ML19308C448
| ML19308C448 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/09/1979 |
| From: | Hebdon F, Weiss S NRC - NRC THREE MILE ISLAND TASK FORCE, Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| TASK-TF, TASK-TMR NUDOCS 8001240604 | |
| Download: ML19308C448 (42) | |
Text
_ _ _ _
O I!-
T, vm 4
N U CL E A R R E G U L AT O R't CO MMI S S ! O N O
I f
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IN THE MATTER OF:
THREE MILE ISLAND l
SPECIAL INTERVIEWS l
l l
INTERVIEW OF SEYMOUR H.
WEISS I
i 0
l
?00R~0RGINAL Place -
Washington, D.
C.
Date.
Wednesday, September 5, 1979 Pages 1 - 42 O
1..
(202)347 3700 ACE -FEDERAL REPORTERS,INC.
OfficalReponen 444 North Ccpitel Street 8 0 01 2 4 0 hp Wcshington, D.C. 20001
/
NATIONWIDE COVERAGE - DAILY
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' CR 6772 AR 1
1 UNITED STATES OF AMERICA 2
NUCLEAR REGULATORY COMMISSION 3
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _x Interview of:
5 SEYMOUR H. WEISS 6
- - - - - - - - - - - - - - - - -x 7
8 September 5, 1979 6935 Arlington Road 9
Bethesda, Maryland Conference Room 10 11 f
1, 12 The interview of SEYMOUR H. WEISS was convened at.1:40 O
ia e-14 Present were:
FRED HEBDON, FRED FOLSOM, and SEYMOUR H.
15 WEISS.
16 17 18 19 20 21
! O 22 23 24
'ce Federst Reporters, Inc.
25 1
x
2 l
I EEEEEE15 2
WITNESS:
Examination by:
Page:
3 O
4 SEYMOUR H. WEISS Mr. Hebdon 3
5 6
7 8
4 10 Resume of Mr. Weiss found after page 5 of the transcript.
i 11 Copy of " Transcript Notes from the Davis-Besse 1 Project Manager's Daily Telephone 12 Log, December 23, 1978" found following 13 P*ge 24 of the transcript.
O 14 15 l
16 17 18 19 I
I, 20 l
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22 23 l
24
.Feder:4 Reporters, Inc.
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'3 P3QgEER1NEg 2
When upon, 3
SEYMOUR H. WEISS O
4 was called as a witness and, having been first duly sworn, 5
was examined and testified as follows:
6 EXAMINATION 7
BY MR. HEBDON:
8 Q
Have you had an opportunity to read this letter 9
from Mr. Rogovin in the Special Inquiry group?
10 A
Yes.
II Q
Do you ha"e any questions or comments.concerning 12 that?
[]
13 A
No.
I4 Q
Do you understand it?
15 A
- Yes, i
16 Q
All right.
17 Would you please state your name?
18 A
Seymour Weiss.
19 Q
What is your current position?
20 A
I am the section leader of Section C of the Reactor
~
2I Safety Branch of the Division of Operating Reactors.
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0 What is your relationship to the Reactor Safety.
23 Branch and the Division of Systems Safety?
I 24 We'are in different divisions that handle the A
?wFederd f.eporters. Inc.
25 licensing and permit applications.
We would handle operating l
4 1
reactor work.
2 O
Would your involvement from a technical perspective 3
be similar to theirs, just a different time in.the licensing 4
process?
5 A
Pretty much.
The Reactor Safety Branch has in it 6
the disciplines, let's say, that the Reactor Systems Branch has 7
in DSS.
8 Q
Okay.
Fine.
What was your position in late 1978?
9 A
I think it was the same.
10 0
Would you describe your employment history, including 11 positions held at the NRC?
12 A
I have been in industry about, I guess, 19 years, n) 13 working -- working -- starting out workin? fr* ALCO Products
(
14 on small package reactors for the Army, as expcrimental 15 nuclear engineer.
Worked at the Westingaouse Bettis Laboratory, 16 doing reactor physics, experiments and z.nalyses, and 17 worked with the light water breeder as aell as some of the 18 destroyer projects, and worked in the reactor theory on I
l9 reactor development section at Bettis.
20 I then came to the NRC in the summer of 1973, 21 where I worked in the Office of Standa rds Development.
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22 I worked there three years, and then 1 came to the Reactor 23 Safety Branch as a section leader.
l l
24 O
So it would be about 1976?
kee.Feveret Reporters, Inc.
l 25 A
Yes.
I l
1
5 1
Q What is your educational background?
2 A
Bachelor's Degree, Mechanical Engineerina at New York 3
University.
Master of Science and Engineering at Union
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k-4 College, and Ph.D. in electrical engineering from the University.
5 of Pittsburgh.
6 Q
Approximately when were those degrees, just roughly?
7 A
How about if I give you a resume that has that?
8 O
All right, that will be fine.
9 A
Bachelor's Degree was 1958.
10 Master's Degree was 1961.
II And Ph.D. was in 1975.
12 O
Okay.
~( )
12 MR. HEBDON:
This is a document entitled " Resume, 14 Seymour H. Weiss." I'd like to have this included in the 15 i record.
16 (The document follows:]
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2s l MS-Fedefti Reporter, inC.
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RESUME
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th Seymour H. Weiss Professional Experience:
5/76 to present, USNRC, Office of Nuclear Reactor Regulation Duties: Section Leader in the Reactor Safety Branch of the Divigion Provides technical supervision and review of of Operating Reactors.
the work of nuclear, fuels, and systems engineers in Section C of the Reactor Safety Branch conducting evaluations of license amend-ments for operating nuclear reactors.
Coordinates and provides technical supervision of various generic problems associated with Serves as the safety of the reactor core and reactor systems.
supervisor of reviews pertaining to accident analyses, emergency core cooling systems, and related safety features.
Responsible for planning, coordinating and reviewing safety evaluations of design and performance of nuclear reactor cores, reactor systems and selected engineered safety features, and assessments of accident analysis. Member of ANS 5.1, Decay Heat Working Group.
7/73 to 5/76, USNRC, Office of Standards Development Duties: Nuclear Engineer in the Engineering liethodology Standards Branch of the Division of Engineering Standards.
Responsible for the development of nuclear reactor standards, codes, and criteria relating to reactor safety and for advising other NRC divisions in related reactor safety matters. Represents the Division of Engineering Standards in technical supervision of contractual assistance work performed by National Labs and private contractors and in technical meetings with NRC and industrial representatives relating to reactor standards, codes, and criteria.
3/62 to 7/73, Westinghouse Electric Corp., Bettis Atomic Power Laboratory, West Mifflin, Pennsylvania 2/67 to 7/73, Senior Scientist, Reactor Development and Analysis, Reactor Methods Section Os Duties: Development of methods of nuclear analysis. Writing of Evaluation and specifications detailing programming requirements.
verification of nuclear design and analysis programs.
Determine
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_- needs of projects in these areas and supervise implementation. Co-author of the PAXO3 program description manual WAPD-TM-888(L).
Also responsible for (1) the development of Bettis capability in computer aided electrical network analysis, and (2) evaluation of the nuclear design and depletion model used in A1W-3 prototype (Enterprise core depleted at NRF in Idaho).
3/62 to 2/67, LWB Project (formerly called LSBR), Experimental Physics Section Duties:
Experimental reactor physics, analysis and reactor opera-tions. Responsible for all analysis perfonned on experimental cores.
Assigned and supervised a group of four or more scientists in the performance of calculations, u;ing the LWB design model and more sophisticated Bettis codes, for comparison of experiment with calcu-lation. Maintained liaison between experimental and design groups regarding contemplated design model changes and effects of changes.
Co-author of U-233 core hazards report.
Specified handling and glove box procedures for U-233 fuel.
O em to 4m, tecterer, Un4.ersity of eittsburoh, Eiectrice, En2 neerino i
Department. Taught Direct Energy Conversion and Fundamentals of Electrical Engineering.
1/59 to 3/62, Experimental Nuclear Engineer, ALCO Products, Inc.,
Schenectady, New York Duties:
Experimental reactor physics, reactor operation, processing and evaluation of experimental data, development of data processing codes, error analysis, and report preparation.
Experience in experi-mental program planning. Contributed to performance of flexible critical experiments, end-of-life core physics experiments, Zero Power Experiments, and development of SM-1 A Core I integrated startup test program. Wrote summary report of SM-1 physics measurements through end of core life.
Education:
O sish schoei: Broax 8198 school of science - 9/so to e/s4 College:
New York University - College of Engineering - 9/54 to 6/58, B.M.E.
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Graduate Schools:
(1) City College of New York - School of Technology - Graduate Division, 9/58 to 1/59.
(2) Union College - Graduate Engineering Division, 9/59 to 5/61, M.S. in Engineering - Thesis
" Measurement of Neutron Flux and Spectrum for Use in Radiation Damage Studies on Army Reactor Pressure Vessels."
(3) University of Pittsburgh - Electrical Engineering Department, 9/67 to 12/75, Ph.D. in Electrical Engineering -
Dissertation
" Gaseous Reactor Concept Utilizing Direct Energy Conversion Methods."
4 Other Education:
(1) Union College - Graduate Engineering Division, 9/61 to 2/62, Course in " Theory of Functions of a Real Variable."
(2) Bettis Atomic Power Laboratory, 4/64, Fortran School.
(3) Bettis Atomic Power Laboratory, 4/66, Applications of Calculus of Variations to Reactor Theory.
(4) Bettis Atomic Power Laboratory,10/66, Statistics for Scientists and Engineers.
(5) Bettis Atomic Power Laboratory, 3/67, Advanced Reactor Kinetics.
(6) George Washington University, 3/74, Fault Tree Analysis.
(7) USAEC,1/75, BWR/PWR Radwaste Systems Course.
(8) USNRC, 6/78, Personnel Practices Seminar (9) Civil Service Commission, 2/79, Management of Time.
(10) USNRC, 3/79, Interpersonal Skills Workshop for Managers.
i
6 1
BY MR. HEBDON:
2 O
I'd like to ask you a few questions about an 3
incident that occurred in Davis-Besse in September of 1977.
()
i 4
I'm particularly interested in the knowledge or the under-i 5
standing of that incident that you had prior to the accident 6
at TMI, specifically prior to March 28, 1979.
What knowledge 7
did you have concerning an incident that occurred at Davis-l 8
Besse in Sectember of 19777 i
l 9
A As far as I could recall, very little.
In going l
10 !
back through my records, and my recollection is very little of i
^
11 what took place at Davis-Besse at that time.
For one thing, 12 l at about the time that occurred, our branch was organized on a
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13 somewhat different basis than it is now.
We were more or less 14 on a functional basis, and right now we are on a vendor basis, 15 so particular sections right now handle particular vendor's 16 reactors, and mine essentially handles B&W and CE reactors.
I I
17,
I think that in 1977, when the Davis-Besse events t
occurred, we were organized on a functional basis, such that you had three section leaders.
One was basic expertise.
His primary specialty was fuels and materials.
c 21 Another w s reactor physics and reactor design.
('~)
22 Another was ECCS systems and transients and 3
23yaccidentanalysis.
24 h 0
Which group were you assigned to?
niv4ederal Reporters, Inc.
25 A
My specialty would have been in the area of reactor l
7 1
physics core design, transient and accident analysis.
i 2
0 In your capacity of reviewing transient accident l
3 analysis, did you become involved with that incident at Davis-
-s l
4 Besse?
5 A
No, we didn't.
I 6
0 Would it have been normal for your group to become 7
involved with that sort of an incident?
8 A
I don't think so.
I think it would probably have 9
been looked at, number one, by the Division of Systems Safety, 10 DSS, because the plant had not been transferred to DOR yet, 11 and I would presume it was one of the system's problem, and I 17 don't think the section would have' looked into it.
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13 Q
Do you recall attending a briefing by Jerry Mazetis s_,
14, concerning his trip to the Davis-Besse following the September 15 24th, 1977 incident?
16 A
No, I don't.
77 When would that have occurred?
1 Q
It would have been about the 30th of September, 1977.
A My notes, the file I have with me, don't go back that far, but I have prior to this meeting skimmed through M
what I have, I don't think I have that on my records.
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O All right.
23 d A
I don't recall going to that briefing.
i 24 Q
In a more general sense, at that particular point nee Fe1 erat Reporters, Inc. g 25 h in time, if a plant had not yet been transferred from the h-l o
8 Division of Project Management to the Division of Operating I
2 Reactors, what would your involvement have been with any sort of incidents or any sort of problems that developed?
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4 A
We would not have gotten involved.
5 0
So it would have remained entirely within DSS?
6 A
Yes.
For them to handle that sort of situation?
7 0
8 A
Yes.
Unless for some reason they requested DOR 9
involvement.
10 0
Okay.
And to your knowledge, they didn't in that 1
II I particular case?
12 A
That's right.
There was another incident that occurred at Davis-I3 O
Id Besse on November 29th, 1977, that was essentially a cool-15 down transient that resulted ia a fairly low pressurizer 16
- level, U
Prior to March 28th, 1979, what knowledge did you j
have concerning that incident?
A Very little.
It would have been about March 29th that Creswell's board notification reached me, and at that 2',
point I started going into it.
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22 h Q
Okay.
Were you involved with the review of a "g
proposal by Toledo Edison to use a dual setpoint for generator
.9 24 -
level at Davis-Besse in certain transients?
. 4 cr Feieral Reporters, Inc. f j 25 A
Yes.
That review was done in my section.
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9 1
0 Approximately what timeframe was that review done?
2 A
Late November-December '79.
For better dates I 3
could go back to the testimony I gave to the Presidential
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Commission.
It's in that --
5 O
Maybe we can get back to that a little later.
i 6
What was the reason for the dual setpoint?
j i
7 A
Davis-Besse was still doing start-up tests and one 8
of the tests they wanted to do was the loss of offsite 9
power tests, and I&E, Inspection & Enforcenant, had some l
10 concern as to whether the procedures were okay, and whether 11 there was an unreviewed safety guestion.
12 Now the procedures -- procedural aspect in doing
()
13 tne test, they were changing their procedures, at least as 14 far as I recollect, they were changing their procedrzes, 15 such that the operator would take control of the e < feed 16 pumps and maintain the steam generator level at a nominal of, l
'7 l I think,.35 inches instead of letting it automatically go to d
120 inches a
O Do you recall the reason for that change?
A Right.
This is going to be part of the other.
2i J O
Okay.
1
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22 A
The unreviewed safety cuestion was brought about u
23 by I&E because they were saying that, well, should NRC approval 24 be obtained for this procedural change, as well as the nee vrwe neponen, ene.
25 reason for the procedural change.
That being they would l
10 1
lose pressurizer level indication because of the rapid cool-
'2 down, because the aux feed pump flow was so great, would cause 3
the system to shrink, and you would tend to drain your
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4 pressurizer.
i 5
O So then it was your understanding that part of the 6
concern was this concern about loss of pressurizer level 7
indication; is that correct?
I I
8 A
I think at the time we were not at all concerned 9
about loss of level indication.
We were concerned about the 10 actual loss of level.
11 O
In the pressurizer?
I2 A
In the pressurizer.
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13 O
Actually having the pressurizer go dry?
14 A
Right.
15 Q
So that was one of the concerns, the question of f
16 the possibility of emptying the pressurizer is one of the concerns you were considering as part of the review?
I 4
A Yes.
~
Q Why was this particular issue being reviewed by your group?
Had Davis-Besse been transferred to DOR at that point?
21 A
I don't recall.
I would have to see if I could pull p
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22 f;lout the date of transfer.
?
23 ll I would guess -- I don't recall.
If you want me to --
1 24 is that -- if that date is important, I can try --
me.Feseest Reporters, inc, j 25 i
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BY MR. FOLSOM:
f t
2 O
Can you supply it later?
p 3
A It's in the transcript.
,s 4
0 It's in the transcript?
I did look it up.
5 BY MR. HEBDON:
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6 0
We can find that later.
j i
7 A
To the best of my recollection, I think it had just l
8 been transferred.
9 Q
That's no problem at all.
We can get it out of the 10 transcript later.
l 11
[ Pause.]
12 What were the conclusions of your review?
What
)
13 were the results?
14 A
Well, we had a number of staff meetings and we had 15 l a number of -- or at least one phone call with the licensee l
16 i and the vendor that I recall.
i 17 il Me concluded that relative to it being an unreviewed 1
safety question, if-the operator was taking manual action only to preserve pressurizer indication level indication, that this was okay, but if the operator was taking manual 2I control in order to survive a particular transient, then g
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22 h this would be an unreviewed safety question.
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Now if he were taking the manual action to prevent l'
24 emptying the pressurizer, would that have been an unreviewed 4c14tdera6 Reporters, Inc.
25 safety issue?
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A Right, to prevent actual emptying of the pressurizer.
2 O
Now did you have any analyses performed, or did you 3
perform any analyses to determine whether or not he would in Q
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fact empty the pressurizer during any particular transient?
4 5
A Davis-Besse had submitted analysis which I presume 6
had been performed by B&W that the pressurizer, while it did l
l 7
lose level indication, did not drain, was not emptied.
8 0
Did you analyze or did you review those analyses?
9 A
It was reviewed in my section, yes.
10 0
Who actually performed that review?
l 11 A
Gene Imbro.
l 12 l Q
Okay.
Did he conclude that the analyses were correct?
8
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13 A
Yes.
14 l Q
So then it was the conclusion of your section that 15 B&W's analysis that showed that the pressurizer would not 16 empty during these transients was in f act accur wte?
U A
Yes.
.3 i O
Okay.
Do you have any idea or were you aware of why they had-initiated this particular change in the first place, why 2;. they were concerned about this loss of pressurizer level u
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22 l indication?
23 ll
'A No, I don't think so.
24 O
Did you ever hear any connection with that particular AcaJederal Reporters, Inc.
25 problem and the November 28th, 1977 incident I referred to i.
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13 1
earlier?
2
.A I don't recall that it was brought up.'
I don't 3
remember.
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4 Q
Okay.
I i
5 Were you aware of concerns that had been raised 6
by James Creswell of I&E Region III relating to this issue 1
7 of pressurizer level?
I j
8 A
Not at the time, no.
l 9
O Are you aware of those concerns now?
I 10 A
Yes.
11 Q
When did you become aware of them?
12 A
I would guess within the last month or so.
()
13 Q
Now I think you started to mention earlier that 14 you had received this board notification and that was back 15 l in. January of this year?
16 A
I received his board notification on the day of the U.
TMI event, so it was March.
I O
March 28th?
A I guess that would have to be when it was.
0 All right.
Do you see see any connection between 2 '.,
the concerns that were raised by Mr. Creswell and the issues
()
22 that you reviewed in the course of Jeviewing the 35 inch 23 level as opposed to the 120 inch level?
.24 A
No, because we'didn't look at the level indication
- aca Feieral Reporters, Inc.
25 at all.
We were concerned about the steam generator level
F 14 1
and the operator controlling either to 120 inches or 35 2
inches, and their eventual addition of a dual setpoint i
3 indicator which would do this automatically.
The -- we had t')
4 concerns whether the pressurizer would drain or whether the 5
pressurizer would overflow and go out one of the relief 6
valves, but we never addressed level indication as such.
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Q But you don't feel that your consideration of the j
i questionofwhetherornotthepressurizerdrainsisassociatedl 8
9 with the concerns that Mr. Creswell had?
10 A
I think that is correct.
I think Mr. Creswell's 11 concerns were level indication value to the reactor operator, 12 and at the time we didn't look at that at all.
m 13 0
- Okay, i
14 l
Do you recall ever having any discussions with i
15 people in the I&E regions, particularly I&E Region III, 16 l concerning the potential safety significance of pressurizer c
level indication?
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.c A
Okay, I am aware from what was presented to me at the Presidential Commission deposition that statements were made by Region III inspectors that we were looking at the 2:
pre.ssurizer level indication.
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22 To the best of my recollection, we never addressed s
i 23 1; pressurizer level indication.
We were concerned with the dual 24 '
setpoint indicator steam generator level, and operator
, e.co. Federal Heporters, Inc.
25 control of the aux feed system to maintain a certain steam l
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15 1
generator level.
2 O
But from what you've said, you were concerned about 3
the actual pressurizer level?
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A Yes.
l l
5 0
But you were not concerned with pressurizer level 6
indication?
I I
7 A
Correct.
8 Q
Now it would seem that it would be not too hard 9
for somebody to confuse those two issues.
Prior to the TMI 10 accident and all the things we've learned from that, what J
11 would you consider -- what would you have considered to be i
I 12 the safety significance of a loss of pressurizer lavel
()
13 indication?
14 A
I would not have been -- well, I can't really say.
15 It's easy to say, well, I would not have been concerned with 16 it, thinking in terms of absolutely divorcing TMI from my mind, U
and I would have to think that I would have looked at operator interaction in a transient and been concerned with it.
c It's a hard thing to say.
I don't know.
Q Well, the thing I'm trying to get a perspective 2;
for is if in the course of a conversation with someone, L
22 possibly someone at the I&E Region at I&E Region III about 23 this issue of steam generator level and the dual setpoints, 24 :
and all the rest of it, if the cuestion had come up, if the
- Aes Federst Reporters. Inc. '
25 issue had come up concerning whether or not a loss of l-i
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16 1
pressurizer level indication low, having to go off scale low, 2
if a question as to whether or not that was an unreviewed 3
safety issue had come up, how hard would it have been for
.)
4 you to come up with an answer?
Is it something you would have 5
had to go off and analyze for some period of time?
Or is it 6
something you feel you could have answered off the top of your 7
head, or how hard would it have been for you to develop an 8
answer to that question?
i i
I 9
A My notes indicate that a meeting was held with myself 10 Gene Imbro from my staff, who was reviewing the Davis-Besse i
11 1 submittals, Guy Vissing, who was the Davis-Besse project i
12 manager, Brian Grimes and Mr. Fairtile as to this issue, and
,~
13 my very cryptic notes indicate that we concluded that, as I 14,
stated earlier, if only to preserve indication, manual action 15 was not an unreviewed safety question.
I 16 l If manual action was required to survive a tra'hsient, l
17 then that would represent an unreviewed safety question, so I
.1 apparently we had some staff discussions to formulate this
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conclusion.
Q Let me try one more question.
I guess I'm still a u
2, little confused about how that conclusion you just stated, if n
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22 h manual action is required just to save pressurizer level 23 [ indication, that that's not an unreviewed safety issue, how 24,
does that differ from saying that simply the loss of
..ce Fritral Reporte's, Inc. I 25 fl pressurizer level indication low is not an unreviewed safety a
1 i
I 17 1
issue?
2 A
We were looking at it under the context of the test, 3
they are undergoing this loss of offsite power test, and the O
4 question arose -- I presume it was from Region III -- as to I
5 whether or not this was an unreviewed safety question.
6 The meeting that we had on December 18th with the i
7 participants that I just mentioned to go over that came to the 8
conclusion that it was an unreviewed safety question only if 9
manual action was required to' survive a transient.
10 0
Which you concluded it wasn't, the action wasn't II required?
I2 A
That's right.
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O But I guess I'm still a little confused.
You're Id saying that if manual action is required to maintain pressurizer 15 level, and only to maintain pressurizer level indication, that i
I6 ! that is not an unreviewed safety issue?
17 Now I still don't understand how that differs or if I
it differed.
Is that different from saying that loss of pressurizer level indication is not an unreviewed safety issue?
Is there a difference between those two statements?
.:q A
I guess I'm a little hazy about the point you are 22 0
(-
trying to make.
23i O
Well, you are telling me that -if the operator has 24 i to.take action simply to-save pressurizer level indication,
- ce.Fejeral Reporters, Inc. ;
25 that's not an unreviewed safety issue.
Is that correct?
- it N
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18 1
A Under the context of performing this test over this 2
one time, yes, I think that is true.
3 0
Well, wasn't he proposing this as a much longer term 4
solution?-
5 A
I wouldn't say it was a very long term solution.
I 6
They were proposing that they would install a dual setpoint 7
level indicator which would automatically do the same thing.
i 8
I think the licensee was also stating that they could do this 9
test and lose level indication without any problem.
They 10 didn't want to do it that way because they wanted to do it i
11 the way the plant would eventually be modified to operate 12 with the dual level setpoint.
(])
13 0
I guess that gets me back to the same point again.
14 If the utility came to you and said, "We can do this test, 15 but we're going to lose pressurizer level indication in the 16 course of it, is that all right?" -- if they had asked you T7 that question, what would your answer have been?
.i a
A That's kind'of clouded by the-circumstances as they are today.
O Well, try to divorce that answer as much as you it 2; j can from Three Mile Island.
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22 i,
A It would not have been an automatic "it's okay."
We 23 h would have looked into it to see if there were any problems 24 with that, and.that is, I think, what we did at the time, to
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25 the best of_my recollection.
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-Q Okay.
2 BY MR. FOLSOM:
3 O
My point that I wanted tor: raise is when do you cross 3
4 the line between manuni operation to preserve pressurizer 5
level indication to tl.e point where manual operation becomes l-6 essential to preserve the plant from damage, if that's what I 7
understand the dichtomy to mean?
8 A
well, the licensee's analysis that he had submitted 9
had indicated that there was a substantial amount of water 10 left in the pressurizer -- I don't recall the levels any more.
11 If it had been at all conceivable that he could pull a bubble 12 '
into the primary system, that's where we would have drawn the I'N 13 line.
V 14 BY MR. HEBDON:
15 0
You would have found that to be an unreviewed safety 16 [
issue?
17 A
Yes.
.c BY MR. FOLSOM:
O My second question is:
Is it your response in I'
this instance was totally confined to an interim test situa-21.;
tion between the time of the inquiry and the establishment
()
22 of the dual setpoint indicators?
- I l
23 N A
Yes.
h 24
.Q Is that correct?
. eke-Fedevel Reporters. Inc.
25 A
Yes, if I understand your question.
We looked into lsI
,1
20 1
-what the operator's procedures would be, what he had to do, if 2
it were feasible that he could do this.
That was part of our 3
review.
4 Q
All of this is for a temporary situation, is it not?
5 A
Yes.
6 Q
Was it not?
7 A
Yes.
8 BY MR. HEBDON:
l 1
9 O
Now let me clarify something now.
They had intended 10,
to go to a permanent design that would have automatically 11 maintained this dual level?
j 12 A
Yes.
I
()
13 Q
Did you review that?
14 A
The actual device itself, I think, was reviewed by 15 the plant systems branch, who do reviews of electrical compo-I 16 [ nents.
i 17 O
Did you review the design consideration of having d
an automatic dual level setpoint?
A I think that was done in the section, yes.
2; O
Now you mentioned that you were reviewing this 21 procedure primarily for the purposes of allowing them to do
()
22 this test.
There was going to be some period of time between 23 when they cor.pleted the test and when the permanent installa-g 24 tion would be installed?
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25 i
A
'Yes.
F l
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21 1
Q Do you recall approximately how long that was going 2
to be?
3 A-No, I don't..
I don't.
4 O
During that ;oriod of time the plant would be 5
operated, I would assume?
f 6
A
- Yes, 7
O If an incident or accident were to occur, was it f
8 your understanding that the operators would go ahead and use 9
that same manual dual level setpoint?
9 I
10 A
Yes.
II O
So then it was more than just the test itself; they 12 were going to be using those dual level setpoints as an l
I
)
13 interim solution until they could make the permanent fix at I
l I4 l
some point in the future?
i 15 A
Yes.
16 0
Okay.
17 Do you recall any discussions that you had with Mr. Kohler about any safety issues associated with this e
particular review of the dual level setpoint issue?
2-A With Mr. Kohler?
2; O
With Mr. Kohler or anyone else from I&E Region III.
/nd 22
.A
.If I'm not mistaken, I had a number of phone n
23[- conversations, I guess, with Mr. Kohler, and Mr. Creswell, and 24 I think with Mr. Streeter of Region III.
Ace Federal Reporters, Inc.
25 g
Do you recall approximately when these phone l
U 1.
22 1
conversations occurred?
2 A
No.
Mr. Kohler, Mr. Creswell, and Mr. Streeter 3
apparently do have those phone calls logged in, at least 4
they so indicated to the Presidential Commission.
5 O
Do you remember the approximate timeframe?
Was 6
it January, December?
7 A
Yes.
8 0
So it was some time around the end of 1978, early 9
1979, prior to the Three Mile Island accident?
10 A
Yes.
{
11 0
Could you recall for us the content of those 12 phone calls?
l
()
13 A
We had a conference call with -- off the record.
14
[ Discussion off the record.]
15 MR. HEBDON:
Back on the record.
16 l THE WITNESS:
We had a conference call with Davis-i 17 Besse and Region III on December 23rd.
Transcript notes of this conversation are in Brian Grimes' Presidential Commission transcript.
But one of the things we were concerned about at the conclusion of this conference call was whether Davis-Besse 2; q had looked at the loss of offsite power when they did their I_ss)
- 22,!
small break ECCS calculation, small break LOCA.
23 And it seems to me that the conversations I
'l 2^
subsequently had with Region III were relacive to whether, you Ace Federal Reporters, Inc.
25 know,.this particular item had been resolved.
I
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23 l
1 BY MR. HEBDON:
2 O
This question of whether or not they had done an 3
ECCS analysis, assuming a loss of offsite power?
j
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4 A
Yes.
w i
5 O
What was the conclusion of that?
What did you 6
conclude?
7 A
We had asked them if they had not done it, they 8
should do that analysis and submit it to us.
9 0
Was it your assumption that the ECCS analysis had I
10 been done without assuming a loss of offsite power?
i II A
I wasn't sure.
I would have thought it had been 12 done, assuming a loss of offsite power, but when the question
(}
13 was asked of them, they didn't really know, or they didn't 14 have the right people there to respond to that question.
15,
O So it was your understanding then that this was 16 the basis of the concerns by I&E Region III?
17 A
Well, when I&E Region III asked or would have asked
' us did we have anything ti.'.t's unresolved, this would have
.c been one of the unresolved items, according to my notes.
2; The other~one would have been the commitment to O
submit for review the dual setpoint indicator.
q i-( 22 O Would it be possible to get a copy of these notes 23 0 that you were just referring to? 24 .A Okay. They are in the Presidential Commission ( Aca.Fe12ral Reporters. Inc. h; 25 0 transcript for Brian Grimes. I have no problem with that. i; l 4
24 1 0 okay. If we could, it says here -- just to make a 2 copy of this, and we'll return the original to you. 3 MR. HEBDON: For the record, this is two pages of I 4 notes. The title is " Transcript Notes from the Davis-Besse 5 1 - Project Manager's Daily Telephone Long, December 23, 1978, 6 Concerning the Steam General Dual Level Set Point Modification," 7 and I'd like to have those included in the record. l 8 [The document follows:] i 9 i 10 i t l 12 (:) '3 14 i 15 16 l 17 l .c, 21.r. !i 22 l ( )- 22 ll 24
- An+rsersi neponers, inc.
25 -i i \\
Transcript. Notes. from the Davis Besse-1 Project Manager 's Daily Telephone Concerning the Steam Generator Dual Level Set Point Modification: (December 23, 1978) b.2. Conference consisting of Brian Grimes, Ed Jordan, Don Kirkpatrick, Symors Weiss, Mort Fairtile, G. S. Vissing concerning TEco letter dated 12/23/78 December 22, 1978 - Steam Generator Level Set Points 73 l Questions which were developed: ~ What h'appens when operator fills SG and pressurizer goes solid? l. What is the effect of over cooling transient - when SG becomes filled with 2. cold HPI? Do we exceed TS limits on cooling? Is it for loss of What is the ' basic reason for the 35" level setpoint? '. 3. offsite power or for loss of feedwater? 4. Concern for operator going to 35" during a LOCA. Concern for Loss of offsite power in which operator goes to 35" set point and Does 10 min. operator action effect ECCS analysis? 5. then we have a small break. Call to Region III, Tom Tambling, John Streeter, Richard Knop with 12/23/78 the people previously mentioned. Discussed above. 12/23/78 Conference chil to TECo, IE & 00R IE III - Tambling, Streeter, Knop IE Hdgr. - Ed Jordan, Don Kirkpatrick TECo - C. Domeck, E. Novak, Plant Rep. DDR - Grimes, Weiss, Fairtile, Vissing 1. Would system go solid on HPI? Answer HPI pump shut off head is 1600F#" Would not go solid. 2. Consequences of overcooling transient. This Answer - Plant is designed to sustain 40 loss of power to coolant pumps. is similar to overcooling trrnsient. Have had loss What is reason for controlling SG at 35"? Operation is at 40". 3. of feedwater @ 40% power and controlled at 35" by operator action. Operator can quickly take this kind of action. Concerning implications on seall break LOCA. Does natural circulation test 4 gS. show better heat transfer capability? Are they still in ' compliance with the () - regulations with regards to LOCA? TECo says that natural _ circulation flow test showed greater flow than anti-B&W says there.is no change with analyses for small break. cipated. go into 35" level set point when How can TECo be sure that operator is can there is a small break? TECo says operator has good indication, alarms, etc. TECo says that B&W tells them that with RC pumps running the 35" level setpoint
.c is OK for small break. What happens when the plant has a loss of off site power and then a small break? - TECo could not answer this. They said that it would be cleaner to change to auto set point at 35" but this would take a long time and money with new ECCS analysis. They think the dual level is quickest method. Brian Grimes said that the design should consider loss of offsite power and (])~ then a LOCA. TECo said they have not analyzed that event. We asked if this analysis is required under 50.46. We asked them to talk to B&W to check and make sure if this analysis should be in the record. Conclusions (1) We asked TECo to look at theloss of offsite power and then a small break - make sure they are in conformance to 50.46 (ECCS) (2) We requested within a commitment to have a (1) a design within 60 days of permanent modifications, (2) 50.59 determination within 60 days for the mod with copy of safety analysis, (3) schedule (as early as possible - before next refueling) of the modifications. We concluded interim fix OK and TECo has no' restraints. Also let us know outcome with talk to B&W concerning compliance with ECCS 50.46. () I C
25 1 BY MR. HEBDON: 2 Q So when they asked you then if you had any unresolved 3 concerns, were you at that point considering this concern of ('~)' 4 whether or not the pressurizer emptied? 5 A No. That had been resolved prior to that phone 6 conversation on the 23rd. I 7 Q So that issue was already resolved? l 8 A Yes. 9 Q Would you have told them that that issue was i 10 already resolved? j i 11 A They were on the phone with us. 12 0 Did that come up in the conversation? Did the issue ! () 13 of whether or not -- 14 A [ Witness reviewing documents.) t 15 I think it did. I'm not sure if it's really -- if 16 l it's covered in here in these notes. The -- prior to this ui conference call, Davis-Besse had submitted analysis that showed 11 11 what level pressurizer went down to. That was discussed. And I assumed that I&E would have known that we didn't have any question on that. 2; j Q At this point in the analysis, at this point in () 22 your discussion, would you -- might you have had or did you 23 make a statement similar to your comment of a few minutes ago 24, to the ef fect that if the manual action is required, simply ,W-Feierst Reporters, Inc. l 25 l to save the pressurizer level indication, then it's not an Jli
f 26 1 unreviewed safety issue? Would that sort of a comment have 2 come up in the course of that conversation? 3 A I think so, yes. (_-) 4l 0 Okay. j 5 I'd like to go on to some specific questions about t 6 the dual level setpoint analyses. We've been reviewing a j l 7 fair number of documents that have been provided by B&W, and inj l 8 some of those documents they make some comments concerning j i 9 some aspects of their small break LOCA analyses, and I'd like 10 to try and discuss some of these with you, because they discuss;-- i i 11 ' they seem to imply that they had some concerns that they felt i 12 l the NRC may or may not have been aware of. r), (_ 13 First of all, were you aware or are you aware 14 that the B&W small break topical report was based on a 32 foot auxiliary feedwater level? 15 ' o 16 l A I was not, no. 17 Q Have you reviewed that topical report in any detail? ,e A No, I haven't. O So you're not particularly familiar with it? A No, I'm not. 21 O Would you have reviewed it in the course of reviewing e /~5 I (_) 22 l the dual level setpoint analysis? 23 A Well, if B&W did not make a submittal incorporating 24 the loss of offsite power into their small break LOCA, as ,we.FedtrM Reporters, Inc. 25ll apparently they didn't, it would have been handled by a phone 0
27 1 call from them indicating that the analysis that's on record 2 did assume loss of offsite power. 3 Q Do you know if it was finally concluded that the nU 4 analysis assumed a loss of offsite power, or was it finally l i 5 concluded one way or the other? 6 A To the best of my recollection, in talking with 7 the people who did the review, the analysis was done assuming 8 loss of offsite power. f. 9 0 Okay. 10 And the dual level setpoint, they assume a 10 foot 11 level for the small break dual set setpoint, small break j 12 half of the dual setpoint. Do you know where that number came 13 from? 14 A No, I don't. I 15 O Did you review in the course of yourureview the 16 dual level setpoint? Did you review the adequacy of that 17 I 10 foot level? l A I don't recall really the details of that review. .c O If the B&W small break topical report had assumed a 32 foot level, and you were using a 10 foot level, in the 2:. dual level setpoint, would that have caused any concern to you? () 22 I A Yes, I think we would have asked them to clarify i 23 ] why the different assumptions were made, and what is the 1 24 i effect of the difference on the analysis. ,,.ce-Fe feral Reporters. Inc. f 25 ; Q Would it have required very extensive reanalysis l 8
28 1 in your mind to justify the 10 foot as opposed to the 32 foot? 2 A I can't_say. I don't know. 3 Q Okay. (- 4 Would it have required re-review or relicensing of l 5 their topical report? 6 What I'm trying to get a feel for is how big a l 7 problem would this have been for them if they had done their 8 original topical report assuming 32 foot level, and now they i 9 wanted to use 10? How hard would that have been to get that ~10 approved? Would it have been a big deal? Would it have been I 11 a minor phone call? Would it have been a meeting to chat 1 12 about things, or what? i 13 A Generally, as the LOCA models are changed, mode.1 14 l changes are reviewed by DSS as opposed to DOR. I 15 Q Do you know if B&W or the NRC have ever assessed 16 the small break assuming that the reactor coolant pumps l continue to operate? 17 .c A Well, that's getting into more recent events. I had not looked at B&W small break analysis prior to seeing some Z, of the submittals that are coming in relative to the Bulletins 21 & Orders and Lessons Learned questions. 22 Q If someone had told you back, say, in mid-1978 or 23 : late 1978, that the B&W small break LOCA analyses did not 24 ' consider or did not include an analysis of that transient Ace-Federal Reportees. Inc. 25 with the pumps running, would that have been a concern? I N ,u
o 29 1 A With the primary coolant pumps? 2 0 Running, with them operating. l l 3 A~
- yes,
%) 4 0 Would that have been a particularly significant l I 5 concern? 1 6 A I can't answer that. I don't know. I 7 0 Okay. j i 8 Did you consider the generic implications of the 9 dual level setpoint? 10 A Yes. One of the things we were concerned with was i II this problem, a Davis-Besse problem, was it generic to all i 12 B&W plants. 13 In conversations with B&W, in December of ' 7 8, I Id think it was early '78, had indicated that the aux feed pumps 15 for Davis-Besse had 50 percent greater capacity than the 16 l other B&W plants, and that's what caused the rapid cooldown E U and shrinking of the primary system. E' O So then it wasn't so much a problem of the level to which it fell, but the rate at which it fell? 2-A That's right. 2; j Q So then you did consider this for other B&W plants 22 and decided it would not be a problem for then because of the U 23 ! slower filling rate? i 1 .d j A-Yes. Ace-Federst Reporters, Inc. } 25 ] g red like to go on and ask you some general questions .i
i l 30 1 that are. associated just with the overall licensing and i 2 review process. 3 Does DOR review plant procedures? \\s l 4 A I don't think so, no. l 5 .O Does your group review plant procedures at all? l 6 A very little. We might get into some questions on 7 tests, as to how a particular start-up test is being done, to 8 see if it's measuring the right thing that we had specified, l 9 when we said do a certain test. We might get into'looking at procedures if Inspection l 10 II Enforcement asked our opinion. l 12 Q Would you routinely review operating and emergency l O is procedures 2 I4 l A I don't really know how to answer that. If we 15 had to make a decision on a safety problem and one of the 16 factors was did an operator have enough time to run to hit a I7 button or turn a valve or do that particular operation, we 2 would look into it and say did he have other responsibilities at the time? You know, what is going on? Can he do that? I .O But your review of procedures then would be 01.ly U from the perspective if the operator were required as part of 22 an analysis or part of an accident scenario to take a specific 1 23h action? 24 A If the licensee is going to take credit ior nes-Federst Reporters, inc, ! 25 l operator action in an event that DOR is reviewing, then we l' d
31 1 might look at not all of the procedure, but the areas of 2 interest. 3 Q All right. 4 What is your perception of the relationship that 5 exists between I&E and NRR? l 6 A I think it's quite good. There's no antagonism, l 7 there's no jealousy between the two organizations. We, I l l 8 think, are on a fairly good standing with Region III people. l 9 - There's no hesitancy on their part to call us up and ask us i 10 advice on a particular area they are looking at. 11 We would not hesitate to call them up, since we 12 are not at the plant, and they are, what do they see? If i () 13 it's something we need to know. 14 0 Is there a difference in your review proceduras 15 and philosophy between safety-related and non-safety-related 16 l systems? I 17 l A If a system is not taken credit for in the safety i analysis, presumably you don' t need the system to assure reactor safety, so we probably would not look at it. 20 O Do you ever consider whether that system could g 21 either as designed or as potentially operated, could interfere - () 22 with the safe performance of other systems? 23 lll i A I can' t say what goes on in the plants being n 24 initially licensed. When the plant comes in for its construc-cre Federal Reporters, Inc. 25 tion permit review or its operating license reivew, and they
32 1 have a piece of equipment, that's not safety-related, or 2 _that they've added a piece of equipment, I really don't know t 3 how that would be looked at by the people doing review. j (v~1 4 But if a plant is ir. existence and operating, and 5 it's a DOR plant, and they say, "We are going to add a piece t 6 of equipment or" -- well, we would look at the ramifications, 7 if it works, if it doesn't work, if it interferes; you know, I 8 that type thing. 9 I don't know what happens -- i 10 Q I'm really more concerned from your perspective in 11 DOR than I am licensing for now, because obviously that's 12 the area you are familiar with. If somebody says they are G(_) 13 going to put in a new piece of equipment, and it's not safety-14 related, would you look at how that system could interact 15 with systems that are safety-related and possibly interfere 16 with their performance? 17 A Yes, I think we would. 0 You do routinely address that issue on how non- ~ safety-related systems could interfere with safety-related A systems? A I can't say for sure, because I can't recall j 2; ;{ I (~j) 22 j examples, but-I think we would, yes. , y 23 [i O Okay. 24 What is the basis for deciding whether or not a nceJederal Reporters, Inc, ; 25 ' system is afety related?
J 33 1 A If it's necessary for the ~ safety of the plant, which 2 is equal to the same thing as saying the safety of the public. 3 BY MR. FOLSOM: (~) s i 4 0 What triggers tha$ question? What makes that ques-x- l 5 tion surface? Is it automatically considered with respect '6 to every piece of. equipment that is added on? 7 A I don't really know how to answer that. 8 O Does it depend on somebody asking the question? 9 A I don't know. A plant wodld come to DOR with tech 10 specs already written, with the safety analysis already Il performed and reviewed. 12 It is predetermined for us what is a safety system, 13 what criteria is it meeting. Id BY MR. HEBDON: 15 Q Do you ever have questions where one of your 16 reviewers would come forward and say, "This particular system 17 is not safety-related, but I think it should be"? 2 A I have not had that happen to me. O Is it your perception that the designation of systems as safety-related is applied in a consistent and E;. rational manner? 22 A I think it is, yes. il 23 ( Q So in your experience, you don't know of any cases 24 y or you have not found it common that a system is safety-related, ActJederal Reporters, Inc. - 25 and then another system that's necessary for its functioning,
34 1 for example, would be considered to be non-safety-related. 2 An example that's been b'rought to our attention l 3 is the' fact that the diesel is safety-related, but on some .g_ 's_) i 4 plants the air-start system for-the diesel isn't safety-related,' 5 and that would seem to be a somewhat inconsistent application 6 of that rule. 7 From your experience, have you found similar types l 8 of inconsistencies? l 9 A Yes. 10 0 Do you know r.' any other precursor events that II are relevant to the acv_ dent at TMI? i I 12 A Well, I guess the Rancho Seco event. () 13 Q Approximately when did that event occur? Do you 4 14 recall any of the details of it or why you feel it's a i 15 precursor? 16 A It was highlighted in Creswell's board notification. II Q Okay. Any others? IE A No. Q Do you have any additional information that might 21, be relevant to our incuiry in the events surrounding the o. 21, accident at TMI? ) 22 A I don't think so. Based on the testimony I gave j 23 j;. at the Presidential Commission deposition, T. tried to pull l 24, together as much as I could, and I don' t think I have anything i l,,ce sawai nemners, inc. 25 else. ~..
35 1 Q Okay. Do you have anything else to add? 2 A No, I don't. 3 MR. FOLSOM: Let me ask a couple of questions, if I
- ( s.
I ) uj 4 may. 5 BY MR. FOLSOM: 6 Q You are responsible for reviews on B&W, Combustion 7 Engineering, and research and test reactors? I I 8 A Yes, that's correct. 9 Q And you're looking at reloads, operating reactor r 10 safety problems, core problems, fuel, thermohydraulics, and 11 emergency core cooling systems? I 12 A Yes. () 13 Q What part do event reports from operating reactors 14 play in your education, if you will? 15 A We get monthly summaries of LERs that come in, which 16 is consistent with the LER number, the plant number, the 17 component, and a very, very brief description of what occurred. T-Those are reviewed on a semi-formal basis within the branch. In my section, I skim through it, and then I route it around L to my reviewers. 21 O And what do you do with the information that 22 i surfaces in these reports? n 23 fi A If you're seeing something that catches your eye 24 as occurring a number of times, I might ask a reviewer to look Ace Federst Reporters, Inc, 25, into it, number one, what is happening? Is it a safety problem? i u
36 1 And how does it impact on the work we do? 2 O Next question: 3 If it is, what do you do? l 1 .2 4 A We would identify it as an immediate issue to be j 5 brought up through management, depending on the significance, i 6 would determine on the time scale of the action we take. I 7 Q I'm a layman here. Don't you go down and bang on l 8 B&W's desk and say, "What the hell are you going to do with 9 this?" 10 A Yes, sure. 11 0 2ou don't do it, but somebody does it? 12 A
- Yes,
,r3 13 O Now explain that procedure for me. ' f 14 A We have never had the instance where one of my 15 reviewers has gone through and said, "Here are six events 16 ! that have occurred, and they could be a problem because of l Uh such-and-cuch," and that is what I am asking them to do. ll 2 BY MR. HEBDON: I Q You mentioned that your reviewers review the LER K smnmary, the monthly summary. Is that what I underctood you to 21 say? es ( ) 22 A Yes. I was going to continue. 23 H Q Okay. 24 1 A We do get its the branch a copy of all the LERs as Ace.Fz1 cts Reporters, Inc. l 25 :l submitted by the liccasec. The branch chief will then segment --
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37 1 send us those that pertain to CE and B&W reactors. The 2 other section, Westinghouse and the other section of boilers. 3 I would then go through these to see if there is any problem I w/ 4 that I think should be looked at by a reviewer in our area. 5 BY MR. FOLSOM: I' 6 Q Now is this the same procedure that you followed 7 last year and the year before? I i 8 A Yes. i 9 O How many LERs per month or annually do you see? i 10 A I would guess I see four or five a day. i Il O So would you say there is a considerable beaten 12 zone of experience that is covered by the LERs? 13 If you are an Army man, you will know what I mean, I4 or a Navy man. 15 A I don't. 16 O The beaten zone is the area where the bullets hit I7 the target, and if they form a close pattern, you begin to get d some concept of whether the rifleman is a sharpshooter or if it's all over the target, it doesn't have much significance, 2 except that he's not. hitting the mark. But if it forms a 2; ' pattern, then it would alert you to a problem, would it not, O sf 22 that might be a generic problem? .I 23! A I think that's a fair generalization. The primary 24 responsibility for the LERs remains with Inspection & Enforce- > Ace Fecerst Reporters, Inc. 25 ment. Inspection & Enforcement would do a review if problems i
38 1 turn up, or whatever their criteria.are, they feel they need 2 additional assistance, there would be a transfer of respon-I 3 sibility to either DOR or DSS. j ,_v \\] 4 Q Once again, as a layman, the hallmark of NRC in its 5 regulatory scheme is redundancy? 6 A Yes. 7 Q Now aren't you in a position to have a redundant 8 safeguard with respect to LERs, duplicating perhaps but i 9 backstopping what I&E does? l 10 A Yes, I think that's true. II Q How effective do you think that has been? 12 A I can' t judge that from what happened at TMI. We're (') 13 missing something. (m,/ 14 O All right. Coming down to specifics, then, did 15 anybody in your shop have any reaction to the TMI LERs? 16 There was one beforehand, I mean on the Davis-Besse two. 17 A The September-November -- c Q Yes, there were two at Davis-Besse and one at TMI. A I can't vpcak for the branch, but I'm pretty sure no one in my section did. 2I Q Would you have expected them to? I mean is this 22 something you could have anticipated they should have seen? .i 23 h I'm not asking you to damn your operation, but in hindsight. .t 24 4 A Well, I'm sure there was some branch review of what - Ace-Federst Reporters, Inc. 25 went on. It.was not done in my section. I'm not sure if I b
39 I can pinpoint actual reviewers who might have looked at 2 anything in that area. 3 O' To your knowledge, was anything done as a result, O initiated? 5 A I think since Davis-Besse was not a DOR plant, it 6 would have been difficult for DOR to get into what DSS was 7 doing. 8 O Is that a shortcoming? 9 A Well, you're getting into the fundamental, if there j 10 are any problems of the organization, whether the organization 11 should be the way it is. 12 My personal feelings are -- I don' t know if this is () I3 something that belongs on or off the record -- O Put it on the record. That's what we're struggling 15 for. 16 A My personal feelings are that allowing a group to devote itself to operating reactors and not to other reactors that are going through the licensing process gives and ~' highlights the operating reactors more than the old system before we were reorganized into a Division of Systems Safety
- ' ;j and Division of Operating Reactors.
() 22 I guess one needs to take what we have now and build 9 3
- !j upon it, and try and learn from what happened at TMI.
Where 6 7 nce Federsi Reporters, Inc. oS. h Q I don't want to develop a lecture here, but assume ll
40 1 something in a plant that is physically completed, but has not -2 had its license transferred to your division yet, it's still 3 in the, qualification stage. O 4 A Right. 5 q And something appears in an LER with respect to that 6 plant that your staf f would recognize as a generic problem 7 applying to plants that are in your management. 8 A Yes. j 9 The LERs come to our branch regardless of whether 10 the plant hasbeen transferred. i 11 O All right. Then the strict compartmentalization of 12 the organization is not necessarily a basis for sloughing off i ~ () 13 review of something that's gone on in another compartment 14 because the machine is going to be operated the same way in i 15 both departments. Am I making any sense to you on that? 16 A Let me try to interpret what you're saying, j-17 Q Please. O A For a plant, for instance, Davis-Besse, before its p transfer, if there is a particular LER that comes out on some piece of equipment or a system, it would come to me regardless 21 of whether Davis-Besse were in DOR or not, because it is a () 22 B&W plant. 23 If I were sharp enough, hopefully I would catch it 24 and-see the generic implications and alert my people to this. ,,aFetierst Fleportets, Inc. 25 ! And let's say that takes place, the reverse, we have feedback,
41 1 operating feedback, in which we try to feed back to DSS what 2 we have learned. I find that that is nice to complete the 3 paperwork trail, a documentation trail of those feedback memos. O 4 Generally we talk to DSS on man-to-man, reviewer-to-5 reviewer basis a lot quicker and a lot earlier than any kind 6 of formal feedback memo or documentation takes place. i i 7 O Do you feel that that is an adequate system, the i 8 talking? 9 A Yes. 10 BY MR. HEBDON: 11 Q You say that you review about four or five of these i 12 LERs per day? Physically approximately how long are they? () 13 How big are they? Are they couple of pages, 20 pages, 50 14 pages, on the average? 15 A Generally only a page or two. 16 Q Approximately how much time do you devote to that !7 ! activity? d .c A Oh, I guess I spend about no more than an hour a day. It depends on the type of LER. 2; O If you received an LER that was say, an LER supplement 21 that contained 50 pages of descriptive material and analysis g a 1 em() 22 h and graphs, and that sort of thing, would you read it all, or 4 23 0 would you just skim through it, or how would you review it? 24 =j A If it had to do with an area that was outside what aceJedera! Repo tets. Inc, I I 25 j the branch-I'm in reviews -- i
d 42 1 Q B&W plant? 2 A Let's say it's a B&W plant, it's an area that the 3 engineering branch or the plant systems branch would review, i ( -4 clearly not related to what we look at, I wouldn't spend very-5 much time at it. 6 Q Okay. 7 A If it were having to do with environmental. evaluation j 8 or siting evaluation, which is not what we look at, I would not 9 spend very much time at it. l 10 Q Let's say if it were in your area. 11 A If it were a significant LER with analysis, 50 pages l 12 or so in my area, I would skim through it and ask the reviewer ' () 13 to go through it in detail. 14 Q Okay. 15 MR. HEBDON: Do you have any more questions? 16 MR. FOLSOM: Not that I can think of at the moment. l MR. HEBDON: Do you have anything else to add? 17 d ~. 5 THE WITNESS: No, I don't. i 4 ' MR. HEBDON: Thank you very much. That completes the interview. 21 [Whereupon, at 2:50 p.m., the interview was ( _) 22 adjourned.] 23 l[ 24 i pa Faierst Remrters, inc. ; ) 25 ' l il -.}}