ML20062G703
| ML20062G703 | |
| Person / Time | |
|---|---|
| Issue date: | 08/11/1982 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | |
| References | |
| ACRS-T-1123, NUDOCS 8208130139 | |
| Download: ML20062G703 (97) | |
Text
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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS SUBCOMMITTEE ON EXTREME EXTERNAL PHENOMENA 9
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August 11, 1982 PAGES:
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Washington, D. C.
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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 3
SUBCOMMITTEE ON EXTREME EXTERNAL PHENOMENA 4
Room 1406 1717 H Street, N.W.
5 Washington, D.C.
6 Wednesday, August 11, 1982 7
The Subcommittee met, pursuant to notice, at 8
1400 p.m.
9 PRESENTL 10 DAVID OKRENT, Chairman MYER BENDER 11 J. CARSON MARK CHESTER P. SIESS 12 JESSE C. EBERSOLE PAUL G. SHEWMON 13 NRC STAFF MEMBERSs JIM KNIGHT 15 L. REITER D. GUZY 3
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MR. OKRENTa The meeting will now come to 3
order.
4 This is a meeting of the Ad visory Committee on 5
Reactor Safeguards, Subcommittee on Extreme External 6
Phenomena.
7 I am David CXrent, the Subcommittee Chairman.
8 Other ACRS members who are present or may be present 9
during this meeting are Mr. Mark, Mr. Siess, Mr.
10 Ebersole, perhaps Mr. Render.
11 The purpose of the meeting is to discuss the 12 ACRS recent recommendations on evaluation of seismic 13 design margins f or earthquakes more severe than the 14 SSE.
15 This meeting is being conducted in accordance 16 with the provisions of the Federal Advisory Committee 17 Act and the Government in the Sunshine Act.
Dr. Richard 18 Savio is the Designated Federal Employee for the 19 meeting.
20 The rules for participation in today's meeting 21 have been announced as part of the notice of this 22 meeting previously published in the Federal Resister on 23 Wednesday, July 21, 1982.
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24 A transcript of the meeting is being kept and 25 will be made available as stated in the Federal Register l
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1 notico.
It is requested tha t each speaker first 2
identify himself or herself and speak with sufficient i
3 clarity and volume so that he or she can be readily 4
heard.
5 We have received no written statements from 6
members of the public.
We have no requests for time to 7
make statements from members of the public.
8 It would seem to me, in looking over the 9
material for this Subcommittee meeting, tha t the brief to comment made by the ACRS in its report on Perry, in the 11 reported dated July 13, 1982, perhaps is a way of 12 opening up the subject to see what the staff has to 13 say.
In that report we said we recommend that the 14 Applicant and the NRC Staff conduct studies to evaluate 15 the margins to accomplish safe shutdown, including 16 long-t'erm heat removal following an earthquake of 17 somewhat greater severity and lower likelihood than the 18 safe shutdown earthquake.
He believe it is important I
19 that there should be considerable assurance that the i
20 combination of seismic design basis and margins in the 21 seismic design is such that this accident source 22 represents acceptably low contribution to the overall 23 risk from this plant.
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24 It is the same thing as stated in similar 25 letters, but I think that is maybe one succinct way of O
ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345 l
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1 doing that.
2 Okay.
Wh y don 't we begin with Mr. Knight and 3
see what the Staff has to say.
4 MR. KNIGHT Just to prove that there is 5
something to this business of great minds running in 6
similar channels, we have chosen an excerpt from the 7
Perry letter as perhaps being fitting as a way of 8
setting some context for the discussions.
9 (Slide.)
10 MR. KNIGHT:
Similar f rom the standpoint of 11 the working folks, so to s pe ak, is the excerpt from the 12 Wolf Creek letter for the we hope that during our 13 talk today we could focus to some extent on two things.
14 One is, of course, a discussion of the lower 15 probability, more severe earthquake.
The other, 10 however, I think from the standpoint of implemen ta tion, 17 it is really the question of needed modifications made 18 to the plant.
The logistics, I suppose one might say 19 the practicalities, one might say, of setting about to 20 make plant modifications require that you have some 21 standard thst you decide that at some juncture the 22 margins that are available or the situation that exists 23 is not acceptable and you now must go in and make
()
24 hardware changes.
But in order to do that, the designer 25 has to have a standard.
He has to know, am I going to i
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I have some different spectral intensity?
Am I going to 2
have some different spectral shape?
All the a questions 3
have to be answered.
O 4
(Slide.)
5 MR. KNIGHT 4 We go a little bit deeper, and we 6
talked about the sargins.
Are we talking margins on 7
code limits?
Would it be within keeping with the 8
philosophy expressed in the letters if we perhaps looked 9
at more advanced analytical techniques that still 10 demonstrated margin-to-f ailure on some presumably 11 technically founded basis, but distinctly different from 12 what is utilized in tne basic design.
13 If we get into testing of equipment and 14 looking at fragility limits, I can see questions arising 15 as to whether some sort of statistical level would be 18 acceptable.
If I tested 17 pieces of identical switch 17 gear to failure, would I take the 84th percentile or 18 would I be looking for 99.9 confidence they would 19-survive at some level?
These all become very real 20 questions.
~
21 Do you have acceptable fragility if you have 22 what some have already tagged as the SSSE, super-SSE?
23 Would I perhaps be willing to look at something less
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24 than the criteria that we apply for functionability?
In 25 other words, where I absolutely refuse to accept relay O
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1 chatter for the design basis, do I accept relay chatter 2
as long as t here wasn't structural f ailure?
They are i
3 all very reasonable questions.
fg V
4 Last, if I really were talking about very high 5
motions at very high stress levels, would I also depart 6
from such criteria as dasving or anything else that has 7
again a technical basis but one which would be different 8
from that which is applied in our usual practice ?
9 HR. OKRENT:
Could I offer a comment?
10 HR. FNIGHTa Yes.
11 HR. OKRENT:
I am speaking for myself, but I 12 have little doubt that the Committee did not intend to 13 suggest that you should strive f or the same margins, 14 whether there is a lower probability, more severe 15 earthquake that you strive for for the SSE, and so --
16 3R. KNIGHTS That is a most significant point 17 to us.
18 HR. OKRENTs So your point on inelastic, 19 certainly if by including inelastic you can show that 20 various things do well, that would cover those systems 21 and components, in my opinion, and it was couched in 22 terms of risk, you notice.
23 So where a question is appropriate to answer
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24 s ta tistically, that is also, it seems to me, an avenue 25 that one would follow.
Not all things may be O
ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345
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1 susceptib.le to the statistical determination, but some 2
may go that way.
3 And on the last point, if chatter doesn't 4
affect the ultimate thing you are trying to achieve, 5
then it is all right.
If chatte r bothers something, 6
there is no unique answer there, I suspect.
7 MR. KNIGHTa No, absolutely not.
And again, 8
as I say, in essence I suppose one might say we finesse 9
the question in the usual applications by simply saying 10 no relay cha tter.
Therefore you don't get involved in 11 auestions about which systems would interact with other 12 systems.
13 MR. OKRENT:
That's right.
So if you had a 14 system that could chatter, you would have to show that 15 it is okay.
16 MR. KNIGHTa Conceivably it might well be an 17 unanswerable question.
18 MR. OKRENT:
Hell --
19 MR. KNIGHTS But again, from our standpoint, 20 these doors are all opened and we've got.to be able to 21 provide sufficient guidance to a utility to be able to 22 say, all right, this is an increment, I guess, in our 1
23 view, an increment above and beyond what was anticipated
()
24 as being an adequate design basis for this plant, an 25 adequate licensing basis, and now provides sufficient 1
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1 guidance so that their goal is something we can come to 2
g rips wi th.
3 HR. SIESS:
On the question of inelastic, our O
4 experience has shown that structures, at least, show a 5
lot more resistance than is easily calculated.
You 6
can 't always calculate some of those margins that I 7
think are there.
But certainly NUREG/CR-0098 is an 8
attempt to formalize, what shall I say a recognition of 9
inelastic beha vior.
10 Wouldn't it be a legitimate guidance to 11 looking at some of those things?
12 HR. KNIGHTa Well, again, I think it might.
13 Again going to my personal opinion, I certainly think it
(
14 would be.
15 HR. SIESS:
And if it was actually intended --
16 well, it was intended, I think, for the SEP planth, but i
17 there is nothing wrong with applying those principles to 18 any plant that has already been designed and built, is 19 there?
20 HR. KNIGHT:
I certainly don't believe there 21 is, but as we all know, the experience in the past has 22 not always been the best when the Staff has taken a 23 particu.lar tack and then come before the Committee af ter
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24 the fact.
I think it ought to do this job the way it 25 should be done.
We really need to develop prior to
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1 ALDERSoN REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345
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1 laying the requirement on a utility the ground rules.-
2 There may well be situations that have to be treated 3
that way, but it is in my opinion a very significant 4
effort that could evolve here that would ha-e no 5
foundation for acceptability until we started examining 6
the end product.
That is really an unacceptable 7
situation.
8 MR. SIESS:
Well, in some of the previous 9
instances -- I guess North Anna was one, was it not 10 what we got back when we asked about margins was simply 11 the margins calculated stress, and those were usually 12 fairly substantial for piping, as I recall, and some 13 other things.
On 14 But even if those margins are not substantial, 15 that does not mean we still don't have margin for 16 inelastic behavior which affects both the forcing 17 function and the behavior, right?
18 MR. KNIGHT 4 That is certainly true.
19 If you remember, for instance, taking North 20 Anna, there were some items.
One that comes to mind are 21 the hold-down pumps that blew the pumps that were 22 designed to normal design numbers, like 1.01.
23 I guess I should stop in mid-sentence there.
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24 To me we were answering a somewhat different question 25 there, however.
There we were saying, all righ t.
This O
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1 plant, were we to specify seismic motion for this plant 2
today, it would be somewhat greater than that for which 3
it was designed.
So let us look at the margins that are 7-V) 4 a vailable and discern whether or not we would have a 5
satisfactory situation today.
And in that context, just 6
saying the margin is 1.0 many would argue is indicative 7
of a to tally sa tisf actory answer.
You met your' goal, 8
and it would be satisfactory if you designed ridht up to 9
the line today, and you would have made it.
10 The more recent letters I believe are asking a 11 significantly different question.
That is, regardless 12 of where your SSE f alls with regard to our best 13 judgment, our best exercise of technology today, we 14 ought to be able to demonstrate that an event of more 15 severity, albeit less probable, could still be, 16 tolerated.
It is that somewhat nebulous requirement 17 that I think is going to give us a great deal of 18 dif ficulty in trying to implement it.
You are in the i
19 posture again of going to the utility and saying, well 20 you've got to do something better without being able to l
l 21 explicitly tell them what their goal is.
22 MR. BENDERa I think that might be a 23 misinterpretation, Jim.
I think what we have been t ()
24 asking is if it happened, here are the margins, as
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25 opposed to saying that you really need to show that O
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1 margins exist.
But there surely is some reserve in 2
these structures, and even though we may not ever expect 3
to use them, knowing that they are there increases our O
4 confidence in the event of mistakes and other things.
5 MR. SIESS:
Maybe we need a research project 6
to find out what the seismic margins are.
7 NR. BENDER:
In.fset, the Livermore people 8
tried to do it, and so far I haven 't seen any good 9
results from it.
I saw some very bed use of it 10 yesterday when TVA made their presentation.
11 MR. KNIGHTS I am not sure you were here in 12 the room or not.
I said some of the key words that led 13 us perhaps to think in error, but that any needed 14 modifications be made, given some time period.
l 15 MR. OKRENT:
Well, let 's talk a little bit 16 more about the point because it is sort of central to 17 the whole discussion.
It may De that when you look in 18 detail at a specific plant in all aspects with regard to 19 accomplishing safe shutdown -- and that means not 20 getting into a situation where you have a LOCA as 21 well -- in all aspects, given earthquakes of 22 increasingly low probability, the plant can accommodate l
23 this.
It may in f act have inelastic distortion, and l ()
24 some components may be ir retrievably lost that you don't 25 need for this purpose, and so forth and so on.
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1 But just to say that it is.!esigned for some 2
safe shutdown earthquake, if the probability of the safe 3
shutdown earthquake lies in the range of, let's say, one 4
1000, one to 10,000 per yea r, that may not be enough if 5
you haven't taken the de tailed look.
6 Now, after you take that detailed look you may 7
decide that the plant in all respects is adequate for 8
less probable earthquakes although there will be 9
nonelastic behavior.
Ckay.
10 On the other hand, in some cases you may say 11 it is good in 98 percent but there are some things, 12 based on the existing information, we can't say how it 13 will be 20 years from now when it has aged and so N
14 forth.
So we are going to need some research or some 15 special testing on similar things in order to see what 16 we can estimate, let's say.
17 Or it may also be that in some cases you say 18 these things are pretty much going at their limit now, 19 and if we want to have them capable of taking a somewhat t
20 less likely ea rthquake, we need the modification.
21 Now, up to now, so f ar as I am aware, there 22 has not been a really systematic look.
Things have been l
23 sampled.
And usually when structures are sampled, if
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24 they were seismic class 1 originally, they seem to come l
l 25 out in good shape.
I don't recall seeing any.
You have
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1 a liquefaction question you talk about, they increase 2
it, but aside from that, everything I have seen seems to 3
indicate structures turn out pretty good even without CJ 3
4 inelastic.
5 So the guess is on my part that it is other 6
areas.
And in fact, you have found that there have been 7
some parts of plants that were supposed to be seismic 8
classified, that the cable trays and so forth weren't so 9
well supported and so forth and so on.
But there has 10 not been the systematic look to see that you can with a 11 sufficient degree of assurance get the sufficiently 12 small contribution to risk.
13 I could foresee an approach to this that 14 didn't start out trying t'o specify in detail what had to 15 he met for a specific plant.
It might be that one 16 approach was to have someone write down how would I do l
17 this if I were going to do it?
What are the practical 18 ways of trying to include inelastic effects and so 19 forth?
What kinds of criteria do I think are 20 appropriate to measure against?
What do I expect to be 21 the things for which I have sources of information, and 22 where do I know already that I am limited, and sort of, 23 you migh t say, lay out a kind of a proposed detailed
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24 work scope for the thing for people to look at and 25 reflect upon which would be, from my point of view, a O
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1 thinking kind of effort.
2 And it should not start with a detailed 3
prescription.
4 MR. SHEWHON:
Dave, that sounds like it might 5
be a good research program, but I have difficulty seeing 6
it as a licensing criterion.
7 MR. OKRENT:
Well, it is the sort of thing 8
that the people who are expert in the struc tural seismic 9
area certainly, like, well, some of the consulting 10 companies -- I won't name any single one or two because 11 I might leave one or two out and their feelings would be 12 hurt -- they could in fact lay it out for the structural 13 part.
They might need some assistance from people who i
14 think about moving parts and electrical systems and so 15 forth, but I think what I am suggesting is the sort of l
16 thing that e xperienced people in the field could 17 propose.
18 MR. SHEWMON:
The experience'd people wrote the 19 code, and you said you thought the structures part was 20 pretty good where they looked at it before.
21 MR. OKRENT:
What I said was when they looked 22 at structures that were designed for a certain 23 earthquake and examined how much larger an earthquake
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24 would the structures take before they thought there was 25 a reasonable likelihood of failure.
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1 MR. SHEWMON:
Not failure going plastic, 2
usually.
It is inelastic.
3 MR. OKRENT:
Failure to serve its function.
O 4
MR. SHEWMON Not just tha t going plastic in 5
some small part.
That is not failing to perform its 6
function at all.
That is where the margin comes in.
7 MR. OKRENT:
There are different looks people 8
hav e taken, Paul.
I was going to taka a different 9
measure, which was f ailure to serve its f unction, and 10 then one usually gets a very considerable capability in 11 the structures.
12 When I see analyses done, it is more than 13 just --
14 MR. SIESS :
The regulations for SSE simply 15 require tha t they remain functional.
16 MR. KNIGHT:
That's correct, yes.
17 MR. SIESS4 It does not require they stay 18 within any specified allouable stresses at all.
19 MR. SHEWMON4 Jim is beyond this fire drill 20 which he does sort of yearly now, or more f requently, 21 and what we get is ratios of stresses, at least with 22 regard to structural things, and that is usually where 23 it goes plastic, I suspect, or gets up to some yield
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24 stress.
25 MR. KNIGHT:
And the reason is one I think of O
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1ot pure ecoao=1c=-
rae taro== tioa ta t 1-2 available to design houses, if you will, is some 3
calculated stress in accordance with the code, and you 4
vant to compare that with the code limit.
That type of 5
information you can-extract rather readily.
6 If you are getting into a situation where you 7
are now going to say let's go back and look at this 8
beast on what amounts to a really different basis, and 9
if you want to start with structures, you pretty well 10 have to turn an entirely diff erent group of people. loose 11 to sit down and start doing a far more sophisticated 12 analysis.
And I am sure that the Staff and the group 4
13 could argue about the nuances of the methods of analysis-0 14 for some extended period.
15 16 17 18 19 20 21 22 23 24 25 O
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1 MR. BENDER:
The people who were at Livermore 2
last year did an investigation of the design of the 3
plant.
They even interpreted it for more severe
}
4 earthquakes.
I really didn't have a great deal of 5
enthusiasm for that progrhm, but having done it, they 6
did develop some methodolooy for looking at margins.
7 It seems to me that you might at least look to 8
see whether they learned enough in that to provide some 9
way of making such assessments.
I don 't disagree with to Paul.
I think we may be pushing the requirements more i
11 than we need to.
But nevertheless, if the methodology 12 is there we may as well be prepared to use it.
13 I am more concerned about the fact that we O
14 vill find some mistakes tha t have to be looked at, we'll 15 see an earthquake that is worse than the SSE.
But 16 that's my perspective.
17 MB. KNIGHTS I think, just to play on that for 18 a moment, if I may, I know one of the thoughts often 19 voiced on the Staff is that there seem to be two lines 20 of thought about gaining margin.
One is to increase the 21 basic level of the seismic input, and in some 22 discussions they say, well, we found things wrong here 23 and things wrong there, and these things aren't as good
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24 as you think they are.
25 But if that is the problem, then you have to O
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1 go after the specific problems.
You have to have better 2
design quality and control.-
You have to have better 3
perhaps field inspection, if those things a re 4
occurring.
And increasing the level of the seismic 5
input for the design basis really isn't going to ask the 6
question.
7 Just one other point that came to my mind 8
while I was speaking, I know I do it myself, I've 9
probably done it before the Committee a number of times,
~
10 but I think most of us immediately grapple with the 11 basic structures, because it's pretty straightforward 12 and it's fairly easy to get a handle on and discuss.
13 In doing such a study, I am more concerned 14 with the situation where a plant has, say, a number of 15 pieces of equipment that have been tested to a certain 16 level.
They have that.
That's all they have, is that 17 test record.
j 18 To go back now and say, well, oka y, but what i
19 could it really take -- well, I'm by no means demeaning 20 the thought or demeaning the effort.
I'm just kind of 21 thinking out loud about the problem.
You run into a 22 number of practicalities.
23 Well, you could argue, you could take similar
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24 pieces of equipmente In many cases similar equipment 25 doesn't exist.
You can empanel a group of people to O
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1 look at it and speculate on what it could take, given 2
what it was tested a t.
I doubt that would be a 3
satisfactory solution.
4 I think there was an effort to discern after 5
the fact the fragility of this equipment.
That is a 6
significant undertaking and it has problems that are 7
quite a bit more severe than, say, going back and 8
re-analyzing a structure, albeit that can be a pretty 9
severe task in itself.
But at least it is doable.
10 I think we can dig out the capability of a 11 piece of equipment to take some more severe -- something 12 more severe than you can discern from what was already 13 done.-
Clearly, we can look at the actual test that was Akl 14 perform and in most cases discern that the input was in 15 fact more severe than the basic design requirement, just 16 because of the prs ticalities of testing.
17 But to go above that I think is a problem that 18 has very significant import from the standpoint of the 19 plant and the money involved.
One of the things I 20 intended to finish up with today is that I think we are 21 in an area thst starts raising rather significant policy 22 questions.
23 (Slide.)
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24 By no maans do I intend to stand here today 25 and talk about the direct relationship or implications, l
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1 but there are these things nappening.
The Commission 2
has spoken on the safety goal.
We in the Staff are very 3
properly charged with giving very much deeper O
4 consideration to what may or may not be backfit and how 5
that should be handled in accordance with the 6
regulations.
7 And the question arises, if we are looking at 8
seismic events in excess of those that on* develops by 9
meeting the regulations, and we insist that they do meet 10 the regulations when the plant is licensed, that raises 11 the question of whether what we 're talking about is even 12 de facto modifications.
13 MB. OKRENTa Let's look a little bit about 14 this question of meeting the regulations for the 15 moment.
Let's think a little bit historically in this 16 regard.
17 There was a time, I believe, when the Staff 18 and the Applicants and maybe the ACRS looked upon the 19 SSE as being a sufficiently low probability event that 20 if the plant just met it, as it were, you were providing 21 a sufficient level of safety in that regard, met it with 22 not too much margin, and so forth.
23 I think it is not too hard to go back into
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24 some of the memoranda and so forth that were written.
25 The first time the Staff spoke about maybe the SEE ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE S.W., WASHINGTON, D.C. 20024 (202) 554-2345
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1 through one plant was 10 per year, and the Committee 2
took enough notice to write a memo to whoever was the 3
Director of Regula tion at the time asking him to tell us 4
more about it.
5 In fact, I just happened, in reading the Grand 6
Gulf construction permit letter in' preparation for a 7
Subcommittee meeting coming up not too long after this 8
one, I find that by chance I appended a remark to that 9
CP letter saying I didn't think the SSE for Grand Gulf
-6
-7 10 was a 10 or 10 event, which presumably some 11 people were saying back in 1974.
12 MR. KNIGHT:
Right again.
13 MR. OKRENTs Or I wouldn't have appended the 14 comment.
15 Now we're not talking about that frequency any 16 more.
It is somehow the same SSE.
So you could say, 17 well, it 's the same reg ula tion.
But now the Staff comes
-3 18 in, when they provide a rough estimate they say 10
-4, and in the next breath they say, but there are marg 19 10 ins that give us an acceptably low risk.
They may not use 20 that kind of ph rase, but that's the implication of it.
21 22 In fact, there may well be these margins, but 23 if.you haven't really assured yourself everything you
)
24 need with regard to safe shutdown then you may be l
25
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t ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE-, S W., WASHINGTON, D.C. 20024 (202) 554 2345
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1 relying more on luck than is appropriate.
2 The point I'm trying to make is, this question 3
of meeting the regulations is a little bit fuzzy.
One's 4
knowledge of what the frequency of the SSE, or at least 5
one's opinion on the frequency of the SSE, because I'm 6
not sure it's knowledge, has shifted.
And so there is 7
acre inportance in being assured tha t the margins exist 8
in everything you need for safe shutdown.
9 Now, I agree with you, you can ' t just analyze 10 big structures, things that have been (;ualified by 11 shaking.
There may be some difficult situations, and in 12 some cases it may not even be practical to devise a 13 simulator kind of test.
()
14 That doesn't mean, I think, that one shouldn't 15 know that there are these possible situations and let 16 expert opinion look at this and sa y, yes, I think this 17 is okay, or no, in the same way you looked at things 18 related to fires and made judgments and in some cases 19 made very expensive fixes being required, even though 20 there were differing opinions.
21 There is nothing truly quantitative in that 22 judgment as to what you must or must not do with regard 23 to a fire.
That doesn't prove anything about the
()
24 other.
You've had more than one serious one.
25 So I think myself that it is relevant to look ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345
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I to understand the problem to see in which areas you are 2
in pretty good shape, in which areas you can examine 3
yourself theoretically, which areas perhaps will lend 4
themselves to related kinds of experiments, or maybe 5
where experiments have been done, and which areas are 6
grey areas, and narrow the things that can be done.
7 My intuition is there are going to be small 8
pieces here and there that may at least reflect areas of 9
concern, and then ma ybe by looking at the experiences 10 the Japanese are having on their shakers and so forth, 11 this will tell you you need not be concerned on tha t 12 one, but you need to go back on another one.
13 I don't think we know the answer today, but if 14 you don 't look you won 't know it in three years, 15 either.
16 MR. KNIGHT:
Well, I think clearly, a t least 17 from my point of view, as you say, myself when I first 18 f ined the Staff and got involved and started asking 19 around, what is the level of likelihood, or whatever 20 word we used a t the time, as f ar as this big earthquake
-6 21 goes, the numbers you mentioned, 10 were bandled 22 about.
23 I guess from our point of view, however, those
()
24 were secondary to the fact that there was a regulation 25 in placa and we met the regulation.
I suppose the next ALDERSON REPORTING COMPANY,INC, 400 VIRGINtA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345
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1 step was to go beyond that, perhaps, to the history of 2
that regulation and how it was developed.
I am not 3
reslly competent to speak to that.
4 Yes, sir?
5 MR. SHEWMON:
We are talking here as if the 6
SSE is an immutable number that indeed somebody, people, 7
can agree on.
Yet I've also sa t in this meeting and 8
heard a fair discussion about what the SSE should be, 9
and it changes from year to year.
And just as with 10 Grand Gulf, Dade can go back and say, well, he didn't 11 think that was good enough.
But also with Grand Gulf, 12 as some of the others, it gets increased over the last 13 several years.
14 Is there anything in the Staff's thinking when 15 they consult their ouija board or their crystal ball, or 16 wherever they get their SSE's, to say that it should 17 increase to he something proportional to this one in 4
18 10 or one in 10 ?
It seems to me that 's a t least 19 as good a part of this question as beating on the 20 structure or whatever else.
l 21 MR. KNIGHTa I'd like to ask Leon Reiter to 22 comment.
23 MR. REITERa My name is Leon Reiter, f rom the
()
24 Staff.
l l
25 There really is no such criteria, some number O
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()
1 that we look at.
I think the kind of numbers that have 2
been presented in the past have been nrmbers which asked 3
what are estimates of the SSE, these are usually made 4
from a quick survey of the studies.
We typically say
-3 4
5 and they typically come out 10 10
- However, 6
this is not to say that this is meant to be some sort of 7
rigorously defined factor which should be used in some 8
calculation to determine adequacy.
9 MR. SHEWHON:
To what extent do you find that 10 the SSE does define adequacy?
If MR. REITER:
Appendix A.
There is no 12 probability in Appendix A.
Appendix A refers to the 13 operating basis earthquake.
14 MR. SHEWMON:
I have the other side.
The 15 plant must be designed to take an SSE and close down 16 safely.
But I'm not getting what criteria the Staff 17 uses in selecting the value of the SSE.
Could you 18 briefly help me on that?
l 19 MR. REITER:
Yes.
The SSE is generally 20 defined as an earthquake based on an evalua tion of the 21 maximum earthquake potential.
With respect to the 22 western United States or areas where there are clearly 23 defined faults or clearly defined structures, then an
()
24 estimate is made of what we think the estimated maximum 25 potential of those structures are.
OU l'
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1 ER. SHEWMON:
Yes, I've been through that.
Go 2
ahead.
r' 3
MR. REITER:
In the eastern United States,
(_)
4 where these structures -- we cannot identify these 5
particular structures, then we take what is called the 6
tectonic province approacn.
There we take large areas 7
of usually similar geometry, and the regulation says we 8
take the largest historical event within that, and we 9
assume that it occurred near the site.
No mention is 10 made of probability there.
11 MR, SHEWMON:
So you come up with what you 12 think is the most reasonable maximum earthquake, and 13 another group of experts comes up and says, gee, that 14 will happan at least every 1,000 years, not once every 15 million years; is that right?
16 MR. REITER:
No.
What comes up is a 17 discussion as to wha t is the appropria te tectonic 18 province that one considers.
In other words, there may to be a discussion as to what the boundaries of that 20 province are.
21 Very often utilities will define a very small 22 province.
We say, no, that 's not the appropriate 23 province; we think the larger province would be more
()
24 a pp rop ria te.
In d oing that, once you increase the size 25 of the province or change the size of the province, then O
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1 a different earthquake or a larger earthquake will 2
become the controlling earthquake.
3 There is, however, a stipulation in the 4
appendix that came in as a result of controversy over 5
Seabrook that s,ays, if seismological or geological data 6
warrant it, then we can have an SSE which is larger tnan 7
the historical earthquake in that particular location if 8
it is felt -- these are the words of the regulation.
If 9
it's warranted in that case, we have something larger.
10 In some cases we have that, we have done that.
11 But there is no specific criteria which links 12 the SSE to some probabilistic number.
13 MR. SIESS:
I was going to say, you are right, O
14 there is nothing in the regulations that brings in 15 probabilities, but probabilities have been brought in by 16 us and others as we began to look at PRA.
And when you 17 bring the probabilities in, it turns out that it is not 18 very terribly low.
t 19 HR. REITER.
Dr. Siess, again I think those l
20 probabilities -- that we've been asked these questions 21 for years already.
I go back to an old discussion we
-5 l
22 had with Dr. Okrent.
The 10 number as far as I know
(
23 came in with the Greenwood plant.
But since then, in
()
24 all my six years with the agency, in talking to all the 25 people, I'm not aware of that kind of a number creeping
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1 in again.
'l Again, that number was not-- the numbers
-3
-4 3
10 or 10 were meant to be some sort of rigorous 4
criteria which are applied to some measure of safety.
5 They were meant to say, what is the survey of what some 6
people -- what is opinion, I think Dr. Okrent put it, as 7
to what the SSE, the return period of the SSE would be.
8 That at various times has been applied to the size of.
9 the earthquake, density, magnitude, peak accelera tion, 10 spectrum, the whole range of things.
I 11 So I think the Staff has expressed some 12 concern in attempting to use those rather vague answers 13 in some sort of a rigorous way to define acceptability 14 or lack of acceptability.
15 MR. SIESS4 You'll find the same problem when 16 you try to use PRA in any rigorous way.
But let's face 17 it, if=I believed that the SSE was a true threshold j
18 value and if beyond the SSE something disastrous was 19 going to happen to my plant, and if I take the best 20 evidence I can find, which isn't all that good, that 21 tells me that the probability of exceeding the SSE is as 22 much as one in 1,000, then I think I've got a right to 73 be concerned.
()
24 I don't think it is a real threshold value for 25 an awful lot of the plant.
I know it isn't for O
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1 structures, because we have a lot of experience with 2
structures that hsve been designed for these 3
I can't say the same thing for every 4
single thing in tha t plant and there are an awf ul lot of 5
things in that p1snt tha t are going to get shook.
6 Now, the regulations don't help us here.
The 7
regulations assumed that the SSE was the maximum 8
earthquake potential, and most of us I think felt in 9
those days tha t that was the maximum earthquake you 10 could have.
11 NR. OKRENT No, I don't think ro.
And in 12 fact, Leon, I don't have a copy of Appendix A as it was 13 adopted in '73 handy, but ny recollection is that for O
14 the eastern U.S.
the intent was to look at the 15 historica1' record, and also to allow for the limited 16 history available.
17 HR. SIESS:
Right.
18 ER. OKRENT And I think words of that sort
(
19 were included in Appendix A as adopted.
20 I would argue that in the phrase, whatever it 21 was, that said the limited history, one could equally 22 well within the context of those words have said, 23 wha tever is the historical maximum intensity in my
()
24 tectonic province, I will add one-quarter of an ami, 25 I'll add one mmi, I'll add two mai, to allow for this.
O ALDERSON REPORTING COMPANY,INC.
400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345
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And one could have set up a recipe of this sort and ss 2
still have been within the words of Appendix A.
3 MR. SIESS:
And you might have been still one 4
in 1,000, Dave.
5 MR. OKRENT So I don't think it's anything as 6
was originall written that led one down a clearly 7
defined path to the SSE, even given that we had some way 8
of defining what the provinces are, which is not always 9
easy.
to ER. REITER:
Excuse me, Dr. Okren t.
I think 11 rou would be correct if the only sta tement in Appendix A 12 was what you had said, that we're looking at the eastern 13 United States, it's limited to seismicity.
- However, 14 there is a very prescriptive part in Appendix A that 15 comes after, which says how you deal with the tectonic 16 province.
17 Do you take the maximum intensity, maximum 18 earthquake that will occur -- it doesn't say you take it 19 and then you weigh it according to the historical 20 seismicity.
It didn't sa y it a t that time.
21 So I think there was -- I think you're righ t 22 that the feelino in the past, there was some of this 23 feeling th a t the eastern United States there is a
()
24 vagueness limit to historical data, and the general 25 feeling -- and this I get f rom conversations with people O
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1 at that time -- vis the way to deal with this was to 2
take these large tectonic provinces and assume that the 3
largest earthquake in the large tectonic province is 4
going to occur near the site.
5 I think that was the attempt to deal with the 6
problem of limited historical data and a lack of 7
knowledge in the eastern United States.
Once we had 8
that, it was fairly prescriptive where it would go, 9
except, as I said, in extenuating circumstances and in 10 the Seabrook amendment which was added later.
11 MR. OKRENTs The words are in the earlier 12 version which permitted one to choose an earthquake 13 larger than what was historical in the province.
I 14 think at that time his thinking was that this was a
-3 15 really improbable earthquake compared to 10 per 16 year.
-5 17 We have the Greenwood case where 10 18 there was a little element of shock.
So large? That was 19 a shock, not so small.
l 20 So it was in that context that this approach i
21 was developed and was being used, let's say, in the-late 1
22 sixties and early seventies.
And with the change in 23 thinking, and also with the tendency to take this same
()
24 recipe but move to smaller and smaller provinces, and 25 you take the 100 square miles or 500 square miles in the O
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middle of the state and call that a province, sort of.
[}
2 We have departed from the basic approach to 3
risk, if I can call it that, that people thought they O
4 had in the early seventies, when Appendix A was 5
formulated and adopted, or formulated and second and 6
third drafts were adopted.
7 MR. KNIGHT:
I guess in part what we are 8
seeing here f.s the interplay or the effect of practice.
9 I think it's something we ought to keep in mind, that to regardless, if you will, of the background, in practice, 11 at least in my view, utilities have had every right to 12 presume that if they got their construction permit and 13 some level of -- for seismic design was stipulated, 14 agreed upon, developed, whatever word you want to use, 15 they now had a fixed design basis they could proceed to 16 design to and they would be done.
17 In some ways I know I was involved in numerous 18 arguments with the utilities tyself, when plants would 19 come before you and say, well, if we were to do it today 20 it would be somewhat different.
We have often argued 21 that, well, this isn't really backfit; it's just better 22 technology or better understanding of the problems.
23 MR. SIESS:
Jim, how can you make -- you're
()
24 making a distinction be tween, let 's sa y, some of the 25 plants we have recently reviewed, where these questions OV ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345
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1 have come up, and some of the older plants, the SEP 2
plants.
I'm not sure.
Were only those before Appendix 3
A?
4 MR. KNIGHT:
I believe so.
Leon has been 5
immersed in that program.
6 MR. REITER:
Certainly phase one and phase J
7 t w'o.
8 MR. SIESS:
Those were before Appendix A?
9 MR. OKRENTs Before Appendix A was adopted 10 there was a draft version of Appendix A.
11 MR. SIESSs There were six draft versions that 12 vent around for as long as the GDC's.
We went around 13 with this same kind of a problem on tornadoes.
We
(')
l 14 didn't require any tornado west of the Rocky Mountains, 15 because we thought the probability was so low that we 16 didn't need to worry about it.
17 And at some point somebody pointed out that 18 the probability of a tornado on the West Coast was one 19 in 1,000 or something like that, which was about the 20 same probability -- the probability was about the same 21 as it was east of the Rocky Mountains.
But the l
tornadoes just couldn't be as big or as strong.
And 22 23 that changed our perception of designing for tornado 24 loadings on the West Coast, and we made a change in the 25 rules, right?
l l
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~
1 HR. KNIGHT:
I guess I have to stop there (d'
2 wasn't, if I remember correctly -- and someone please 3
correct me if I'm wrong.
Are you using the word " rule" O
4 in terms of the regulation, as opposed to the reg 5
guides?
6 MR. SIESSr Whatever applied.
It might be a 7
reg guide, but it's applied like a rule, or a standard 8
review plan or somathing like that.
9 MR. KNIGHTS I think tornadoes vare covered to under GDC.
11 MR. SIESSs The probability of a safe shutdown 12 earthquaka has changed with time once we began to think 13 probabilistically, but we haven't made any changes in
(
14 the regulations.
And that is your problem now.
We are 15 trying to get people to look at consequences of an 16 earthquake greater than an SSE, and you are having 17 difficulty finding a regulatory framework in which to do 18 it, right?
Is that the essence of the problem?
19 HR. KNIGHTS Yes, it is, indeed.
- Actually, 20 there are two.
One is the framework and the other is 21 the criteria that we would ask them to meet.
22 HR. SIESSa Now, NUREG/CR-0098 only applies to 23 structures?
I believe it does.
It does include some --
(])
24 it does include some -- no, the inelastic behavior stuff 25 is all based on structures.
That was Newmark and Hall,
()
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(~h 1
and I can't believe they would have much to say about V
2 components.
I would include piping.
3 MR. KNIGHTS That was my hesitation.
I was O
4 thinking of piping as a structure.
5 MR. SIESS4 Functional survival of electrical 6
components and pumps and valves.
But that would 7
certainly be an adequate guideline if you really had to 8
do it, and that doesn't require
-- that permits 9
inelastic analysis.
It also permits modified spectrum, 10 doesn't it?
11 MR. KNIGHT:
Yes.
12 MR. SIESS4 But as Dave has indicated, and I 13 think we sort of agree, there is not likely to be much
)
14 problem in the structures.
It's more likely to be in 15 the equipment.
16 MR. KNIGHT:
I think if we're going to venture 17 into this type of thing, I would very much hope we would 18 give it enough prior thought to have some uniform 19 approach.
And as I said, I freely stand here and admit 20 that the Staff is very much at ends as to what that 21 approach ought to be.
And I am most reluctant to charge 22 off and try -- you know, try one approach here and one 23 approach there, and in essence say, wait 'til we see the
()
24 results and we'll decide whether it's good enough or 25 not.
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1 That's an extremely difficult posture, I 2
think, to put both the Staff and the utility in.
3 MR. BENDER:
Jim, have you looked at the stuff 4
Livermore did last year?
5 MR. KNIGHT:
Certainly.
I can't say I 6
personally have been through it all, but I am aware and 7
the Staf f has been tracking it.
8 MR. BENDER:
I guess the point I'm trying to 9
make is, if there was any use in that work it was in 10 developing some kind of methodology that might be 11 applied.
12 MR. OKBENT4 Did they incluae inelastic 13 deformation in their analysis?
14 MR. BENDER 4 No, they didn 't.
But at least 15 they 9howed how to look at certain kinds of margins.
16 MR. KNIGHTa We need to go back and look a 17 little harder there.
18 MR. SIESS:
Is this part of the SSMRP?
19 MR. BENDER:
It's in the other part, the load 20 combination.
l 21 MR. KNIGHT:
Load combination work, right.
22 MR. SIESS:
The Zion PR.4 look s a t the effect 23 of earthquakes on components, but I don't know to what
()
24 extent their fragilities were based on qualifica tion or 25 whatever.
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1 MR. KNIGHT:
My brief look at some of tha t 2
says they primarily went out and looked under some of 3
the information developed under SSERP, 4
MR. SIESS If I look at Zion, where it takes 5
three times the SSE to start failing components, I get a 6
lot of comfort.
But I don 't ha ve the slightest idea of 7
how much of that to believe.
8 MR. OKRENT It's expert opinion, mostly.
9 MR. KNIGHT:
Mostly.
You've got a number of 10 tests that were perf ormed on off-the-shelf equipmen t, 11 that were perf ormed for some of the original missile 12 projects, and I would think we could probably have a lot 13 of debate about exactly how to apply that.
14 I personally feel that it's indicative of the 15 performance of a class of equipment, but trying to say 16 that I can now take that information and apply it to 17 component Z at the Perry plant --
18 MR. BENDERS Well, the least you could do, I 19 would think, would be to do a couple of typical examples 20 to see how something would be treated, if you were l
21 concerned about trying to show that you had more margin i
22 than was originally in the design, and develop some 23 understanding of what the next thing is you'll take into O
24 ac=ount.
25 It's not looking at it as though you had made ALDERSoN REPORTING COMPANY,INC, 400 VRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345 t
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1 a design mistake, you hadn't put as much strength in the 2
insulation as you thought.
What would you do to aph 3
accommodate it, other than beef it up?
You might be 4DE 4
able to show that the next stage of degradation wasn't 5
all that bad.
6 I'm not sure that you would come to the right 7
conclusion, but if you haven't tried it there's no sense.
8 in th rowing up your hands.
9 MR. KNIGHT:
I'm certainly not -- on behalf of 10 the Staff, we're not throwing up our hands.
We are 11 openly trying to say that we see problems in trying to 12 get on with thic recommendation, and we are certainly 13 searching for all the advice we can get as to what was O
14 the point of the recommendation and what is viewed as an 15 adequate response, so we can be on target as much as 16 possible.
17 MR. BENDER:
I think I would like -- I'm just 18 speaking for.-myself at the moment, but the letters 19 you 've had so f ar are not intended, at least in my mind, 1
20 to have represented a generalized request to look a t 21 designs beyond the SSE specified.
I think the questions 22 were raised specifically that related to a particular l
l 23 feature of a particular plant.
24 I don't see any reason to say tha t it should i
25 be a generalized requirement.
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1 MB. SIESSs There are enough plants to be 2
pre tty general.
3 Jim, will a plant like Perry have to do a 4
PRA?
5 MR. KNIGHTS I think so.
Do the folk s f rom 6
Mr. Thadani's shop know offhand?
7 MR. BUSLIKs Arthur Buslik from the 8
reliability risk assessment branch.
I don't really knov 9
whether this applies to near-term operating license 10 plants, whether it is required.
11 MR. SIESSs Let's assume for a minute they 12 would be.
Some plants are.
Would they have to include 13 seismic in there, or are external events excluded?
Akl 14 MR. OKRENTs It is being done in some and not 15 in others.
16 MR. SIESS:
I asked if it was required,
/
17 because if they have to do a PR A and they have to do 18 seismic, do they know enough to do that if they don't 19 know enough to do what you think we ask for?
And if 20 they do know enough, it seems to me you would satisf y a 21 lot of the questions.
i 22 MR. KNIGHTS Maybe to go back to the first of 23 your questions, I don't believe that, at least under
/O
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24 current recommendations, that they would be asked to do i
25 external events.
The question of whether or not they're ALDERSoN REPORTING COMPANY,INC, I
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1 going to do something notwithstanding, if they do I 2
don't believe it'll includa external events at this 3
point.
At least current thinking says they would not do 4
external events.
5 Part of the reason for not doing external 6
events, although the principal reason there, I would say 7
-- and I'll look for Leon to correct me -- is the 8
difficulty in handling the seismological aspects.
But 9
certainly, in all the discussions we've had internal to 10 the Staff, the other side of the question -- in other 11 words, how do you get these fragilities, how good are -
12 they, how do you use them -- was part and parcel of 13 deciding that the technology really isn't going to apply O
14 there.
15 It's being done, certainly.
Some people are 16 certainly worthy of taking a shot at it.
How it would 17 be viewed once it's done and how you use it is an 18 entirely different question.
I think we're all 19 grappling with that.
20 HR. SIESS:
I guess.
Do the levels of 21 ignorance vary that grea tly ?
22 HR. KNIGHT:
I'm sorry?
23 MR. SIESS:
I wonder if our levels of
()
24 ignorance vary that greatly.
25 MR. OKRENT:
Could I offer a couple of ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W WASHINGTON, D.C. 20024 (202) 554 2345
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1 comments on what you just said?
My impression is that 2
right now in the NREP outline of initiating events, 3
external events that are not included, I think that is a
)
4 f undamental error that the staff will be making if it 5
stays along that path.
6 It may have been reasonable when they began 7
IREP.
It was already unreasonable by the time they 8
finished IREP, because in fact other groups are 9
including external events.
They are turning out to be 10 potentially as important as any others, according to the 11 results more important.
12 And I think it's not going to be an untenable 13 position for the NRC to take.
They may take it, but I O
14 think they will find if they stay with it that they will 15 have been wrona.
16 The other comment I would like to make is, if 17 there's a Commission safety goal which in f act one wants 18 to use to measure for backfitting and one wants to use 19 in considering ATWS or things like that in some suitable 20 way, it is also, it seems to me, the same policy which 21 one has to ask himself about when you talk about 22 seismic.
23 Again, I think the Commission will find itself
()
24 at some point in an untenable position if it sort of 25 puts its head in the sand with regard to the seismic O
ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345
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(])
1 issue and says, we have a regulation, and so forth.
My 2
own opinion is it's better to have the information to at 3
least know where there may be spots that are not very 4
well known and say, okay, our overall judgcent is these 5
are not likely to be too important, than just to retreat 6
to a legalistic one.
7 I think the legalistic one will become feet of 8
clay for some reason or another at some point.
9 HR. MARK Dave, I guess I don't think of what to Knight has been describing for us as an attempt to 11 retreat and hide behind the verbiage of legalistic --
12 HRe OKRENT:
I don't want him to be pushed 13 into that.
~
14 MR. MARK:
But I think I at least would feel 15 it is quite necessary to agree with.him that what we 16 have put, and he has shown on the slide, in our letters 17 on Wolf Creek and Perry and possibly Midland 2 is really 10 very vague advice.
It was perhaps vague because we l
19 didn't exactly know what we thought should be there.
It 20 sort of came up in part, if not entirely, from the 21 coming to feel that the SSE was not a highly improbable 22 event.
It may ha as probable as core melt or something 23 else, and it is proper to take some account of that in 24 some form.
25 But we did n ' t write any words which gave a l
l ALDERSON REPORTING COMPANY,INC, l
400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345
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1 recipe that could be followed in licensing.
The needed 2
modifica tions -- well, needed for what?
Well, for a 3
con siderably larger carthquake, albeit less frequent.
4 Well, how much larger?
5 Now, we have not talked, remember, about how 6
we might in any way quantify those words.
I think it 7
might be possible, but I don't believe we have done ite 8
If you step up the ami by one unit, what do you do to 9
the, oh, more or less standard curves of frequency if 10 you go from ami 8 to ami 9?
11 MR. REITERs In the eastern United States, you 12 go into the realm of almost f antasy for the most part.
13 But those particular numbers for the eastern part of the 14 United States are almost in the realm of fantasy.
But 15 you 're saying if you go up one unit of intensity?
16 MR. MARK That was my question.
17 MR. REITERa Studies in the East indica te tha t 18 if we double the acceleration -- and that's what happens 19 when we increase the intensity by one --
20 MR. MARK:
One unit of intensity?
21 MR. REITERa Doubles the accelera tion, doubles 22 the intensity.
It seems to lead to an increase or a 23 decrease in risk by a factor of five.
()
24 TR. MARK:
You mean decrease in frequency.
25 You double the acceleration with the unit and you O
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()
1 decrease the frequency by a' factor of five.
2 MR. REITER:
Righ t.
Tha t is a ballpark 3
estimate.
But again, it can't be applied to all ends of 4
the intensity scale.
5 Dr. Mark, there's something here I feel I must 6
say.
I think as a seismologist on the Staff, we all 7
understand the concern about the SSE.
I think what 8
concerns us is, one, the use of these vague 9
protabilistic numbers that have been supplied in the 10 past to somehov assess the adequacy of the SSE on the 11 one hands on the other hand to come up with using
-5
-6 12 numbers like 10 10 to describe some future 13 goal which we have to arrive at.
14 We have enough problems in trying to determine 15 what the level of the SSE is without trying to determine 16 rigorously what a 100,000 million year earthquake is.
17 It is just beyond the state of the art.
I think we 're 18 perhaps fooling ourselves with these numbers that we 19 gain some safety from them.
20 MR. MARKS I'm aware of the problems.
I'm 21 also awa re that if you told me you thought you knew the
-5 22 number exactly for 10 I wouldn't believe you 23 anyway.
()
24 (Laughter.)
25 MR. MARK:
Perhaps that's only because I've ALDERSOM REPORTING COMPANY,INC, 400 VIRGINIA A%E., S.W., WASHINGTON, D.C. 20024 (202) 554-2345
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1 heard you talking about the problems.
2 However, I think that we have some work to do, 3
Dave, as well as suggestions to make, before we can make 4
an operable suggestion, one that somebody could take 5
back to Wolf Creek and say, this is the level to which 6
we ask you to push this.
7 MR. BENDER:
Well, I agree with Carson, 8
really, that we do need more guidance than we have 9
given.
And I do not want to set the thoughts I have as to specific guidance I would give today, but there a re 11 approaches that could be considered.
12 First of all, some f ractions of these plants 13 that we made comments about, if you were to reevaluate 14 the earthquake today you would probably assign a 15 different earthquake than you did when the original 16 licenses were given.
That would be one way to decide 17 wha t level you ' vere trying to address.
18 Another is to consider a thought we ha ve had 19 many times, that we ought to have some floor on the 20 earthquake that is considered.
And some of these plants 21 are designed to a pretty low seismic level.
So going up 22 to some specified floor and using that as the basis for 23 finding out what the margin is night be a useful
()
24 exercise.
25 I guess the third is to look at this research O
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46 1
work that was done and see whether it provides any clue 2
as to how to go about looking at things.
I think it 3
would show you where to look.
Most pieces of hardware, O-4 there's so darn much margin tha t you could ignore it 5
altogether.
There are just a few places where the 6
margin is sufficiently low to make it worth lookino at 7
carefully.
8 I would be inclined to say that, using that 9
work as a reference, you might find some guidance as to 10 what to concentrate one That is about the best 11 commentary I can make at this stage of the game.
12 MR. OKRENT:
Well, Carson, I would say that 13 the sentence, wherever it is in the Perry letter, we 14 believe it is important that there should be 15 considerable assurance tha t the combinat$on of seismic 16 design basis and margins and seismic design are such 17 that this accident source represents an acceptable low 18 contribution to overall risk at this plant that is a 19 fairly specific kind of general guidance.
20 It doesn't say how to do it at all, but it l
21 does, I think, suggest, at least to me, what one should 22 look for in this regard.
23 MR. MARK:
Look, Dave, I don't disagree with O
24 rou.
I enink ene war it was phrasea in verry was 25 perhaps the best.
The way it was phrased in Wolf Creek O
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didn 't come out with tha t line of thought.
It just 2
said, considerably bigger --
'3 MR. OKRENT:
Well, as Mike suggested earlier, 4
Wolf Creek was designed with two different design bases, l
5 so the Committee was suggesting you might want to take a 6
harder look at the part that had lesser design basis.
7 MR. MARK The'way it was said in Perry could 8
be harmon2 zed with the present form of the safety goal.-
9 MR. OKRENT:
Yes.
10 MR. MARKS The way it was said in Wolf Creek l
11 doesn ' t accomplish that.
12 MR. OKRENT:
I think the intent was the same, 1
l 13 but this particula r sentenec, I think, could provide, as l
14 I say, the general thought.
15 Can I make a comment?
It is conceivable to me 16 tha t for f uture plants people will find it useful in l
17 trying to meet this objective, if that's your objective, 18 not to have the same design basis seismically speaking 19 for all parts of the plant, because they may know that 20 if I design it for a.2g earthquake for the containment 21 I can handle whatever it is, 1g or something, with no 22 point going to larger numbers, whereas there may be 23 certain actuators or relays or something which, if I O
24 heve designed it for.29 end shook it for the 25 corresponding, I don't have much margin, at least I O
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()
1 don't know I have much margin, so I have to qualify it 2
for something substantially larger or whatever I'm going 3
to do.
4 In other words, in order to try to meet this 5
kind of objective, it may mean that for the future plant 6
you k'eep it in mind as you go around.
And getting back 7
to the point that Bender made, there may be certain 8
specific areas where things are the most sensitive, so 9
those in fact have some additional either analysis by 10 inelastic to show they're all right or some additional 11 support or wha teve r, or it may be less support.
I don't 12 know which is better, in fact.
I'm not trying to enter 13 that argument at all, myself.
14 MR. BENDERS It's worthwhile to remember that 15 in many cases we have designed it in such a way that we 16 are making the structure too strong.
It has been a 17 disadvantage.
18 MR. OKRENT4 It's conceivable.
19 HR. SIESS:
Dave, I certainly agree with the 20 Perry statement, that it represents a goal that is not 21 unreasonable.
But it seems to me that the way that is 22 stated the only way you can satisfy it is by doing a 23 PRA, because when we use the words " contribution to
()
24 risk" to me that conjures up PRA.
Now, maybe it doesn't 25 to everybody, but --
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()
1 MR. OKRENT:
Can I comment on that?
I think 2
PRA including seismic in principle would be part of 3
doing that.
But you might be able to use other PRA's 4
and decide that your plant is similar enough to these 5
others so that you know what the rest of the plant is 6
like, for example.
And you might look at seismic under 7
a portion of the PRA for different plants.
8 But having done a PRA of the kind I've seen, 9
let's say for Zion, I don 't think they looked hard 10 enough a t the specific plant to provide necessarily the 11 assurance that the actuators and the valves and so forth 12 that you need, and the small lines and so forth, are 13 okay.
It was a generic kind of fragility study that was O
14 used in Zion, and one has to go back in that area, I 15 think, and do some more thinking.
l 16 MR. EBERSOLE:
May I ask a point of f
17 cla rification?
To me this thing sort of resolves 18 itself.
I realize we've designed LOCA mitigation 19 systems to survive earthquakes, but we've never admitted 1
20 we're going to have a loss of coolant accident, at least 21 formally.
Those strange words, " coincident, not caused 22 by."
23 If we are going to get with the business of I) 24 mitigating LOCA's and earthquakes, I think we should be 25 forthright and say that, and be very specific, because
)
ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, O C. 20024 (202) 554 2345
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()
1 life gets a lot tougher when you say that.
You've got a 2
lot more things to worry about than a simple reactor 3
trip and the shutdown cooling function.
And I'd like to 4
see some policy statements in fact clarifying very well 5
indeed that that's the way we're going to go.
6 Then we're going to have to look at a much
,7 larger field of seismic margins.
8 hR. MARKS Competence of equipment to cope 9
with the loss of coolant accident coincident with the 10 following earthquake.
If, on the other hand, we deny 11 this combination, then we've got a much smaller field of 12 problems to work with, and I think a much better chance 13 of showing we can do it without gross costs involved.
14 As a matter of fact, we just heard yesterday, 15 it turns out in their seismic analysis, unlike th e --
16~
one of the earlier plants, where they had pinion ty'pe 17 pumps that turned out to be weak, th at by some simple 18 braces that prevents the swinging of these things 19 apparently they have fixed that.
We're going to look at 20 it again.
Their problem was based on relay chatter and 21 pump performance.
22 This is a housekeeping problem that may get 23 into fine detail.
We may find two dollar items right at
()
24 the plant we didn't look at tha t really are Achilles' 25 heels, in spite of our heavy investments in heavy O
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and so forth, that some of the little gingerbread in the 3
design just won't work.
4 5
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8 9
10 11 4
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13
, 14 15 16 17 18 19 21 22 23 24 25
' O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W WASHINGTON, D.C. 20024 (202) 554-2345
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52
()
1 MR. KNIGHT:
I certainly agree that as a 2
result of such effort you find Achilles' heels, weak 3
links, that the effort has paid off handsomely.
)
4 You bring up another point, though.
It gets 5
back to this business of the nitty gritty, I suppose one 6
might say, of regulation.
You end up chasing your tail 7
a little bit.
8 As you say, if I find that the first time 9
around there were some really weak link and now the next 10 weak link is, you really need some point where you 11 decide, okay, there certainly is always going to be 12 something which is a weak link, so to speak, but its 13 capacity is so much larger than whatever the goal is O
14 t hat we are right.
15 MR. EBERSOLE:
I think there may he a general 16 recipe that the weakest link ought to be the costliest 17 link, that you shouldn't wind up with some poorly 18 designed, $2 items.
19 MRe KNIGHT:
That is certainly f undamental 20 cost-benefit.
21 MR. BENDER:
There are only 15 members of this 22 Committee, so there are only 15 opinions.
23 (General laughter.)
()
24 MR. BENDER:
But I think that a pretty good 25 approach is to try to compare the hardware that exists ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345
53
(])
1 in these plants that we have asked about, the hardware 2
in the plants that have the higher seismic design 3
requirement taan the one that we are talking about, and 4
seeing what the sienificant differences are, and tha t a t 5
least would give you some feeling for the problem.
6 I guess I am not in a position to say how much 7
of that needs to be done, but it will turn out that most 8
of the equipment is about of the same class.
I would be 9
much more concerned about trying to jack up the seismic 10 requirements at Diablo Canyon from what it is now than I 11 am from trying to suggest that you take another look at 12 Summer, if I can use the extreme.
13 While I don't like to ask you to do more O
i 14 analysis, I don't think the problem is all that big.
15 MR. OKRENT:
I think Mike 's point about trying 16 to see what was done at some of these higher design 17 basis plants, that could be an interesting way.
18 We are going to have to break in a few minutes 19 because I think there is another meeting that begins at 20 2:30.
21 I would like to suggest that maybe we get 22 together again for an informal discussion in a couple of 23 months or something.
Maybe we'll have developed some
()
24 possible approaches or,whatever, and maybe somebody can 25 do a little looking one way or another as to --
()
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MR. KNIGHTS I think that would be most 2
useful.
We will certainly be 1 coking at how to tackle 3
thic problem.
4 As I said, I am convinced myself that there 5
are some significant policy questions here.
We will 6
also be trying to come to grips with those.
I just 7
don't want to be in a posture of saying yes, we will 8
come back and lay out a plan for you.
9 MR. SIESS:
Dave, why don't we have a meeting 10 with SSMRP and instead of listening to them telling us 11 What they have been doing, we can ask them some 12 questions sbout seismic margins and see if they can 13 answer them.
I 1 earned more about the project than what 14 we were in the previous meeting.
15 MR. OKRENT:
We have already had a suggestion 16 for how we might have some future meetings.
17 I might just say as a point of information 18 tha t a t the LMFBR safety meeting in Lyons, the issue l
19 that stuck out in my mind is the one that the French who i
l 20 were the furthest along emphasized, that for seismic 21 design they seem to be thinking in terms of possibly 22 putting one of their isolation devices under future 23 LMFBRs to facilitate seismic design.
O 24 MR. SItSS:
Thet v111 he1p the expert merxet 25 in Ca11fornia.
l O
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55 O
"a oxaz"r.
1t 1= tatere tiao-2 MR. EBERSolEs Is that because they do not 3
vant to admit TVA?
4 MR. OKRENT:
No.
You have to keep certain 5
functions going.
They have a thin wall system, you have 6
sloshing possibly, and either because it -- I just 7
wanted to know if you were focusing on that.
8 MR. KNIGHTS Is that published?
9 MR. OKRENTs There will be a proceeding on one 10 of the papers given by one of the French engineers.
11 MR. SHEWMONa Wha t stuck out as most important 12 in Dave's mind will be published only in the minutes of 13 the meeting.
14 HR. OKRENT:
Well, I guess we had better then 15 thank the Staf f f or coming down and talking about this, 16 and we will adjourn then this subcommittee meeting.
17 (Whereupon, at 2: 25 o ' clock p.m. the meeting 18 of the subcommittee was adjourned.)
19 20 21 22 23 24 25 0
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TU C N RE N TORI CD.WISSICN This is Oc cartif*/ that the accached proceedings befers the O. in the matter cf: ACRS/Subcor.mittee on Extreme External Phenomena Qate cf Proceeding:
Auaust 11, 1982 Occket liu=ber:
Place of Freceeding:
Washington, D.
C.
were held as herein a pears, and that this is the original transcrip thereof for the file of the Ccc:sission.
Jane N. Beach Official Reporter (Typed)
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O RiE SITE-SKCIFIC PORTIONS OF die PLMT, INCUJDING VITAL ASECTS OF THE ULTIiWiE HEAT SINK #0 ASSOCIATED SYSTEMS, EE DESIGED FOR A 0.12 9 EARTHQUAKE, #0 AE BEING EANALYZED FOR #1 EARTHQUAE REPESBRED BY SITE-SKCIFIC ESPONSE SPECTRA THAT ARE ENCOMPASSED BY EGULATORY GUIDE 1.60 SKCTRA ANCHORED AT A ZER0-KRIOD ACCELERATION OF 0.15 9.THE STNEARD PORTION OF vie PLANT, ON DIE OBiER HAND, WAS DESIGED FOR A 0.20 9 EARTHQUAKE WIU1 THE USUAL PARGINS OF SAFETY #0 THUS h0JLD DE EXPECTED TO WITHSTN0 A CONSIDERABLY LARGER EARTHQUAKE WIE10lK FAILING IN SUCH A MANNER AS TO CAUSE A SEVEE ACCIDEtR.
WE D0 t0T HAVE CONFIDENCE THAT ALL VITAL ASKCTS OF TE ULTIf%TE HEAT SINK
- 0 ASSOCIATED SYSTBiS HAVE MARGINS SUFFICIBIT TO PROVIDE #1 APPROPRIATE LEVEL OF ESISTANCE TO A LOWER PROBABILITY, IDE SEVEE EARTHQUAKE. E ECOWB0 EiEEF0E TliAT T[ SEISMIC MARGINS INHERENT IN die CCtPONEfES OF THE ULTIf%TE EAT SINK AND ASSOCIATED SYSTEMS BE ItNESTIGATED FURTH
- 0 THAT ANY NEEDED MODIFICATIONS BE MADE BEF0E THE PLANT RESLPES OP AFTER THE SECOND REFUELING.
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O E EC0ftEND ilMT TE APPLICNff #0 TE NRC STAFF CONDUCT STUDIES TO EVALUATE DE f%RGINS AVAILABLE TO ACCOMPLISH SAFE SHlfTDOWN, INCUElilG LONG-TERM HEAT f00 VAL, FOLLG11NG AN EAREQUAE OF SGB&MT GREATER SEVERITY AND LOWER LIKELIHOOD THN1 TE SAFE SHUTDOWN EARTHQUAKE. lE BELIEVE IT IS IMPORT #1T THAT THEE SHOULD E CONSIDERABLE ASSURANCE TIMT lliE COMBINATION OF SEISMIC DESIGN BASIS #0 MARGIllS IN lliE SEISMIC DESIGN IS SUCH 111AT THIS ACCIDENT SOURCE PEPRESENTS AN ACCEPTABLY LG1 CONTRIBLIT T0 lliE OVEPALL RISK FROM THIS Pl#1T. WE EC0ftB0 THAT ANY EEDED MDDIFI-O cations BE ranE BEFORE mE PLANT ESLES OWRATIQ1 FOLL0 DING THE SEC0tB EFUELING. WE WISH TO E KEPT INFORMED ON lliE PROGESS NO ESULTS OF TliESESlUDIES.
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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS t
W ASHINGTON,0. C. 20555 f
n July 13,1982 Q
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Honorable Nunzio J. Palladino Chairman U. S. Nuclear Regulatory Commission Washington, DC 20555
Dear Dr. Palladino:
SUBJECT:
ACRS REPORT ON THE PERRY NUCLEAR POWER PLANT, UNIT 1 During its 267th meeting, July 8-13, 1982, the Advisory Committee on Reactor Safeguards reviewed the application of the Cleveland Electric Illuminating Company (Applicant), acting on behalf of itself and as agent for Duquesne Light Company, Ohio Edison Company, Pennsylvania Power Company, and the Toledo Edison Company, for a license to operate the Perry Nuclear Power Plant, Units 1 and 2.
The plant is to be operated by the Cleveland Electric Illuminating Company.
A tour of the facilities was made by members of the Subcommittee on the morning of June 28, 1982, and a Subcommittee meeting was held in Cleveland, Ohio on June 28 and 29,1982 to consider the application.
During its review the Committee had the benefit of discussion with represen-tatives of the Applicant, the NRC Staff, and members of the public.
The Committee also had the benefit of the documents listed.
The Cornittie enmmented on the application for a permit to construct this plant in its reports dated December 12, 1974 ard May 12, 1975.
The Perry Nuclear Power Plant is located in Lake County, Ohio near Lake Erie approximately 35 miles northeast of Cleveland, Ohio and 21 miles southwest of Ashtabula, Ohio.
Units 1 and 2 use General Electric BWR-6 nuclear steam supply systems with a rated power of 3579 MWt and a Mark 'III pressure suppression containment system with a design pressure of 15 psig. Construc-tion of Unit 1 is about 83% complete and Unit 2 is about 43% complete.
Because loading of fuel for Unit 2 is scheduled for May 1987, the Committee does not believe it appropriate to report at this time on the operation of Unit 2.
Our review included the management organization, technical support staff, status of operational staffing, and the training program. This is the first nuclear power plant to be operated by the Applicant.
The plant staff has a minimum amount of boiling water reactor (BWR) nuclear background.
We agree with the NRC Staff on the urgent need for additional personnel with BWR.
experience within the operating management.
The Applicant should fill the O.
position of Superintendent of Plant Operations in the near future.
Experi-enced senior technical support personnel should be included in the staffing plans of the Applicant.
This matter should be resolved in a manner satis-factory to the NRC Staff.
We wish to be kept informed.
7 wyin
t, Honorable Nunzio J. Palladino July 13,1982 As a result of adverse experience on the Perry project several years ago.
O the Applicant restructured its quality assurance procedures and its quality control end assurance organization.
The revised organization has been reviewed and audited by the NRC Staff. We wish to receive a report from the NRC Staff which discusses design and construction problems, their disposi-tion, and the overall effectiveness of the effort-to assure appropriate quality.
The Applicant has committed several technical staff members to matters related to probabilistic analysis and studies of systems interactions.
We believe that efforts of this sort by the operating utilities are to be i
encouraged.
The Mark III suppression pool dynamic loads have been identified as an Out-standing Issue in the NRC Staff's review.
The NPC Staff has provided the Applicant with a proposal for the appropriate design basis loads, and it appears that the Perry design will be able to accommodate these loads.
Additional concerns with the design of the Mark III containment have been recently brought to our attention.
The NRC Staff is currently assessing these issues for impact on the Mark III design. We will continue to discuss with the NRC Staff, on a generic basis, Mark III suppression pool dynamic loads and other additional Park III issues.
Hydrogen contro) systems for Mark III containments are being developed by the Mark III Owners Group.
Efforts by this Owners Group are being directed toward the development of a hydrogen ignition,ystem which makes use of distributed ignition sources.
The NRC Staff has indicated that they will be able to meet with the Committee on this matter in the near future.
We expect to review this system on a generic basis.
Acceptability of this system is designated as a License Condition.
/ We recommend that the Applicant and the NRC Staff conduct studies to ::valu-l
(
ate the margins available to accomplish safe shutdown, including long-term j heat removal, following an earthquake of somewhat greater severity and lower t
l likelihood than the safe shutdown earthquake.
We believe it is important i
j that there should be considerable assurance that the combination of seismic
/ design basis and margins in the seismic design is such that this accident
! source represents an acceptably low contribution to the overall risk from this plant.
We recommend that any needed modifications be made before the plant resumes operation following the second refueling.
We wish to be kept informed on the progress and results of these studies.
O o
~.
l Honorable Nunzio J. Palladino July 13,1982
(.
1 During our review, the NRC Staff identified a number of other License O
Conditions Confirmatory Matters, and Outstanding Issues which remain to be resolved.
Except for the issue of turbine missiles, we are satisfied with i
the progress on these topics, and we believe that they should be resolved in a manner satisfactory to the NRC Staff.
We wish to be kept informed con-cerning resolution of the turbine missile issue, and wish to receive a technical report which discusses and evaluates the problems involved.
If due consideration is given to the recommendations above, and subject to satisfactory completion of construction, staffing, and preoperational testing, the ACRS believes there is reasonable assurance that the Perry Nuclear Power Plant, Unit I can be operated at power levels up to 3579 MWt j
it.hout undue risk to the health and safety of the public.
Sincerely, P. Shewmon j
Chairman j
i References i
l 1.
Cleveland Electric Illuminating Company, " Final Safety Analycis Report, Perry Nuclear Power Plant, Units I and 2." with Amendments 1-6 2.
U. S. Nuclear Regulatory Commission, " Safety Evaluation Report, Perry l
Nuclear Power Plant, Units 1 and 2." USNRC Report NUREG-0887, dated May 1982 3.
Famorandum from D. Houston /J. Kudrick, NRC, to A. Schwencer/W. Butler, i
NRC,
Subject:
Summary of May 13, 1982 telecon with John Humphrey -
Concerns about Grand Gulf Mark III Containment, dated May 18, 1982 i
4.
Letter from John M. Humphrey, Humphrey Engineering, Inc., to L. F.
Dale, Mississippi Power and Light,
Subject:
BWR-6/ Mark III Contain-ment Design Issues, dated May 8,1982 l
l l
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/p as:%'o UNITED STATES
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NUCLEAR REGULATORY COMMISSION s
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I :I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS mswincios.o. c. rosss
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July 13,1982 Honorable Nunzio J. Palladino Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Dr. Palladino:
a
SUBJECT:
ACRS REPORT ON THE SUITABILITY OF THE CLINCH RIVER BREEDER REACTOR PLANT SITE During its 267th meeting, July 8-10, 1982, the ACRS reviewed NUREG-0786,
" Site Suitability Report in the Matter of Clinch River Breeder Reactor Plant" and considered the suitability of the proposed site for such a plant.
The matter was also discussed on June 24, 1982 at a joint meeting of the Subcommittees on Clinch River Breeder Reactor and Site Evaluation.
During both meetings, we had the benefit of input from representatives of the NRC Staff and the Department of Energy (Applicant).
We also had the benefit of the documents listed below, as well as a direct discussion with the author of Reference 4.
Tne proposed Clinch River Breeder Reactor (CRBR) plant site is located in Roane County in east-central Tennessee, approximately 25 miles west of Knoxville and within the city limits of Oak Ridge, Tennessee.
The site consists of approximately 1,364 land acres on a peninsula formed by a meander in the Clinch River.
It is bounded on three sides by the River and on the north by the Department of Energy's (DOE) Oak Ridge Reservation. The site property is owned by the Federal Government, and the portions of the site required for constructing and operating the plant will fall under the custody of DOE.
The CRBR plant will be a single-unit electric power plant with a liquid sodium-cooled loop-type breeder reactor utilizing a fuel of mixed uranium-plutonium oxides.
With the initial reactor core, the design power will be 975 MWt, and the net output will be 350 MWe.
DOE has requested a Limited Work Authorization (LWA-1) to begin nonsafety-related site preparation activities.. It is rcquired in 10 CFR 50.10 that, before an LWA-1 can be granted, an Atomic Safety and Licensing Board (ASLB) must determine that there is reasonable assurance that the proposed site is a suitable location from a radiological health and safety standpoint for a nuclear power reactor of the general size and type proposed. Our review was O
made in response to an NRC Staff request in connection with the required ASLB determination. The NRC Staff carefully defined the scope of the review to consider whether the site is suitable for a reactor "of the general size and type" of the CRBR; the current design of the CRBR plant itself was not evaluated.
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'C iionorable Nunzio J. Palladino July 13,1982
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Among the topics considered in this review were the location and distribu-O tioa of Populatioa round the site; the 2eoio9x. seismoio27. aad hydroiosi of the site; an assumed Site Suitability Source Term; and the risks to be expected from a plant of the CRBR type.
As part of its approach to trying to make the risks from an LMFBR coctparable with those from a light water reactor (LWR), the NRC Staff provided 8 review criteria for CRBR core disruptive accidents.
We believe that this appears to be a-reasonable first approach but also believe that at the construction permit stage substantive assurance will be needed that such criteria are being met.
We wish to note that we do not necessarily agree witn all the LMFBR Design Criteria specified in Appendix A of NUREG-0786.
The NRC Staff appears to have accepted the Applicant's assertion that a CRBR type plant would not represent an undue. hazard to the K-25 Plant.
We recommend that the Staff confirm through an iddependent assessment that the potential effects of a CRBR type plant on the K-25 plant are acceptable.
,'With regard to the seismic design of thid plant, we believe it is important that the combination of seismic design basis and margins in the seismic j
I design be such that this accident source represents an acceptably low contribution to the overall risk from the " plant.
We believe this matter
$ will warrant detailed examication at the construction permit stage to assure
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' that necessary margins are available for all important systems and compon-(ents.
The NRC Staff has concluded that the CRBR plant can be designed and con-structed in such a manner that it will present no greater risk to the health and safety of the public than an LWR plant neeting current safety criteria.
We believe that the proposed site is suitable for such a plant.
Sincerely,
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P. Shewmon Chairman R_eferences 1.
G.~S. Nuclear Regul: P ry 1, h sfon, " Site Suitability Report in the Matter of Clinch Rizer Breese Reactor Plant," NUREG-0786, dated June 1982, Revision to March 4,1977 Report O
2.
Letter from J. R. Longenecker, 00E, to P. Boehnert, ACRS, concerning earthquake recursion relationships, dated July 7,1982 3.
Handout from NRC Staff (undated) titled, " Review Criteria for CRBR Core Disruptive Accidents" 4.
Letter from T. B. Cochran, National Resources Defense Council to
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P. Shewmon, ACRS, dated July 7,1982
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
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msm NGTON. D. C. 20555 Jur.e 8,1982 Honorable Hunzio J, Palladino Chairman U. S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Dr-Palladino:
SUBJECT:
ACRS INTERIM REPORT ON MIDLAND PLANT, UNITS 1 AND 2 During its 266th meeting, June 3-5, 1982, the Advisory Committee on Reactor Safeguards reviewed the application of Consumers Power Company for a li-cense to operate the Midland Plant, Units 1 and 2.
This application was also considered at Subcommittee meetings held on April 29, 1982 in Washing-ton, D. C.,
on May 20-21,1982 in Midland, Michigan and on June 2,1982 in Washington, D. C.
On May 20,1922 members of the Subcommittee toured the plant.
In the course of these meetings the Committee had the benefit of discussions with representatives and consultants of Consumers Power Company, Babcock and Wilcox Company, Bechtel Corporation, the Nuclear Regulatory Commission Staff, and members of the public.
The Committee also had the benefit of the documents listed below.
The ACRS reported on June 18, 1970 regarding the construction permit ap-
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plication for the Midirer.d Plant; on September 23, 1970 regarding several amendments to the application; and on November 18, 1976 regarding applica-ble generic matters.
l The Midland Plant site is located on the south bank of the Tittabawassee River adjacent to the southern city limits of Midland.
The main industrial complex of the Dow Chemical Company lies within the city limits directly l
across the river from the site.
There are about 2000 industrial workers within one mile of the site, and the estimated 1980 population was about 51,400 residents within five miles of the site.
This makes the Midland site one of the more densely populated sites at distances close to the Plant.
Each of the two Midland units employs a Babcock and Wilcox designed nuclear steam supply system rated at 2468 MWt with a stretch power rating of 2552 MWt.
The Midland Plant is unique in that the heat generated will be used not only to produce electricity but also to produce process steam for the Dow Chemical Company plant via a tertiary system.
O The Midland Plant has been the subject of several major problems related to quality assurance during plant construction.
One of these problems relates to the soil fill under several, safety-related structures.
The L.
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Honorable Nunzio J. Palladino June 8, 1982 O
eerici #ci = rei t4#9 to soii riii a v i a to excessive tti = eat ae some cracking of these structures, and have also introduced questions concerning, the adequacy of protection against liquefaction of the granular portions of the fill in the event of strong vibratory motion accompanying an earthquake.
The Applicant has proposed and is implementing, under close surveillance by the NRC Staff, remedial measures with regard to the foundation deficiencies.
We are 91nerally satisfied with the approach being taken, subject to confir-mation of the overall quality assurance program and the seismic design basis.
Both of these items are discussed below.
With regard to quality control of design and construction, the report of the ERC Staff's Systematic Assessment of Licensee Performance (SALP) review for the period July 1,1980 to June 30, 1981 revealed deficiencies in the instal-lation of piping and piping suspension systems, in the pulling of electrical cables, and in the handling of problems relating to soils and foundation.
Deficiencies by the Applicant in the handling of soils-related matters have continued to occur, subsequent to issuance of the SALP report.
We believe that the NRC Staff is handling the corrective actions for specifically identified quality assurance deficiencies in an appropriate manner.
In view of the overall concern about Midland quality assurance the NRC should arrange for a broader assessment of Midland's design adequacy and construction quality with emphasis on installed electrical, control, and mechanical equipment as well as piping and foundations.
We wish to receive a report which discusses design and construction problems, their disposi-tion, and the overall effectiveness of the effort to assure appropriate quality.
Our reservation concerning seismic design relates to the lack of adequate assurance tnat the Midland Plant will be capable of accomplishing shutdown heat removal for low probability earthquakes more severe than the safe shutdown earthquake (SSE).
The Midland seismic design basis at the con-struction permit stage corresponded to a MMI VI, peak ground acceleration of 0.129, employing a modified Housner spectrum.
For the operating license review, the NRC Staff has reevaluated the original seismic design basis and the Applicant and the NRC Staff have agreed on the use of site-specific analyses which have led to increases in the design response spectra for frequencies above about 2 cycles /sec.
Historically, no earthquakes stronger than the newly proposed SSE have O
occurred within 200 miles of the Plant.
However, expert opinion differs widely on the exceedance frequency of the proposed SSE and on th severity at the site of earthquakes whose likelihood is less than 1 in 1 or 1 in 5
10 per year.
w Honorable Nunzio J. Palladino June 8, 1982
[capabil'ity of the plant, as originally designed, to withstand the rev The Applicant is currently reevaluating by selective audit the seismic
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SSE.
Measures taken to assure safe shutdown in the event of an earthquake include thh use of dewatering to reduce the potential.for soil liquefaction, We recommend that all systems and components important to decay heat removal i
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be carefully evaluated for their ability to accomplish necessary functions I
in the unlikely event of lower-probability, more severe earthquakes in order to provide the necessary degree of assurance.
This matter should be re-solved in a manner satisfactory to the NRC Staff.
We wish to be kept informed about the resolution of this matter.
We believe that any recom-mendations for changes in the plant resulting from this evaluation should be implemented by the end of the second refueling outage.
The Applicant has agreed to provide core exit thermocouples, a hot-leg-level measurement system, and subcooled margin monitors as instrumentation to detect inadequate core cooling.
Consumers Power Company also plans to include a remotely operable vent on top of both inlet loops to the steam generators; however, Consumers has not committed to supply a high point vent on t'ne reactor vessel head.
This matter should be resolved in a manner satisfactory to the NRC Staff.
The ACRS recommends that the Applicant review further the potential for providing indications of water content or l
level within the reactor vessel.
The staff of the Applicant includes many personnel who have had nuclear l
power plant experience.
However, operating experience with this B&W type power reactor is limited, and the NRC Staff is requiring that at least one person having experience on a large commercial PWR be included on eau shift for one year. We support the NRC Staff position.
The Applicant's experien',e with the operation of nuclear power plants should, in principle, place Consumers in a favorable position to provide continuing, careful oversight of the operations at the Midland Plant.
In view of some prior adverse operating experience at the Palisades Plant however, we recommend that the NRC Staff institute an augmented audit of operations at Midland, at least during the early years of operation at power.
We have reviewed the evaluation made of the tertiary process steam system for use by Dow Chemical Company.
This system appears not to impose any unacceptable impacts either on the safe operation of the Midland Plant or on the people working'at the Dow Chemical Company.
n The Applicant has undertaken an effort to have a probabilistic risk assess-V ment (PRA) performed for the Midland Plant and stated that the results will be available in the fall of 1982.
We believe it desirable to have plant-specific PRAs performed for each commercial nuclear power plant and that
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Honorable Nunzio J. Palladinn A,w G. 1982 m
it is particularly appropriate for the Midland Plant because of its rela-U tively ~high, close-in population density. We wish to have the opportunity to review the Midland PRA with assistance from the NRC Staff, and to offer comments or recommendations as appropriate.
We do.not believe that this review need delay licensing of the Midland Plant for. operation.
Recently, questions have come to light in connection with B&W plants con-cerning the availability of natural circulation in the presence of an interrupfed or continuing small break loss-of-coolant accident.
We wish to see a proposed NRC Staff resolution of this issue.
The Applicant described an extensive systems interactions study being undertaken for the Midland Plant.
We wish to be informed of the results of this study.
We believe that, in view of the population density near this plant, addi-tional prudence is appropriate for the Midland Plant in the resolution of the ATWS issue and other Unresolved Safety Issues.
We endorse the participation of Dow Chemical Company plant personnel in emergency procedures developed on the basis of an assumed failure at the Midland Plant.
Similarly, there should be active participation by Midland Plant personnel in emergency procedures developed on the basis of an assumed failure at the Dow Chemical plant. The Applicant and the NRC Staff should promote continued coordination of these types of relationships, as well as those involving appropriate state and local groups to assure that the capability for an effective emergency response is developed and main-tained.
With regard to the eleven items identified in the ACRS Supplemental Report on Midland Plant, Units 1 and 2 dated November 18, 1976, we have the follow-ing comments.
The issues related to vibration and loose-parts monitoring, potential for axial xenon oscillations, behavior of core-barrel check valves during normal operation, fuel handling accidents, effects of blowdown forces on core internals, LOCA-related fuel rod failures, and improved quality assurance and in-service inspection for the primary system have all been resolved or are in a confirmatory stage of being resolved. Separation of protection and control equipment has been accomplished in an appropriate manner; however, the safety implications of control systems remains an Unresolved Safety Issue directly applicable to Midland.
Resolution awaits completion of the NRC Staff Task Action Plan A-47.
The effect of ECCS i
induced thermal shock on pressure vessel integrity has been resolved in part; however, the Unresolved Safety Issue on pressurized thermal shock 1
O will apply.
Environmental qualification of equipment remains a generic j
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H2ncrabic Nuazio J. Pal 16dino June 8,1982 O
issue yhich is under review by the NRC Staff and whose resolution will apply to the Midland Plant.
Instrumentation to follow the course of an accident has been resolved in part by the development of revised Regulatory Guide 1.97.
We do not believe that licensing of the Midland Plant for operation need await further resolution of any of.the eleven issues dis-cussed above.
The various other matters identified by the NRC Staff as open or confirma-tory in T.he Safety Evaluation Report should be resolved in a manner satis-factory to the NRC Staff.
We wish to be kept advised concerning resolution of the turbine missile issue.
Ine ACRS believes that, subject to satisfactory completion of construction and staffing and if due regard is given to the comments above, the Midland Plant, Units 1 and 2 can be operated at power levels up to 5 percent of full power with reasonable assurance that there is no undue risk to the health and safety of the public.
We defer our recommendation regarding operation at full power until we have had the opportunity to review the plan for an audit of plant quality and the proposed resolution of the question regarding natural circulation in i
the presence of a small break LOCA.
Dr. Kerr did not participate in the Committee's review of this matter.
Sincerely,
\\.
P. Shewmon Chairman
References:
T.
Consumers Power Company, " Midland Plant Units 1 and 2 - Final Safety Analysis Report" including Amendments 1-43 2.
U.S. Nuclear Regulatory Commission, " Safety Evaluation Report Related to the Operation of Midland Plant, Units 1 and 2 " NUREG-0793, dated Pay 1982 3.
U.S. Nuclear Regulatory Commission, "NRC Licensee Assessments "
NUREG-0834, dated August 1981 l
4 Letter from J. Cook, Consumers Power Company, to J. Keppler NRC,
Subject:
Midland Project Response to Draft SALP Report, dated m
Fay 17,1982 5.
Letter from J. Cook, Consumers Power Company, to J. Keppler, NRC,
Subject:
Midland Project Quality Assurance Program Update, dated April 30,1981 h
's Hancrable Wnzio J. Palladino ~
June 8,1982 6.
Letter from J. Hind, NRC, to J. Cook, Consumers Power Company, y
Subject:
Systematic Assessment of Licensee Performance (SALP),
dated April 20, 1982 7.
Letter from J. Cook, Consumers Power Company, to H. Denton, NRC, Subject :
Summary of Soils-Related Issues at the 'Hidland Nuclear Plant, dated April 19, 1982 8.
Letter from K. Drehobl, Consumers Power Company, to D. Fischer, ACRS.
Subject : Midland Project Soils Information, dated April 12, 1982 9.
Statenent of Ms. M. Sinclair to ACRS, dated June 4,1982 10.
Letter from B. Stamiris to Dr. D. Okrent and ACRS Members,
Subject:
Midland OL Review, dated May 29, 1982 11 Letter from M. Sinclair to Dr. P. Shewmon, ACRS,
Subject:
Midland OL Review, dated May 28, 1982 12.
Statement by Dr. C. Anderson to ACRS Midland Plant Subcommittee dated Fay 20-21. 1982 13.
Statement by Ms. M. Sinclair to ACRS Midland Plant Subcommittee dated Fay 20-21, 1982 14.
Letter from B. Stamiris to D. Fischer and ACRS Members,
Subject:
Soil Settlement and QA Issues, dated May 20, 1982 15.
Letter from M. Sinclair to Dr. C. Siess, ACRS,
Subject:
Midland Soil Settlement, dated April 26, 1982 4
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'4 UNITED STATES p
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WASHINGTON, D. C. 20555 h
May 11, 1982 0
Honorable Nunzio J. Palladino Chairman U. S. Nuclear Regulatory Commission Washington, D. C.
20555
Dear Dr.,
Palladino:
SUBJECT:
ACRS REPORT ON THE WOLF CREEK GENERATING STATION, UNIT NO.1 During its 265th meeting, May 6-8, 1982, the Advisory Committee on Reactor Safeguards reviewed the application of Kansas Gas and Electric Company (KG8E), Kansas City Power and Light Co. and Kansas Electric Power Coopera-tive, Inc. (Applicants) for a license to operate the Wolf Creek Generating Station, Unit No.1.
The Station is to be operated by KG8E. A Subcommittee meeting was held in Emporia, Kansas, on April 21-22, 1982, to consider this project.
A tour of the facility was made by members of the Subcommittee on April 21, 1982. During its review, the Committee had the benefit of discus-sions with representatives and consultants of the Applicants, Westinghouse Electric Corporation, Bechtel Power Corporation, the Nuclear Regulatory Commission (NRC) Staff, and with members of the public.
The Committee also had the benefit of the documents listed below.
The Committee commented on the constructioh perrit application for this plant in its report dated October 16, 1975.
The Wolf Creek Generating Station is located in Hampdon Township, Coffey County, Kansas.
The site is in eastern Kansas approximately 53 miles south of Topeka, and 100 miles east-northeast of Wichita.
The nearest population center is Emporia, Kansas, 28 miles west-northwest of the site (estimated 1980 population of 25,019).
The Wolf Creek Generating Station will be the first commercial nuclear power plant in the state of Kansas.
It should be assured that state and local agencies are qualified to respond to possible emergency situa-tions associated with the operation of the Wolf Creek Generating station.
The Station will use a Westinghouse, four-loop, pressurized water, nuclear steam supply system having a rated power level of 3425 MWt.
Unit 1 em-ploys a cylindrical, steel-lined, reinforced, post-tensioned concrete containment structure with a free volume of 2.5 million cubic feet.
The Wolf Creek Generating Station uses the Standardized Nuclear Unit Power O
Plant System (SNUPPS) design.
It is one of two plants built to this design.
The Committee reported on the operating license application of the other plant (Callaway Plant Unit No.1) in its November 17,1981 re-port to,you.
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Hencreble K
- 10 J. Palladino May 11, 1982
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The Wolf Creek Generating Station is the first nuclear power plant to be Q
operated. by KG&E.
The Committee reviewed KG&E's management organization.
l experierice, and training programs.
We were favorably impressed by the general competence and attitude of KG8E's personnel.
Nevertheless, we wish to emphasize the importance of KG&E's building a strong in-house capability for analyzing and understanding the nuclear-themal-hydraulic behavior and systems perfomance of this plafit.
To strengthen the shift structure during the initial period of operation, i
KG8E pla6s to augment each shift with a consultant who is an experi-enced, previously licensed PWR operator.
These consultants will serve for a period of one year after startup.
In addition, KG&E has retained the services of a consultant with considerable ccmmercial nuclear experi ~
ence to act as a technical assistant to the Plant Superintendent through the initial loading of fuel.
We believe the technical assistant to the Pl ant Superintendent and the " experienced operator consultants" should be retained until the operating organization has developed an experience base involving those operational duties of importance to public safety.
This experience base should be defined by the NRC Staff in consultation with operational experts and incorporated into the regulatory requirements instead of using arbitrary operating time periods as a basis for measuring skill. We encourage the practice of assigning the Senior Reactor Operator (SRO) candidates to extended tours of service at operating nuclear power plants, and recommend that others in the operations staff participate in such a program to the extent practical.
KG8E has proposed, as an alternative to a Shift Technical Advisor (STA),
that at least one SRO on each shift have the training and background required for en STA.
This approach appears to us to meet the need which originally led to the requirement of an STA. However, it is not clear that the level of training given to the SR0s will correspond to that intended l
for STAS, and we recommend that the Staff review this matter carefully.
l The site-specific portions of the plant, including vital aspects of the ultimate heat sink and associated systems, were designed for a 0.12 g i
l earthquake, and are being reanalyzed for an earthquake represented by l
site-specific response spectra that are encompassed by Regulatory Guide 1.60 spectra anchored at a zero-period acceleration of 0.15 g. The standard portion of the plant, on the other hand, was designed for a 0.20 g earth-quake with the usual margins of safety and thus would be expected to withstand a considerably larger earthquake without failing in such a manner as to cause a severe accident.
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1 Hencrabic Hunzic J. Felladino May 11,1982
/ We do not have confidence that all vital aspects of the ultimate heat sink
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[. and ass.ociated systems have margins sufficient to provide an appropriate 1evel of. resistance to a lower probability, more severe earthquake.
We
> recommend therefore that the seismic margins inherent in the components of the ultimate heat sink and associated systems be inv.estigated further and that any needed modifications be made before the plant resumes operation after the second refueling.
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Other issues have been identified as Outstanding Issues, License Conditions, and Confirmatory Issues in the Staff's Safety Evaluation Report dated April 1982; these include some TMI Action Plan requirements.
Except as noted above, we believe these issues can be resolved in a manner satis-factor.y to the NRC Staff and recommend that this be done.
We believe that, if due consideration is given to the recommendations above, and subject to satisfactory completion of construction, staffing, training, and preoperational testing, there is reasonable assurance that the Wolf Creek Generating Station, Unit No.1 can be operated at power levels up to 3425 MWt without undue risk to the health and safety of the public.
e Sincerely,
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P. Shewmon Chairman
References:
1.
" Final Safety Analysis Report for Standardized Nuclear Unit Power Plant System," with Revisions 1-8.
l 2.
" Final Safety Analysis Report, Wolf Creek Generating Station Unit No.1," with Revisions 1-8.
3.
U.S. Nuclear Regulatory Commission, " Safety Evaluation Report Related to the Operation of Wolf Creek Generating Station, Unit No.1," NUREG-0881, dated April 1982.
4.
Written statement by John M. Simpson, Attorney for Intervenors, Re:
Emergency Planning Procedures and Plans - Wolf Creek Plant, dated April 22, 1982.
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[Ote:vq#c UNITED STATES p-NUCLEAR REGULATORY COMMISSION y 4 '"$[o,I ADVISORY CCMMITTEE ON REACTOR SAFEGUARDS
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WASHINGTON, D. C. 20555 I
March 9, 1982 O
soaoradie wuazio a r 11 diao Chairman U. S. Nuclear Regulatory Commission Washington, DC 20555
SUBJECT:
REPORT ON CLINTON POWER STATION UN'IT 1
Dear Dr. Pa11adino:
s During its 263rd meeting, March 4-6, 1982, the Advisory Committee on Reactor Safeguards reviewed the application of the Illinois Power Company, the Soyland Power Cooperative, Inc., and the Western Illinois Power Cooperative, Inc. (Applicant) for a license to-operate the Clinton Power Station Unit 1.
The plant is to be operated by the Illinois Power Company.
A tour of the facility was made by members of the Subcommittee on the morning of February 25, 1982 and a Subcommittee meeting was held in Decatur, Illinois on February 25-26, 1982 to consider this applica-tion. During its review the Committee had the benefit of discussion with representatives of the Applicant and the NRC Staff.
The Committee also had the benefit of the documents listed.
The Committee commented on the application for a permit to construct this Station in its report dated April 8,1975.
The Clinton Power Station is located in DeWitt County in east-central Illinois about 6 miles east of the city of Clinton and 22 miles north-northeast of Decatur.
Unit i uses a General Electric BWR-6 nuclear steam supply system with a rated power level of 2894 MWt and a Mark III pres-sure suppression containment system with a design pressure of 15 psig.
Construction of Unit 1 is about 85% complete and Unit 2 is about 35 compl ete.
Construction of Unit 2 has been deferred indefinitely, and the Applicant's motion to sever the Unit 2 proceedings from Unit I licensing proceedings has been granted.
Consequently, both the Committee and the NRC Staff hanc limited this review to Unit 1.
The Committee's review included an evaluation of the management organi-zation, the operational staff, and the training program.
The Clinton Power Station is the Applicant's first nuclear station and staffing for l
plant startup and operation is not yet complete. The Applicant, however, has made considerable progress and has a well-established training pro-gram.
The NRC Staff will continue to monitor the Applicant's progress l
and expects to complete its review before fuel loading.
O PP iceet is currenti, restructurins the construction end oPeretionai The A i
quality assurance and quality control organization in response to NRC Staff concerns. The revised organization will be reviewed and audited by the NRC Staff.
The Committee wishes to be kept inforned on this matter.
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Honerable Nunzio J. Palladino March 9, 1982 The Mark 111 sup'pression pool dynamic loads have been identified as an Outstanding Issue in the NRC Staff's review.
The NRC Staff has provided the Applicant with a proposal for the appropriate design basis loads, and O
it appears that the C1tnton design will be able to accommodate these loads. The Committee will continue to discuss, on a generic basis, the Mark III suppression pool dynamic loads with the NRC Staff.
Hydrogen control systems for Mark III containments re being developed by the Mark III Owners Group.
Efforts by this Owners Group are being directed toward the development of a hydrogen ignition system which makes use of distributed ignition sources.
The NRC Staff has indicated that they wil1~be able to meet with the Committee on this matter in the near future.
The Committee expects to review this system on a generic basis.
Acceptability of this system is a License Condition.
('The Applicant, in response to NRC Staff requirements, has reevaluated certain safety-related systems of the Clinton design using the ground motion parameters that describt the site-specific spectra equivalent to a design basis earthquake of M equal to 5.8.
The Applicant has reana-b lyzed what he believes to be the limiting structures and components
) using this new response spectrum and has concluded that all Seismic 5 Category I structures will withstand the design basis earthquake.
Work by the Applicant is continuing.
The Committee believes that specific attention should be given to the seismic capability of the emergency AC power supplies, the DC power supplies, and small components such as actuators and instrument lines that are part of the decay heat removal system.
This matter should be resolved in a manner satisfactory to (theNRCStaff. The Committee wishes to be kept infomed.
In its Safety Evaluation Report dated February 1982, the NRC Staff has identified a number of Unresolved Safety Issues as being applicable to Clinton as well as a number of Outstanding Issues, Confimatory Issues, and License Conditions.
We believe that if due consideration is given to these matters and to our recommendations above, and subject to satis-factory completion of construction, staffing, and preoperational testing, there is reasonable assurance that the Clinton Power Station Unit I can be operated at power icvels up to 2894 MWt without undue risk to the health and safety of the public.
g Sincerely,
\\.
P. Shewmon C
Chairman References T.
lil~iWs Power Company, et al., " Final Safety Analysis Report, Clinton Power Station Units 1 and 2" with Amendments 1-12.
2.
U.S. Nuclear Regulatory Commission, "Saf 3ty Analysis Report Related to the Operation of Clinton Power Station Unit 1,"
NUREG-0853, dated February 1982.
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[Q Cth%o, UNITED STATES Q*
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^j NUCLEAR REGULATORY COMMISSION 2
ADVISORY COMMITTEE ON REACTGR SAFEGUARDS
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WASHINGTON, D. C. 20 55 August 11, 1981 O
SS The Honorable Nunzio J. Palladino Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555
SUBJECT:
REPORT ON ENRICO FERMI ATOMIC POWER PLANT UNIT NO. 2
Dear Dr. Palladino:
During its 256th meeting, August 6-8, 1981, the ACRS completed its review of the application of the Detroit Edison Company (Applicant) for a license to l
operate the Enrico Fermi Atomic Power Plant Unit No. 2 (Fermi-2). A Subcom-mittee meeting was held in Washington, DC, on July 24, 1981 to consider this proj ec t.
A tour of the facility was made on July 15, 1981.
During its re-view, the Com:nittee had the benefit of discussions with representatives of I
the Applicant and the NRC Staff.
The Committee also had the ber.efit of the documents listed. The Committee reported on the construction pennit applica-tion for this unit in its report dated March 9,1971.
i The Enrico Fermi plant is located in Frenchtown Township, Monroe County, Michigan.
The nearest population center is the city of Monroe, Michigan about 5.5 miles west-southwest of the site.
Fermi-2 is equipped with a General Electric BWR-4 nuclear steam supply system with a rated power leve? of 3292 MWt and has a Mark I pressure suppression containment with a design pressure of 62 psig. The Applicant has performed a detailed evaluation of the containment's ability to withstand LOCA and relief valve hydrodynamic loads as required by the NRC for the Mark I Containment Program.
As a result of this evaluation, extensive modifications were required and are underway. However, since the evaluation was perfonned prior to the issuance of the NRC report delineating the Staff's acceptance criteria (NUREG-0661 - Safety Evaluation Report, Mark I Containment Long-Term Pro-gram - Resolution of Generic Technical Activity A-7), the design has not yet been shown to be completely in confonnance with this report.
The Applicant has made a commitment to perfonn a plant unique analysis on the basis of the NUREG-0661 criteria and other requirements established by the Long-Ters l
Program, including in-plant confirmatory tests to assess loads resulting from safety relief valve operation. The Applicant will submit this analysis to the O'
Staff for audit review upon its completion.
Subject to the results of this analysis, the NRC finds the App 1tcant's evaluation generally acceptable.
This matter should be resolved in a manner satisfactory to the NRC Staff prior to full power operation. We wish to be keptNinformed.
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Honorable Nanizio J. Palladino August 11, 1981 O
We note,that Detroit Edison has acted as its own architect-engineer for this proj ect.
The Applicant stated that this arrangement will result in a valu-able carry-over of knowledge as people transfer from construction to plant operation activities.
The NRC Staff has reviewed t,be Applicant's organi-zation and management structure and has expressed some concern about the personnel transition. The Staff recommends 1 hat care be taken to assure that quality of construction and safety of cperations are not compromised during the transition. We concur in this recommendation. To address a concern over a lack of commercial nuclear power plant operating experience, the NRC Staff is requiring that the control room staff be augmented with vendor personnel during startup. We recommend that the NRC assure that these personnel remain on site for a period of time which permits the necessary operating experience to be obtained by the Applicant's Staff.
Tne Applicant described the program and the philosophy for training of personnel. Training has a high priority and a training simulator has been ordered to aid in this effort.
The simulator will be used for operator training and will also be used to train other plant personnel including managers and supervisors.
It will also be used to test ATWS operating procedures.
The NRC has reviewed the Applicant's ATWS procedures and finds them generally acceptable.
The NRC should assure that the ATWS procedures and the associated simulator training are well coordinated.
The Applicant discussed provisions to address station blackout. In the event of a loss of all offsite AC power and loss of all onsite emergency diesel generators, the Applicant can call on a sel f-starting turbine-generator located onsite. While we recognize that this additional power source further lowers the robability of a station blackout, we recommend that the NRC Staff assure that procedures exist to address a station blackout event and that operating personnel are adequately trained in the use of these procedures.
We wish to be kept infonned.
Construction of this unit has taken a longer than usual time owing to fi-nancial difficulties and the impact of the TMI-2 accident.
4 a result, the
- Applicant has been required to perform a seismic reassessment of the struc-l tures, systems, and components required for safe shutdown based on currently accepted NRC design response spectra.
This reassessment is still under way.
Preliminary results indicate that there is sufficient margin in the original l
designs to meet the NRC requirements and that only minor equipment changes will be required.
This matter should be resolved to the satisfaction of the NRC Staff.
O The NRC has begun review of the Applicant's emergency planning.
Because of the plant's location, interaction with Canadian authorities is neces-sary.
Responsibility for this interaction rests with the offices of the Federal Emergency Management Agency.
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Honorable Nunizio J. Palladino August 11, 1981 0
The NRC Staff proposes to require the installation of core thermocouples in Fermi-2 as specified by Regulatory Guide 1.97, Revision 2, "Instrumenta-tion for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident."
The Applicant has not yet agreed to this requirement.
The ACRS supported use of core thermocouples in BWRs in its letter of November 10, 1980 to the NRC Executive Director for i
Operations, but called attention to the need for further study to detemine the appropriate vertical location of such them9 couples.
Since most of the i
information of interest froc thermocouples may be obtainable from a small number of thermocouples placed in a more accessible location, we recommend that this requirement be reevalurted.
The Applicant's security plan was discussed.
We note with approval that security guards will be Detroit Edison employees.
As part of the NRC Staff review of plant fire protection provisions, the Applicant simulated a control room fire to demonstrate that a fire external l
to the control panels will not result in a loss of redundant shutdown func-tions.
The NRC Staff has identified what it believes to be deficiencies in the test and the Applicant has responded in a recent sub@tal. We believe this item should be resolved in a manner satisfactory to the NRC Staff.
Other issues have been identified as Outstanding Issues in the NRC Staff's Safety Evaluation Report dated July 1981. These include some TMI Action Plan requirements.
We believe these issues can be resolved in a manner satisfac-
{
tory to the NRC Staff and recommend that this be done.
The Committee believes that if due consideration is given to the recommenda-tions above, and subject to satisfactory completion of construction, staff-ing, and preoperational testing, there is reasonable assurance that the Enrico Fermi Atomic Po eer Plant Unit No. 2 can be operated at power levels up i
to 3292 MWt without undue risk to the health and safety of the public.
Sincerely, i
J. Carson Mark Chairman i
IiO "ereraaces:
j l.
Detroit Edison Coapany, "Enrico Fernf Atomic Power Plant Unit 2 Final Safety Analysis Report," Volumes 1 - 11 and Amendments 1-37.
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2.
U.S. Nuclear Regulatory Commission, " Safety Evaluation Report Related to -the Operation of Enrico Fenni Atomic Powar Plant Unit No. 2," USNRC l
Report, NUREG-0798, dated July 1981.
3.
U.S. Nuclear Regulatory Commission, " Safety Evaluation Report, Mark I Containment Long-Term Progra.a - Resolution of Generic Technical Activity A-7," USNRC Report, NUREG-0661, dated July 1980.
I
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION I
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASH!NGTON D. C.20555
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March 18,1981 Honorable Joseph M. Hendrie Chaiman U.S. Nuclear Regulatory Commission Washington, DC 20555
SUBJECT:
REPORT ON VIRGIL C. SUMMER NUCLEAR STATION UNIT 1
Dear Dr. Hendrie:
During its 251st meeting, March 12-14, 1981, the ACRS completed its review of the application of the South Carolina Electric and Gas Company for a license to operate the Virgil C. Summer Nuclear Station Unit 1.
This proj-ect was considered at subcommittee meetings on February 26-27,1981 i n Columbia, South Carolina, and on March 11, 1981 in Washington, D.C.
A tour o' the facility was made by members of the Subcommittee on February 26, 1981.
During its review the Committee had the benefit of discussions with repre-sentatives of the Applicant, the NRC Staff, the U.S. Geological Survey, and of the documents listed.
The Committee reported on the construction permit application for this plant in a letter to AEC Chairman Schlesinger dated November 15, 1972.
The Summer plant is located in Fairfield County, South Carolina, about 26 miles northwest of Columbia, South Carolina.
The nearest community with more than 1000 residents is Winnshore, about 15 miles to the northeast.
The plant is adjacent to the Monticello reservoir, which provides cooling water for the main condenser, as well as the ultimate heat sink.
The Summer plant employs a Westinghouse, three-loop, pressurized water, nu-clear steam supply system.
The containment is a cylindrical, carbon-steel-lined, prestressed concrete structure having a design pressure of 57 psig.
At the construction permit review stage, some of the ACRS consultants were reluctant to accept the position of the Regulatory Staff and its consul-tants that the-1886 Charleston earthquake could be clearly localized in the Charleston area with regard to recurrence and recommended that a somewhat increased seismic design basis be employed. -The ACRS supported the Regula-t'ory Staff position favoring a safe shutdown earthquake (SSE) acceleration of 0.15g.
However, in separate reports to the AEC dated May 13,1971 and May 16, 1973, the ACRS urged initiation of a seismic research program in-tended to provide a better understanding of the likely causes of earthquakes near Charleston as well as several other areas in the eastern United States.
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Considerable research has since been undertaken in the Charleston area, and an improved understanding of the possible causes of earthquakes in the east-ern United States has been developed.
However, there still exists more than one theory with regard to the source oi the 1886 Charleston earthquake.
Honorable Joseph P. Hendrie March 18,1981 Since the construction parmit stage, a new issue has arisen with regard O-to the choice of seismic design basis; namely, the potential for a moderate earthquake at the site resulting from reservoir-induced seismicity.
The Applicant has studied seismic activity in the vicinity of the Monticello reservoir since it was filled in 1977, and combined the results of those studies with information about the local geology and hydrology in arriving at the conclusion that a maximum near-field earthquake magnitude of 4.0 should be considered in evaluating plant safety.
The NRC Staff and its consultan.ts have concluded that a near-field magnitude of 4.5 should be used.
However, one member of the NRC Staff disagrees with the majority Staff position, suggesting that the available information does not rule
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out a somewhat larger reservoir-induced earthquake, and that a near-field earthquake having a magnitude of 5.0 to 5.3 should be used for assessing seismic safety.
The ACRS consultants agree that there does not exist a very good basis for choosing a specific near-field event, and generally support the use of a near-field magnitude of about five for evaluation of the plant.
Because it is difficult to judge that the probability of significant e'xceed-ence of the original SSE is sufficiently small, the ACRS has requested, and l
the Applicant has provided, information that indicates there is sufficient l margin in the original design to cope safely with accelerations considerably larger than the SSE of 0.15, including those which might arise from a near-9 l
field, magnitude 5 earthquake.
The Applicant's results to date regarding seismic design margin are reassur-ing.
The ACRS recommends that these studies by the Applicant be extended to include all systems and components whose function is important to the assurance of the continuing removal of shutdown heat.
Such studies need not be completed prior to operation of the Summer plant.
The discussions relative to the seismic issues at the Summer Nuclear Power i
l Station raise certain questions that we believe should be addressed.
These questions, which largely pertain to emargency preparedness, include the ability of certain key systems to function after a major seismic event.
Included among such systems are the emergency alarm features to alert the public to an accident in the plant, meteorological and field radiation mon-itoring networks, communications, and emergency evacuation routes.
As a result of the continuing microseismic activity induced by the reservoir, the Applicant has, at NRC request, agreed to continue seismic monitoring for r
at least the next two years.
L'e recommend that the NRC Staff assure that
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the monitoring program is not halted prematurely.
l In its review of the Applicant's organization and management, the NRC Staff r
has identified several areas requiring at(ention, including the size of the
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engineering organization and the adequacy of experience with nuclear power k
reactors within the company, including hands-on operating experience within
Honorable Joseph M. Hendrie March 18,1981
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the operating organization.
The Applicant has taken steps to obtain the services of outside groups to provide additional technical capability for the short tenn while the needed in-house capability is developed.
Care should be exercised that, as part of this effort, sufficient technical breadth and independence exists among the members of the Nuclear Safety I
Review Committee for the plant.
We have previously recommended that probab'ilistic safety analyses be per-formed for all plants in operation or under construction.
We believe that this reco,mmendation is applicable to this unit, but that such studies need not be perfonned prior to licensing of the plant.
During construction of the essential service water intake structure and pump heu:e, settlement well beyond that predicted was experienced.
While the settlement of the structures appears to have halted, the NRC Staff is still evaluating information addressing the stability of the subsurface materials and foundations of the intake structure and pumphouse.
This matter should be resolved in a manner satisfactory to the NRC Staff.
The ACRS believes that, if due consideration is given to the items mentioned above, and subject to satisfactory completion of construction and preopera-tional testing, there is reasonable assurance that the Virgil C. Summer Nu-
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clear Station Unit 1 can be operated at power levels up to 2775 MWt without undue risk to the health and safety of the public.
Sincerely, J. Carson Mark
. Chairman
References:
1.
South Carolina Electric and Gas Company, " Final Safety Analysis Report, Virgil C. Summer Nuclear Station," Volumes I-XX and Amendments 1-22 2.
U. S.. Nuclear Regulatory Commission, " Safety Evaluation Report Related to the Operation of Virgil C. Summer Nuclear Station, Unit No.1,"
USNRC Report NUREG-0717, dated February,1981 3.
Letter from J. Devine, USGS, to R. Jackson, NRC, in response to an NRC request for update on USGS information concerning occurrence of earthquakes similar to the 1886 Charleston event, dated December 30, 1980 4.
Memorandum from A.. Murphy, Site Safety Research Branch, NRC, to R.
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Jackson, Chief, Geosciences Branch, NRC,
Subject:
Rccommendation of Maximum Reservoir-Induced Earthquake at the V. C. Summer Nuclear Station, dated February 6,1980 5.
"Conments from the Palmetto Alliance, Inc., by Michael Lowe on V. C.
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Summer Operating License Application R'eNiew by the NRC Advisory Com-(
mittee on Reactor Safeguards," dated February 26, 1981 6.
" Testimony Before the Advisory Committee on Reactor Safeguards Related to the Virgil C. Summer Nuclear Station," Ms. Ruth Thomas, received February 26, 1981
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UNITED STATES NUCLEAR REGULATORY COMMISSION
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W ASHIN GT 6N. D. C. 20555 I
ADVISORY COMMITTEE ON REACTOR SAFE 2UARDS
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r December 11, 1979 C
Honorable John F. Ahearne Chairman U.S. Nuclear Regulatory Comission 1
Washington, DC 20555 SUa3ECT: INJERIM LO4 POWER OPERATICN OF SEQUOYAH NUCLEAR POWER PIANT, i
LEIT 1
Dear Dr. Ahearne:
During its 236th meeting, December 6-8, 1979, the Committee considered a proposal for interim, low p,wer operation of the Sequoyah Nuclear Power Plant, Unit 1.
At its 229th meeting, May 10-12, 1979 and also at its 228th meeting, April 5-7, 1979 the Committee had considered aspects of the application of the Te'nnessee Valley Authority (hereinafter referred to as the Applicant) for authorization to operate the Sequoyah Nuclear Power Plant, Units 1 and 2.
A tour of the fecility was made by members of the Subcomittee on January 24, 1976 and the application was considered at Subcommittee meetings on March 12, 1979 and on November 5, 1979. During its review, the Comittee had the benefit of discussions with represente-tives and consultants of the Applicant, the Westinghouse Electric Corpora-tion, and the Nuclear Regulatory Commission (NRC) Staff. % e Committee also had the benefit of the documents listed. % e Committee reported on the application for a construction permit for this plant on February 11, 1970.
The Sequoyah Nuclear Power Plant is located on the west bank of the Tennessee River in Hamilton County in southeastern Tennessee approximately 17 miles northeast of the center of Chattanooga, Tennessee. Construction on Unit 1 is essentially complete and construction of Unit 2 is about 90%
complete. Each unit will utilize a four-loop pressurized water reactor nuclear steam supply system having a power level of 3411 MWt and an ice condenser systm enclosed within a free-standing steel contalment vessel which is surroc.n&d by a reinforced concrete shield building. % e ice condenser system is similar to that used in the McGuire Nuc1 car Station and the Donald C. Cook Nuclear Plant. The Applicant has modified the ice condenser system as a result of the operatire experience gained in the Donald C. Cook Nuclear Plant. %e Applicant and the NRC Staff have grade plans to monitor the performance of the ice condenser containnents at the Sequoyah Nuc1 car Power Plant (Generic Item 63 in the ACRS report, " Status
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of Generic Items Relating to Light-Water Reactors: Rep)rt No. 7," dated March 21, 1979). We Committee recommends that such plans be implemented.
p Honorable John F. Ahearne December 11, 1979 The Segmyah Nuclear Plant will utilize 17x17 fuel assemblies. A surveillance program has been developed by the NRC Staff to follow the behavior of these assemblies, and data are being obtained frem several plants not in operation in which such assemblies have been installed for test. Experience to date has been satisfactory. W e Committee wishes to be kept informed of the results of the various 17x17 assembly inspections and test programs now under way.
The Sequoyah site is considered by the NRC Staff to be within the Southern Valley and Ridge tectonic province. W e maximum historic earthquake within this txtonic province is the 1897 Modified Mercalli Intensity (MMI) VIII earthgake in Giles County, Virginia. During the construction permit review, the NRC Staff concluded that a modified Housner response spe:tre anchored at 0.16g was acceptable as the safe shutdown earthquake. Since that time, the NRC Staff has adopted methods which would characterize an MMI VIII earthquake with the more conservative response spectre specified in Regulatory Guide 1.60 anchored at 0.25g.
The Applicant, in response to NRC Staff recomendations, has evaluated the Sequoyah design using a site-specific safe abutdown response spectre developed from North American and Italian strong m tion records of appro-priate magnitude and epicentral distance and has compared the probability of the safe shutdown earthquake being exceeded at Sequoyah to that at C,-}
other Tennessee Valley Authority plants that meet the Standard Review Plan. It has been concluded that the risk of exceeding the present design spectrum and the risk of exceeding the site-specific spectrum are comparable and that the probability of exceeding the safe shutdown earthquake is not appreciably different from that for other plants in this region. The NRC Staff has reviewed the Applicant's evaluation and has concluded that the Sequoyah plant is adequate to withstand the effects of the safe shutdown earthquake without loss of its capability to perform required safety l
functions. W e NRC Staff, to verify their judgments regarding structural and companent design margins, has performed an audit of the design margins l
in representative critical sections of the reactor and auxiliary building structures and in representative components required for safe shutdown.
The Comittee recomends that this program for the quantification of the seismic design mrgin be continued and expanded to the extent necessary to ensure that all structures and equipnent necessary to accomplish safe shutdown do indeed have some margin. Similar recorxmndations have been made by the Comittee for the North Anna Power Station, Units 1 and 2, and the Davis-Besse Unit 1 in its reports dated January 17, 1977 and January 14, i
1979. % !s matter should be resolved on a schedule and in a m nner satis-l factory to the Staff.
The Dnorgency Core Cooling Systems (ECCS) for the Sequoyah Nuclear Plant incorporate the Upper Head Injection (UHI) system. W e NRC Staff has completed its review of the Westinghouse Electric Corporation ECCS eval-uation model for plants equipped with UHI, and the Committee in its April
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12, 1978 report on the McGuire Nuclear Stattbn has concurred with the
5 Honorable Jonn F. Ahearne Decenber 11, 1979
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1 Staff's conclusions. % e NRC Staff has completed its review of the application of this approved evaluation model to the Sequoyah Nuclear Plant and goncurs with the Applicant.
The Committee has been reviewing the circumstances relating to the recent accident at the Three Mile Island Nuclear Station Unit 2 and has made recomendations for improvements in plant design and operating procedures which should be considered for all pressurized vater reactors. Se Comittee is continuing its review of the implications of this accident and expects to provide additional recomendations. It is expected that these recormfendations will be considered and implemented as appropriate by the NRC Staff. % e Committee wishes to be kept informed.
The NRC Staff has identified a number of outstanding issues, confirmatory issues, and licensing conditions, not related to T.4I-2 accident consider-ations, which have not been specifically addressed in this report. Rese issues should be resolved in a manner satisfactory to the NRC Staff.
Various generic problems are discussed in the Comittee's report, " Status of Generic Items Relating to Light-Water Reactors: Report No. 7," dated March 21, 1979. %ose problems relevant to the Sequoyah Nuclear Plant should be dealt with by the NRC Staff and the Applicant as solutions are found. D e relevant items are: 54-60, 63-65, 69, 71, 72, 74, and 76.
The NRC StafI has not completed its twiew of the Sequoyah Nuclear Power Plant application for a normal operating license at full power, and various implications of the Three Mile Island accident on the Sequoyah Plant remain to be decided. The ACRS has not completed its own review in regard to these matters.
The Applicant has proposed a program of interim low power operation to provide improved operator training and the development of additional ex-perimental information on the behavior of a nuclear unit and its systems under transient conditions. We Applicant has proposed a special test series which includes the following:
1.
A tural circulation following a simulated reactor trip.
l 2.
Natural circulation following a simulated loss of offsite l
power.
3.
Natural circulation with loss of pressurizer heaters.
4.
Effect of steam generator isolation on natural circulation.
5.
Natural circulation at reduced pressure.
6.
Cooldown capability of the charging and let3own system.
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Honorable John F. Ahearne
'4-December 11, 1979 7.
Heat removal following a simulated loss of onsite and ofifsite AC power.
O 8.
bstablishment of natural circulation from stagnant flow conditions.
9.
Baron mixing and cooldown.
The NRC Staff plans to review the proposed experimental program in detail to assure itself that all safety-related aspects cre being dealt with appropriately. The Comittee wishes to be kept informed.
The NRC Staff edvised the Committee that it will require that WA's emergency procedures for Sequoyah be reviewed by Westinghouse. The NRC Staff also stated that an acceptable emergency plan will exist prior to reactor operation.
The Committee believes that there is' reasonable assurance that the Sequoyah Nuclear lower Plant, Unit I can be operated on an interim basis up to power levels of about five percent of full power without undue risk to the health and safety of the public. Subject to approval of the detailed test program by the NRC Staff, the Committee recomends approval of an interim low power license for the purposes proposed.
Sincerely, t
Max W. Carbon 01 airman
References:
1.
Tennessee Valley Authority, " Final Safety Analysis Report, Sequoyah Nuclear Power Plant," Volumes 1 to 13, and Anendments 1 to 61.
2.
U.S. Nuclear Regulatory Commission, " Safety Evaluation Report Related to the operation of Sequoyah Nuclear Plant Units 1 and 2," NURIG-0011, March 1979.
3.
Letter from L. M. Mills, WA, to D. B. Vassallo, NRC, dated October 31, 1979, containing revised responses to the Lessons Learned Requirements.
4.
Intter, L. M. Mills, TVA, to L. S. Rubinstein, NRC, dated October 30, 1979, containing responses to ACRS questions.
5.
Intter from L. M. Mills, WA, to L. S. Rubinstein, NRC, dated October 23, 1979, containing information on natural circulation in Sequoyah, Unit 1, Q
and Diablo Canyon, Unit 1.
- 6., tetter from L. M. Mills, WA, to D. B. Vassallo, NRC, dated October 12, 1979, containing responses to ACRS recommendations.
e l
y 5-December 11, 1979 Honorable John F. Ahearne Letter from L. M. Mills, WA, to D. B. Vassalle, NRC, dated September 7, 7.
1979, containing responses to the Short-Term Recorxnendations of the Lessons I4arned Task Force.
O 14tter from L. M. Mills, WA, to D. B. Vassallo, NRC, dated July 12, 1979, 8.
containing responses to NRC-IEE Bulletin 79-06A and ACRS recommenoaticrs.
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/ p W E:o og UNITED STATES E
g NUCLEAR REGULATORY COMMISSION i
r.
8 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON. D. c. 20555 January 14, 1977 O
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i i
g Honorable Marcus A. Rowden Chairman U.S.. Nuclear Regulatory Comission Wash'ington, DC 20555
Subject:
REPORP (N DAVIS-BESSE hm POWER STATICN, UNIT 1
~ - "
Dear Mr. Rculen:
At its 201st meeting, January 6-8, 1977, the Advisory Comittee on Reactor Safeguards completed its review of the application by the Toledo Edison Conpany and the Cleveland Electric Illuminating Co:tpany for a license to operate the Davis-Besse Nuclear Power Station, Unit 1.
Members of the Comittee visited the plant on May 18, 1976, and a subcomittee meeting was held in Washington, D.C. on December 21, 1976. During its review, f
f the Comittee had the benefit of discussions with representatives and
- d. -
consultants of the Applicant, the Babcock and Wilcox Company, the Bechtel Corporation, and the NRC Staff. The Comittee also had the benefit of
.t the documents listed. The Comittee reported on the application for a construction permit for this unit on August 20, 1970.
The Davis-Besse Nuclear Power Station, Unit 1, is located on the south-western shore of Lake Erie about midsay between the cities of Toledo and Sandusky, Ohio. The minimum exclusion distance is 2400 ft. The low population zone, with a radius of two miles, included about 870 people in 3
the 1970 census. The nearest population centers are Toledo (1970 popula-F tion 383,818) and Sandusky (1970 population 32,674), both about 20 miles from the plant.
?
The nuclear steam supply system employs a Babcock and Wilcox pressurized water reactor similar in most respects to those first used in the Oconee Nuclear Station.
This system differs from the Oconce units other similar units in that the steam generator loops are ra, and several ised about 30 ft above the level in the original plant arrangement. Although this change was made to eliminate the need for internal vent valves, four such valves are provided because of their beneficial effect in reducing steam binding following a postulated loss-of-coolant accident.
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- i Honorable Marcus A. Rowden January 14, 1977 The proposed power level for the unit is 2772 MWt, as co m ared to 2633 MRt proposed at the construction permit stage. This higher power level is the same as that proposed for the Rancho Seco and Three Mile Island,
' Unit 2 reactors, both of which have been reviewed by the NRC Staff and i
the Comittee and found acceptable.
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The structures and cog onents of Davis-Besse, Unit 1, were designed for a Safe Shutdown Earthquake (SSE) acceleration of 0.15g at the foundation 4
l i
level. Because of changes in the regulatory approach to selection of seis-mic design bases, the Comittee believes that an acceleration of 0.20g r-l
- i' would be more appropriate for the SSE acceleration at a site such as this I
in the Central Stable Region. The Applicant presented the results of
)
preliminary calculations concerning the safety margins of the plant for i
l an SSE acceleration of 0.20g. The Comittee recomends that the NRC Staff review this aspect of the design in detail and assure itself that signif -
cant margins exist in all systes required to accoglish safe shu of I
the reactor and continued shutdown heat removal, in the event of an SSE at this higher level. The Comittee believes that such an evaluation need i*
not delay the start of operation of Davis-Besse, Unit 1.
The Comittee wishes to be kept informed.
l-e performance of the Emergency Core Cooling Syste (ECCS) has been 4
evaluated using a Babcock and Wilcox evaluation model applicable R the I ! t raised-loop configuration. The NRC Staff has reviewed these evaluations l-and has determined that certain assumptions regarding return to nucleate t'..
boiling do not comply strictly with the provisions of Appendix K to 10 CFR Part 50. The NRC Staff is also reviewing several other areas
!d relating to ECCS performance. These matters should be resolved in j
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a manner satisfactory to the NRC Staff.
6 4
'/}.
In conjunction with the evaluation and assessment of the impact of routine waste releases from this plant, the Comittee recomends
.s 1
that the NRC Staff provide leadership in encouraging the development of imprcved environmental radiation surveillance capabilities on the part of the State of Ohio and appropriate local regulatory agencies.
The Comittee notes that post-accident operation of the plant to maintain safe shutdown conditions may be dependent on instrumentation and electrical equipr.ent within contain:ent which is susceptible to ingress of steam or water if the hermetic seals are either initially
.,b e
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a Honorable Marcus A. Rowden January 14, 1977
\\ 0 defective or should becca defective as a result of damage or aging.
The Comittee believes that appropriate test and maintenance procedures should be developed to assure continuous long. term seal capability.
The Comittee recomends that, prior to comercial power operation of y
Davis-Besse, Unit 1, additional means for evaluating the cause and likely coarse of various accidents, including those of very low probability, should be in hand in order to provide improved bases for timely decisions concerning possible off-site emergency measures.
to be kept informed.
The Committee wishas
-e The question of whether the design of this plant must be modified in order to comply with the requirenents of NASH-1270, " Technical Report on Anticipated Transients Without Scram (ATWS) for Water-Cooled Reactors,"
remains an outstanding issue pending the NRC Staff completion of its review of the Babcock and Wilcox generic analyses of ATWS.
The Comittee recommends that the NRC Staff, the Applicant, and the Babcock and Wilcox Cogany continue to strive for an early resolution of this matter in a manner acceptable to the NRC Staff. The Co:raittee wishes to be kept f
informed.,
Davic-Besse, Unit 1, has installed a bypass loop containing two manually operated valves around the decay heat removal systen suction lire iso-lation valves.
The normally closed bypass valves would be opened in the event of a spurious closure of one of the decay heat removal system suction line isolation valves during system operation.
The Comittee recommends that further attention be given to the means egloyed for iso-lation of the low pressure residual heat removal system from the primary system while the latter is pressurized, an3 that reliable means be developed to assure such isolation.
i isfactory to the NRC Staff.This matter should be resolved in a manner sat-I
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il The Committee wishes to be kept inforred.
The Comittee sJpports the NRC Staff program for evaluation of fire pro-tection in accordance with Appendix A to Auxiliary and Power Conversion Systems Branch Technical Position 9.5-1, " Guidelines for Fire Protection for Nacicar Power Plants."
high priority to the cogletion of both owner and staff evaluations and recomendations for Davis-Besse, Unit 1, and for other plants nearing com-pletion of construction in order to mximize the opportunity for 1:rproving fire protection while areas are still accessible and changes are more feasible.-
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g NUCLEAR REGULATORY COMMISSION j
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o
WASHINGTON. D. C. 20555 l
January 17, 1977
' O Honorable Marcus A. Rowden Chairman U.S. Nuclear Regulatory Comission Washington, DC 20555 SUa7ECT: REPORP ON NORIH ANNA POWER STATICN, UNITS 1 AND 2
Dear Mr. Rowden:
At its 201st meeting, January 6-8, 1977, the Advisory Cormittee on Reac-tor Safeguards conpleted its review of the application of the Virginia Electric and Power Conpany for a license to operate North Anna Power 1.
Station, Units 1 & 2.
'Ihis project was also considered during a Subcoe-mittee meeting held in Washington, D.C., on January 5, 1977. The Com-mittee previously conpleted a partial review of this project at its 198th meeting, October 14-16, 1976, as discussed in its report to you, dated O'
October 26, 1976. During its review, the Comittee had the benefit of discussions with representatives and consultants of the Virginia Electric and Power Conpany, the Westinghouse Electric Corporation, the Stone and Webster Engineering Corporation, and the Nuclear Regulatory Comission, (NRC) Staff. The Conmittee also bad the benefit of the documents listed.
In its report of October 26, 1976, on North Anna, Units 1 & 2, the ACRS had not conpleted its review of the adequacy of seismic design bases and seismic design; loss-of-coolant accidents and emergency core cooling; quality assurance and control of on-site fabrication and installation; asymmetric loads on pressure vessel structures arising from certain pos-
?e tulated pipe breaks; and plans for upgrading protection against fires.
[ The NRC Staff has now conpleted its review of the Stafford fault zone and concluded that the available geological and seismological information supports the conclusion that the Stafford fault zone is not capable with-irs the meaning of Appendix A to 10 CPR Part 100, and that the available information does not warrant any change in the previously approved seismic design bases for North Anna 1 and 2.
Representatives of the U.S. [
Geological Survey concurred that there exists no definitive information showing significant movement during the last million years and that the O
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fault is not capable. Consultants to the ACRS concur with this interpre-tation. hhile they generally find the current design bases acceptable for 6
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Honorable Marcus A. Rowden January 17, 1977 y
the already constructed North Anna plants, they have recon-ended that, f O
in, view of the uncertainties of knowledge concerning the sources of eirthquakes in the Eastern United States, a minimum safe shutdown earth-quake (SSE) of 0.2g acceleration should be utilized for new plants for I
which construction permit applications are submi,tted in the future.
ne Applicant presented partial informhtion concerning the calculated Y
safety factors during safe shutdown earthquake conditions for some of the engineered safety features. The Comittee recomends that the NRC [
Staff review this aspect of the design in detail and assure itself that significant margins exist in all systems required to acconplish safe shutdown of the reactors and continued shutdown heat removal, given an SSE. The Comittee believes that such an evaluation need not delay the start of operation of North Anna 1 and 2.
The Comittee wishes to be j
kept informed.
The NRC Staff has now conpleted its review of emergency core cooling t.-
system performance and found it to be acceptable. The Comittee con-curs.
The NRC Staff has conducted and is continuing extensive investigation of construction activities of North Anna Units 1 and 2.
These investi-(
gations have been separated into four phases:
i 1.
16vestigation of specific allegations made by three j },'
individuals of faulty construction practices; 2.
a detailed inspection of certain safety-related piping not J.
directly implicated in the original allegations but whid was potentially subject to similar problems; s
3.
detailed monitoring of the nondestructive preservice l
baseline-examination of selected welds in safety-related piping by the Licensee and his contractors; and i
4.
inspections of the performance of selected conponents in specific piping systems during the preoperational testing program.
The NRC Staff has concluded that various items of non-conpliance with NRC requirements have occurred and has defined a program to remedy the matter.
O The Comittee has had the benefit of a review and evaluation of this matter by its wn consultant, who supports the adequacy of the NRC e
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Honorable Marcus A. Rowden January 17, 1977 investigations and has made several r'ecomendations, including one O
rel'ated to a program to ascertain that significant deficiencies do not exist in safety related piping systems. The ACRS concurs. 'Ibe Committee wishes to be kept informed regarding resolution of these recomendations.
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W ne NRC Staff has reported that the matter of asymetric loads on pres-sure vessel structures is essentially resolved. The ACRS has had the benefit of meetings of an Ad Hoc Working Group on this general subject,
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in Toronto on August 5,1976, and in Ies Angeles on December 1,1976.
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The Comittee agrees that, subject to final evaluation by the NRC Staff, f
this matter is in an acceptable status for North Anna 1 and 2..
The Applicanti is in the process of studying fire protection measures at the plant in accordance with the guidelines of Appendix A to Auxiliary and I
Power Conversion Systems Branch Technical Position 9.5-1.
The NRC Staff
' (' Q has stated that, as a plant about to come into operation, North Anna 1 and 2 will be given priority in the evaluation of fire protection matters, and
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of operation on the second fuel cycle. Tae Comittee finds this approadt that most, if not all inprovements will be inplemented prior to the start j-
'l' to be acceptable.
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- O ae Co-ittee notes that rest-accideae overation of the 91 ne to aint in safe shutdown conditions may be dependent on instrumentation and electrical ft equipment within containment which is susceptible to ingrees of steam or water if the hermetic seals are either init;ially defective or should be-I co
- re defective as a result of damage or aging. The Comittee believes
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that appropriate test and maintenance procedures to assure continuous long-term seal capability should be developed.
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The ACRS believes that, if due regard is given to the items mentioned h
above and in its report of October 26, 1976, and subject to satisfactor l
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co:rpletion of construction and preoperational testing, there is reason y t
able assurance that the North Anna Power Station, Units 1 and 2, can be operated at poser levels up to 2775 IGt without undue risk to the health and safety of the public.
l' S0cerelydyo's,e. 1 ~
1-p1.
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M. Bender Chairman op 5
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- 2 Honorable Marcus A. Rowden January 17, 1977 4
l Attachment Report of W.R. Gall, ACRS Consultant, dated January 3,1977,
Subject:
Review of Allegations and Inspectors Findings as Reported in NBC In-vestigation Report $50-338/76-28, 50-339/76-16 I
North Anna, Units 1 and 2.
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REFERENCES:
k' 1.
horth Anna Power Station, Units 1 & 2 Final Safety Analysis Report,'
with Amendments 1 through 60 2.
Safety Evaluation Report (NUREG-0053) related to operation of North
~h Anna Power Station, Units 1 and 2, with Supplements 1 through 5.
Virginia Electric and Power Conpany (VEPCD) letter Serial No. 338 3.
to Mr. Benard C. Rusche, ONRR, NRC, dated Novenber 24, 1976, on environmental testing'of safety related instrumentation.
}
I"4 4.
VEPCO letter Serial No. 350 to Mr. Benard C. Rusche, ONRR, NRC, dated November 30, 1976, forwarding a document entitled, " Safety Related l?
Equipment Tenperature Transients During the Limiting Main Steam Line 2 '.l, -
ii Break,"
5.
VEE letter Serial No. 346 to Mr. Benard C. Rusche, ONRR, NRC, dated
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Novenber 30, 1976, on measures considered for use at North Anna te overpressurization events.
I 6.
VEPCO letter Serial No. 316A, dated December 3,1976, re model testing l[t of LRSI punps.
!l-7.
VEPCO letter Serial No. 298/102276, dated December 16, 1976, contain-ing information on IOCA effects on reactor fuel. (Westinghouse P10-
'?
PRIETARY).
Hw 8.
NRC letter of Decent >er 14, 1976, from D.B. Vassallo to Dr. Dade W.
I c.
Moeller, Chairman, ACRS, subject " Staff Report - Assessment of the l
Stafford Fault Zone."
3 9.
NRC meno dated Decenber 2,1976, from Dudley Thonpson and Boyce H. Grier
, if to Ernst Volgenau, I&E, subject, " Transmittal and Evaluation of In-t vestigation Report, No. 50-338/76-28, 50-339/76 North Anna Units 1 and 2."
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- 10. VEPCO letter Serial No. 371, dated December 9,1976, forwarding a copy, of VEPCD's reply to E. Volgenau re I&E Investigation Report Number 50-338/76-28 and 50-339/76-16.
- 11. NRC letter dated December 6,1976 from E. Volgenau, ISE, to VEP00 Atta: Mr. T. Justin Moore, President referring to the ISE investi-gation of construction activities at North Anna 1 and 2 forwarding a " Notice of Violation", and a " Notice of Proposed Inposition of Civil Penalities."
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Honorable Marcus A. Rowden
-5r-January 17, 1977 4
REFERENCES (con't)
Subject:
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' " Investigation of alleged discrepancies in the construction and quality control program for piping installation at the North Anna Power Station."
- 13. VEPCO letter serial 390 to Dr. Dade W. Moe11ef, Chairman, ACRS, for-warding a copy of Mr. T. Justin Moore's letter of December 23,1976 to l
M Dr. Ernst Volgenau re the North Anna investigation.
- 14. VEPCD letter Serial No. 391, dated January 4,1977, providing infor-t mation re concerns related to auxiliary power and containment systems.
- 15. North Anna Environmental Coalition (NAEC) letter dated January 5,1977, I',*..
to Dr. Dade W. Moeller and Dr. David Okrent, ACRS, requesting that certain items be made a part of the record of the January 6-8, 1977, ACRS meeting.
- 16. NAEC letter dated January 7,1977, to Dr Dade W. Moeller and Dr. David j
Okrent, ACRS, adding two additional items to the list submitted in the NAEC letter of January 5,1977.
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