressure in the steam lines has gone down
~
16 significantly, you can either go on the auxiliary boiler to 17 run your main feed pump turbines or you can go to the decay 18 heat mode.
That's the long cooldown process without the 19 MR. DENTON:
What's the auxiliary boiler?
20 MR. JAIN:
It's an oil-fired boiler that we use for 21 station heating, for example, when the plant is not running, 22 or v9 also use it for initial warming up of the plant's 23 chilled steam for the turbine.
It is shut down during normal 24 operation.
25 MR. DENTON:
If you had a loss of all feedwater, is
4 l
45 1
that the system that you try to bring into operation?
2 MR. JAIN:
We normally don't 3
MR. MYERS:
I believe that the start-up time on the 4
auxiliary boiler is quite significant and I'm not sure if we 5
actually consider it able to be brought on 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day.
6 ft's able to be started, but i beileve it takes a while to get 7
that system going, and it would actually be 1 don't want to 8
say a detraction froE the main recovery mode, but it would be 9
much less reliable and swift to recover feedwater than the 10 start-up pumps or recovering the aux turbines.
11 MR. DENTON:
Do you happen to remember the
%A,-
12 conclusion of the reliability of the aux feed system that EDS 13 reached?
14 MR. JAIN:
As to what we should be doing?
~
15 MR. DENTON:
No, how reliable, say, the system was 16 likely to be, either in this state or upgraded.
17 MR. JAIN:
As far as the numbers?
19 MR DENTON:
Yes 19 MR. JAIN:
I don't r ememb er it, but the report has 20 the numbers in there.
I don't remember it right now.
21 MR. MYERS:
That was submitted.
22 MR. DENTON:
l'Il look at that later.
Let's see if 23 there are any other questions from the staff or other 24 parties.
And if not, I propose we take a caucus for the staff 25 and mull over what we have heard and see if we have some more
1
,I 46 1
questions.
But before we break, we'll see if anyone has any 2
further questions.
3 is there any other information you want to tell us 4
about the system that you've got or actually plan to take?
5 MR. WESSMAN:
One quick question.
Maybe it's 6
outside the scope of what you-all wanted, but can you 7
summarize some of the changes that are going on with the main O
feed system?
9 1 understand you have had some difficulty with 10 governors on main feed pump turbines.
11 MR. MYERS:
We can probably give you 30 seconds.
We d
h 12 converted the complete main feedwater pump control system in 13 this last outage.
It's a General Electric system 14 modification.
We have had some problems, as a matter of fact, 15 prior to last Sunday's event.
We even did special 16 instrumentation of this system specifically to try to nail 17 down if any control failures occurred, and that's hopefully 18 when we get into with the fact-finding team, intu the 19 machinery, that instrumentration is going to give us quite a 20 bit of input.
But it's a complete change-out.
Our main 21 feedwater pump control guru is right now working with our 22 team, the fact-finding team, back there.
We can respend to 13 yois more in the future, or if you have specific que s t i ons, we 24 can probably get specific answers today for you on that.
But l
25 the staff here that Information is not available.
l
Y 4,
47 1
MR. DENTON:
Okay.
(Jhy don ' t we break until say a 2
quarter after 12:00.
That will give us time to see if there 3
is other information or questions we might have on what you 4
have told us.
5 CRecess.3 6
7 S
9 10 11 1
12 13 14 15 16 17 18 19 20 21 R3 23 24 25
i
)
', I i
)
48 1
MR. DENTON:
Let's resume the meeting.
We don't 2
have any additional questions to ask.
What you told us this 3
morning has been very useful to help us reconstruct events and 4
actions that you have taken.
5 What we will do is coordinate with our team at the 6
site to be sure that any follow-up questions that we ask of 7
you are coordinated with' them.
I think some of the types of 8
questions that come naturally to mind are whether the EDS 9
study that was done for you sometime ago used what I would 10 call generic failure rates as opposed to plant-specific 11 failure rates, and whether if you were using plant-specific 12 performance data over the past couple of years or you had 13 gotten a different result, and the question of whether it is 14 safe to resume operation with the type of failure rates that 15 we are experiencing is t,h e type of question that leaps to 16 mind.
17 But I don't have any specific requests to make for 18 additional information of you at this time.
What I will do is 19 coordinate with Mr. Rossi and others to be sure that we're not 20 having two different arms of the NRC asking for information, 21 so l*ll just mention these items as one that I think follewed 22 fece the discussion we hada namely, you had a reliability 23 study done, it looks like the system was getting upgraded, 24 you t h o u g h.t it was being upgraded, you worked on the valves 25 and control systems and those kind of things.
And then we had
49 1
both systems that didn't work, and that naturally leads to the 2
question of why the difference between what you had expected 3
from a previous pRA analysis and real life, and that's a 4
scenario that we will be looking into back here and with 5
Mr. Rossi.
6 MR. MYERS:
Harold, I would like to make one cemment 7
that I think you can feed back to your staff now, and if the 8
report doesn't stress it we should note that in the EDS pRA 9
that was done, plant-specific failure rates were used, as a 10 matter of fact, and because of that we did see specific 11 differences between generic and industry, and that started D
,j 12 some of our follow-on work with Limotorque operators and some 13 other consultants to actually -- in that comparison of our own 14 failure rates individually to industry's.
And we took some 15 action there.
16 So I believe that maybe doesn't stand out in the 17 report but it should be in there somewhere, the specifics.
10 MR. DENTON:
De you have a feel for whether yeur 19 plant-specific experience up to a few weeks age would give the 20 same estimate that was used in the EDS report, whether 21 patterns of changes in the reliability of 22 MR. MYERS:
I think tha patterns that changed we saw 23 were improvements.
Like I said, the main contributors were 24 quite well defined and, in our corrective actions, were 25 noticeably turned, both in valve operation and in governor
e 50 1
performance, which were the main contributors to system 2
unavailability.
3 in those particular areas I think we can say that in 4
going back and revisiting the data now, l*m sure we will see 5
past performance had reflected that.
Of course, the exact 6
caaJa of the two turb no; being unavailable is a major focus 7
of our investigation, and with the factfinding team activity 8
at the site, both of us are anxious to get in there and A
9 validate our findings.
And ! think hopefully, that will shed 10 a lot of light back on the analysis to reflect whether it was 11 something that should have been foreseen or something outside
. A i) j 12 the procedures of PRA.
13 MR. DENTON:
That is something we'll have to await 14 for respective efforts to be completed.
15 All right, if there are no more questions or 16 comments, I appreciate your coming in on such short notice.
17 Thank you.
1:3 CWhereupon. at 12.20 p.m no meeting was ad,. ve r eo. J 19 20 21 22 23 24 25
r-r e
6 5 1
CERTIFICATE OF OFFICIAL PEpORTER 2
3 4
5 This is to certify that the attached proceedings 6
before the United States Nuclear Regulatory Commission lo the 7
matter of:
g 9
Name of Proceeding: Meeting Between the NRC Staff and Toledo-Edison Company Concerning AFW Systems 10 11 Occket No, P-
)
12 Place-Bethesda, Maryland 13 cate:
Monday, June 17, 1985 14 15 were held as herein appears and that this is the original 16 transcript thereof for the file of the United States Nuclear 17 Regulatory Commission.
13 (Signature)
(Typed Name of Reporter)
And Riley 20 21 22 23 Ann Riley & Associates, Ltd.
24 25
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J 9 I ENCLOSUPE 2: ATTAOfENTS SUPPORTING QUESTION 20 i
l QUESTION 20, PLEASE PROVIDE THE FOLLOWING INFORMATION FOR EACH YEAR SINCE 1980: (A) THE NUMBER OF 10 CFR 50,72 REPORTS. ANSWER. YEAP NUMBER OF 10 CFR 50.72 REPORTS 1980 12 1981 14 1982 7 1983 18 1984 14 1985 (T0 JUNE 19) 11 MARKEY/IE JUNE 21, 1985 a
.JEF \\ QUESTION 20. (B) THE TOTAL NUMBER OF LICENSEE EVENT REPORTS, SPECIFYING THE NUMBER ATTRIBUTABLE TO PERSONNEL ERROR, EQUIPMENT FAILURES, AND DESIGN OR FABRICATION ERRORS. ANSWER.
SUMMARY
OF LERS FOR DAv!S BESSE FOR 1981-1984 CAUSE STATED IN LER ABSTRACT NO. OF DESIGN / YEAR LERS PERSONNEL EQUIPMENT FABRICATION OTHER 1981 79 33 37 8 1 1982 68 28 23 13 4 1983 74 19 38 10 7 1984 2? 8 7 6 1 TOTALS 243 88 105 37 13 MARKEY/IE JUNE 21, 1985 0
gj-v S 00ESTION 20. (C) THE NUMBER OF UNUSUAL EVENTS, ALERTS AND ABNORMAL OCCURRENCES.
- ANSWER, THE FIFTEEN (15) DAVIS-BESSE UNUSUAL EVENTS ARE AS FOLLOWS:
DATE DESCRIPTION 04/13/80 FLOOD WATCH 06/25/80 MINOR FIRE IN CONTROL ROOM 07/29/80 EYE INJURY NON-NUCLEAR RELATED 02/04/81 LOSS OF METEOROLOGICAL INDICATION 02/19/81 LOSS OF METEOROLOGICAL:lNDICATION 12/16/82 OVERTURNED GASOLINE TRUCK IN AREA 01/18/83 LOSS OF CONTAINMENT INTEGRITY (NO. 2 MSIV LEAKING) 12/17/83 REACTOR COOLANT LEAK (LETDOWN SYSTEM ISOLATION VALVE PACKING LEAK) 01/17/84 LOSS OF METEOROLOGICAL INDICATION 02/21/84 LOSS OF METEOROLOGICAL INDICATION 03/02/84 MAIN STEAM LINE SAFETY VALVE STUCK OPEN 05/02/84 LOSS OF METEOROLOGICAL INDICATION 05/06/85 LOSS OF METEOROLOGICAL INDICATION l 05/16/85 RCS LEAK / PRESSURIZER SPRAY VALVE PACKING 06/09/85 REACTOR TRIP / LOSS OF MAIN FEEDWATER/ AUXILIARY FEEDWATER MALFUNCTION MARKEY/IE JUNE 21, 1985 e
A ~ QUESTION 20C. (CONTINUED), i THE DAVIS-BESSE ALERTS ARE AS FOLLOWS: NONE. i l l THE DAVIS-BESSE ABNORMAL OCCURRENCES ARE AS FOLLOWS: 1 1981 - NO REPORTS. 1982 - IN NUREG-0090, VOL. 5, NO 2, INFORMATION WAS REPORTED ON l DAMAGE THAT OCCURRED TO THE-STEAM GENERATOR AUXILIARY FEED-i j WATER HEADER IN NO 1 AND NO. 2 STEAM GENERATORS. THIS l INFORMATION APPEARED IN THE UPDATE SECTION-OF THE REPORT UNDER ABNORMAL OCCURRENCE 76-11, STEAM GENERATOR PROBLEMS. I i 1983 - NO REPORTS. I j 1984 - NO REPORTS. t 4 I 1 i MARKEY/IE i JUNE 21, 1985 1 I'. . _ -,.. ~., _ _ _ - _ _., _ _ _ - _ _. _.. _ ~. _,
OUESTION 20. (D) THE NUMBER AND CAUSE OF REACTOR TRIPS. i ANSWER-1 OVER THIS PERIOD OF FIVE AND ONE-HALF YEARS, FORTY (40) REACTOR TRIPS OCCURRED,' THE CAUSE OF EACH TRIP AND ITS DATE ARE SHOWN IN THE FOLLOWING TABLE, A LIST OF CAUSE DEFINITIONS IS PROVIDED AT THE END OF THIS TABLE. FOR 1984 AND 1985, REACTOR TRIPS WERE ANALYZED TO DETERMINE IF AUXILIARY FEEDWATER PROBLEMS OCCURRED d AFTER THE TRIP, THOSE ARE MARKED WITH AN ASTERISK, "*". THIS ANALYSIS WAS NOT DONE FOR OTHER YEARS.. I DATE CAUSE i 02/05/80 HARDWARE BELIEVED CAUSED BY MAIN STEAM ISOLATION VALVE CLOSURE. i a 03/27/80 HARDWARE ELECTRICAL FAILURE IN HEj(10DCONTROL
- SYSTEM, i
04/07/80 HARDWARE HIGH FLUX DURING MAINTENANCE ON i a HUMAN ERROR FEEDWATER HEATER CAUSED PRESSURE INCREASE IN CONDENSER AND THE AUTOMATIC INTEGRATED CONTROL L SYSTEM (ICS) ATTEMPTED TO CORRECT. i MARKEY/IE JUNE 21, 1985 c -s,-.- ,. -. =,,.,, ,m-
4. t GUESTION 20D.'(CONTINUED) 4 s DATE CAUSE 11/06/80 NOT PROVIDED BELIEVED TO BE DUE,TO ELECTRICAL s PROBLEM IN SWITCH YARD. 4 l 11/07/80 HUMAN ERROR PERSONNEL ERROR COMPLICAT'ED BY ICS CONTROL. l 11/12/80 HARDWARE ELECTRICAL FAILURE. .l 03/11/81 HUMAN ERROR ELECTRICIAN ERROR LED-TO MAIN STEAM ISOLATION. VALVE CLOSURE, 04/25/81 HUMAN ERROR DURING STARTUP, TRIPPED ON LOW STEAM j GENERATOR LEVEL DURING TEST OF MAIN \\ FEEDWATER BLOCK VALVE. STARTUP~ FEED l PUMP DID NOT RESPOND QUICKLY ENOUGH. i 05/12/81 HARDWARE ELECTRICAL FAILURE'. i 06/24/81 HARDWARE LOSS OF DC RESULTED IN LOSS.0F NON-NUCLEAR INSTRUMENTATION. i MARKEY/IE i-JUNE 21, 1985
1 QUESTION 200, (CONTINUED) i DATE CAUSE I 07/30/81 NOT PROVIDED 09/02/81' NOT PROVIDED MANUAL TRIP. i 10/22/81 PROCEDURE TRIP DURING USE OF SURVEILLANCE TEST PROCEDURE. 12/28/81 NOT PROVIDED HIGH FLUX TRIP (POSSIBLY SPURIOUS) i WHILE EXERCISING RODS. 10/28/82 PROCEDURE CONTROL VALVE TESTING. I 10/29/82 NOT PROVIDED MAIN FEED PUMP TRIP DURING SURVEILLANCE TESTING. 11/08/82 HARDWARE FAULTY MOISTURE SEPARATOR-REHEATER l LEVEL SWITCH. 01/15/83 HUMAN ERROR UNTIMELY CORRECTION OF AXIAL FLUX i l IMBALANCE. i MARKEY/IE i JUNE 20, 1985 s
e QUESTION 20D. (CONTINUED) DATE CAUSE 01/18/83 HARDWARE ELECTRICAL FAILURE, BLOWN NON-NUCLEAR INSTRUMENTATION FUSE CAUSED HIGH PRESSURE TRIP. 01/31/83 HUMAN ERROR DUE TO FEEDWATER FLOW REGULATING VALVE BEING CYCLED. 04/10/83 HUMAN ERROR NEGATIVE AXIAL FLUX IMBALANCE. 05/10/83 HARDWARE ELECTRICAL FAILURE OF AN INVERTER. 07/25/83 NOT PROVIDED ACTUATION OF STEAM AND FEEDWATER RUPTURE CONTROL SYSTEM FOR UNKNOWN REASON. 10/02/83 S/G LEVEL TRIP DUE TO MANUAL FEEDWATER CONTROL. 10/02/83 HARDWARE ELECTRICAL FAILURE IN INTEGRATED CONTROL SYSTEM (ICS). 10/03/83 HUMAN ERROR ICS IN COMBINATION WITH HUMAN ERROR. MARKEY/IE JUNE 20, 1985 i L.
= i l QUESTION 20D. (CONTINUED) : DATE CAUSE 10/15/83 HARDWARE CONTROL PROBLEM WITH MAIN FEED PUMP. i i 11/09/83 HARDWARE ELECTRICAL FAILURE IN CONTROL ROD l DRIVE SYSTEM. J 11/14/83 HARDWARE ELECTRICAL FAILURE IN CONTROL ROD r i i DRIVE MODULE. 12/17/83 HARDWARE ELECTRICAL FAILURE IN REACTOR PROTECTION SYSTEM ANTICIPATORY J iRIP CHANNEL. 01/09/84 HARDWARE. ELECTRICAL FAILURE IN THE CONTROL ~ ROD DRIVE SYSTEM. 03/02/84* HARDWARE MAIN STEAM ISOLATION VALVE CLOSURE. 06/24/84 HUMAN ERROR OPENED WRONG TRIP BREAKER DURING SURVEILLANCE TESTING. t i j MARKEY/IE l JUNE 20, 1985 i i
j - QUESTION 200. (CONTINUED) i DATE CAUSE i 09/11/84 . HUMAN ERROR FEEDWATER MOISTURE SEPARATOR-REHEATEP LEVEL SWITCH BUMPED. 09/12/84 HARDWARE STEAM AND FEEDWATER CONTROL SYSTEM i ELECTRICAL FAILURE, 2 01/15/85* S/G LEVEL FEEDWATER FLOW CONTROL VALVES WERE IN MANUAL AT ZERO POWER. j l 03/21/85 HARDWARE OSCILLATIONS IN FEEDWATER CAUSED BY MAIN FEED PUMP, i l 04/24/85 HARDWARE FLUX TO FLOW IMBALANCE SIGNAL, } l 06/02/85 HARDWARE TURBINE TRIP ON HIGH VIBRATION DURING i
- TESTING, i
4 i 06/09/85* HARDWARE LOSS OF MAIN FEED PUMP. 1 MARKEY/IE JUNE 20, 1985 i i
f i t 00ESTION 20D. (CONTINUED) ;
SUMMARY
THE OVERALL TRIP
SUMMARY
BY CAUSE FOR THOSE REACTOR TRIP 1 SINCE 1980 IS AS FOLLOWS. l HARDWARE.................. 20 j HUMAN ERR 0R............... 9 STEAM GENERATOR........... 2 ) PROCEDURE................. 2 j HARDWARE AND HUMAN ERROR... 3 0THER..................... 6 TOT A L...................... f4 0 A i r 4 r i i i i i l l MARKEY/IE JUNE 20, 1985 1 ^6 ~
QUESTION 20D. (CONTINUED). CAUSE DEFINITIONS 1 HARDWARE THE SCRAM WAS A DIRECT RESULT OF A SYSTEM, SUBSYSTEM, MAJOR SYSTEM COMPONENT, OR COMPONENT FAILURE OR FAULT. HUMAN ERROR THE SCRAM WAS THE DIRECT RESULT OF AN INCORRECT PERSONNEL ACTION. BOTH ANY COMBINATION OF HARDWARE AND HUMAN ERROR THAT (HARDWARE & CAUSE A REACTOR SCRAM, NOTE THAT THE ONLY REQUIPE-HUMAN ERROR) MENT IS THAT THE SCRAM WOULD NOT HAVE OCCURRED WITHOUT BOTH ACTIONS. PROCEDURE REACTOR SCRAMS THAT ARE A DIRECT RESULT OF PROCEDURAL DEFICIENCIES. S/G LEVEL REACTOR SCPAMS THAT OCCUR WHILE THE STEAM GENERATOR LEVEL CONTROL SYSTEM IS IN MANUAL CONTROL. UNKNOWN REACTOR SCRAMS FOR WHICH NO CAUSE CAN BE DEFINED, l.E., WHEN THE SYSTEM INITIATING THE SCRAM IS TESTED AFTER THE PLANT HAS BEEN STABILIZED NO FAILURE OR FAULT CAN BE FOUND. MARKEY/IE JUNE 20, 1985 O
QUESTION 20D, (CONTINUED) CAUSE DEFINITIONS NOT PROVIDED THE SCRAM CAUSE COULD NOT BE DETERMINED BASED ON INFORMATION PROVIDED IN THE LICENSEE'S LER, i 't MARKEY/IE JUNE 20, 1985 1 4
QUESTION 20. (E) THE NUMBER OF ALL ENFORCEMENT ACTIONS INCLUDING A BRIEF DESCRIPTION OF THE ISSUE, THE SEVERITY CLASSIFICATION AND ANY FINE LEVIED. ANSWER. TWO ESCALATED ENFORCEMENT ACTIONS HAVE BEEN TAKEN AND COPIES OF THE ACTIONS ARE ATTACHED, A COMPUTER PRINTOUT OF ALL CITATIONS, INCLUDING NON-ESCALATED ACTIONS, HAS BEEN PROVIDED. A. A NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF civil PENALTIES IN THE AMOUNT OF $90,000 WAS ISSUED ON NOVEMBER 21, 1984 AND INVOLVED SEVERAL SEVERITY LEVEL lli VIOLATIONS. BOTH TRAINS OF THE CONTROL ROOM EMERGENCY VENTILATION SYSTEM (CREVS) WERE REMOVED ON APRIL 23, 1984 THROUGH MAY 7, 1984 WITHOUT VERIFYING THE OPERABILITY OF REDUNDANT EQUIPMENT. THIS RENDERED THE CREVS INOPERABLE IN VIOLATION OF TECHNICAL SPECIFICATION REQUIREMENTS. SEVERAL EXAMPLES OF FAILURES TO FOLLOW PROCEDURES REGARDING THE OPERATION OF THE STARTUP FEED PUMP WERE ALSO IDENTIFIED. IN ADDITION, THREE EXAMPLES INVOLVING CHANGES IN THE FACILITY AS DESCRIBED IN THE SAFETY ANALYSIS REPORT WERE IDENTIFIED. NO WRITTEN SAFETY EVALUATION OF WHETHER THE CHANGE INVOLVED MARKEY/IE JUNE 21, 1985
i QUESTION 20E, (CONTINUED) 2-A CHANGE IN THE TECHNICAL SPECIFICATIONS OR AN UNREVIEWED { SAFETY QUESTION WAS PREPARED AS REQUIRED BY 10 CFR 50.59. THE civil PENALTY WAS PAID ON JANUARY 14, 1985, B. A NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTIES IN THE AMOUNT OF $13,000 WAS ISSUED ON JUNE 20, 1980 AND INVOLVED THREE ITEMS OF NONCOMPLIANCE 4 INCLUDING AN OVEREXPOSURE TO RADIATION DUE TO INADEQUATE CONTROL OF-ACCESS TO A HIGH RADIATION AREA NEAR THE REACTOR
- CAVITY, THE VIOLATIONS WERE NOT CATEGORIZED ACCORDING TO SEVERITY LEVEL BECAUSE THEY PREDATED THE EXISTING ENFORCEMENT POLICY.
THE CIVIL PENALTY WAS PAID ON ] JULY 10, 1980, i i i i 1 I i 4 l MARKEY/IE JUNE 21, 1985 y . ~, ..--r-. .,.-.--r-m e y,--w.---. o r,,__-___,. cr v.y _. m_.v_-~,y- .m, -, - _ _,,_ _..- -- y-w ,e ,.c., .~m,
NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTIES Toledo Edison Company Docket No. 50-346 Davis-Besse Nuclear Power Station License No. NPF-3 EA 84-95 An inspection conducted during the period June 11 through July 27, 1984 at the Davis-Besse Nuclear Station identified a number of violations of NRC requirements. These violations relate to the licensee's inability to recognize the importance of design basis and technical specification requirements for equipment operability, to ensure procedures which define requirements for equipment operability are followed, and to ensure that the requirements of 10 CFR 50.59 are satisfied. The violations involving equipment operability also relate to the licensee's failure to take effective corrective actions once problems have been identified. To emphasize the need for the licensee: (1) to recognize the importance of design basis and technical specification requirements for equipment operability and to ensure that these requirements are met when equipment is removed from service, (2) to ensure that procedures which define the requirements for equipment operability are followed, (3) to ensure that appropriate reviews are conducted in accordance with the requirements of 10 CFR 50.59, and (4) to ensure that adequate corrective actions are taken to preclude repetition of identified problems, I propose to impose civil penalties in the cumulative amount of $90,000. The base civil penalty for Item I is $50,000. The base civil penalty amount for Item II is $40,000 because two of the violations occurred prior to the revisions to the recent Enforcement Policy. In accordance with the General Policy and Procedure for Enforcement Actions, 10 CFR Part 2, Appendix C, and the Policy as revised, 49 FR 8583 (March 8, 1984) and pursuant to Section 234 of the Atomic Energy Act of 1954,'as amended, ~ 42 U.S.C. 2282, PL 96-295, and 10 CFR 2.205, the particular violations and the associated civil penalties are set forth below. I.A. Technical Specification 3.7.6.1, " Control Room Emergency Ventilation System," requires that two independent control room emergency ventilation systems shall be operable. A system is considered operable when it is capable of performing its specified function (s). Technical Specification 6.8.1.a requires that written procedures be established, implemented and maintained covering the activities specified in Appendix A of Regulatory Guide 1.33, November 1972. Appendix A specifies typical safety-related activities that should be covered by written procedures. This includes procedures for operation of the control room emergency ventilation systems. Administrative Procedure (AD) 1839.00, " Station Operations," requires that, prior to removal of safety-related equipment from service, operability of redundant safety-related equipment must be verified by inspection. In addition, this procedure requires that the 84iT270344 841121 PDR ADOCK 05000346 G PDR
Notice of Violation 2-NOV 211984 applicable technical specification action statements be evaluated prior to the removal of the safety-related equipment from service. Contrary to the above, both trains of the Control Room Energency Ventilation System were recoved from service on April 23, 1984 through May 7,1984 without verifying the operability of the redundant eouiprent or evaluating applicable technical specification action statements. This rendered the Control Room Energency Ventilation System inoperable in violation of technical specification requirements. B. Technical Specification 6.8.1 requires that written procedures be be established, implemented and maintained covering the activities specified in Appendix A of Regulatory Guide 1.33, November 1972. The activities specified in Appendix A, Section A, Administrative Procedures, include procedure adherence, shift and relief turnovers and log entries. Appendix A, Section C, " Procedures for Startup, Operation, and Shutdown of Safety Related PWR Systems," list the feed-water system as requiring instructions for energizing startup and shutdown of the system. Contrary to the above, on June 24, 1984, the licensee failed to start the startup feed pump in accordarce with the applicable sections of the approved procedures (SP 1105.27 and SP 1106.?7) for operation of the startup feed pump; failed to log the starting of the startup feed pump in the reactor operator's log; and improperly initialed the trip recovery procedure (PP 1102.03) indicating that the startup feed pump was started per an approved procedure (SP 1106.27). In addition, on June 25, 1984, the licensee failed to shutdown or restore the startup feed pump to normal in accordance with the applicable sections of the approved procedures (SP 1105.27 and SP 1106.27); improperly initialed the plant startup procedure (PP 1102.2) indicating the startup feed pump was stopped per the approved procedure (SP 1106.27); failed to perform an adecuate turnover regarding the status of the startup feed pump system; and failed to properly sign off the completien of Section 8 i of the startup procedure (PP 1102.2). Collectively, these two violations have been evaluated as a Severity Level III problem. (Supplement I) (Cumulative Civil Penalties $50,000 assessed ecually among the violations.) II. 10 CFR 50.59(a)(1) states that the license may make changes in the facility as described in the safety analysis report...without prior Comissien approval provided that the proposed change...does not involve a change in the technical specifications incorporated in the license or an unreviewed safety question. t i f ( 5 --,._,-n-n ~ ,,---..,_------,---r-.
I Notice of Violation NOV 211534 10 CFR 50.59 requires that the licensee maintain records of changes in the facility to the extent that such changes constitute changes to the facility as described in the safety analysis report. These records shall include a written safety evaluation which provides the bases for the determination that the change does not involve 7 an unreviewed safety question. l Contrary to the above, in the following instances, the licensee made changes in the facility as described in the safety analysis report without preparing a written safety evaluation of whether the change involved a change in the technical specifications or an unreviewed safety question. i (1) On November 1, 1983, the licensee removed one of two Emergency Diesel Generator (EDG) ventilation supply fans from service without j preparing a written safety evaluation and without realizing this action represented a change in the facility as described in the Updated Safety Analysis Report (USAR). The USAR' describes the EDG ventilation supply as containing two 50% capacity fans. (2) On December 19, 1982, the licensee initiated a Facility Change 1 Request (FCR) that was implemented on May 24, 1983 which changed the position (to open) of the Startup Feedwater Pump (SUFP) suction valve ' during power operation without preparing a written safety evaluation. The USAR describes the valve as closed during power operation. (3) On March 8, 1984 and May 4, 1984 lead shielding was hung on decay heat system piping changing the loading of the safety system as described in the FSAR and without preparing a written safety evaluation. Collectively, the above violations have been evaluated as a Severity Level III problem (Supplement I). (Cumulative Civil Penalties - $40,000 assessed equally among the violations.). Pursuant to the provisions of 10 CFR 2.201, Toledo Edison Company is hereby I required to submit to the Deputy Director, Office of Inspection and Enforcement, U. S. Nuclear Regulatory Commission, Washington, DC 20555 and a copy to the Regicnal Administrator, U.S. Nuclear Regulatory Commission, Region III, i l 799 Roosevelt Road, Glen Ellyn, IL 60137, within 30 days of the date of this Notice a written statement or explanation, including for each alleged violation; (1) admission or denial of the alleged violation; (2) the reasons for the violation, if admitted; (3) the corrective steps which have been taken and the i results achieved; (4) the corrective steps which will be taken to avoid further violations; and (5) the date when full compliance will be achieved. Considera-tion may be given to extending the response time for good cause shown. Under j the authority of Section 182 of the Act, 42 U.S.C. 2232, this response shall be submitted under oath or affirmation. ,... -. _. - _ - -.,,, -.,,.,.... ~, -
r Notice of Violation - 4.. NOV 211?34 Within the same time as provided for the response required above under 10 CFR 2.201, Toledo Edison Company may pay the civil penalties in the amount of $90,000 or may protest imposition of the civil penalties in whole or in part, by a written answer. Should Toledo Edison Company fail to answer within the time specified, the Deputy Director, Office of Inspection and Enforcement will issue an order imposing the civil penalties proposed above. Should Toledo Edison Company elect to file an answer in accordance with 10 CFR 2.205 protesting the civil penalties, such answer may: (1) deny the violation listed in the Notice, in whole or in part; (2) demonstrate extenuating circumstances; (3) show error in this Notice; or (4) show other reasons why the penalties should not be imposed. In addition to protesting the civil penalties, in whole or in part, such answer may request remission or mitigation of the penalty. In requesting mitigation of the proposed penalties, the five factors contained in Section V(b) of 10 CFR Part 2, Appendix C, as revised 49 FR 8583 (March 8, 1984) should be addressed. Any written answer in accordance with 10 CFR 2.205 should be set forth separately from the statement or explanation in reply pursuant to 10 CFR 2.201, but may incorporate statements or explanations by specific reference (e.g., citing page and paragraph numbers) to avoid repetition. Toledo Edison Company's attention is directed to the other provisions of 10 CFR 2.205, regarding the procedures for imposing a civil penalty. Upon failure to pay any civil penalty due, which has been subsequently deter-mined in accordance with the applicable provisions of 10 CFR 2.205, this matter may be referred to the Attorney General, and the penalty unless com-promised, remitted, or mitigated, may be collected by civil action pursuant to Section 234c of the Act, 42 U.S.C. 2282. FOR THE NUCLEAR REGULATORY COMMISSION / .hMr ~
- jDames G. Keppler Regional Administrator Dated at, Glen Ellyn, Illinois this at* day of November 1984 i
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ri LJ - 0 ALL IHOPOSEU CHAf4015 OH MODlilCA110NS 10 PL AIJ ie 5 Ybit MS OH [0UIPMLNI IHAl AFFECI.NUCLEAM SAfEIY AND () - HECOMMtND WRIIT[N APPHOVAL OR DISAPPROVAL OF CHAN GES OR MODIFICATIONS 10 THE STATI0tJ SUPER lhif t:0[tJI . CONTRARY TO THE AHOVE, TEMPURARY MODiflCATIONS lD ASSOCIAl[D WlIH NONCONFORMANCE REPOhTS IJCR-232-M1, - (LIMITORGUE VALVE MODIF IC A T IONI, NCR 342-81 (SLkV - ICE WATER VALVL MODIFICATIONS, AND NCR M3-bl (AUXI ($ - LIAav FEEDWAl[R PUMP ST[AM LINE MODIF I C A l l utJ ) WLPC - NOT RLVIEWED HY THE STATION REVIEW U0460 AND A HE - C OMME ND A T I ON CONCENNING THL MODIF IC AIION'S ACCIPIA gg DILITY WAS NOT MADE TO THE STATION SUPE R IN i t hDi f41. 5 1 10 CFR S0, APPENDIX M, CRITfkION XVil, A0 IMPLLNEN gg - IED BY THL 10LEDO CDISON OPERAllNG QA PR6GkAM AND fliC FSAR SECTION li.2, HEQUIRI IHAI THE APILICANI - SHALL PROVIDE RECORD STORACT CONSISifNI WIIH APPLI gg - CAULL RLGULATORY Rf 4UIRE NL f415. TH( T0llie0 LDISON - QA l*ROGRAM COMMITS IO ANSI f14 5. 2. 9 - 19 7 4 AND NEGULA IORY Gulb[ 1.8d, REVISION 2, OC100LR 19 f f., WITH A rd gg LXC[P110N S PE C I F Y I NG A TWO HOUN F IHE PROilCTION R - ATING FOE RECORD SIGRAGE FACILITIFS. CONThANY TO INE A00VL, htCUPUS OF AUDIIS, AUDITCH ANb LC INSPE gg - CTON QUALIFICATION /CLRTIFICATION AND CAlthwATIONS - WLR( rdO I P R O V i t'[ D ThL RfGUIRED PkOILCTION. 5 1 II - 10 CFR S0, APP [hDIX b, CRITLRION II STALLS IN PART - o *INE PH06 HAM SHALL PROVIDE F OR I N00C ih l N A T I Ord AN D T H A l tJ I NG 3F PERSONNEL FEPFORPING ACTIVITIES AFF[ gg - CIltJG uuAL II Y AS 74ECESSARY 10 ASSUHL THAT LUITAbLE - PROFICI[tCY 15 ACHIEVED AND MAlhTAINED." Tite DAV IS-utSSE UPDAft0 SAIETY ANALYSIS REPORT (USARI 1 74 gg - SECTION 13.2.1 $1 AILS THAT TH[ ikAINIldG PRubkAH IS - OESCRII2ED AND admit.ISTERLD BY THE AD 1828 SF LIE S - OF PROCfDUH[S. SL C I I Ote 13.2.2.2 Of IHE USAR I fiC LU gg - DLS CONMIlftLNIS THAT CHEMISihY Ar.D ttEALTH IHYSICS - (C&HPS PfRSONNLL AHL PHOPfHLY 1 RAINED a f;U t'AINTAIN - PHOFICIENCY lia THLIR H[bulRED JOH SmILLS I Hk OU ble gg CONilNUED I M A l fa l fdG. SECTION 13.2.2.3 0F THE USAR - INCLUDLS C O P M l l filtJ I S THAT MAINTENANC[ FL R SONNL L AR - [ Ph0PLHLY T H A I N( D IO PERFGHM IHfth JOHL A l.D IHAT gg I HL Y R L M A l ti PHOFILILNi IN THL R L O U II:[ U J0tt SN ILL S. ADMIf4ISTF AT [v[ PHOC[ DUNE AD 1826.00 18[RSONNEL T R Alfal4b P h "J b R A M I NIQUIHLS INIIIAL THAINihG A fv0 0 0f4 gg - I l hu l tdG I F A I N I t.G FOR CAHP PLH59hhf L PLR AD IM2r.82 - Abu FUN ALL MAINTENANCE PtRS0hNfL P[H At AhPA.al. CONTHAkV 10 THL Ah0VL, litt MASILH IR A INIt.G StHLU II - ULL F 09 19P4 DID NOI I DL f4T I F Y THAT ANY IN111AL Ok - C ute l l f4 U I N G I H A l t4 I NI, HAD ULEN SCH[bOLED IN 19H4 t OR - CAMP F E k s chN[ L OR fok ELECINICAL M A l tJ I t f4 A ?.C L P! H 5 gg - ONNi t. 5 1 - 14 CF9 SO, APPENUlX H. CRI1(HION NIX, A% IPPLI PL foi gg - [U tY T Ht IPL[D0 EUISON OPERAilONAL QUAllif ASSUHA - NCE PRU6 HAM HEQUIHEb IHAT f:C A SON [ $ li[ L ST At L ISHED - TO ASSHR[ THAT TOOL %, GAUCES IN?lPUMENTS AND OTH[R (g - eL Atuk ItJG Ahu ILSTING DEVICES Utfo IN ACTIvlIILS - dFLLildb NHAllff Ahl PN0fthLY CONIAOLLtD, C A L IliH A - 100 AND AbJU$lt u A l SPL C IF ILO PLHI4US IO MAINTAIN () - ACCUHACY h!IHIN NECESSANY LIMIis. C ore l P Ak Y 10 IHC
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LJ - 14 U IN At t.hb A.\\L C W i lee OHAWING 5 [ r" I t s l'-/ t. IP 19 - O liC Ilois Fit 0M ACTUAL!ON Of IHL flkt PHbiltil6fthlN - It:KLER SYSTEMe CONIRARY 10 THE All0VE e NUMLR9US EL i, ~' EC TRICAL JUNCil0N HORES WERE NOT MAINIAINTH IN ACC OkD Af4CE WITH DRAWING H-269AS IN THAI T HE SL 5 A F t' I Y - RELATED JUNCTION H0XES WERE NOT PReiECTED FHOM IHL () - t!NE Ph0TLCTION SPHINHLER SYSl[M. 5 I 10 CFR fl.214Al STAliSe IN PART: "[ACH I Nill V IOil A l e CUNPOR Ai!Ots e PANTNLHSHIPe OR LNIIIY SULJtLi IO TH [ RE6ULATIONS IN THIS DAHi SHALL ADGPT Al Pn 0 Pit l Ai[ PHOCEDUNES TO: (1) PROVIDE FON: (Al E V A L U A il r40 D gg - E V I A l lOrJS Ok... I (2) ASSU1E THAT A DIRECION UR Nf5 P ON S i ti Li 0FFIC[H 15 INf0RMID II THE CONSIRDCTION O - R OPERATION OF A F ACILIIYe Oh ACilVIIVe OH DA5IC C gg OMPONENT SUPPLILO FOR SUCH F ACIL ITY OR HA51C ACilV - IIY: t!) FAILS TO COMPLY...e ORI (Ill CONIAlfJS A - OffECT.= Cir.I R AN Y TO IHE A tiO V f HE QU IHi t'E N T S e THE NI - foL(OWING UttICILNCILE WERI IDI NTIF If D IN THf PNor - EDU>ES A D 0f' T E D t; Y IoLEDO EDISON COMPANY 8110C0) PU - NSUANI ID in CFH 21 28tal; (Al NO ASSur#AN(C 15 PR - OVIDED THAT PQl[N!!AL in CfH ?! NfPORTAplE I llit S [ - DtNilFItD tty l0LLOG [DISO(J COMPANY PEN 56YJ8 L WouLD bL FORWAPutp 10 NUrttAP FACILITY L NG ] tet L R I NG ENFE I 3 FGN E V A L U A T i bil e A ', REQUIN[O hY THf T[DCO GA MANU - All (H) I NS T R UC T I OPJS FROVIDED hY IEDC0 PHOLEDUNL Q - AI elbo, GA REVIEW OF NONCONFONMANCf NLP0HIS, HE SU 4 - L il 0 IN HYPASSING THE Pk0 GRAMMATICALLY Nf0DIPFD NF - [ EVALUAi!Of. OF POTI NII AL ht POMI At!LL DLVIA110N!I ( Cl NEITH[k bAl 41$0 NOR NFE PROCEDud[ ifE-00To PRO OI - CLSSING 0F NCHL, SlJH S e AND SD0hs, PROVIDED F 3R PRE - PARATIOfe A l.D H A I N T E t A *JC C PF HECONDS OF IHC RESULIS 4 - UF EV ALUA TIONS PERF ORMI D PUN 50ANI 10 10 CIR 21. OI - I f4 Abuli!Oise IHL5[ PROCEDU'4[ S PhDVlp[O 310 a.L T A IL L D - CHITERIA UF 0tJ WHICH 10 ttA$f THE EVALUATIONI 406 I - HL I f.L i h UC I l 0 N S PROVIDfD FUR P!G T IF I C A T I ON OF A PES $h - PO4:,II LE CCriP ANY CF F I Cf:k AltEAHED IJ LLAVE TH[ Uri - [ R M I N A i ! O f4 GF RfPORIAH]L11Y TO THAT Rf 5POPJSIlitt 00 - MsANY ofFICIN. T Hl. INSIHUCTIONS PHAVIDIu tuR 740l! gg - F IC A I IO'J OF A HE 5P0fAS!HLE COMPANY OFFICLk 010 f40 T - PhotIDE ASSURAtsCE THAT IHL RLGu!Rrp N011tICAIIGN W - A5 f4 A DE AND 010 IJ01 PkUVIDE F OR THE PREPAHATIL4 AH - 0 HalNTLt4ANCL Of MLCOPDS TO ASSURF COMF L I Al.Cf WIIH THL P49 VISIONS OF 10 CFR 21.?ltAltfli If I h4 uulu - MLNIS RLVIL'kED li Y THE I N L P t. C I GH PH OV IOl la A *.5UE A NC E O - Thai NOT IF IC AIIONS M ADL l ts 1Hf C ')M K I L L i r.f. PUN LU A ra - T I6 10 CFA 21.fitAlt23 WUULD MfE1 Titt hEdulhf6 TNT - 5 0F 10 CFM 21.714b)(2) Aflu (331 A f2t' ONL OTHER tNA lI - e-LE. + 0500034684013 06/25/1984 06/29/L9h4 i S 3 0 - 169 NRL INSPECTORS WE RE NOT POSITIvfLY IDE..I IF IE D - Niellf klLULSilh6 ACCLSS iP TH( PEnit CILP AHLA.
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NL I HI A l A% Iul' t ( L D l'UMt WAS S10PPID Nt. IHI Al'1 h u o VLli PMDCLbuME ($P 1806.2738 FAILLO 'O f*t Nt sh4 Al: 4 - DEQU4ll IDH t40VL H RLGARDING THL S T A ll.S Of inil SIAH1 () UP ft[D PUMP SYST[H& At40 FAILED 10 PROPENLy SIbla 0 - FF INE COHlLET'ON Of SECTION 8 0F THE SIARIUP PROC - EDUNE IPP 1102.23. GD e 3 1 - 10 CF R 50.b9tAlitt h i A l[ $ IHAT IN[ LICENSEf NAT HA - Mt CHANGt5 I fd THE F ACILI T Y A5 GTSCHIHfD IN THf SAF g) - [i Y ANALVSIS HLi' ORT...WITH0ui PRIOR CONFI% LION APP - 00 VAL PROVIDED THAI THE PMOPOSLD CHANGF.. 00E5 N0f INVOLVE A CH A r.GE 1% THE TECHNICAL SP[ C IF I L A I I Of;5 gg - INC0RPORAi[D IN THE LICENSE ON AN UNR[VitWLD S All I - Y QULSTION. 10 rF R $0.L9 REGUlH[S IHAI THL L ] L(: fJ S - EL PAINIAIN RLLONDS OF Cif A NGL S Itd THL #ACI.Ilv TO gg - THE EMifhi IHAT SUCH CHANGLS CONSTIluit CHANGES 10 - THE FACILIIT AS DESCRIDED !N THE SAFETY A.. A t y s i s - RfP0ki. THJ5[ HECOHOS SHALL INCLUUt A WHillfN SAF gg - tiY EVALUATION kHICit PMOV ICE S IHI II A Sf 5 F 06. THI DE - I L R M I N A i l 0tJ IHAI IHL CHANGL bots NOT INV0tVL Are UN - REVIEWED S Af f t v uuE N T I0tJ. CONTRAhY TO IHL ALOV[e gg - liJ Itti FOLLJWING I4SIANCLS. THE LICft4SEE MADE CHAN V GLS IN THE FACILITY AS DE SCh if'[D I f. IHf SAF E IT A f1 A LYSIS R[ PORT WIIHotti PRfPARING A WRifit N SAFliY LV gg ALUATION Of WNilHEN THf CH ArdGE INV0LVLD A (HAN0L I N TleE TLCHNICAL SPECIFICATIONS ON A fd UNRIVIEWED SA F L i v QUE S T I OfJ. (I) ON NOVt PU[R 1, 1983, THE LIC[N gg SEE HEMuVED ONE OF IWo EMfhGENCY DIESEL GE f.[ h A T Ok - 800G3 VENTILATION SUPPLY FANS FROM SfRVICE WITHOUT OP E P AN it#G A WFITiffa SAFEiv LVALUATION AND WITHOUT gg H E A L I/ I tog THIS ACIION REFRESEfAIED A CHANG; Ild THE - tACILIlY As OL SCRIllCD IN THE UPDAIED SAf[ly ANALY - SIS REPORT (USAP). INE USAP OL SC H illE S THE EDG VLN gg - i t L A T IOPJ StfPI L T AS CONTAINIf4G IWO SOE CAPALITY f A tJ e S 1 - 10 CFH 50.59tApt!) ST Alf S Illa T IHL LICfNSET MAY MA - nt CHANG [$ Ile IHL FACILIIV AS 0[$CRlbfD les illE 5AF - CIT ANALVSl$ N100MT...WITHOUT PRIOR C0PMISS10N APP gg COWAL PROVIDED THAT T HE Ph 9f'05t h CH A f40t...tsOf S t;0 T - INVOLvl A CHafJLE Ils IHL TE CHlil C AL SPECIFILA110NS - lieC OH PJ H A ll b IN IHL LICEN5L ON A tJ Ut;hlV lt Wt b S At t i gg - Y OtaES110N. 10 CFh 50.'.9 NiculbES IHAT IHo L IC[t.S - LL MAlf41A]N PLLOFDS Of CHAP.GL$ Ita Ilif fACILIIY 10 - i t.f L X ii hi IHAT SUCH C H A ttG t '. CLh%IllllTI CHANGL' TO gg IHL FACILITY AS O( 5CR ittE D IN Itti S Af t if A s. A L T r i L - Na P sh i. THf50 14 E C OF D S SHALL I t.Cl ubt A b h l I I L 's SM - IIT E V At te A T ION WHICH PHOVIDES THt i a5f t i u. IH! tit - I L M M IN A i t et: Itt A T THt CHANGL D0i5 t40 f INVOLbl AH 68 5 - Rt v iE WI D mal [TY QUL ?.T ION. CONINAMV 10 Thf AHOVt e - IN IHf 70LLOWING I te S T A NC E S. THL LICI.NSE E MICL C H A F' - G '. S IN THf fACILItt AS DESCRlett. Ife IFE $Af[IY ADA - Lysis 6 front WITHoul Pk t P A H i rJG A VHillita SAftTY fV - A L U A I I Oil Gt kHtIHLR THL C H AtJG1 IP;VOLVf D A L it A h 44 I - ts THE TECHNICAL U L C ll l C A T I Old". Oh A fd uteh1 V ; E Wi lf
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5 1 - IECHNICAL SPEC IF IC A T IONS 4.9123 AND 4 6 5.1 STAT - E IN PART: " T Hf [PURG[NCY VENIILATION SYS1(M SERV - 10!NG ThL $10kAGE Pu0L ANEA SHALL ttC DEPONiihAILD - OPESAHLE Al LEAST ONCE PER 51 DAYS WHENEVEd IMHADI - AIED Fu[L IS IN THL SIUHAGL P 0 0 t.. " CONihAHY TC IH - [ AliOV[ e Os4 NOVEMBEH 20. 19ete DUNING CGhE ALT [ HAT I Gt!S e THL EMLHGENCY VLNTILATIQN SYSTEM SLRVILING 1 - HL STORAbt IOOL AREA HAD NOT PLEN DIM 0NSTR ATLD DPt 8 - R4blE SINCE AUGUST 19 1984. THE SL COND [ tier 6f NCY - VENTILAIION SYSTEM WAS NUI OPI H Al6LI t'LCAUSL OF MA INTINANCE. THIS CONDITION WAS Ih0MPILY CutNLCIID - BY IHL LICL NSt t F OL L OW !!1G IliE NIlt IC A T IOf; D V THt NR C INSPLC10h. 4 3 0 - TECHNICAL SPE CIF IC A T ION 3.9.2 ST A T[S IN PAhl: "LA CH SOURCL RANGE NEUIR0f4 FLUM P 0f4110h SHALL 80 OfMO - NSTRAl[D CILRADLL WITHIN LILHI tel h00R$ PFION 10 b THE INITIAL START OF C OR [ OP[ RATIONS". CONihAFT I - 0 INE Ah0VL e Or4 NOVENGER 20, 19H4, THE LICLNS[L WA S CON 0uCIIf4G CORE Ali[ RATIONS AND THE AUDILLI INDI CAil0N PORTION OF THE SOURCI R AP;Gf NEUTHON FLUX MO - N110MS WAS UISCOVLNLD TO UL JNOPLRAt*LL. SURVLILLA NCE TEST $091.01. NFWISION 9e *SOHNCE RANGf F UtiC T ! Of4 A L T[ Sis THAT WAS PERFORMED TO OEM0feSikATE SOU a NCE RANGE NEUIRON FLUX M0t4ITOR OPERAHILITY WAS GLI [RMINLD 10 til INAD[QUATF. THIS CONDITION WAS PROM - PILY ConhtCTED DY THL L ICL f4SE L FOLLOWING Iu[Nili !C - AIION HY THE NHC INSPLCTOR. 4 1 - 10 CFH 50, APPEN0!X be CRITEkl0h XII, SIAltS IN PA - RT: "Mf A SUhf 5 SHALL DE ESTAULISHfD TO AS$uRE THAT gg I N =. I k U FI N T S USt h IN ACTIVITIES AFitCIIfib UUALITY P
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.E WASHINGTON, D. C. 2033s e% i %, '%.. "o* JUN 2 0 ISc0 Docket No. 50-346 EA-80-37 Toledo Edison Company ATTN: Mr. Richard P. Crouse Vice President Nuclear Edison Plaza 300 Madison Avenue Toledo, OH 43652 Gentlemen: The findings of a recent inspection of the radiation protection program at the Davis-Besse Nuclear Power Station, particularly with regard to the overexposure to radiation which occufred near the reacto. cavity on April 30, 1980, indi-cate that sufficient attention has not been given to the control of access to high radiation areas. The inspection findings were discussed with members of your staff at the conclusion of the inspection. On June 4, 1980, the Director of our Region III Office met with you to discuss the circumstances surrounding the April 30, 1980, overexposure and to discuss the three apparent items of noncompliance identified during the inspection. These noncompliances are set . forth in the Notice of Violation attached as Appendix A to this letter. In our view, the items of noncompliance in Appendix A demonstrate a lack of effective radiation exposure control. The potential for a significant per-sonal exposure in reactor cavities was described in IE Circular No. 76-03, " Radiation Exposures in Reactor Cavities," dated September 10, 1976. You were sent a copy of this circular. In addition, a week before the occurrence, an NRC inspector had discussed the potential hazards in the reactor cavity and cautioned your radiation protection management personnel regarding high radia-tion areas generated by incore instruments removed from the reactor core. On May 6,1980, NRC inspectors requested an oral response to IE Circular No. 76-03 at the May 8, 1980 exit meeting. Your written response involving changes to increase Chemistry and Health Physics management control, increase training, and require the use of two different types of high range survey meters for entry into locked high radiation areas is acceptable. Please ensure, however, that these changes and any other changes necessary to control exposure in the reactor cavity and other hazardous areas are promptly and fully implemented. We consider the April 30, 1980, overexposure to be very serious not only because the actual dose of 4.76 rems exceeded the regulatory limit, but also because of the potential for an extremely large radiation exposure. We are particularly concerned that to some extent this overexposure resulted from CERTIFIE0 MAIL RETURN RECEIPT RE00ESTED i n n i i) om ,,a , i p j 1-t f " a v~
Toledo Edison. Company your failure to act upon a high priority facility change request written in June 1978 to install permanent barricades at reactor cavity acceM. points and the April,1980 request to relocate the temporary barrier to the proper loca-tion. Consequently, we propose to impose civil penalties in the cu tulative amount of Thirteen Thousand Dollars (513,000) for these noncompliances. Appendix B of this letter is the Notice of Proposed Imposition of Civil Penalties. You are required to respond to this letter, and in preparing your I response you should follow the instruction in Appendix A. In addition to inadequate preparation and planning, weakness in communication between the senior chem and rad tester and his supervisor, preoccupation with j some other assigned radiation protection responsibilities, and a sense of urgency to get the job completed before a plGnned electrical outage all appear 3 to have contributed to the incident. In responding to the noncompliance items in Appendix A, you should specifically address your plans for strengthening your controls related to the preparation for and management of work in high radiation areas. Your written reply to this letter and Notice of Violation and the findings of our continuing inspections of your activities will be considered in determining whether further enforcement action, such as additional civil penalties or orders to suspend, modify, or revoke the license, may be required to assure i future compliance. In accordance with Section 2.790 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations, a copy of this letter and the enclosures will be placed in the NRC's Public Document Room. Sincerely, i 1 ,n Vi tor Stello, J., Director Office of Inspection and Enforcement
Enclosures:
1. Appendix A, Notice of Violation l 2. Appendix B, Notice of Proposed Imposition i of Civil Penalties 4 cc w/ enc 1: (See next page) 1 - - --~ ' "*"^ ~~ ~ ~* ~ ~ "~~
4E Aopendix A NOTICE OF VIOLATION This refers to the inspection conducted by representatives of the Region III (Chicago) Office at the Davis-Besse Nuclear Power Station Oak Harbor, Ohio, of activities authorized by NRC License No. NPF-3. During this inspection conducted on April 21-24, and May 2 and 6-8, 1980, the following apparent items of noncompliance were identified. Item 2 is a violation; Items 1 and 3 are infractions. 1. 10 CFR 20.101(b) requires that during any calendar quarter, the dose to the whole body of any individual working in a restricted area not exceed 3 rems. Contrary to this requirement, during the second calendar quarter of 1980 a senior chem and rad tester received a dose exceeding 3 rems when on April 30, 1980, he entered the normal sump tunnel leading to the cavity beneath the reactor vessel while the incores were in the withdrawn posi-tion beneath the vessel. His TLD indicated a dose of 4.70 rems. A TLD worn earlier in the calendar quarter indicated a dose of 0.055 rems. This is an infraction. (Civil Penalty-54000) 2. 10 CFR 20.201(b) requires that surveys be made as may be necessany to comply with 10 CFR 20 regulations. One of these regulations, 20.101(b), sets dose limits for individuals in a restricted area. Contrary to the above, exposure rate evaluations made by the senior chem and rad tester during the April 30, 1980, entry to the normal sump tunnel leading to the cavity beneath the reactor vessel were not adequate to ensure that dose limits of 10 CFR 20.101(b) would not be exceeded. This violation resulted in an overexposure to radiation and had the potential for causing a substantial radiation overexposure. (Civil Penalty-55000) 3. Technical Specification 6.8.1.a requires implementation of procedures contained in Appendix A of Regulatory Guide 1.33, November, 1972, which includes radiation protection procedures. Procedure HP 1601.05.1, Section 6.3.4, titled " Job Planning and Preparation" states, "A major portion of the occupational radiation dose is received during maintenance inspection, refueling, and non-routine operations. The following actions should be carried out if applicable:... 7. Minimize personnel radiation exposures by planning for access to and exit from work areas...." S 3{' Q& " ! l y @O f ~
Appendix A. Contrary to the above, regarding the entry to the reactor cavity en April 30, 1980, proper planning for access to and exit from the werk area to minimize personal radiation exposures was not evident, in that neither the assistant shift supervisor nor the senior chem and rad tester reviewed the previously conducted limited survey of the area which was recorded and on file nor were either aware of the construction of a temporary door in the wrong location in the normal sump tunnel, altnough the construc-tion error had been noted in the Health Physics Log. This is an infraction. (Civil Penalty-54000) This notice of violation is sent to Toledo Edison Ccmpany pursuant to the provisions of Section 2.201 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations. You are hereby required to submit to this office, within twenty-five days of the date of this notice, a written statement or explanation in reply, including for each item of noncompliance: (1) admis-sien or denial of the alleged items of noncompliance; (2) the reasons for the items of noncompliance, if admitted; (3) the corrective steps which have been taken by you and the results achieved; (4) corrective steps which will be taken to avoid further noncompliance; and (5) the date when full compliance will be achieved. i e
i + i 3 Accendix 8 NOTICE OF PROPOSED IMPOSTION OF CIVIL PENALTIES ? l Toledo Edison Company Docket No. 50-346 r This Office has considered the enforcement options available to the NRC, including administrative actions in the form of written notices of violation, i civil monetary penalties, and orders pertaining to the modification, suspension, 1 or revocation of a license. Based on these considerations we propose to impose civil penalties pursuant to Section 234 of the Atomic Energy Act of 1954, as amenced (42 USC 2282), and to 10 CFR 2.205, in the cumulative amount of Thirteen Thousand Dollars ($13,000) for the specific items of noncompliance set forth in Appendix A to the cover letter. In proposing to impose civil 1 penalties pursuant to this section of the Act and in fixing the proposed i amount of the penalties, the factors identified in the statements of considera- ) tion published in the Federal Register with the rule making action which j adopted 10 CFR 2.205 (36 FR 16894) August 26, 1971 and the " Criteria for j Determining Enforcement Action," which was sent to NRC licensees on December 31, 1974, have been taken into account. 4 Toledo Edison Company may, within twenty-five days of the date of this notice, i pay the total civil penalties in the cumulative amount of Thirteen Thousand Dollars (513,000) or may protest the imposition of the civil penalties in wh' le or in part by a written answer. Should Toledo Edison Company fail to o I answer within the time specified, this office will issue an order imposing the i civil penalties in the amount proposed above. Should Tol'edo Edison Company elect to file an answer protesting the civil penalties, such answer may (a) i deny the items of noncompliance listed in the Notice of Violation in whole or i in part, (b) demonstrate extenuating circumstances, (c) show error in the i Notice of Violation, or (d) show other reasons why the penalties should not be imposed. In addition to protesting the civil penalties in whole or in part, such answer may request remission or mitigation of the penalties. Any ritten ) answer in accordance with 10 CFR 2.205 should be set forth separately from. ) your statement or explanation in reply pursuant to 10 CFR 2.201, but you may incorporate by specific reference (e.g., giving page and paragraph numbers) to j avoid repetition, i l Toledo Edison Company's attention is directed to the other provisions of 10 CFR 2.205 regarding, in particular: failure to answer and ensuing orders; l answer, consideration by this office, and orders; request for hearings, 4 hearings, and ensuing orders; compromise; and collection. i Upon failure to pay any civil penalty due which has been subsequently determined in accordance with the applicable provisions of 10 CFR 2.205, the ) matter may be referred to the Attorney General, and the penalty, unless compro-mised, remitted, or mitigated, may be collected by civil action pursuant to Section 234c of the Atomic Energy Act of 1954, as amended, (42 USC 2282). t I k
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NOV 2 i 1c84 Docket No. 50-346 EA 84-95 Toledo Edison Company ATTN: Mr. John P. Williamson Chairman and Chief Executive Officer Edison Plaza 300 Madison Avenue Toledo, OH 43652 Gentlemen: This refers to the safety inspection conducted by Messrs. W. G. Rogers and D. C. Kosloff of the Region III staff during the period June 11 through July 27, 1984 of activities at the Davis-8 esse' Nuclear Power Station authorized by Operating License No. NPF-3. The results of the inspection were discussed on July 13, 1984 during an Enforcement Conference held in the Region III office between Mr. R. P. Crouse and others of your staff and Mr. C. E. Norelius and other members of the NRC staff and on October 2,1984 during a meeting between Mr. W. A. Johnson and others of your staff and Messrs. R. C. DeYoung and J. G. Keppler of the NRC. The following violations were identified during the inspection. On May 7, 1984, both Control Room Emergency Ventilation System (EVS) chiller control switches were discovered in the "off" position. This rendered both Control Room EVS trains inoperable. Your program failed to recognize the technical specification requirements for the operability of the equipment and your program failed to ensure that procedures were followed to verify the operability of the equipment. On November 1,1983, one of the two ventilation fans for the Number One Emergency Diesel Generator was removed from service. You failed to recognize that removal of this ventilation fan from service represented a change in the facility as described in the Updated Safety Analysis Report (USAR). This change affected the design basis requirements for equipment operability. In addition, the required review in accordance with 10 CFR 50.59 was not conducted. On December 19, 1982, you initiated a Facility Change Request that was implemented on May 24, 1983 that changed the position of the suction valve to the startup feed pump to the open position instead of closed as required by the design basis analysis for flood protection. On May 14, 1984, you determined one auxiliary feedwater pump was inoperable as this valve was open contrary to USAR requirements. You immediately closed the suction valve and modified procedures to control the opening and closing of this valve. During recovery activities following a unit trip on June 25, 1984, the suction valve was routinely used for unit startup. On July 1, 1984, you again discovered the suction valve was open rather than CERTIFIED MAIL SAk%g344Bhg 46 RETURN RECEIPT REQUESTED DOCg OS ppR PDR gla l G. t
o Toledo Edison Company 3 ACV :! ; G34 You are required to respond to the enclosed Notice and you should follow the instructions specified therein when preparing your response. Your response should specifically address the corrective actions you will take to increase management involvement and oversight and to reduce personnel errors. Your reply to this letter and the results of future inspections will be considered in determining whether further enforcement action is warranted. In accordance with 10 CFR 2.790, " Rules of Practice," a copy of this letter and the enclosure will be placed in the NRC Public Document Room. The responses directed by this letter and the accompanying Notice are not subject to the clearance procedure of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511. Sincerely, Original signed by 3 James. G. K?ppler James G. Keppler Regional Administrator
Enclosures:
1. Notice of Violation and Proposed Imposition of Civil Penalties j 2. Inspection Report No. 50-346/84-15(ORP) cc w/encis: T. O. Murray, Station Superintendent Harold W. Kohn, Ohio EPA James W. Harris, State of Ohio Robert H. Quillin, Ohio Department of Health i i i i 1 r 1
Toledo Edison Company 2 NOV 211934 closed. A review of these occurrences determined that an adequate 10 CFR 50.59 review was not conducted, that approved procedures for operating the system werc not followed, and that operators failed to implement the corrective actions you initiated following the discovery of this problem on May 14, 1984. In addition, a recent Performance Appraisal Inspection identified additional deficiencies with regard to the conduct of reviews in accordance with the requirements of 10 CFR 50.59. This inspection also identified two examples when on March 8, 1984 and May 4, 1984, lead shielding was hung on decay heat piping and no safety evaluations in accordance with 10 CFR 50.59 were performed. These events indicate the need for significant improvement in your ability:
- 1) to recognize the design basis and technical specification requirements.for equipment operability and to ensure that these requirements are met when equipment is removed from service and 2) to ensure that procedures which define requirements for equipment operability are followed.
These events also indicate the need to ensure that adequate corrective actions are taken to preclude repetition of identified deficiencies. During the September 23, 1982, Systematic Assessment of Licensee Performance (SALP), we identified a weakness in your ability to recognize design basis requirements for equipment operability. The NRC Region III staff restated this concern during an Enforcement Conference on March 9,1983 and again during the October 28, 1983 SALP. As a result of the March 9, 1983 Enforcement Conference, you committed to implement a Comprehensive Corrective Action Program to address these and other concerns. You also assured us that other administrative measures were being implemented to deal with these problems. However, your corrective actions have been ineffective as evidenced by your failures to recognize design basis requirements for safety-related equipment / systems. To emphasize the need for the licensee: (1) to recognize the importance of design basis and technical specification requirements for equipment operability and'to ensure that these requirements are met when equipment is removed from service, (2) to ensure that procedures which define the requirements for equipment operability are followed, (3) to ensure that appropriate reviews are conducted in accordance with the requirements of 10 CFR 50.59, and (4) to ensure that adequate corrective actions are taken to preclude repetition of identified problems, I have been authorized, after consultation with the Deputy Director, Office of Inspection and Enforcement, to issue the enclosed Notice of Violation and Proposed Imposition of Civil Penalties in the cumulative amount of Ninety Thousand Dollars ($90,000) for the violations described in the enclosed Notice. The violations have been categorized in the aggregate as two Severity Level III problems in accordance with the General Policy and Procedure for Enforcement Actions, 10 CFR Part 2, Appendix C, and the Policy as revised, 49 FR 8583 (March 8, 1984). The base civil penalty for Item I is $50,000. The base civil penalty for Item II is $40,000 because two of the violations identified occurred prior to the revisions to the recent Enforcement Policy. 9 l l
e i ,f i l f i I l 4 i i 4 ') ENCLOSURE 3 Attachment Supporting Q-23 i 4 -l 1 i l i i i l ) { l ?
EllCLOSURE 3: ATTA0fENTS SUPPORTIfE.0LESTION 23 -.}}