ML20236P826
| ML20236P826 | |
| Person / Time | |
|---|---|
| Issue date: | 08/07/1987 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | |
| References | |
| ACRS-T-1605, NUDOCS 8708130070 | |
| Download: ML20236P826 (180) | |
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1 e ORG'NA- UNITED STATES NUCLEAR REGULATORY COMMISSION 1
a IN THE MATTER OF: DOCKET NO:
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 328TH GENERAL MEETING f
O LOCATION: WASHINGTON, D. C. PAGES: 1- 135 DATE: FRIDAY, AUGUST 7, 1987 nlj j -
I $ 7J LES si >- s -
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ACE-FEDERAL REPORTERS, INC.
O Official Reporters 444 North Capitol Street
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NATIONWIDE COVERAGE ,
CR31970.0 BLW/sjg 1 1 UNITED STATES OF AMERICA O 2 NUCLEAR REGULATORY COMMISSION j ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 3
328TH GENERAL MEETING 4
1 5
3 Nuclear Regulatory Commission 6 Room 1046 1717 H Street, N.W.
Washington, D. C.
7 g 8 Friday, August 7, 1987 9
The 328th General Meeting raconvened at 1:00 p.m.,
10 Dr. William Kerr, chairman, presiding.
11 )
ACRS MEMBERS PRESENT:
12 13 DR. WILLIAM KERR DR. FORREST J. REMICK 14 MR. JESSE C. EBERSOLE DR. HAROLD W. LEWIS 15 MRlCkRL E MICHELSON DR. DADE W. MOELLER 16 DR. DAVID OKRENT MR. GLENN A. REED 17 DR. PAUL G. SHEWMON DR. CHESTER P. SIESS 18 MR. DAVID A. WARD MR. CHARLES J. WYLIE 19 20 21 22 23 llh 24 25 ACE FEDERAL REPORTERS, INC.
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h PUBLIC NOTICE BY THE UNITED STATES NUCLEAR REGULATORY COMMISSIONERS' ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 1
~
FRIDAY, AUGUST 7, 1987 l
The contents of this stenographic transcript of the '
proceedings of the United States Nuclear Regulatory Commission's Advisory Committee on Reactor Safeguards (ACRS), as reported herein, is an uncorrected record of the discussions recorded at the meeting held on the above date.
No member of the ACRS Staff and no participant at h this meeting accepts any responsibility for errors or inaccuracies of statement or data contained in this transcript.
O
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01 01 2 BLWbw 1 PROCEEDINGS 2 (1:00 p.m.)
3 DR. KERR: The improved Westinghouse Standard l 4 Plant. Mr. Ward.
5 MR. WARD: I don't have a great deal to say.
6 Briefly, sometime back, a review of the l
l 7 Westinghouse Standard Plant Design was begun by the Staff 8 and the ACRS was participating in that to start. The review
, 9 came to a halt for a couple of reasons. The Staff can 1
10 explain those, if it is necessary to have an explanation.
11 And the reorganization of the Staff has caused, I think, 12 some delay, but now I understand the Staff has a definite
(]
v 13 schedule for proceeding with a review. And that is what we 14 are going to hear today f rom Mr. Kenyon, who is Project i
15 Manager for this review by the Staff. I f
16 And that is all I have to stay, and I invite you (
17 to go ahead, Mr. Kenyon.
18 MR. BERKOW: The Project Director has 19 l responsibility for the review of the licensing applications l
20 j for all the standard plant reviews, including the 21 Westinghouse RESAR SP/90, the GE advanced boiling water 22 reactor final design approval, design certification 4 23 application scheduled for September and the Combustior, q (Jl 24 Engineering Advanced System aided design, which is scheduled 25 to be submitted starting this month, and the EPRI ALWR l
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BLWbw 1 requirements document. This is the first interaction we 2 have had with the committee since the reorganization on this i
3 project, and I am sure we will be down here meeting with you 4 from time to time to go over the status and the progress on 5 all of these projects.
6 Today, as Mr. Ward mentioned, we are here to give 7 you a status report on the review that is going on in the 8 RESAR SP/90, and it is a status report as opposed to a 9 discussion of technical issues or SER content.
10 Tom Kenyon is the PM, and he will be giving that
.1 status report.
12 As was also mentioned, in the past few years, for
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v 13 a variety of reasons, there was very little progress made on 14 this review; howeve r , I am happy to report at this time that 15 l the review is back on track. We have the necessary 16 resources and priorities pretty much to get the job done on 17 the established schedule.
18 In that ragard, there is an interesting twist on 19 ,
the format of the review. As you may recall, this 20 l application was the first application that the Staff ever 21 j accepted in pieces back in 1983, module by module.
22 It was a novel concept at that time. Since then, 23 ljustaboutallofthestandardplantsthatwehaveunder
~
/T 24 review or will have under review are following suit and are
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25 submitting it on a module by module basis.
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l 01 01 4 BLWbw 1 At the time we reactivated this review a few 2 months ago, as it turns out all of the modules were in 3 house, so we are now doing an integrated review, more or 4 less conventional review, of the type that we always do.
5 So we made the complete circle.
6 At this point, I would like to turn it over to 7 Tom, who will give you the status of the project.
8 MR. REED: I believe the House of Representatives 9 night before last passed the standardized plant bill that l
10 they had, and I guess it has to go through the Senate, but 11 in view of the fact that it appears that you might now get a 12 congressional national bill pushing on standardized plants,
(~)T 13 it seems to me, it is very, very important that we l
14 standardized, not incorrectly, but correctly.
I 15 Now I know way back 20 years ago after the Ginna I
16 : Tant was created, Westinghouse says this is our 17 sv_adardized plant, and then af ter -- af ter 12 holes were 18 shot through it, it was changed to something else. It seems 19 ! to me there is a great burden, since in this country, we l
20 l standardize on four plants, four manufacturers, obviously, I
21 that we get this thing done right.
22 A healthy attitude of the Staff is to pursue with -
23 great vigor and intelligence and competence this getting the
~ 1 24 i standardized plant right.
(t )T 25 MR. BERK 0W: How could anybody argue with that?
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YLWbw 1 (Laughter.)
2 DR. SHEWMON: A wise man wouldn't try.
f.
3 (Laughter.)
4 MR. BERKCW: Tnat sounds good to me.
5 With that, I will turn it over to Tom.
6 (Slide.)
7 MR. KENYON: , Good [dternoon.
8 I an' the Project Manager on the RESAR SP/90 9 project. My name is Tom Kenyon.
10 Today,.I intend to go tirough a little bit of the 11 review history on the project and then let you know what our 12 current review schedule is and the current review status of
{}
13 where we are today.
14 In October of 1983,:as w)s pointed out already, 15 Westinghouse submitted their application for a preliminary 16 design apprc/31 for the: advanced PWR. As has been pointed 17 out, the modules that they began submitting, the first 18 module coming in in '83 -- the first two modules came in in 19 b'83. The modules were different in format. Instead of q0' 1 following the normal format of the SRP, they were in and of l 21 themselves supposed to be completely describing the specific l
l 22 system or area.
1 23 For instance, Module 5 was intended to discuss
(} 24 l the aspects of reactor systems in Module 10, discuss the 25 aspects of containment systems.
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BLWbw 1 'Because the format of the submittals were 1
- 2 ~different, the Staff decided to approach the review in a 3 slightly different way.
4 The original approach was for the Staff to review 5 each module as it was received and then to issue a draft SER 6 on the individual models, after which we intended to meet 7 with the ACRS to discuss the results of our review of each 8 module. When all of the modules were submitted and the 9 review process had been gone through, the Staff intended to 10 issue an integrated final SER that would discuss the results 11 of the review and iesue -- make our decision on the PDA.
(}. 12 The entire schedule that we had originally agreed 13 to follow, would have resulted in a final integrated SER 14 sometime in early 1986. Obviously, we haven't met that 15 schedule. No draft SERs have been issued to date, although 16 some SERs on individual sections of modules have been 17 written, 18 Westinghouse made their final submittal in March 19 of 1987, and when the Staff reorganized in April, 20 approximately 50 percent of the review Staff that had been 21 working on the project in the 1984-85 time frame were no 22 longer available to complete their review.
'23 When I tool over in April of 1987, as a result of
(} 24 the reorganization, I realized, since all of the modules 25 were now in and since we had lost so many of the original ACE-FEDERAL REPORTERS, INC, 202-347 3700 Nationwide Coserage mig-3364M6
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BLWbw 1- review staff, that performing module by module. review was no 2 longer necessary nor was it really practical.
3 Some of the SERS that has been written, because 4 they were written in the '84 '85 time frame, had expressed a 5 need to review modules that had not been submitted at that !
6 time, in order to complete the review.
7 Since'all of the modules are now in, we are now k 8 doing what I call an integrated review of the project, much 9 like we have reviewed most of the FSAR and PSARs in the 10 past.
11 We are currently working toward a schedule that l \
12 will produce a draft SER, a draft integrated SER, in April l
l 13 of 1988, at which time the Staff will be ready to discuss l
l 14 the results of our review with the ACRS. We propose to meet l
l 15 with the ACRS sometime in mid-1988, and we expect to issue a 1
16 final integrated SER describ'.g the results of our review 37 and make a PDA decision by the end of '88.
I 18 MR. MICHELSON: How many volumes will be in the 19 draf t SER?
20 MR. KENYON: Pardon?
21 MR. MICHELSON: How many volumes in the SER {
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22 itself? How many volumes?
23 MR. BERKOW: One. {
}
24 MR. MICHELSON: That would be the SER for a 10 or 25 15-volume set of modules. Is it the same modules we are now l
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BLWbw 1 receiving one at a time? We won't have to look at some new 2 material other than the SER?
l 3 MR. KENYON: That is our intention.
4 We are only going to have one volume. I 5 MR. MICHELSON: It will be fairly long.
l 6 MR. KENYON: They're about 400 or 500 pages, l
l 7 because they are supposed to be covering the entire design.
I 8 It will be somewhat.like the typical SERs 9 MR. MICHELSON: Clearly more than one 10 subcommittee meeting would be needed. Probably two or three 11 would be my guess, and they would jump from April to June.
{} 12 That is full ACRS, I had interpreted.
13 MR. BERKOW: Yes.
14 MR. MICHELSON: That's a fast jump; isn't it 15 MR. BERKOW: That is dark target. It could 16 change, as we move along.
17 MR. MICHELSON: I just wanted to understand.
18 Thank you.
19 MR. KENYON: The current review status, as has 20 been mentioned many times today, is that the review has been 21 mentioned many times today is that the review has been 22 reactivated. The review is essentially in a hiatus in the 23 last year or so due to more pressing priorities and resource
{} 24 restraints.
We have established a new review team consisting 25 l
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'2 the draft SERs that were written on the individual modules 3 have been redistributed to the Staff for NRC management 4- concurrence where recurrence and modification is necessary.
5 In some cases, we have already received the concurrence'on 6 some of the -- on the more complete draft SERs. In other 7 case, such as where the reviewers are new to the project,
'8 the reviewers are using these draft SERs as additional input 9 to their review. ;
.10- MR. MICHELSON: All of this work is coming under 11 the standardization policy. That policy has never been 12 issued, has it?
.13 MR. BERKOW: Not yet. There is another paper 14 that is either on its way or has gone down to the
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15 ;
Commission, and there is also some rules being prepared.
4
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16 MR. MICHELSON: Hopefully, this work -- as best j l
17 you know, it is in compliance. ;
18 MR. BERKOW: Yes. It is in compliance; yess 1
19 MR. MICHELSON: Thank you. I j
20 MR. KENYON: Right now, the Staff is preparing {
I l 21 requests for additional information and sending it to l 22 Westinghouse, and I expect the typical technical review l l 23 meetings to begin within the next month or so.
24 MR. MICHELSON: The final standardization policy, l
25 do you tnink it would be approved before the start of this j l
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BLWbw 1 final review? Is that in that time frame?
2 MR. BERKOW: Prior to the completion of the PDA 3 review?
4 MR. MICHELSON: Yes, prior to next April.
5 MR. BERKOW: It depends on whether they put it 6 out for public comment or not. If they decide to put out a 7 proposed policy statement for comment, then I think that 8 they could take some time. There will be a 60-day comment 9 period, and I would anticipate a lot of comments.
10 MR. MICHELSON: I am thinking of the basic 11 question of completeness of design that will be thrashed 12 out, I assume, in a standardization policy, but if you don't 13 thrash it out, and you come to this, and you say the design 14 is incomplete --
15 MR. BERKOW: Except that this is a PDA rather 16 than an FDA, so the completeness of design issue is not 17 quite as important.
J 18 MR. MICHELSON: This would be a two-stop approval I 19 process?
20 MR. BERKOW: Yes.
21 MR. MICHELSON: Thank you.
} 22 MR. KENYON: In summary, what we are basically 23 saying is that the Staff will not be prepared to meet to 24 discuss the results of the review until early to mid-1988, 25 although to some people it seems like going to an integrated Ace FEDERAL REPORTERS, INC.
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01 01 11 BLWbw 1 review process may take a little longer to produce results.
2 I think it is going to prove to be more efficient 3 as well as result in more comprehensive SERs. The SERs, we 4 have said before, will be similar to the ones you have seen 5 in the past on the NTOLs. Perhaps not as much detail.
6 That concludes my presentation.
7 Are there any questions?
8 MR. WARD: I think in order for the committee to 9 do its job on some reasonable schedule, ten months from now, 10 I think we are going to need some sort of interaction at 11 least at the subcommittee level between now and then, either 12 with me or with the applicant or both.
13 I guess we had a briefing three or four years ago 14 on this design as it stood then, but there are a lot of new 15 faces here in the committee who would be involved in it, and 16 I rather expect there have been some changes in the design.
17 So I think we will need some briefings from the applicant i 18 and the Staff between now and next June.
19 MR. HERNDON: The committee, I believe it was the l
20 full committee, had a fairly comprehensive briefing. There I
i 21 !wasaquiteextensivehandout. I think it would be good f
22 background material for the base material, as far as the 23 design features, and so forth. Your other concerns, I guess l
l 24 you probably are looking for something in draft form as soon l
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V. j BLWbw 1 total draft SER. I think we can probably try to accommodate 2 that.
3 MR. .KENYON: The reason that I set up the 1 4 schedule the way it is for a draft SER and the final SER is 1
5 to accommodate the need for the ACRS to review our work. I 6 am not sure what time frame you are meeting, thinking of 7 meeting. My own personal opinion is that at this point it -
8 probably wouldn't be a good idea to meet until sometime in l 9 the beginning of '88.
10 Now with Westinghouse, the design is quite far-11 enough. In f act, I believe they have even started with the
(~ 32 final design, and they may wish to comment on that. I don't 13 know.
14 MR. MICHELSON: How many modules are still i
15 missing't 16 MR. KENYON: They are all in.
17 MR. MICHELSON: I don't recall getting all the 18 modules.
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19 MR. KENYON: There were originally supposed to be 20 16. One of them was combined with another, Modules 6 and 8.
21 MR. MICHELSON: How many should I have now
\
22 MR. KENYON: 14 23 MR. MICHELSON: I will recheck. I don't think I I l
24 have them.
25 MR. KENYON: We can get your copies.
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l g00202 13 BLWbw 1 MR. MICHELSON: Let me check first.
2 MR. BERKOW: The last one came in March, I 3 believe.
4 MR. KENYON: Module 10, March of '87.
5 MR. MICHELSON: That's right.
6 MR. KENYON: Module 14 was not submitted because 1 7 it was a description of the initial test program that 8 decided that it would be better to put that off until the 9 FDA stage, when they were further along in the design.
I 10 MR. MICHELSON: Would they be numbered 1 through i l
11 14, or are there numbers missing?
12 MR. KENYON: 6 and 8 are listed as Module 6 and l']
v 13 8, so you are missing Module 8. And there is -- Module 14 14 is missing. There are four volumes on Module 16 and two
(
15 ,
volumes on Module 5, as I remember. The other thing we j 16 should be sure of is that you have incorporated all of the l
17 amendments. {
l 18 MR. MICHELSON: I don't reall ever getting the !
i 19 amendments. l
} (
l 20 ; MR. KENYON: Let me know what you've got.
21 MR. MICHELSON: There's no way for me to know i
! I 22 what the amendments are. If I have a list of your f
23 amendments.
} I 24 DR. KERR: Why don't you let him know what you l (~
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j 25 don't have. i t
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.BLWbw 1 MR. BERKOW: Perhaps we can provide a list of the
.l 2 --
l 3 MR. MICHELSON: I can't, because I don't know how i
(. 4 many amendments. l l l
[ 5 MR. BERKOW: We will provide you with a list of {
l-l- 6 the amendments.
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L 7 DR. KERR: Any other questions or comments j l' {
8 MR. WARD: I guess I am still thinking that we 9 need for the committee and the subcommittee to get a little I better handle on this process -- maybe early next year.
10 11 Glenn, you started this review. You had the 12 subcommittee. Where do you see us, the state of knowledge 13 of the committee? ;
14 MR. REED: Are you pointing at me as a member of 15 the subcommittee?
16 MR. WARD: As a member and former chairman. I 17 MR. REED: My review goes back to beingi involved 18 as a private industry member of the Westinghouse Utility k
19 Technical Review Committee.
20 MR. WARD: Do you think the committee has hear 21 enough about this design, so that we should just wait for 22 the Staff's SER draft? {
23 MR. REED: No. I think there needs to be more 24 initiation. I don't think the committee has focused on it 25 enough.
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\~ 02 l BLWbw 1 MR. WARD: We will be talking with you. I think 2 you will need to plan something, maybe not next month but 3 maybe sooner than the beginning of next year.
I 4 DR. OKRENT: Is'there a comparison available of ;
)
5 this proposed design, the Sizewell design, the Convoy design 6 and most recent French design that the Staff could bring in 7 and not only show differences but discuss reasons and merits 8 and demerits?
9 MR. KENYON: As far as I know, there isn't one 10 now. Maybe Westinghouse has one. Can you answer that?
11 MR. SHANNON: We have no document that provides 12 specific criteria or comparison between SP-90 and any of the
}-
13 plants you mentioned with the possible exception of 14 Sizewell. We have designed this plant based on our existing 15 designs and then evolved into that design, based on the 16 probabilistic risk assessment basis, design features and 17 enhancements that enhance the safety of this plant.
18 We have not necessarily considered what other 19 countries have done with their designs in the design process 20 of this plant.
l 21 '
DR. SHEWMON: Do you have a document comparing l 22 this design with the last Westinghouse plant that went into i
23 commercial operation and what the changes are?
24 MR. SHANNON: The last ones were either Shearon-
{~}
l 25 Harris or Vogtle's.
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BLWbw 1 DR. SHEWMON: Do you have a document compared .
2 with anything built before?
3 MR. SHANNON: Not a specific document that goes 4 into the kind of detail that I think the committee is !
l 5 looking for. What we have done is taken this as an l 6 evolutionary step from our previous design.
7 DR. SHEWMON: What the hell is it evolving from, 8 when you say you don't have a comparison from anything, but 9 you say it is evolving?
1
( 10 MR. SHANNON: We don't have specific documents.
11 If there is information on specific systems and 12 subsystems --
13 DR. SHEWMON: Pardon me for interrupting your {
14 train of thought.
15 DR. OKRENT: No, you didn't interrupt. You 16 contributed to the discussion. I don't expect to be (
17 reviewing the matter, but if I were, I, myself, think I 18 would benefit from the workmanlike job by both Westinghouse, 19 working par -- not with Westinghouse -- with the Staff, -
20 which showed these comparisons and including, indeed, the 21 last one mentioned by Dr. Shewmon, namely, the most recent 22 Westinghouse design to receive a CP and to understand from 23 Westinghouse.the reasons for changes in design, and I hope, 24 just some results of some PRA calculations with the 25 committee, as we know -- doesn't believe our prima facie ACE-FEDERAL REPORTERS, INC.
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2 And also comments on why features others have 3 found relevant, whether it be the way the Germans build
- 4. their primary system or whatever or.the British or whatever, 5 why these should or shouldn't be in the Westinghouse design.
6 I guess North America and South America, e 7 together, are an island, but there is a world around us.
8 MR. MICHELSON: How do you propose that this 9 design fits into the EPRI work now being done? The design 10 criteria that is supposed to be used for future lightwater 11 reactors? Are those criteria supposed to be incorporated in .
12 this design?
13 MR. BERKOW: No, it wasn't by intent. This k
14 design was submitted, obviously, long before the EPRI 15 requirements document was even started, and while CE and GE 16 claimed that they will, to some extent, adhere to those 17 requirements, Westinghouse, I don't believe, has made ;
18 that --
19 MR. MICHELSON: You're making a connection.
I 20 MR. BERKOW: It is likely that a lot of the 21 Westinghouse design would conform to some of the EPRI 22 requirements, but we don't know that, and we are not
{
23 attempting to make any comparison between the EPRI i
/ 24 requirements document and the Westinghouse --
l 25 MR. MICHELSON: The ACRS has to review this asd a ACE. FEDERAL REPORTERS, INC.
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BLWbw 1 stand-alone case for future use without comparing it to what 2 it ought to be, and so forth, which we felt we were going to 3 do when we reviewed the EPRI requirements document.
4 MR. BERKOW: There was never any intention to 5 have this conform to the EPRI requirements document.
6 DR. OKRENT: But you could at least look again -
7 and'see where they were the same, where they were different -
l 8 and why, and I must confess, I don't understand the staff 9 not automatically doing this, f
10 I would suggest to the committee that they write 11 a letter to the Chairman of the Commission now, a year 12 before whatever it is, you might be reviewing Westinghouse, O
13 urging whatever comparative information you think would be 14 helpful, giving them plenty of advanced notice.
15 MR. BERKOW: For your information, at the present f
16 time, the Staff doesn't intend to make a specific l 17 comparison, even between the GE and CE reviews and the EPRI l 18 requirements document.
1 l 19 We are going to review these documents against -
1 20 the informal guidance standard review plan, whatever other l
! 21 guidance we have.
22 DR. OKRENT: I would put that in the letter too. .
23 MR. MICHELSON: That wasn't the original plan.
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24 MR. BERKOW: The original plan was the schedules {
l l
25 were going to be compatible and GE submittals were going to l ACE. FEDERAL REPORTERS, INC.
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BLWbw .1 follow the EPRI submittals by three months. That is no ;
2 longer the case.
3 MR. MICHELSON: What is the burden on us to look 4 at these things on a case-by-case basis without ever going 5 back and revisiting the fundamental requirements?
6 MR. BERKOW: The EPRI requirements document is 7 really intended more for the next generation of water '
8 reactors than for the generation that we are reviewing now.
9 MR. MICHELSON: I thought this was sort of the
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10 next generation.
f 11 MR. BERKOW: This is an interim -- ,
12 MR. MICHELSON: The next next generation.
13 MR. BERKOW: All of these designed, if not 14 totally completed, were well under way before the first EPRI 15 requirements was written.
16 DR. OKRENT: Look at how it compares with the 17 Yankee Rowe then.
18 DR. KERR: This is not a standard plan .
19 MR. REED: They would be embarrassed, perhaps.
20 MR. BERKOW: It is a standard plan.
21 DR. KERR: I thought standard plans, by 22 definition, were the next generation.
23 MR. KENYON: Westinghouse, as I understand it, is 24 working on the next next generation, as you called it.
25 DR. SHEWMON: One could argue that whatever is ACE FEDERAL REPORTERS, INC.
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L ['30 02 02 20 1 l 'LJ l l BLWbw- 1 sold next is the next generation, but let me ask a different 2 question.
l.
l 3 Do you know how this compares with the Japanese 1
( 4 Westinghouse PWR plant? Is that the next generation l
!- 5 compared to this also?
1 1
j 6 MR. BERKOW: I have to defer to Westinghouse.
7 MR. FONDAFETTI: The submittal to the NRC is l
l l 8 basically identical to the Westinghouse APWR that we are l
l 9 developing for Japan, with a few minor exceptions, where the ,
l 10 requirements are different. We had to make some design-11 modifications to meet NRC requirements.
12 DR. SHEWMON: That puts it one cut ahead of the 4 13 last GE plant they reviewed, as I understand it.
14 Thank you.
15 MR. WYLIE: You say it is basically the same 16 design as the Japanese?
17 MR. FONDAFETTI: Yes; basically the same.
18 MR. WYLIE: Basically?
19 MR. FONDAFETTI: As a very simple example, MITI 20 does not have reactor level instrumentation. Westinghouse 21 SP/90 has reactor level instrumentation. So just those 22 differences, but the reactor vessel steam generation reactor 23 coolant pump safety systems are the same.
24 MR. WYLIE: Is this true of safeguards?
(O'T 25 MR. FONDAFETTI: Again, there are a few, I would l )\CEJFEDERAL REPORTERS, INC.
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2 MR. WYLIE: Minor. -What about trains?
3 MR. FONDAFETTI: The same.
-4 MR. WYLIE: They are identical?
5 MR. FONDAFETTI: Identical diesel capacity is 6 identical --
7 MR. WYLIE: Identical diesels.
8 MR. FONDAFETTI: Yes.
9 DR. SHEWMON: Identical number of welds in the 10- primary system.
11 MR. FONDAFETTI: Yes.
12 MR. WARD: I thought there was going to be a
(~}
s-13 difference in the number of power trains from the review we 14 had two years ago or something like that, I thought there 15 was'a difference.
I 16 MR. FONDAFETTI: The Japanese design has two j
(
17 diesel generators. The design we submitted for SP/90 has l
1 18 two four fluid trains . l l
19 MR. WARD: Thank you. j i
20 Any other questions?
21 (No response.) l i
22 Okay. Thank you very much, Mr. Kenyon. Mr. I l
23 Chairman.
{}
24 DR. KERR: Will you give consideration to Mr.
25 Okrent's suggestion?
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'G BLWbw 1 MR. WARD: Yes.
2 DR. KERR: The next item is entitled " Leak Before 3 Break Criteria." Mr. Shewmon is the cognizant subcommittee 4 chairman, according to my notes.
5 DR. SHEWMON: There was a meeting of the Metal {
6 Component Subcommittee, and before we get onto the item l
7 today, we heard about the revision of 1.99 Reg Guide 1.99, 8 which has to do with the temperature at which primary l I
9 systems can be pressurized and this is under revision now, 10 because of the acquisition or the accumulation of more data, !
l 11 and one point of general interest -- this has not come out 12 yet.
13 And so it has been kicking around the Staff for ;
i i
14 at least a year now, because apparently, they are finding j
{
15 that they will have to raise this pressurization -l
{
16 temperature, or the referenced ductility temperature in many i l
17 of the operating plants, and this is enough to be causing 18 them some concerns or causing the operator some concerns, so j 19 that we will not bring a report back to the committee on 20 today, but we will later on. I i
21 l t
1 22 l l
23 l 24 25 i l
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BLW/bc 1 DR. MARK: What kind of temperature are we to 2 think of?
3L DR. SHEWMON: It starts low and finishes high.
4 It would be lower than 350 because that would get them into 5 trouble with the pressurized thermal shock. The changes 6 have come about because with nickel, they find that nickel 7 has an effect in addition to copper.
8 And if you talk to a good classical metallurgist, 9 he will tell you that nickel is nice because it helps give 10 you toughness through the whole wall of the pressure vessel.
11 And so some vendors I think, CE being one of them, wanted to 12 do what a good, classical metallurgist said would be 13 beneficial.
14 And so they put nickel through them. And now the 15 people, some people are ruined the day that they made that 16 decision, just like they rue the day they put that pretty, 17 shiny copper on the outside of the weld wire, because it 18 wasn't rusty.
19 I think what they are talking about is, before 20 j the end of life, they will have difficulty. And, in 21 general, it will make operation somewhat -- they will have l 22 to heat up the BWR somewhat higher before they can 23 pressurize.
t j
(] 24 DR. MARK: Four-fifty or something like that.
25 DR. SHEWMON: No. Let's say 200, or in there.
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BLW/bc l~ DR. MARK: Kece it hotter than they have ;
i 2 previously felt necessary.
i
'3 DR. SHEWMON: Yes. The main thing that we talked l l
I 4 about had to do with GDC-4. This is the general design rule 5 that has to do with dynamic effects of postulated pine i I
6 breaks in high energy lines.
7 And what we have been doing is to accept the idea l
8 that you can demonstrate that lines would leak before they 9 would break.
10 And if you can do this, then you can get rid of 11 some awkward, heavy equipment in there which makes 12 inspection or observation of the lines difficult.
[}
13 We have reviewed this matter before. And since 14 we felt it would increase plant safety by making things more 15 accessible for inspection, we have approved it.
16 The problems -- and Carl and I were the two 17 members there -- the problems of potential concern have come 18 with matters of what are the sizes of leaks that would now 19 be postulated to establish environmental effects or a static 20 pressurization, or other steam leaks that were credible for 21 compartments and subcompartments.
L 22 The adequacy of the design criteria of supports l
l 23 for heavy components, like steam generators, in place of
{} 24 large seismic events -- and we have written a letter on that 25 before.
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I P O 03'03 25
'V BLW/bc 1 . And what we will hear about today is a summary of 2 the changes that have come about since the last time we 3 heard about this, mostly the public comments that were 4 obtained when GDC-4 went out for review.
5 The staff wants to get the GDC-4 rule and the 6 case they are developing with the public comments out as 7 final. They also have a standard' review plan, 316 or 363, 8 which talks about the leak before break evaluation 9 procedures which people must go through in order to show
{
10 that, indeed, a line qualifies for such special status.
I 11 And these will be developed in the presentation. ,
l-12 Where is the schedule for the agenda for the l N'~)%
13 presentation? Do we have one? Okay, 10.1. And, unless, l
14 Carl, you have some other comments, or are there questions, 15 go achad.
l 16 MR. REED: I'm trying to remind myself, and I '
17 probably can't do it very well, where we stood back a couple -
18 of years ago when this leak before break thing came through 19 and we wrote a letter.
20 I was very concerned when we wrote that letter ;
21 that this could be interpreted. It might be an 22 interpretation that BWRs as well as PWRs could fly -- leak 23 before break treatment. {
l I {
24 And I was assured, no, that wouldn't happen j 1
25 because in the rule or policy, there was this rigorous l
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'v' BLW/bc 1 treatment or evaluation that would take place on every
- j. 2 pipeline with respect to the BWR to decide whether or not it 3 could qualify for leak before break.
4 Now, where do we stand? And what are we trying J
{
5 to achieve here? Are we trying to put the BWR in the same ]
I 6' category as the PWR? j l
i 7 DR. SHEWMON: As I understand the history, and 8 Bob Bosnak can comment on this later when he gets back, they l 9 initially said that they would not allow -- I think they 10 continue to say that they would not allow the application of 1
11 this to pipes in which stress corrosion cracking had been i A ]? observed or was credible.
(.) {
13 That originally excluded any of the BWR primary I l 14 system. Since then, the staff has come out with a revised 15 document on when they think stress corrosion cracking is j i
16 credible and when they are willing to go to a much longer j l
17 inspection period because they think it would be ruled out.
18 And if you do two of three possible things, one 19 being stress adjustment, one being material which is not j 20 prone to this, and a third being a hydrogen water treatment, j i
21 if you do at least two of those, then they're willing to 22 give you this longer inspection period and say that the j 23 stress corrosion cracking is not likely. j 24 And I believe taat, under those conditions, they 25 would then allow a leak before break to be used for those )
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2 MR. REED: If you did two out of three of the l 3 corrective measures.
l 4 DR. SHEWMON: Yes.
l 5 MR. REED: I think in any corrosion, you have to -
l 1
I i
6 have a long history of unf ailed samples before you decide l
l 7 you have a good ally or that you are on safe ground.
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8 Now, it seems to me that, in addition to those l
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l 9 three, you ought to have a long history of unfailed pipes in l
l 10 the field. Or uncracked pipes in the field.
l 11 I'm concerned. I would like to know exactly what l
12 we might approve here. I will be watching it very closely.
l j
13 MR. EBERSOLE: I have a question about the l 14 consideration that appears to diminish as pipe sizes get 15 small, and the thesis that seems to be behind that.
16 If a small pipe breaks, so what? It doesn't l
17 matter because only the following takes place. There is a l
l 18 diminished -- you don't get too much fluid out.
l l
l 19 DR. SHEWMON: It is my impression that leak .
l l 20 before break cannot be applied to lines under about six i
l 21 inches.
l l 22 And the way this rises is that you have to be 1
23 able to tolerate a crack which is so big that it will leak 24 at 10 times the detectable limit and still have a stable
[}
25 crack.
l ACE FEDERAL REPORTERS, INC.
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l O'03 03 28 BLW/bc 1 And as you get down the smaller pipes, that crack 2 has to be just too big to be both stable and provide that 3 large of a leak.
4 And so, as they have gone through and reviewed l
l 5 things, we are always talking about lines that arc above six l
l 6 inches, although they do not call out six inches explicitly l
l 7 in the rule. -
l l 8 MR. EBERSOLE: Is that to say that you consider l
l 9 six-inch line and below can fail?
I l 10 DR. SHEWMON: No. It says that we will assume i
11 that they do fail and, therefore, we must have pipe whip 12 restraints and design for dynamic effects on these. It is 13 only for the bigger ones that they can do this.
l 14 MR. MICHELSON: I think it is sx inches and l
1 j 15 above.
l l
16 MR. BOSNAK: There's no prohibition effectively l
l 17 because of the leakage detection. It's somewhere around six l
l 18 inches.
I
! 19 MR. MICHELSON: Six inches and up.
! 20 DR. SHEWMON: Is that all of the questions then?
l 21 Okay.
I I 22 MR. BOSNAK: I am Bob Bosnak, Deputy Director of 23 the Division of Engineering in the Office of Research.
l 24 The staff has been working on this area for some l
l 25 five years, and this is, we hope, another milestone in what l
ACE. FEDERAL REPORTERS, INC.
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f'70 03 03 29 k) m l l BLW/bc- 1 we think will achieve safer plants. j 4
2 What we are seeking, we are seeking your i
l 3 approval, first, to publish in the Federal Register the 4 final amendment to GDC-4. About a year ago, you saw the l
l 5 proposed rule. The rule itself is unchanged from what you 6 saw then.
7 However, the supplementary information is where 8 we have addressed all the public comments, where we have 9 addressed the CRGR meeting comments, and where we have 10 addressed the ACRS subcommittee on Metal Components.
11 (Slide.)
{} 12 We are going to get into all of those changes 13 during the course of the presentation.
14 The second aspect is we would like a letter of 15 approval from you to issue for public comment, standard 16 review plan section 363. And as you have heard from Paul 17 Shewmon, 363 is, if you like, the recipe for implementation:
18 How do I go about implementing the procedures for a leak 19 before break?
20 Now, we did have a number of public comments and 21 during the subcommittee presentation, I went through all of 22 them. What we are going to do here for the sake of brevity 23 is to dwell on those that resulted in comment and change to
{} 24 the rule.
25 We had some 28 letters. Fifteen dealt with ACE FEDERAL REPORTERS, INC.
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BLW/bc 1 equipment' qualification, 15 of the 28. One of those 15 2 represented 32 utilities.
3 (Slide.) l 4 The first thing I would like to comment on is 5 that we have received several comments that are factors and 6 you heard the factor of 10 on leakage detection capability 7 and a factor of 2 on the leakage -- the unstable crack.
8 That those were far too conservative.
9 The staff considered those and decided that there
.10 was too much uncertainty. And we desired to stay with the 11 factors of 10 and 2.
12 However, we did change the margin on the lows 13 that you use for the crack stability analysis. We had a 14 factor of 1.4 before and we have reduced that to 1.0 if-you 15 use the absolute values.
16 Before it was an algebraic summation with the 17 1.4. So, ef fectively, it has not changed. But that is one 18 change in the process that you have to go through to I i
19 demonstrate leak before break. I
(
I 20 The next point though with, again, the equipment l- 21 qualification. Because we had so many comments and, as I i
}
22 said, 15 of the 28 letterc, and one of those represented a l 23 number of utilities, we.are saying that even though we want i
24 containments to be designed the way they were, we want ECCS 25 to stay the same, we will accept on a case by case basis '
I ACE FEDERAL REPORTERS, INC. l 202-347-37(K) Nationwide Cmerage mk336-6M6 )
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(_) I BLW/bc 1 and, again, the utility that wishes to demonstrate this will !
l 2 have to come in and make an argument based on the ,
1 3 engineering environment in the particular area that it is 4 talking about.
l 5 So we are, in effect, opening the door, if you 6 will, on a case by case basis. One of.the things that the 7 lawyers have commented on when we started with this process, 8 and this includes the limited scope rule, the limited scope 9 rule, if you recall, refers only to PWR main loops.
10 It was that the staff has some inconsistencies 11 and that we are retaining the double-ended pipe rupture for
{} 12 containment design, ECCS, and equipment qualification.
i 13 And, again, in the broad scope rule, this was one 14 of the major set of comments. So we are permitting this 15 and,.egain, the lawyers told us that if we have this on a 16 number of cases, we would probably have to come back and 17 have another amendment to the rule, because we cannot make a 18 rulemaking by exception.
19 But we do have this one area that we have opened 20 up.
21 The next change is the third bullet on this 22 particular slide. Where we said originally that you had to 23 use archival material, saying that if there is available--
24 and we do have procedures in the standard review plan that
}
25 you just can't go out and find information from -- on sale l
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2 You do have to go through a detailed process to l
l 3 show that you have lower-bound information, material '
4' property information.
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.BLW/bc 1 We changed the original 750 creep temperature to i
2 700 for ferritic and 800 for austenitic.
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3 MR. REED:- On the second, I'm not sure if I '
4 understand what you're saying.
5 MR. BOSNAK: We're saying you should use what you 6 always used before for equipment qualification. However, we.
7 will consider on a case by case basis if someone wishes to 8 come in and make an argument for something less than the 9 environment that you would get from a double-ended pipe.
x.
10 break. The staff will consider such an application.
11 ,
MR. REED: I want to relate that again to PWRs 3 12 and BWRs. You're saying that the PWR main loop, and so on,.
(O
". 3 they don't have to go through this, but the PWR up until now l
14 has had to make a case. q 15 MR. BOSNAK: What we're saying is the broad scope 16 rule refers to both Ps and Bs. Broad scope rule is what 17 we're dealing with now. That refers both to PWRs and BWRs,
{
l 18 high energy lines. 1 l 1 l 19 But there is a screening process that you have to i I l l 20 go through to determine that the lines that you are looking i l I 21 at, the high energy lines, whether they are in a P or a B, l l
22 are not subject to a number of things. And one is stress 23 corrosion cracking.
1 24 We have included water hammer, erosion, erosion, L 1 j 25 corrosion, cavitation.
l I
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\) 04 04 BLW/bc 1 We have a list of mechanisms that if they are 2 present in the system, you would expect to have one of those ..
3 phenomenon, then you cannot apply leak before break.
4 MR. EBERSOLE: It includes water hammer?
5 MR. BOSNAK: It does include water hammer.
6 MR. MICHELSON: This is both inside and outside 7 containment. It is outside containment where the 8 environmental -- because of the low level of qualification.
9 DR. OKRENT: How do you show that you will not 10 have a water hammer in the system that you can't show is 11 always full? That's the first question.
12 And I suppose, in the second case, even in a
) {'/}
s 13 system that is full, that there aren't --
14 MR. BOSNAK: It may be very dif ficult. In a case 15 like you mentioned, I don't know how you would be able to 16 show that it is a mean. Nothing, of course, is ever 100 17 percent. You would have to see histories of similar 18 systems.
19 If I were a utility trying to make the argument, I
20 that's what I would do to indicate why I have a line that is 21 not going to be, say, paritially filled, where I'm not going 22 to have valve closure problems.
23 DR. OKRENT: With what confidence does the 24 utility have to make the argument for the staff to accept?
} I 25 You earlier said --
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%-) i BLW/bc 1 MR. BOSNAK: It is extremely unlikely and we are i i
2 in the realm of probability of somewhere in the order of 3 probably 10 to the minus 4, 10 to the minus 6. We don't --
4 the staff does not have at this point in time definitive 1
5 criteria. !
6 What we are trying to point out is -- that was 7 one of the other comments that I think we had before. If we 8 did, we would be able to write a reg guide on the subject.
9 But we don't have that now. But we are pointing 10 out in the standard review plan that if you have a system, a 11 high energy system, and it is prone to these kinds of 12 situations, do not go in and request leak before break.
13 DR. OKRENT: You just used the word " prone", and 14 that is quite different than --
15 MR. BOSNAK: Unlikely.
16 DR. OKRENT: I must confess, from what you've 17 said, I have no way of judging what role water hammer --
18 MR. WICHMAN: We do have definitive criteria.
19 This is based on the resolution of USIA-I water hammer and 20 the NUREGs that emanate from that. There.are a number of 21 operational and hardware recommendations.
22 Those NUREGs that have to be complied with.
l 23 Also, we look very hard at operational history.
(} 24 In Westinghouse reactors alone, you have over 400 reactor 25 years of operation. And we'd look at all of those things l
I ACE-FEDERAL REPORTERS, INC.
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, 36 l 7^)TO
\- 04 04 BLW/bc 1 before determination is made as to whether or not a certain 2 system is subject to severe water hammer and not which would 3 likely degrade the piping to the point where the assumption 4 of a leakage size flaw would be not valid for the leak 5 before break purposes.
- 6. DR. OKRENT: I believe you said 400 years of i
7 history?
8 MR. WICHMAN: In Westinghouse reactors lone, yes.
9 DR. OKRENT: I would argue that is not a very 10 reassuring amount of history. If, in fact, the event could 11 lead to core melt, or even worse, a large release, and I
{} 12 don't know at this stage whether those in fact are included 13 if, first, you have decided that you can't have water hammer .
I 14 and, hence, you can't have the pipe break.
15 And s you modify whatever it is, certain kinds of 16 environmental requirements. And there is a chain of
(
17 dependent failures which are never included. /
l 18 I would say furthermore that those NUREGs did not i
19 include a family of possible transients wherein you could go 20 l from full to partly full. Whether it starts with steam and j l t I
21 then liquid, or liquid to steam, for which history is no l
l l l 22 valid indicator because we are talking about events that are i I {
( 23 relatively rare. But not necessarily known to be one in a I l
l 24 milli < ,n. I 25 MR. W1CHMAN: So rare in fact that I'm aware of i
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BLW/bc 1 all'of the sevore water hammers that have ever occurred,
}
2 that it only breached the pressure boundary into cases. One 3- being Indian Point III many years ago.
4 MR. BOSNAK: Again, on these --
5 DR. OKRENT: When you say " rare", I think that is 6 incompatible --
7 MR. WICHMAN: You said " rare". I didn't.
8 MR. MICHELSON: Extremely unlikely.
9 DR. OKRENT: You can go back and look at the 10 transcript. Nevertheless, to allude to the staff's writeoff 11 of water hammer --
12 MR. BOSNAK: We haven't written it off.
13 MR. WICHMAN: We don't write off water hammer.
14 We look very carefully at water hammer. Operational 15 procedues, hardware and history. I don't know what more one q 16 can do.
17 And if we have any doubts at all, we do not 18 include that piping in leak before break. We are very I
19 careful because this is an emerging issue and we are very 20 careful of how we apply this technology. l 21 DR. OKRENT: Could you refer me to the NUREG 22 where I can find a study of hypothetical situations that 4
23 lead to change in phase?
l
( 24 MR. WICHMAN: I have to defer to Mr. Bosnak.
25 That came out of Research. L. Circus, I believe, was (
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2 MR. BOSNAK: Resolution of A-1, I-think your l
3 point is that you can never totally eliminate water hammer.
1 4 DR. OKRENT: If you go back and look at the l
l 5 committee's-comment at that time, they said there was a need L
6' for study of transient situationn. j h 7 The point I'm making is back then, the committee l
8 didn't accept that the problem was fully studied. And the 9 staff is just merching along as if it was.
10 I won't raise it any more.
11 MR. MICHELSON: Maybe it is worthy to clarify 12 though that during the subcommittee, and maybe you need to 13 repeat here, tell us how you treat the problem, the off 14 normal condition such as loss of off site power, station ,
(
15 blackout, some of these other classical events that we're 16 looking at, from the viewpoint of being potential sources of
(
17 water hammer on restart. f 18 For instance. And are those kind excluded. And 19 if so, why? j
(
20 MR. BOSNAK: Again, on a system by system basis--
21 and I don't think we covered this in the subcommittee.
22 MR. MICHELSON: We certainly raised the question.
23 I think I remember what the answer was. {
{
24 MR. BOSNAK: If there is a probability because of 25 any kind of feature that you might have, including off site l
l
)
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l BLW/bc 1 power, if the system is capable of having a water hammer, l
l 2 then it would be ruled out.
l l 3 MR. MICHELSON: That would be included in the t
I 4 analysis where you are proving it is extremely unlikely.
l~
5 You would have to include such events in that proof process.
6 MR. BOSNAK: I think you would have to go through 7 and include potential events.
8 MR. MICHELSON: That was my understanding. Thank 9 you.
10 MR. EBERSOLE: I want to just mention again the 11 subject for the small pipe failure and the general practice 12 of the panelist to consider commissions for small pipe on a 13 radioactive release context. And then dismiss it if you l
14 didn't get enough curies out or something.
15 The industry is filled with general purpose type 16 I open electrical apparatus in the form of relay boxes, I 17 which is all sorts of things that are virtually fully 18 veitilated. Subject to water from any direction.
19 I don't care whether a pipe is one inch or one 20 j and a half, or even three quarters or whatever, if it's got 21 enough energy in it to pitch water around, then somebody has 22 to look at the potential impact of that, with also the
(
23 attendant philosophy that small valves don't need to have a 24 lot of discipline in testing. j 25 MR. BOSNAK: I just want to point out and cover ACE FEDERAL REPORTERS, INC.
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BLW/bc 1 this at the end that we maintained, besides the comments j 2 that we made earlier with respect to small pipe probably not 3 being able to qualify for leak before break because of the l
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4 leakage detection, but we have kept the leakage cracks that l 5 are currently in standard review plan 362. That has not ]
1 6 .been changed. j 7 And irrespective of whether you have leak before
-8 break or not, the words are in the new standard review plan. I 9 MR. EBERSOLE: Apart from whether -- were they in ;
1 i l 10 the first place adequate to cover this business of system.
l l 11 interaction from the small pipe releases on to this critical 1~
l 12 equipment?
) {
13 MR. BOSNAK: The criteria is adequate. Your l l
14 question gets into how they are implemented. That may be !
j 15 another issue.
l 16 I believe the criteria are adequate. They do 17 cover flooding, the environmental qualifications.
i i 18 MR. EBERSOLE: The ultimate insult, to have a 19 sewer pipe dump water on a scram challenge. That has, in 20 fact, nearly happened. If it hasn't, in fact.
l 21 MR. BOSNAK: A sewer pipe would be -- is not a 22 high energy line.
23 MR. REED: I would like to get clarification on l
{} 24 the 400, where after years of Westinghouse reactor operation l 25 and how it relates to the number of water hammer events.
l
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~BLW/bc 1 I certainly feel that there have been 400 reactor -
2 years on the order of maybe 50-100 water hammer events.in .
3 feedlines.
4 Is that what we are agreeing to?
5 MR. MICHELSON: The subcommittee said 168.
6 MR. REED: QED, it almost seems to me, in 7 feedlines, is still very bad history. And you can almost .
8 rule out --
9 MR. BOSNAK: I do agree with you. I think you 10 will have problems with feedlines, depending how they are i 11- laid out, how they are engineered.
12 MR. REED: Will J-tubes be considered by you?
13 MR. BOSNAK: It's one way of preventing water 14 hammer, but not necessarily a cure. (
15 DR. SHEWMON: You're not allowing anybody --
16 MR. BOSNAK: If somebody came in and said they 17 had J-tubes installed, and that's all I'm going to do, is 18 that enough -- the answer would be no.
19 MR. EBERSOLE: It's not prohibited, per se.
i 20 MR. MICHELSON: Correct me if I'm wrong. There's i
21 nothing preventing me from qualifying a feedwater line for a l
22 leak before break if I can satisfy you of the extremely low 23 likelihood of water hammer.
( 24 MR. BOSNAK: That's right. That's the way it l 25 works. {
l {
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( k) I BLW/bc 1 DR. SHEWMON: And a few other things. ;
i i
2 MR. BOSNAK: There are a lot of other things that ;
1 i
3 are 4n there. )
l 4 MR. MICHELSON: There's i.Jthing excluding {
l 5 feedwater. !
l 6 DR. SHEWMON: I would rather not make the case 7 for getting one through. I 8 Let him get on with his presentation.
i 9 MR. BOSNAK: The next slide contains several i
10 other areas that have been changed. First of all, it deals ;
11 with a new procodure for limit load analysis that was not
(' 12 available at the time the proposed rule was developrd. !
V) 13 There are within the ASME now material that )
14 covers submerged arc wells and shielded metal arc welds.
15 The next bullet deals with the fact that we had a 16 comment saying are you going to apply reg guide 145, outside 17 containment.
18 And we're saying outside containment if you can l
l 19 control it, schedule operator walkdown would be accepted.
20 And that, again, if someone wants to use the reg l
l 21 guide 145, they can do it. But, outside containment, it is 22 not required.
23 If I have piping subject to IGSCC, what do I do?
{} 24 Or can I do anything with it?
l 25 We're saying, on a case by case basis again ACE-FEDERAL REPORTERS, INC.
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'd BLW/bc 1 because we have to look at each one independently, if they 2 have two mitigating methods and one is the either mechanical 3 stress improvement or IHSI, and, secoadly, the water 4 chemistry, hydrogen water chemistry, provided that I started 5 with a piping system that had flaws that are smaller than 10 6 percent, and that is what that paragraph really refers to, l 7 given the idea that I have done this early on in service and l ?
I 8 will not have had larger cracks.
l l 9 Now we have added the last sentence and this was
{
l (
10 discussed at CRGR and we put that into the standard review l
11 plan. It was not in an earlier version. (
{} 12 In fact, if I have repaired pipe by weld l 13 overlays, leak before break does not apply.
l l 14 MR. REED: That whole paragraph doesn't give me l
l 15 much comfort. It seems to me that f rom a history point of i
l 16 view and length of time of service point of view, and length 17 of time of proved testing -- and let's say you're doing 18 laboratory work with the number of samples -- it is not 19 sufficient to allow leak before break application to the 20 BWR.
21 Where has it been used? One place? Hydrogen?
22 What is a sample removal? l 23 MR. BOSNAK: The stress improvement is probably
{} 24 the mitigating treatment that's going to be the more effective, but we want both.
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'}00404 BLW/bc 1 in NUREG 0313.
2 But if you do have compressive stresses on the 3 inside of the pipe, there is good service experience that {
f 4 you are not -- that your crack is not going to grow. It 5 will not progress. '
l 6 And we're not saying that you can do this on any 7 It has to be practically new or virgin pipe 304.
pipe.
8 Again, the material is subject to IGSCC.
s 9 MR. REED: That is an awful liberal 10 interpretation. I think the boiling water reactor people 11 would applaud it, but I wouldn't.
12 I think you just don't have enough history of 13 noncracking and nonpropigation of pipe cracking and proof of 14 materials and proof of environment and stress.
15 You really never know what stress is.
16 MR. BOSNAK: If you perform either IHSI or the 17 mechanical stress, you do have on the inside surface a 18 compressive stress, and that's what you're looking for.
19 MR. REED: Around elbow and everywhere, at all 20 temperatures?
l l 21 DR. SHEWMON: Welds is the only place you have l 22 sensitization, which is another thing you have to have.
l 23 Around elbows is irrelevant.
l
{} 24 MR. WICHMAN: This comes directly from NUREG 0313, rev. two, which this committee has reviewed and 25 I
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- BLW/bc- 1 approved'previously. So you're not seeing anything new 2- here.
3 MR. BOSNAK: We think they're very tight. In 4 other' mrds, we are applying this in a very limited way to-
.5 BWR' piping.
6 DR. SHEWMON: There is a variety of'NRC reports 7 that Mr. Reed can read on this subject, or certainly they
-8 are not limited to NRC.
9 Why don't you go on? If he really wants to read 10 up on-that area, I can help him.
11 MR. BOSNAK: The last bullet is before fatigue'
.12 crack growth-analysis. It is not necessary and it has.been 13: . removed because the crack stability analysis that is 14 required 1to be performed, in effect, bounds the fatigue 15 ' analysis.
16 DR. OKRENT: Would you be willing to go to jail.
17 if a plant were permitted to use the leak before break, the
- l. 18 bullet'on the previous slide, and then it turned out to l:
19 fail? i L 20 MR. BOSNAK: Which bullet are we discussing?-
l
! 21- DR. OKRENT: The BWR application that Mr. Reed i
! 22 was talking about. I'm trying to see what your own level of l
23 not only confidence but willing to accept the buck is here.
i l'
l.
24 MR. BOSNAK: What we're really saying is, first l 25 of-all, if I start with good material --
l t
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BLW/bc 1 DR. .OKRENT: I'm just trying to see whether --
2 MR. BOSNAK: I'm saying yes, if I have done a 1
3 compressive stress improvement on the inside surface of the l 4 line, I will not have crack growth. And so the answer is 5 yes.
6 MR. REED: The people who go to jail, as you 7 notice in Chernobyl, are those at the plant.
8 MR. MICHELSON: One of the things to worry about 9 a wee bit here is if you get the break, there are locations l
10 where that break and subsequent whip can also penetrate l
11 primary containment at the same time.
1 i 12 We haven't put any -- we said protect the primary 1
{-
1 13 containment from the pressure viewpoint, but not from a pipe l
l 14 whip viewpoint, as I. understand it.
1 j 15 Isn't that correct?
16 MR. BOSNAK: No. We're saying as far as 17 containment is concerned, you protect it for all of the 18 dynamic effects. And that hasn't changed.
19 MR. MICHELSON: I missed a big point then. I 20 didn't find the point that if the whip -- if the full 21 , guillotine occurred, and the whip can penetrate containment, 22 that you must take the full guillotine and keep the i
23 restraints?
{} 24 I didn't pick that one up.
It's there, but again we only 25 MR. BOSNAK:
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0 04 04 47 l BLW/bc 1 postulate breaks at finite locations; they are not breaks 2 everywhere.
3 MR. MICHELSON: You go back to the old materials.
1
! 4 The ground rules are that you do postulate such breaks if i
5 they penetrate primary containment, just like if they go off 6 the walls on the outside of containment. There is a similar L
7 rule for'inside containment.
f 8 You take a break there if it's going to penetrate l
(
{
9 containment.
10 MR. BOSNAKt They are finite break locations, and 11 those are the things that are protected against.
p) 12 MR. MICHELSON: It is slowly degraded from the t
{
13 old O' Leary level.
14 MR. BOSNAK: You want to think about the positive 15 effects. We're getting rid of pipe whip restraints. We're-16 getting rid of jet impingement shields. And, later, we will 17 discuss what we are doing with respect to these 2,000 kip {
18 snubbers and very high strength threaded f asteners we are I
19 relying on to do something for us. {
l 20 And I think we are putting our faith in something {
l 21 that is really not going to perform. So there are a lot of l l
22 positive benefits. And some of the things that you are j i
! 23 alluding to are, I think, not as positive. )
I l
24 (Slide.)
l 25 Now, additional comments. The first came about I I
1
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BLW/bc 1 as a result of the staff review. We decided that we would 2 apply this only to class one and class two. We are not 3 applying it to class three piping. And, obviously, we are 4 not applying it to high energy lines that are something less 5 than class three.
6 So this is a further typing, if you will, and the 7 reason for it was, as far as in service inspection is 8 concerned, the only place where you get either surface or 9 volumetric is in the class one, class two systems.
10 So we wanted to have something more than just 11 visual examination.
12 MR. EBERSOLE: I recall in the case of the great 13 big main steam lines that carry 200 pounds or thereabouts, 14 it is critically important that the main steam isolation 15 valves function to isolate discharges from one boiler and 16 not have discharges propigate f rom two boilers.
17 This has led to long arguments about whether, l 18 since these valves generally come out two at a time on the 19 sides of the buildings, whether there should be something 20 like armor plate or something to protect one valve from the 1
21 disruptive effects of something happening in the pipe i
ll 22 nearby. The other line.
l l 23 Is this to say that we will not have to have that l'
24 kind of protection any more?
l 25 MR. BOSNAK: No, we're not getting into that l
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2 MR. EBERSOLE: This is jet impingemet and 3 atmospheric effects.
4 MR. BOSNAK: If a valve was impacted by the 5 dynamic effects of the postulated pipe break and that is now 6 gone, those dynamic effects are effectively'gone.
7 MR. EBERSOLE: So that says two of these 32-inch 8 valves'can be side by side with no isolated provisions, one 9 from the other.
10 MR. BOSNAK: As far as isolation one from another 11 is concerned, you're talking about arrangement in layout.
12 MR. EBERSOLE: Yes. I don't know of any. They 13 might be the lines that emerge from just tube stream 14 generators, and that's all they are.
15 MR. BOSNAK: You're always trying to protect y
16 against things that might happen and the arrangement and 17 layout of the steam line is such that you can remove 18 effectively your isolation capability from an arrangement 19 point of view, and is not a good way to do it.
20 MR. EBERSOLE: And it says point to where the 21 rule is so that I have accounted for the jet impingement 22 effects, and you are writing them off, aren't you? {
23 MR. BOSNAK: We are writing them off with respect f
{
24 to -- l 25 MR. EBERSOLE: Anchors.
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.BLW/bc 1 MR. BOSNAK: That's correct.
2 MR. EBERSOLE: That's all.
3 MR. BOSNAK: The second bullet on this slide is {
4~ the--all of the discussion on heavy component support 5 redesigned. And I have two or three slides that will follow 6 this. That gets into that. That was where the committee 7 wrote a letter. We had three of our consultants that wrote 8 letters with respect to heavy component design.
9 That is what that particular thing is about. And 10 I am going to describe those here in just a few minutes, 11 Now, the next item came about as ' a result of our 12 discussions here two weeks ago with the subcommittee in 13 comparing what is now on the books in standard review plan 14 362.
15 We thought we needed to provide some guidance 16 with respect to pressurization for structural design, the 1
17 structures protecting essential equipment.
18 So that if I have a piece of essential equipment, 19 we don't want to locate a whole number of essential pieces 20 of equipment in the vicinity of a pipe which is now not 1
21 postulated to break.
22 And I don't know if that is something akin to 23 what Jessie Ebersole was talking about. But, again, you l l l 24 want to avoid those kinds of situations. You don't want to l l 25 run a high energy line in the vicinity of the control room l
l I l
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(_I BLW/bc 1 'just because you know now or you are confident that it .is 2 not going to break. You don't make these kinds of what I 3 would call dumb arrangements, decisions, just because you 4 have leak before break.
I 5 MR. WARD: You say you don't do that. How is 6 that controlled?
7 MR. BOSNAK: By the review --
8 MR. WARD: How is it controlled in regulations?
9 MR. BOSNAK: By the review that is given to the 10 layout of high energy and moderate energy piping.
11 MR. WARD: So there is something in the standard 12 review plan that caused --
13 MR. BOSNAK: There is a standard review plan, 14 361, that gives this kind of guidance. And we're talking 15 about, again, we're talking about this application to 16 existing plants, plants that are already built and 17 operating. And for new plants.
18 And I think the kinds of things that you are 19 talking about with respect to arrangement really impact and 20 perhaps they need to be covered for new plants better than l 21 they are now.
l 22 But that is where this arrangement factor has to 1
l 23 be taken care of.
(} 24 In the late sixties, early seventies, what people l 25 used rather than to dictate that they got a poor
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BLW/bc 1 arrangement, they postulated pipe breaks. And as a result of 2 postulating pipe breaks, we ended up with all of this 3 equipment that is in the plants, that we believe do not lead 4 to safe operation.
5 They, in fact, can impede normal operations and 6 they definitely impede in service inspection. So there has 7 to be a better way of doing things rather than to rely on 8 unfortunately postulated breaks. And we don't postulate 9 them everywhere; we postulate them at certain points in the l 10 piping system.
12 DR. SHEWMON: We've got all of these wonderful l 13 seismic issues yet that we haven't even touched on. Let's l
14 get on to the goodies.
15 MR. BOSNAK: Piping repaired by weld overlays are 16 excluded. And one of the CRGR recommendations was that we 17 include something within the standard review plan indicating l
18 that administrative controls may be needed.
19 The idea behind that was that field experience is 20 such that the attitude of some plant managers, at least, to l 21 leakage is -- has not been the best. In other words, there 22 has been no reaction.
23 So, if you have an area where you have postulated 24 leak before break and you have leakage detection, you want 25 to react to it and not put it off.
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.r0 04 04 s 53 LU BLW/bc 1 DR. SHEWMON: Bob, we have to go a little faster l -
2 if we're going to make it.
3 (Slide.)
l 4 MR. BOSNAK: Now we're getting into the area of -l
{
! 5 the heavy component supports. If you recall, we had a j 6 letter from Bob Kennedy. We had one from C.K Chou, and we 7 had one from Spence Bush that indicated that particularly in 1 8 redesigns, and what we are talking about here is redesigns 9 of heavy component supports, and we are not making vast 10 changes in the structural steel associated with the heavy 11 component support.
12 We're talking about two things. We're talking (A
_)
13 about snubbers and we're talking about high strength 14 threaded fasteners. Numbers that we're required as the 15 result of the postultted pipe breaks are in the order of {
16 2,000 kip, somewhere above that.
17 They are very large. They are very difficult to $
l 18 maintain. And they are obviously very difficult to test. !
\
19 And what a number of utilities are anxious to do I 20 is to replace these numbers with smaller capacity, better
)
I 21 engineered, better designed snubbers than they have, and 4 i
22 again to remove the high strength fasteners which, in some j l
23 cases, have failed in place, with lower strengths that are 24 not subject to the stress corrosion cracking because, again,
(].
s j
j 25 the load has gone away. [
}
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BLW/bc 1 And with the limited scope rule, what we are 2 pointing out is that this provision can do this in this 3 limited sense. And it has been in effect since May 6, 1986.
4 5
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7 8
9 10 11 1 12 13 14 15 16 17 l .
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BLW/bc 1 So we are saying that when we have demonstrated l- 2 .in a particular piping system in a particular plant, 3 eliminating water hammer and all the other ideas, that we 4 can go ahead and make this change and eliminate the dynamic 5 offects --
6 MR. REED: Just a minute. I guess that clearly !
i 7 says that you are talking about pump. casings and vessels of 8 all reactors.
9 MR. BOSNAK: We are talking about reactor coolant 10 pumps. We are talking about reactor pressure vessels, the 11 steam generator. We're talking about -- {
l 12 MR. REED: Piping excluded.
13 MR. BOSNAK: Yes, we're taking about large 14 component supports.
15 DR. SHEWMON: If they break loose, then you can 16' bust the piping that way. {
17 MR. BOSNAK: The concern was, if I change out 18 these snubbers, particularly the ones up on the steam 19 generator, and if I have a . seismic event that is beyond my f 20 design basis, that this could cause some problems. And they 21 stem from the fact that some of the work that was done at 22 Lawrence Livermore was done to justify the coupling of SSE 23 and LOCA. And the models that they used assumed as soon as
.- 24 the support failed, that the pipe would also fail; which was l
25 very conservative, obviously. j I
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l l F10 06 06 56 l N) 1 BLW/bc 1 It was not real, but it was satisfactory for the 2 work that they were doing.
3 MR. EBERSOLE: Vhat are you going to.do about 4 those enormous supports you put in at North Anna based on 5 the thesis you get upsets due to instantaneous 6 circumferential failure of the suction line and have 7 overturning moments of the vessel? {
8 Do you recall that plot?
9 MR. BOSNAK: We're getting into what happens with 4
{
10 support such as that. I will get to your question. I think (
11 it's on the next slide.
12 (Slide.)
13 With respect to this, again, redesign, our ,
i l
14 talking about existing plants that have these things, new 15 plants are a little bit different with respect to the 16 snubbers.
17 Now, obviously, you can remove and replace the 18 high strength fastener material much more easily than a I i
19 snubber redesign, where you have to go back and reengineer i
20 perhaps the whole system.
I 21 So we're saying use the existing SSE and use l
22 today's current design requirements, the seismic j 23 requirements for peak broadening for generating the floor i
24 response spectrum. You use what you use today for the 25 inputs, you use what is in the TSAR. J l
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BLW/bc 1 You were talking about snubbers. You have to 2 show that what you are putting in is better in the way of 3 functional reliability. And we have had several situations, 1
4 Trojan is one, where we did have problems with a large l 5 hydraulic anubber that locked during normal operation.
t 6 These are the kinds of things we're trying to 7 prevent. And we're saying the last bullet on the slide:
8 You cannot change out and reduce the structural 9 capacity of the columns, pedestals, hangars, skirts, all of 10 the structural members that are in the plant that support 11 these components. We're only talking about the snubbers.
}
J l 12 MR. MICHELSON: In building a new plant, what do 13 j you say?
14 MR. BOSNAK: I'm going to get to that. That's 15 coming.
16 Again, if you do have pipe ruptures in systems ,
(
17 which don't qualify for leak before break -- and this is 18 true currently in some of the branch lines, and the PWR 19 area, and you do have to go through and look at -- from an 20 independent design and verification procedure -- that you -
[ l 21 I haven't introduced design and fabrication errors.
22 Now, here is the -- this is the key here that 23 we're going to cover when we get to Bob Kennedy's material.
(3 24 But you look at the downgraded, if you will, snubbers that
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25 are left. And you say, even though I expect that they're ace FEDERAL REPORTERS, INC.
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2- I'm trying to show that even if these things f ail l 1
3 in a seismic event well beyond the design basis earthquake, ;
4 and we're talking about something on the order of perhaps 5 two to three times, that I'm not going to lead to a rupture 6 of the primary coolant loop.
7 I can take into account the structures that are 8 there in the facility. I can take into account the 9 stiffness of the piping that is still attached to these l
10 heavy components.
11 I do that and there are several analyses that 12 have been done, and that's why I say we don't have to start 13 from scratch. Quite a bit of work has been done and you can 14 build on that. But you have to go through and demonstrate ;
15 this.
16 This is one of the points that Bob Kennedy had in 17 his presentation here two weeks ago.
l 18 Now, with respect to the high strength fasteners, 19 all you need do is replace them and, again, we're talking l
l 20 about something that was of the order of 200 Ksi. We go 21 down to probably half that ultimate capacity.
22 And, again, you use the original seismic loads in i 23 evaluating the new fasteners.
)
! Carl Michelson's question with regard to future (7 24 l %/
25 plants, we are not saying that a utility has to go through ACE FEDERAL REPORTERS, INC.
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0l06 06 59 BLW/bc 1 all of these procedures. We expect at that time we're going 2 to have in effect the seismic hazard will be updated, so 3 we're not' going to have the plants that exist along the East 4 Coast that have an SSE, a .1 or .15. They will be up in the 5 order of .2, .3, .25.
6 And, therefore, we will have built into the
{
7 design itself the capability of doing what we're talking 8 about here. So that is what we're doing on new plants.
9 (Slide.)
10 Now, Bob Kennedy, again, two weeks ago, rather 11 eloquently presented to the subcommittee, and we tried to 12 get him here for the committee's presentation today, bu c he 13 had already a prior commitment, but what he said, and we are 14 quoting from his slide, that:
15 If I do one of the two things, I can eliminate 16 the pipe rupture loads from the design basis. j
(
17 And this is for new plants, the first bullet. l 18 And the second one, as you recall, the past live, that is I
L 19 exactly, you use the word credible, we didn't use " credible" 20 because we are actually postulating a failure there. (
I: j 21 Perhaps we are a little more conservative than l l l 22 Bob was. But -- and he understood that we would adopt one l
l l
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23 for future plants and the second one for operating plants. j 24 And on that basis, he said that he would prefer 25 to have used the seismic load factor or reducing allowable
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2 However, the staff discussed that and felt that j 3 if we do it in this particular area, it's going to raise l
4 questions about' other components that perhaps have a higher j l
5 risk of leading to core melt. I i
6 So we prefer to follow either of these two {
.7 (indicating). l i
8 MR. EBERSOLE: I noticed in two, you limit that 9 to rupture reactor coolant loop --
10 MR. DOSiiAK: What paper we're concerned about was 11 the reactor coolant loop leading to a LOCA. If we are ;
g 12 talking about some other area where you're not going to have 13 the potential to lead to core melt, the primary thing was 14 that you could lead to core melt.
15 MR. EBERSOLE: Let me go on. Suppose I don't 16 pick reactor coolant loop piping, I pick a main steam line.
17 What that does with these new two-steam generator plants, it 18 permits facing containment with a potential discharge of 19 both boiler contents, both steam generator contents, unless 20 we are successful in intercepting the reverse flow of one of 21 i the two. And that is an active function to do that.
22 You can always argue it is far from perfect 23 because of the valve problems. If you dump the contents of g 24 two steam generators in a containment, I believe the 25 pressure and temperature and other loads exceed that of the ACE-FEDERAL REPORTERS, INC.
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1 2 MR. BOSNAK: But the containment is designed for j l
i 1
3 such things.
l 4 MR. EBERSOLE: The contents of both the steam _)
i 5 generators? Am I correct? Is that correct? l 6 I'm not aware of that.
l l 7 MR. BOSNAK: It is not designed for the content 8 of both of the steam generators.
9 MR. EBERSOLE: That's what I'm saying. There is 10 a new potential brought out by the fact that you must have 11 active equipment which is not reliable, which is 12 sectionalizing valves. At least claim you only get the 13 discharge of one, but you can have two.
i 14' MR. BOSNAK: Well, we said reactor coolant-loop.
15 I think you can also show without too much trouble that the 16 steam'line is also not going to rupture.
)
17 We don't say it there because this was their 18 concern. What I'm pointing out is this -- these are the l
19 comments that we received from the three consultants. They 20 were concerned about --
e+ l l- 21 MR. EB3R3 OLE: Everybody is beyond the primary 1
22 , loop.
g 23 MR. BOSNAK: Loss of coolant accident --
24 , MR. EBERSOLE: It's the old classic. I'm not
, 25 personally sure that secondary rupture with a potential for ACE. FEDERAL REPORTERS, INC.
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f'90 0'i 0 6 62 BLW/bc 1 having contents of the two steam generators discharge 2 doesn't impose a greater risk.
3 MR. REED: I can address that for you and I think 4 clarify _it.
5 MR. BOSNAK: What we have here now as a result of 6 the subcommittee meeting -- and I'm going to go through 7 these rather quickly -- (
8 (Slide.)
9 MR. BOSNAK: Continue. We did say that there are {
10 these other standard review plants plans and this has got to 11 be worked into them.
12 (Slide.)
13 And we are ending up saying: Here is the design 14 criteria. Where you have leak before break, that's what 15 you've got to do. And where you don't have leak before 16 break, that's what you've got to do (indicating).
17 As we said here, what you are eliminating is the j l
18 dynamic effects, the containment ECCS and EQ equipped for q 19 the thing that we talked about earlier, on a case by case 20 basis, do not change. . j l
21 Leakage cracks are still there. And that is what !
l 22 you use for environmental qualifications.
I l l 23 (Slide.) (
! (
24 The pressurization for structural evaluation (
(' 25 involving essential equipment -- and that is required and l (
! ACE-FEDERAL REPORTERS, INC. I
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}D BLW/bc 1 has been placed into 363, the last page of your hondout is 2~ the additions that-we have made to' 363 to cover this.
3 (Slide.)
4 So, here, again, where this CRGR recommendations, 5 you have seen them all before. This one -- these two I 6 didn't talk about, but I think they are self-explanatory.
7 This one, as far as review and approved, we 8 didn't have those words for the limited scope rule. And 9 what the legal staff, OGC, wanted us to clarify is the fact ,
)
10 that scweone has already gone through this exercise. l l
11 The primary loop does not have to come in and do 12 it all over again. And that is what that thing is meant to
- 13. clarify.
14 (Slide.)
15 So to close, again, what we are looking for is a 1 16 letter indicating that we can go ahead and publish in the i
17 Federal Register the final amer.dment to GDC-4 and issue for l 18 public comment the new SRP Section 363.
19 DR. SHEWMON: Thank you. Any other questions? l l
20 MR. MICHELSON: One small question to.make sure I I
\
21 understand a comment made much earlier. l 1
22 If I have a high energy line inside of 23 containment and I apply one of the presently postulated 24 breaks at certain locations, if that break were to cause a 25 whip that damaged containment, would I need to leave the l l
l ACE FEDERAL REPORTERS, INC. !
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BLW/bc 1 support shields in place?
2 MR. BOSNAK: If you can damage containment --
\
3 MR. MICHELSON: It includes the jet dynamic l
4 effects as well as static effects. ;
I 5 MR. BOSNAK: That is correct, that would stay. j 1
6 MR. MICHELSON: I missed that somehow earlier. I l 7 didn't realize you were having the dynamic effects there as 1
8 well.
1 9 DR. SMENMON: Thank you very much.
10 That is all we have. I will bring forward two 11 letters. One will be on the 363, which we think this should
{} 12 go out for public comments; and the other will be on GDC-1, 13 which will be slightly longer, but not a lot.
14 DR. KERR Comments or questions?
15 MR. MICHELSON: The only question on GDC-4 is we l
16 haven't seen the final version as near as I can tell. What ;
17 we reviewed --
18 Mh. BOSNAK: We gave you, Paul, this afternoon 19 the final version, as we publish it. And the only 20 difference from the copy that you have there, there are a 21 couple of -- I call them clarifications to take care of the 1
l 22 OGC problems, but you have it now.
1' l
23 DR. SHEWMON: This is in the handout with the big I
{} 24 three changes that the staff is making are those that are 25 written in by hand.
) /(CE-FEDERAL RFPORTERS, INC.
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! .I 0 07 07 65 BLW/bc 1 MR. MICHELSON: We are sure that those changes l
2 will stay.
3 MR. BOSNAK: Yes.
4 DR. KERR: Anything else?
5 (No response.)
6 DR. KERR: We will take a 15-minute break until 3 7 o' clock. k 8 (Recess.)
9 DR. KERR: Our next session here is Nuclear Power 10 Plant Operating Experience. Mr. Ebersole is our i
11 subcommittee chairman.
(k 12 MR. EBERSOLE: We have here right at the very {
l 13 beginning five events that we are going to formally present:
14 And if we have time left over, we will discuss a few others, 15 peripheral events.
I L
16 These are all based on, first of all,-the July l
(
17 16th telephone conference between myself and Glen Reed on I
{
18 the staff, and subsequently a meeting in Bethesda in which-- I i
19 at which we picked up the events selected today and decided l l
20 to give you a little peripheral information on the other l
21 events, which we may or inay not get to. j l
22 We're going to jump into the meeting here 15 j l
23 minutes early, and it might give us a chance to pick up the j 24 set of fascinating things that have happened in the field in 25 the last two months.
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BLW/bc 1 But, first, we're going to get to the main course 2 events. And I will turn the meeting over to you, Pat, to do 3 that. !
4 MR. BARANOWSKY: I am Pat Baranowsky from the 5 Events Assessment Branch. We do have hot items to present 6 today, and I have shown them on the viewgraph. There is a l
7 list of items and the presenters. j i
8 The first item will be presented by Tom Peebles 9 from Region 110. It has to do wit.h reactor trip breaker {
10 failure at the McGuire plant anv. aome subsequent 11 malfunctions that occurred with similar reactor trip 1
(q./
12 breakers. t
{
13 And with that said, I will start right off with 14 Tom.
15 MR. EBERSOLE: These are the kinds of breakers 16 that are supposed to have no common mode failure source, and 17 it turns out that this may, in fact, be that. They're in j 18 the reactor trip breaker circuits. And we are passing 19 around some photographs given to me that show the complex <
20 nature of this equipment on which we depend on two, one of 21 which must work. l l
22 Go ahead. .
23 MR. PEEBLES: I am the leader for the events that I 1
24 we are going to discuss. There are three other members of f
25 the IT here and a member of the licensee's staff from Duke ACE-FEDERAL REPORTERS, INC.
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BLW/bc 1 Power.
l 2 Initially, we start out July 2, 1986, reactor I l
3 control rod drive time tests were being conducted at McGuire !
4 unit two as part of startup testing following the refueling 5 outage. l l
6 Af ter several rod banks had been successfully ]
7 tested to be reactor trip breaker failed to open during a ;
8 manually initiated trip from the main control panel. ,
i 9 The trip attachment had overheated, shorted and 10 opened. And the -- to the STA circuitry opened. Three 11 operators in the control room stated that the reactor 12 tripper lights show that the reactor trip breaker had g
13 tripped.
14 Subsequent observation at the breaker cabin in 15 the events recorded printout both indicated it had not 16 opened.
17 Failed reactor trip breaker was found when a 18 technician observed that the reactor trip breaker was still 19 closed and there was smoke in the area. Pressing the manual 20 ,
trip break - pressing the manual trip plate on the trip l
21 l breaker did not open it.
I 22 The RTB was subsequently opened by manipulating 23 the manual spring charging handle. It was then removed from 24 the cubicle.
ggg 25 Licensee continued attempts to isolate the ACE FEDERAt. REPORTERS, INC.
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'BLW/bc l' problem and to assure that the plant was in a safe 2' condition. The failed reactor trip breaker was taken to the 3 test room and placed on a_ bench. The inspection of the 4 reactor trip breaker show that the STA coil was ourned and 5 had opened. Reactor trip breaker was then electrically (
6 closed and successfully opened by deenergizing the_under-7 voltage trip attachment.
8 It was electrically closed, mechanically bound 9 and failed to trip during the manual trip force test using a 10 push force gauge. And then continued to duplicate the l
11 original scenarion of the technician raanipulating the handle
, 12 and the breaker opened. {
13 One further test of the RTB for opening was l
l 14 successful. he RTB was then in quarantine. McGuire unit 15 one bypassed -- was then used to replace the failed breaker.
l 16 RTB and associated circuitry was tested.in place l
17 satisfactorily and no abnormal conditions were found. )1 18 The other three breakers on unit two were then !
l i
l 1
l 19 tested and inplaced satisfactorily. Licensee determined J 20 that the problem was isolated to the one reactor trip 21 breaker and the other breakers were operable. )
22 The rod drop tests were then resumed. A
! 23 methodical procedure for inspecting the breaker for the 1
24 mechanical binding in erroneous position indication was 25 developed by representatives from Duke Power, reactor trip j l
l ace-FEDERAL REPORTERS, INC. i 02-347-3700 Nationwide Coverage tun 336-6646 l l
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BLW/bc 1 breaker vendor and the IT.
2 The mechanical inspection revealed a looseness of 3 'the center paw lever of the pole shaft, in particul ar a 4 broken weld between the pole shaft and center pole lever.
5 This broken weld in the general looseness 6 contributed to approximately five degrees of angular 7 rotation and a skewing of the main drive link in its roller 8 with respect to the closing cam in its mounting frame.
9 It was hypothesized that che skewing of the main 10 drive link was such that a good contact on the mounting i I 11 frame could result in jamming of the reactor trip breaker in
{
l I 12 the closed. position. {
- l. 13 This was the probable cause of the breaker's l
14 failure to open and the focus of further efforts of the I
15 investigation.
16 The electrical inspection of the internal breaker
(
17 wiring was completed with no. abnormalities found.
18 Subsequent attempts to duplicate this failure of the trip I
19 were minimally successful.
20 All parties then agreed that dismantling of the {
{
21 breaker and further testing should be conducted under l 22 laboratory conditions. The breaker is to be shipped to the 1
23 vendor's laboratory. There the causes will be further 24 investigated with the licensee in participation with the 25 NRC. .
I l ACE FEDERAL REPORTERS, INC. !
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L P 10 07 07 70
.b u BLW/bc 1 (Slide.)
2 MR. PEEBLES: These are the slides you all have.
3 And, basically, the cause was initially attributed --
l 4 obviously, the long-term investigation by Westinghouse and 5 Duke is going to come up with a better conclusion.
6 But, initially, it is believed to be a 7 combination of high cycle wear and a broken weld, appear to 8 have led to the jamming of the mechanism. The cause of the 9 erroneous status light indication is not known and what is 10 known is that it is not under the single break cubicle and 11 it is being verified.
"s 12 (Slide.)
('J 13 As we talked about the licensee replaced it with 14 a bypass breaker for unit one, the bench test of the B 15 breaker were conducted in duplication of binding was 16 minimally successful.
17 Electrical controls and contacts were properly 18 wired and operating. The smoke was caused by the burnup of 19 the shunt trip coil when the breaker failed to open, and the 20 smoke was not sufficient, even itself, as smoke detector in 21 the room and was minimally noticed by the technician who was 22 in the area.
23 Licensee short-term actions, the licensee and
{} 24 Westinghouse are continuing to investigate these problems.
The licensee, until he understands the problem with breaker 25 ACE FEDERAL REPORTERS, INC.
i 202-347-3700 Nationwide Coverage 800-336-6646
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I l'
e 0 07 07 71 BLW/bc 1 indication, is going to verify that the breaker opened 2 before he resets the breaker.
3 The information notice has been issued.
4 Information number was 8735.
5 The second event that began was on 7/24/87, 6 visual inspection of welds by the licensee during a TM on a 7 Catawba breaker, using lessons learned from the McGuire 8 incident, resulted in notice of several problems with welds-9 on the pole shaft.
10 One weld indicated a small crater crack. One a 11 quarter inch long lack of fusion at the weld that attaches 12 the center pole over the pole shaft.
13 And also there was rubbing on one side of the li- closing cam against its adjacent supporting frame member.
15 IT visited Catawba on 7/30, inspected welds, concurred with 16 the licensee observations. Licensee has disassembled the 17 breaker and is permitting the NRC staff to use the shaft in (
I 18 this briefing. (
l 19 .
Subsequently, the NRC is sending the pole shaft j l \
20 ~ to Franklin Research for evaluation of the welds. Franklin I
{
21 Research will be requested to develop criteria to inspect l 1
22 welds on reactor trip breakers at the other plants. I 23 (Slide.)
24 Ther-e was another event --
25 MR. MICHELSON: You're leaving that one? This ACE FEDERAL REPORTERS, INC.
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V BLW/bc 1 was the first occurrence of this type of failure?
2 MR. PEEBLES: There have been several problems 3 with these breakers in closing. There has not been problems (
4 with this breaker in opening in the past. However, the main 5 drive link, as we will show you, is the single feature in 1 6 the breaker.
(
7 MR. MICHELSON: The failure of the weld before (
- f. - {
8 and that sort of thing?
9 MR. PEEBLES: There has been a failure of the
{
10 weld before but it has not resulted in failure of the 11 breaker to open before.
( 12 MR. MICHELSON: Thank.you. j
\_ j 13 MR. PEEBLES: The McGuire breaker, they were {I 14 performing six-month surveillance tests on the undor-voltage 15 trip attechment. And it is called a trip margin test.
16 The clearance between the under-voltage trip 17 attachment and the trip bar was excessive. With insertion 18 of a thickness gauge of 70/100ths, the breaker did not trip.
19 We took a look at the breaker and obser.ed a repetition of 20 the test and concurred with the licensee observations.
21- And af ter we verified the clearances and the 22 tolerances, we removed that under-voltage trip attachment.
[ 23 There was no abnormal signs of wear on it. This under-24 voltage trip attachment is one that had been installed in 25 1983. 1 ACE FEDERAL REPORTERS, INC.
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r"]O 07 07 73 4 l NJ l BLW/bc 1 When the under-voltage trip attachment had had a 2 problem, new under-voltage trip attachment was installed, l
3 retested and determined acceptable.
I 4 Now, this new under-voltage trip attachment had i
5 considerably better clearances than the other one did. The 1
6 IT reviewed the maintenance history and determined there 7 were no previous problems. The failed under-voltage trip f 8 attachment was one of the first batch of under-voltage trip 9 attachments, in 1983.
10 Tolerances and all the dimensi.>ns will be checked l
11 prior to disassembly of the original breaker at the
( 12 Westinghouse lab. The original breaker had two to three-()3 13 thousand cycles on it. .
14 The Catawba breaker, which has a crack in the 15 weld but it is not all the way broken, had 1,600 cycles on 16 it. Breakers, since 1983, when they went into the every 17 six-months testing and every six-months PM and monthly {
{
18 testing, had been averaging 250-260 cycles a year. {
1 19 l MR. WYLIE: Do you know what the mechanical j l I 20 l design life of the breaker is? l l
21 MR. PEEBLES: We discussed that question a couple i l
22 of times, Jim Thomas f rom Duke Power, I don't remember the l I
23 exact number. l I
(* 24 MR. THOMAS: The information I have seen in the m
i 25 design verification test of the 416 has a breaker, not l
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BLW/bc 1 necessarily as a reactor trip breaker, has verification 2 tests that my interpreta' tion is a cumulative of about 8,000 3 cycles for design life. I think there is a 5,000 life, 4 5,000 cycle test required by ANSI and I believe the original 5- breaker that was tested also had an extra 3,000 cycles on it 6 for some new L-testing.
7 I know the design life is substantially more than 8 the cycles that the breaker had seen at McGuire.
9 MR. PEEBLES: The under-voltage in the shunt are 10 allowed up to 2,500 cycles.
11 MR. THOMAS: That is correct. That is the amount 12 of cycles that was tested in the actual verification test.
13 That design, that was redone in 1983 and '84.
I 14 MR. THOMAS: 250-300 is our estimate.
15 MR. WYLIE: These apparently wore out earlier 16 than 15 years. If they were made properly, they would wear 1
17 out in 18 years.
l 18 MR. PEEBLES: We have the front panel of the l
19 breaker. We just put that up so you can see the manual trip 20 plate. This is the bar. The manual charging bar. You can i 21 see why they were jiggling it. That is, again, a fairly l
l 22 structural part of the breaker and that is why they can open 23 it in that position.
(} 24 (Slide.)
This is the breaker without the front plate on L 25 1
1 14CE FEDERAL REPORTERS, lNC.
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.U f BLW/bc 1 it. I'm not sure what you can see other than the fact that 2 the auxilliary-contacts.are up here. They're fairly beefy 3 contacts; they are not microswitch contacts. They have got 4 a considerable gap between them.
5 When we're looking at the testing, the main 6 contact blades were I believe open 5/8th of an inch before
- l l 7 the breaker indication -,tually changed. {
8 MR. EBERSOLE: I would like to have the committee 9 take a look at that figure to get a flavor of the l
l 10 complexity. There are all sorts of influences that make the l'
l 11 operation less simple and then less reliable.
l 12 There is the issue of whether we should depend on 13 two of these to clear the scram system without diversity in 14 the form of this simple -- to open exitation circuits and 15 provide the power to the breaker. :
16 DR. KERR: Is there any indication that this 17 machine, which was effective when it was installed, was when {
l 18 it was installed, or does it appear that whatever has {
19 occurred -- is there an indication that this breaker was 20 defective when installed? l i
21 And does it appear that breakage or whatever had i 22 occurred in installation --
23 MR. PEEBLES: That is an interesting question, i
G 24 Jim Thomas may add to this. It almost appears that, as we
, U 25 said, there are several potential contributors, one of l
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BLW/bc 1 which, as you will see in the pole shaft, is the welds are 3 2 of questionable quality.
1 3 You can't tell at this point. Like I said, we're )
1 4 going to take the extra -- and have it sent off to Franklin ;
1 5 Research in order to ask them to' document it and to see the i 1
6 condition of the welds and how long they think they would I
7 last, et cetera. ,
1 8 MR. THOMAS: I think the information we have 1
9 today based on the initial test and inspections is that we 10 definitely have poor quality welds from a visual inspection 11 standpoint.
rT 12 We are awaiting'the full weld analysis for the U
13 formal test program to give us more information there. It j 1
14 dcfinitely appears that the welds could have been of the !
l 15 higher quality for this particular breaker, and that the 16 cyclic wear appears to have contributed to the eventual 17 fatigue that caused that particular weld to break.
18 All of that is inconclusive at this time. It is 19 going to require substantial laboratory testing analysis to 20 really firmly come to that conclusion.
21 DR. KERR: Would a good QA program have spotted 22 the low quality welds?
23 MR. PEEBLES: It did spot the low quality welds l 24 on the pole shaft at Catawba. That is after -- it is also l /~N0 l 25 after you were put on notice that there was something you l
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BLW/bc 1. .ought to look for.
2 DR. KERR:. In selecting inspecting equipment 3 before it's put in operation at all.
4 MR. THOMAS: They were manufactured under the 5 quality control program as it was in the early. seventies.
6 The recent manufacturered breakers that Westinghouse has 7 supplied for replacement parts to Duke show a substantial 8 visual quality improvement. The welds are far better in the 9 overall visual appearance of the breaker.
10 It gives indication that the quality procedures !
{
11 and the quality control in the manufacturing process have- (
12 been substantially improved since the early seventies.-
I 13 MR..EBERSOLE: I'd like to ask the question while 14 you're up there. What you see happening here is 15 investigation of these breakers to find out what is 16 pointedly wrong with them. And presumably an action taken to l
17 improve that point of failure.
18 They still will be here as a residual and there i
19 is a dependency, one of which is much clearer to get the 20 reactor home. I have been a long time unhappy with this
'21 state of affairs which can be obtained by intercepting i
i 22 the raodus -- q l
I l 23 I guess I would like to get the reaction of how i l
24 comfortable you would be as an electrical engineer if we had l
25 this diversity in your plants. l l l l
l ACE-FEDERAL REPORTERS, INC.
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BLW/bc 1 MR. THOMAS: We.took a long look back in 1983 2 when the devices gave us some problem. And it's to what 3 type of diverse equipment is available for.this particular 4 application. By diverse, I mean not switch gear, not going 5 to a DB ficky (ph.) or a G.E. switch gear, which is still a 6 switch gear.
7 In 1984, we could not find an acceptable diverse 8 type of mechanism that we felt would significantly increase 9 the reliability or even meet the design requirements.
1 1 10 The closest we came to something diverse was 1
I 11 vacuum contactors, which'are simpler but don't have the l
fq 12 capability to interrupt felt currents that the switch gear l \~)
33 does.
14 MR. EBERSOLE: Not interrupting -- intercepting l
15 the exitation circuit, which is just a few amperes.
16 MR. THOMAS: Yes, but if you were to have a 17 fault, then you ahve got to have something that is not going 18 to malfuction itself and propigate the problem.
19 So we didn't find something that we could insert 20 into the system without still having to depend on the 21 capabilities of a breaker of this type.
- 22. We looked at various types of options, saturated 23 reactors. We looked at the possibility of developing a
(} 24 solid state type of breaker. And, to my knowledge as of 25 this date, we still don't have the technology available in a 1 1
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BLW/bc 1 proven form that provides significant improvement in this 2 area.
3 MR. EBERSOLE: What interrupts the exitation l I
4 curve now?
5 MR. THOMAS: The breakers.
1 1 6 MR. EBERSOLE: What interrupts the exitation 7 curve at this time?
8 MR. THOMAS: The same type breakers. I' 9 MR. EBERSOLE: Same as this?
10 MR. THOMAS: Yes. <
l 11 MR. EBERSOLE: It's just in another place?
12 MR. THOMAS: Yes.
13 MR. PEEBLES: It is not a class one E-breaker. ,
14 MR. THOMAS: It is a DS-416.
{
15 MR. EBERSOLE: The same thing, but not l
16 classified. Thank you. {
17 MR. PEEBLES: I just wanted to show you what this 18 was real quick so you'd know what it was.
19 (Slide.)
20 Here's the large movable blade contact going back 21 in this direction. Here's the insulated link. Pole shaft i 22 comes down through here. The weld that was cracked was l
l 23 here. This is the main drive link in the roller and the 1
24 closure cam.
l 25 If you could see the rest of the breaker, you l
l ACE FEDERAL REPORTERS, INC.
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BLW/bc 1 would see that there is almost nothing in the breaker that 2 is substantial enough in order to keep the.large contacts in 3 the two large opening rprings from opening the breaker other 4 than the main -- link. And that's why we concentrated in 5' that area.
6 MR. MICHELSON: Where is this gap that you put l I
7 the 70 mil thickness gauge into? It is not on that picture?
8 MR. THOMAS: It is a different part of the i 9 breaker. It is a UV device that does not relate to the 10 present problem, 11 MR. PEEBLES: It is a tolerance that you have to 12 worry about when you are adjusting it. That's why you have 13 to look at tolerances when you're taking the other one {
{
14 apart. i 15 MR. THOMAS: It had no relationship to the 16 jamming.
)
17 MR. PEEBLES: Not in this case.
) 18 MR. MICHELSON: It does prevent the breaker from 19 opening. I 20 MR. PEEBLES: It could. ,
i
.21 MR. MICHELSON: I thought the 70 mil thickness
) ;
22 gauge prevented the breaker from opening.
23 MR. PEEBLES: That is correct. ]
(
24 MR. MICHELSON: Is that correct? {
( j 25 MR. PEEBLES: Jim, right at this spot, right? j l
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BLW/bc 1 MR. THOMAS: Yes. Let me clarify. The test that k '
2 we were doing is to determine how much margin the UV device 3 has to trip the breaker. We want to maintain a certain i 4 margin in over-travel. That means the amount the UV device 5 lever pushes the trip shaft beyond the point that it trips.
6 We inserted a shim that estricted the over-travel 7 to a certain level and, at that point, the UV device did not 1
1 8 trip that breaker that one test.
L
! 9 We were never able to get it to repeat.
1 l
10 MR. PEEBLES: We tried it several times later.
l
! 11- MR. MICHELSON: It just ' ailed that one time?
I i
j r~s 12 MR. PEEBLES: One time.
I' j 13 MR. MICHELSON: Okay.
I j 14 MR. THOMAS: It did trip with a margin of i
l 15 19/1000ths versus 70/1000ths, still with a restriction of 16 19/000ths it tripped, so we still had margin. We didn't l
l 17 have it. We wanted to maintain --
l 18 MR. PEEBLES: Obviously, it changed somewhere 1
19 since 1983. So you do have to take a look at that.
20 MR. WYLIE: Why do you test these --
21 MR. THOMAS: Most of the tests are associated 22 with the tests of this UV device that were as a result of l 23 the UV problems in 1983. We have to periodically take the 24 breaker out of service and make many tests related to the UV
(~}
\J l 25 device. Each one of them causing the breaker to trip.
1 l
l A PE. FEDERAL REPORTERS, INC.
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BLW/bc 1 Plus, we have some periodic type on line
'2 requirements that are more frequent than they were prior to 3 '83.
4 MR. PEEBLES: It'is part of the license condition 5 for unit two. That's part of the answer.
6 MR. WYLIE: Is that peculiar to just McGuire 7 plants or the McGuire plant or the Catawba? They're all 8 tested 300 times?
9 MR. PEEBLES: It is part of the regulation in the 10 latest Vendor Manual.
11 DR. KERR: Is there any possibility of f
12 eliminating the unver-voltage trip?
13 MR. PEEBLES: Making it like a regular breaker. I 14 MR. THOMAS: The under-voltage trip has nothing 15 to do with the breaker failure that we had at McGuire. We 16 had a mechanical binding of a link in the pole shaft, which j i
17 has nothing to do with the under-voltage device or the {
)
18 tripping action of the under-voltage device. i 19 It is just that when the under-voltage device {
20 told it to trip, the breaker in another link bound up the !
I 21 under-voltage device has been supplemented with the shunt j l
22 trips so that, in actuality, in the actual operation day to ;
23 day, the shunt trip trips, but'it is outrun by the shunt.
24 So the shunt really is what is causing the 25 breaker to trip.
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BLW/bc 1 DR. KERR: The under-voltage trip causes you to 2 have to test more frequently than you otherwise would have 3 to test it and, therefore, adds to wear and tear on the 4 breaker.
5 MR. THOMAS: That is correct. The test 6 associated with the under-voltage causes us to test it more.
7 But eliminating the under-voltage device rather than the 8 test, I don't think would be what I would recommend.
9 The under-voltage does provide diversity in trip 10 within the mechanism; reducing the number of tests we are 11 required to perform on those under-voltage devices I think 12 is a good idea.
13 MR. BARANOWSKY: We need to move along. There is 14 going to be an AIT report that will be coming out. Two 15 reports. I think it might answer some of these questions on 16 technical details. /
1 17 MR. EBERSOLE: I'm going to claim the last 18 question for myself as subcommittee chairman. It seems 19 something is coming out of this discussion that might be i 20 worth considering. You mentioned the exciter circuit is 1
21 controlled by the same sort of breaker and would be, l
l 22 .therefore, subject to the same sort of problems.
23 On the other hand, this is evidence of wearing 24 out. And if you use -- if you have it triggered by wires 25 and a couple of circuits, it would act only when it fails to 1 (
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~BLW/bc 1 scram, then you would rarely then, if ever, have to do 2 anything. You would have a fresh, clean breaker back on the
- 3. exciter circuit where you would have these by test.
4 Do you follow me?
5 MR. THOMAS: You're talking about the actual 6 control circuits and the rod control system and if the --
7 set -- !
l l 8 MR. EBERSOLE: Would not be an instantaneous l
l 9 backup but based on signals that you had failed to scrum.
10 MR. THOMAS: We looked into that in '83 and l
l 11 actually tested a Catawba MG set by interrupting the l
l ('N 12 control, essentially telling the MG set to stop its output.
l ' %.)
l 13 And that is a feasible thing to do. It is not l-l 14 extremely complicated. We had our safety analysis people l
!- 15 review what benefit we would get from a PRA standpoint in 16 regard to increasing the probability of a trip with that 17 system added.
18 It came out a little surprising in that the 19 complexity that it added, even though it wasn't significant, 20 almost offset what gains we made.
l l 21 We made a decision not to install this diverse l
l 22 trip in that it did not appear to be cost benefit in terms l
l 23 of providing substantial increase in reliability.
24 MR. EBERSOLE: You would suggest that it has
)
1 25 negative effects. I can see a unilateral effect --
l l
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.BLW/bc 1 MR. THOMAS: It did not prevent the reactor from 2 tripping.- It complicated the mechanisms involved in trip, 3 increasing ~the probability of testing failures.
4 MR. EBERSOLE: What'were those mechanisms?
5 MR. THOMAS: The fact that we got more equipment 6 that we would have to periodically test and have to devise 7 schemes where you could test them independently of each 8 other, that would periodically be put in the system in a 9 mode that is less safe than normal.
10 MR. EBERSOLE: Do you have that study in print 11 that we can see?
p 12 MR. THOMAS: Yes.
O 13 MR. EBERSOLE: I would like to see that if l
14 possible. Thank you. Go ahead, Pat.
15 MR. BARANOWSKY: Our next item is by Mr. Richard 16 Lobel. It has to do with some procedural inadequacies 17 regarding the switchover circulation of the Westinghouse 18 plants.
19 MR. LOBEL: Before I start on this topic, I would 20 like to make a slight correction of something I talked about 21 last time; I talked about three events in terms of 22 management procedural inadequacies, and one of them was at
)
l 23 Oyster Creek and had to do with blocking open vacuum 24 breakers.
25 And I stated that this was discovered by the next l
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l 0 08.08 86 BLW/bc 1 shif t and the utility feels very strongly that we make the 2 point that it was discovered on the same shift. It was 3 discovered -- although this is under some dispute, still, by ,
4 somebody, in connection with somebody coming on duty early. l 5 But it was discovered on the same shift.
6 7
8
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0 09 09 87 BLWbur 1 The topic for this talk is a problem that has 2 arisen with procedures for switch-over during a large break 3 LOCA from taking suction during injection from the RWST to 4 operator actions to take suction from a sump when the RWST 5 runs low.
! 6 This issue arose as part of a Florida Power &
l 7 Light Turkey Point design basis reconstitution effort that l
8 they have underway, where they have gone back and looked at l
l 9 some of their important safety systems and done a thorough 10 review of the bases for the system and the procedures that I
l 11 accompany them.
l l 12 And as part of this, they looked at the time that l
13 was allowed their operators for switching suction from the 14 RWST to the sump. Their procedures allowed them to turn off 15 all ECCS pumps while they were making the switch-over, and 16 they asked Westinghouse for the basis for this 10-minute 17 number.
1 18 Westinghouse went back and looked and found that l
l 19 the basis was somewhat lost in antiquity, so they restudied l
i 20 l this 10-minute number on the basis of the information that i
21 ; is available toaay, and they found that they couldn't I 22 support the 10-minute number.
I l 23 Their analysis showed that two minutes was more 24 like the available time for -- this is during the switch-25 over for all ECCS pumps to be shut off. The reason for this ACE-FEDERAL REPORTERS, INC.
3 202-347 3700 Naticaide Cmerage Mn 336-6646 L_ _ ___
0 09 09 88 BLWbur 1 is when the calculations were first done, the basis for the 2 10 minutes was that-they assumed a simple pot boiling model 3 'where the water in the core boiled away at a certain rate, l 4 depersding on the decay heat, at about 15 or 20 minutes after 5 the larCe break LOCA.
l 6 When they went back and looked again, they took 7 into account new data from tests in Japan and calc"lations 8 with more accurate codes that showed that not only would 9 they bet the boil-of f but they would have water leaving the 10 break from entrainment and, from the difference in heads 11 between the colder water in the downcomer and the warmer 12 water in the core, would be driving water out the break.
13 Turkey Point, the utility for Turkey Point ther.,
14 rather than depend on operator action in two rainutes, 15 changed their procedures so that they left the high pressure i
16 pumps running. So they had a continuous source of water 17 turned off the low head pump, switched the suction to the
'10 sump to the low head pump, started that pump, and then 19 switched the high head pump, so there was a continuous 20 fsourceofwatertothevessel.
21 Let's see, we asked Westinghouse what other 22 plants might be affected. They gave us a preliminary list, 23 and we had preliminary conversations with the other
(
24 utilities, and it turns out that all the utilities involved l 25 have made some change to their procedures to accommodate L
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2 Some were done earlier. Some were done.as part 3 of the update of the emergency operating procedures, updates 4 af ter Three Mile Island or af ter an SEP review, and some
, [. 5 l51 were done as a result of the findings.that were subsequent 2
6, f tc the Turkey Point discovery.
7 Westinghouse sent a letter to all of their ,
- '; , t 8 customers, not just the ones that they' felt might have a b '
9 problem, and so at this point it appears that the utilities 10 t -- that all the utilities that seemed to be involved have
, 11 ]s made some change to their procedures to accommodate this,
("g 12 I some of them are still studying the problem in T.) r j i 13;.l(1.termsofpossiblesinglefailureshnddon'thavefinal
>I 14 s changes to the procedures, but all have accommodated it to e 15 some extent.
el
<' 16 ' ' Questions.
17 MR. WARD: hhat' sort;of cha6ge -- you were f s s
18 specific,about the change, the procedures made at Turkey 19 Point. '
s 20 flR. LOBEL: They arq similar type changes. They 21 arebasicalfychangeswhereonetrainiskept in operation 22 or one pump is'tept in operation so that you have continuous
., 23 flow.- '
t1 1 1 24 They s2ry in how they accommodate.. single
- %-)1 l 25 failures. One utility is still talking about having one i,
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BLWbur 1 train operable while another train is switched, and that 2 obviously does not meet the single failure criterion. So l 3 they are still looking at how they will accommodate the 4 single failure criterion, but they are all basically )
5 similar, trying to keep flow going at all times.
6 MR. WARD: Do the high pressure injection pumps l 7 draw directly off *.he sump in all Westinghouse --
i L 8 MR. LGBEL: They have to be piggybacked back to l
l 9 the low head pumps, but when you get the signal from the l
l 10 RWST, you still have enough inventory so that you have some l
! 11 time -- especially with a lower capacity pump, you still 12 have some time to make the switch.
)
l 13 MR. EBERSOLE: Can you say something about the 14 level of anticipated damage if the wrong procedure were 15 followed?
16 MR. LOBEL: If the existing procedure at Turkey 17 Point had not been changed, it appears the core could 18 uncover enough so that it is Westinghouse's estimate that 19 the 2200 degree Fahrenheit limit would be exceeded. l 20- DR. KERR: Is this best estimate code or by the 21 codes required for -- in Appendix K?
22 MR. LOBEL: It is really neither. What they did i 23 was they took -- they wrote a new type of program to account
/} 24 25 for the inventories and the flows of the system, NPSHs of the pumps, and did calculations. I suppose it is more of a ACE FEDERAL REPORTERS, INC.
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F90 0 9 0 9 91 lbBLWbur l 1 best estimate. It is not an Appendix K calculation.
I 2 But it is a rather simple calculation, and the i 3 estimate of exceeding 2200 is based on core inventory, not ;
1 k
4 on a calculated temperature.
- 5 MR. WARD: How important is the assumption about L 6 the decay heat level at the time the calculation is made?
l 7 If the decay heat power is low, you will of course have --
l, 8 you will have a lower froth level for cooling in addition to (
1 l 9 having lower power.
10 MR. LOBEL: The earliest you can get the switch-l 11 over is around 15 -- that you need the switch-over is 15 to 12 20 minutes. It is my understanding that is usually based on 13 assuming you have the pumps operating at runout conditions. -
14 So you really have more time than that. So the 15 or 20 15 minutes is probably conservative.
16 MR. WARD: I guess I am suggesting that low decay 17 heat power may be a worse condition than a higher decay 18 power.
k 19 MR. LOBEL: You have less froth -- l I
20 j MR. WARD: Less steam cooling of uncovered rods.
I 21 MR. LOBEL: It could be to that extent. I don't j l
l-22 know. That is probably true, but then you would have to j l
23 consider that you probably wouldn't lose as much inventory, j I
24 also, under those conditions. You probably have to do a 25 more detailed calculation than what has been done so far, ACE. FEDERAL REPORTERS, INC. j l
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(_) l BLWbu r 1 and the utilities have all taken the tack of trying to ,
l 2 maintain flow rather than doing detailed calculations of j 3 temperature.
j 4 MR. EBERSOLE: To get a handle on tie frequency l 5 of the notification, communication, and action systems, when ;
l 6 did the first notice come out from Westinghouse and when was i 7 the last fix made?
8 MR. LOBEL: I am not sure when Turkey Point first 9 discovered the problem. They had discussions with 1
l 10 Westinghouse, with their engineering people. They came to l
11 the conclusion that they needed to change their procedures 1
12 in March of this year.
1
(~
Q l 13 Turkey Point notified the Staff. Florida Power &
1 l 14 Light notified the Staff via 50-72, a call to our Operations l
l 15 Center. Another plant also did that, and we started I
16 discussions with Westinghouse and the utilities subsequent l
j 17 to that.
l l
18 MR. EBERSOLE: Does Westinghouse when it sends j.
l 19 out the notices to do these things, doesn't the Staff get a l
l 20 copy? -
! 21 MR. LOBEL: No. We asked Westinghouse for a copy l
22 when we became aware of it through Florida Power & Light.
I 23 MR. EBERSOLE: Westinghouse will find a problem l
l
{} 24 and send it out to all of their owner-operators and not tell 25 you that it exists, and then there will be a rabble of ACE. FEDERAL REPORTERS, INC, l
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BLWbur 1 response to it. a n'd you won't hear about it until somebody {
2 recognizes it and sends it back to you.
3 Is that a logical communication system?
4 MR. LOBEL: Is it logical?
5 It is not preferable from our point of view.
I 6 MR. EBERSOLE: Is there indicated here a change 7 in the communication process tnitt says that the designers,
{
l 8 owners, vendors, whomever, when they find something they l
l 9 should fire off a copy to you people if they tell owner-10 operators?
l 11 MR. LOBEL: That is something that should be l
12 discussed.
13 MR. EBERSOLE: And then it can be given i 14 consideration.
(
15 MR. LANNING: It should come under the auspices 16 of the Part 21 reporting.
17 MR. MICHELSON: Why didn't it get reported?
18 MR. LOBEL: This is speaking for Westinghouse a 19 little bit. I think their reasoning would be that they --
20 and the reason this list of plants was tentative is that I
l 21 i they knew what set of plants should have the problem based {
I I
22 on their original design, but they didn't know what changes {
l 23 had been made to the plants, and so they were not in a j 24 position to speak for the utilities.
25 MR. MICHELSON: They either have to do it l
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l O.09 09 94 BLWbur 1 themselves, Westinghouse, or they have to assure themselves ;
i l
2 that the utilities potentially involved do it. The law 3 requires that one or the other report.
4 MR. LANNING: It is my understanding that I 5 Westinghouse did not provide generic emergency operating 6 procedure guidance'in the phase of the switch-over, and that 7 part of the procedures was developed by each utility.
8 So it really fell -- they had to report it, 9 right.
1 10 MR. MICHELSON: Report it under 50.55(e)? i 11 MR. LANNING: Yes. j 12 MR. EBERSOLE: On a real generic basis, does the 13 Staff contemplates any action to be taken here in view of
. 14 the apparent commonality of this communication, what I would i
15 regard as a communications deficiency? Is there anything
[
16 contemplated at this time to do something about this?
17 MR. LOBEL: It has not been discussed in a lot of 18 detail yet. You have to underst.and that this is the group 19 who first looks into the ="ents, and we have discussed this 20 with Reactor Systems Branch and NRR, and they are going to 21 pursue this further, both looking at the technical aspects 22 and I think looking at these aspects, also.
23 MR. EBERSOLE: How is our schedule?
24 MR. BARANOWSKY: We are doing okay.
) 25 MR. EBERSOLE: I think then that -- I think the ACE-FEDERAL REPORTERS, INC.
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BLWbur 1 subcommittee would be interested in hearing about the 2 follow-up on a generic basis. It sounds to me like it 3 offers an opportunity to have funny things happen between 4 the vendor and the owner-operator that the Staff never hears I
5 of.
6 I was disturbed to hear you say that Westinghouse .,
f 7 didn't know what the owner-operators were doing to the 8 Westinghouse plants without closure back to Westinghouse 9 that told them what they were doing.
10 MR. LOBEL: I am telling you what Westinghouse J 11 says. I am not really in a position to know the details of c 12 the interactions between the utility and the vendor.
13 MR. EBERSOLE: Any questions or comments on this?
14 (No response.) j 15 MR. BARANOWSKY: Our next presentation is by Gary 16 Carter, Events Assessment Branch. He will be described an j l
17 event that occurred in a power plant in which there was a l
}
18 degradation. 1 I
19 ,
MR. CARTER: I am Jerry Carter. i l
l 20 l The Pilgrim utility in June of this year was 21 making a modification pursuant to the 50.59 review of the l
22 containment spray nozzles and header system. Because of !
t
) 23 some recommendations by GE and from the MARK I initiatives,
/~~} 24 they had decided to remove the existing nozzles from the V
25 upper and lower spray within the drywell and replace it with
)
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BLWbur~ l a plugged nozzle; i.e., six out of the seven little holes in 2 the nozzle would be capped, and this would reduce.the flow
'3 through the nozzle in order to have a better flow 4 distribution within the drywell and also to provide adequate 5 cooling for their drywell wall.
)~
6 'When they removed the nozzles, they found that k
} 7 rust had accumulated within the header and in the nozzles
)
8 themselves to a depth of perhaps approximately a half an 9 inch.
10 When we first heard of this, we were very 11 concerned about the generic implications. Were there other 12 utility plants with plugged spray nozzles in the containment 13 area, and was the testing that was done to verify the 14 operability of the spray system adequate?
15 The piping within the drywell is carbon steel.
4 16 Where the environment is logically going to be somewhat 17 humid, there is an occasional instance that has occurred 18 where water has been injected through the spray system into 19 the drywell.
I 20 This happened for a short duration at Pilgrim, 21 and upon investigation we found that this has also occurred 22 at several other plants.
23 DR. SHEWMON: Is this tap water feedwater --
24 MR. CARTER: Water from the LPCI.
25 DR. SHEWMON: It doesn't sit there normally?
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l BLWbu r 1 MR. CARTER: Normally, the pipe is. dry, but l
2 because of the test requirements there is water downstream l
l' 3 of the isolation valve just sitting in the water in the
! 4 piping. It is not a filled system, but it is not dry t
l 5 either.
L 6 DR. SHEWMON: And it is exposed to air?
l 7 MR. CARTER: It is exposed to air.
l 8 One of the concerns was that the air in the 1
9 system with the wet carbon steel piping might be the source
! 10 of the rust. We asked several utilities if they would make 11 a quick check of their nozzles and their piping.
r^\ ' 12 A check was made at. Peach Bottom of approximately U
13 10 nozzle locations, both in the drywell and the wetwell.
14 They found a slightly rusted surface, but not: anything that 15 would indicate the large amount of flakes that had occurred 16 at Pilgrim.
17 We.also had a similar piece of information from 18 Cooper. Cooper had had an inadvertent spray actuation in 19 the past. They did not also have any accumulation of rust 20 at that point in time.
21 The utility, to determine the operability of 22 these nozzles, injects air into the piping system 23 periodically and makes a qualitative test of nozzle air
/% 24 flow. There is no quantitative test of flow distribution
(~) )
25 either with air or water with these nozzles.
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' \ ~)S0 10 10 BLWbur 1 Pilgrim had successfully tested by air the 2 operability of these nozzles back in, I think it was, 1982.
3 1984 is when the actuation occurred. There was an {
4 actuation, as.I mentioned, at Cooper in 1980, and they did 5 not have any rust.
6 Pilgrim is currently still studying the problem, 7 trying to determine the source of the rust.
8 There is one interesting thing that has come up j
9 since the original event with the inerting of the 10 containment area. You should have a minimal amount of 11 oxygen, but they are saying in addition that you get minimum 12 migration of the. nitrogen, displacing the oxygen in the 13 actual header itself. Because the nitrogen has not swept 14 into the piping system, they do not expect that there is a 15 large amount of rust f rom this source, but they do not L 16 currently know the source.
l 17 The Staff is currently intending to look at the {
l 18 specific material of the spray systems, as opposed to my l
l 19 statement that it is carbon steel, to see if they can
{ '
20 determine some subtlety that might have led to the problem.
l 21 Currently, we do not know precisely where the
\
22 rust came from other than it may have been there and caused {
l 23 by the actuation of the spray system at Pilgrim and moved it 24 to the piping system, where it settled out. I
(~}
\_) k l 25 If that is the cause, it appears to have been a l
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'k) l BLWbu r 1 plant unique problem rather than a generic concern that we j 2 originally had.
3 MR. REED: The low pressure safety injection i 4 water is unborated?
i 5 MR. CARTER: Correct.
6 MR. REED: Do you have any idea what the pH is?
7 MR. CARTER: I do not know, but it is not 8 borated.
9 MR. REED: It seems to me the pH would be l
l- 10 important to evaluate.
l 11 MR. MICHELSON: It would be neutral. There are
, 12 no additives to it.
13 MR. CARTER: It is coolant water.
14 MR. EBERSOLE: You suggest this has perhaps no 15 generic implications. Now, let me ask you this. This was 16 'inerted containment. All of the light bulbs an; inerted.
17 So you would expect in the beginning that carbon steel would 18 not give much trouble, hopefully rarely.
19 Let's extrapolate that to spray rings forever --
20 I like the new big MARK III containments. I think they have l
21 spray rings in them -- and the fact that they are not 22 inerted and examine whether or not the containment sprays, 23 if they are critical or more critical in the safety context 24 as they are here, whether we should expect or not expect to
(~}
As
- l. 25 get sprays when we need them.
1
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l r ** 100 i t}01010 BLWbur 1 Have you looked at carbon steel spray rings 2 wherever they may be, especially where they are in ordinary 3 damp atmospheres, or do you intend to now inquire?
4 MR. CARTER: I cannot speak for the Reactor 5 Systems Branch that is doing the event follow-up in detail 6 beyond the point where we had it. I would suspect they
(
7 would, but I don't know for a fact that they are or not. I 8 DR. SHEWMON: I had thought I asked you earlier 9 if this water on steel was exposed to air, and you said 10 "yes."
11 MR. CARTER: Yes. Pilgrim is an old plant.
g 12 DR. SHEWMON: He is talking about inerted plants.
13 Pilgrim is not inerted.
14 MR. CARTER: Pilgrim originally operated in a 15 normal atmosphere.
16 DR. SHEWMON: Normal atmosphere, meaning air?
17 MR. CARTER: Yes. It then became operational 18 with an inert atmosphere, nitrogen. That would have been
(
19 injected into the drywell area. I would suspect that the 1 20 header piping itself would have remained primarily air if q 21 there was oxygen in there until it was consumed.
22 Speculation on my part.
I 23 I think the question that is being raised here is !
l 24 in the future under MARK III --
25 DR. SHEWMON: I don't care what his question was. I i
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BLWbur 1 I want an answer to my question.
2 MR. MICHELSON: Did you look at the configuration l 3 to see what happens when you do the periodic testing of j
{
4 those valves?
5 MR. CARTER: The piping -- j 6 MR. MICHELSON: To the ring header outside of 7 containment. There are two isolation valves just outside l 8 the drywell. ;
i, 9 MR. CARTER: That is correct.
10 MR. MICHELSON: Aren't they horizontal? Do they 11 drop down into the header? !
12 MR. CARTER: There is a horizontal drain line
<O 13 that comes off the side, but there is water that would be l 14 trapped, as I understand it, between the valve and in the l 15 drywell.
16' MR. MICHELSON: What about the trapped water when 17 you are doing the checking of the valves to see if it is 18 operable? Does it drain down into the header and lay there?
19 It depends on the configuration.
20 MR. CARTER: In the particular --
l l 21 MR. MICHELSON: There are two problems with these 22 in the past. One has been the drainage of water during a 23 valve test. The other has been the valve leakage itself, 1
{} 24 leaking by slowly, just laying in there, adding to the 25 corrosive environment.
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BLWbur 1 MR. CARTER: The utility is intending as part of
.1 2 their fix to this to install additional drain lines to try 3 to remove the water that does'get past the valve, either by ,
l 4 leakage or during the testing. ;
5 MR. REED: I don't know how well this reactor is I
6 shielded and the neutron leakages on the reactor, but you i 7 are telling me this is an inerted containment. Then therein 8 lies the potential in some nook or cranny to produce nitric l
.9 acid, and these things have happened in the early days, in-1 10 pile testing.
11 Is anyone looking at whether or not -- what the 12 atmospheric conditions are inside the wall while operating 13 with the inert containment? Is there any development of 14 nitric acid?
15 MR. CARTER: Aga i n., I am not aware of that, but I 16 cannot answer that question.
17 MR. REED: It might be significant, and it might 18 be there is more neutron leakage on Pilgrim than the others.
19 MR. MICHELSON: Also, the piping's condensers are 20 hanging on the containment. They have a tendency to 21 condense out.
22 DR. SHEWMON: If you are going to form nitric 23 acid...
24 MR. BARANOWSKY: To give an evasive answer to 25 your question, Mr. Reed, the licensee has not completed ACE FEDERAL REPORTERS, INC.
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01010 103 BLWbur 1 evaluating why they think the rust occurred, nor is our 2 e"aluation complete. Mr. Carter is just trying to tell you l
3 what we know today. It is still ongoing. We are not going i
4 to drop it. l 5 MR. EBERSOLE: That means we will have a follow-6 up later on.
7 Are there other questions?
8 I guess this particular spray system, if I recall 9 correctly, is not of very profound importance. It is to 10 cope with containment, effecting bypass, and of course it is 11 extremely important if we have suppression bypass. It is 12 the only thing left other than containment venting.
13 Is that correct? What about pipes not in the 14 drywell but over in the donut?
15 MR. CARTER: In the case of Peach Bottom, where 16 they looked at that piping, they did not fina the rust. I 17 don't know what they may or may not have found at Pilgrim.
18 MR. EBERSOLE: They haven't looked at it yet.
19 MR. CARTER: I can't answer that. I am not aware 20 of it.
21 MR. EBERSOLE: You would intuitively expect you 22 might find the same thing?
23 MR. CARTER: Intuitively, you would think so. I 24 would think it might even be worse.
25 MR. EBERSOLE: It might be worse?
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BLWbur 1 MR. CARTER: It is above the water, but the 2 temperature should be lower.
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3 MR. WYLIE:- The spray sparger.
4 MR. CARTER: On this particular system it j 5 supplies either the drywell or it can be used to spray over 6 the wetwell. t 7 DR. MOELLER: The top spray didn't appear to be I
8 going into the drywell.
9 MR. EBERSOLE: It is.
10 MR. CARTER: There are.two spray headers. The 11 figure you have is a schematic that does not fully represent 12 Pilgrim. The upper ring is just at the shoulder of the 13 bulb. The other one is down about 30 feet. They noticed j 14 that at Pilgrim both of them rusted, but at a lesser amount 15 in the lower header.
16 MR. EBERSOLE: All right, we will go on. I 17 MR. BARANOWSKY: The next item on our list ;
18 involves a loss of reactor coolant inventory event during 19 maintenance activities at the North Anna Power Plant. That 20 will be discussed by Mr. Leon Engle, Project Manager for i
}
21 that plant.
)
J l
22 MR. EBERSOLE: And he would mention the Millstone i 23 3 aspects of the same problem?
f 24 MR. BARANOWSKY: Let him talk.
25 MR. EBERSOLE: We will see. l
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1 BLWbur 1 MR. ENGLE: I am the Project Manager for the 2 North Anna Power Plant. The event I want to speak about has 3 to do with a loss of reactor coolant system inventory.
l 4 (Slide.)
i
- 5 This event, as opposed to some of the others you 1
6 -hear on operating reactor events, didn't happen in a hurry. I 1
l 7 It took three days to occur. !
I L
l 8 In addition to that, it is a rather unique l
9 combination of differences in system component levels, low l
10 cystem pressures, and residual heat and pressurizer vacuum l
l 11 that hydrostatically decoupled the pressurizer from the rest 1
i e 12 of the system.
(])
l 13 And a third item that should be of some interest 14 to you, that is that RVLIS, which you people kept pushing 15 for long and hard during the post-TMI years, could have been 16 shown to be a very valuable instrument indication in Modes 4 17 and 5. This whole event did happen in Mode 5.
18 MR. EBERSOLE: It is somewhat spectacular to find 19 that RVLIS, when you get into Modes 5 and 6, you can 20 j disconnect it, ignore it, throw it away, whatever, but it is 21 probably as valuable in those modes as any.
22 MR. REED: It won't work because of the pressure 23 differential -- the pumps are on. It throws it off.
24 MR. EBERSOLE: We are shut down, Glenn. We have 25 stopped everything.
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BLWbur 1 MR.-REED: I. thought you said in the higher --
- 2. MR. EBERSOLE: This is 4, 5, and 6. It is down 3 even'when.you are refueling.
4 So the arbitrary -- apparently -- denial of use 5 of RVLIS in Modes 4, 5, and 6 is something I just cannot 6 understand and I think needs to be investigated.
l 7
1 8
9 10 11 12 13 14 15 16 17 18 l 19
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21 1
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25 L I l 1 l ACE FEDERAL REPORTERS, INC. l 202-347-3700 Nationwide Coverage 800-336-6646 l 1
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BLWbw 1 MR. ENGLE: On June 17, 1987, after a 59-day i
2 refueling outage -- so this unit was fairly cold, had a 3 little heat in it, commenced operations to leave Mode 5 and 4 Intermode 4. They initiated two reactor coolant pumps'and 5 after several hours, the A pump tripped, and they determined 6 that the motor had grounded and would need to be replaced.
7 They then secured the other pump, the sea pump, and to 8 facilitate motor removal, they began reduction in system 9 temperature and pressure and cooled down by venting of the 10 pressurizer.
11 Eventually, the pressurizer level was reduced to
(
12 80 percent level and the PORV is closed. A decision was j 13 then made by the management that rather than going the 14 normal course all the say down to mode 5, they would reduce 15 the primary level to about 20 percent, as opposed to the
\
16 normal procedure which would be to reduce it to the mid loop 17 plus 40 inches. I 18 on June 18th, the pump was uncoupled and, of 19 course, a small leak of approximately, now determined to be 20 about 1-1/2 gpm developed up the pump shaft from the s9al f
21 injection.
22 The operators continued to monitor at the 20 23 percent pressurizer level from RCS inventory and went to a ,
I 24 procedure called 10P34, which allows so-called " volume 25 control tank float." This has a pressure of about 30 psig, ACE-FEDERAL REPORTERS, INC.
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.V 4 BLWbw I a nitrogen blanket and provides flow to the reactor coolant 1
2 pump seals. I 3 Now this procedure is very terse. It doesn't say j 4 much. It says, fellows, you can go into this mode. It 5 doesn't say, hey, look out for mass inventory balance. The I I
6 system is rather cold. It just doesn't say you can't stay 7 there forever, because you are not going to be able to. The 8 system is cold, and it leaves them there. I mentioned a 9 little later that this procedure is being modified, and this 10 is, incidentally, a Westinghouse procedure.
11 Okay. The operators assume that they could 12 maintain makeup in this system for balance in inventory due 13 to the seal leakage by watching the pressurizer level at 20 14 percent and watching the makeup control tank. So finally, 15 at 113 hours0.00131 days <br />0.0314 hours <br />1.868386e-4 weeks <br />4.29965e-5 months <br /> on June 22, Westinghouse working on the pump 16 requested that if the pump seal leakage could be reduced, it 17 would help repairs. And the operators decided to raise the 18 pressurizer level to cycle the PORV to vent and then reduce 19 pressurizer level to generate a slight vacuum.
20 i Well, once they cycled this, they rapidly 21 determined that something had been going on that they hadn't (
I 22 thought about, because the pressurizer relief tank pegged 23 I all the way from positive to negative, and they suddenly 24 realized that they had had a vacuum for some period of time.
25 (Slide.)
(' (
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f~')0-11 11 109 V-BLWbw 1 And they immediately began a quick inventory to 2 see what they had. They immediately restored makeup and had 3 ' determined that there were probably gases in the reactor 4 head from the nitrogen that had been used in this volume 5 control tank, and by the time they restored inventory, they 6 had.to add about 17,000 gallons of borated water.
7 Now during the event, there was adequate 8 inventory at all times for our RHR, but a perusal of that l
9 RVLIS data at a later date determined that the RCS level was 10 7.9 feet below the top of the reactor head, about 5.3 feet 11 above the center line of the loops and 8.8 feet above the 12 top of the core.
u 13 Now several things, I think, can be said about
'14 this event, and it goes back to TMI, in part, and that is 15 that using a pressurizer level for monitoring reactor j l
16 coolant system inventory can, at the very least, be 17 misle'ading.
l l 18 It also tells us something about Modes 4 and 5, j L j l 19 where, at the very best, you don't have to do much j l
l l 20 instrumentation to go by, when you are coming down to cold (
l (
l 21 shutdown. l I
l 22 MR. EBERSOLE: What about 6, Mode 67 l
23 MR. ENGLE: To some degree, but I think lesser I l
24 than 4 to 5.
25 Another thing it clearly points out, is an l
l 1
ACE FEDERAL REPORTERS, INC. I 202-347-3700 Nationwide Cmerage WF336-6646 9
i
cE0 11 11 110 BLWbw I accounting problem. You had better damn well take care to 2 see that you are measuring reactor coolant system inventory, 3 keeping check of mass flow balances, and it requires 4 constant surveillance of all available level indication, and 5 finally, if there had been routine surveillance of RVLIS 6 during this period, it could have alerted the operators to 7 this problem.
l l 8 Now I want to point out that RVIIS is not tech i
l j 9 spec. It is tech spec only through Modes 1, 2 and 3. It is l
l-10 not required for 4, 5 and 6.
l-l 11 MR. EBERSOLE: I would like to stop at that point 12 and have you discuss, if you can, how that came to be, when, i 13 in f act, the uppermost value of RVLIS, as evidenced by the 14 TMI 2 case, was, you need it when you are fiddling around at 15 some condition other than power.
16 MR. ENGLE: I cannot answer your question as to 17 how that was determined.
18 MR. EBERSOLE: The other is this not generic 19 across-the-board to all PWRs.
20 MR. ENGLE: I think it is not going to be 21 addressed by the Staff in terms of maybe, should we require 22 RVLIS in 4, 5 and 6.
23 MR. EBERSOLE: For instance, Diablo Canyon.
24 MR. REED: Question. What is the elevation, 25 relative elevations of the volume control tank that is ACE. FEDERAL REPORTERS, INC.
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N""101111 l BLWbw 1 physical elevations and the pressurizer. Are they separated l
l 2 by 100 feet in elevation? You have your fill energy coming l
l 3 from a pressurized volume control tank down in the auxiliary l
4 building, and quite frequently, you will find pressurizer 5 many feet above volume control tanks.
L.
[ 6 MR. ENGLE: Unfortunately, I put these elevations 7 on here, but below this elevation, which includes the VCT, I l 8- don' t have .the elevation.
9 MR. REED: I have never heard a procedure like 10 this in running a PWR. It is really a wierdo, the so-called 11 float thing, but I assume these people did look at the 12 elevations and what 30-pound pressure in the tank would push 13 the water into the pressurizer or into the system.
14 MR. ENGLE: The procedure was intended to do just 15 that.
16 MR. REED: And the elevations verify it can be 17 done.
18 MR. ENGLE: And they can do it, and they did it.
19 MR. REED: It is not greater than, let's say, 50 20 feet of elevation difference.
21 MR. ENGLE: I cannot explicitly tell you the 22 difference, but it is less than 50 feet.
23 MR. REED: With flappers and resistances, and 24 maybe check valves and vertical lines?
}
25 MR. ENGLE: Like I indicated, this pressure puts ACE-FEDERAL REPORTERS, INC.
202-347 3700 Nationwide Coverage W)-336-6646
0 11 11 112 f' BLWbw 1 these guys right out in the wild blue yonder, and it doesn't l
j 2 say, fellows, check anything. Watch it. It doesn't even f-l 3 tell you the time. By the procedures, they could be in this
)
4 condition for days and days and days.
5 MR. EBERSOLE: It could be much earlier, after
[
lT 6 shutdown when there is more heat.
7 MR. ENGLE: That's generally true, and that is a 8 very salient point that you have to realize, is that this 9 unit was a very cold unit, having been shut down for 59 10 days, but by the same token, the operator should have 11 realized it was cold and that it did not have residual heat 12 in there to maintain what t. hey thought was a slight bubble 13 in that pressurizer over this period of time. They didn't 14 have the heat there to maintain it such that the pressurizer 15 would not go into this vacuum.
16 Now the licensee, after this happened, he 17 immediately issued an event description to INPO and began 18 immediately looking at their procedures, and I was down 19 there a couple of days after it happened, and they are 20 looking into these procedures very, very carefully. They 21 are getting together with Westinghouse on it. The North 22 Anna procedures will require, for RCS makeup in this regime 23 that operators are reevaluated and made to realize that l
24 every piece of available instrumentation must be watched 25 under these conditions.
I ACE FEDERAL REPORTERS, INC.
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BLWbw 1 In addition, the procedures will spell out that 2 they maintain an inventory of makeup and let down. And 3 finally, RVLIS is going to be included in that procedure 4 regardless of the fact that their only tech spec through
! 5 Mode 3 and the simulator trading at North Anna for operators 6 is being reevaluated for these conditions, and the simulator 7 will actually include, including RVLIS in the procedure.
l l 8 MR. EBERSOLE: Your statement that it would still l
l 9 be inspected only at Modes 1, 2 and 3, sounds a little l
l~ 10 funny. Why shouldn't it be tech spec clear through?
l 11 MR. ENGLE: Well, unless it becomes a generic --
L l 12 MR. EBERSOLE: I am assuming -- it is on that l f)
\_
13 basis I say that. '
14 MR. ENGLE: If the licensee asks for an 15 amendment, we possibly -- we would probably give it to them 16 gladly, but unless it is a generic issue, and it goes 17 through backfit and for the whole industry, I can't require 18 North Anna to do that at this time.
19 MR. EBERSOLE: How at this time do you figure 20 that we will figure out it is generic? Is the information 21 notice going out all over?
22 MR. ENGLE: Yes. Information notices are going 23 out. The region is evaluating. They are sending a notice 24 out. " Events" has sent -- having Reactor Systems looking
[}
25 into the implications of tech specing 4, 5 and 6, and the l
/(CE. FEDERAL REPORTERS, INC.
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- l. _. .__ _ _ _ - - _ _ _ _ _ . _ _ _ _ _ _ _ _ _ - - _ _ _ - _ - _ -
l ["10 11 11 114 l U BLWbw 1 vendors have all been notified of this event.
2 MR. EBERSOLE: Thank you.
1 3 Any questions?
l 4 (No response.)
l l 5 MR. EBERSOLE: Thank you.
l
( 6 And how are we doing?
l l
7 MR. BARANOWSKY: We are way ahead of schedule.
[
8 You asked if we could mention a few other events 9 that occurred during shutdown that indicate either l 10 questionable procedures or operations. One event that you
- 11 might be familiar with, having read " Nucleonics Weeks,"
('T 12 involved a situation that occurred at the Fermi Power Plant
\_/
13 not too long ago, in which an operator trainee was tasked to 14 record reactor coolant system temperatures as a function of 15 time, and the plant was going through some corrective 16 maintenance on the RHR systems, and that is why they were 17 interested in what might be happening with primary coolant 18 temperature and wanted to keep an eye on it.
19 And this trainee did record those temperatures 20 hour after hour after hour right through mode changes, and 21 no one had ever told him what to do when it got to above 200 22 degrees, so they went to around 210, before someone finally 23 realized that they ought to ask what is happening with the 24 temperature. They were able to make some corrections, but 25 it did show some flaws in the way the operations were ACE FEDERAL REPORTERS, INC.
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2 Another event that occurred at Millstone -- I I
3 can't remember whether it was 2 or 3 -- involved also !
j 4 looking at reactor -- at temperatures. In this case, they !
l 5 were looking at -- if I recall correctly, temperatures in i 1
6 the RHR system, and I think they were looking at a loop that 7- was isolated. I don't have all of the details that clear in I
8 my mind, but the temperature reactor coolant system was l l
9- going on while all the RHR system in that particular loop 10 was indicating a stable temperature. ;
11 So I guess one needs to really think out the j 12 activities that they are doing during shutdown. We see
)
13 quite a few events that involve loss of inventory or 14 increases in temperature and in some cases, boiling in the 15 reactor, when the plant is shut down. I think there is 16 going to be a lot more attention paid to that area of 17 reactor operations in the future. ;
18 MR. EBERSOLE: Okay. i 19 MR. BARANOWSKY: The last event that we will be
} i 20 ! talking about today happened at Brunswick, a BWR. It was a 21 scram, where complications, and one of those complications 22 involved common mode failure of some safety relief valves 1 23 that there used also for ADS functions. In this instance, 24 what happened is, there was a voltage regulator problem at 1
[}
25 the plant, which results in a turbine generator trip, and is ACE FEDERAL REPORTERS, INC.
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.BLWbw- 1 fairly typical for this type of plant, when the reactor i
{
2 scrammed, the reactor water level got down to the level to 3 set point, at least approximately, and initiated some 4 isolations, and at level 2, one expects things like the MSIV I
5 to isolate, which will take away the main feedwater system.
6 The reactor coolant recirculation pump should trip and 7 things like high pressure coolant injection and reactor core 8 isolation cooling systems should actuate automatically.
t 9 (Slide.)
i 10 In this case the MSIVs did isolate, but the 11 reactor core isolation cooling system and the high pressure !
12 coolant injection system did not actuate, and one of the
)
l 13 reactor coolants recirculation pumps did not trip. The {
l 14 operators manually started thoO4 systems when they first !
! 15 started HPCI. The high pressure coolant injection system.
16 That system tripped because of valve alignment problems, l 17 which'were then corrected by the operators and they went on 18 to start that system successfully.
19 The reason why these systems didn't start, as it i
20 turned out, from our post-incident investigation, was the 21 normal variation that one might expect in the 22 instrumentation where the the Level 2 set point actuation is l
l 23 at 118. inches in the vessel, but the instrumentation is set l l- )
24 up so that actuations will occur plus or minus three inches, 25 and the levels that were recorded on the four channels i
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5 i ' og y r'70 11 11 117 ybf L 'BLWbw 1 indicated some variation within that plus or minus 3 inches, L
rh;t 2 so the level had come down just about to Level 2 and hung 3 around there, and the indications were such that two of the l ,
l' 4 < four channels tripped off, actuating some equipneat '
and two 1: l 5 of the four channels did not trip ar.d some of the equipment * ]
l' 6 did not get actuated or isolations did not occur.
7 Now as part of the recovery f rom this event, the 8 operators were using RCIC to control level. They were using 9 safety relief valves in manual to control pressure. It 10 N turrs out that one of the three safety. relief valves that 11 were being used.to control. pressure would not open on demand 12 when they tried to use it. So they were successful in 13 getting to shutdown with the other two that they were using.
14 Initially we were interested in this event, 15 because of the lack of initiation of HPSI.and RCIC, and IC' after we investigated the variations in the instrumentation 17 set points and so forth, we were satisfied that that problem 18 has -- is one of less significance than possibly the fact 19 that the safety relief valve also did not actuate on demand 20 when desired, because the safety relief valve is part of the 1
- j. 21 automatic depressurization system, which is sometimes also l
l
.22 necessary for operation at BWRs and for certain loss of l 23 coolant accidents that are surely necessary.
l 24 We asked them when they restarted the plant to l 25 test the valves, the safety relief valves that are used to I !
I L ACE-FEDERAL REPORTERS, INC.
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0 11 11 118 BLWbw 1 perform the ADS function, and when they did that another j l
2 valve failed, a different one, that they hadn't tried to l
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i 3 use, and at that point, of course, our concern started to j 4 grow.
l 5 The licensee has just had maintenance down on the j l
6 solinoids for the safety relief valves that allowed them to l l
7 operate in manual by Target Rock people, and so what they 8 did was they replaced the newly maintained solenoids with !
l l 9 some older ones tnat they knew to be operating, and they l l 10 took the newly maintained ones off, and they sent them off' ;
o i l
?
! 11 to Wylie Labs, where Target Rock does the testing and the )
l ..
12 maintenance on these valves.
(-
13 Now as it turns out, when they tore down the 14 valves -- first they tested them and actually all of the 15 valves work, even the ones that were stuck during the 1
16 i nc i'de n t. But when they took the valves apart, the solenoid L
y 17 valves, in particular, they found a substance on them that l
18 was identified as locktight to be stuck around the plungers l
l 19 on the solenoids, and apparently, it had worked its self 1
20 free on the two valves that were seized up when the plant l
L 21 was operating. But clearly, there was a substance there J l 22 that could have caused the plungers to get stuck, and they 23 on two other valves found a much lesser degree and 24 apparently older deposit of block -- perhaps from a prior L 25 maintenance which, of course, piqued our interest, because l
Ace FEDERAL REPORTERS, INC.
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_-_ -- __ _ . _ . _ - - - _ - _ _ _ - - - - _ _ , _ _ - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ = _ _ _ _ _ _ - _ _ _ _ _ - - _ - _ .
l l l I 1 \
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l BLWbw here we had two valves f ail that would have been needed in I
l 2 the ADS function and two more that showed signs of simiar !
l \
3 failure mode. And normally, one needs five out of seven of l 4 these valves at that plant for ADS function, for design l l
l 5 basis function, but you need three or four of them even on ;
4 1
6 the best estimate basis anyhow. '
l.
7 MR. EBERSOLE: I would like to comment at this l
8 point. I would call your attention to the fact that the ADS
! 9 valves are the lifeblood, the real root devices of the BWRs.
l l 10 You have the two HPCIs, the turbine-driven pumps. Both of 4 l
1 l 11 them are unreliable. If they can't get water into them, I
l 12 they are up the creek for sure. The inventory will come
)
l 13 down. You go below level, and the ADSes have to work, or 14 you are going to melt the core.
15 And I have long complained about these -- these 16 are weak points in the general rationale of depressurizing-17 this plant. I think this is an example of what could result 18 in serious trouble, if you couldn't depressurize the plant 19 and just get water in it, after we have lost the two 20 unreliable high pressure pumps. They have been improved by 21 diesel-driven high pressure core sprays. I would be less 22 concerned about those, but with the ones with the HPCIs and 23 RCICs, I think they are a standing risk when coupled to this 24 design of depressurization systems.
25 MR. REED: I, of course, challenge the locktignt ACE-FEDERAL REPORTERS, INC, 202 347 3700 Nationwide Coverage MG 33MM6
r"9 120 k )0 11 11 BLWbw I argument. I have made many -- I have been in the record 2 many times as being against pilot, internal pilot-operated 3 relief valves on PWRs and BWRs, because of the fluid 4 environment, namely, oxygen, hydrogen, boric acid 5 potentials.
6 I am very much against and am asking more 7 research on internal pilot-operated relief valves. I do not 8 think they are reliable, unless you have pure water systems.
9 MR. BARANOWSKY: I just happen to have a simple 10 type schematic drawing here that I can show you a little bit 11 better what we are talking about. The solenoid valves let 12 air into the -- nitrogen in BWRs, because we are talking 13 about inerted containment. The problem is with this valve 14 over here. There have also been problems with the Target 15 Rock valves, with the -- with the corrosion, and so forth, 16 along the seeds on the main valve, which have caused them, 17 in numerous instances, to fail to lift at certain pressure 18 set points, and in some cases, quite a bit of pressure is 19 required in order to lift them.
20 So I don't want to suggest that problem is not a 21 real one, because it is being looked at by Target Rock, in 22 terms of material compatibilities and the type of 23 environment that these valves are being operated in.
24 MR. REED: It is being looked at by Target Rock?
)
l 25 MR. BARANOWSKY: Yes.
l'
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.l BLWbw 1 MR. REED: Why haven't we heard about it?
~2 MR. BARANOWSKY: We get so many deficiency I l
3 reports on the safety relief valves that don't meet the {
{
4 specs for testing, that they are giving a serious look into 5 the material situation on the valves.
6 MR. REED: I think I have been toading incident 7 reports and LERs for 20 years on these pilot. operated valves i
8 failing to function. There is always a new cicuse, new 9 chains of material or something.
10~ MR. BARANOWSKY: That is the same observation I 11 have. I can't disagree with you.
12 MR. EBERSOLE: What bothers me is a lack of 13 potential for action that will do something better than have 14 this thing depressurize the boiler. If this were on a PWR, l
15 I would be less concerned, because I would still have the f i'
l 16 secondary system, but I look at this against the background l
l 17 of RCIC and HPCI reliability, and it makes me shudder a L
I 18 little bit. I recall that picture to -- our attention is 19 being the lifeboat or the parachute of the boiler.
20 MR. BARANOWSKY: It is an important system, in 21 terms of the reliability of HPCI and RCIC. In the past, the 22 ADS system has been fairly reliable, and in the risk studies 23 that I have looked at, the big concern has been whether or 24 not operators will use it in time, and this is one of the 25 first times I have seen a hardware-oriented problem of ACE FEDERAL REPORTERS, INC.
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()10 11 11 BLWbw 1 maintenance. It is a hardware-oriented problem, common mode 2 in nature that could take that kind of function away.
3 Actually, to continue on, we were fairly L 4 concerned about this, and we immediately contacted the 1
l 5 regions and Target Rock. The regions contacted licensees.
1 l 6 We wanted to find out what was happening. What happened is, l
[ 7 several valves were tested at operating plants, when they l
8 were in operating conditions that permitted that. We didn't 1
f 9 want people at full power to go popping off valves and start I
j 10 a situation with a stuck-open valve and you need the ADS 11 system, t
s 12 Also Target Rock had several other valves in
[ k'-)
I 13 their shop from different plants that they were.able to-14 inspect and test to see whether or not this potential common i 15 mode problem with a locktight at least on the ADS was out l
16 there all over the place, and we would have to take some 17 immediate action. I 18 Fortunately, what we think we found out is that 19 in the situation of the Brunswick valves, a new technician 20 for Target Rock had worked on the maintenance of that valve 21 and supervision of his work and procedural and 22 administrative controls over what he was doing during some 23 : of these critical portions of the maintenance wasn't as good ,
i 24 as it should have been. He has clearly cautionca not to let
}
25 the excessive locktight drip, but unfortunately, that is ACE FEDERAL REPORTERS, INC.
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L()01111 l BLWbw l what happened.
! l 2 We think that based on examinations that have- i l
l 3 been made of valves, teardowns and tests that have been I 4 down, that it is isolated because of this one technician.
5 At least I can say that there will be continued 6 testing of these valves to assure that, under conditions in 7 which one could expect this type of lock up to occur, it 8 won't have occurred, plus, Target Rock is looking into how 9 they can best modify their procedures, improve their 10 supervisory oversight of the maintenance activities and, in 11 fact, they are also looking at what kinds of postmaintenance i
12 testing that they can do to detect this condition better 13 than has occurred in the past.
14 MR. EBERSOLE: Are they going to keep on using 15 locktight?
16 MR. BARAN0WSKY: Yes. The reason they use 17 locktight is that it holds the nut in place within the 18 solenoid and the valves have been environmentally qualified 19 with the locktight in there. And if you take the locktight 20 , out or put a different substance in there, one then l
21 questions the environmental qualification. I can't say !
22 whether they would make a change in the long term and do a 23 requalification or not.
{} 24 MR. EBERSOLE: There are plastics that hold 25 better than locktight.
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BLWbw 1 MR. BARANOWSKY: They are going to a nylock type 2 nut, or they have one in there, but they feel that they have 3 to use the locktight also. They can't take the chance on 4 that nut coming free.
5 MR. REED: Can you show me on the preceding slide 6 where the locktight was?
7 8
9 10 {
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P^10 13 13 125 tg BLWbur l' MR. BARANOWSKY: In the solenoid there is a 2 plunger, and the plunger opens and closes the valve. It is 3 part of the solenoid plunger. It blocks the valve stem to 4 the --
5 MR. REED: This assembly was sent to Wylie Labs, 6 not just the pilot operated valve.
7 MR. BARANOWSKY: This was sent. They will do a 8 whole maintenance on the whole valve, but in this case --
9 MR. REED: Are you sure which failed? The i'
10 solenoid or the pilot operated relief valve?
11 MR. BARANOWSKY: Yes --
l 12 MR. REED: What evidence do you have to be sure?
! 13 MR. BARANOWSKY: The valves worked all right when 14 they put the other solenoids on. They worked before.
15 MR. REED: After you exercised the valve?
16 MR. BARANOWSKY: Let me'tell you a little piece 17 of information about lock tight. The lock tight sets up in 18 a situation where there is a lack of air. It is an l
19 anaerobic substance that they are using.
20 They had tested these valves when they were going 21 '
up in power, and they worked, but the plant was not inerted 22 at that point. So when they inerted the plant, the lock l
l 23 tight set up. So the valve was functioning.
l 24 .And then when they had the thing not work, they l 25 took the valve off, they put a new one, and it worked with l
t L ACE-FEDERAL REPORTERS, INC.
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0 13 13 126 l BLWbur 1 the new solenoid on there.
l' 2 So that gave us confidence that it wasn't the j 3 valve that had a problem.
4 What we know about lock tight, I guess you would [
1 5 have to say it is somewhat of a deduction because these l
6 valves, when it breaks free it is like a crusty material and
!- 7 it won't hang up again, and that is really what happened.
8 There was enough crusty material to allow us to draw that l
l 9 conclusion.
I. (
10 MR. EBERSOLE: At Browns Ferry there was a great {
11 deal of difficulty getting it qualified. It must remain
\
12 energized at 125 or 250 DC to keep the blowdown function 13 going. That is in the face of the ambient that you would 14 face when you are blowing down. {
15 There was a great deal of difficulty getting it 16 qualified, and finally they did get a recipe, a cookbook 17 thing, some sort of a combination of plastics and whatever, l
18 and got it going with much difficulty in getting quality l
19 control on that particular solenoid. q 20 You must hold voltage on this plant to keep it 21 going down. If you lose that voltage, you will have circuit i
22 failures, solenoid failures, or whatever, and you will lock l 23 up at high pressure, and you are in trouble.
l l
24 I want you to get a nice good view of the l
25 lifeboat importance. ]
l )
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,BLWbur 1- MR. REED: It seems like it would be nice and 2 simple to have air operated valves.
3 MR. EBERSOLE: Yes.
4 MR. BARANOWSKY: We are working on an information 5 notice on this, and we are still in touch with target routes 6 on' additional follow-up. For the most part, we think we 7 have it understood if not finally taken care of.
8 MR. REED: This is the wrap-up of this session.
9 I would like to just make a comment, observation, 10 .that of these five events that were deemed important enough 11 to bring in, three very definitely are design deficiencies, rT 12 desi.gn problems. One is operations, and this one here hangs ;
i L/ :
13 in the balance between maintenance and design.
l 14 MR. EBERSOLE: One may lead to questions. What l
l 15 about the degreed operators versus the nondegreed operators? I l l 16 (Laughter.) I l I l
17 MR. REED: I have been worried that we have l l 18 tended to think that we have gotten into the flogging of l l
l l 19 operations and maintenance all the time, and I keep finding i
20 scads of design deficiencies, and it worries me that there
! l l 21 is no one really focusing hard and flogging the design
{
22 aspects.
l
( 23 DR. SHEWMON: Are design deficiencies as one guy j
! I 24 perceives as one and another one may not, like tax breaks?
f 25 MR. REED: I thought this was quite clear that {
l {
l l
! ACE-FEDERAL REPORTERS, INC.
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t 0 13 13 128 BLWbu r 1 the break was a design deficiency. I 2 Do you agree with that?
3 DR. SHEWMON: I don't agree to any design 4 deficiency. j f
5 MR. MICHELSON: It is not a design deficiency.
6 MR. REED: It could be manufactured not according 7 to design, but they must have had spot welding, right? A 8 tricky point.
9 Switch-over? What are you going to call that?
I 10 That is a design deficiency. Didn't calculate it.
11 Pilgrim's spray header, I say that is a design
{
12 deficiency.
13 MR. MICHELSON: We don't know that. We don't 14 know what the problem is.
15 Are we all done with the --
16 MR. EBERSOLE: I was just going to say we have 17 done better than we expected with the schedule, and, Wayne 18 or Pat, do you have anything else to add to your 19 presentation?
! 20 We figured we didn't have time to cover it. If 21 you do, you can take the prerogative of bringing it up, 22 since we are well ahead of schedule.
23 MR. MICHELSON: Let me ask, for one that wasn't 24 brought up that occurred recently. I don't remember the 25 details. I thought I would hear them.
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i r* 129 k,)90 13 13 BLWbur 1 A condenser, main condenser, the butterfly valve 2 lost its keewai pin or flipped close. Could you send me the i
3 details on that, since I wasn't able to find any? Just send 4 the details?
5 MR. EBERSOLE: That was the case of the bolts 6 backing out, inadvertent valve closing, a water hammer
(
7 occurred with splits and apertures and led to a flooding of l l
8 the turbine. They had designed the turbine hull -- (
l 9 MR. MICHELSON: I thought it was important enough 10 maybe to hear. (
11 MR. EBERSOLE: It is a new water hammer.
-{
12 MR. MICHELSON: Water hammers don't cause things I 13 like this to happen. .
14 MR. BARANOWSKY: It was a water hammer caused by 15 the fact that the large butterfly valve -- I forget how 16 large -- .
17 MR. MICHELSON: It has got to be 30 or 40 inches. .
18 MR. BARANOWSKY: 10 feet. It closed in one or [
19 l two seconds when it normally closes in about a minute.
I 20 MR. MICHELSON: The butterfly aspect is also f
i li l
21 extremely interesting because that isn't the only place.
j 22 So is that available?
l 23 MR. BARANOWSKY: We have a report on that.
1.
24 MR. EBERSOLE: One refinement of that is there l 25 was a certain degree of failure of the circ water system.
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1 0 13 13 130 BLWbu r 1 You had about two to six inches of water on the floor, and l
2 it was designed for an elevation of higher than that. I l
3 think that report ought to be embellished by saying we were 4 perhaps subject to a condenser box failure of X square feet, i
5 where you would have a veritable Niagara and it exceeded the j i
6 design basis of flood level event on the turbine hull floor i
7 and you would have flooding of other spaces where critical l
8 equipment was located, i 9 Are you following me?
l l 10 MR. BARANOWSKY: I am not sure exactly what it 11 takes to get that kind of flooding situation. As of several 12 years ago, most plants -- all plants, I guess -- really went l 13 back and looked at the plant design for very large break --
i 14 I think the severance of the large pipe.
l 15 MR. EBERSOLE: If they did that --
l 16 MR. MICHELSON: Surprises have occurred since I
17 then on smaller breaks where they thought they had all the l
l l
18 pathways figured out and they found some new ones.
i l 19 MR. BARANOWSKY: That is why we are in the l
l 20 business of looking at operating experience events. We see 1
l 21 something that doesn't look right based on what we think we l 22 should see.
l l 23 But in this case it really worked out to be the 24 design was adequate for the event and somewhat expected.
25 MR. EBERSOLE: If you have any other, as I said, h ACE FEDERAL REPORTERS, INC.
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0 13 13 131 3LWbur 1 events that you could just mention here to further color the ;
2 active situation out on the field, mention them. I have a 3 few here if you don't.
4 MR. BARANOWSKY: We see so many, my mind is -- I 5 don't think I have any that I could point out. I think 6 these are typical. I will be trying to cover the spectrum 7 of plant shutdown equipment and operations.
l 8 MR. EBERSOLE: 1 will mention a few, and you can l 9 embellish them in detail better than I can.
l j 10 Maine Yankee, 6/13, we had this unhappy state of 11 affairs where you had loss of all service water, loss of f
12 ECCS containment isolation valves, and quite a bit of 13 isolation equipment for ten minutes.
1 14 That is an event that can the plant -- can it l
l 15 tolerate the total loss of water flow into the plant? You 16 don't get bearing seizures, diesel failures or anything 17 else. We continue to get this total loss of power to 18 critical systems on both trains.
1 19 MR. MICHELSON: What was the cause?
20 MR. EBERSOLE: They were doing overcurrent device l
21 calibration on the 40-watt breakers with the plant in hot 22 They had a disabled breaker.
l shutdown.
23 On 7/20/87, breaker trip, breaker at McGuire, I
24 which we had covered.
l 25 Coming up to a very interesting one indeed here -
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BLWbur 1 - not this one -- Brunswick 2, we lost both pressure cooling 2 systems. Reactor operators were the cause. I believe it is 3 valves.
4 One that struck me as being interesting is we 5 have the case that I was really unfamiliar with the details, 6 where the component cooling water system has an event which 7 is protected so that it closes in the event radioactive 8 discharge occurred from that event.
9 So there was an event that had radioactive 10 discharge from the tank, and the leak, I believe, that was 11 causing the radioactive discharge began to pressurize the 12 tank. There was a potential of blowing the tank-up, and j 13 then you could lose component cooling, and that is a sad 14 state of affairs for certain designs, not necessarily all of 15 them, because that is the coupling recirculation treated 16 water system that gets the heat out of the critical 17 components in the plant.
18 I don't know whether it was that critical design 19 or not, but this is -- this occurred -- this is a pickun 20 from a Part 21 report issued in 1984. It just came to light 21 here by a process I didn't understand. I picked it up in 22 the daily reports here recently along with other reports.
23 Kiwani. Now I ask if we are so -- I take this up 24 in connection with the BWR ATWS potential due to I 25 radioactivity from the dump volumes.
! ACE. FEDERAL REPORTERS, INC.
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BLWbur 1 Another case where the
- ntastically unnecessary 2 logic of preventing a little leak of radioactivity induces f 3 plant challenges, severely challenging safety. I would not 4 want to have a little radioactive leak from the component 5 cooling water tank lead to an event that would cause you to 6 blow the tank.
7 MR. REED: I think that thing ought to be against 8 the code. If they don't have a safety valve on that tank 9 and they have a trip valve on the discharge --
10 MR. EBERSOLE: Westinghouse advised a number of I
11 plants to remove the internals from the CCW surge tank f
12 relief valve and ensure that the vent remains open by -- if 13 the CCW system was designed as a closed system without 14 automatic containment isolation --
I 15 DR. SHEWMON: Why don't you give it to her for j l
16 the record? {
{
17 MR. EBERSOLE: There may be CCW systems that are j l
18 closed if you close that vent. I don't know that. But if I l
19 there are, I think it deserves a very careful look. l l
20 This is points that may be affected, designs one l l
21 and two, a whole string of them.
22 An incident here of some significance. l l
23 Deficiency report. 6/26, Philadelphia Electric Company 24 filed a significant deficiency report on the potential 25 hydraulic valve operators. It sounds as though this is a l 1
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, . O.13 13 134 BLWbur 1 generic problem.
2 MR. BARANOWSKY: The Limitorque operator 3 hydraulic lockup problem is being addressed as a generic l 4 problem. l l
l 5 MR. MICHELSON: What do you mean by " hydraulic"?
l l 6 MR. BARANOWSKY: Hydraulic lockup of the limit L
L 7 switch. As I recall correctly, it will over-torque the l
l 8 valve and possibly shear the stem. . That is what they call 9 it.
10 There is a grease in the limit switch mechanism
)
l 11 which hydraulically locks up so that the switch doesn't work 12 properly. It cannot relieve when the switch is supposed to l 13 change positions adequately because of viscosity and things 14 that happen when the valves are placed in operation,in terms 15 of changes to the viscosity associated with the grease.
16 I cannot claim to be an expert on it. AEOD, Eric 17 Brown, some people in our Generic Communications Branch, are 18 following it. There is a testing program and design 19 modification program to fix that up.
I 20 l MR. EBERSOLE: The solution was here to cut two 21 slots to let the grease out. What bothered me was that 22 there are thousands of these valves.
23 Is it implicit in this report that many of them 24 may be affected?
25 MR. BARANOWSKY: I think it is true that they ACE FEDERAL REPORTERS, INC.
l 202-347-3700 Nationwide Coverage fd10-336-6M6 3
"'10 13 13 135
-b BLWbur 1 might be affected, but what we have found is that it is not i
2 the kind of thing that happens every time you exercise the 1
3 valve. It is not always understood as a random type 4 failure, and that is why it is being pursued.
5 MR. EBERSOLE: I have no further material. So as 6 far-as I am concerned, the subcommittee report is firiishede 7 DR. REMICK: I look down to the Staff and I see, 8 Ray, the next item is the future ACRS activities.
9 (Whereupon, at 4:45 p.m., the ;ommittee was 10 adjourned.)
11 j
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CERTIFICATE ~OF OFFICIAL REPORTER
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This- is to certify that the attached proceedings before the UNITED STATES NUCLEAR REGULATORY COMMISSION in the matter of:
NAME OF PROCEEDING: ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 328TH GENERAL MEETING DOCKET NO.:
PLACE: WASHINGTON, D. C.
DATE: FRIDAY, AUGUST 7, 1987 were held as herein appears, and that this is the original transcript thereof for the file of the United States Nuclear Regulatory Commission.
(sigt L (TYPED)
BARBARA L. WHITLOCK Official Reporter ACE-FEDERAL REPORTERS, INC.
Reporter's Affiliation O
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, r 33SA3 S:?/90 RKE'Y STA"US R3:? ORT I
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A :?resen a ion "o '::1e ACRS :Tu:.:. Committee "1omas Kenyon, Projec; Yanager !
Augus; 7,1987 O
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- R3 VIEW EIST03Y i
- CUR 3ENT RsVIEW SCH37Ls O
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REVI]W HISTORY
- Initial Submi;ta:. -
Oc:ober 1983 o
- Final Submi;tal -
March 1987 Review Approach
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Original -
- V:odu:.e-by-modu:.e Curren; -
- n;egra;ed O
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. CURRENT REVIEW SCHEDULE
. Draft SER -
April 1988
- ACRS Meeting -
June 1988 O
- Final SER -
October 1988
- December 1988 PDA Decision Date -
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Insert to SRP 3.6.3 Plant design and arrangement considerations to assure protection against
() fluid system piping failures are discussed in SRP 3.6.1. SRP 3.6.2 covers the size and location of postulated breaks, postulated leakage cracks and how postulated breaks and leakage cracks affect structural design and equipment qualification.
When leak-before-break has bcs., demonstrated to be applicable as provided for in this section, designers are cautioned that safety related equipment lo-cations should continue to be chosen with care as discussed in B.1. of ASB 3-1 of SRP 3.6.1. Safety related equipment should not be concentrated in the vicinity of high energy lines irrespective of whether leak-before-break is applicable. Routing of high energy lines in the vicinity of control rooms is not acceptable.
When leak-before-break has been demonstrated to be applicable as provided for in this section, the following provisions of SRP 3.6.2 remain applicable:
- 1. Postulation of leakage cracks in piping qualifying for leak-before-O breek is stil, required (see e.3.c. of xEe 3-1 ef SRe 3.6.2). <
Leakage cracks in high and moderate energy piping not qualifying for leak-before-break are also required.
- 2. Evaluation of structures protecting essential equipment is still j required (see D. s,e.(4) of MEB 3-1). Design of these structures I includes all dyr.amic effects of high energy piping not qualifying for leak-before-break; however, for piping which does qualify for leak-before-break, only the pressurization associated with an area equivalent to the cross section of the controlling pipe and a linear release time of three second should be assumed, unless justification is provided for alternative leakage conditions frrm an evaluation of
- other potential sources, such as flanges, bolted covers and valves. j l
- 3. Environmental qualification of safety related equipment is specified
~
in B.1.c.(5) of MEB 3-1. Relaxations in these requirements wTien l n leak-before-break is demonstrated can be authorized as discussed in U -
the Supplementary Information of the final broad scope amendment to l GDC-4. I
t..
L' MCGUIRE 2 - FAILURE OF REACTOR TRIP BREAKER TO OPEN ON DEMAND L.((( JULY 2,-1987 PROBLEM JAMMING OF BREAKER MECHANISM CAUSED FAILURE TO OPEN.
STATUS INDICATOR LIGHT ON CONTROL PANEL LIT, ERR 0NEOUSLY TELLING OPERATORS THAT BREAKER OPENED.
CAUSE A COMBINATION OF HIGH CYCLE WEAR, AND A BROKEN WELD APPEAR TO HAVE LED TO JAMMING 0F MECHANISM. CAUSE OF ERR 0NE0US STATUS LIGHT INDICATION IS NOT YET KNOWN.
SIGNIFICANCE
([)' FAILURE'0F BREAKERS TO OPEN COULD LEAD TO ANERR ATWS 0NEOUSEVENT STATUS LIGHT FUNCTION COULD PRECLUDE DETECTION OR DE CORRECTIVE ACTION BY THE REACTOR OPERATOR.
DISCUSSION ,
JULY.2, 1987, PLANT IN HOT STANDBY, CONTROL R0D DROP TESTING IN PROGRESS TECHNICIAN IN BREAKER ROOM OBSERVED SM0KE IN.VICINTY TRIP BREAKERS
- - OPERATOR TURNS MANUAL TRIP SWITCH IN CONTROL ROOM:
TRIP LEVER ON MECHANISM ALSO PUSHED REACTOR TRIP BREAKER "B" WOULD NOT OPEN OPEN LIGHT ON CONTROL PANEL FALSELY INDICATED "B" BREA OPENED "B" TRIP BREAKER WAS THEN OPENED MECHANICALLY BY J0GGIN CHARGING HANDLE O
FOLLOWUP:
L .O-
LICENSEE REPLACED BREAKER "B" WITH BYPASS BREAKER FROM UNIT 1 AIT AT SITE-7/7-10/87 e BENCH TESTS OF BREAKER "B" CONDUCTED:
l DUPLICATION OF BINDIflG WAS MINIMALLY SUCCESSFUL L ELECTRICAL CONTROLS AND C0tlTACTS PROPERLY WIRED AND OPERATING FINDINGS JAMMING IS PROBABLY DUE TO EXCESSIVE LATERAL PLAY'Irl THE MAIN DRIVE MECHANISM PERMITTING CONTACT WITH THE MECHANISM
, FRAME AT THE CLOSING CAM 1
CAUSE OF FALSE INDICATION BY STATUS. LIGHT HAS NOT BEEN POSITIVELY DETERMINED BECAUSE CIRCUITS WERE ENERGIZED LIMITING TROUBLESHOOTING PRELIt11 NARY CONCLUSION IS THAT FALSE INDICATION WAS NOT CAUSED BY THE BREAKER FAILURE, BUT MAY BE IN THE EXTERNAL
. WIRING
[])
SM0KE WAS CAUSED BY BURNUP OF SHUNT TRIP C0ll WHEN BREAKER FAILED TO OPEN LICENSEE SHORT TERM ACTIONS LICENSEE AND WESTINGHOUSE (BREAKER MANUFACTURER) WILL CONTINUE TO INVESTIGATE THESE PROBLEMS (DRAWING AND FIELD WIRE CHECES, BREAKER LAB TESTS)
NRC VENDOR INSPECTION PROGPAM WILL FOLLOWUP DN BREAKER FAILURE INFORMATION NOTICE HAS BEEN ISSUED q
)
a
- O-1
._- _ __a
CATAvTA 1 Q
ON 7/24/87 VISUAL-INSPECTION OF THE 6 WELDS ATTACHING THE 3 POLE LEVERS T0.THE CENTER SHAFT INDICATED A SMALL " CRATER" CRACK, A ONE INCH LONG LACK OF FUSION IN THE WELD THAT ATTACHES THE CENTER POLE LEVER TO THE POLE SHAFT, AND RUBBING 0F ONE SIDE OF THE CLOSING CAM AGAINST ITS ADJACENT SUPPORTING FRAME MEMBER.
7/30/87 AIT VISITED CATAWBA, INSPECTED WELDS, CONCURRED WITH LICENSEE OBSERVATIONS.
LICENSEE WILL DISASSEMBLE BREAKER AND PERMIT NRC STAFF TO USE POLE SHAFT FOR ACRS BRIEFING SUBSEQUENTLY, NRC IS SENDING POLE SHAFT TO FRANKLIN RESEARCH FOR EVALUATION OF THE WELDS O
- FRC WILL BE REQUESTED TO DEVELOP CRITERIA T0 INSPECT WELDS ON RTB BREAKERS AT OTHER PLANTS e
I i
l x:)
1
MCGUIRE 1 O
ON 7/27/87 WHILE PERFORMING THE TRIP MARGIN TEST, THE CLEARANCE BETWEEN UVTA AND THE TRIP BAR WAS INADEQUATE. WITH It!SERTION OF A THICKNESS GAUGE (0.070"), THE BPEAKER DID NOT TRIP AIT ARRIVED AT SITE AND OBSERVED REPETITION OF THE TESTS Ot!
7/28/87 AND CONCURRED WITH THE L!CENSEE OBSERVATIONS UVTA WAS REMOVED, A SPARE UVTA WAS INSTALLED, RETESTED, AND DETERMINED ACCEPTABLE AIT REVIEWED MAINTENANCE HISTORY AND DETERMINED THAT THE WERE NO PREVIOUS PROBLEMS. THE FAILED UVTA WAS ONE OF THE FIPST BATCH OF UVTAs WHICH WERE MANUFACTURED SUBSEQUENT TO THE SALEM ATWS TOLERANCES IN ALL DIMENSIONS WILL BE CHECKED PRIOR TO DISASSEMBLY OF ORIGINAL BREAKER l
\
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2 33
AGENDA FOR ACRS MEETING ON AUGUST 7, 1987 l .
PLANT EVENT PRESENTER McGUIRE REACTOR TRIP BREAKER FAILURE T. PEEBLES 1
WESTINGil0VSE ECCS RECIRC SWITCHOVER R. LOBEL PROCEDURAL INADEQUACY PILGRIM POTENTIAL DEGRADATION OF CONTAINMENT SPRAY SYSTEM J. CARTER N. ANNA LOSS OF'RCS INVENTORY L. ENGLE BRUNSWICK SCRAM W/ COMPLICATIONS AND SRV P. BARAN0WSKY COMMON MODE FAILURE l)
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- j. 7- MCGUIRE 2 - FAILURE 0F REACTOR TRIP p*-- BREAKER TO OPEN ON DEMAND L JULY 9, 1987 Prov.a sg +1ly
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WESTINGHOUSE - POST LOCA ECCS SWITCH 0VER PROCEDURES b,
PROBLEM PROCEDURES PERMIT ALL ECCS PUMPS TO BE STOPPED FOR UP TO 10 MINUTES FOR MANUAL SWITCH 0VER FROM RWST TO THE CONTAINMENT SUMP, ONLY ABOUT 2 MINUTES MAY BE AVAILABLE, CAUSE NONCONSERVATIVE CORE BOILOFF CALCULATION DID NOT CONSIDER CONTINUED FLOW FROM BREAK.
SIGNIFICANCE ECCS INTERRUPTION COULD LEAD TO CORE UNC0VERY DURING SWITCH 0VER AND POTENTIAL FOR EXCEEDING 10 CFR 50.46 LOCA CRITERIA.
DISCUSSION TURKEY POINT EMERGENCY LOCA PROCEDURES REVIEWED AS PART OF DESIGN
($) BASIS RECONSTITUTION EFFORT.
PROCEDURES PERMITTED ECCS PUMPS TO BE SWITCHED OFF FOR UP TO 10 MINUTES FOR REALIGNMENT TO COLD LEG RECIRCULATION (SUCTION FROM CONTAINMENT SUMP).
ADEQUATE TIME WAS THOUGHT TO BE AVAILABLE BEFORE CORE UNC0VERY BASED ON POT BOILING MODEL.
FLOW OUT OF BREAK DUE TO DENSITY DIFFERENCES BETWEEN CORE AND DOWNCOMER WAS FOUND TO OCCUR DURING POST REFLOOD AT LARGE SCALE TESTS IN JAPAN, (CCTF)
LICENSEE IS REVISING PROCEDURES S0 THAT HPI REMAINS RUNNING WHILE LPI PUMPS ARE REALIGNED S0 THAT ALL ECCS FLOW IS NOT STOPPED I FOLLOWING A LARGE BREAK LOCA, WESTINGHOUSE SENT A NOTIFICATION LETTER TO ALL CUSTOMERS, A LIMITED NUMBER OF PLANTS AFFECTED.
FOLLOWUP
- NRR IS REVIEWING THIS ISSUE WITH RESPECT TO CHANCES MADE TO ls) EMERGENCY PROCEDURES AND SUPPORTING ANALYSES, s.
(]~ )' UPDATE PILGRIM - POTENTIAL DEGRADATION OF CONTAINMENT SPRAY SYSTEM PROBLEM RUST FOUND IN THE PRIMARY CONTAINMENT SPRAY HEADER AND N0ZZLES.
SIGNIFICANCE SPRAY PATTERN AND FLOW RATE MAY NOT BE THAT ASSUMED.
INTEGRITY OF HEADER MAY BE COMPROMISED.
BACKGROUND PILGRIM SHUTDOWN AND DEFUELED.
SPRAY N0ZZLES BEING PLUGGED (6 0F 7 HOLES).
CONTAINMENT SPRAY IS A REQUIRED SYSTEM AT PILGRIM.
($)
- QUALITATIVE AIR TEST OF CONTAINMENT SPRAY SYSTEM AT PILGRIM IN 1982.
SPRAY PIPING IN BWRs IS CARBON STEEL.
SPRAY PIPING IN PWRs IS STAINLESS STEEL.
INSPECTIONS MADE AT OTHER FACILITIES:
PEACH BOTTOM - NO SIGNIFICANT RUST FOUND CCOPER - NO SIGNIFICANT RUST FOUND BROWNS FERRY - NO SIGNIFICANT RUST FOUND CONCLUSIONS OPERABILITY TEST FOR CONTAINMENT SPRAY SYSTEMS MAY NOT BE ADEQUATE.
FOLLOWUP STAFF TO ASSESS ADEQUACY OF OPERABILITY TEST.
LICENSEE TO COMPLETE EVALUATION OF CAUSE OF RUST. ]
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_N09TH ANNA, UNIT 1 LOSS OF'RCS INVENTORY (MODES 485)
P9OBLEM: LOSS OF RCS INVENTORY (MODES 485) i CAUSE: RCP SEAL LEAKAGE 8 INADEQUATE PROCEDURE FOR DRAINING RCS SIGNIFICANCE: OPERATOR RELIANCE OY PZR LEVEL FOR RCS INVENTORY CAN BE MISLEADING. ROUTINE SURVEILLANCE OF RVLIS DATA COULD HAVE HELPED MITIGATE LOSS OF RCS INVENTORY DISCUSSION:
t*
JUNE 37, 1987 (59 DAY REFUELING OUTAGE) NA-I COMY.ENCED OPERATIONS TO LEAVE MODE 5 AND ENTER MODE 4
- AFTER SEVERAL HOURS RCP-A TRIPS DETERMINED RCP-A MOTOR GROUNDED 8 WOULD NEED TO BE REPLACED -
SECURED RCP-C (ALL RCPs SECURED)
TO FACILITATE MOTOR REMOVAL, SYSTEM TEMPERATURE AND PRESSURE REDUCED 8 C00LDOWN AND VENTING OF PRZ CONDUCTED PZR LEVEL REDUCED TO 80% LEVEL 8 PORVS CLOSED PZR LEVEL REDUCED TO 20%-
JUNE 18., 1987 RCP-A MOTOR UNC00 PLED SMALL LEAK (APPROXIMATELY 2 GPM) DEVELOPED UP PUMP SHAFT FROM SEAL INJECTION OPERATORS MONITOR 20% PZR LEVEL FOR RCS INVENTORY MAKEUP TO RCS WAS VIA VOLUME CONTROL TANK (VCT) " FLOAT" OPERATORS ASSUME MAKEUP FOR BALANCE INVENTORY DUE RCP-A SEAL LEAKAGE CAN BE COMPENSATED BY MONITORING PZR LEVELS 8 MAKEUP BY VCT l
AT 0133 HOURS ON JUNE 21 l SEAL LEAKAGE BE REDUCED , 1987, WESTINGHOUSE REQUESTED PUMP l OPERATORS DECIDED TO (1) RAISE PZR LEVEL, (2) CYCLE RORV T0 l
VENT AND (3) THEN REDUCE P2R LEVEL TO GENERATE P2R VACUUM ON PORY CYCLING, IMMEDIATE INDICATION OF ALREADY EXISTING PZR i
(y' VACUUM 9
OPERATORS REALIZED P2R LEVEL OF 20% NOT ACCURATE INDICATION OF RCS INVENTORY IMMEDIATE CHECK MADE OF AVAILABLE INFO MAKE UP RESTORED TO RCS AND RX HEAD VENTED FOR GASES :
APPROXIMATELY 37,000 GALLONS OF BORATED WATER NEEDED TO REESTABLISH RCS INVENTORY DURING EVENT, ADEQUATE INVENTORY TO MAINTAIN RHR RVLIS DATA INDICATED RCS LEVEL 7.9 FT. BELOW TOP OF RX HEAD, 5.3 FT ABOVE CENTERLINE OF LOOPS, AND 8,8 FT ABOVE TOP OF CORE USE OF PZR LEVEL FOR.f,0NITORING RCS INVENTORY CAN BE MISLEADING MONITORING FOR RCS INVENTORY (MODES 485) REQUIRES CONTINUOUS REVIEW OF RCS INVENTORIES REQUIRES CONSTANT SURVEILLANCE OF ALL AVAILABLE LEVEL INDICATION -
ROUTINE SURVEILLANCE OF RVLIS DATA WOULD HAVE ALERTED OPERATORS TO RCS INVENTORY LOSS '
f-q FOLLOW UP LICENSEE ISSUED EVENT DESCRIPTION ON INPO NETWORK LICENSEE REVISING PROCEDURE 1-0P-3,4, UN!T SHUTDOWN FROM COLD SHUTDOWN CONDITION (MODE 5)
- 200*F TO COLD SHUTDOWN (MODE 5) 5 140*F)
REVISED PROCEDURE WILL REQUIRE REVIEW FOR RCS MAKEUP AND LOSS INVENTORY AND ROUTINE SURVEILLANCE OF ALL AVAILABLE LEVEL INDICATION
- USE OF RVLIS INDICATOR TO BE SPECIFICALLY SPELLED OUT IN REVISED PROCEDURE (MODES 485). TS PRESENTLY REQUIRE RVLIS OPERABLE MODE 1-3 ALL PLANT OPERATORS PEING MADE AWARE OF EVENT
- LICENSEE'S HUMAN PERFORMANCE EVALUATION SYSTEM (HPES)
INVESTIGATION UNDERWAY HPES REPORT WILL BE SENT TO INPO
- 1 LESSGNS LEARKED WILL BE INCORPORATED IN LICENSEE'S TRAINING PROGRAM 3
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(- l' PEGIOil II a NRR MET AT SITE ON 6/24/87 AND DETERMINED NO FURTHL'R INVESTIGATION NEEDED PRIOR TO NA-1 RESTART 4
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J7 BRUNSWICK 1-SCRAM WITH COMPLICATIONS AND SRV COMMON MODE FAILURE PROBLEM: ,
RCIC AND HPCI FAILED TO AUTO START AND ONE RECIRCULATION PUMP FAILED TO TRIP AFTER REACTOR SCRAM FROM 100% POWER. ONE SAFETY RELIEF VALVE FAILED TO MANUALLY ACTUATE, ANOTHER FAILED DURING RESTART TESTING.
CAUSE:
NORMAL VARIATION IN BISTABLE SET POINTS CAUSED SOME AUTOMATIC ACTUATIONS AS LEVEL APPROACHED LEVEL 2. CURED LOCTITE FOUND ON PLUNGERS OF S0LEN0ID VALVES USED TO ACTUATE SAFETY RELIEF VALVES (SRV'S).
SAFETY SIGNIFICANCE:
EQUIPMENT FAILURES AND ANOMALOUS BEHAVIOR COMPLICATE SCRAM REC 0VERY.
COMMON MODE FAILURE OF SRV's USED IN ADS.
(~j DISCUSSION:
PLANT SCRAMMED ON TURBINE LOCK 0UT WHILE TROUBLESHOOTING VOLTAGE REGULATOR PROBLEM.
MSIV ISOLATION OCCURRED FROM CHANNELS A2 AND B2 (BUT NOT Al AND B1)
AS LEVEL APPROACHED LEVEL 2.
. l LEVEL 2 SET POINT IS 118" 3". RECORDED LEVELS WERE 117", 121", .j 119", AND 120" 0N THE FOUR CHANNELS OF THE ANALOG TRIP SYSTEM. !
HPCI TRIPPED DURING FIRST MANUAL START ATTEMPT DUE TO VALVE ALIGNMENT ERROR.
ONE OF THREE SRV's BEING USED IN MANUAL FOR PRESSURE CONTROL !
FAILED TO OPEN ON DEMAND.
l SRV's USED FOR ADS FUNCTION TESTED PRIOR TO RESTART. A SECOND j SRV FAILED TO OPEN MANUALLY. BOTH VALVES HAD JUST BEEN REBUILT BY l TARGET ROCK TECHNICIANS AT WYLE LABS. .
l ALL SOLEN 0ID VALVES ON ADS CHANGED OUT AND SENT TO WYLE FOR 1 EXAMINATION.
l 4* DISASSEMBLY OF SOLEN 0ID VALVES REVEALED LOCTITE RESIDUE ON PLUNGER ASSEMBLY OF TWO FAILED VALVES AND TWO OTHERS.
l
ADS VALVES MAINTAINED BY TARGET ROCK AT OTHEP, FACILITIES TESTED SATISFACTORILY.
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- TECHNICIAN WHO PERFORMED BRUNSWICK SRV MAINTENANCE HAD NOT WORKED ON OTHER VALVES.
FOLLOWUP:
ALL PLANTS'HAVING SIMILAR VALVES MAINTAINED BY TARGET ROCK WERE CONTACTED AND AS APPROPRIATE VALVES WERE TESTED.
TARGET ROCK PLANS TO MODIFY ADMINISTRATIVE CONTROLS AND CLARIFY :
PROCEDURES. POST MAINTENANCE TESTING IS BEING REEVALUATED.
INFORMATION NOTICE ON EVENT IN PREPARATION.
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