ML20141M340

Transcript of 388th ACRS Meeting in Bethesda,Md on 920806. Pp 1-159.Supporting Documentation & Viewgraphs Encl

ML20141M340
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Issue date:	08/06/1992
From:	Advisory Committee on Reactor Safeguards
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ACRS-T-1925, NUDOCS 9208110368
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OFFICIAL TRANSCRINOF PROCEEDINGS O- .

Agency: Nuclear Regulatory Commission Advisory Committee on Reactor Safeguards Til}C: 388th ACRS Meeting Docket No.

O

LOCATION

Bethesda, Maryland -

DATE:

Thursday, August 6,'1992 PAGES: 1 - 159 p

4 i

AC3835CO)yNASE .

fcrthe Ufe ci te Co ntes I

L

' ANN RILEY &~ ASSOCIATES, LTD.

!'O ^i j Q (, "" t ,>

1612 x s< ".w. se>te 300 Washington, D.C. 20006

-(202) 293-3950 j 9208110368 G20906 PDP ACR5 T-1925 PDP

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L.

PUBIC NOTICE.BY THE UNITED STATE NUCLEAR-REGULATORY COMMISSION'S ADVISORY COMMITTEE ON REACTOR SAFEGUARDS DATE: ^"9"" '

O The contents of this transcript _of_the-procee' dings of the United States Nuclear _ Regulatory Commission's Advisory Committee on Reactor Safeguards, (date)

August 6, 1992 , as' Reported =herein',-~are a_ record of

~

the discussions recorded at;the meeting held-_~on':the aboves -

date.

~

This-transcript has not been_ reviewed,.-corrected -

or edited, and-it may contain inaccuracies.

{-_.

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i i

b-I l'

OG 1 UNITED STATES OF. AMERICA 4

2 NUCLEAR REGULATORY COMMISSION-'

l 3 ***

i-

! 4 ADVISORY COMMITTEE ON REACTORLSAFEGUARDS=

l j S 1

! 6 388th ACRS Meeting I

l 7 i

8- U.S. Nuclear Regulatory Commission L

f 9 7920.-Norfolk Avenue 10 Conference Room P-110 j.

l 11 Bethesda, Maryland l.

[ 12 i 13 Thursday, August 6,l.1992 14 4.

4 -

[ 15 The above-entitled proceedings cor.menced at-- 8: 30-h 16 o' clock a.m.,-pursuant to notice,. David Ward,= chairman,-

f 17 presiding.

l.

p 18 l _

~~4eo .

j. 19 'PRESENT FOR THE ACRS FULL COMMITTEE: ,

20 .D.-Ward l :P.=--Shewmon: -

i jj 21 J. Carroll JI. Catton i

j, 22- W.~Kerr .T. Kress-23 H.' Lewis C. Michelson ,.

E. Wilkins _C.-Wylie-h .25 ~ W. Lindblad . R.- -Fraley '

i: x -

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( 1 PARTICIPANTS:

2 T. Murley J. Wilson 3 T. Boyce W. Russell 4 T. James A. Chambers 5 M. Rubin C. Berlinger 6 C. Morris E. Rossi 7

8 9

10 11 12 13

, 14 15 16 17 18 19 20 21 22 23 t

24 25 O ANN RlLEY & ASSOCIATES, Ltd.

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V _

3 i 1- P R O C E;E'.D I N G S 2

[8:30 a~.m.]

3 MR. WARD: The meeting-will-now come to order.

4 This is the first-day _of-the.388th:-Meeting of the 5 Advisory Committee.on Reactor Safeguards, 6 During todayfs meeting, the committee'will discuss 7 or hear reports on the following:' One, Inspections, Tests, 8 Analyses, and Acceptance _ Criteria,'or ITAAC, for selected-9 ABWR systems; 10- Two, the supplement to Generic-Letter 83-28',

11- Required Actions Ba sed on Generic Implications of the Salem 12 ATWS Events; 13 Three, the proposed: Revision 3.to Regulatory Guide 14 1.101, Emergency Planning'and Preparedness; 15 Four, proposed ACRS reports ea several ratters,--

16 including, especially-today, the NRC-Severe-Accident 17 Research Program Plan;-

18 Five, proposed'ACRS positioneon ITAACs for

19. selected:ABWR systems, and.the supplement to_ Generic-Letter -

20 83-28. -

21 The meeting.will be:conductediin accordance with

.22 provisions'of the FederaliAdvisory Committee 1Act.-

23 - Mr. Raymond. Fraley is ; the designated Federal-24 Official for the initial; portion-of the: meeting.-

25 We have received 4no' written 1 statements or requests-ANN: RILEYL &iASSOCIATES,'Ltd;

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4 1 for time to make oral statements from members of the public 2 regarding today's sessions.

.f 3 A transcript of portions of the meeting will be -,

4 kept, and I request that each speaker use one-of the 5 microphones, and identify herself or himself.

6 I have a number of items of current interest to 7 begin with, before we do our first technical agenda item.

8 First, we understand that the Commission has-not 9 yet voted on the AP600 testing issue. -They reportedly were 10 awaiting a staff response.to the 7CRS letter of_last month, 11 perhaps somewhat more recently_than that. __

12 I understand that we have just today-gotten, this 13 morning, hand-carried a copy of: tlA +?.aff response, _ but 14 haven't had a chance'to look at it yet, but theyLwill have 15 it copied,.and passed out,'so-that it will be available.

- 16 MR. WILKINS

There._is,_in our_RES Research 17 Committee report on this same subject,.which, if'I read it 18 correctly, comes to.a slightly different position than we 19 _did. In fact, a considerably different position.'than.We 20- did.-

21. MR. WARD:

Yes,'that is. correct.

! 22 MR. WILKINS: ~ And that may be. causing some l' ~23 confusion.

24 MR. WARD: I think-the-Commission has to_look at 25 all the advice it gets:and"make.up-its mind.

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l S

1 Another item, I altest regretLto.have to say this, 2 .but this.is Bill Kerr's last meeting.; In. fact, he'will be 3 standing up and leaving early on Friday afternoon, I guess, 4 or early Friday evening.

5 Our. intention,-if we can persuade him to-attend-on 6 Thursday night, is to honor him with a dinner.at 6:15 at 7 Positano's, not Bill's favorite restaurant, but'I think'he 8 will find it acceptable.

9 MR. CARROLL: We already took,him to his favorite 10 restaurant earlier this week.

11 MR. WARD:~ Let's see, a-third-item,~INPO has-12 invited us down for a general briefing on'their activities.

13 This meeting is= scheduled in Atlanta-for< August:24th. So-14 far, Michelson,.Wylie and Wilkins have signed.up, and 15 expressed-interest in attending. I think=it would be a--good. I 16 idea for more people to attend that, possiblyjthe date could.

. 17 be changed.if necescary. It seems to.me.it isiquite useful.

18 LI would like to: introduce Dr. Vernon Hodge,--who is' 19 going to:be with'us.

20 Vernon,._would'you stand up.

21 -Vernon is with us on rotational-assignment from 22- NRR, and he will be with us, aslI understandfit,~in the-23 months-of August and September. He'has been'with NRC-since.

24 1975. He.has just been working.in'the' Generic?-

25 Communications Branch., He has-broad experience within the ANN RILEY & ' ASSOCIATES, Ltd.

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6 b

U 1 agency and NMSS, IE, and NRR.

2 Before NRC, he vnrked at the National Reactor 3 Testing Station, and has de- io some work in physics at the 4 University of Idaho, and at UCLA.

5 MR. WILKINS: Fortunately, he was at NRTS out at 6 Idaho, so that compensates.

7 MR. WARD: I think Dr. Hodge is from the 8 government, and he is here to help us, so let's make us of 9 his experience and talents in-that next two months.

10 Bill Lindblad is with us, but still as a member 11 expectant, so we'have to treat him, and he has to treat us 12 accordingly.

13 Jay Carroll just pointed out to me this morning an 14 interesting item that'was in the ANS Magazine, Nuclear News, l 1E in the July-1992 issue, and it is a little summary of a l

16 report from a national research council ad hoc committee.

17 The report is, Nuclear Power Technical and Institutional 18 Options for the Future.

l 19 Apparently this was a study requested by Congress, 20 and it has some interesting recommendations about really

( 21 where DOE should be -- A number of things, but the most 22- interesting ones were on where DOE should be-spending its 23 money on its support of types of reactors to be developed, j 24 It was chaired by John Ahearne, which doesn't surpri e'any of us. He does a good job.with this sort of ANN RILEY & ASSOCIATES, Ltd.

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7 1 thing.

2 There is ' a report and, Ray, I think the me:abers -

3 would be jnterested in getting_ copies of that. If.you-would 4 -arrange that,:ve would appreciate it.

3 5 I would like to remind _you-that the meeting.is 6 scheduled to continue until about-3:00 on_ Saturday. . You 7 know we often schedule the meeting to continue'to 3:00 on

~

8 Saturday. The difference is we are serious'about it this 9 time, so I hope -- we gave you--some advance notice. :We got 10 -a lot of letters that were belo over from lhst month.

11 'MR. SHEWMON: When did we get'.the adva'nce notice'?L 12 MR. WILKINS: About 10 days ago.

13 MR. WARD: Right, 10 days ago.. I think a memo-14 went out to everybods.

15 If you need help with transportation' changes, I 16 think Tanya stands ready to help,'I believe.

17 Let's'see, we are supposed-to have here a list-of:

10 letters. Has that been: passed-out?

19 MR. BOEHNERT: Yes.

20 We would likeito' inaugurate a-new.

MR.-WARD:

21 ' procedure. We had some difficulty _last_ month with'an 22 important letter and we'd like to. inaugurate.'aln'ew-23- proceduru.

24 .In a' nutshell the procedure is=this. :If we have a:

25 letter to be' presented to the committee for= consideration,-a O .

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k-4 I; 8' 4

l( ) 1 letter that the schedule demands _ require that we get out,

f. 2 that actually be issued on Saturday -- in~other words, we p

j- 3 don't have the default option of if we don't like the j 4 letter, have trouble with it or don't finish it of deferring i 5 it to the next month, we'd likelto have t. draft of that l 6 letter made available to the members before close of I 7 business on Thursday so that-the members can take it back to

. 8 their hotel rooms and read these drafts.

! 9 Then on Friday morning, probably usually the'first t

! 10 thing but some time Friday morning we will have a short j 11 session not to discuss in depth each letter but just to get i

12 at least a soft vote around the table on those-letters that l _

13 have to go out, that.we perceive haveito go out,-a soft; vote-14 to determine whether the draft which we have: been presented-i d

15 is. good enough for our paragraph:by paragraph consideration..

i 16 If not, maybe the general' problems can be pointed.

p.

[ 17- out to the author and would'have to go back and work on it

18 Friday to get a better draft.

I- 19 That is the procedure we want=to-follow.

I-l 20 Now in-view of that-newl procedure, let's look at 21 the-long list of letters that we?have'got in: front of us 22-f --=okay,'it's just the'first page, butathere are'10 letters-i

[ 23 there..

24 MR._WYLIE: Mr.LChairman, where you have a meeting 25 on Friday that the; letter hinges-on, whatJdo'you doLabout

[

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9

( 1 that? Does the subcommittee chairman then anticipate what 2 he thinks the committee is going to do?

3 MR. WARD: Yes. I think we'll have to do that.

4 MR. SHEWHON: We are putty in your hands, as 5 always, Charlie.

6 MR. WILKINS: An alternative is for the 7 subcommittee chairman to write two letters, one of which he 8 says what he thinks he would like to do and then the other 9 is what he thinks the committee would do if they reject his 10 sound advice.

11 MR. WYLIE: A lot of that depends on the 12 discussion at the meeting.

13 MR. CARROLL: True.

14 MR. WARD: It does and I agree that presents a l 15 difficulty but I think we'd like to adopt this.

l 16 MR. MICHELSON: Those agenda items for which we l 17 must get a letter out at a given meeting,'they always have-18 the first hearing on Thursday. We reserve ~for Friday things 19 that don't have to get out at that time anyway. That_would.

20 alleviate it.

21 MR. WARD: That would help. That would help. To 22 the extent that we can schedule the agenda, that would be a 23 good way to do it.

24 We can't always do that but when we can,;okay._

25 Hey, Fay, would you take a note on that? I.think t%

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10 that's a good point.

/

) 1 2 MR. FRALEY: Yes, we will try to do that.

3 MR. WARD: Let's look at this list now and I don't 4 want to take too much time but if we can take three or four 5 minutes and look at this list and decide which of these 10 6 items, 10 letters, really needs to get out this month.

7 I think the first one does.

8 MR. WILKINS: Yes.

9 MR. WARD: I think the second one does.

10 MR. MICHELSON: It is already pretty well done.

11 MR. WARD: In some cases the drafts are already 12 available on some of these so there is no particular problem 13 but I am just -- let's see, the third one, the policy 14 issues.

15 Charlie, what about that? Do you perceive that to 16 be a problem?

17 MR. WYLIE: I know the Staff would like to have 18 it but I don't know how possible that is. Perhaps the Staff 19 could comment on that.

20 MR. WARD: Let's give it a definite maybe here for 21 now. It's available anyway, so it's sort of moot, I think, 22 but okay.

23 The proposed Reg Guide -- I don't know, Paul.

24 Number 4, I guess it's a holdover from last month. I would 25 sure like to get it out.

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i 11 1 MR. SHEWMON: It's March 1, yes.

l 2 MR. WARD: It's not exactly a top orgent_ issue.

3 MR. SHEWMON: I don't think it will be 4 controversial. I would like to see a reading of it.

l 5 MR. WARD: We will give that a yes, and 106,-does l

l 6 that need to get out?

(

7 MR. MICHELSON

Well, it should get out pretty __

i 8 soon but it waited this long. . It could wait longer, I 9 guess.

10 It's virtually ready.

11 MR. WARD: Okay,-but it's just a matter of what's 12 needed, not what is possible.

13 Okay, selected ITAACs. I think that needs to get 14 out.

l 15 MR. CARROLL: Yes.

16 MR. WARD: The~EPRD, Charlie, should that?

l 17 MR. WYLIE: -They are issuing in August their 18 report.

19 MR. WARD: All right, that-is' desirable, I guess; 20 83-28?

l 21 MR. WYLIE: Yes. They would like to have-that-22 thing. We have got a draft of that.

23 MR. WARD: You know,.I think there'is.asdifference

! '24 in the Staff would like to have and we want to cooporate' 25 with them and we can, but when it's -- there's a difterence-bq ANN RILEY & ASSOCIATES, Ltd.

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12 1 between that and " urgent."

(

2 MR. CARROLL: I think we need some kind of a code.

3 A is urgent; B ir desirable; and C --

4 MR. KRESS: We won't miss the plar for.

5 MR. WILKINS: I suppose they are all desirable.

6 MR. WARD: Let's see, 83-28 is the --

7 MR. WYLIE: We have a draft available and --

8 MR. WARD: Oh, we do?

9 MR. WYLIE: A proposed draft.

10 MR. BOEHNERT: If we don't write a letter th3s 11 month, it will be issued without our comments.

12 MR. WARD: Charlie, you said a draft is available 13 now?

Q 14 MR. WYLIE: Is in typing.

15 MR. WARD: Marginal requirements -- anybody know 16 anything about that? Lewis isn't here.

17 MR. HOUSTON: I don't know what the priority is.

i' 18 There is a SECY out that I assume we will be responding to.

19 MR. WARD: How about soon?

20 MR. MICHELSON: There's another comment back

21 there.

l 22 MR. SAVIO: EDO holds the report on the progress.

23 The EDO I think wants to respond to that in. September, j 24 That is the regulations part.

25 MR. WARD: This is part of a follow-up on that, i .O .

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13 1 okay, and Ernest, Reg Guide 1.1?

2 MR. WILKINS: I don't know what the Staff priority 3 on that is.

4 MR. DURAISWAMY: That one I think NUMARC has 5 scheduled a workshop on that. The Staff would like to have 6 ACRS comments I believe before that.

l 7 MR. CARROLL: B.

8 MR. WARD: We will give that a " desirable," okay.

9 Okay, is there anything else that should come up 10 for general discussion before we go to the next agenda item 11 on ITAAC?

12 [No response.]

l l 13 MR. WARD: Okay -- inspections, tests,-analyses,

(

(, 14 and acceptance criteria.

l 15 We had a subcommittee meeting yesterday, a joint l

t 16 subcommittee meeting in which we had presentations which 17 gave from both the NRR staff and from GE, which gave us a 18 general picture on the status of the. development of ITAACs.

19 Then in addition to that, the general 20 presentation, GE described to us -- we'd scheduled a little 21 more of an in-depth review, not great de;th but with some 22 depth a review of five sample ITAACs. There are five ITAACs 23 which we -- of the system, of systems.

24 These are five out of what are or will eventually.

25 be about 85, and we selected this not as a way of providing Oi V

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14

() 1 -- making an in-depth review of each ITAAC itself, but as a 2 way of sampling the system, tha process.

3 We had been requested to review the ITAAC program 4 and process by the Commission in the report, comment to them 5 on what we thought about it. So that was the purpose of the 6 meeting yesterday, to begin that review and we want to 7 complete the review over the next two hours here this 8 morning.

9 We've scheduled presentations from both the staff 10 and from GE. We won't attempt to go over -- in i Ot, even 11 yesterday, we didn't cover all five of these system ITAACs, 12 but we covered sort of three, plus fractions of another one.

13 What we have planned to do this morning is to go over just O

(/ 14 one of the ITAACs in some detail, and then I think GE bas 15 the slides for the others,.and we can sort of flash these up 16 on the screen, go through them, and give you an opportunity 17 to ask questions; in other'words, go through those rather 18 quickly.

19 So that's the plan for this. morn'.ng, and there is 20 information on this behind Tab 2 in your. book. A number of 21 us_were at the meeting yesterday, and--it always helps me to

~

22 hear things twice, and some of you will be hearing this for.

23 the first time.

24 I think this is a very important topic. This is 25 key to success in the Commission's certification program, so rT

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[/}

w 1 we need to give it our attention.

2 Do any of the other members who attended yesterday 3 have anything they would like to say at this time?

4 [No response.]

5 MR. WARD: Okay. As we go along, I invite the 6 other members, you know, to ask questions or comment to 7 expose the non-subcommittee members to the issues which you 8 consider to be important coming out of the review yesterday.

9 Okay. Well, let's begin. We have Tom Murley with 10 us this morning, and I understand Tom would like to make a 11 few introductory comments. Tom? j 12 MR. MURLEY: Thank you, Dave. I do have just a 13 few minutes to talk to the committee about where we stand l'h

(_,/ 14 with regard to the overall ABWR review. Perhaps it can help 15 you put in perspective what you eve going to. hear later this 16 morning and also plan your long-ti m reviews.

17 The staff is going to issue a draft final safety 18 evaluation report on the ASWR in'about one month from today, 19 and I should stress that it will not be a final product, but 20 it will be a full scope FSER. That is, the breadth of the 21 material that we'll ultimately publish will be outlined in 22 this draft that we're publishing, but it's going to contain 23 about 340 open items.

24 Some of the items arc, we know, very important to 25 you and to us. For example, the PRA material, the severe J.n ANN RILEY & ASSOCIATES, I'd.

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16 1 accident material, and a large-number of-the~ITAAC will be 2 missing because we did not have the submittals in time'for:

3 our cut-off date,'which I believe was in.May some time, for 4 'us to prepare this draft.

5 There remains a good deal of work to be done_yet 6 by GE and by the staff to produce a final' product.. I_would 7 categorize this work in two~ main areas. We have to 8 systematically work off the'open items andlthis. depends most 9 strongly on how fast GE.can complete the design work and j 10 complete the response to_our questions and answers.

11 The second aspect of:thetwork that remaine to be-12 done is what I call'a comprehensive final review for quality 13 and consistency, and this must be done both by GE in their_ l 14 shop and here by the staff, and I'll talk just a minute 15 later-how we're going to'do that.

16- I realize it_is an imposition on:the committee to-17 ash that you begin reviewing a draft document that has many 18 open items. Nonetheless;_I believe that'with all the review 19 that has been-done upito date on the.ABWR, that-it will save ,

20' time-in-the long runLto oeginfreviewing the: full scope. draft 21 FSER in order'for you to:~ identify:the: areas that:Lyou want to 22- focus-on and where;you think.more' work or more:information- ,

L 23 -is needed.

24. Therelis one. area in particular, and-that's'ITAAC, 25 the topic you're going to hear about today,..that requires-a' *

.b x) .

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17 1 lot more work before we will be satisfied even. But we b[~h 2 believe the format and the scope are generally correct, so I 3 think it's worth your effort and time to focus on the 4 direction that we're heading and the format and how we 5 intend to make our final safety judgments even though some 6 of the details may not be correct or there may not be all 7 the detail that we think we need.

8 Now, we have had internal to the staff two 9 independent groups review the overall licensing process that 10 we're using to implement'Part 52. We've also asked these 11 groups to look at specific information in the SSAR, in the 12 Tier 1 design document and the ITAAC, and we've also had a 13 group look at the inspectability. aspects of the ITAAC.

O)

(m, 14 The more senior group we call the graybeards, and 15 this was a very experienced group consisting of Jim Snierak, 16 Ed Jordan, Stu Ebneter, Themis Speis, Bill Kane, Guy 17 Arlotto,-and Joe Scinto.

18 They met over a three-month period, from May l

19 through July, and they've just really finished their work, 20 and they have given me some letter reports on their work.

l l 21 They have given some very important conclusions. It was a l 22 group that I set up and they were advisory-to me.

23 Let me just read a couple conclusions that they 24 gave. Overall, the committee found that the design 25 certification process could provide sufficient information O ANN RILEY & ASSOCIATES, Ltd.

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1 for the staff to make a final ?afety determination on the GE 2 ABWR-design. However, brsed on the material reviewed by the 3 graybeard committee, the informatfoa that has been provided 4 thus far by GE is not sufficient for the staff to maketa-5 final: safety decision, and we agreed with that. T 6 The committee also noted a number of specific-7 areas of weakness or deficiency.

8 By the way, I will make sure that the ACRS gets 9 these letters so you have them as well.

10 The committee noted:some QA type problems, where 11 the material in the-SSAR.did not agree.with the-material in 12 Tier 1.

L 13 MR.. CARROLL:' We noted' a ' few of those, tio.

! 14 MR. MURLEY: Yes, I'm sure-you-did..- -In general, 15 the Graybeard committee concluded that the ITAAC did not -

16 contain sufficient' detail,-and C one that's very important~is,.

-17 they felt that-in the ITAAC, there.was an over-reliance on 18- process inspections: versus ~ observable' testing. - I'll'come4 -

19 back to that in a minute,:becau k that's an.-important point'.- -

20 Finally, the Graybeards'made alrecommendation to 21 me that re form a special review group.of technical .

l- 22 treviewers and regional field experienced-people not-

! .23 heretofore.. involved with tnisLreview,:to perform a-hundred 24^ percent review of the GE ABWR-ITAACs in order to, ensure that'-

25 - the problems-identified:are corrected, -

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19 1 MR. MICHELSON: I didn't quite understand. What 2 was that last comment? How was it made, that it should be 3 done?

4 MR. MURLEY: Yes, it was a recommendation to me.

5 MR. MICHELSON: That there be a hundred percent 6 review?

7 HR. MURLEY: That there be a hundred percent 8 review. We are going to do that. I'll talk about that in s 9 just a second.

10 There was a second group, primarily from the 11 regions. It was chaired by Bill Beach from Region IV, and 12 it had representatives from each of the regions-as well as 13 from headquarters, and they focused primarily on another 14 question, although very closely tied into this, which is how 15 can the staff make inspection findings of acceptability 16 based on the Tier I and the ITAAC material that we have in 4 17 the draft?

18 They concluded also.in -- that,-in general, the 19 Part 52 process is sound and wcrkable, and-the second group 20 had difficulty reaching a conclusion on the' level of detail 21 that-is needed in Tier I and ITAACs. - Some thought that with 22- the changes that we were making,: that it would be acceptable 23 and there was a minority opinion.that. felt that even with 24 the changes that we're contemplating,.that there's not 25 enough detail to reach final safety judgments.

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-20 1 -Then they made numerous suggestions for changes in 2 the Tier I and the ITAAC material. Here again, we will-send 3 all of these reports to.the Commission and to the ACRS. ,

4 Following the last of the Graybeard 4

5 rocc7mendations, I am setting up another group. This is 6 going to be under John Craig, who, you recall, has not been 7 associated with this design, but he his in charge of the C License Renewal Branch, and he's a very capable person.

1 9 He will be_headina a group composed of-10 headquarters and regional people, and they're going-to 11 perform a hundred percent review of the Tier I and the ITAAC 12 material as it is revised over the.next several months.

13 Now, keep in mind, we're going.to have parallel activities 14 going on.

15 We're sending comments-back and forth all-the time 16 now to General Electric. They're revising material'all the 17 time. As this material becomes more or less in' final shape -

18 according to ou reviewers,--then Craig and his people will 19 also be reviewing it in_' parallel to make"sure that'there's 20 consistency between all theLITAACs, between what we say in.

21 -our SER and what's in the application, the SSAR,:and what's.-

22 in. Tier I and:ITAAC.

23 He's also-going to be looking-at--thel inspect-24 ability-~. aspects of this. Now,-this-la'something:that the 25 Committee has recognized, but I'dclike.to emphasize it.

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() 1 2

This is the first time that any of the licensing reviews that's been done by the staff, that I know of,-has really 3 emphasized how the regional fniks are going to inspect these 4 findings and how they're going to conclude that a particular 5 requirement has been met.

6 And it's become clear to us -- I mean, I 7 recognized it when I set this Graybeard group up, that we 8 needed some people with broader wisdom and with regional 9 experience to take a look at it, and their findings have 10 reinforced my belief, and that is that we need people who 11 have actually -- actually are going to have to be in the 12 field doing inspections of pre-op testing, inspections of 1 13 startup testing, and inspections of construction, even, to O(_/ 14 make sura th. that a reviewer back here in headquarters 15 thinks is a requirement and a way of meeting that 16 requirement, is really achievable in the field.

17 I don't fundamentally see any problems in doing '

> 18 this, but it -- that has been a weakness in our review up to 19 now. Now, GE, we understand, has budgeted and will be doing 20 the same type of QA review for the entire SSAR, as well as 21 Tier I and ITAAC, this Fall.

22 Finally, we understand that NUMARC will be 23 conducting an industry review of the material in the 24 application, and this is principally Tier I and ITAAC, we 25 understand. This NUMARC industry group will be using i

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'l 22-1 utility people, architect engineer people and other vendor '

2 people over the next-two or-three months.

3 It is, if one wanted to characterize it, I think 4 it's an industry Graybeard review, and we welcome it. We 5 think it-is very useful for the utilities to take a look and 6 to make sure that they are satisfied with the ITAAC as well. ,

7 Now, --

8 MR. WARD: Will the industry group have -- and GE, 9 for that matter, have access to the draft SER at that time?

10 MR. MURLEY: Yes.

11 MR. WARD: Okay, in both cases they will.

12 @. MURLEY: Yes. They'll have access to.the l

13 draft SER. and.they will - =we'll send them copies, of 14 course, of the -- our Graybeard review and our independent 15 reviews as we do them.

l l 16 It's going to be a very busy six months, and we, 17 of course, will work with the ACRS.here(all during this

18. time. I'm not predicting how long this will'take.-

There's

19 just too many variables, but it's very high priority for the L

L 20 staff, the NRR staff.

21 MR. WARD: 1Well, you've got:a--drop-dead date, iso-- ,

22- you'*e asking soma sort of prediction on how:long-it'will; 23 take, right? I'mean,_you want-to issue'the-FDA'by the end 24 of the year?

25 MR. MURLEY: ' Yes, that's-our target. I'm not sure EO . ANN RILEY &: ASSOCIATES, Ltd. - -;

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1 we can make it.-

2 MR. WARD:' That's what'I wanted-to know. Is that '1 l

3 what you're saying?

]

4 MR. MURLEY: Yes, yes. I'm not going to-sign off 5 until my staff and I are satisfied, and~we're quite'a ways

\ ..

6 from being satisfied now with the' quality.. I think we 7 finally got the scope, our arms around the scope of the ,

8 thing.

9 What we're finding is that there are some gaps ind-10 some errors and mistakes and that sort of' thing'. -

11. MR. iCCHELSON: ~ One.of the issues,-of course,-is 12 whether the ITAAC should come before the FDA. . Do you have 13' some views on that?'

14 MR.-MURLEY: .I'sure do. I have=very ' strong views 15 on that. As we have gotten into this, we've concluded.and-l 16 I, personally, have concluded;-that~I can't--make a safety 17- judgment on a' design without:theLITAAC.. That's-how strong I' 18 feel-about-it.

19 So,-the notion:of.'decouplinglthem.or splitting, I

.20- think, is a theoretical--idea, and-asEweiget:into it, I'm noti h -21, inclined!to do that.

L 22- MR. - ' SHEWMON : . -One of the things thatlcame up 23 yesterday -- aid ~other members..c'an probablyLspeak:with more 24'  : authority on' detail 'was-the11arge number'of numbers that-

- 25 . were. coming-in Tier lI.- Thatlmeans-then thatuit's;goingLto O .

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() 1 change some of these numbers and you've got to go through a 2 rule change, and so it would be nice if you really didn't 3 think they were essential to safety, not to have numbers to 4 check against up there. Have you gotten into this at all?

5 MR. MURLEY: Yes, that's going to be part of the 6 review, the systematic review, that the Craig group does.

7 To be consistent, I mean, you're exactly right. There must 8 be some numbers in there, temperatures and flows and that 9 sort of. thing. But we're going to systematically go through 10 and ask ourselves, do we need all these numbers, or could we 11 put it in a Tier II document?

12 You're right, that's going to be part of our 13 review.

( .

14 MR. MICHELSoN: one of the fundamental issues that 15 came up yesterday during the Subcommittee discussions was 16 the question of what do we do about references that might be 17 appropriate for Tier I, references to-codes and so forth?

18 Also, you could use references to the SSAR to 19 . avoid putting numbers directly into the Tier I. The-e'.9 20 still a lot of fleribility then of changing the numoers back 21 and forth in the process. I gather that the discussion --

22 and correct me if I'm wrong -- that GE was-objecting to 23 putting numbers -- putting references into the Tier I 24 material because somehow it brought the Tier II material 25 into Tier I in some kind of legalistic rulemaking fashion.

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() 1 2 of that?

I wondered if you'd like to comment on the status l

3 MR. MURLEY: Yes. In general, this was a comment 4 that the Graybeards had as well, that there was-an 5 inconsistent reference to codes'and standards, for example.

6 What we're going to do is, in the Tier I and ITAAC 7 documents, we're going t . reference the codes and the 8 standards that we're requiring but the actual version will 9 be in Tier II, so that au that -- keep in mind that the 10 deference between why we have a' Tier I and a Tier II.is that 11 the Tier I material is that which cannot be changed, except 12 by rule.

13 So, it is meant to be the distillation e' those O 14

(_j/ features of the design that are so important that_you don't 15 want them changed easily. But that's not the say that Tier 16 II material is not important. It is, in my mind, just as 17 important. The only difference is that-it can be changed 18 through a process, the 50.59 type process that still 19 requires a careful review before you change it, but it 20 doesn't change the -- it doesn't introduce a.new un-reviewed 21 safety issue. That's the criterion.

22 MR. MICHELSON: Would you care to comment on the 23 question of referencing a section of the SSAR where you 24 would' find the numbers that you might wish to use?

25 MR. MURLEY: If I understand you correctly, the l

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26 IT 1 SSAR is Tier II.

V 2 MR. MICHELSoN: Yes, but in Tier I, instead of 3 saying 20 degrees or whatever the number might be, you can 4 indicate the section where the number can be found, and that i 5 allows the number to vary from time to time.

6 MR. MURLEY: No, no, if we reference it in Tier I, I 7 it effectively becomes -- I'd have to think about that. I

8 think it effectively becomes part of the rule.

9 MR. MICHELSON: That was the question on the

'0

. table, and by the same argument, if you reference a code, 11 the code becomes a part of the rule, is the argument.

12 MR. MURLEY: Yes, the code does, but it may say, 13 we'll nave to word it properly so that as versions of the

() 14 code come into play, you don't have to go back through the 15 rulemaking, and I think we can do that.

16 MR. RUSSELL: I was going to clarify that the 17 approach that we're proposing -- and this is going to be in 18 a Co mission paper which is on its way up to the Commission, 19 as it relates to codes and standards -- the example of 20 citing the ASME code in the Tier I material is a good 21 example 22 We would not include the particular version of the 23 code or the addenda of tN, code in Tier I. That would, 24 instead, be in Tier II. What we're proposing is that a COL 25 applicant would follow the version of the code or addenda of (3

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27 1 the code which is in effect based upon the time of COL 2 application which had been approved by the NRC.

3 This way, later editions of the code prior to --

4 application are covered. We are not creating an-5 inconsistency then between Tier I material and Tier II 6 material. So, that is the approach that we're taking.

7 In some instances, we may wish to specify a 8 particular code version addenda. The only example of that 9 that I'm aware of is dialr>gue that's going on with respect .j 10 to the codes and standards which would be in effect for 11 construction of the reactor pressure vessel because we are i

12 quite specific on details of the pressure vessel in the 13 design certification, and we may 9 ant to keep that i 14 standardized.

15 So, there may be some instances where we, indeed, 16 for good reasons, want to cite a particular code or standard 17 with its version'in Tier I. .I think those are goi'ng to be 18 very few and far between. Most will-be the details in Tier i 19 II.

20 MR. MICHELSON: The question was. raised as to 21 whether or not an inspector can-inspect _any - _he can 22 inspect the plant against_any of the Tier II material.

23 MR. MURLEY: . Absolutely..

24 MR. MICHELSON: _ Of course,_the_ Tier _I.is_ [

25 mandatory inspection as I would see it.

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28 1 MR. MURLEY: Yes.  ;

2 MR. MICHELSoN: But Tier II is optional. Any part 3 of that has to also be met. For instance, Tier I drawings 4 don't show much in the line of a flow sheet. Tier II I 5 drawings have a complete flow sheet. If the inspector finds 6 a missing component that is not shown in Tier I, but is 7 shown in Tier II, I think you can question that.

8 MR. MURLEY: That's a very good point. Tier II is 9 part of the application and it is part of the certification.

10 It's just not -- it doesn't have what we all issue- 1 11 preclusion. It has material -- as Tier I does, but it is an  :

1; essential part of our overall finding, and so the material 13 in Tier II is essential.

14 Now, as part of our review of inspectability, we 15 came upon this very question, and that is that the-16 inspectors who are carrying ou& their normal inspection 17 activity, just like we~do now during construction, they'are 18 going to have detailed drawings and P& ids in their hands, 19 and it raised the question, how do these' detailed drawings 20 and P& ids and detailed documents relate to what we 21 certified? And we don't'have at this stage a clear 22- connection between those.

23 So what we're-going to require,;and we're 24 proposing this to-the Commission '-- we've got a paper that: -

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29 1 Commission that will require the COL applicant to produce

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2 what we call a bridge document, and this bridge document is 3 going from the Tier 1 and ITAAC material to the details that 4 are used to construct and conduct pre-op tests and that_ sort 5 of thing. They're going to have to certify -- the COL 6 holder will have to certify that these detailed drawings are 7 consistent with Tier-1. i 9 Of course, we'll do our own overview check, but 9 we're going to lay that on the COL holder, so that we will 10- have high confidence that what the inspector is using is in 11 fact consistent with what we've approved back here in 1992.,

12 MR. WARD' You said consistent with Tier 1. Do 13 you mean that, or is this Tier 1 and Tier 2?

14 MR. MURLEY: Yes, both.

15 MR. WARD: Okay.

16 MR. MICHELSON: It's hard to be consistent _with 17 Tier 1. There's not much in that drawing.-

18 'MR. WILKINS: It's hard to be-inconsistent.

I.

l 19 MR. MICHELSON: _ It's hard to'be inconsistent with 20 Tier 1.

21 MR. RUSSELLt If I could clarify with respect to 22 the Tier 2 material, Tier _2 may be changed. Provided the 23 change does-not impact the certified design that is in Tier 24 1 and does not constitute an unreviewed safety question, you-25 can use a 50.59_like process, b' ANN RlLEY & ASSOCIATES, Ltd.

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So if we presume that they have done that, they have done an engineering analysis of that, we would expect 3 that the P&ID which is in Tier 2 would be changed to reflect 4 that modification that they had made such that the inspector 5 could still use the P&ID or the bridge document to do the 6 physical walkdown to ensure that the plant as built matches 7 as intended or as designed. The only difference in Tier 2 8 is that there is a permissive that a change may be made 9 under certain conditions.

10 So the approach that Tom was discussing with the 11 bridge document, we're essentially going to be looking at 12 more in engineering; that is, how was the engineering done 13 to convert the certified design into the drawings and 14 details that are released for field construction?

15 Then there will be another bridge that goes from 16 the details in the certified design regarding ITAAC pre-op 17 testing, in fact their typical test procedures for system 18 turnover at the end of construction, for system testing 19 during pre-ops. We're going to be expecting that those will 20 be through this bridge document approach certified to be 21 consistent with the design certification and the Tier 2 22 material as changed such that you have indeed a good track 23 through the various levels.

24 It's really the change mechanism that strongly 25 differentiates between Tier 1 and Tier 2, where in Tier.1, f'b l \~)

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[,,)T 1 we're saying you'd need to go back to rulemaking to do that.

2 It does not mean that Tier 2 is not important or that Tier 2 will not be in fact used for inspection.

4 MR. MICHELSON: There is little if any difference b between what you're proposing and what presently has been 6 done in terms of inspection.

7 MR. RUSSELL: As it relates to that aspect of it.

8 There are other elements that would be different. We're 9 proposing a sign-as-you-go process, where we would sign off 10 on things as they're completed. We are proposing to have a 11 different structure with resources on site so that we're not 12 delaying activities with a new approach to how that would be 13 managed, t

r~N k ,)

s 14 But the process would be very similar to the 15 current experience, with the new requirement for us to make 16 a finding that ITAAC have indeed been satisfied and 17 documenting the basis for that.

18 You know, we would not do it 100 percent, but we 19 would audit sufficiently to conclude that the ITAAC had been 20 completed, and then we would notice that in the Federal 21 Register at the appropriate time.

22 MR. MURLEY: Okay. That concludes my opening 23 remarks.

24 MR. WARD: Okay.

25 MR. MICHELSON: I have one other follow-up i

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I question. We haven't heard much about how you propose to i

2 inspect or monitor the 50.59 change process as it's 'toing 3 on. Do you care to give a few words on that?

4 MR. RUSSELL: What I would like to propose, since 5 that is a key element of the draft Commission paper which we 6 expect to get to you in about a month, is that I'd like to I

7 get the description of what the 50.59 like process is to 8 you.

9 We are still having dialogue at this time with oGC 10 and we have not finalized the paper. So while I could-11 discuss what is our. current thinking, it may change as it 12 goes through the review based upon some legal nuances.

e 13 MR. MICHELSoN: It is covered there.

14 MR. RUSSELL: It-is in the paper, and we'd be back 15 to you within about two months and could discuss it at both 16 subcommittee and full committee.

17 MR. WARD: Let me ask you-a question. If in this 18 50.59 like review the analysis shows;that there'is an 1- 19 unreviewed safety. question.--

20 MR. RUSSELL: Yes.

21 MR. WARD: -- what'happens?.

I 22 MR. RUSSELL: okay. Let me. clarify one thing that 23 we are attempting v.o accomplish both as it. relates to what-24 would be- the definition of an unrelated safety question, 'and 25 we're proposing a broader definition thun has been usedlin-i -

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33 1 the past. We are proposing, in fact, to incorporate the 2 staff safety evaluation into the process from the standpoint 3- of identifying what could constitute an unreviewed safety l

4 question. l 5 This'is particularly important in the DAC area, l

6 where analysis methods are relied upon to reach a safety l 7 conclusion, and we're into process controls-where we have -

8 reviewed a particular process.

9 So we're suggesting that to the extent the staff's 10 safety evaluation explicitly documents material in Tier 2 11 which the staff has relied upon in reaching its safety 12 judgment, that that material may not be changed without 13 approval of the staff.

14 Now, that would be an additional -- it's a lower 15 level of approval-than rulemaking, but it would'still 16 require staff approval.

17 MR. MURLEY:. Dave, with regard to your question of' 18 what happens-if we determine that the 50.59 change does in i 19- fact introduce an unreviewed safety issue,'then.what we .

20 would -- of course, NRC has. final' judgment on that matter, 21 and it would require an exemption, or they would have to --

22 if we didn't want to issue an exemption, we would have to-23 make them change'the design, change it back to an acceptable 4 24 -design.-

25 MR.. WARD: Canweissue'anexemptionwithout] going-

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() I through the rulemaking process?

Yes. We can make an exemption to this 2 MR. MURLEY:

3 certified rule, just like we can any other of our rules.

4 MR. RUSSELL: Let me clarify also. Jerry pointed 5 out a very good feature of what we sre talking about with 6 respect to an unreviewed safety question.

7 An unreviewed safety question would be handled 8 through the normal amendment process. That is, you have a 9 combined operating license so you have a license and that 10 can be amended so that the approval process, should you run .

11 into the situation where you wish to change something that 12 is in tier two that the staff has relied upon, that would be 13 within the scope of an unreviewed safety question, they 14 would then submit an amendment which would be reviewed 15 through an amendment process to identify, for example, an 16 alternate method of performing a calculation associated with 17 a design acceptance criteria.

18 That amendment process would have the normal 19 notice opportunity for public participation. Whether it is 20 prior to effectiveness or after effectiveness would be 21 subject to whether we found it had a significant hazard or 22 not.

23 MR. WARD: Okay, thank you very much, gent) amen.

24 That's correct.

25 MR. MURLEY: One final point. Bill mentioned

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( ) 1 amendment. I mentioned exemption. Either one could be used 2 to deal with a change like this.

3 KR. RUSSELL: Generally the exemption would be 4 against a requirement that might be in tier one, that is an 5 exemption from the rule, where an amendment would be a 6 change to tier two through an amendment process. That's the 7 tying in the Staff Safety Evaluation and so what it means is 8 tnat the Staff needs to be very careful in how it writes its 9 SER from the standpoint of clearly identifying material 10 which it has relied upon in reaching its safety conclusion 11 because it is that naterial which would result in a need for 12 an amendment.

13 If we are silent, if we say do a review and we use 14 a standard review plan type approach and we only identify 15 the exceptions and we do not identify those things which we 16 found acceptable, if we are silent by virtue of that silence 17 they may make a change pursuant to 50.59 and do it without 18 prior Staff review and approval.

19 MR. MICHELSON: Are you going to have some kind of 20 a flag in the SER?

21 MR. RUSSELI : We are looking at how we physically 22 implement this now but we are principally doing it through 23 the safety evaluation process and if you recall the approach 24 we took on the DAC for piping, we were very explicit with l 25 section by section on methods, what we concluded, et cetera, l

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l 35 1 what we found acceptable.

2 MR. MICHELSON: But you are also explicit in a j 3 number of other places where I don't think you intended to i

4 be that firm. If you are using language format to identify, .

S that's tough. If you have a flag of.some sort, I can't miss I 6 it.

7 MR. RUSSELL: We will take that under 8 consideration but right now it would be basically the 9 . concept is a language _one. We-have to look at how we 10 implement that but we are essentially saying that there is 11 material which is in tier-two-which is critical to the 12 Staff's reaching a safety finding where you are talking 13 about analysis to be performed in the-future.

14 We recognize that there could be a new analysis 15 method that's developed in the future. We-would want to 16 have the ability to take advattage_of that or the licensee 17 may wish to take advantage of vhat and so we have identified 18 a process that would allow that to occur b'ut it would occur 19 based upon a Staff review and approval with a notice 20 process, which would be a normal amendment.-

,_ 21 MR. MICHELSON: Could you tell me roughly what you '

l .

think the criteria might be in determining whether you get 22 i-23 an exemption to-tier one, or-you have_to do a new rulemaking .

24- for tier one? -j 25 MR.' RUSSELL: _It-is:the same criteria that' exists.

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for exemptions now. We would have to make the same findings.

MR. MICHELSON: But what roughly is it though? I 3 mean what differentiates between getting an exemption and 4 having to go through a new rulemaking?

5 MR. RUSSELL: Those aspects are covered both in 6 the paper, form and content paper, and I think we'll have a 7 number of meetings. I would like to discuss those when we 8 have counsel available also to answer questions because we 9 are working jointly with OGC.

10 MR. MICHELSON: All right.

11 MR. WARD: Let's turn now to a discussion'of- ,

12 ITAAC.

13 Who is going to lead off? Tom Boyce?

14 MR. BOYCE: Yes.

15 MR. WARD: All right. Tom?- Let's go ahead with 16 your presentation.

17 (Slide.]

18 MR. BOYCE: Good morning. My'name is Tom Boyce.

19 I am a Project Manager at NRR. I am here to discuss'ITAAC 20 for the GE ABWR. This is a follow-on to the subcommittee 21 presentation of yesterday.-

22 (Slide.)

23 MR. BOYCE: I-will go through these slides fairly 24 quickly.- I have tenlof them.- Stop me if you have any 25 questions. In-view'of the time, I'll go through them

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. 38 I relatively rapidly.

2 This is a summary of where we are. I would start 3 with this bullet: ITAAC development continues to be 4 iterative.

5 Dr. Murley referred to several ongoing efforts in

' 6 regard to the review of ITAAC. GE discussed with the 7 subcommittee yesterday some of the init.latives that it is 8 taking in order to correct inconsistencies and to upgrade 9 the quality of the material.

10 This is an understatement -- ITAAC implements  :

J 11 several aspects of 10 CFR, Part 52.

12 (Slide.]

, 13 MR. BOYCE: This slide shows some of the SECY 14 papers that have gone into the development of ITAAC. Most 15 of ITAAC has been developed and' discussed in these SECY 16 papers in a very high level-form.

17 The last bullet here notes that the I&C-and-the 18 human factors DAC is expected later this month, also the-l .

19 form and content paper which-doesn't pertain directly to 20 this was alluded to earlier.and it will be coming out-21 shortly.

22 MR. WARD: It is this' form and' content paper that

-23 was going to include the description of the.50.59?

24- MR. BOYCE:. That's-correct'.

l 25 (Slide.]

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MR. BOYCE: The reviews proceeded in this fashion:

GE has submitted the ITAAC in three stages, and we've held 3 senior management meetings to discuss them, approximately 4 every six to eight weeks. It started with nine pilots in 5 September of 1991.

6 After extensive discussions and back and forth 7 type of comments, Stage 2 was submitted in April of 1992, 8 and the full ITAAC submittal was submitted in June of 1992.

9 MR. WILKINS: Full means full as perceived by GE, 10 or does the staff concur that it is full?

11 MR. BOYCE: GE submitted what they considered a 32 full ITAAC. In light of the ongoing reviews that we have 13 had, we are supplementing the information in that, but I

(_, 14 believe the format and scope, as was alluded to earlier, is 15 probably complete.

16 MR. WILKINS: Thank you.

17 MR. WARD: I was confused by what you said. You 19 said you're supplementing the information.

19 MR. BOYCE: We are providing a great deal of 20 comments.

21 MR. WARC: You're commenting back to them and you 22 want them to supplement.

23 MR. BOYCE: That's correct. My terminology was 24 not exact.

25 MR. WARD: So there's going to be a Sttg' IV.

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40 (J 1 MR. BOYCF.: I'll let GE answer that one.

2 MR. JAMES: We agree that the Stage 3 that we 3 submitted has all of the sections that we believe are 4 necessary, but do need further submittals to reflect i 5 incorporation of comments as we go. But there won't be 6 another stage.

7 The original source of the Stage 1, Stage 2, Stage 8 3 was, we sat together with the staff and agreed that there )

9 should be a sequenced submittal that would incrementally  !

10 cover all the systems that had to be covered. We just 11 numbered them 1, 2, 3.

12 MR. WARD: So it's a scope thing rather than a 13 detailed?

/'s V 14 MR. JAMES: No. We intended Stage 3 to be the 15 final submittal of a three-stage process that by the time it 16 wa.3 complete, would have submitted the entire ITAAC for 17 ABWR.

18 MR. WARD: All right.

19 MR. MICHELSON: I assume, in other words, any 20 remaining questions or problems ought to fit within the 21 content of the presently proposed ITAACs?

22 MR. JAMES: That's correct.

23 (Slide.)

24 MR. BOYCE: This slide shows where.some_of the 25 comments are being fed back to GE. Draft SERS on'the

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41 1 control room and the I&C DACs, as I alluded to earlier, will 2 be provided shortly, and you have seen the draft FSERs on 3 the DAC areas.uf piping and rad protection.

4 Comments on the Stage 2 submittal are being 5 provided as part of the August draft FSER, and comments on 6 the Stage 3 submittal are being provide via separate 7 correspondence. I'll make sure you guys have -- the ACRS-8 gets copies of those comments.

9 As was alluded to earlier, the-Graybeards are 10 reviewing or-have reviewed the ITAAC material independently.

11 There is also an effort headed up_by John Craig.that will be 12 doing a one hundred percent review of the rest.of the ITAAC.

13 as a followup. There is some-interaction with GE and NUMARC 14 scheduled for later this month.

15 [ Slide.)

16 MR. BOYCE: This slide talks about'some of:the 17 things that have been alluded to earlier. The staff has 18 identified various inconsistencies in the submittals between 19 the'SSAR, the design' description and'thelITAAC. -I'think the 20 Subcommittee came across several of those'in-their review 21 with GE also.

22 GE right now, we believe,fis resource constrained, 23 and this has resulted in delays-in submittals and' 24 resolutions of issues. We:should' note that GE must have 25 ' timely _submittals in' order to close the many,'many Lo  : ANN t RILEYI ASSOCIATES, Ltd. ;

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outstanding issues on ITAAC, and support the schedule for FDA.

3 (Slide.]

4 MR. BOYCE: This slide shows some of the types of 5 ITAAC that have been developed and the iterative development 6 procoso here. The bulk of the ITAAC are c- .ained in 7 Systems ITAAC for systems of the design. -

elected to 8 develop the ITAAC on a systems approach, as opposed to an 9 SRP based approach that the staff reviewed the SSAR to.

10 Their generic ITAAC for generic concerns across 11 systems -- and these are cross referenced to the systems 12 where appropriate in Section 3 of that Stage 3 submittal.

13 The staff is considering COL ITAAC for varicus procedural 14 requirements. An example is given here of training.

15 Interface ITAAC are specified in Section 4 of the 16 Stage 3 submittal and those are for site-specific elements 17 of the 6,asign. An example is given in Part 52 and here of 18 the ultimate heat sink.

19 As we've discussed previously, there are DAC for 20 selected areas of the design.

21 MR. WARD: Let me make a comment for the benefit 22 of the members who weren't at the Subcommittee meeting, that 23 the ITAAC which we reviewed yesterday and which we'll 24 partially review today, are only of the first type, the 25 systems ITAAC. I think we will perhaps have some question

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1 or want to consider whether it will be necessary for us to 2 do come sort of review of the other types, particularly the ,

3 generic ITAAC in order to ::omplete what the Commission has '

4 asked us to do. That's something to think about. i 5 MR. CARROLL: I might add to that that generic 6 ITAAC could be things like welding procedures, environmental 7 qualifications, and could you name some others, Tom?

8 MR. BOYCE: I think there are nine generic ITAAC 9 that are listed in the Tier I, Stage 3 submittal. One of ,

i 10 chem is welding,-one of them is equipment qualification 11 which contains three subelements, seismic qualification, ,

12 environmental qualification, and qualifications-for': urge 13 withstand capability and electromagnetic-' interference. '

14 Also, in' generic ITAAC are implemented some of the 15 DAC requirements for the I&C area, the safety system logic I

16 and control, instruments setpoint methoJology and software -

17 - I think it's.being renamed computer development process.

18 Those are some of the examples.

19 MR. LINDBLAD: Tom, where you have this.page on- f 20 design certifications,-and you' refer to licennee procedural 21 requiremer.ts such as training,-is=this training for design 22 or training for operations?

23 MR. BOYCE: It would bn training for operation.

24 As the human factors elements'that go into t' raining are 25 being incorporated into the control room-DAC,.where _

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44  ;

1 appropriate, and otherwise covered in the review of the SSAR 2 material that will be shown to you later in the draft FSER  ;

3 that will be forthcoming. [

4 MR. LINDBLAD: Thank you.

5 (Slide.)

6 MR. EOYCEt The next few slides discuss the .i 7 selected aspects of the ITAAC, and some issues that we are 8 still wrestling with. This discussion talks about tha 9 relationship of the design description'to the ITAAC.

10 In Section -2 of the: tier or.e submittal there are- '

11 two compononts for.any system. Onetis a description of that 12 system, and that is called the design-description, and'the 13 second half of that is the ITAAC, which actually will- l 14 confirm the elements in the design description,cand will be i

15 the basis for the staff's safety determination under:52.103.

36 It should be noted that the ITAAC will be used for 17 the fuel load decision, and. subsequent' facility modes to.the' ,

18 design. In other words, once they'are not that'is what they.

19 are going to be used for,Jthat 52.103. finding, and-the 20 design description wil1~be-what carries on,-and is what'is- I 21 going to be compared in that- 50.591like process- that' was 22 allucted to earlier.

23 This last bullet discusses afGE proposal that:

24 certain' systems.should.have design' descriptions with that 2 5 -' . ITAAC: based on theisafety significance. of the system. The O ANN RlLEY & ASSOCIATES, Ltd. . .

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45 1 ground work for that was laid in SECY 91-178, and a follow-2 up SRM, where the Commission gave us guidance that we should

, 3 take a graded approach to the level of details based on the  ;

4 safety significance of the system. We ars still evaluating 5 that.

6 In addition, the staff is still evaluating, as l:

"I indicated in this third bullet, whether all elements of-the s 8 design description are going to require a corresponding  ;

9 ITAAC. The subcommittee also found instances where the

, lo deciqu description would contain design eloments, but they '

11 would not necessarily have a corresponding.-ITAAC, and we are U right now trying to establidi a consistent methodology ,

13 throughout the ITAAC for trestruent of those sorts of issues.

14 [ Slide.)

15 MR. BoYCC: This is another topic for some 16 requirements that cannot_be met prior _to fuel load, but yet 17 we have traditionally _ required in the past.. An-' example is '

18 given of start up testingj and initial power testing. .You 19 can't do these prior to fuel load.

20 The situation.is very-analogous to what we did

'21 under Part 50 with an operating license. We' issued an .

22 operating license' prior to fuel load, but this_ testing-also 23 occurred after fuel load there,.and the way it was handled-24 was as a condition of the license, we are: proposing,:under l- 25~ Part 52, to handle it exactly the same.way, and-these ANN RILEY & ASSOCIATES, Ltd.

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4G I conditions of the license would be tied to the COL.

V) 1 2 (Slide.]

3 MR. BOYCE: This particular slide discusses the 4 treatment of what the ACRS and the Commission have terned 5 non-traditional items. They refer to insights from PRA, and 6 various sovere accident issues resolutions that have come 7 before you, such as SECY 90-016, and that lineage of SECY 8 paper type of issues.

9 Right now, these insights have been incorporated 10 into the SSAR in accordance with the Commission's direction.

11 An argument can be made that implicit confirmation of these 12 issues will occur since the ITAAC verify the design as 12 Jimted in the SSAR, and the tier one elements are just O)

\_ 14 extracts of the SSAR.

15 However, it is not easy to find those when you are 16 actually just looking at the tier one laaterial. As a 17 result, the staff has requested GE to develop a cross-18 reference of these SSAR issues to the ITAAC. GE has 19 provided an example of one of those cross-references in 20 their stage three submittal. It is in Appendix B to the 21 stage three submittal. We also provided it to you and the 22 Commission in SECY 92-214.

23 '211dt.)

24 MR. BOYCE: This last slide addresses something 25 that Dr. Murley and Mr. Russell addressed, and that was the

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1 requirements for FDA, whether or not we should approve ITAAC 4

2 as part of that process.

3 This concludes my portion of the brief.

4 MR. MICHELSON: Could I ask a general question?

5 It is not clear to me, and I am cure you or GE can 6 clarify how you are going to-handle-the generic questions of 7 analysis, and determinations-of pipe break considerations 8 outside of primary containment.

9 We discussed that to some extent yesterday for the  ;

10 main steam chase, for instance, as it-penetrates through the l

~

11 control--building.. But the same. question arises in every 12 single system that you are dealing with,.so it looks~like it 13 is:a generic one. 3 14 We didn't find it previous under piping design.

15 We were assured it was under some. kind of' building design, 16 and we looked at the control building design: yesterday, and .!

17 it wasn.'t in there, but where will I find this, eventually?

18 MR. BOYCE --Mr. Michelson,-I wish"we hid been able-19 to address that question here with-our plantisystems folks

-i 20 . yesterday.- I know it-came up:twice early-on,-.and then I-r 21 think we: just had a timing problem. Where the' question didn't

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i 22 come up when the right people were here.

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23 I would have to' defer:that question until I can 24 get our plant. systems' folks ~back, but I_was.only preparedL--

25 MR. MICHELSON:- Maybe-we'can just'get.a GE reply a h - ANN - RILEY & ASSOCIATES, Ltd.-

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() I to it, which is good enough?

f i 2 MR. JAMES: The building features that are needed t

3 to withstand the pipe ruptures are a part of the building 4 ITAACs, is the wall 1.5 meters, as was required by the 5 subcompartment Snalysis.

6 MR. MICHELSON: As pointed out yesterday, of 7 course, we didn't find requirements for, for instance, the ,

8 main steam lines, and feedwatar lines that pass through the 9 control building chase. The building has to be analf ted for 10 the pipe break, and pipe whip affects in that area.  :

11 MR. JAMES: That-does get back to the discussion 12 we had yesterday. The pipe break analyses have been done.

13 MR. MICHELSON: What ITAAC is going to assure that 14 it is done, or do we need to even check?

15 If it doesn't appear in any ITAACs, it means it is-16 not important enough to check?

[

17 MR. RUSSELL: Not correct. My understanding, and 18 we have some structural-audits that are_ going on, and_there 19 is re-analysis being done now of both the control building 20 and the' reactor building.

21 My understanding of the approach is to postulate l

22 breaks of the largest size. pipe in a particular_ area to 23 perform.the analysis now, considering._ venting area,:and 24 loads that would occur on.that structure,-and'to assure that 25 the structure has the capacity,.at least'asidesigned, to be C:) .

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49 (v) 1 able to handle those loads.

2 So my understanding !s that that analysis has been 3 done on a bounding break basis, e.nd has resulted in features 4 in the structures that, if implemented, would be 5 satisfactory to ensure that the structures could handle the 6 subcompartment loads.

7 I believe the methods being used are identically 8 the methods in the SRP for subcompartmerit anala is.

9 MR. MICHELSON: My question was simply how are we 10 going to verify this later?

11 MR. RUSSELL: What we are doing now in the SER 12 review is reviewing that analysis and reaching conclusions i 13 as to whether the analysis was satisfactory. If the l

[h 14 analysis is satisfactory, then it's sufficient to confirm

(_/

15 that the design as built conformed to that analysis, which 16 is what the purpose of the ITAAC is for each structure.

17 So the ITAAC for the structure would specify the

( 18 general layout in the structure, the size of the walls, et 19 cetera, which would give you the venting areas, the wall 20 thicknesses and the capacities.

21 MR. MICHELSoN: The problem with that approach is 22 when you did the analysis, you had to assume where your

! 23 supports were, for instance, and you don't know that today, 24 at least not in any of the information I've seen come 25 through in Tier 2.

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50 1 MR. RUSSELL: The approach -- again, what I'm 2 talking about is the subcompartment analysis for the breaks.

3 The issue of whether a pipe whip would be sufficient to 4 cause a structural failure needs to be addressed at this 5 point in time. The issue of whether a pipe whip would 6 damage other equipment in the room is being addressed by 7 physical separation to show that there is capability to 8 perform the necessary actions from another space.

9 MR. MICHELSON: I think you are missing my point.

10 I could determine the pressurization effects without knowing 11 where the supports are. I cannot determine the pipe whip 12 effects without knowing where the supports are.

13 Now, you make certain assumptions today in your 14 analysis. Twenty years from now, a fellow builds one of 15 these plants and the inspector simply has to go in and 16 ensure that the support that was assumed in the analysis is 17 really there in the plant now.

18 MR. RUSSELL: Let me suggest that we take this 19 issue up at the subcommittee meeting on the 19th, and we'll 20 be prepared to address it with the right staff here.

21 MR. MICHELSON: Okay. It'll come up again then.

22 Okay.

23 MR. WARD: What happened yesterday? I thought we 24 were going to get this answered with the staff yesterday.

25 MR. MICHELSON: Well, we ran out of time, o

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51 1 MR. BOYCE: It came up twice --

2 MR. MICHELSON: Yes.

3 MR. BOYCE: -- but it came up in the context of 4 SLC and RHR, and then when our plant systems folks arrived, 5 the question didn't come up during that time interval they 6 were here.

7 MR. MICHELSON: It came up ic. che context of the 8 main steam and feedwater lines through the control building 9 when I asked a question about the pipe whip analysis that 10 would be done now to determine those thicknesses and what 11 verification would be performed later to verify that your 12 assumptions are still valid. I think that verification is 13 needed, and I don't find it required in any of the ITAACs so 14 far.

15 MR. WARD: Okay.

i 16 MR. MICHELSON: That's -- and i just wondered 17 where it was or where it would be, or is it needed.

l l 18 MR. RUSSELL: To the best of my knowledge, we have 19 not addressed that issue yet.

l 20 MR. WARD: We'll hear about it on the 19th.

21 MR. LINDBLAD: Tom, in your treatment of non-22 traditional items, you asked GE to incorporate insights from 23 PRA into their standard safety analysis report. Are.they 24 expected to comment on areas that have small significance'to 25 safety as reflected in PRA and that you would confirm that, v

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52

[\m/) 1 that they are weak influential items?

)

2 MR. BOYCE: In your question, you asked me whether f 3 PRA insights are going to be incorporated in the SSAR. GE 4 right now, I believe, is providing us a list of where design 5 1.7 sights have been incorporated into the SSAR. I think that 6 was part of a June 30th cubmittal.

7 In terms of the ITAAC, ce are looking primarily 8 for some of the more important insights and things like that 9 that would be appropriate for treatment in a rulemaking 10 environment in Tier 1, and so those tend to be the very high 11 level type of insights that we would get.

12 An example was given in SECY 92-214 where the 13 vacuum breakers were considered a very essential part of the b)

(_ 14 design basis analysis, and that their function was critical 15 to that analysis. In other words, they're a significant 16 contributor to core damage frequency if they malfunctioned.

17 So based on that sort of insight, we would be 18 going back and taking a close look at what we could do to 19 write an ITAAC to confirm that that particular component was 20 built correctly and would function correctly in its intended 21 environment. So that's the sort of insight that I expect to 22 see in ITAAC.

23 MR. KERR: What does function correctly mean?

24 MR. BOYCE: Well, I think that case, it just means 25 it would open under a differential pressure, although I'm

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l 53 1 not sure of that.

2 MR. KERR: With any particular reliability? I 3 mean, once out of !00 --

4 MR. BOYC11: Well, the question of reliability, the 5 PRA insights -- well, PRA handt over reliability numbers to 6 a program like design reliability assurance --

7 MR. KERR: Well, I'm not interested in what PRA 8 does; I'm interested in what the reviewer is going to 9 require or whoever is inspecting is going to require that 10 the valve do. Just t'.at it be capable of_ opening, or that 11 it open 50 times out of 100 on demand?

12 MR. BOYCE: Up to this point, we have not tried to 13 verify relia'Jility numbers in ITAAC. They tend to be --

14 MR. KERR: What is it that you are doing, then, to 15 try to verify it?

16 MR. RUSSELL: This is Bill Russell from the staff.

17 There are two distinct and different places th'.t insights 18 from PRA are provided. One is as it relates _to physical 19 design, which would be the tracking into ITAAC.

20 There are'others which relate to the reliability I

21 of the component to perform.and how important is that 22 component. That is a part of.the reliability assurance-23 program which ensures that it's-not only capable of _

24 performing, but-is_ reliable'in; performing that, and we have-25 no.t yet developed completely either.the licensing controls o ANN .RILEY- & ASSOCIATES, Ltd.

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54 NJ) or what type of-ongoing demonstration would be necessary.

1 2 We bolleve it's going to be a combination of

3 surveillance testing, maintenance kinds of activities. We 4 do not expect that we would have reliability testing 5 performed for components.

6 MR. KERR: It's not a question so much of 7 necessarily reliability, but it seems to me that unless you

8 assume that any component that is appropriately designed can 9 achieve the reliability

• hat is "equired, that the required 10 liability might have some influence on how much you pay for 11 a particular component.

12 Take a valve, for example. You can buy cheap 13 valves that have low reliability; you can pay much more and 13 4

\m / 14 get higher reliability. But if ycau buy a very cheap valve, 15 I'm not sure that a reliability assurance program is going 16 to achieve the re11 ability that one hopes e- needs to 17 achieve.

l 18 So it seems to me that in the spirit of this 19 thing, that one needs to make some decisions in_ choosing the 20 components.

21 NR. RUSSELL: There is no. disagreement with that 22 at all. In fact the details of.the design of.the valve have 23 been extensively discussed. In fact at one point in time it 24 was proposed that there be an air operator on the valves 25 such that the valve could be remotely cycled and tested and V

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55 1 put limit switches on it.

)

2 The Staff felt that that-introduced a larger 3 potential for error in that-the valve could be left in the )

4 wrong position or lock open and it was suggested that that 5 was not appropriate, that a valve like a check valve with 6 the particular design that has been proposed is more 7 appropriate.

8 I~ don't think we are talking in this case about'a-9 valve that you are going to be able to buy off the shelf.

10 This has got some rather particular design features

~

11 associated with it but the question equally applies to or

~

12 more so to passive plants where we are: relying on very

~

13 highly reliable but fewer components.to perform functions, 14 particularly_ check valves under a low differential'. pressure.

15 This is essentially a check valve that we expect.

16 to operate under relatively low differential 7tessure and be 17 able to re-close and there have been things done in the 18 design to minimize cycling.

19 MR. KERR: You are usingl terms, Bill, that=say vou 20 expect it to be able'to open and close.

l l_ 21 MR. RUSSELL: Yes.

-22 MR. KERR: I don't'know what that means-in terms 23 of a PRA. I raised the issue; simply!following up. You have 24 -_some_ insights'from aLPRA and presumably?the-insightigives 25 you some idea of the reliability thatithe component needs to

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4 56 1 have if the risk is to be within some appropriate region.

2 MR. RUSSELL: The insight that we have from the 3 PRA is a little bit different in this case.

4 We have actually done sensitivity studies on 5 suppression-pool bypass and how much bypass-can be 6 accommodated without having an early containment failure, 7 and the bypass occurs-by having a partially open or failed i .

8 open relief valve, so the insight is that_-this is a very 9 important valve from the standpoint of containment 10 performance but we have not specified at this point a  :[

~

11 reliability number for it. We-have not addressed whether 12 prototype testing is needed on some' basis. Those are. issues 13 yet to be addressed in our review.

(~

V) 14 Right now we have only addressed it from the' 15 standpoint._of its-capability of function and how important 16 is that function, both from:the standpoint of providing a.-

17 relief path so you don't fail-the diaphragm floor between

~

18 the drywall and the wetwell-and secondly being able to close V .

l 19 so that you.get the suppression functionfand=no-bypass or 20 little' bypass.

21 'We have not addressed the reliability: aspects'at 22 this point in-time.

23 _MR.=KERR: _ It seems to me that'a-' fairly-important'

24 insight of the PRAs is being neglectedfif that is the. case 25 because --

yn O .

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l 57-1 MR. RTISSELL: Not neglected. We just have not come 2 to grips with~it yet..

3 MR. KERR:. Then I should revise it to say that' 4 "has not yet been come to grips with."

5 MR. RUSSELL: Yes. In other words,-if you have 6 got a valve _-that-has a very high reliability,--low. failure _

l 7 rate, how do you' demonstrate that?

8 MR, KERR: I mean I can't,-I don't have the anstar

)

9 to the problem. _I-am simply saying that if youoare using.

10 insights from PRAs, and some of the insights-that are very-11 important are quantitative ones that you either address them 12 or you decide you can't address them and maybe-the. answer is

( 13 that independently what-the PRA tells _you, youEcan't do 14 anything about valves.

15 I don't know, but if you need to do something, it l 16 seems to me that one of-the things you nsed-toido is-17 quantitate it.

l 18 MR. RUSSELL: At.this' point in time =most of the 19 insights are relatingLto what I'will characterize'as 20 deterministic type requirements _-for'the component or-an 21 approach. -l 22 ~We-discussed this as it related tolinsights from

-23 the-control. room design, oncimportant' tasks to be--. performed' 24= 'and how you ensure'through afdeterministic process ~that you t-25' do not-introduce the potential-or minimize the potential =for LO: - ANN RILEY & ASSOCIATES, Ltd.:

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1 human errors in carrying out that activity.

)

5' 2 It's a similar situat' ion here. -We have not at

3 this point proposed reliability testing to demonstrate '

i

4 component reliability. We are using in essence generic data ,

5 from experience and extrapolating that. .When you get to a 1

l 6 unique design or design that is very important, we may i

7 require diversity or a different capability.

8; We have that situation with software. We cannot 9 -quantify the reliability of software and'we have taken'a i 10 position that there shall be diversity because we cannot 11 assure the kinds-of reliability that we think would'be 12 needed based upon PRA insight so different approaches are 13 being taken on a case by case basis.

O 14

\_/ What we are trying to do isLfind out"from a system 15 performance standpoint what is important and using the PRA-

16. to provide that. type of insight.

17 What we do about it varies from a couponent'or an-18 insight basis and what we are trying to dofis make-sure that 19 those important-features get- factored into the ITAAC where--

20 appropriate, where it is a design-related feature which can 21 be-verified with an ITAAC.

22 MR. BOYCE:- I would just like toiadd that we have 23 asked GE.to provide the roadmap. .We're wrestling in. l '

-- 1 24 -. parallel.with this, but1without seeing.anoinitial proposal, 25 it is difficult for usito. argue.in the' abstract as to' 1()!

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59 1 exactly what we would_like. Possibly GE could comment as to 2' what their intentions are.

3 MR. JAMES:- I.cannot' constrain myself, Mr.

4 Chairman.

5 We submitted that and I think_you know it, that-- ,

6 that whole package has been submitted to you that provides-7 the roadmap on the severe. accidents.

8 There is a five page table in there that shows 9 what the important characteristics of the plant are~in: terms 10 of minimizing plant risk and it identifies where the-'ITAAC, 11 in which ITAAC those various plant characteristics are 12 confirmed.

. 13 MR.-BOYCE: -I am sorry,-Tony, you're right. You' 14 have submitted that. Perhaps you could-offer some insight 15 into the reliability question that has been brought up.

16 Maybe that's a better way.tc put it.

17 MR. KERR: I am more' interested in what the Staff 18 plans to do than I am in what GE plans to do because--I think 19 to some extent what GE plans _to.do will depend'on what7is-20- required, 21 l MR. CATTON: -Does GE have'to. place a' reliability i

- 22 on that, and why aren't they required-to_ meet it?

-23: MR. CHAMBERS: -We are, and.that's what the whole I

24. ' reliability assurance program is.- -

25 MR. CATTON:- So what's the problem?

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60 1 MR. CHAMBERS: There is no problem. Every 2 component has a reliability assumption'as it's-modelled in-3 the PRA. We in:the Staff, if we don't agree.with what we 4 have initially proposed for reliability, which is usually

~ '

5 based on past history,_we negotiate, come_up with-an 6 acceptable reliability and that is either. based _on past 7 history or some expectation-for future demonstrations and 8 that's what it ends up being.

9 MR. CATTON: So what was the reliability for the 10 vacuum breakers?  !

11 MR. CHAMBERS: I am not sure what it was for the 12 vacuum breakers. We can get that-for_you. It's all~-- all .i 13 that stuff's in the PRA. It's all=in Chapter 19.

14 MR. CATTON: Is the number in the ITAAC?-

15 MR. RUSSELL: No, it is not. That l's the-issue 16 and, as I said, we are requiring a' reliability assurance 17 program to ensure that to the' extent we can that-the-18 reliability.that is achieved during operations'is' consistent.

~

l 19 with the assumptions in the PRA, or if not,_that there are-20 changes made to improve 7the reliability, but that-is an-21 ongoing activity-with-time. '1 l

22 It is not something'that;is conducted prior to 23 fuel load with an'ITAAC. There is aLdifferentiation-24 between.the two, butiPRA insights'go either into:the 25 reliability assurance program where-you get reliability _

l i

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.61 1 kinds of numbers and insights which can be based upon 2 operating ~ experience with time after you' select._ vendors, 3 components,-.have some experience, and what we are saying..is-4 we are using_ historical experience to make judgments-as to ,

5 whether those are reasonable.for initial startup, initial -!

6 fuel load. That's in the PRA review.

7 MR. KERR: Is it expected that a-supplier who ,

8 supplies components will be asked to guarantee that-a-9- particular valve will'have a demonstrable reliability?

10 MR. JAMES: 'Yes.

11 MR. KERR: Good. -That will-not,,however,;be part 12 of that ITAAC.

13- MR. JAMES: In general, reliability-is'not part of 14 1TAAC. The important reliability' aspects, the ongoing,.

15 forward-year reliability aspects are beyond the scope of

, 16 ITAAC.

17 MR. KERR: .Thank you.-

18- MR.~ MICHELSON:- Are-you saying-'you're going to do 19 'a reliability oriented procurement?. .You'reigoing'to specify 20 a reliability-number, for-instance,-for a valve,.and procure' l 21 accordingly?

22 Mk. CHAMBERS: .Where it's important,. clearly that: ,

23 will be something that's in a procurement spec.

24 MR. MICHELSON: Okay, good.

25 MR. JAMES: It's-just part;of the bigger picture o ANN RILEY & ASSOCIATES,- Ltd.-

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l a-62 1 of commitments that are made in the SAR that-have to be 2 translated into' procurement tiocuments.

3 MR. MICHELSON:. Is that committment'in-the SAR for l- 4 motor-operated valves; that-they will be procured to the PRA i _.

} 5 reliability numbers? Not to my knowledge, but my knowledge i

6 is extremely limited.

~

i 7 MR. WARD: I think that's what'he said, he said-

+

8 they would.

{ 9 MR. MICHELSON: Yes, I'm sure he did say that they-10 would.

l

. 11 MR. WARD: Now, where are they going to say 12 they're going to do it?

l 13 MR. RUSSELL: Let me clarify it. To the extent 14 there are' insights as to what should be in a-reliability .

h 15 assurance program, we're addressing those now, as it relates 16 to design. The COL applicant is'the.one that's-' going to i: 17 have to submit-the-plans for how they're going to implement,

~

[ 18- both in the design stage,-and in-the operations' stage, the

! 19 reliability assurance program..

) 20 - The. text on the reliability assurance program

! 21 covers both. There is a limited-. aspect that-is being

-22 covered now, and'it's essentially a reasonableness test.--Do 23 you expect that you can achieve that. reliability or:not,..and 24 what is;the staff judgment on it, and-that's aLpart of the PRA review.

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r 1 We are not putting reliab'ility numbers in ITAAC.

2 They are going into a reliability assurance program which 3 will be supplemented by the COL applicant, which will 4 describe how they're going to handle such things.as 5 procurement, et cetera.

6 MR. MICHELSON: Where's the committment to procure 7 to the PRA reliability numbers? Where will that be found?

8 MR. RUSSELL: It will be discussed in the COL 9 applicant's submittal.

10 MR. MICHELSON: That's too late. In the 11 certificate, where will that be?

12 MR. RUSSELL: It will not be.

13 MR. MICHELSON: There is_no requirement.-

14 MR. RUSSELL: There is no. requirement that gets

-15 down to individual components. .It's vendor-specific.

16 MR. MICHELSON: You' won't implement that 17 requirement until the guy comes up for: COL,_but'isn't that -

18 - hasn't finality already caught-you?

19 MR. RUSSELL: No.

20 MR. MICHELSON:- Can you implement 1such a 21 requirement-at --

22 MR. RUSSELL:- We will have insights-as to what's 23 important and we've asked for a-particular_ appendix to 24- identify-what the important assumptions were as to what j 25 needs to be achieved. .That's;in_the design-certification;

'0 l_

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- 1 that's in Tier II material, and I believe it's' Appendix 19-2 K of the SSAR'submittals.

3 MR. MICHELSON: That becomes'a committment to' meet 4 then those require 9ents?

1 5 MR. RUSSELL: In Tier II, there are processes that 6 you could make changes to it. We're also-talking about

7 having them update the PRA, based upon procurement decisions 8 that they make at the time of the COL.

9 MR. MICHELSON: Now, will that be found in the 10 reliability assurance program section of the SSAR?L l' MR. RUSSELL: The description of the reliability 12 assurance program describes that process,=yes.

i 13 MR. MICHELSON: I know '.t describes the process, i 14 but I' don't. recall it. explicitly _says,-now that the

15 equipment purchased-for the plant will-have t7 meet the 16 reliability. numbers expressed in the PRA, which is-what 17 we're asking to see.

18 MR. RUSSELL: In' fact, if youllook at the grandson-19 of 90-016,~we'L3 identified al policy issue that is

20- essentially to'have the=PRA updated byLthe COL' applicant, 21 based upon decisions.that they are making on equipment

22 procurement to address-this11ssue.

23 MR. MICHELSON:' But that doesn't answer my.

, 24 question, though.

25 MR.-RUSSELL: -If they procure equipment-that can't -

i RO

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-65 1 achieve the reliability, that will haveLto be factored back

-2 into the PRA model and show that it's still-acceptable.

3 MR. MICHELSON:- Butitnat-committment must be in 4 the SSAR, or you'can't enforce it later.

5 MR. RUSSELL: - - But it's not - one that we would -

6 necessarily impose on GE. At this point'in; time,-we're 7 trying to make judgments as.to,whether they have made-8 reasonable assumptions-about what can beLachieved'or not, 9- and their reasonable basis for choosing-that.

10 The. COL applicant-is the oneithat's going to have-11 to implementLthat_ process..

12 MR. MICHELSON:' Yes~,;but thei-implementation isn't 13 the question;,the question is where does.the requirement 14 appear? Implementation ~.is later,.obviously.

i 15 MR. RUSSELL: _.It's in the.' general description of -!

16 the reliability assurance program.

17 MR. MICHELSON:- Okay,,gocd.

18 MR. CARROLL: . I guess one pointithat is relsted to 19 this is that last< month,'we looked at the' license 1 renewal 20 program or. license renewal rule;in the Reg Guide, Land liti 21 became. apparent that-it was-!very similar,-but-not identical 22 to'the-maintenance, rule. I guess:our' letter that=we'll 23 finally _get-out'.this' month is going.to: comment-onithat.; lAre 24 we introducing now, a thirdikind'of thing'thatra licensee .

25' will have-to_ meet, namely, the. reliability: assurance program-ANN RlLEY:& ASSOCIATES, Ltd.-

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that may not be -- that may be similar.but not identical to the other two?

3 MR. RUSSELL: The answ r to that.is yes, and we 4 believe that the maintenance program will constitute a 5 significant portion of meeting the overall reliu_ility 6 assurance program.- That is unique, because we are also 7 requiring a PRA as a part of the design certification and 8 licensing basis of the plant, and we've required a 9 reliability assurance program for. future plants..

10 It's basically as building the PRA with time, to 11 ensure _that the-plant matches what was licensed.

12 MR. CARROLL:,-Okay,.all I'm suggesting is that 13 there-ought to be better coordination between the-14 reliability assurance program and the other two, than there 15 were between the first two, because-it looks to us like a i f

16 licensee may have to do a lot of work to meet two different '

17 rules where-if somebody had thought about it.up_ front,-there 18 wouldn't have been those problems.-

19 MR. RUSSELL: That-is a subject of a separate:

4

20. meeting.

21 MR. CARROLL: . I'understard._

22 MR. WARD: All right, let's---- is that-all you-23- have?

24 -MR. BOYCE: That's all..

25 MR. WARD: Before we go to GE,_let's take our l

\

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[V ) 1 morning break and return at 10:20.

2 [Brief recess.]

3 MR. WARD: Okay.- We now come to that part of the 4 program where we hear from GE. I'm not sure, Mr. James, 5 that we're going to have time to go over the SLCS ITAAC in 6 any detail, but we want to hear your general comments, 7 including -- you probably want to -- might want to make some 8 comments on what you heard narlier today, and we'd like to 9 hear those.

10 MR. JAMES: Well, I had planned to start my-11 presentation, Mr. Chairman, by saying that I gave you 12 yesterday the GE defensive response to some of the staff 13 comments and was not planning to repeat those today.

/3 (m,/ 14 MR. WARD: Yes. Unless you have any new ones. I 15 remember them from yesterday.

16 MR. JAMES: No. We have the same defensive 17 responses.

i 18 MR. WARD: Okay.

! 19 MR. JAMES: I think, though, by way of l

20 elaboration, the only point I would like to make is there's 21 been a lot of discussion of inconsistencies in the ITAAC 22 treatment. I agree there are some, but my-view is.the 23 majority of them are being driven by a couple of causes.

24 One is that the SAR is still a moving target.

25 As we have heard in the last day and a half, there t

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1 ara quite a number of areas of the SAR that are still 2 undergoing revision, and, frankly,'I view ITAACs as the tail 3 on the dog, and the dog is still wiggling, so we're having 4 trouble keeping the tail straight. That's part of the 5 problem we've got here.

6 The other part of the_ problem that I think you've 7 even seen here is I think this whole issue of Tier 1. One 8 of our big challenges is education.

9 Every time I -- I've been going to meetings like 10 this for about a year, and every time you get into meetings, 11 any meeting like this with any engineers, it's clear that 12 there is not a uniform understanding of what Part 52 is all 13 about and what's intended in ITAACs and what Tier 1 is all

(

V 14 about.

15 So ve have a tremendous education job on our hands 16 with not only GE staff, NRC staff, but all others that come 17 into contact with this process,-and that's part of the 18 challenge we face, I believe.

l 19 Having given you those responses, let me --

20 MR. KERR: Well, I hope you aren't going to retaove 21 all the fun that this committee has in picking up 22 inconsistencies. I mean,.think of how dull it would be if 23 we couldn't find any.

l 24 [ Laughter.]

25 MR. WILKINS: They've got a whole' group at GE f%

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69 i i

g

() 1 2

which creates these inconsistencies, some for the ACRS and some for the NRR staff, you know.

. 3 [ Laughter.) ,

4 MR. JAMES: Oh, yes. We'll keep you guys busy.

5 No problem.

6 (Slide.)What-I have done at your suggestion, Mr.

l 7 Chairman, what I plan = to do was -- and introducing myself ,

f 8 for those members that weren't here yesterday, my name is 9 Tony James from GE. I had planned to briefly go through at  ;

4 10 your suggestion the material'that I covered yesterday'at the 4 11 subcommittee meeting, with a slight modification.

[ 12 _[ Slide.)

13 MR. JAMES: This was the agenda'from yesterday.- I l ,9

[ 's_f i

14 would plan to go through the first couple of items, which l 15 gives you an overview of the assumptions that have been i 16 driving GE's preparation of Tier 1 material, and then

. 17 hopefully, if we have time, at least flash up the standby-p 18- liquid control system version of Tier 1Ujust tofgive those l

19 members'of the committee that werent here yesterday an- -

20 overview of what it looks like, and then,< depending on time 21 and interest, - address - those ' or leave them - for ai future date.-

i 22 So'if that meets with your approvalo--

23 MR. WARD:' Tony,1I'd likeLto save -- we have until 24 eleven o' clock for this topic,1b ut: I'd like to save'at-least 25 five minutes, perhaps-a couple =more than that,'for-committee I ( ,[ -

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( 1 discussion at the end. So if you could plan on finishing at 2 ten minutes of.

3 MR. JAMES: Okay. I will do my best.

4 (Slide.]

5 MR. JAMES: The first chart I think probably 6

doesn't need much discussion because it's obvious from the 7 discussions here that I think most people understand the 8 basic issues and structure of the design certification.

9 What's been driving GE is this column here.

10 That's the-approach we have been using, and this is the 11 basis for the approach. It's clearly a two-tiered pro.~ess 12 that we all understand.

13 We view Tier 1 as reserved for the top level O)

(m, 14 information, a subset of the stuff that's in the SAR. We 15 believe it's justified by these references.

16 Clearly ITAAC, including the various versions of 17 ITAAC that I'll describe in a minute -- the design 18 acceptance criteria, the generics and the interface ITAAC, 19 as well as the regular ITAAC -- are in Tier 1, and that's 20 exclusively called for by Part 52, 21

- These two here form a sort of matched set that I 22 l really are very important, and misunderstandings en these l 23 are the source of many difficulties in the discussions that 24 we have on ITAAC.

25 First off, our view is that ITAAC verified

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() 1 2

conformance with the Tier 1 design. We've got'a top level Tier 1 design, and.we'think ITAAC are aimed at confirming" 3 that design, and Tun think this is in effect called for by-4 Part 32 with words like "will-operate in conformance with 5 the design certification."

6 But_a very important-adjunct or corollary to that i

1 is that the massive information that is_not verified <by.

C ITAAC is verified by.the in place existing =Part 50 QA. i 9 processes._ Part 52 invokes-Part 50, including Appendix B i

10 and all of the QA and verification that; flows from Appendix 11 B. This is a very:important assumption that's driving our 12 split between Tier 1 and Tier 2.

l l 13 We say in effect all of today's; processes, all of

/~}

(_/ 14 them, exist, and ITAAC are in addition to today's processes 15 and are aimed at'the top level part of the design.

16 That split between wh'at is coveredEunder ITAAC_in 17 Part 52, and what should be. covered-underLthe existing 18 verification and NRC inspection' procedures under.Part 50 is 19 a source offa lot of-discussions.-

l 20 [ Slide.]

21 -MR. JAMES: -What I have tried to do here is 22 summarize the elements that are included in: tier one because o 23 what we are really trying to dothere is put in place the.

, 24 tier one material for ABWR. We-keep on using shorthand,_and l'

l- 25 I.am. calling it:ITAAC,-but.really what we.are doing is l

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-1 trying to find, in the stage three material that was 2 submitted, the tier one material for the ABWR.

i 3 These are our view of what the various elements 4 are in tier one, if I could go through them briefly.

5 The design descriptions _that we have talked about, 6 they are the summary of the design. and that will become the 7 certified design in the rule. Tier one includes the 8 inspections, tests, analyses and acceptance criteria that we 9 have been talking about., They are aimed at verifying that 10 the specific. features of the as-built facility comply _ with 11 the certified design.

12 The DAC, which was _also a subject of the 13 discussions today, but, in our view, DAC'aro sort of on=an 14 ITAAC that includes some of the design process.- ~ When the 15 design details were not'available,.the DAC are the tier _one, l

16 as are both the interface ITAAC, and the gent.ric-ITAAC. U 17 The interface ITAAC are those that are t

18 specifically called for by Part 52 for_ portions of'the plant' j 19 that are outside of the scope of the Gesign for.which-.we are 20 seeking certification.

21

~

The-example we always use=is-the ultimate heat'

! 22 sink. Part252 calls for interface-ITAAC that will-verify 23 that site specific' features comply with the-_ requirements of 24 'the! certified design.

25 'In addition,.'of_courre, Part 52 calls for a' O

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73 1 scoping. preliminary design for these features that were not 2 -held to, but at least must be described to support plant'PRA 3 studies.

4 Then the generic ITAAC, again, which we haven't 5 been discussion today, but'those.are otar ITAAC that verify 6 generic aspects.of the as-built facility.

i 7 Fcr naspie, tne e es that we have been discussing 1 8 are welding, environentalt qualification, instrument set-9 points, generic subjects'like that.

10 So we see all those elements as being en important 11 part of tier one.

12 (Slide.]

l 13 MR. JAMES: The next chart tries to summarize the 14 approach we have been using, the approach we used to develop ,

15 'the material that is-on the table now'.

'16 Also,-very early=in?the cycle, we. decided-or a big 17 point of discussion was, what would-:be the basic structure-

't 18 of the material, and there;were:really two options. .One was 19 the SAR structure, and the other;was the system-by-system--

l 20 structure that GE and other vendors-use for! design and'other 21- activities..

22 .We reached consensus very early'on that.a-system-

'23 by-system approach made.a lot of, sense.-,The trouble with l' 24' .the SAR approach is that it-tends:to be fractured,-different-l~

l 25 subjects are--discussed'in:different chapters for a given.

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-74 1 system.

2 So we have proceeded on'a system-by-system 3 approach, and I don't think there is any controversy on that 4 point at all.

5 Within that structure,-we are using a graded 6 approach to systems that reflect safety significance, an 7 important core cooling-system would got fairly detailed-8 treatment. A non-essential plant support system might get 9 very light treatment,-light to almost verging on zero at 10 times.

11 Within those assumptions, these are the steps that-12 we go through to prepare the material. The first thing we:

13 do is say, okay, the design description comes first. Before 14 we.can decide what ITAACs are required,'we have to decide =

15 what-features of the design are going to be included in tier 16 one.

17 We have been preparing those._with the guidance 18 that tier.one really is reserved for principal design bases, 19 and design. features..

20 So, havingLdone that, we then say, okay, now.we

'21 can move.to step two,.which'is prepare an ITAAC. table for'

22. each of the systems. :It derives from-step one, from the 23 -design' descriptions but, as'Mr. Boycc pointed out, itfisn't;  !

l 24 necessarily one-for one. . We, haven't had a. rigorous.~one.for 25 one- of every statement here .gets an- ITAAC here. . There.isi L

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4 75 3

1 another judgment process in here that_ says we 'will pick the l 2 most.-important features to be! subject-to.ITAAC.

i':

3 Having done that,'then we obviously move into 4 preparation of other tier.one entries ~as'needed, the 3

! 5 generics, and the other entries, j - .

j 6 I guess'I will.say it again because it is~so

~

7 important to what we have-been-doing, the whole thing is

{

8 driven by..the tiered approach:to-certification,_and an-9_ underlying sort of bedrock assumption for everything we have i

10 been doing is-that the existing Part 5'O verification ,

j 11 processes are still applicable,_and play a *x role-in-Part-  :

i

12- 52.

I 13 We view ITAACs-as in addition to, and not-14 supp.lanting any'of the stuff thatlis_done-today,: and that 15 is, I will say it again,-an important assumption that we l

16 have been making. That isn'tividely agreed to yet. "

17- As part of.the iterations'we l have been having with

18 the-staff, there is some reluctance to rely:on Part 50 to.do _

[ 19- verification.. There is sometimes_an attitude.of,fif.it has-20 to be' verified,-it has.to be an ITAAC,sright?

21 Our. answer is,_no,.only the top level stuff'is 22 covered in ITAAC, everything else=gets-verifled by_ existing 23 processes. '

24 [ Slide.]. -

{ 25 MR. JAMES: The next1two charts, Iftried to-ifNV 4

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l l 76 l-1 summarize from the first chart.for the design descriptions 2 what are the characteristics that:we have L'ied to pick out, 3 and used to guide us when preparing the design descriptions.

4 The top' level', principal design bases and design ,

5 features, it clearly is drawn from the SAR descriptions.

6 Rule No. 1, there is nothing~1n the tier one design 7 description that hasn't been drawn-from the SAR description.

8 We introduced no new tectnical subjects. _

1 9 I have covered that one, it is system-based, based 10 on the level of detail. It contains only :information in the ,

11 SAR, does not address any plant operating conditions. By ,

12 the wey Part 52 is written, ITAAC must be completed by_the 13 time of fuel load. So nothing beyond-fuel load could be ,

14 covered by ITAAC.

15 We had included-numerical information in the 16 design descriptions to the extent necessary,to-~ identify.the 17 principal design bases and features. This topic we 18 discussed this morning,- tier one is-self-contained, and #

19 avoids any' direct-references.-to tier two. documents for.the-20 very reason-that we talked about this': morning.. It elevates 21 those tier two documents'to' tier-~one' status.

22 MR. CARROLL: Why don't you say'a little more--

23 about-that, what your-lawyers are telling'you? -

24 MR. JAMES:- Our' lawyers are saying in'effect iffin-

~25 Tier 1-you make-a reference to another document, then by i

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/ ) 1 reference, that entire document is elevated to Tier 1

'x_/

2 status. That's the input we're getting from our legal 3 people. So we seriously avoided, I have to admit with a few 4 slips of the net, but we're avoiding any references to Tier 5 2 documents.

6 I understand the staff legal people concur with 7 that assessment, as I understand what Mr. Russell said this 8 mo '.ing.

9 MR. SHEWMON: One question on that. Then Tier 2 10 would define the particular version of the ASME code that 11 was used, because otherwise, you know, you sure aren't going 12 to go through a GE rulemaking for the ASME boiler and 13 pressure vessel code to change their code.- That's under

~s

(_) 14 their control, not yours or the NRC'E.

15 What bothers me some is to say it's now part of 16 Tier 1, but somebody else controls it, not you, unless you

( 17 want to say it's a particular version.

18 MR. JAMES: I guess I'm not quite sure what the l 19 question is.

20 MR. SHEWMON:- Well, I'm bothered by your saying 21 that everything in Tier 2 referenced in Tier 1 that is in

! 22 Tier 2 becomes part of Tier 1. I guess that's okay if you l l i 23 referenced a particular version of the code only because 24 that's invariant with time.

25 MR. CARROLL: -I think what Bill said this morning

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1 is that Tier 1 might say the pressure vessel is going to be-2 designed to the ASME code. Tier 2 --

3 MR. SHEWMON: Pressure vessel is a poor one.

4 MR. CARROLL: Okay.

5 MR. SHEWMON: Take piping.

6 MR. CARROLL: All right. Piping is going to be 7 designed to the ASME code. That's all it would say.

8 MR.-JAMES: In Tier 1.-

9 MR. CARROLL: ' In Tier 1.

10 MR.. JAMES: Yes.

11 MR. CARROLL: And in Tier 2,- it would:. simply.say 12 what version,-what addendum, et cetera, et cetera, but thah 13 would allow you in the future-to-make~ changes if-that were 14- desirable..

15 MR. WARD: Yes.

-16 MR. SHEWMON: But you are referencing what,_then, 17 in Tier 1?

18 MR. JAMES: -It would just say ASME~ code, piping 19 code. It might even:say Section 3. But-it would'be-a very-20 general reference.

21 MR. MICHELSON:- Right now, it even says class in-22 the-case of some of those-flow diagrams.- Whether you do-

, 23. that, I-don't-know,Lin:the future, but?right now,- you'do 24 indicate.classEas well.

25 MR. JAMES:- Yes. .I-think we intend to retain O' c ANN RlLEY & ASSOCIATES, Ltd.-

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( 1 that.

2 MR. SHEWMoN: I guess it your lawyers are happy, 3 we can live with that.

4 MR. JAMES: Yes. But it's worth noting that the 5 lawyers -- the utilities -- I'll get into this in detail-in 6 a minute, but there is a counter-view that putting just a 7 reference to the ASME code in Tier 1 when it comes to 8 acceptance criteria makes it a very mushy acceptance 9 criteria that is in vulnerable to all sorts of ques,tioning 10 and intervention at the time of ITAAC sign-off. But I'll 11 get into that in a minute.

12 MR. MICHELSoN: But if everybody. understands the

~

13 process, it's clear that you inspect to the Tier.2 level f%

h 14 requirement and can inspect to anything in Tier 2. Tier 1 15 is just a higher level -- it's supposed to be mushy and 16 general. I think you purposely put mushiness into it.

17 MR. JAMES: Let me get to my chart that gives you 18 the utility's view of the world. There is a counter-view.

19 But vou are right, and.it's a very important point 20 that I have to stress at every meeting, that every 21 commitment that we make in the'SSAR is going to get 22 verified. The only debate is is it verified by Tier 1 ITAAC 23 _

process or is it verified by existing procedures?'

24 Also, under existing procedures, the NRC staff has 25 all of the remedies and actions, corrective action processes p

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80-j h 1- that they've got today. So that's an important point that U

2 I'd-like to stress. -It's not like if it doesn't get covered-3 in ITAAC, it doesn't get-verified. That.is absolutely not 4 the case.

5 The last point here -- remember-now, I'm-still 6 talking about the d>cign descriptions only; I haven't gotten 7 to ITAACs yes. The design descriptions,'they may include

-8 simplified P& ids, one-line diagrams, et cetera, et cetera, 9 and --

10 MR. MICHELSON: .One small comment that I noticed I

~

11 . failed to make yesterday but I think you should take'into 12 consideration. That is, generally 11n your Tier 1 flow-13 diagrams, you are careful to indicate that'the valve type 14 -can be changed. You use a different symbol, in other words.

15 But you, in the case o.' standby liquid _. controls, still ,

16- showed globe valves on the injection end, and'I'm -- I 17 suspect when you start looking at valve requirements, you 18- may.wish to change your valve. type,-or you might,-in which

-19 case you'd have to go back andl change the rule, and I don't 20- know why you'just don't.show the.other symbol 1for both;--

21 for the injection valves as well..

22 - There is something sacred about1a : globe valve -

23 apparently_that that's the.only kindsyou wculd accept for 24 that application. That would be the inference'of showing it.

25 that way.

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-81 1 MR. JAMES: And that's. exactly right. Our- .}

2 strategy on those figures was to.be-neutral-on valve type if' -

3 there was no-requirement-for a particular valve type.-

4 MR. MICHELSON:: Well, you must 'have had some 5 special reason to belie've a-globeois the only way to go.-

6 MR.. CARROLL: No leakage.

7 MR. JAMES:l Yes. The leakage-characteristic.- -!

8 MR.EMICHELSON:- Globetvalves-leak.--

9 MR'. JAMES: They don't~ leak as --

~ '

10 'MR.;MICHELSON:- But there are other special, valves 11 that they could use.--Just.:for'your edification, globe'is T

12 the-way I.-readL-it; -[

13 MR JAMES: :And_you,are exactly right. That's Dv 14 exactly right. Globe!'is there, and it i s there because:we -

15 -.want the globe valve:to be there. If we were neutral,'we 16 would have shown the' neutral sign.

17 '[ Slide. ) .

p '

L 18: MR. : FAMEG : -The next. chart is the.same. summary'as

" 19- the previous one, but nowffor the ITAAC. I*11-describe the

. form of the:ITAAC in.a: minute.- But"initermsfof1 selecting 21- entry,.we've cel'ected them..so they'reialmedLat_ confirming.

22 that the as-built facility. complies with the certiited' 23 . design.1 2 41 ,Again, system-based,.lsame_ discussion.

25 Numerical values may have ranges-.or tolerances. -

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82 1

Again, there's seme conflict with that in the material you 2

have because we put in some numbers in Stage 3 coupled with 3

a blanket caveat in the front of the report that says if 4 1 rangen weren't specified, then they were accepted industry 5

practices for the particular parameters under consideration.

6 our lawyers gasped at that and want us to go back 7

and put wherever there's a number, right where that number 8

is, to put the range. but the concept in thet it's okay to 9

have ranges or tolerances that reflect the realities of 10 life.

11 I've covered this politt. The ITAAC process by 12 definition in Part 52 endo at fuel load, so post fuel load 13 testing, which is quito a consideration because there are a '

14 lot of things checked during the hot functional tests not 15 covered by ITAAC, and the guidance -- you know, this is 16 really a repeat of what I've said before, but we're really 17 doing no more in most cases than just picking pieces of the 18 verification processes that would have been gone through 19 before.

30 MR. MICHELSoN:

21 In the area *2t r'Jes of tolerances, I'm not sure that those uce ranges appear in 22 the SSAR as well. They should be.

23 You diouldn't inver.' n-ranges or new ideas in the ITAAC, so you need to go back ar.

24 fix your -- the ranges in your SSAR.

s 23 In many cases, you didn't state the range, and now ANN RlLEY & ASSOCIATES, Ltd.

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83 1 you are statitq the range.

2 MR. BOYCE: In some cases, the SSAR, as it's 3 typically developed, would list nominal values, and when 4 you're writing an ITAAC, you're trying to get enveloping 5 values, so in some cases where the ITAACs specify enveloping 6 values, the SSAR has had -- GE is going back to reexamine 7 whether SSAR needs to be able to justify the enveloping 8 value in addition to the nominal value.

9 MR. MICHELSON: Yes. I-think it's very important 10 in the SSAR, when you have to do certain kinds of diesel 11 loading analyses and so forth, to recognize what the voltage 12 is going to.be on the bus, and you'd use the-lowest 13 tolerance end of-the-range,.for' instance, but if it wasn't 14 stated in the SSAR, you'd use the nominal.value, and if 15 you're using nominal, you had better verify that that 16 nominal-voltage is available, and you can't put that lower 17 end tolerance on it unless-you've.done'the analysis on it 18 that way.

l 19 That would be my view'Lat least.,

20 MR. JAMES:

. A lot of the key safety-numbers are )

i 21 dependent on minimal or max, one-sided numbers, like, for 22 example, core flow from the pumps. We assume 4200 so the-23 ITAAC is going to say that it has to be 4200 or greatet. I 24 think'many of the key safety parameters in i.he SAR are 25 minimum numbers-and we've got to exceed them.

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84 1 MR. MICHELSoN: As long as it's clear what you're l 2 doing, I see no fault with it. It's just that it isn't

3 always clear yet. It takes a little more QA control.

i

4 [ Slide.)

i

! 5 MR. JAMES: I'm conscious of the clock here, so i

6 I'm trying to move ahead to get through this material. This 7 briefly describes a typical Tier I entry that -- if you look l 8 at the 100 systems we've got, snywhere between a half and -

9 - in the design description -- anywhere between a half and

10 five pages-of text, anywhere between five and five plus 4 1 11 figures, the inspections, tiasts, analyses that.go with each j i 12 of those systems are in a table form like this where we t

l 13 said, here's the design committment, here's the ITA you're

} 14 going to do, here's the acceptance criteria you're going to 15 use to judge whether this was successful.

{

16 A typical system will have anywhere between two f

17 and 20 rows of entries here, 2 to.20 individual rows across 18 here in this standard table.that we put-together. So that's 4 19 what you'll see in the Stage 3 material. The next two l

20 charts summarize GE's view of the status of on this chart, 21 where we utand in terms of the GE NRC review.

22 I think it's-pretty much compatible with what Tom 23 Boyce said. earlier. WeLhave submitted.-- there's.a little 24  : approximate in here -- 100 percent of the proposed Tier I 25 material, and like I said yesterday, that approximate isn't-O ' ANN RILEY &L ASSOCIATES, Ltd.

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85 1 a euphemism for like 50 percent; it means that we think 2 we've got in like 98 percent.

3 We've got in about 115 system entries, plus other 4 generic items that we talked about. There's-some ongoing 5 work on the roadmaps and other cleanup items.  ;

6 Interactions to date indicate that we think we've 7 got consensus on basic scope and content. The size of the j i

a box is about_right, but many details are open, many, many 9 details are open. Just as an= example of that,-we got 1

10 recently from the Staff, from their Electrical Systems j-11 Branch, 37 pages of detailed questions on some of the 12 electrical systems. That calibrates you to the level of 13 discussion going on.

14 NRC is currently reviewing the Stage 3 items and  ;

15 Mr. Murley covered it pretty well in terms of the Graybeard 16 report and the regional office activities that are. going on. -

17 We absolutely agree with the staff that there's-going to'be r

I 18 intensive interactions over the-nextLfew-months..

19 (Slide.)

20 MR. JAMES: The next. chart'is something that we 21 haven't talked about yet, but this is.the parallel industry L

22 GE review going ~on. This review is being. conducted under 23 the auspices of the NUMARC' organization; andLthere are ,

~

24  : people-from the utilities, et cetera, involved. .IReviews'are-25 going on in parallel.

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i 86 1 We've covered approximately -- 20 systems so far 2 in detail, and we've got further major activities scheduled 3 in August and September. So, we aren't through yet, but 4 there are several trends that are pretty clear.

5 There are some changes dropping out as a result of ,

6 .egal -- I should add, I guess, up here, lawyers. There's a 7 couple of lawyers involved in .:he review, and they're making ,

8 comments that there are some changes needed strictly for 9- legal reasons. This thing was written by engineers, so it 10 tends to be a bit-loose in terms of tight legal 11 interpretation, so they're making legal changes.

12 Frankly, I don't think they're making much 13 technical impact. The utility input, this is the major 14 input from the utilities on the acceptance criteria in that 15 righthand table of the ITAAC table.- The utilities are 16 looking for very, very crisp, preciae untabiguous, 17 measurable acceptance criteria.

18 .When you get into re'.riews of.Part.52 with the -

19 utilities, their immediate focus-is ITAAC signoff.is. going L 20 to be the basis for fuel load, and the only. basis for a 21 hearing at.the time of-the fuel load is~whether you did'or l

22 didn't meet:ITAAC. If the acceptance criteria are mushy, 23- then that provides.an opportunity for a hearing-at the time

24. of the fuel load.

.25 So, their whole focus -- not entire focus,'but ANN 'RlLEY & ASSOCIATES, Ltd.

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87 1 their major focus is on that back end use of ITAACs and 2 whether there are chinks in the armor, so to speak, that l

3 would allow in the unnecessary second hearing.

4 MR. LEWIS: Just our of curiosity, does that 5 include statistical criteria?  !

6 MR. JAMES: Not if it's written in a way that an 7 engineer can crank through some formulae and if the number 8 is bigger than 50, go, with the number less than 50,.no go.

9 MR. LEWIS: In other_words, in order -- it's okay 10 if there is a formula, regardless-of how mushy the .

t 11 interpretation is. This a serious' question, because this 12 has come up recently.-

13 A statistical formula is never -- it can'be a 14 precise formula, but it' describes uncertainty because that's 15 what statistics is. So, I'm a little concerned. This has 16 come up in a recent case in which-there's an ongoing battle [

17 -- I hate that word -- there's a battle between the two 18 contestants about whether uncertainty can be expressed'in a 19 statistical way, in such a way as to meet a regulatory 20 requirement. That's what we're talking about here.

21 _

okay, but you haven't thought through that issue. .

22. ' MR. JAMES: I think it's -- I guess'I'm not a 23 statistician, but if the point is where reasonable '

24 statisticians could:get into a debate over whether --

25. MR. LEWIS: Is that an oxymoron? 1 O ANN RlLEY & L ASSOCIATES, Ltd.- '

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88 1 MR. JAMES: As to whether the numbers are big or-2 small, now, if it's that sort of situation.

3 MR. LEWIS: I see, so there will be a jury of 4 statisticians? okay, fine, I understand.

5 MR. JAMES: You wouldn't be happy with that, I can 6 tell you. Their view is, go out in the field with a tape 7 measure and measure it. Is it two feet or isn't it? If it 8 is, got if it isn't, fix it. That's the level of 9 preciseness that they're looking for.

10 MR. LEWIS: Yes, okay, fine, thank you. In other 11 words, you really haven't thought this through.

12 MR. JAMES: Not on a statistical -e whether 13 statistical acceptance criteria would avoid that problem, I 14 guess, no, we haven't thought it through.

15 HR. LEWIS: Thank you.

16 MR. SHEWMON: You don't_want the inspector out in 17 the field to get involved in the discussion.

18 MR. LEWIS: You know why I'm asking this question ~.

19 MR. JAMES: 'I am really trying to move.here.

20 Another one of the utility inputs is that they-21 think we probably have.got a little too much material _in l

L 22 some of the non-safety systems,_-turbine-island _ systems, for 23 example, that are unrelated to' safety,_and.they might have a 24 point, and we will be discussing'with the staff:whether-we

'25 can reduce some of that.

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l 89 1 Perhaps the second most important direction in

[JD 2 which the utilities are headed, they don't like the generic 3 ITAAC that we have prepared, for the same reason as this, 4 and, let's talk welding, for example. We have a welding 5 ITAAC, a generic ITAAC proposal on the table, and the 6 utilities think it is nothing but trouble.

7 Their first view is that welding is a massively 8 detailed process with operator training, and millions of 9 Welda, and different veld materials, and weld records, and 10 how you inspect X-rays, and how you do UT, et cetera, et 11 cetera. It is a massively detailed process. That clearly 12 should be covered by today's procedures, today's QA 13 procedures, and is not a subject for ITAAC because it is

/'D k ,)

m 14 almost impossible to come up with crisp, precise acceptance 15 criteria.

16 In that right-hand column of ITAACs, if you wanted 17 crisp acceptance criteria for welds, you would, in effect, 18 write a welding manual that would have thousands of entries 19 in it.

20 MR. CARROLL: Would they have the same objection 21 to equipment qualification?

22 MR. JAMES: Yes.

23 MR. CARROL': They do?

24 MR. JAMES: That is exactly right, they do. In 25 fact, you could repeat the speech I just made on welding, f3 L)

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l 90 1 and it would apply.

2 MR. CARROLL: That is, for example, where you are 3 going to put seiomic requirements. Do they object to having 4 seismic requirements as generic requirements?

5 LR. JAMES: If they can be written in terms of the 6 acceptance criteria, which I don't think they can. Again, 7 there are various options, depending on what the equipment 8 is, whether you do pipe prototype testing or analysis, the i

9 various options when you get down to the regulations.. I 10 think they would say, no, that is a process that is best 11 covered by Part 50.

12 In all fairness,.I should emphanize that we have .;

l 13 not discussed these with the staff yet. 'I anticipate a fair i 14 amount of discussion.

15 MR. CARROLL: Why don't we just leave..these until If later. We are running out of-time,1Mr. Chairman.

17 [ Slide.)

13 MR. JAMESt. I had.a summary which-I probably don't 19 need to go through except that I-think we will all-agree the red box le.probably right. -We are making. progress, but we

~

20 21 have a lot of work to do'. That.is the-bottom-line.

22 Do you Want me to stop?

23 MR. WARD - No.-'We just pointed out to the: members 24' that'weren't here that attached to this' package are ther 25' ITAACs, and'these are the complete ITAACs,-five. complete i A

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ones, that is what an ITAAC is.

these, and this is a sample of five.

There are about 85 of 3 MR. CARROLL: One thing that might be useful for 4

4 us to have is, yesterday you marked up the ITAACs and design 5 descriptions as we were offering you comments.

6 MR. JAMES: Yes.

7 MR. CARROLL: Do we have a set of those, Paul?

8 MR. BOEHNERT: No.

9 MR. CARROLL: Can we have a set?

10 MR. JAMES: I would be happy to. I think what I 11 had better do is, they are in my terrible handwriting, and 12 what I will do, I will rewrite it so that it is legible, and 13 send a copy to the secretary, f

(m},/ 14 MR. MICHELSON: Thore are matters of inconsistency 15 and so forth that everybody admits that there is going to be 16 a QA on this thing.

17 MR. WARD: Unless you have some particular 18 interest, our task, as I see it, is to make some judgment or-19 form some opinions about the process. We have only looked r

20 at five out of 85, and if the process is depending on us to l 21 scrub each of them, it is not going to happec l

j 22 Thanks very much.

23 Let me just take the remaining four or five 24 minutes that we have to discuss what the committee might-25 want to say in a letter. With Carl's help, I have a few f)

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j 92 j 1 comments here, and I am going to go off to draft the letter, 2 so let me see if the committee thinks this is about what we- l 3 ought to say.

4 First of all, as far as for these system ITAACs 5 that we have looked at, we think that the scope and the 6 format are about right, and the process can be made to work.  !

j 7 The second comment is that there remains a large 8 content problem. Although the-form and scope are okay, the 9 content, as indicated by our sample, and other things we 1

1 10 have heard, there is a problem with consistency, accuracy, 11 and completeness, and a real mhjor what I call'a QC program 12 is-needed, and we have heard that there'is to be one. The .

r 13 staff, as Murley told us, has a major effort.- GE has one, 14 and the industry has a separate one. l 15 Murley went so far as to say that if completion.of j 16- that QC program means the FDA is. delayed, sobeit. I guess I i

17 would say we. agree with that. That is not a point of making  ;

18 the letter,_but we can argue about that.

I j 19 Another point was that --

l

- 20 MR. CARROLL

I guess I would like'to hear from :1 j

21 Tony one more time your arguments as to why ITAACs_are not-22 needed at the time of FDA?

{- 23 MR. JAMES:- Our'v3ew is that ITAAC really_ derived

24 _directly.from the certified design, and-that onceLyou have 25 done the-design view, and

have the design settled,'that.is

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93 l 1 all done as part of the SAR review, and ITAAC can be 2 developed later, because they are nothing else than the 3 distillation of what is in the SAR.

4 There is an exception to that on the DAC ITAAC.

S We agree on the DAC ITAAC that that is necessary to form the 6 basis for a safety finding, but not the other more routine 7 ITAACs.

8 MR. MICHELSON: But you still intended to do it 9 before date of certification?

10 MR. JAMES: Yes.

11 MR. MICHELSON: Somewhere between FDA and 12 certifications is when you would do your ITAACs, if I 13 understood you correctly?

14 MR. JAMES: Yes, they would.have to be completed .

15 to support the certification hearings, yes.

16 I should state,_still, that the GE position is 17 that we think it can-be decoupled, but we are-not currently 18 requesting decoupling. We think it is achievable on a 19 coupled basis.

20 MR. WARD: I didn't-cattch thet. He thinks they 21 could be done --

22 MR. MICHELSON: Later.-

23 MR. WARDS. - after FDA, but before certification?

24 MR. MICHELSON.- That 's right. -

25 -MR WARD: -Does that make-a big window?

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94 l

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MR. MICHELSON:

work on it, certainly.

They would have a few months to 3 MR. WILKINS : I also heard him say that they are 4 not at the moment fussing about this because he thinks he 5 can meet the schedule to do it before FDA.

6 MR. WARD: And you heard what Hurley said, didn't 7 you.

8 MR. MICHELSON: I wanted to 11,1 sten to add only 9 that the committee did endorse the idea of doing it after 10 FDA in their September loth letter, and maybe we want to 11 reconsider that endorsement, because that letter still is on 12 the table, and the commission hasn't decided yet. If we 13 want to change our mind, this would be the letter to do it f

(_}/ 14 in.

15 MR. WARD: That is a good point.

16 MR. dICHELSON: One way or the other.

17 MR. WARD: Another point in the letter is that it 18 is not clear how fire, flooding, pipe breaks and safety 19 analysis assumptions 1or events outside containment are 20 committed to in the ITAAC process, and we are promised that 21 we will hear more about that in a meeting later this month.

22 MR. MICHEISON: So we just kenp it open.

23 MR. WARD: A couple of other items, this issue 24 about how the ITAAC assumes.or assures the procesa some sort 25 of a closure on the reliability assumptions that were made C')

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in the PRA, I think, is a point we need to discuss in the letter. I haven't figured out exactly what we are going to 3 say about that, but --

4 MR. MICHELSON: I would propose to keep that one, 5 put it in the letter, but keep it open to hearing uore about 6 it. I think everybody ought to think through that one 7 first.

8 MR. CARROLL: The answer I have heard'from the 9 staff and GE is that_that is what the reliability assurance  ;

10 program-is, and it would be separate from ITAAC.

11 MR. WARD: I guess I would like~to know a little 12 more about that also. We are going to try to get a look at 13 this appendix,_19(k) to the SAR, and see what it actually_

) 14 says.

15 MR. KERR: A reliability assurance program is fine-16 if it can work. -I haven't seen the reliability _ assurance 17 program, and I am not sure.you can assure the reliability'of

18 something whose reliability is assumed to be ten to the ,

, 19 minus four per demand on an individual plant; basis.

20 MR. KRESS: You certainly will~not'be.able to do i

21 it at the ITAAC stage.

l 22 MR._ CARROLL: I_ guess all'I'm pointing out is that i 23 I think it's a separate subject from the subject of ITAAC.

24 MR. WARD:- I'm-not sure.- I.think'it'sfcomplex,

-25 it's difficult but I mean just waiting until after the COL -

.()

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96 1 is hardly -- if you have got some component that doesn't 2 come anywhere near close to satisfying the requirement 3 you're in a big hole.

4 MR. CARROLL: Didn't Tom or Bill say that they 5 werE going to tell us more about that?

6 MR. WARD: Well, yes, but I think the point of our 7 letter is that this deserves a lot of attention.

8 MR. MICHELSON: One other thing perhaps you should l

9 add and GE pointed it out -- there are quite a few open 10 items on the ITAACs yet. You know, it isn't a question of 11 larguage or anything, it's just not there and it's still' >

12 open and it will appear later, if I understand the process 13 correctly and we probably should point that out to the -

14 Commission.

15' MP. WILKINS: That we recognize that these are' 16 some of the 80 ITAACs that we haven't seen yet or open 17 items.

18 MR. MICHELSON -- To these 57 l 19 MR. CARROLL: Virtually all the ITAACs are in now.

4 20 That's what the Stage 3 submittal was'about.

]

21 MR.' WARD:- Another time --ilet me get one more

. 22 item -- would be whether or not~we want-to'look'at a sample-l 23 of.other types of ITAACs and we'll bring that point up;in

^

-24 the letter. We'll decide.what we want~to do about it.

25 -I can't totally judge these ITAACs without-knowing lQ

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97 1 what the implications of the generic ITAAC are to the system 2 ITAAC.

3 -MR. WARD: All right.

4 MR. LEWIS: Just before we brush this reliability 5 thing under the table --

6 MR. WARD: I don't like what you said. We don't 7 intend to brush it under the table --

8 MR. LEWIS: No, ilo --

9 MR. WARD: I think we are going to make it a point-10 in the letter.

11 MR. LEWIS: I understalid.

12 MR. WARD:- Okay.

13 MR. LEWIS: I just think-it is more important that 14 it came up during the questioning. As you_know, I have a 15 fixation on diesels'and so I looked at the-line-that -

16 describes the--reliability testing on-the-diesels.

17 MR. CARROLL: 'We discussed it yesterday.

18 MR. LEWIS:- You did? Well,-I wasn't there.-See, 19 the design commitment is that the manufacturer has conducted 20 reliability tenting on the units which is one'of the 21 requirements. The test is that the manufacturer's test 22 document shall be visually-inspected, which doesn't give me 23 the comfort.I really-wish.I could get:from it.

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98 1 that the required reliability has boon performed and that 2 the diesel generator has passed the test requirements.

3 MR. CARROLL: What that's about, Hal, we asked 4 about it yesterday, is that there is Reg Guide on testing of 5 diesels, diesels by type or new diesels. Charlie?

6 MR. WYLIE: To be better characterir.ed, it would 7 be acceptance tests for the diesel.

8 MR. CARROLL: And all that means, all that really-9 means, and it's not clear at all the way that it's stated, 10 is that they will do the tests-as prescribed by this Reg 11 Guide.

12 MR. LEWIS: But given the fact.that we have been-13 told that these thine3s have to be very precise,_it has to 14 say that, and then as you know accepting the Reg Guide 15 produces an entire array of lasues which are not-made clear.

16 There is a-substantive issue here'and I just'want i 17 to make sure it was treated.that way.

18 MR. MICHELSON:- The Reg: Guide' appears only in tier 19 two, of course.

20 MR. CARROLL: But this is an invitation. for'a 21 smart lawyer at:the tier one level to really.have'a' field-22 ' day.

23 MR. LEWIS: Or even a: smart something'else.--

24 MR.-WILKINS: LAgain we;are talking oxymorons.

25 -MR. WARD: We will wrap up'this topic then..

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l 99 j

() 1 MR. MICHELSON: There may be other items that 2 members wanted in the letter.

3 MR. WARD: Oh, yeah, yeah, yeah, yeah. Excuse me.

4 MR. MICHELSON: Any other items?

5 MR. WARD: You'll have an opportunity. We'll get 6 you a draft before the end of the day.

7 Okay, let's go to the next topic then, which is 8 the discussion of generic letter 83-28 and I believe Mr.

9 Wylie has some comments to lead off.

10 MR. WYLIEt Thank you, Mr. Chairman.

11 The background material for this part of our 12 meeting is under Tab 3. The purpose of this part of our 13 meeting is to review a proposed supplement to NRC Generic

%,) 14 Letter 83-28 and to consider the differing professional 15 opinion filed by Mr. Charles Morris of the Electrical 16 Systems Branch of NRR.

17 Generic Letter 83-28 was issued following the 18 failure of reactor trip breakers to open on demand at-the 19 Salem plant in February,1983.

20 The Staff has issued a proposed Supplement 1 to 21 Generic Letter 83-28 for public comment which expired August 22 3rd.

23 The Supplement 1 to Generic Letter 83-28 would 24 inform the licensees that life testing of reactor trip 25 breakers and periodic replacement of breakers or components I -

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100 1 as originally described in Generic Letter 83-28 are no 2 longer needed.

3 A differing professional opinion was filed by Mr.

4 Charles Morris of the NRR Staff regarding the issues covered 5 by the proposed Generic Letter Supplement 1. Mr. Morris 6 requested to pursue his differing professional opinion with 7 the ACRS by letter of June 18,-1990. At that time the ACRS 8 agreed to defer consideration of that request pending a 9 review and decision by the CRGR.

10 The control and Electrical Power Systems 1

11 Subcommittee met August 4 this week with :nembers of the NRC j 12 Staff including Mr. Morris, and reviewed the proposed 13 Supplement 1 to Generic Letter 83-28 and heard and -

14 considered the differing profossional opinions expressed by 15 Mr. Morris.

16 We have arranged this morning to have members of i

17 the staff summarize their background history land actions ,

18 leading up to supplement.1 and_for Mr. Morris to summarize 19 his differing professional pinion on the proposed-20 supplement.

21 Are there any comments from members of the 22 subcommittee?

a 23 [No response.)

24 MR. WYLIE:. Hearing none, _ suppose we proceed, and 25 I believe Carl Berlinger of the_ staff is going to make the O ANN _RlLEY & ASSOCIATES, Ltd.

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101 1 initial presentation. Carl?

2 [ Slide.)

3 MR. BERLINGER: Good morninf. My name's Carl 4 Berlingor. I'm presently chief of the Generic 5 Communications Branch, Division of Operational Events 6 Assessment.

7 The subject of my presentation or my portion of 8 the pressntation is to summarize Generic Letter 83-28 9 Supplement 1, some of the bacRground information, and to 10 briefly describe how we arrived at the.c'ecision to issue 11 this supplement.

12 (Slide.)

13 MR. BERLINGER: The Generic-Letter 83-28 requested .

14 licensees to implement long-term corrective actions in' 15 response to the-Salem ATWS event.- Some of.the major actions 16 included in the generic letter included auto-actuation of-17 shunt trip attachment for all trips in~ response to.all-18 reactor trips, implementation of preventive' maintenance and 19 surveillance testing programs,-and Actionz4.2.3-and 4.2.4-20 with, regard to performance of life testing and the period 21 replacement. based on life-expectancy: coming out of the life 22 testing. program.

23 As it turns out,_ licensees _do not_ fully impleme,4 24- . life testing and-periodic replacement. programs andJdo not 25' feel', based on some of the_ actions that_had been h ANN RILEY & ^ ASSOCIATES, Ltd.

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102 1 implemented, that these were necessary.

2 MR. CARROLL: Why don't you amplify a little more 3 on that, because I think it's important to what the 4 Westinghouse owners group actually did do.

5 MR. BERLINGER: The Westinghouse owners group did 6 undertake a study involving life testing, and tl ey did cycle 7 reactor trip breakers up to, I think it was 5,000 cycles.

8 Their program ended at that phase, indicating that'they felt 9 that trip breakers could be replaced, I believe it was 10 between 1,000 and -- or 2,500 cycles. 2,500. And in fact, 11 that program was completed in approximately 1989, something _

12 in that time frame.

13 MR. CARROLL: So-when you say did'not.fally.

14 implement in that bullet, what it really means;is they did 15 do this testing, they rubmitted the results, and the staff-16 has never --

17 MR. BERLINGER: Accepted those results.

18 MR. CARROLL: -- accepted those results. And your 19 problem with the. testing was?

20 .MR. BERLINGER: 'Well, the problem expressed by 21 members-of the staff was that the testing did.not go for a 22 long enough period of time, number of cycles-to' failure, and 23 it did not --

24 MR. CARROLL: BUt in the 5,000 cycles,~other than 25 nornal maintenance and lubrication, the. breaker l performed ANN' RlLEY &- ASSOCIATES, Ltd.

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105 1 flawlessly?

2 MR. MORRIS: I don't think so. The breaker was 3 used as the vehicle for exercising the under voltage 4 attachments, and they were the subject of the test and they 5 performed admirably 5,000 cycles or more.

6 MR. LEWIS: When you say that they hadn't done 7 cnough testing, did the generic letter specify how much 8 testing they should do?

9 MR. BERLINGER: I really didn't hear the middle of-10 your question.

11 MR. LEWIS: You said the staff didn't accept the 12 results because they hadn't done enough_ testing, and I 13 wonder against what criteria-the not enough was judged.

14 MR. ROSSI: I believe the generic letter asked 15 people to do life testing and establish a replacement 16 program based on the life testing --

17 MR. LEWIS: I see.

18 MR. ROSSI: -- and' submit it to the' staff for.

19 approval. What basically happened is'the industry submitted ~

20 something and there were-lots of questions 1back and forth 21 and we never reached closure on the issue.

22 MR. LEWIS: Yes.-_--I-was just trying-to find out 23 what.the criteria were.- You would interpret --

24 .MR. ROSSI: I don't know the criteria.

25 specifically. .I-don't know whether Carl;does or not.

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MR. LEWIS: But when you use the term life testing, that suggests that they should test to the end of 3 life or something like that. Was that the criteria? j i

4 MR. RocSI: I don't think that criteria was l 5 provided in the generic letter.

6 MR. LEWIS: I see. Okay. Fine. So in a sense, 7 it was sort of a discomfort, not based on anything that was B laid down in the generic letter?

9 MR. ROSSI: That's correct.

10 MR. LEWIS: Okay. Fine.

11 MR. WILKINS: Let me ask a question. The second 12 line says the GL 83-28 requested licensees to implement. It 13 doesn't say required licensees to --

14 MR. ROSSI: Yes. Let me answer-that question, 15 because there has always been a moderate amount of confusion 26 over what we can do in the NRC with a generic letter or a 17 bulletin.

18 A generic letter can go out or a bulletin can go 19 out and it can request licensees to take certain actions, 20 -like in this case, it can request life testing of the 21 breakers, it can request that licensees change their design-22 to automatically actuate the shunt trip every time you get a 23 reactor trip signal.

24 The only thing that a generic letter or bulletin 25 can require of the licensees is that they submit under oath On V

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105 1 and affirmation a written response -- that can be required 2 -- saying whether they are or are not going to do the 3 requested actions and the reasons for not doing them if they 4 so choose. So a generic letter does not make it a 5 requirement; it's a request. Then the usual situation is 6 that licensees will come in and generally do what is-7 requested, but they do now and-then do something that's 8 alternate to that.

9 So a-generic letter or a bulletin in terms of 10 actions to change the design or to do life testing do not 11 have the status'of a rule or regulation.

12 MR. WILKINS: Do I infer from that that the .

13 licenses have responded to you on Item 4.2.3 and 4.2.4, or 14 that they have_not responded? ,

15 MR. ROSSI: I think that the conclusion is'that we ,

16 -- that the staff never agreed that theylhadfadequately 17 responded to the life testing, or_to periodic _ replacement 18 -programs, and we have done~a review to decide what further 19 guidance we needed to formally lasue to-licensees on how -

20 they could do it.

21 The bottom l'ine of this presentation is that when 22 we looked at whether we needed to issue further guidance'on-t 23 doing the life testing and having.a replacement-program,'we 24 concluded-that the programs-were no-longer necessary, so.we~ o 25 concluded they're.not necessary,,so we' don't need-to really  :

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I 106 1 discuss anymore, what the criteria for the life testing were 2 and that sort of thing. We have concluded, for reasons that 3 Carl will discuss, that it's just not necessary to do those i i

4 any longer.

5 MR. WIIIINS: Thank you. f 6 MR. CARROLL: Let me go back to one point here.

7 Would the Salem event have been a non-event if the first 8 star under the second bullet had been in place?

9 MR. ROSSI: You have to make an assumption,'but -

10 it's likely that it would have been, yes'. ,

11 MR. CARROLL: That's the real'important thing that.  !

12 was done out of all'of-this.

13 MR. WYLIE: _One other-. thing up to Ernest's 14 question, I believe that all the licensees have actually .

15 implemented the first two.

16 MR. CARROLL: The first two,'oh, yes.

17 MR. BERLINGER: The only actions which were 18 _ requested that licensees did not fully:-implement were.these-19 two regarding life testing and_ periodic = replace: rent 20 Programs. The other --

21 MR. KERR:. . Wait'a_ minute. What you--said: earlier, 22 the.' difference was that the staff'never'did decide whether-23 they:had-been= implemented or not.: 'New.you're-saying that. '

24'- the staff' decided that they=hadtnot been implemented.-  !

25 MR. ROSSI:i Weididn't:reallylcome:down.today to _

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107 1 review exactly what we did in term of the staff's review of 2 the life testing and the periodic replacement program. I 3 think I'd like to summarize it as one where we never came to 4 closure with the industry on it, and we were in the process 5 of deciding whether we needed to issue additional guidance 6 to tell the industry what they needed to do to satisfy the 7 staff.

8 That process was not completed. So, what -- in 9 the process of trying to develop further guidance on how to 10 corplete items 4.2.3 and 4.2.4, we concluded-that-they 11 wer n't necessary any-longer because the automatic actuation 12 of *.he shunt trip and the prwentive maintenance and 13 surveillance testing programs that were implemented by_the-14 licensues have appeared to be sufficient to improve the 15 breaker reliability such that those last two items are no 16 longer needed.

17 So, we didn't really come to talk about'all_the 18 details of the life testing and the periodic replacement-19 prcj,e because we never reached closure. And to.say 20 whether they'didn't do - ..the industry didn't do enough or:

21 whether.the-staff-was requiring ~too much or.we were too 22 vague or whatever, that process was just nevericompleted, in 23 my opinion.

24 MR. LEWIS:- I-have worse. memory thanimy friends, 2 25 so remind.me. ; Salem was'nine years _'ago, _ Was it one'or more-l l

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108

() 1 2

than one breaker that failed at Salem?

MR. ROSSI: Both breakers failed.

3 MR. IIWIS: Both breakers, that's what I seem to 4 remember. Were they near the end of their life? I'm trying 5 to understand the relevance of life testing to Salem.

6 MR. ROSSI: I don't know that there was ever a

' determination that they were near the end of their life.

8 MR. LEWIS: So life testing was sort of put in as a

9 an add-on, independent of what caused the Salem event.

10 MR. ROSSI: Life testing, in my opinion, was put 11 in as something that, in general, is good to do, and 12 probably wouldn't hurt and probably would improse the 13 reliability of the breaker, q,) 14 MR. LEWIS: Thank you.

15 MR. ROSSI: The big problem at Salem was that the 16 under voltage trip attachment did not have enough margin to 17 reliably trip the breaker, and so v~ asked licensees to 18 implement design changes to aute  ; ally actuate the shunt 19 trip, and we also asked them to do further things to 20 maintain the under voltage trip attachment better so that it 21 would be core reliable.

22 i:R. LEWIS: I'm jus' trying to get oriented. It 23 was sort of lips painting the car after the engine failed, 24 because it was laid up anyway.

25 MR. CARROLL: But a breaker, Hal, is sort of like O

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l 109 j 1 a classic car. 'igt know, you can make a breaker -- the same 2 breaker, at least it, terms of its frase, work forever.

3 MR. LEWIS I know, it's like George Washington 8s l.

4 original ax. I know that very well.

i 5 MR. SHEWMON: 7et's go on.

6 MR. BERLINGERt As part of the staff's followon'to 7 determine what action or additional action was needed,-if  :

8- any, we reviewed operational experience data, spanning a 9 pe'riod from 1986 to early '91, and those_-- that data was .

10 the data found both in NPRDS and'the:LER data system.

11 In the period noted here,-it was only reactor trip 12 breaker failure to open on demand. That was one_that.

13 occurred at McGuire in 1987,-and that. failure was primarily 14 associated with cracked weld-and misalignment-because-of the 15 cracked weld, and it really was not an end-of-life type of. -

16 failure. -

17 Subsequently, there's been a recent event at Palo 18 Verde, Unit-III, in-March of this' year, and that' event:also j 19 was not' associated with end-of-life-type-' failure.

20 MR. CARROLL . It also-was on!a unique breaker that 21 no one-else haa.-- -

. R MR. BERLINGER: It was a different' breaker:from.

23 what anycne else in1 the industry is: using- for a:. reactorstrip q 24 breaker. . As-:a matter of fact it's-the-only plant:that-has 25 that model; breaker installed as a ; trip breaker.

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110 1 [ Slide.]

2 MR. BERLINGER: As a result of > o review that was 3 undertaken t. look at reactor trip breaker reliability the 4 Staff concluded the following things.

5 One was that actions that had been completed as a 6 result of the generic letter actions that had been 7 implemented had been very effective in reducing -- well in 8 improving reliability in the trip breakers at the 9 reliability or actually looking at the failure rate, the 10 failure rate had gone fror ten to the minus two, ten to the 11 minus three range to the ten to the minus four, ten to the 12 minus five range of railures per demand.

13 In addition, in using a PRA model or the PRA 14 models from NUREG-1150, the core damage. frequency for AWS 15 events versus reactor trip breaker failure rates was studied 16 and I have attached in the handout slide -- let me just 17 throw that out for a second.

18 [ Slide.]

19 MR. BERLINGER: This slide is meant to show that 20 with failure rated of the reactor trip breaker on the ten to 21 the minus two range or higher that the core damage frequency 22 clearly was affected directly by reactor trip breaker 23 reliabiitg *ut below ten to the minus two, ten to the minus 24 three raari t'e c. ore damage frequency was relatively 25 unaffect% by rip breakers.

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1 MR. LEWIS: You know, I dontt want to start a 2 debate but let me stipulate ---incidentally,'you'should -

3 spell affected correctly next time around. --I -- w a n t t o 4 stipulat( chis is the wrong way to use a-PRA because what 5 you are doing is comparing--the effect of'the reactor trip 6 breaker failures with all the other background things 7 instead of with the safety goal and we are supposed to be -

8 moving into the safety goal era, and it may be.that 'way up f

i i 9 there when it is more than comparable with the other things, 1

j 10 it may still not require help'or it'may be that it.should ao 11 much further down, because it's-the wrong-thing-to compare i 12 it with, but that's not your job. I just'want to say that.

13 MR.-BERLINGER:- It is well beyond my--area of 1' 14 expertise so I do not want to-get into-an argument with you.-

{ 15 MR. LEWIS: We are not getting into an' argument..

16 MR. CARROLL: _ You probably also should have added, 17 Hal, that this isn't particularly meanihgfulLunless-I know A8 what sort'of uncertainties are associated with'--

9 -MR. LEWIS: I~didn't'say that but you'did.

'O MR. SHEWMON:- Can we go on now?. '

21 -[ Slide.)

22 MR. BERLINGER: Looking at the conclusions drawn 23- from utilizing that chart we concludeditherefore that 24 further actions to address end of life degradation 'as it ~

25 ~affected reactor trip breaker reliability _were not g

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-112 1 jurtified.

2 Therefore the generic letter supplement proposes 3 to remove the two-requirements for actions 4.2.3-and 4.2.4.

4 MR. WILKINS:- But you just got through telling me 5 those weren't requirements.

6 MR. BERLINGER: Pardon me?

7 MR. WILKINS: .You just got through telling em that ,

i 8 those are not requirements.

9 MR. BERLINGER: I stand corrected. If I-said 10 requirements, I meant actions requested.

11 MR.-WILKINS: -I-was-talking about on the slide.

12 MR. RoSSI: Over the years we have had a great 13 deal of difficulty getting-everybody on~the-Staff to.-use the 14 right word and as a matter of fact'I'am sure you will. find 15 the generic letters that have been issued that say " required 16 actions" but legally we have gone through this with our 17 General Counsel a number of-times and in-actual fact the 1 18 actions are only requested and we apologize for-the error in l

19 the slide.--

20 MR. LEWIS: Are you suggesting your staff is not.

21 trainable?

22- MR.-BERLINGER: No,.we just forget.

23- MR. -- WILKINS : ' I've made my point -- go ahead. i j

24- MR. BERLINGER: -I.had-nothing more that?I want to 25 .say.

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i 113 )

, 1 I want to give sufficient time to-Mr. Morris to 2 -express his DPV/DPO.  !

3 - MR. CARROLL:' People may be' curious'about the'last 4 thing in the package, the Union Electric letter.

5 MR. BERLINGER:. When I was here on Tuesday at that 6 point in time-the comment period had ended. .We only had one ,

7 public comment that.had been submitted from Union Electric.-

8 It was generally in support _of the. action ~ proposed-that 9 issued a supplement.

10 Since.then, as of thic morning, wel received.one- l 11 comment letter from the Ohio CitizensifortResponsible Energy 12 and they oppose the issuance of the supplement.

13 Unfortunately the basis for'their opposition as-stated in

. . 14 their letter are unfounded, clearly I. don't think they 15 understood that_in fact the problem here-is that' life 16 testing hadn't been implemented so it has no real effect on' 17 the action's that are being-taken.

18. MR.-LINDBLAD: Tom, regardless of_whether the 19 supplement ^is_ issued,'is it the Staff's' intention to; respond 20 to the people who have reprrted to you what their actions 21 were with regard _to life testing?.

22 . MR.. ROSSI:

.I think there was a statement in the.

23 generic letter that addresses that someplace in here. -We do 24 not intend to go back to'each licensee.on this.

25 The generic letter will closeLthe items 4.2.3 and:

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- 114 1 4.2.4, I mean that's our intent. We will issue the generic 2 letter and that closes everything.

3 MR. LINDBLAD: Does that say that the action-of 4 the licensees was sufficient or not?

E MR. ROSSI: No, I think it says that the items are 6 not -- that the requests in the original generic letter are 7 no longer necessary and then I believe -- I don't know that 8 I have the generic letter in front of me here -- it's got 9 some words that to the extent that licensees may haIe made 10 commitments to programs for periodically-replacing reactor 11 trip breakers or components in response to generic letter 12 83-28, they may review and modify these programs taking into 13 account their plant-specific operating experience,

.O V 14 maintenance programs and root cause determination programs 15 for reactor trip breakers.

16 That tells them that they can go back and review 17 what they told us with respect to these two items.

18 MR. LINDBLAD: And those actions are at their own 19 discretion?

20 MR. ROSSI: At their own discretlon but this 21 generiu letter will close the issues-on these two items.

22 MR. BERLINGER: Any further questions?

23 MR. SHEWMoN: No, I don't believe so.

24 MR. BERLINGER: Thank'you very much.

25 MR. WYLIE: Now Mr. Morris.

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115

/ h 1 MR. SHEWMoN: Please proceed when ready.

V 2 MR. MORRIS: My name is Charles Morris, and I am 3 from the Electrical Systems Branch in NRR, and I think I can 4 accept most of the blame for the staff being disinclined to 5 accept Westinghouse's testing, and for insisting on a life 6 test.

> 7 The reason for insisting on a life test is that.I 8 thought it would be refreshing to have some hard data. 11 9 had reviewed various PRAs, sections ot PRAs, and I was 10 constantly brought up against the use of engineering 11 judgment.

12 As I said on the 4th, my own judgement with 13 respect to reliabilities, and wear, and other parameters of (O

ms / 14 devices, even those with which I am familiar is suspect, and

! 15 I am very skeptical about other people's engineering I

l J6 judgment, particularly when they have to have a number.

17 I know there are problems with testing, but I 18 thought it would be refreshing to have a number with respect 19 to the breakers. I had no particular reason to suppose that 20 the breakers were unusually unreliable, or risk significant.

21 The Commission decided, in 1983, that matter for 22 me. They gave me the issue on a platter, and I thought, 23 let's run with it. We really.ought to have one hard fact.

24 I don't care, as I said, for engineering judgment, and 25 reasonable conclusions because nature is all to often

(~T U

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116

-1 unreasonable.

2 Given that set of circumstances, I decided when 3 the~ staff proposed around '87, when I joined the branch I am 4 in now, to substitute for the life test something that you-i 5 could vaguely describe as on-going qualification, I thought 6 -that was unsatisfactory, and though tactful, it wasn't a 7 rigorous way to resolve the matter.

8 so I suggested that we should do a proper life-9 test and, at that point, my life became very interesting, 10 and I hope that that will come to an end this afternoon or 11 this morning, because we have really gone on with this long-12 enough.

13 I think I have given enough reason for.how I:'got

~

14 us into the DPo situation, and why we_are where we are 15 today, and my writing has been, shall-we say, extensive on-16 the subject, because I-was swimming against the stream, and 17 I thought I should have a reason ~for everything. I f _ my -

18. reasons are invalid, that is-just my-fault, but I thought 19, they should be based on something.

20 I-don't-want to use up my whole.20-minutes, 21 because I have written just about everything I have thought 22 on the subject, and if:LI.have made errors, they'are not

-23 . graven.in-stone, but they~certainly,are down on. paper.

24 There is a-paper;tra11. Everywhere-I vent wrong, that was 25 just my fault.

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117-p

! 1 So we come to.what I am objecting to now. The 2 staff has decided that we had one failure based on the NPRDS' 3 source. When-I~ consider the questions that occurred to me 4 in trying to design a-life-test, and all the uncertainties 5 and the matters that I thought could not be resolved until t

6 we at least stuck our tow in the water and begun-the

o

7 testing, I found it miraculous that the NPRDS, which I have l b been told was constructed for other purposes, should-9 suddenly function as an error-free data source.  !

{

1

! 10- I don't know all the steps that are taken in i

i_ Il gathering the data, preparing the breakers, interpreting the 12 cause of failure, and so on. .They.are:various, I am sure.

]

! 13 Precisely because I don't know'how NPRDS functions, and 14 because of'the difficulties I had in constructing a life

15 test, I was inclined to be skeptical.

16 There is no doubt that we:only found one failure, 1 f

17 and even by the most pessimistic,-in terms'_of'my realism,.

18. not in terms of nature,'that there might not be more than 19 ten, it'was, at least..in its useful life, reliable:enough,.

20 if you accepted the limit on the failure on demand-of ten to-21 the minus'three.

22- We instituted maintenance.and surveillance,land it-23 should have had a--beneficialieffect.on the failure rate,1 and 1 24 it certainly seems to have had on the useful life of the 25 breakers.

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(_/

1 However, there were many more failures, slow 2 failures to trip, delayed trips, and these were usually 3 caused by the under-voltage attachment failing to exert 4 enough torque.

5 That suggests to me that perhaps the maintenance 6 and surveillance program contributes to the reliability, 7 perhaps not enough. The most useful device was the shunt 8 trip device with its, basically, much greater torque, which 9 made it easier, or more sure that the breaker would trip.

10 of course, the shunt trip is an active device, and 11 so, on occasion, the staff has put too much faith in passive 12 devices as being fail safe. There is no question that the 13 shunt trip device will not fail safe, and it has the whole

(~))

(_ 14 train of components, it has to function in order to have the 15 impulse.

16 But the fact remains that with the shunt trip 17 device, and with the maintenance contributing uncertain f

l- 18 fractions, we seem to have, if we can believe the~NPRDS l

l 19 results, a device which is reliable enough-in its useful 20 life.

21 However, devices wear out like everything else and 22 I do not know where that wear-out will be. The NPRDS sort 23 covers five years. In those five-years, each of those 24 breakers tested will have been cycled between 100 and 500 25 times. Now that is.a very short useful life. As a matter es U ANN RILEY & ASSOCIATES, Ltd.

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119 1 of fact, you wouldn't expect too many failures with those 2 breakers even if they had not been maintained over-100'to 3 500 cycles.

4 I don't think it's extraordinary that we have only-5 found one or two or even if we had'aore. It doesn't mean 6 the breakers aren't reliable in their useful life but we get-7 to wear-out.-

8 MR. SHEWMON: Sir, could you tell me what the 9 regular testing is for these?- We both agree that it is 10- important that these things function._ We both agree that.

11 there is many different kinds of conditions out there and we-12 both agree there's several whys to skinLa' cat usually and so-13 my question is really-getting~at to what extent is there a (3

(,/ 14 testing program or surveillance program'in place which is-15 demonstrating the safety.of these even if it's not 16 generating quite the qu:mtitative kind of.date?

17 I'm somewhat concerned.about talking.-about~a' life 18 test because what somebody does in'a laboratory flexing 19 souethin- aay not reflect what;happens out;in'the plant as 20 well, sosif there is a good surveillance program it might .,

21 actually get us more' reliable data.

22 MR. MORRIS: I-never supposed that;,it would give 23' us all the-answers. It would give us anianswer under-a j 24- given set of conditions and-if we then compared she.

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! 120 1 1 got to. find some way to bring the plants up to the lab i 2 results.

3 MR. SHEWMON

Could you tell me though what-kind-

4 of a survoillance' program is in place or a testing program, i~ 5 if you'll let me use that word, for_t5ose that are actually- -

i e 6 in the plant?

. 7 MR. MORRIS: You have people here who probably L 8 know more about that than I do,-but I'll tell you.what I- -

9 understand are the maintenance and surveillance programs.

10 They have monthly checks en these-breakers.where l 11 they measure the trip torque, where-they measure the time to 1

12 trip. -If they have a failure-or.a delayed trip they'do a-13 root cause analysis, and they lubricate it-or they adjust it, 14 get-it back into the -- within specs from time to trip or 15 replace the broken thing like which-is frequently the

16 undervoltage-attachment.

17_ MR. CARROLL: . Or worn parts.

i .

18 MR. MORRIS: . There are a number of anecdotes which 19

~

are unspecific which1say we think-this failure was-due to^

20 wear. Wear of what,_why they-thought ro I-don't know.

l 21 _You know,'.it's'like-having the' Bible on a [

22 postcard. . There's a lot of stuff you would_like:to know but-23 it isn't there.

4 24 MR. CARROLL:' That-was one of-your comments. -l 25 Tuesday was-that you:really;can't-get-that information by:

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H 121 1 looking at NPRDS records because it's usually a very crypt'ic 2 description of-what the problem was.

3- MR. MORRIS: Right and they frequently say, at _

4 least in a dozen cases, we don't know.what happened. -We 5 sent the-breakers'back to the manufacturer for him to tell 6 us.

7 Presumably there was a follow-up IER or whatever-8 report but it doesn't always appear in-the source and-I 9 don't recall seeing~it.

10 MR. MICHELSON: What fraction of the breakers out 11 there are being reported in NPRDS?-- Do you have any fool for 12 that?-

13 It's a voluntary system =and it has varying, -

-34 reporting and some utilities fo'cus in one area and others in 15 other areas.

16 MR..ROSSI: I' don (t think we have the answer to 17 that. He may have it in all his stackEof . paper but Is 18 don't --

19 -MR..MICHELSON:- -Being a. fairly critical.itemb I am 20 hoping that it's a very high percentage of reporting.

21 MR. ROSSI: Wait a minute, I think-I'de.::have-it.

22 It'is in the' letter, I believe,:that I. signed to Bill

-23 Russell on December 12th. It's."While.the-completeness of:

24 reporting ;across thei entire .5000- components perJplant, scope

, 25 of NPRDS has been estimated at 65 to 70 perce'nt by=AEOD,' it -

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l-f l-122 1 is reasonable to expect more complete reporting of safety 2 significant failure modes-and safety-significant equipment a

3 such as complete failure to-open of the reactor trip '

4 breaker" so it is somewhere, apparently according to this, L 5 the one sentence that I found here, it's estimated at above

- 6 65 percent.

7 MR. MICHELSON: I don't think what you said says 8 that. What you said is that of the 5000. components or i

9 something.

10 MR. ROSSI: It is 65-70-percent;and_then_it goes 11 on to say --

12 MR. MICHELSON: But.-I want to know how many_of-the 13 breakers are being reported, not;how many components are-O l Q 14 being reported. It could be'only 10 percent of the breakers p 15 are being reported although.60' percent of the components are 16 being-reported.

17 MR._ROSSI: _

Okay. Well,-_-I can1t answer your._

18- question from the= material I have'here.

19' What I.can tell you,? however, is-the reactor trip '

20 breakers' failure.s to :open would he riade' known _ hy : the t

21 resident in almost every case.

22 MR. MICHELSON: 2 Well,;that shouldibe an=LER,_

23 shouldn't..it?- .

L 24( MR. ROSSI: -

_It might=not have to'be:an[LER under 25 the LER rule because:it Is:a singleLcomponent failure but'

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I 123 J) 1 scmething like this we would know about through-the resident 2 and reactor trip breakers are such a high visibility item-  !

L 3 that it is extremely _unlikely that a failure would have 4 occurred'over this time frame and we not be told about it.-

5 MR. MICHELSON: That was what I was hoping that 6 maybe you were going-to tell me it was nearly 100 percent

'7 reporting becausa of the importance of that component.

8 MR. WILKINS: I gather from some things that I 9 have read that Mr. Morris has written.that failures to open 10 are not the only kinds of failures.

11 MR. MORRIS: I hated to pollute the picture with 12 failure to close because it sounda-like I was looking for- '

13 any path to force them to do a life test, but'there^1s no.

(h,

(_)

I 14 denying that there would be a bonus from a life test on 15 failtres to open because you can't open the breaker until 16 you can close it and there are something likeffour times as 1? many failures to close at.all as; compared.to even delayed: ,

l 18 failures to open on the breakers and I think it islan awe-19 inspiring circumstance that we have: breakers used-in safety 20 functions that have,such'high failures toiclose rates.

21 Now what saves us I believe is that there are few 22' breakers that'are cycled as often_as the RTVs.

l 23 MR. ROSSI: In the case'ofithe reactorLtripL j 24 breaker'the onlyl safety-function.it has is a failure to open-l 25 so that's-wny we-look attfailures to_open.-

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124

) .1 MR. LEWIS: Well, failures to close tell you some 2 things.

3 MR. ROSSI: They tell'youLsome things about where 4 the problems are and they have to be fixed by the utility 5 but our concern from a safety standpoint is failure to open.

6 MR. LEWIS: Well, not really, because'if you are 7 interested in the health of the organism you should look at-8 any symptoms of abnormality because they may find their way.

9 into -- really, that's a bad philosophy.

10 In a debate, you wouldn't defend it. ,

11 Can I ask Carl's' question in a different way? Mr.

12 Berlinger quoted failures per demand as triple-01 to fcur 13 below 1 and that's a numerator divided by a-denominator 14 presumably. The numerator is the number of failures 15 observed. Where did the denominator como from?

16 MR. BERLINGER: I believe that the' denominator 17 comes from an estimate of the number ofzcycles that these 18 breakers are routinely tested,.and.it may in fact-be much 19 higher than that if the maintenance individual at the plant 20 is having'some kind ofLa problem and he may bang the breaker 21 maybe half a dozen times.

22 MR. LEWIS: If we' are talking about a minimum of-

_23 one failure, that's talking about- 100,000 tests.- Have-there 24 been 100,000 tests in which.there were no failures?:

25 MR. ROSSI:t - No, I am sure there has not been that l

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125 1 number of tests.

2 MR. LEWIS: I am sure of'it too. That's-why I am-a 3 asking the question.

4 MR. WILKINS:- Mr. Morris's second paragraph in the  ;

5 August 6th memo that was stuck in front of you recently says 6 21,000 tests, which is between ten to-the fourth and ten to X 7 the fifth, 8 MR. LEWIS: Indeed it'is.

9 MR. MORRIS: The reason for that number is'that we 10 have something like-I believe 35' plants, 7 breakers per, 12 11 type tests per year, 1 cycle per test, 4200 tests, 5 years,.

12 21,000. I 13 Of course if they bang them as often as-they.may, 14 we might have 200,000 cycles in there.- i 15 MR. ROSSI: Well, somewhere An'your material there 16 is a letter from Ashok Thadani-to Bill Russell-dated 17 November-12th, 1991 and I believe you:have it there.-

18 Apparently what that letter does is it assumed,.for one 19 case it assumed all plants participate in NPRDS and that ,

20- there are-30 breaker challenges per year.and-it took the 1.1 - total number af plants in years and so-forthtand that's -1 22 where.they got the two times ten.to=the minus fifth as a 23 failure on demand.

24 Then in'the other case what was-assumed-was-25

'25 percent of the plants participate.in-the'NPRDS and-there are O ~

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-r 126-1 10 breaker challenges per year so they were.those two point 2 calculations is what was done but he assumed in one case-25'

'3 percent participating and I think that would_have to.be low l i 4 for something like reactor trip breakers. ,

} 5 MR. LEWIS: Fine, okay.

i 6 MR. ROSSI: It's probably more like - --I'd gueas i

4 7 it's up in.the area of 80 percent at.least.

}-

Now I understand the numbers.

~

! 8 MR. LEWIS:

9 MR. MICHELSON: It'wouldn't be hard to verify what i

i 10 the real number is if you. wanted to.

11 MR. ROSSI: Oh, I'm sure we can go and getjthat.

12 We-may even have the-information in_all the paper that:ve-F 13 have with us. It's just I don't-know it right.off the top 14 of my head where to find-it.

[' 15 MR. CARROLL: No, I think that's--the point Mr.

16 Morris-made Tuesday is that the real number of: actual s

17_ breaker openings may be much higher than;this.

18 MR. MICHELSON: It was the_ participation-I was 19- -questioning, not the number.

20 MR. MORRIS: The participation in-the NPRDS I 21 plotted:a curve somewhere in there. showing something like 40 22 percent of;the_ licensees reported no: failures of any-kind,-

23' not just. failures to open.- I:put-_all the failures ~.in to get 24: a biggeridata base-andLI. normalized.itiby dividing by-the, '

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127  ;

( 1 that we didn't bias it against the people who had been in-2 operation for 20 years.

3 I had a very steep curve, hitting zero--at about 60 4 percent of the licensed PWRs who should have been reporting 5 failures and that was'the_ basis for my remark that some 6 people have been remarkably fortunate and others conversely.

7 MR. LEWIS: 'Am I correct in interpreting your; 8 position as saying that we can be fooled by these numbers 9 which show remarkable breaker reliability because 10 essentially all~these breakers are new? Is that essentially 11 your point?

12 I don't mean you in a sense but, you know,-but 13 sort of 500_tries and you're really better than_that, most-14 people seem to agree, is that the point?

15 MR. MORRIS: What I'm saying I think'is-that I am 16 the-Thomas -- I hesitate to say St. Thomas of the nuclear 17 power industry. - I am excessively skeptical perhaps, but-I 18 can think of all.the reasons I don't believe thsse reports 19 and part of that comes from reading the. sort.

20 When you focus on the failures and not on the 21 successes you see a-long list.of repairs that have_to bee 22 frequently.made to theso breakers and it is perhaps a biased 23 point-of view, but I'think itlis a conservative one as long 24 as-we don't waste recources trying to cater-tolit.-

25 I think I have' lost your question,-listening to my l

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128 1 own echoes.

1 MR. LEWIS: It's okay. I understand what's going 3 on. I'm trying to understand (a) how good the numbers are 4 we have been given about reliability. -I have come to 5 understand that. Now I am trying to understand your concern 6 because it is claimed that the data given normal skepticism 7 about whether everything is reported and all that, that 8 there are hard data, to use the original term, which sav 9 that the reliability is somewhere around ten to the -f 10 fourth or something like that per demand, and you have luat.e 11 the point that in fact with a few times a month or something 12 like that, over five years of NPRDS, these things have been 13 run 100 to 500 times were the numbers you used, and that'100 14 to 500 times is really not much if the reliability of a 15 breakar is of the order of ten to the minus four in its 16 lifetime and then you are raising the question that that 17 means -- I am interpreting you -- that there are failure 18 modes which simpty haven't-shown up yet because these are 19 generally breakers that have not b^en run very.far through 20 their life and therefore somebody ought to run some through 21 their life and find out when the life effects begin to 22 appear.

23 Is that your position?

24 MR. MORRIS: Put even more succinctly than I 25 could, yes.

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i 129 )

'() ~1 MR. LEWIS: Aw, shucks! You don't have to be nice 2 to me.  ;

3 MR. MORRIS: One always tries.

t 4 MR. SHEWMON: Bill Lindblad. -

5 MR. LINDBLAD: Mr. Morris, have you proposed this ,

6 as a research project at some time?' ,

7 MR. MORRIS: Somewhere. deep in.those tablets.of' 8 mine that you have piled somewhere before you I suggested 9 some time ago that if it was not political have the industry _

10 -test those breakers that we might do it or.we might do itLin 11 cooperation, and as I remarked then =in a deathless sen+.once, 12 it went down'an-infinite line. No response. No echo.

13 MR. LINDBLAD: I am-sorry. ,I'm a new innocent-b (ms/ 14 here. Is there a process for-the NRC' Staff.to? identify to I

15 the Research branch where.they feel there is.a need for 16 research and that is --

17 MR. MORRIS: ' I:am told there is.

18 MR. LINDBLAD:. And you followedithat?

t 19 MR. MORRIS:- I haveLnot. ILhave tried l in other 20 occasions with a: lamentable lack of success, uno doubt 21 because I don't express myself.very well..

22 MR. CARROLL: One of the things we discussed withm i

23 you on Tuesday was your notion lof what.this program would 24 be. I think you talked about maybe 30 breakers ---

25 MR. MORRIS: Right.

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e 130 1 MR. CARROLL: Why don't. you describe that a little 2 bit. That may be of interest.

3 MR.- MORRIS: Well,-having insisted-on the life  ;

4 test, at one point in the DPO resolution they_ passed that 5 hot potato to me. They said, well, if you wanted to test 6 them, how would you do it?

7 Based on my somewhat limited _ experience in tests,-

8 I said, well, I read a book on statistics and-you know what-9 a little knowledge can do to you -- I decided,-well, we'll 10 have 30 breakers because at that point the sample I 11 population, and the-sample results in the population results 12 should-be pretty _close and-we'll cycle _these breakers every 13 five or ten minutes depending--on.the1 temperature sensors to O() 14 show any overheating of coils or' motors and so.on which-15 would allow criticism later of the environment'or..the-16 -circumstances being abnornal and-thereby rendering the 17 results invalid.

18 So we put-these in a cabinet, 30 of'them, cycle =

19 them automatically, collect the data"with sensors -- the 20 ~ timing' data, and feed it'into a_ computer.and plot the number-21 of delayed trips.

22 .HR. CARROLL: Well, and also'haveia consistent-23 crew of technicians who understood-the breakers and did-24 first rate maintenance on them.

25 MR. MORRIS:- Yes. .As'I said-I.thinklon Tuesday I LO ANN JRILEY & ' ASSOCIATES, Ltd.. -

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a 131 f^)

v 1 thought that was one of the two make or break features of-2 the test, that we would have to have the right people 3 interpreting the failures and doing the repairs or anybody 4 could point at it, as I did to the poling, and find logical 5 or procedural' defects.

6 It is the heart of the program. I don't know how 7 you go about selecting a man who has all the virtues 8 required but we would have to try.

9 MR. SHEWMON: So these life tests would be with 10 maintenance done every month?

11 MR. MORRIS: Yes, sir, or on the proper numbef/ of-12 cycles that we would have to -- there we again get into the

/ 13 uncertainty of the number of cycles that are actually O

( ,) 14 experienced.

15 MR. LEWIS: Unfortunately, yua know, there are 16 certain failure modes you never pick up with the accelerated 17 life testing. In the real world things often-fail because 18 the grease jells or somebody who is supposed to re-grease it 19 finds it's too hard to get in there and he just closes it up 20 again and that you only learn, as Paul hau emphasized, there 21 is a real world out there where people do things with 22 varying degrees of cleverness and capability and the_ general 23 rule that I know of on redundant systems where you have a-24 pair of breakers or a pair of engines or something-like that

(

25 is if you keep the same maintenance crew from getting its

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l 132 I hot little hands on both element of a redul: dant system, you

[~

n s) 2 have gone a long way to assuring reliability.

3 The worst thing you could do is get a good crew in 4 to take the both apart at the same time.

5 MR. ROSSI: It might be worthwhile just describing 6 a little bit more what happened in the two breaker failures 7 that we know about. There's the one in McGuire 2 that 8 occurred in 1987 and there the breaker mechanically bound in 9 the closed position and the main cause was that they had 10 excessive lateral movement of a linkage because of a broken 11 center pole lever to pole shaft weld, so that one was a weld 12 breaking problem and it was attributed to substandard 13 welding during fabrication. They had porosity, lack of

(~N

( ,/ 14 fusion, and inadequate extent of welding. That one was not a 15 life -- it was not a wear-out kind of problem.

16 The one that recently occurred at Palo Verde, that 17 one occurred March 31st, 1992. That one was-caused'hy 18 misalignment of marking contacts, inadequate lubrication of 19 the moving contact linkage, and improper assembly of-on.

20 insulating link.

21 In the case of the Palo Verde one, Westinghouse 22 had put out information that generally told about?this 23 problem but it wh a little bit ambigaous-so Palo Verde 24 didn't follow the' Westinghouse guidance. That one also wan l l 25 not due to wear-out.=

1 l

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f 133 1 MR. LEWIS: That was maintenance.

2 MR. ROSSI: Yes, it was. bad maintenance.

3 MR.-CARROLL: And going back to Salem, Ernie, 4 didn't I just hear the other day that the Salem-people had a 5 lot of guidance on what to do with these breakers but they 6 interpreted ~it to mean that:you don't-really apply these 7 techniques.unless you have a problem.

8 Is that what I heard? ,

9 MR. ROSSI: I think we generally said that:the 10 - Salem problem wiss; due to thellack of: a lot of _ margin;in the' .

11 -undervoltage trip attachment for tripping-the breaker plus 12 that:-it had not had proper:' things:like lubrication and 13 naintenance over,the; years-and again this is myfmemory of -

14 it. - I did not go'baickland look up the details on' Salem.

15 'MP.. CARROLL

Okay.

16 MR. .SHEWMON: Charlie Rhave we'come to an end?-

~ 17 MR.:LINDBLAD:- 'Mr.. Morris, do(ILunderstand from--

18 .some of your ramarks that perhapsicruciallto thisTall is you 19 areiuncomfortable with the way:the>NRC relles on.the-20 statistics ofgreliabilityfanalysis?to: draw conclusions?- Is y 21 that perhaps at the ---you' don't-have a real1-- or.you'are1 22 -nottdirected!at the, breaker itself,in thinking that iti.is a

-23 bad break.er, but you don'tilike-the relianceion theo 7 24 reliability-statistics that have been?used?3 "Is that the

~

25 ' basis- of Jyour problem?

r i(

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1-134 1 MR.-MORRIS: I hope _you'll forgive me if I answer 2 with just one word, but truth suggests I should: yes.

3 MR. SHEWMON:_ Sometimes that-is an. excellent 4 answer.

5 MR. MICHELSON:' Has anyone made an estimate of 6 what such a test program would cost?. You know, that number 7 of breakers and time required, all that? You have to weigh 8 that a little bit in considering other safety issues that 9 need money and which one it should go to.

10 MR. ROSSI: I think that those-estimates have been 11 made but I -

12 MR. MICHELSON: Do you have any idea --

13 MR.-ROSSI:- I don't know them off the top of my.

14 head.

15 MR. MICHELSON: I mean we're. talking a=few million-16 bucks, I'm-sure,:aren'_t we?

17 MR. ROSSI: I. don't think it'_s:real inexpensive.

18 MR. MICHELSON: No, 30 breakers, getting them, 19 finding them or_ buying them -- I don't'know how you do that 20 but that alone is a substantial cost.-

21 MR. MORRIS:- You might suspect the cost.. figure of' 1

22 $2 million, as I'did,-'because.that'was my estimate, a very i

-23 crude estimate,-and three.' sources who quoted.on the-cost ~of l l

24 running-that' test.came in within $50,000 of.that.

25 'MR.-MICHELSON: But clid.-that -include ? the cost ~ of n

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135 l (vj 1 the breakers?

2 MR. MORRIS: Yes, and it did not include the 3 salvage value of the breakers because I don't see why they 4 shouldn't be as useful as they were or more useful after 5 having been cycled if they were refurbished by the 6 manufacturer.

7 MR. MICHELSON: They're proof-tested now.

8 MR. MORRIS: I mean this would be a real burn-in.

9 h. MICHELSON: But it is a couple million dollars 10 at least, then, is that right?

11 MR. MORRIS: Two million.

12 MR. MICHELSON: Thank you.

l 13 MR. LEWIS: There are other ays to do audits.

(3

, (_) 14 I'm slightly -- there are many questions but the time is l

l 15 short. There are many ways to do audits. There's the way l

16 the IRS does it. That is, one can get some estimate of the 17 reliability of the population out there in_the real-field by 18 throwing darts at a board, pulling some breaker at random 19 out of the cxisting population and taking it apart carefully

! 20 and checking all the tolerances and specs of the state of 21 the grease and, you know, just doing a real audit on a few 22 random breakers in the field. I think that would tell you 23 more than accelerated _ life testing, a personal view, because 24 there are so many things that are time-related and not 25 number of trips related that are involved in life testing.

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136-1- Maintenance is-always there and other things and a 2 certain amount of auditing would not be that expensive.

3 MR. ROSSI: Well,La lot of thatiis probably:goingj 4 on today because recognize that what is done today is-the_.

5 reactor trip breakers are periodically-tested and'they find 6 various p'roblems that are incipient to them of being 7 failures to:open and theyifix those problems'and~they: report 8 them back to the vendor and the. vendor then comes-back with ,

9 'the information based on that, so this sort of-audit of-the.

10 operation of the breakers;is occurring today. :l 11 It's probably not;as formal and systematically as 12 we as a regulatory body could= conceivably-make it but-I 13 guess our view is that it is doing the- jobi that' needs to .be__

-14 done.

15- MR. LEWIS: We are not communicating'really, 16 because I understand that that;is being. dons:on a regular-17 basis but that is just sort of' fun'ctional tests.--It's the-18 difference between an-Army physical-for1 induction =and a real -

19 physical exam, you'know, checkingEwhether:something, works-is-20 different'from taking!it apart and seeing;if it!is really-1

. +

21 working right. That's the-difference between-us.-

22 MR.-SHEWMON:: Are we readylfor~ lunch,iCharlie?=

-23 MR.-WYLIE:--Yes.--Are.thereiany other questibns;or . . .

-24 comments?_ _

25 MR.-MICHELSON: -Are weiwritingLa: letter lon_this?

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I 137 1 1

MR. WYLIE:

1 I'll have a letter.for the_ committee g 2 to.cu sider. -a a

3 -MR. ROSSI: 'When;wonld'be the -- when could we- ]

4 expect to receive = a letter?- 1 5 MR..WYLIE: Well,lit depends on when the committee-6 gets it out.

7 MR. SHEWMON: With luck,;the end of next_ week.

8 MR. CARROLL: :MiddleEof next' week. ,

9 MR..ROSSI: :Okay, so'it~is in the near term,

10. because what we would do is as soon:Las we.get your letter =

11 i and'take it..into account and'; assuming that it doesn't change 12 what.we intend to.do,.I-guee3 our-next;stepawould be since 13 this generic _ letter has~gone out-for:public commentland:-it [

14- has' been to CRGR, . it's : been; through . all of the hoops and ,

15 this is'another hoop that it'is~ going through:-today,.our q 16 intent barring anything that'you might say-in your letter- t 17 would be to send a memorandum to the; commission i telling them-18 that we intend to issueLthe'generih. letter /inisomething.--like a 19 _10 days _if the'y don't have a'probilam withlit.

20 That's;just part of ourfstandard procedure now for- -

21 generic lettern.

22 MR. SEEWMON: -Fine,-thank,you.

23 MR. WYLIE: I'think we:would(like to'thankLthe h

24 -Staff and,Mr. MorrisiforJtheir participation today.

25 MR. MORRIS:- Thank-you for[ listening.

u .

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I 138 1 MR. SHEWMONt Thank you. Okay, we'll adjourn until l 2 1800 then.

4 j 3 (Wher/aupon, at 12:03 p.m., the hearing recessed i 4 for lunch, to reconvena this same day at 1:00 p.m.) .

5

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139 1 AI'TEFNOON SESSION (1:00 p.m.)

") The first thing on the afternoon MR. SHEWMON:

3 session will be a presentation on emergency planning and

, 4 preparedness, in particular Reg. Guide 1.101.

1 5 Ernest.

1 g

6 MR. WILKINS: The item before us is to review l 7 Revision 3 of that Reg. Guide. The earlier revisions of.

1 8 that Reg. Guide refer to NUREG-0654, which contains methods l 9 that are acceptable to the staff for establishing what are l 40 called emergency action levels. I am sure the speaker will-11 define that term.

12 The purpose of Revision 3 is to provide what the 13 staff considers an acceptable alternative to the conditions 4 14 of NUREG-0654, and that's the NUMARC_ document that was- 1 15 prepared with considerable interaction between NUMARC and  ;

16 -the staff.

17 Our decision will be whether we'want to endorse 18 the position of the staff which says that this NUMARC- '

19 document provides an scceptable alternative to what is 20 already in NUREG-0654.

21- MR. SHEWMON: There was not a subcommittee 22 meeting?

23 MR. WILKINS: There=was no. subcommittee meeting..

24 MR.-CARROLL: We have.been briefed on this? '

25 MR. WILKINS;. As'a matter of. fact, we saw the-O ANN RlLEY & ASSOCIATES, Ltd.

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.- - . - . .. - .. . . w. . . - . . = . . _ .--

i 140 1 version of this before it was sent out for public comment 2 some months ago and decided at that time to- wait for the 3 public comments before we viewed the matter in; detail. ,

4 I will steal a little of their thunder and say 5 that the public comments were generally quite favorable.

]

6 There was one negative one which was sort of canceled by a 7 rather similar one in the opposite direction from a rather 8 similar type of respondent.

9 With that I will turn the podium'over to the 10 gentleman from NRR, Mr. Jangochlan.

11 MR. JAMGOCHIAN - Good afternoon. My name-is-Mike 12 Jamgochian. I'm from the-Office of~Research.

13 MR. WILKINS: Sorry. I said NRR.  :

14 MR. JAMGOCHIAN: Close enough. .

• i To my right I have a representative from NRR, Rick A6 Hassalberg. He is here to ansvar any difficult.questiens 17 you may have. .I handle the easy questions.-

18 We plan on going-through the slides that have been i 19 passed out to you;one at.a t'ime. #

20 (Slide.)  ;

21 MR. JAMGOCHIAN: The-purpose of the meeting is to 22 publish Revision'3 of Regulatory Guide 1.101, which endorses 23 an alternative emergency action level scheme to Appendix 1.

24 of NUREG-0654.

25 To give you ajlittle bit of' background, Reg.' Guide ANN lRILEY & ASSOCIATES, Ltd.

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141  ;

1 1.101 was published approximately 1980 and has always 3

2 provided an amplification of the emergency planning- i

} i j 3 regulations contained in 10 CFR Part 5047 and Appendix E.

i 4 It currently endorses IWREG-0654.

5 Appendix E requires an emergency action level 1

6 scheme for the classifications of emergencies. There are l

f 7 currently four classifications of emergencies unusual-8 event, notification of unusual event, alerts, site area i

l 9 emergency and general emergency, the last being fhe most-10 severe.

j 11 (Slide.]

12 MR. JAMGoCHIAN: The-third slide provides the 13 background. When NUREG-0654 was published it listed all of -

14 the emergency planning standards-that are codified in 10 CFR-15 5047'with a listing of how to meet those standards.

-16 Also in 0654 Jt gave example initiating conditions 17 for each of the emergency classifications. If'I'might read 18 fp m the Code of Federal Regulations:-

19 Emergency action'. levels' based not only on onsite 20 and offsite radiation monitoring information, but also on 21 readings from a number of sensors that-3-dicate a' potential 22 emergency such as pressure in containmerJ., c ad the responee 23 of the. emergency-core cooling system for-notification of-  ;

24 offsite agenciesLshall--be described.

25- -So licensee has to set up a scheme of' emergency

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l 142

() 2 I action levels upon which he can then judge the type of emergencies that may exist.. This scheme is also discussed l 3 with state and local government so that everybody knows what ,

4 they are declaring from.

5 The development of an alternative emergency action 6 level scheme is an industry initiative which reflects 11 7 years of experience.

• 8 This scheme was developed by NUMARC with  ;

9 significant input and interaction from NRC. FEMA was also 10 involved in this development.

t 11 It was sent to the ACRS in late-1991 but the ACRS 12 deferred review until1after the public comment.

13 We received ten public comment letters and they 14 were evaluated. Four were from state governmental entities; 15 six were from utilities or companies working with utilities.

16 -All of the comment letters; fundamentally agreed with the EAL 17 scheme and endorsed that 2AL scheme, except for cne._ As was

. 18 pointed out, that one was sort of counterbalanced by another 19 commenter that thought just the opposite.

20 Also, the' staff made a presentation of this Reg.

21 Guide.to a meeting in Florida:in May to-the State-Radiation 22 Control Program Directors meeting. That generally met with 23 favorable response'from ht'e states.

24_ The staff-receivedfappJoximately 40 to_50 phone 25 calls from state governmental entities as well. We were-() -

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143 1 sort of disappointed that we didn't get follow-up letters, 1

2 but most of the verbal correspondence on the phone was 3 explanation or amplification as to what we were doing, and 4 most were very positive and felt that this was a step in the 5 right direction.

6 For slide 4 I will turn to Rick so that he can go 7 into t'a more technical details.

8 [ Slide.)

9 HR. HASSELBERG: Good afternoun. My name is Rick 10 Hasselberg. I am the Project Manager for the Ap-600 11 advanced reactor but in a previous life I was a member of l

.1 12 the technical review group for the NRC which reviewed the 13 emergency act @h ~

O ome called NUMARC NESP-007.

V 14 One vrtt tv r bt Mckground before I get into the 15 bullets en slide 4. As Mike already mentioned, the industry l

16 and the NRC have 32 3r 12 years of experience working with 17 Appendix 1 to NUREG-0654.

18 It is a common misconception that the tables in 19 the back of Appendix 1 give example emergency action levels.

20 They do not. They give initiating conditions for which 21 emergency action levels should be developed.

22 Some licensees over the last 12 years or so have l 23 developed very sophisticated emergency action level schemes 24 in response to the requirements of NUREG-0654.

l 25 Some have incorporated mode applicability; fission 1

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144 1 product barrier challenge / breach schemes. Some have linked 2 their emergency action level schemes with entry into their 3 symptom-based emergency operating procedures. Some-4 licensees, on the other hand, have stuck almost verbatim to 5 the wording of the example initiating conditions in 0654, 6 Appendix 1. So there is a wide variety of emergency action 7 level schemes out there in the industry at the moment, 8 As Mike mentioned, this is an industry initiative 9 and the staff has had considerable interaction with NUMARC.

10 We are quite pleased to say that-NUMARC has incorporated all 11 of our commerits and concerns in Revision 2 of the-document.

12 The methodology was also tested in several 13 different situations against accident-scenarios, exercise 14 scenarios, against currently approved BWR/PWR emergency 15 action level schemes and compared point by point. It was-an 16 in-depth study against the guidance of NUREG-0654, Appendix 17 1.

18 We the staff have concluded that NESP-007 provides 19 example initiating conditions that are an acceptable 20 alternative methodology to.the example initiating conditions 21 listed in Appendix 1 of NUREG-0654.

22 Additionally, we~fcel that the.NUMARC guidance 23 document provides these example EALs for each of thesa 24 . initiating conditions and.we think that those are very 25 good.

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i i 145 1 Of course the NUMARC has not redefined unusual 2 event or alert in any way. We have the same classification  ;

i 3 levels that we have always had.

4 (Slide.)- l i 5 MR. HASSELBERG: Similar to a plant. technical i

! 6 specification document, you will see that the NUMARC scheme j 7 has example EALs and also provides a technica1' basis and i

8 defines an operational mode applicability for when these ,

9 particular emergency action levels and initiating conditions f

l 10 are applicable.

11 Similar to an owners group emergency operations

12 procedure guidelines manual, each example emergency action.  !

13 level in the NUMARC methodology has'one or more recommended

! 14 threshold values for BWR or PWR1 facilities..

i 15 One of the features that I particularly like as a -

16 member of the reviewing group is that NUMARC methodology l 17 clearly defines what a fission product barrier' loss is and-18 what a potential loss is.

i 19 You may recall that'in NUREG-0654,-Appendix 1 one 20 of the definitions for a general emergency is the loss of j: 21 two of the-three_ fission product barriers with a potential

[ 22 loss for the third, but there was no definition-for what a ,

23 -loss or potential-loss was. The'NUMARC document clarifies 24 .that.and I would-be; happy to go over that with you in.as 25- much detail as you think jou would-like. -

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! 146 ,

i 1 (Slide.)

! 2 MR. HASSELBERG The'NUMARC scheme-presents EAL 1

3 thresholds that utilize many of the same observable and  :

i i 4 quantifiable parameters that the operators now use in their ,

! 5 symptom-based emergency operating procedures: temperatures, l 6 pressures, levels, flow rates, subcooling margin, radiation

! 7 levels, isolation-system status, status of the critical 8 safety functions. The same thing that they use in ,

! 9 responding to other emergencies. .;

10 Another of the features that I particularly like--

11 is that the NUMARC document clearly limits the delay time l ,

. 12 for the recognit'on of failed mitigation efforts. In other ,

! 13 words, how long do you wait before you come to the 14 conclusion that the reactor vessel is-not refilling or that. l l 15 the diesels are'not developing their rated speed and l 16 voltage? How long do you decide to wait before you declare 17 that a fire is not under control? Those-kinds of things are ,

18 clearly laid out in this guidance document.-

t 19 -{ Slide.] '

\ . . .

20 MR. HASSELBERG: I have two little introduction i-

! 21 slides here-to get us'into the technical discussion.

'. 22 If I was to trv-to define the NUMARC EAL-23 methodology, I would say that what we have in front of us is-24: an event-based emergency, action level classification system 25 that incorporates a fission product barrier breach scheme b .

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-147 1 but is not limited to a fission product barrier breach 2 scheme.

3 There are four different ways or four different 4 recognition categories that NUMARC proposes. You could come 5 to the conclusion that you have an emergency through 6 abnormal radiological conditions; hazardous conditions such 7 as fires, floods, earthquakes, chemical spills, security 8 events; systems important to' safety malfunctioning;.and of 9 course the fission product barrier challenge / breach scheme.-

10 (Slide.)

11 MR. HASSELE3RG: Having divided these emergencies 12 into four prefixes that they call recognition categories, 13 they then matrix these recognition categories with the four 14 emergency event classifications and_you end up with 16 what 15 NUMARC calls generic identifiers. I list 11 of them on this 16 example:

17 Generic identifier AU would stand for an unusual 18 event based on abnormal radiological conditions.

19 HA is just another example. It would be an alert 20 based on hazardous conditions..

21 _SG is a general emergency-based on system 22 malfunctions.

23 Both NUMARC and the NRC agree that licensees 24 proposing to change their ' approved- EAL schemes to . the NUMARC .

25 . methodology would not.be required to reproduce these 16- -

Q '

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4 148

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generic identifiers. If they have a system which they are more comfortable with, that would be fine. But they will 3 bave to demonstrate to the NRC staff's satisfaction that 4 their proposed EAL scheme incorporates all of the NUMAltC 5 example initiating conditions and that their propostd EAL 6 threshold values have a reasonable technical basis 7 comparable to those in the NUMARC methodology.

8 Any questions, gentlemen?

9 MR. LINDBLAD: Lot me-just aske an observation.

10 You speak-of these levels being event based. Is that to 11 distinguish from condition based, or does an event have'to 12 be a precipitous action?

13 MR. HASSELBERG: For example, you would call a 14 loss of coolant accident event based. That's your-15 conclusion. But as we have learned in our emergency 16 operation procedure development, you don't get a LOCA alarm 17 on the main control board. You have to come to the 18 conclusion that you have a loss of coolant accident through 19 cartain symptoms. It could be containment pressure; it-20 could be-loss of pressure in the reactor. coolant system; it.

21 could be excessive makeup capability.. _These are the-22 symptoms of a loss of coolant accident..

23 MR. LINDBLAD: A condition such as'a discovered 24 ' lack of equipment qualification doesn't: generate any of 25- these?.

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149 o  ;

1 MR. HASSELBERG No , sir. That would fall under 2 either a technical specification or somo non-emergency 3 reporting event.

l 4 MR. LINDBLAD: But that's an event, not a i 5 condition?  ;

a 6 MR. HASSELDERG I'm strictly limiting my- l 7 presentation today to emergency classifications.  :

8 MR. LINDBLADt Thank you.

9 'MR. WILKINSt- Let me make an observation and ask 10 one of the.other gentlemen a question. In your regulatory f 11 analysis portion of this you sometimes say the.NRC staff'  :

12 agrees with NUMARC and in other. cases you.say NUMARC I 13 addressed this issue and'in some. cases you say "we concur." r 14 Are those all the-same things or-are there some subtle 15 differences that I don't understand?~

16 MR. JAMGOCHIAN In.all-cases'NUMARC and the NRC i 17 are in agreement. Fundamentally this regulatory analysis = .

18 has been under developmentfor quite sometime~- When NUMARC p

19 developed their first' document there were-many_ areas that 20 the staff did not agree with them and took exceptions. At-  ;

e  :

that time NUMARC either modified-their' document or explained-21 p 22 their rationale to the staff. .The end product'is, though',_

23 that in'all_ cases'the NUMARC document has addressed the -

ll 24 concerns _ c,f the staff.

l

.25 MR.'WILKINS: And;the varying languagefthat'is O ANN RILEY & ASSOCIATES, Ltd.-

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used is irrelevant. They all mean the same thing.

MR. JAMGOCHIAN: Yes, sir.

3 MR. WILKINS: Which says that you agree.

4 MR. JAMGOCHIAN: Yes, sir.

5 MR. WILKINS: I might tell the Committee that I 6 have read this document, which is quite a document. If I 7 have one cor. ment, it is that I wish it had been shorter, but 8 I think that is a standard comment that many of us have made 9 in a lot of other circumstances.

10 It appears to me that the staff has identified a l I

11 great many areas in which the NUREG-0654 has been altered a 12 little bit. As the gentleman has just said, staff is 13 satisfied that in every case the situation is well in hand.

/~

km ,T) 14 I was unable to find any of these areas in which I 15 had any radical disagreement with that conclusion.

16 MR. SHEWMON: Are We going to hear from anybody 17 else or is that it?

Io MR. WILKINS: The only things that are on the 19 agenda that was presented to me were the presentation of Mr.

20 Jamgochian and the presentation of Mr. Erickson.

21 Was NUMARC planning to-make a presentation or just 22 here to answer questions?

23 MR. NELSON: My name is Alan Nelst'n. I'm a Senior 24 Project Manager with NUMARC.

25 In the process that has unfolded over the years we O ANN RILEY & ASSOCIATES, Ltd.

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have worked with the NRC almost weekly over the last couple of years. We are here to support their endorsement of our 3 document. We felt it was appropriate in working together 4 that the NRC staff present it to the ACRS.

5 I think it van in March of 1989 that NUMARC came 6 in and presented a preliminary review to ae subcommittee 7 where we were encouraged to pursue our efforts. This is a 8 culmination of all that work.

9 MR. SHERMON: Let me bring up one question which 10 is certainly irrelevant for now but intrigues me. The 11 Europeans have accepted another way of rating levels of 12 emergencies. I guess I read somewhere that the U.S. NRC, if 13 they have not decided to accept this way of ranking things,

(~h)

(, 14 then they are seriously considering it.

15 MR. CARROLL: Using it on a trial basis.

16 MR. SHEWMON: What does that mean for this? Is it 17 totally independent of this? Does it n.ean it all has to be 18 rewritten with Revision 5 next year? Or what?

19 MR. HASSELBERG: Was the question to the staff, 20 sir? ,

21 MR. SHEWMON: Yes.

22 MR. WEISS: My name is Eric Weiss. I'm with AEOD.

23 It is my understanding that the staff put forward a 24 Commission paper not long ago that recommended to the 25 Commission that we not adopt the international event O

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severity scale in this country. We have been participating in the development of that scale but for a variety of 3 reasons we plan to continue to use our existing-four levels 4 of emergency classification rather than adopt the 5 international event severity scale with seven.

6 If I am not mistaken, I believe there is under 7 consideration a proposal to rank events in this country at 8 alert or higher for the purpose of communicating with the 9 European community but it would not be done by licensees; it 10 would be done by senior NRC staff after the event had 11 occurred: and it would be for the-purpose of keeping-12 communications open with the European community rather-than 13 to be used in this country;for emergency response.

() 14 our system works very well. It has worked for a 15 number of years. There would be'a lot of negative 16 repercussions, I think, associated with changing at this.

17 time.

18 MR CARROLL:. We are going to have a briefing on.

19 this in the next couple of months on the recent SECY paper 20 the staff put forward.

21 MR. SHEWMON: Fine. That's all I had.

. 22 Will we: write a~1etter on this? At-least a 23 Fraley-gram or something?-

24 MR. - WILKINS : : It would.be a letter.to EDO, I:

25 believe.-

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l 151 1 MR. SHEWMON: You will draft a note on this?

2 MR. WILKINS: I an in the process of drafting it.

3 I have not heard any comments from my colleagues that would 4 indicate that we should not endorse the position of the 5 staff.

6 MR. CARROLLt You haven't given us a chance yet.

7 MR. WILKINS: I just sat back to see who was going 8 to ask questions.

9 MR. SHEWMON: You aren't usually shy, Jay.

10 Any other comments?

11 MR. WILKINS: Give him a chance.

12 (Laughter.)

13 MR. CARROLLt I too have read the final version of 14 this. I commend NUMARC-and the utility people who worked on 15 it and the staff. I think it's a real step forward. There 16 have been a lot of problems with thees EALs in the past.

17 Alan, how extensively do you-feel this'is going to la be embraced by the utility industry? Is everybody going to 19 go along with this? Or ten percent of the' utilities?

20 MR. NELSON: We are running a workshop on the-21 implementation of the NUMARC methodology with the support of 22 the NRC. That workshop will_be in:St. Louis September 22 23 and 23. We expect a substantial attendance of approximately 24 200_' people _from the industry as well as NRC regional folks-25 that are interested-in coming, n

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154 1 In answer to your question, a hand waving of those 2 that might be interested in moving toward this early on is 3 showing somewhere in excess, I would say, of 50 percent.

4 That would be in the first couple of years as the process 5 moves along.

6 MR. WILKINS: I might have commented earlier that 7 any utility could have done this without the Reg. Guide 8 having been revised. Then they.would have had to convince 9 the staff on an individual basis that that approach was 10 satisfactory. This revision does it wholesale.

11 MR. NELSON: What we wanted to ensure was a 12 consistency in the determination-of an event so that from 13 one region to another or within regions the incident was 14 classified consistently. I think that is one thing we have 15 done.

16 We at NUMARC have agreed that we will maintain 17 this document as a living document. That means to say that 18 as we gain _ experience with it through the years we will be 19 coming back to the staff and working with them to revise it 20 and keep current with the operating experiences we.know' 21 today. We think that is a proactive step forward and 22 welcome the opportunity to do it.

23 MR. CARROLL:- 'I think the staff referred to using 24 the guidance _in some exercises.and some other ways of-25 testing whether it really works. Can you amplify on_that a-L O ANN RILEY & _ ASSOCIATES, Ltd.

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155 1 bit?

i 2 MR. NELSON: We had a series of our task force

3 members. Two of them are here, Roy Brosi and Steve Lavie, a

who had a major part in the participation in that

) .

4 5 industry /NRC review. What we did was run eight table-top 6 scenarios for PWRs on one day and eight PWR/BWR th9 next

~

7 day. We did it right around the corner from here, at-the f

8 Maryland Bank Building. j

i 9 MR. CARROLL

This was at the NRC's Incident -l 10 Response?

11 MR. NELSON: That's_-right.:

12 We walked down that.particular_ scenario-for each 13 of the different types of_ events, whether-it was-a site, a 14 rad event, a hazard event,:or a fission product barrier.

15 evant, to test m.are the NUMARC-methodology lined up with i

16 Appendix 1 to see if there were correlations and if.there 17 were.any incidents that may be missing from'one'or the other 18 and how in relationship'would we' classify;that as an unusual 19 event, an alert, site nd general emergency.

20 I guess _thos'e 16 scenarios-in itself indicated to 21 us,that we were going in-the_right direction. i Yee- .. we 22 needed to make some improvements, but overall we-were. pretty.

much in step with~where we needed.to go.,

23 -

That was pretty much the__V&V,: so to speak,_'as we

~

24

~

- _ :j 25 performed it that gave us the confidence we were in the ANN ' RiLEY & - ASSOCIATES, Ltd.~

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k 156 1 right direction.

2 Steve reminded me that we did do a pilot with PP&L 3 at Susquehanna. They had developed a BWR NUMARC package.

4 We sat down with the staff and they evaluated that.- As a 5 lessons learned they are now incorporating those lessons and 6 at some point in time would make maybe the first submittal.

7 We plan hgain to do a PWR table-top V&V type as a l

8 pilot with Duquesne Light at Beaver Valley in the near 9 future. i 10 People are getting hands on. We were 11 incorporating those as we had' developed it and it brings um 12 to where we are today.

13 MR. CARROLL I know-the emergency planning folks ,

14 write these things and they understand them. They know what 15 they mean. But in the final analysis the guy who has to 16 decide what he has got is some poor shift supervisor in the 17 middle of the night. How have you made-sure-that you have- 1

- 18 human factored thin thing for the shift supervisor?-

c 19 MR. NELSON: The task force members themselves are 20 rade up of shift supervisors, STAS'and the-complement. This-21 package was' developed for and_has been tested-with the-22 operators that run-the plant. We found.from the feedback we 23 received from them that it was user'. friendly for the 24 - operators; we learned from their experiencer: nit'has been >

25 developed in a method that they:like and understand.

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We went to them first. We did ten interviews at five PWRs and five BWRs, two from each region, and 3 interviewed operators and requal trainers alike to find out 4 their experience from 0654 and what could we take from that 5 to build into a new methodology. So we started from a basis 6 from the operators, pulled it together and tested it with 7 operators, and now the final document.

8 MR. SHEWMON: Bill.

9 MR. LINDBLAD: What does the industry think Will 10 be the impact on the total number of events reported? Does 11 the industry t.olieve that the number of events reported will 12 go up or go down?

13 MR. NELSON: The document itself does clarify a 14 number of the actions that thu -perator would take in 15 declaring it. There would be substantially fewer unusual 16 events. We feel justifiably so. In that particular case wo 17 might see less unusual events and less alerts but we have 18 justified them within the document.

19 MR. JAMGOCHIAN: When the staff published this 20 proposed guido wo focused on that one point quite 21 significantly throughout the document and in our forwarding 22 letter to state and local governments that indicated that 23 the staff believes that there would be substantially less 24 notification of unusual events. Nonetheless, notification 25 of state and local governments and the NRC under 5072 would p

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158 1 continue but not in the realm of emergencies as such.

2 MR. HASSELBERG: I have a list of the unusual 3 events here from 0654, Appendix 1. What NUMARC proposed to 4 do and what the staff endorses is that if an unusual event 1

5 under 0654, Appendix 1 is also listed in the Code of Federal 6 Regu'ations as a defined non-emergency, there is a j 7 diff, ence in classification. Either it is an emergency or 8 it isn't an emergency.

9 They recommended, for example, that a contaminated 10 man being transported to the local hospital did not need a 11 national federal notification. There is still a one-hour 12 non-emergency requirement for notification. State and local 33 officials may not be notified unless they choose to be.

It That would be worked out individually between the state and 15 the utility.

16 Likewise, a power plant shutdown required by tech 17 specs performed within the time period allowed by tech specs 18 is a non-emergency. A one-hour report is still made but it 19 does not need to be an unusual event. We agree with NUMARC.

20 That's where you get a lot of your unusual events.

21 MR. NELSON: They all fall into that category.

22 I would like to add that one of the things that we 23 looked at closely which still pertains to both 0654 and 24 NUMARC in that the offsiae response stays the same. They l 25 will'do the same if they are notified of an unusual event, O ANN RILEY & ASSOCIATES, Ltd.

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] 159 1 alert and site general emergency. So their reaction to l 2 NUMARC is exactly the same as what they would do today.

! 3 MR. SHEWMON: That brings us to the end of this i I- 4 discussion. We will write a note. We thank the staff.and

! 5 NUMARC for coming in and working with.us en this.

f 6 I suggest that we take a seventh inning stretch in 4

i

! 7 place if you want-to, but essentially'we will go on in a l 8 couple of minutes, as soon as:the room clears, to reading ,

e

.i 9 letters. The first one on the list.is the severe. accident- -

l 1

. 10 program. ,

i 11 (Whereupon at 12 35 p.m. the - recorded portitm of D

l 12 the meeting was adjourned.)

I ou 13 *

, 15 16 17 .-

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• 18 ,

19 20 21 22

, 23

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1 b)

(. REFORTIR'S CdRTIFICATE i

This is to certify that the attached proceed-ings before the United States Nuclear

' Regulatory Commission in the matter of NAME OF PROCEEDING 388th ACRS Meeting DOCKET NUMBER:

PLACE OF PROCEEDING: Bethesda, Maryland were held as herein appears, and that this is the original era script t h e r e o f f o t- the file of the United States Nuclear Regulatory Commission taken by me and thereafter reduced to typewriting by se or under the direction of the court report-ing company, and that the transcript is a true and accurate record of the foregoing proceedings.

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R Q)

l 1 ($3) f REPORTER'S CERTIFICATE This is to certify that the attached proceedings before the United States Nuclear Regulatory Commission in the matter of NAME OF PROCEEDING: 388th ACRS Meeting DOCKET NUMBER:

PLACE OF PROCEEDING: Bethesda, Maryland were held as herein appears, and that this is the original transcript thereof for the file.of the-United States Nuclear Regulatory Commission taken-by me and thereafter reduced to typewriting by me or under the direction of the court reporting company, and that the transcript.is a true and ,

O' accurate record of.the foregoing proceedings.

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h V

ABWR DESIGN CERTIFICATION 8/5/92 ACRS SUBCOMMITTEE REVIIW

() ABWR DESIGN CERTIFICATION R/5/92 ACBl_$UBCOMMITTEE REVIEW REACTOR BUILDING COOLING  :

SUMMARY

OF TIER 1 WATER SYSTEM (RCW) ENTRIES DESIGN DESCRIPTION ENTRIES:

- SAFETY AND NONSAFETY

- THREE INDEPENDENT COOLING LOOPS

- TRANSFERS HEAT FROM SAFETY /NDNSAFETY EQUIPMENT TO UHS

- SAFETY PORTIONS ARE SEISMIC I - QUALITY GROUP C

([) - RCW FLOW INCREASES FOR POST-LOCA HEAT REMOVAL FROM RHR AND D/G SAFETY AND NON-SAFETY EQUIPMENT CAN BE ISOLATED FOR ,

POST-LOCA

- SIMPLIFIED SYSTEM DIAGRAMS (3)

ITAAC ENTRIES:

- BASIC SYSTEM CONFIGURATION RCW LOOP INDEPENDFNCY MYDRAULIC CAPABILITY-SAFETY AND NONSAFETY. ISOLATION AUTOMATIC START OF 2ND PUMP AND 3RD HX AFTER-LOCA O RDR-1 8/5/92-

ABWR nesign occument 2.11.3 Pleactor Building Cooling Water System .

Design Description The Reactor Building Cooling Water (RCW System distributes cooling water during niious plant operating modes, as v i as ditring sht.tdown, and during post LOCA operation of the various safety .n ns. The sptem removes heat from plant auxiliaries and tmnsfers it to the , ate Heat Sink (LHS) sia the Reactor Senice Water (RSW) System. The RCW System removes heat from the i ECCS equipment including the emergency diesel generators during a safe reactor chutdown cooling function.

I The ROV system is designed to perform its required safe reactor shutdown j

cooling function following a postulated loss of coolant accident / loss-of offsite power (LOCA/ LOOP), assiiming a single active failure in any mechanical or l elecuical RCW subsystem or RCW support system. In case of a failure which . l disables any one of the three RCW dhbions, the other two divisions meet plant safe shutdown requirements, including a LOCA or a LOOP, or both Redundant isolation valves are able to separate the essential ponions of the RCW ,

cooled components from the nonsafety-related RCW cooled components during l a LOCA, to assure the integrity and safety functions of the safety-related parts of the system. The isolation valves to the non essendal RCW System are automatically or remote-manually operated, and their posidons are indicated in Q the main control room.

Each RCW dhnion includes two pumps which circtdate RCW through the; various equipment cooled by the_ RCW System and through three heat exchangers which transfer the RCW heat to the UHS via the RSW System.

Each RCW dhhion Main Control Room (MCR) instrument indication includes ,

main loop surge tank level, main loop radiation and RHR HX Gow and ,

temperature. MCR controlincludes all MOVs and AOVs show on Figure 2.11.3.

Nonnal surge tank MUWP makeup is automatic or MCR controUed.

The three RCW train conGgurations are shown on Figure 2.11.3. The RCW

' System provides three similar complete trains ( A, B and C) which are mechanically and electrically separated, The RCW pumps and valves for each RCW dhistor. are supplied electrical power from a different dhsion of the ESF power system.

The RCW ASME Code classifications for different portions of the system are.

indicated on Figures 2.hl.Sa-c.The safety-related portions of the RCW dhnions are designed to Seismic Category I and Quality Group C, and are heated Seismic l - Category I structures.

Q_

1

. 6/1/92 2.11.3 l

, ,..--x.,.

, , , . s. . , -- .-.. .. _ - :. - _; - -,,,-,.-w ., a , .

ABWR oesign Document During various plant operating modes, one ROY water pump and two heat exchangers are normally operating in each division. Flow balancing provisions O are included within each ROY disision.

Pump design parameters are:

RCW A/B RCW C Design pressure (psig) 200 200 Design temperature ('F) 158 158 Discharge flow rate (gpm/ pump) 2 5,700 2 4,800 Pump total head (psig) 2 80 2 75 Heat exchanger capacities are each: 2 45E6Btu /h 2 42E6Btu /h Connections to a radiation monitor are provided in each division to detect radioactive contamination resulting from a tube leak in one of the RHR exchangers, fuel pool exchangers, or other exchangers.

The RCW pumps and heat exchangers are located in the lower floors of the control building. The equipment cooled by the RQV divisions are located in the reactor building, turbine building, and radwaste building, (Figures 2.11.3a<).

Tables 2.11.3b, c, d show which equipment receives RCW flow dwing various plant operating and emergency modes. The tables also indicate how many heat exchangers are in senice in each mode.

O v During nonnal plant operation, RQV flows through equipment which is normally operating and requires cooling and all ECCS equipment, except RHR heat exchangers and ESF diesel generators, as shown by open or closed valves in .

Figure 2.11.3.

If a LOCA occurs, a second RCW pump and third heat exchanger in ear.h loop are placed in senice. Automatic or remote operated i:,olation valves will separate the RCW for the LOCA required safety equipment from the nonsafety-related equipment, if a RQV surge tank low water level signal occurs. The primary containment ROV isolation valves automatically close if a LOCA occurs.

O 2 6/1/92 2.11.3

ABWR oestan occument Anci a LOCA, the following sequence will be followed:

O m u ow eenaewr1iee a o>e RCw Srsiem u avawe ie ihe instrt. ment air /senice ali (LA/SA) compressors, the CRD pumps and CL*W pumps, RCW flow to these nonsafety components is maintained (Figure 2.11.3). Flow is automatically shutoft to other non essential equipment after the LOCA.

(2) If the operator determines after the LOCA from essential RCW instrumentation, that the integrity of the non safety RCW System to the above mentioned compressors and pumps has been lost, he can shut the re. note operated non essentialisolation valves shown in Figure 2.I1.3.

If the surge tank water level reaches a lowlevel, with or without LOCA, indicating -

loss of water out of the RCW System, isolation valves in the supply and return piping to the non essential equipment will automatically close, including the compressors and pumps mentioned abovt Without a LOCA and with low surge tank standpipe water level, all running RCW pumps trip. For post LOCA, both _

RCW pumps continue running with low' surge tank strndpipe water level.

The RCW/RSW heat exchanger design basis condition occurs during post.

LOCA cooling of the containment via the RHR heat exchangers.

The RCW pumps have the flow capacity.to deliver required flow to the ECCS equipment in each division and the above mentioned compressors and pumps if the isolation valves cannot be closed.

After a LOOP, the RCW pumps isolation valves and their controllogic are .

automatically powered by the emergency diesel generators. I A separate surge tank is provided for each RCW division. Normal makeup water source to the surge tank is the Makeup Demineralized Water (MUWP) System.

For LOCA conditions, the Suppression Pool Cleanup (SPCU) System provides a- _

backup surge tank water supply.

Inspections, Tests, Analyses and Acceptance Criteria Table 2.11.Sa provides a definition of the inspections, tests, and/or analyses together with associated acceptance criteria which will be and undenaken for the RCW System.

.O 3 6/1/92 2.11.3

.w~ .

'O O O

"", Table 2.11.3a: Reactor Building Cou.it,g Water (RCW) System i.,

insper tions, Tests, Analyses and Acceptance Criteria inspectiens. Tests. Anslyses Acceptance Criteria Certified Design Commitment

1. Inspection of construction records will be 1. The system configuration conforms with

1. System configuration, including key Figure 2.11.3.

1 components and flow paths,is shown in perforrr:3d. Visual inspection (VI) will be Fegere 2.11.3. performed based on Figure 2.11.3.

2 ' Tes's and VI of the three independent trains 2. Piant tests and VI cot: rm proper

' 2. Three RCW trains are machanically and independence of three RCW divisions.

electrically independent, will be conducted which wiliinclude Independent and coincident operation of the three trains to damonstrate complete

. d4?sional separation.

3. Limited system hydreelic tests w?!! be 3. The results ccnfirm that the RCW has the?

3. ,During various modos of operation,the water flow capability specified by the

' RCW System has adequate hydraulic. . conducted accorafing to avaihebia nonnuclear host plant conditions.The t> *s certified design commitment, including

! capability for plant auxiliaries and the .. safe shutdowa operation with 1 RCW .

will demonstrate a safe plant shutdow primary containment required for safe division out of service.

E shutdown following a design accident or -- with one RCW division out of service.

transient.These safe shutdown requirements are satisfied with only .any 2

- of 3 RCW divisions operating.

4. VI of theinstelled RCW System and RCW 4. Isolation velves are proper!y located Js ,

-_ I. -. isolation valves as shown in Figure 2.11.3 shown in Fp~2.11.3 and are -

H can automaticsily or remote manua!!y . preopera'ional tests as follows will be corppleted:' demonstrated to operate eutt,mst;cally or -

- sepsr:Ae the RCW forthe essential remote manually to isolate RCW foe non-

- efiuipment from the RCW for the none essential from RCW for essential H essendal equipment.

a. Remote-menual operetion of the

Isolation valves from the mein control- equipment cooled by the RCW System.

room.

J~

' b. . During simulated LOCA conditions, a simusted LOCA condition will be - 1 combined with a simulated RCW surge tank water level signal to automatically close the isolation volves.

- l i -

.j {

c. . . A LOCA signal will shut RCW isolatio" i

Gi

' va.ves which will shut off RCW flow to .

- all non-essential equipment except the IA/SA cornpressors, CRO pumps and CUW pumps

~ "~ -

t :__,. u w_ < g_ -

p g. g o J V l

- a' Table 2.11.3a: Reactor Building Cooling Water (RCW) System (Continued)

Inspections. Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests. Analyses Accepta nce Criteria

5. Without LOCA ud with low surga tank 5. RCW System preoperationa! tests will be 5. The RCW pumps will trip or operate as s!smtpipe water ;evel, both RCW pump" in performed as follows: f >llows:

that division trip. For post LOCA, both RLN pumps will oprrate with low surga tank a. Simulate s surge tank standpipe low s. The running pump (s) will trip on surge standpipe water levei. water levelin the standpipe and tank standpipe low water leval.

confirm the running pump (s) trip.

b. With a LOCA condition signal, both

b. During a simulated LOCA condition RCW pumps will continue to operate and a simulated surge tar'k standpipe with a simulated surge tank standpipe law weist level signal, confirm that low water level signal.

s both RCW pumps will operate.

c. Both RCW pumps start on simulated

c. During low surge tank standp*pe water LOCA signal.

level condition, a simulated LOCA Y' signal starts both divisional RCW pumps.

G. A LOCA will rasult in she automatic start of 6. Tests simulating LOCA/ LOOP conditions 6. . LOCA/ LOOP signal successfully starts the second RCW pump in each division and w!Il be conducted for the RCW System second RCW pump and initiates RCW/RSW

. start !!ow through the third RCW/RSW H: which confirm the RCW and its support Hx flow in each division including the in each division. systems will perform its function under following confirmations-thou conditions. Tests will be conducted During LOCA/ LOOP (loss-of-coolant ' for the RCW, which confirm that after the . a. Regardless of which RCW pump was accident / loss of off-site power) conditions, LOOP, each d' vision of RCW pumps r i operating during normal operation RCW pumps and v61ves are powered by ' veives operates with the same divish - before the LOCA, after the LOACl LOOP the ernergency diesel generators (D/Gi. emergency D/G power and associateC simulation occurs, the first and second control power sources. RCW pump will s+ert automatically, powered by the emergency diesel generator.

b. Regardless of which two RCW/RSW

< Hx's were operating before the t.OCA.

e after the LOCA/ LOOF occurs. the RCW

$ motor-operated valve on the third Hx discharge will open automatically.

i

ABWR Design Document i

Table 2.11.3b: Reactor Building Cooling Water Consumers O oivision i Hot Em-rgency Operating Newmel '

Shutdown Shutdown Standry (LvCA)

Mode / Operating at 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> **" Y (Supp ession et 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (no loss of Components Cont'ition' AC)

""*'^ Pool et 97'C)

RCW.RSW Heat 2 3 3 2 3 3 Exchangers In Service ESSENTIAL W Emergency Die. - .. . -. X X sel Generator A RHR Heat .. X X - X X Eichanger A FPC Heat X X X X X X Exchanger A Others (essen. X X X X- X tist)(2, NON- ESSENTIAL RWCU Het X X X -X .( -

Ex: banger OC/ (naids Drywell* X X X Y, X ,.

Others (nun- X X X X X X essentist)'0 .

(1) (X) . Equiperwent receives RCW in this rrcde

(-) Equiptr>ent oc es not recoise RCW in 'his mode, (2)- HECW refrigereter, rc,om coolere (FPC oump, RHR, RCIC, SGTS, FCS, CAMS), RHR moter and seal .

I costers (3) Drywell(A & C) and WP coolera, (4) Ins ruments and servee air coolem: RWCU pornp cookr, CRD pump uit, and R(P Mg eets, y

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i APWR ossign Document J

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Table 2.11.3c: Reactor Building Cooling Water Consumers .

. Division B:.

~

1

- Hot Emer g e ncy -,

Operating Normel Shutdown - (LOCA)

Shutdown . Stand >y 3,,

i Mode / Operating et 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> et 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> tno los: of II IAC IS"EP''i "

Components Condrtions Pool at 97'Cl -

l ACl RCW/RSW 2 3 3 2 -3 3

HeatEachangers i

in Service 3

! ESSENTIAL Emergency Di -. - - X X l -

i. eet Generator B RHR Heat X X - X X f -

ExchangerB FPC HostEx. X X X. X X X l

changer B l

Othe rs lessen- X X X X X- X'

! tial)W

. NON. ESSENTIAL RWCU Heat X X X- X :X -

! Exchanger

' U tooide Depvall* X X X- X' X. -

i.

, Othern (non- X- X X X X ~X-

. esuntiall'0 l til IX) . Equipment receives HCW in this mode.. -

n.) . Eauipment do.e not receive HCW in thii rnode.-

' (2)- H7tM reMgerator, room coolers (FPC pump, MMR. RCIC, SGTG, FCS, CAMS), RHR motor and seal -

coo'.e.*a.

135 Dryweri(B) und H17 coolors. . ..

- (4) ' Dasctor BuWing samp;ing coolers; LCW sump coolsrv (in drywell and reactor building), MIP MG sets

. and RWCU pump coolers.

7 7

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ABWR Design Document Tab!e 2.11.3d: Reactor Building Cooling Water Consumers .

} Diviilon C --

Hot Emergenc y Operating Normal shutdown . Stoney

' Shutdown . (LCCA) .

mom @rsting et 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> . at 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> . . (no loeo of

'*" Y - (Suppre ssion '

Compone nts Condhions lloss of AC . p ,,, , g .q,_

AC) -

3

'3 RCW/RSW 2 3 2 3 3 HsatExchangers in Service t

E.SSENTIAL I4

[

i-Emergency Die- - -- - - X -X.

eel Generator B j l

h

,1 RHR Heat - X X - X-X ,

Exchanger 3 1

i Othm (eemen- X X X X- X X -

tial)* -

NON ESSF.NT1AL ,

Others (r.on- X X X :X. X X essent;al)f83 5

}'

(1) (X) . Equipment roceives RCW in this n ode.

H = Equipment r.;oes not receive RCWin this mcde.

@/

( (2) HECW refrigerator, rorsm coelors, motor toolers, and mechanical as ' cootere for RHR and HPCF.

(3) Inotrumerit end service air celers, CRO pump oil cooler, redwsste components, HSCR conde naor, and l; - turbine building sampling coolers.

s 4

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ABWR oesign Document O MUWP i

A 3-F' TV M SURGE T ANK RHR Hz (Reactor Build >ng)' U

1. ' (Reactor Building)

L EMERGEN0Y D!G (Reactor BuMng)

M FPC HX _->-

$PCU (beac or Bunding)

MUWP CTHERS (ESSENTIAL)

(Reactor end Contot Building)

- CRD & CUW PUMPS ,__ , g.

7 ,

(Reactor Building)

(NC"I , - Ncl3 Os '

u _. _ - . .)

U _ ._. g _ ._ .. oTuERa <no+esseunAu _ __ _ __ _ ._. _ __ _ q (Reactor, Radweste and Turbine BuildN) -

l 'l I .t b . . , _ . _ . , IA/5A COMPRES$cRS - .___ __y I (Turbine Building) ~ l .

I "d

-N- DRYWELL EQUIPMENT -

(NC NCl2 - (NC NC]2 h CONT AlHMENT - - CONTAINMENT g RCW Hz . ,

eena a"

C Rsw- M

-RSW RCW PUMP.

RcwH1 ~~

(**' * *

'"~~ ~~

ICo"1mt BA141 r,sve - d W Rsw I' acW Hz C $

"= I acen m,,.

asw-----M M ;RSW RCW PUMP (conter sangL

-l

%(%)

Figure 2.11.3a RBCW Division _ A 9 6/1/97-2.11.3

,. .- , , ~., -, . - - , , , ,

. ABWR Design Document - _

O Muw?

NC SURGETANK RHRRx. ** dO' 0#^91 "

( (Reactor B.ntding)

U .

EMERGENCY Oto f

u (Reac'.or B# ding)

~

M U

FPC HX (Raactor B# ding) gg MUWP OTHERS (fSSEtmAy -

(Reactor and Control BvDding)

CLIW PUMP ,

7_ (Reactor Bund;ng)

,(HC I f gej3 l u _ . __ _ __ _ q OTHERS (NOP4 ESSENT1AQ h.--

t (Reactor Redwasie andTudne Building)

I h 'j i i U .

~ O b'- DRYWELL EQUIPMENT -D J NCl2 - (NC NCl2 3(NC CONTAINMENT CONTAINMENT RCW Hs n' 1 tcet@ BJdm\

f RSW O' O MSW- RCW PUMP "8

! RCW Hz

Kevd Bsem)

R SW- .RSW ,

m Rew Hr C (Cent d 9inwi RSW O RSW RCW PUMP (ContalBdidng) f

.O Figure 2.11.3b RBCW Division B-2.11.3 10 -

,ABWR oesign Document .-

2 i

r vuwp-1 . NC 3

i SURGE TANK p

hT M (Raactor Bunding) U t

s RHR Hz T _ _ _

1. ' (Reactor B# ding) i, M L EMERGENCY D'G w._m (Reactor BJding)

M

. SPCU _

i

~ M MUWP

' oTNE RS (ESSEtmAQ

' (Raactor and Control BuGding) 4 F

CRD PUMP b, N em 7- (Rosetot Building) ' lNCl3 ,

(NO I p ,

m _ _ _ _ __ _ q OTHERS (NON.ESSENT1AQ . _ , , . _ , , , , , , . , , ,

i (Reeetor, Radwaste and Turbine BaMng) i I I l I I

" L_______ ixes A couPRESsoRs- ._ _ __J (Turbine BuMng) ir l

4 RCWHz fControl BuMrei g RSW 7' M RSW RCW PUMP -

@"* WNI

! RCW Hz (Centrol Buidmi RS Rsw m Rcw Hz C tceetret a ms Rsw d 4 Rsw RcW PUMP (con w es u ngi L_

1-l Figure 2.11.3c RBCW Division - C -

I 11 6/IS2 2,11.3 1

ABWR DESIGN CERTIFICATION l 8/5/92 ACRS SUBCOMMITTEE-REVIEW ,

i o

5

(])

ENERGENCY-DIESEL-  :

SUMMARY

OF TIER 1 ENTRIES ,

1,

GENERATOR SYSTEM (EGS) '

s i

h DESIGN DESCRIPTION ENTRIES:

l

- SAFETY SYSTEM i - PROVIDES ON-SITE EMERGENCY POWER

! - THREE INDEPENDENT DIVISIONS 1 - THREE-INDEPENDENT FUEL TANKS

- SEISMIC' CATEGORY-I

! - START-UP SEQUENCE / TIMING / LOGIC ,

!($)

l j ITAAC ENTRIES:

i - BASIC SYSTEM CONFIGURATION:

F - LOADING CAPABILITY ~

AUTOMATIC, MANUAL START-

-- START-UP SEQUENCE / TIMING / LOGIC

! - PROTECTIVE TRIPS AND ANNUNCIATIONS  ;

n

' CFC-11 z8/5/92-L f-

!;y 1

[.-

i

,, .n,;,. .<.--a - .

ABWR oasign Document l

2.12.13 Emergency Diesol Gen: rater Systim (Standby AC Pow;r Supply)

Design Description O The Class 1E diesel generators comprising the Division I,II, and III standby AC power supplies are designed to restore power to their respective Class 1E t~

distribudon sptem dhisions as required to achieve safe shutdown of the plant ~

and/or to mitigate the consequences of a lossef<oolant accident (LOCA) in the event of a coincident loss of normal electrical power, Each of the three dhisions of the AC power system has its own diesel generator.

The major loads consist of the following systems for all three dhisions: Residual-Heat Removal (RHR) Sptem, Reactor Building Cooling Water (RCW) System, HVAC Emergency Cooling Water (HECW) System, and Reactor Service Water (RSW) Sptem. In addition, Divisions 11 and III include the High Pressure Core Mooder (HPCF) System loads. (The Dhision I RCIC System is aho part of the ECCS network, but is steam driven and theiefore does not present a significant load to the diesel generator.)

Each Class IE diesel generator, with its auxiliary sptems (i.e, Fuel Oil Storage and Transfer System. Jacket Cooling Water System, Staning Air Sptem.

Lubrication Sptem, and Combustion Air Intake and Exhaust System), supplies standby AC power to various Class 1E loads through the 6.9 kV and 480V systems.

m The 480V sptem, in turn, supplies power to the UPS and battery charger for the k dhision's 120 VAC and 125 VDC safety loads. (The low voltage portion _does not signincantly contribute to diesel generator loading, but is included with "other -

480V loads

• per Figure 2.12.13.) Each h physically and electrically holated from the other divisions. No automatic interconnection is provided between the Class IE dhisions. Each dieselgenerator set is operated independendy of the other sets, and is connected to the utility power system by manual control only during _-

testing or for bus transfer. A failure of any component of one diesel generator set will notjeopardize the capability of either of the two remaining diesel-generator sets to perform their functions. The diesel generators and their essentia1 support equipment are classified Seismic Category 1, and are quahfied for the emironments where located. All components except for the fuel storage tanks and fuel transfer equipment are located within the. Reactor Building.

Each diesel generator unit h rated at 6.9 kV,60 Hz, and is capable of .

automatically starting, accelenting, attaining rated frequency and voltage within 20 seconds, and supplying its loads in the sequence and timing specified in the-plant design documents. In addition, each diesel generator is capable of staning, .

accelerating and running its largest motor at any time after the automatic :

loading sequence is completed, assuming that the motor had failed to start initially. Each diesel generator unit is also reliability tested by the manufacturer.

2.12.13 1 W1S2

ABWR oesign Document The diesel generators start automatically on loss of bus voltage. Under voltage sensors are used to start each diesel engine in the event of a sustained drop in o)

C bus voltage below 70% of the nominal 6.9 kV rating of the bus. Low-water-level sensors and dipell high-pressure sensors in each division are also used to initiate the respective diesel start smder accident conditions. However, the diesels will remain on standby (i.e., nmning at rated voltage and frequency, but _

unloaded) un ess l the bus under-voltage sensors trigger the need for bus transfer to the diesel supply. Manual start capability (without need of DC power) is abo provided.

Each diesel is supplied by its own independent fuel storage tank, which is located in an area protected from natural phenomena. This tank has a fuel capacity sufBcient to operate its diesel for a period of seven days while the diesel generator is supplying maximum post LOCA load demand. A day tank is also provided for each diesel, and is located in the Reactor Building. The day tank has a fuel capacity sufBcient for approximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of full-load c,perations.

i Low level sensors on the day tank actuate dual motor driven transfer pumps to replenish the day tank supply from the storage tank.

The standby AC power supplies are desigred such that testing and inspection of equipment is possible during both normal and shutdown plant conditions.

Each standby AC power supply is composed of a threephase synchronous p generator and exciter, the diesel engine, the engine auxiliaries (including the v.J fuel tanks), and the control panels. Figure 2.12.13 shows the emergency diesel generator system interconnections between the offsite power supplies and the -

dieselgenerator standby AC power supplies for Divisions I, II, and IIL The transfer of each Class IE bus to its standby power supply is autornatic, should this become necessary, on loss ofits offsite power. After the circuit breaker connecting the bus to the preferred power supply is open,large motors are kept on the bus for parallel coastdown and optimal residual voltage decay.When the mitage decays to an acceptable level, major loads are tripped from the Class 1E bus, except for the Class 1E 480V unit substation feeders. Then the diesel-generator breaker is closed when the required generator voltage and frequency are established. The large motor loads are later re applied sequentially and automatically to the bus after closing of the dieselgenerator breaker.

Each diesel generator is capable of being started or stopped manually from the main control room. Start /stop control and bus transfer control may be transferred to a local control station in the diesel generator room. Control room indications are provided for system parameters.

p Each diesel generator, when operating other than in test mode, is indeperld ent V of the preferred power supply. Additional interlocks to the LOCA and loss of-2 6/1/92 2,12.13

I --

l ABWR Design oocument i

4' power sensing circuits terminate parallel operadon tests and cause the diesel ~

generator to rever*. and reset to its automadc control system if chher signal .

j( .Q' appean during a test. A lockout or maintenance mode removes the diesel b generator from serdce.The inoperable satus is indicated in the control' room.

p Dedces monitor the condidons of the diesel generators and ef.fect act:on in-accordance with one of the following categories: (1) conditions to trip the diesel:

I engine even under LOCA; (2) condidons to trip the diesel engine except under LOCA: (3) conditions to trip the generator breaker but not the diesel,'and (4) l- conditions which are only annunciated, i-e inspections, Tests, Analyses and Acceptance Criteris -

- Table 2.12.13 provides a deSnidon of the inspections, tests, and/or analpses l

together with associated acceptance cdteria which win be undertaken for the emergency diesel generaton and their auxi!jary synems.

3

• P k'

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3- . 6/1/92 2.12.13

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I- -e ,i- -s. . , + y , w w -e .r,<,r ,- h -v <. = ,<w , .,e +. .v- .., w d ,c >+-c .w e y --#,,o -, w-, , ,v.,--

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O o Taole 2.12.1*h Emergency Diesel Generator System y

" In3pections, Tests, Analyses and Acceptance Criteria Acceptar.co Criteria inspections, Tests, Anolytes f Certilled Design Commitment Plant tests and verification inspection for Tests and verification inspection wil1 be 1.

1. physicallocation confirm proper

1. The three diesel generator trains are conducted which will include indepor dent mechanically and electrically independent, independence of three diesel generator and coincident operation of the three trains to demonstrate comple*e divislinal dMsir ns.

separation.

2. See Generic Eq.sipment Qualification 2. See Generic Equipment Qualification

2. All componeets escential to the operation Acceptance Criteria (AC).

verification activities (ITA).

of the diesel generators are Seismic Category I and qualified for the appropriate environment for locations where instaffed. 3a. The maximum loads expected to occur for 3e. Confirmatory inspection will be performed each division (according to nameplate

3. The three diesel generators are capable of to assure the maximum design loads l ratings) shell not exceed 90% of the rated supplying sufficient AC power to achieve expected to occur for each division are safe shutdown of the pient and/or to within the ratings of the corresponding power output of the diesel generator.

f. mitigate the consequences of a LOCA in diesel generator, the event of a coincidem loss of no~nal 3b. Each of the three units shall produce rated power (Figure 2.12.13.). 3th Testing will be conducted by synchronizing power output et 20.8 PF for a period of 224 each diesel generator to the plant offsite hours (momentary transients excepted).

power system and increasing its output Each unit will then experience full loer 8

poew level to Ita fuity rated lood condition.

rejection by tripping the load an d verifying the unit does not trip.

4. Each diesel generetor attains a voltage of

4. Perform a test of each diesel generator to 6.9 kV110%, and a frequency of 60 Hri2%

4f Each d:esel generator is rated et 6.9 kV, confirm its ability to attsin rated frequency three phase. 60 Hr; sad is'espeble of l rifthin 20 seconds after application of a and vo?tege.

attaining rated frequency and voltage start algnal. ,

within 20 seconds after receipt of a start signal. .

B

O o o

~-

Table 2.12.13: Emergency Diesel Generator System (Continued)

G Inspections, Tests, Analyses and Acceptance Criteria inepections, Toots, Analyses Aeooptsace Critoria Certified Design Commitment 4

5. The eutomatic and manuel start sequences 5. Each of the three units starts from each

5. In the evant of a loss of normal power, each automatic and remote manuel signal, then w ll be tested for each diesel generator .

diesel generator unit is espeble of steriing eccelerates and property sequences its (both menuelly and automatically), . unit. loads. Each local manuel signal also starts 4

accelerating, and supplying its Ioeds in the the corresponding unit, but does not proper sequence and timing specified in initlete toed squencing.The automatic the plant design documents. It is also load sequence begins et $20 seconds and capable of recovery following trip and ends $65 seconds. Following spplication of -

restert of its largest foed. each loed, the bus voltage will not drop more then 25% measured at the bus.

. Frequency shell be restored towithin 2% of nominal, and voltage shall be restored to within 10% of nominst within 60% of each food-sequence time intervel.in addition, the unit's largest motor lood shall be Y' i tripped arid resterted after the unit has completed its sequence, end the bus voltags shall recover to 8.9 kV110% a:

6012% Hr within 10 seconds.

8. Block-start capebility is demonstreted v 6. Each unit wn! be tested and the air receiver

. 6. Each dieset gone ator unit !s capable of following one successful manuel start, tank cepecities shall be snelyzed to eseure 3

manustly starting without the need for eccelerotion, and bus energitation for each extemet electrical power.The air receiver its b!ack-stert cepobility la functional.

of the three units without assist from any -

tanks have sufficient copecity for five starts extemel electric power. Following block without recherg;r.g. -

start, each unit's receiver tanks shall have -l sufficient air remaining for four more

' starts.

7. Interlocks forthe stendby AC power 7. While in 's perellel test mode, each unit will -

.7. ' intertocks to the f.dCA and lose-of. power revert and reset to its automatic control :

! sensing circuite terminate peroNI ~ systern will be tested. .

system following individual application of '

' operation tests snd cause the diesel ' a simuisted LOCA signs.1 and a simuleted

.e generator to revert and reset to its . lossepower signet.

2 5 automatic contro! system if either signet 4 appears during a test.

O- O O

" Table 2.12.13: Emergency Diesel Generator System (Continued)

I" 0 Inspections, Tests, Analyses and Acceptance Criteria inspections, Tests, Analyses Acceptance Criteria Certified Design Commitment

8. Using simuisted signs!s, protective 8. Successful circuit testing will be confirmed

8. Devices monitor the conditions of the for the Individual diesel generator

~ diesel generefors, and effect action in interlocks and annunciations will be tested accordance with one of the following to assure they perform their functions, in protective sensors according to the accordance with the four categorical fo!!owing:

- categories: (1) conditions to trip the diesel engine even under LOCA, (2) cond;tions to comfitions de::dbed.

Catenory 1 Sensgf::: Annuncletions and trip the diesel engine except under LOCA, diesel engine trip signals w

• be confirmed (3) conditions to trip the generefor breaker but not the diesel, and (4) conditions which . In combination w!'A a simuisted LOCA .

signal.

are only annunciated. j Cetenorv 2 Sensors: Annuncletions and diesel engine trip signals will be confirmed without a LOCA, but trips will be bypassed when a simuisted LOCA signal is pre-wd.

b Nacorv 3 Sensorm- Annunciations and generator circuit breaker trip signals will be confirmed.

Cateoorv 4 Sensors: Annunciation signeta will be confirmed.

9e. Visual inspectien and calcu!stion of 9s. Tank Inspections and calculations confirm

9. Each diesel has its own 7-day fuel etorsge cepecities for each tank shall be performed. proper capecities of the storage and day tank and its own 8-hour capacity. day tank .

wh!ch is replenished by the storage tank. tanks. These shell be sufficient for full-loed operation of each respective diesel generator for 7 deys, and 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, reef,,cd.d f.

9b. The fuel transfer system shall be tested. 9b. Transfer system operation for each division will be confirmed by actuating both pumps from the day tank level sensors and observing proper flow into the day tanks.

g 3

O O O Table 2.12.13: Emergency Diesel Generator System (Continued) g e Inspections, Tests, Analyses and Acceptance Criteria 0 Acceptance Criterie inspectione, Tests Analyses Certified Design Commitment

10. Visualinspection of manufacturers test

10. The manufacturers test documents shet! documents confirms the required reffability

10. The manufacturer has conducted be visually inspected, reliability testing on the units. testing has been performed. and that the diesel generator has passed the test requirements.

11. The designated Instrumentation is present

11. Inspections will be performed to verify

11. Control indications are provided for D/G in the control room.

l presence of control room Indication for the system parameters. D/G system.

9

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- ~ .

ABWR oesign Document O

~ ~. - -

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I I I I I I I I I I I I HIGH PRESSURE g l N

g *l SYSTEMS I I I I i i i i i i *

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DMSON ll DIVISION 111 DIVISION I

, O Figure 2.12.13 Emergency Diesel Generator System Interconnections 6/142

-8 2.12.13 y V w- - g

• p- g

2 ABWR DESIGN CERTIFICATION

! 8/5/92 ACE S.UBCOMMITTEE REVIEW L

LO i

SUMMARY

OF TIER 1 ENTRIES i

CONTROL BUILDING (CB)

, DESIGN DESCRIPTION ENTRIES:

1

- SEISMIC CATEGORY I STRUCTURE.

! - REINFORCED CONCRETE WITH STEEL TRUSS ROOF i 2 STORIES ABOVE GRADE 4 STORIES BELOW GRADE.

l

- RECTANGLE 24x56x30.5 M.

i - WALL THICKNESS 0.6 - 1,6 M

! MISSILE AND TORNADO PROTECTION PROVIDED i - PROTECT AGAINST INTERNAL'AND EXTERNAL FLOODS i -

RADIATION SHIELDING

- BAS C. ARRANGEMENT DRAWINGS (7) i-ITAAC ENTRIES:

l

[ - BASIC LAYOUT CONFIGURATION

- FLOOD PROTECTION DESIGN FEATURES

! - TORNADO AND MISSILE PROTECTION DESIGN FEATURES-SHIELDING' FEATURES l 1

I i

AJJ-1 -1 8/5/92-i C

i 1

B

ABWR Design Document 2.15.12 Control Building Design Description The Control Building (CB) is the building that houses the main control room, control equipment, and operations personnel for the Reactor and Turbine Islands. The Control Building is located between the Reactor and Turbine Buildings.

In addition to the control room and operations personnel, this building houses the essential elecuical, control and instrumentation equipment, essential switch gear, essential battery rooms, the CB heating and air conditioning (IWAC) equipment, Reactor Building component cooling water pumps and heat exchangers, and the steam tunnel.

The general building arrangement, including watenight doors and sills for doorways where needed for flood control, is shown in Figures 2.15.12a through 2.15.12g.

The CB is a Seismic Category I structure designed to resist seismic loadings and to provide protection for Dooding, tornado wind, and tomado missiles.

The CB is constructed of reinforced concrete with steel truss roof. The CB has two stories above the grade level and four stories below. The building shape is g rectangle. Major nominal dimensions are as'follows:

yi Overall height above top of basemat 30.5 m Overallplanar dimensions (outside) 0 -180* direction 24.0 m 90 -270* direction 56.0 m Thickness of Outer WaB from 8.2m TMSL to 17.15m TMSL 1.0 m i from 17.15m TMSL to 22.2 m TMSL -0.6 m Thickness of Steam Tunnel Walls, Floors, and Ceiling 1.6 m Thickness of Walls supporting Steam Tunnel ,

1.6 m l

The CB is a shear wall structure designed to accommodate all specified seismic loads with its perimeter walls and steam tunnel walls together with their supporting elements. Therefore, frame members such as beams or columns are designed to accommodate deformations of the walls in case of earthquake i

condition. The columns provide venical weight bearing capability. Column sized I

and floor slab thicknesses are also provided in the general building arran ge ment Sgures. With major dimensions defined as listed above for speciBed reinforced

{} concrete materials and design procedures, the dynamic characteristic of the CB 1 6M/92 2.15.12

ABWR oesign occument structure is defined. Seismic adequacy of the detailed site 4peciSc coritrol building design will be evaluated using the dimensional characteristics no O.- above and approved analytical procedures and methodology for dynamic _

analysis of stmcrures. This work will be in compliance with the ACI an codes governing design of reinforted concrete structures and steel str nuclear power plants. Detailed analyses of the site specific control buildi design will utilize appropriate site data for seismic eve'nts, Goods, t and other loading conditions. ,

To protect against external flood damage, the following des!gn features a provided:

Wall thickness below Dood level greater than 0.6m.

(1)

(2) . Water stops prmided in all construction joints below grade.

(3) Watertight doors and piping penetrations installed below flood level.

(4) Waterproof coating on exterior walb.

(5) Foundations and walls of structures below grade are designed with; d

water stops at expansion and constructionjoints. -

(6)

Roofs are designed to prevent pooling oflarge amounts of water..

To protect against intemal flood da nage, the following design features are pimided:

Elevation differences and divisional separations from remainder of the r

-(3)

CF.

(2) Drainage synem to divert water to assigned floor and location.

(5) Sills for doorwap as needed to provide Odd control; p

' . (4) . Watertight doors installed below interna 10ood.levet

,(5) fWall thickness below internal Good level greates thari 0.6ma n

I.nside the steam tunn el is the mainsteam piping, tiie mainsteam drain line, and i

p the feedwater piping.There is no penetratio9 f rom the steam tunnelinto the control building. Any high energy line breaks insids the steam tunnel will ven out to the Turbine Building. All standinjg watei~will collectin the large w>lumes ,

2 in the lower pordons of the steam tunnel at the Reactor Building or Turbine B 114 ins eoas. .

j lD

- 6/1/92. -.j 2 i

' 2.15.12 ~

1- s i<

ABWR oesten Document On Floor Blf, there are fire hose stands and reactor cooling water (RCW)

piping. I' is designed that any rupture of the fire hose stand willleak onto the 4

O fleer and d,ain to the 4200 level by floor drains. Silh will be provided at doorways to prevent the entry of standing water into the control room complex. 1 The RCW piping runs vertically in a concrete pipe chase. No flooding outside  ;

this pipe chase is possible.

i

On the Door where computer room located, there are fire hose stands, RCW piping, and other piping systems. Varying amounts of standing water are -

expected upon a rupture of any of these systems. Maximum water height corresponds to the height of the door sills. Sills will be provided at doorways to prevent water from crossing divisional boundaries. Similar arrangements and -

designs are also provided for other floors for floods protection.

During normal operation, the concrete surrounding the steamline tunnel provides shielding so that operator doses are below the value associated with uncon. rolled, unlimited access. The outer walls of the control building are designed to artenuate radiation from radioactive materials contained within the reactor building and from possible airborne radiation surrounding the control

- building following a LOCA. The walls provide shielding to limit the direct 4 hine exposure of control room personnel following a LOCA. Shielding for the outdoor air cleanup filters also is provided to allow temporary access to the mechanical equipment area of the control building following a LOCA, should it x be required.

The control building is not a vented structure. The exposed exterior roofs and - ,

walls of the saucture are designed for the required pressure. drop.' Tornado dampers are prosided on all air intake and exhaust openings,-These dampers are

- designed to withstand the specified negative pressure, i'

Inspections, Tests, Anstyses and Acceptance Criteris Table 2.15.12 provides a definition of the inspections, tests, and/or analyxs,

~

together with associated acceptance criteria which will be undertaken for the control building.

L 4.- 6/1/92 2.15.12-

. . .. _. . _ ... _ _ _.. .. _ ~ _ _ __ .._ ._ - . . . _ . . _ . . _ . . . . . . .

Table 2.15 contrcl Building E.'

O Inspectione Tests, Analyses and Acceptance Criteria ,

inspections Tests, Analyses Acceptance Criteris Certified Design Commitment

1. Plant walk through* to check and verify 1. Per Figures 2.15.128 through 2.15.12g.

1. Control building general arrangement is shown in Figures 2.15.12a through requirements are met.

2.15.12g.

2. Review construction records and perform 2. For extemal flooding.

2. Design features are provided to protect s. Exterior wall th'ckness below flood against design basis intemel and extemal visualInspections of the flood control features. level greater than 0.6m.

floods. b. Water stop.

4

c. Watertight door and piping penetrations below flood level.

d. Waterproof coating on exterior walls.

e. Foundations and watts of structures below grade are designed with water stops et expansion and construction joints.

f

f. Roofs are designed to prevent pooling of large amounts of water.

Forintemet flooding:

a. Elevation differences and divisional separation of the mechanical functionc from the remainder of the CB.

b. - Drainage system to divert water to sesigned floor and location.

c. Sit!s for doorweys as needed to provide flood protection.

d. - V/stettight doors Installed below Intemel flood level.

e. Wall thickness below intemal flood level greater than 074.

f. Steem tunnri has no penetrations from the steem tunnel into the control building. Any high energy line or

• Plant welk through is intended to include feechCer pig,ing breaks inside the 3 steam tunnel will vent out to the 3 visualinspection of the as-bulit facility and Turbine Building.

(as-needed) dimensional measurements.

O O O M Table 2.15.12: Control Culiding (Continued)

?

O Inspections, Tests, Analyse: and Acceptance Criteria inspections, Tests, Analyses Acceptance Criterie Coetitled Deslon Commitment

3. The Control Building is designed to have 3. Performed dimensionalInspections of the 3. The concrete thickness for1.,o steam adequate radiation shielding to protect Control Building walls. ceiling, floors, and tunnel well, floor and ceiling shall be operating personnel during operation and other structural features. greater then 1.6m.The steem tunnel following a LOCA. Interface struc. ore and control building wall below the steam tunnel should have a e combined thickness of 1.6m, i.e. In any line-of. night from the control room. the f tctal thickness of concrete between the observer and the steam lines must be 1.6m or greater.

4 The CB is designed to protect against 4. Review construction rocords and perform 4 Fortomado design beals tomedo and tomado missiles, visual Inspections and dimensional checka (es-needed) of the tomado protection a. Roof and wells above grade designed p features. greater then 0.5m.

b. HVAC dampers designed for differentiel pressure > 1.46 psi.

c. HVAC dampers have tornado missile -

barriers.

5. ' The CB is designed as a Seismic Category I 5. Plant welk through to check and verify CD 5. Structures have dimensions compatible structure and has mejor dimensions bulloing mejor dimensions including with data in the certified design (Figures defined in the certified design, column sirrs and floor sieb thickness. 2.15.12a through 2.15.12g).

Review finat design record for meterial properties site input date and snelytical procedures and methodology for seismic analysis.Visualinspections of structures and review of so-built documentation will be conducted to assess compilence with the certified design commitments.

6. . The detail structural design will be based - 8. Thecontrolbuildingdesigndocumentation : 6. Confirmation that the as-built duign is in i

on AC3 and AISC codes and will use site will be reviewed. compliance with ACI and A!SC data for seismic events, floods,tomadoes requirements and is based on appropriate '

, u winds and other losding conditions. site design dets.

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n-Note Roof thicknessis 300 mm Steam Tunnel roof thickness is 1600 mm S

3 Figure 2.15.12e CMOL B"'t. DING ELEVATION (90' - 270*)

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v(x O- O-h ELEVATION 17150mm TMSL 10400mm 14300mm momm e to200 m i

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rol Building Floor 82F Figure 2.15.12e C .

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Figwe 2.15.12e C trol BulkEng Floor B4F  ;

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s .4 m . . .w., ,m , y _w w.., ,..- __.o.._________._.___s__.-._m,._. ._..

l- - ur I

PRESENTER: TOM BOYCE PROJECT MANAGER, NRR I

I

SUBJECT:

INSPECTIONS, TESTS, ANALYSES, AND ACCEPTANCE CRITERIA (ITAAC) FOR THE GE ABWR i

l l

\

l 4

August 5-6,1992 O O O

ITAAC FOR DESIGN CERTIFICATIONS

SUMMARY

OF ITAAC STATUS g

• STAFF AND INDUSTRY ARE LDE/ ELOPING ITAAC, WITH SEN!OR MANAGEMENT INVOLVEMENT

• GE ABWR IS THE LEAD DESIGN FOR ITAAC DEVELOrMENT

• iTAAC DEVELOPMENT CONTINUES TO BE ITERATIVE, AND

! r MANY ISSUES UNDER DISCUSSION i

a SOME INCONSISTENCIES HAVE BEEN NOTED IN SSAR/ITAAC l

• ITAAC !MPLEMENT SEVERAL ASPECTS OF 10 CFR PART- 52 l l

l I

e O 9

ITAAC FOR DESIGN CERTIFICATIONS BACKGROUND

• REQUIREMENT FOR ITAAC 'IN 10 CFR '52.47(a)(1)(vi)

• SECY-91-178 DISCUSSED FORM AND CONTENT OF ITAAC

• SEGY-91-210 DISCUSSED RELATIONSHIP OF FDA AND ITAAC l

• SECY-92-053 DISCUSSED CONCEPT OF DAC l = SECY-92-196 DISCUSSED RAD PROTECTION AND PIPING DAC

• SECY-92-214 DISCUSSED CURRENT STATUS OF ITAAC

= SECY FOR I&C AND HFE DAC EXPECTED TO BE ISSUED THIS MONTH O O O

ITAAC FOR DESIGN CERTIFICATIONS STATUS OF THE REVIEW

• GE SUBMITTED ITAAC IN 3 STAGES, WITH SENIOR MANAGEMENT MEETINGS HELD EVERY 6-8 WEEKS

• STAGE 1 - NINE ' PILOTS" SUBMITTED SEP 91

• STAGE 2 - 40 SYSTEMS SUBMITTED APR 92

• STAGE 3 - FULL ITAAC SUBMITTAL JUN 92 1

O O O

ITAAC FOR DESIGN CERTIFICATIONS STATUS OF THE REVIE'N I

= DRAFT FSERS ON CONTROL ROOM AND I&C DACS SUBMITTAL WILL BE PROVIDED IN SEGY PAPER TO BE ISSUED IN AUGUST; DRAFT FSERS ON PIPING AND RAD PROTECTIONS DACS WERE PROVIDED IN SECY-92-196 ISSUED MAY- 28,1992

= COMMENTS ON STAGE. 2 SUBMITTAL BEING PROVIDED AS PART OF AUGUST DRAFT FSER

.. COMMENTS ON STAGE 3 SUBMITTAL BEING PROVIDED V!A SEPARATE CORRESPONDENCE

• 'GREYBEARDS* AND REGIONAL REVIEW AND COMMENTS ON DESIGN CERTIFICATION MATERIAL WILL -BE PROVIDED

= INTERACTION WITH NUMARC SCHEDULED FOR LKfE AUGUST-O O O

ITAAC FOR DESIGN CERTIFICAT!ONS STATUS OF THE REVIEW (CONT) i

= STAFF IDENTIFIED lNCONSISTENCIES IN SUBMITTALS BETWEEN SSAR, DESIGN DESCRIPTIONS, AND ITAAC

• GE IS RESOURCE CONSTRAINED, RESULTING ::IN DELAYS IN-SUBMITTALS AND RESOLUTIONS OF ISSUES

• FSER (AUGUST) TO INCLUDE INITIAL EVALUATION OF ' STAGE '2 I SUBMITTAL; GE MUST HAVE YlMELY SUBMiTTALS TO CLOSE  ;

ISSUES AND SUPPORT SCHEDULE FOR FDA 1 i

q O O O

T i

ITAAC FOR DESIGN GERTIFICATIONS r i

TYPES OF ITAAC i

e

• SYSTEMS ITAAC" FOR SYSTEMS OF DESIGN

• ' GENERIC ITAAC' FOR GENERIC CONCERN 3 ACROSS SYSTEMS.

CROSS REFERENCED TO SYSTEMS WilERE APPROPRIATE 4

i STAFF IS CONSIDERING ' COL ITAAC" FOR LICENSEE PROCEDURAL REQUIREMENTS (E.G., TRAINING, ETC.)

• ' INTERFACE ITAAC* FOR SITE-SPECIFIC DESIGN (E.G., Ui_TIMATE HEAT SINK, ETC.)

i  :

"DAC* FOR SELECTED AREAS OF THE DESIGN

=

4 I

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} RELATIONSHIP 10F DESIGN DESCRIPTION TO ITAAQ -

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= DESIGN DESCRIPTION. CERTIFIED IN ' DESIGN CERTIFICATION RULE

[ WILL CONTROL PROPOSED CHANGES TO; THEL DESIGN. BYJ A FACILITY

i. THAT REFERENCES.THE- CERTIFIED ~ DESIGN:-

! = ITAAC WILL- BE USED FOR' FUEL' L.OAD DECISIONAND LSUBSEQUENT-l - FACILITY MODIFICATIONS- TO 'THE DESIGN l 3

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= STAFF IS EVALUATING WHETHER ALL ELEMENTS .0F DESIGN.

DESCRIPTION REQUIRE A- CORRESPONDING ITAAC. I

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= STAFF IS EVALUATING GE: PROPOSAL THAT CERTAIN SYSTEMS .

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{ -SHOULD HAVE DESIGN' DESCRIPTIONS .WITHOUT CORRESPONDING s ITAAC, BASED ON. SAFETYJSIGNIFICANCE- OF ' SYSTEM "

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ITAAC FOR DESIGN CERTIFICATIONS: 1 TREATMENT OF REGULATORY REQUIREMENTS!NOT IN ITAAC

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• SOME REQUIREMENTS MET AFTER. FUEL LOAD, BUT-PRIOR TO-OPERATIONS (E.G., START-UP AND INITIAL POWER TESTING)

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• THESE ISSUES TREATED AS CONDITIONS OF THE COL;

ANALOGOUS TO PLANTS LICENSED UNDER 10 CFR PART 50 WHERE

TESTING OCCURED AFTER OL ISSUANCE I

= MODIFICATIONS TO THESE PROGRAMS - WOULD BE LICENSEI AMENDMENTS, PROVIDING OPPORTUNITY FOR PUBLIC COMMENT

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b ITAAC: FOR. DESIGN CERTIFICATIONS TREKfMENT OF NON-TRADITIONAL ITEMS

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• INSIGHTS FROM PRA AND? SEVERE 2 ACCIDENT ISSUE RESOLUTIONS-i (E.G., SECY-90-016 ETC.) INCORPORATEDINTO SSAR. -

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• IMPLICIT CONFIRMATION OF THESE ISSUES 'SINCE ITAAC VERIFY.

DESIGN IN SSAR .

• STAFF HAS REQUESTED GE TO ' DEVELOP CROSS -REFERENCEj OF iSSAR- d

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i ISSUES TO ITAAC; EXAMPLE PROVIDED~ IN-~ SECY-92-2141 FOR 1 CONTAINMENT PERFORMANCE ANALYSES -+

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ITAAC FOR DESIGN CERTIFICATIONSi a

'ITAAC REQUIREMENTS FOR FDA

• SECY-91-210 DISCUSSED REQUIREMENTS TO REVIEW ANDLAPPROVE ITAAC PRIOR TO FDA' y

• REV!EW OF ITAAC :IS .RESULTING' IN! AMENDMENTS iTO SSAR;1 FOR. <

I EXAMPLE, JUSTIFICATION OF ITAAC: ENVELOPING:' VALUESXSHOULD1 BE IN SSAR, WHEREAS .SSAR : CURRENTLY- MAY1HAVE ONLYJLISTED: y ri NOMINAL ~ VALUES ..

• RULEMAKING FOR DESIGN CERTIFICATION CANNOT BE INITIATED ,

UNTIL REVIEW 'OF TIER 1 L MATERIAL 'IS : COMPLETE,

• STAFF CONCLUDES THAT DESIGN AND ITAAC REVIEWS '- ARE: :i FUNDAMENTALLY LINKED 1: AND SHOULD-.BE COMPLETED PRIOR :LTOlFDA i

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• COMMISSION . SUPPORTED' THIS CONCLUS!ON lN SRM -ON -SECY-92-037 BY INDICATING THAT THE STAFF MS BOUND 'BY THELSAFETY:-

DECISIONS IN THE FDA

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ITAAC FOR DESIGN CERTIFICATIONS

SUMMARY

OF ITAAC STATUS

• STAFF AND INDUSTRY ARE DEVELOPING iTAAC, WITH SEN'OR MANAGEMENT INVOLVEMENT

• GE ABWR IS THE LEAD DESIGN FOR ITAAC OEVELOPMENT  ;

• !TAAC DEVELOPMENT CONTINUES TO BE ITERATIVE, AND '

MANY ISSUES UNDER DISCUSSION

• SOME INCONSISTENCIES Hr!E BEEN NOTED IN SSAR/ITAAC

• ilAAC IMPLEMENT SEVERAL ASPECTS OF 10 CFR PART 52 f

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., ABWR DEllGN CERTIFICATIQR

\

8/6/92 ACRS. REVIEW O

l l

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I GE PREPARATION'0F ABWR TIER 1_ MATERIAL-A'

SUMMARY

10 I A. J. JAMES I MANAGER,-HECHANICAL i SYSTEMS DESIGN GE' NUCLEAR. ENERGY L 1408)1925-5002-

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i l ABWR DESIGN CERTIFICATION I 8/5/92 ACRS-SUBCOMMITTEE REVIEW

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! GE PRESENTATION ON TIER 1/ITAAC STATUS l AGENDA i o OVERVIEW OF GE APPROACH TO A. J. JAMES j PREPARATION OF TIER-1 MATERIAL i

o STATUS OF ABWR TIER 1 DEVELOPMENT A. J. JAMES j

> - SUBMITTALS TO NRC

O

- INouSTRY (NuMARC> REVIEW l

1 i o EXAMPLES OF TIER 1 MATERIAL FOR A. J. JAMES l FIYE ABWR SYSTEMS

- STANDBY LIQUID CONTROL SYSTEM (SLCS) A. J. JAMES  ;

1 - RESIDUAL HEAT REMOVAL (RHR) J. C. CHAMBERS i - REACTOR BUILDING-COOLING WATER (RCW) R. D. ROBERT 5 HAW

- EMERGENCY DIESEL GENERATORS-(EDG) C. F. CHRISTENSEN-

- CONTROL BUILDING (CB) A. J. JAMES i

o SVHMARY A. J. JAMES'- '

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i AJJ-2

) ._8/5/92

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AB.R'GESIGN CERTIFICATION 3/5/92 ACRS SUBCOMITTEE REVIEW R&$ES FOR GE APPROACH TO PREPARATION OF TIER I NATERIA1 BASIS APPRGACH ENDORSED BY THE Com ISSION (SRN FEB. 15, 1991)

CERTIFICATION IS TWD-TIERED PART 52 STATENENTS OF CONSIDERATION.

TIER liIS RESERVED FOR TOP-LEVEL SECY-90-241,91-178

...INFORMATION..(L r.,'A SUBSET OF THE SAR)

'ITAAC-'(INCLUDING DAC, GENERIC AND EXPLICITLY SPECIFIED IN 10CFR 52. 79(c)

-INTERFACE) ARE; TIER'1 PART 52. . . . AM3 (THE' PLANT) - WILL OPERATE IN LITAAC. VERIFY CONFORMANCE-WITH TIER l' CONFORMANCE WITH TNE DESIGN CERTIFICATION."'

. DESIGN-PART 52' INVOKES PART 50 QA PROCESSES

' VERIFICATION OF NON-TIER l'IS VIA EXISTING-

PART.50-PROCESSES AJJ-3 8/5/92

ABWR'D W GN CERTIFICATION 8/5/92 ACRS-SUBCOMMITTEE. REVIEW-O ELEB M S INCLUDED IN TIER 1 ELEMERI INTENT DESIGN DESCRIPTION (S) THE CERTIFIED DESIGN INSPECTION, TESTS, ANALYSES VERIFY THAT SPECIFIC FEATURES AND ACCEPTANCE CRITERIA 0F-THE AS-BUILT FACILITY (ITAAC) COMPLY WITH THE CERTIFIED DESIGN /

DESIGN ACCEPTANCE CRITERIA AN ITAAC ON THE DESIGN O (o^c> raocess:wata oest5" ocTA2's ARE (LEGITIMATELY).NOT

.AVAILABLE AT THE' TIME OF '

DESIGN CERTIFICATION INTERFACE ITAAC VERIFY THAT SITE-SPECIFIC FEATURE (5)1 COMPLY'WITH-REQUIREMENTS OF THE CERTIFIED DESIGN GENERIC ITAAC VERIFY THAT GENERIC ASPECTS OF THE ' AS-BUILT: FACILITY COMPLY WITH.THE CERTIFIED-

. DESIGN (s.o. ,. EQ)

O AJJ-4 8/5/92.

ABWR DESIGN CERTIFICATION 811/12 ACRS SUBCOMMITTEE REVIG O GE APPROACH o STRUCTURE TIER 1 ON A SYSTEN-BY-SYSTEM APPROACH

- NOT SAR STRUCTURE

- COVER HOST PLANT SYSTEMS o GRADED TREATHENT OF SYSTEMS THAT REFLECTS SAFETY SIGNIFICANCE -

o STEP 1: PREPARE TIER 1 DESIGN DESCRIPTION FOR EACH SYSTEM

- PRINCIPAL DESIGN BASES AND DESIGN FEATURES

!O o STEP 2: PREPARE ITAAC TABLE FOR EACH SYSTEM DERIVES DIRECTLY_:FROM THE DESIGN DESCRIPTION EN1 RIES o STEP 3: PREPARE OTHER TIER 1 ENTRIES AS'NEECED KEY CONSIDERATIONS o DESIGN DESCRIPTION AND ITAAC CONTENT REFLECT THE TIERED APPROACH-T0 CERTIFICATION-o EXISTING PART 50 VERIFICATION PROCESS.ARE APPLICABLE AND PLAY A KEY ROLE IN PART 52 O

' AJJ - 5 '

8/5!92 u-,.-.-. .

i l ABWR DESIGN CERTIF_1 CATION 81511LACRS. SUBCOMMITTEE REHEW

O 1

i CHARACliRullC1_OF TIER 1 ENTRIG

]

I DESIGN DESCRIPTIONS i

I o DESCRIBES THE PRINCIPAL DESIGN. BASES 'AND. DESIGN FEATURES OF i THE FACILITY; DRAWN FROM SAR DESIGN DESCRIPTIONS 3 i-

! i

' o SYSTEM BASED APPROACH WITH LEVEL OF DETAIL GRADED-TO REFLECT SYSTEM IMPORTANCE TO SAFETY- ,

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u 1

o CONTAINS.ONLY TECHNICAL INFORMATION ALREADY COVERED IN TIER ,

1 i 2 (SA*.)

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1

. o DOES NOT ADDRESS PLANT OPERATING CONDITIONS (COVERED BY. .

I TECH. SPECS. )'

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l o INCLUDES NUMERICAL INFO 3MATION T0-THE EXTENT NECESSARY TO j IDENTIFY PRINCIPAL DESIGN-BASES AND FEATURES

[.

[ o SELF-CONTAINED AilD AVOIDS DIRECT. REFERENCES TO: TIER 2-

! DOCUMENTS F .

L o MAY INCLUDE SIMPLIFIED P&IDfSc ONE-LINE DIAGRAMSL GENERAL:

ARRANGEMENT DRAWINGS WHICH ADDRES$1THE DESIGN FEATURES.IN' '

I THE TEXT.OF?THEiTIER11 DESIGN DESCRIPTION.-

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AJJ-6 . l 8/5/92:

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ABWR DESIGSI CERTIFICAT10H 8/5/92 ACRS $llBCOMMITTIE REVIEW CHARACLERISTICS OF TIER 1 ENTRIES ITAAC o AIMED AT CONFIRMING THE AS-BUILT FACILITY COMPLIES WITH THE CERTIFIED DESIGN o SYSTEM BASED AND DERIVED FROM (AND ADDRESSES HOST OF) THE TIER 1 DESIGN DESCRIPTION

()

o NUMERICAL VALUES MAY HAVE RANGES OR TOLERANCES o THE ITAAC PROCESS ENDS AT FUEL LOAD

- POST-FUEL LOAD TESTING NOT IN ITAAC (LICENSE CONDITION) o UTILIZE ELEMENTS OF EXISTING NUCLEAR POWER PLANT

'(]

VERIFICATION PROGRAMS

!AJJ-7 8/5/92-

i ABWR DESIGN CERTIFICATION 8/5/92 ACRS SUBCOMMITTEE REVIEW O TYPICAL TIER 1 ENTRY FOR AN AB)fR SYSTEM l

l DESIGN DESCRIPTION (TYPICAL) ENTRY PER SY_$.If8 I

1/2 - 5 PAGES OF TEXT I

0 - 5 FIGURES, DIAGRAMS' 1

INSPECTIONS. TESTS, ANALYSES AND ACCEPTANCE CRITERIA FOR EACH O SYSTEM TABULATION CONTAINING 2 - 20 ENTRIES CERTIFIED INSPECTIONS, 1 DESIGN TESTS, ACCEPTANCE ,

COMMITMENT ANALYSES _CRLIIRIL_

DERIVED FROM THE WHAT ACTION WILL WHAT CONSTITUTES-SYSTEM DESCRIPTION BE TAKEN T0. ACCEPTABLE VERIFY THE'CDC7 RESULTS OF THE

. ACTION 7 ,

O AJJ-8 8/5/92-

. . . - . . . - . - - . . . - - - . . .- . . . .. . . ~ . . . . .-.

ABWR DESIGN CERTIFICATION 8/5/92 ACRS SUBCOMMITTEE REVIEW

() l GE/NRC REVIEW OF ABWR TIER 1 STATUS:

o GE HAS SUBMITTED 100% OF PROPOSED ABWR TIER 1 KATERIAL

- 115 SYSTEM + OTHERS

- ROAD KAPS AND OTHER CLEANUP ITEMS IN PROGRESS o INTERACTIONS TO DATE INDICATE CONSENSUS ON BASIC SCOPE AND CONTENT

(])

- MANY DETAILS OPEN o NRC CURRENTLY REVIEWING THE MAJOR JUNE 1 STAGE 3 SUBMITTAL

- COMMENTS EXPECTED FIRST RALF 0F AUGUST

- " LOTS OF WORK REQUIRED" o ANTICIPATE INTENSIVE INTERACTIONS OVER THE NEXT FEW MONTHS O

l

- AJJ-9 8/5/92' 4

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! ABWR DESIGN CERTIFICATION l 8/5/92 ACRi_ SUBCOMMITTEE RF. VIEW

!O STATUS: INDUSTRY REVIEW 0F ABWR TIER _1 l

?

l o REVIEW CONDUCTED UNDER AUSPICES OF NUMARC STANDARDIZATICN l

OVERSIGHT NORKING GROUP (50WG)

I

- UTILITIES f

' - EPRI

- INPO i - A/E'S I

I o REVIEWS IN PARALLEL WITH NRC i

- - APPROXIMATELY 20 SYSTEMS COVERED 50 FAR-MAJOR ACTIVITIES SCHEDULED IN AUGUST / SEPTEMBER ~

l O l o SEVERAL TRENDS ALREADY CLEAR:

(

CHANGES NEEDED TO REFLECT LEGAL-SIGNIFICANCE OF TIER 1 ACCEPTANCE CRITERIA NEED-T0-BE MORE PRECISE, UNAMBIGUOUS

- REDUCE THE AMOUNT OF: TIER l' MATERIAL FOR NON-SAFETY-SYSTEMS STRONG DESIRE TO ELIMINATE GENERIC:ITAAC-o ADDITIONAL REVIEWS SCHEDULED FOR AUGUST / SEPTEMBER AND INTERACTIONS WITH NRC BEING PROPOSED FOR SEPTEMBER /0CTOBER 9

O AJJ-10 8/5/92

i

.ABWR DESIGN CERTIFICATION )

815/92 ACRS SUBCOMMITTEE REVIEW i O

IUMMARY o REVIEW OF ABWR TIER 1 MATERIAL NOW MOVING CENTER STAGE-t l

o THE TIERED APPROACH TO DESIGN CERTIFICATION.IS A MAJOR INFLUENCE ON ITAAC, DAC, INTERFACE ITAAC AND GENERIC ITAAC [

SCOPE AND CONTENT.

i o IN ADDITION TO ITAAC, PART 50 0A PROCESSES HAVE AN '

IMPORTANT. VERIFICATION ROLE IN PART 52 LICENSING o GE/NRC CONSENSUS ON BASIC SCOPE OF TIER l'

- MANY DETAILS OPEN '

EXTENSIVE INTERACTIONS THROUGH END OF CY92 o -UTILITY /NUMARC PARALLEL REVIEW-

- SOME-CONFLICTS WILL NEED'TO.BE. RESOLVED-MAKING PROGRESS;:

LOTS OF: WORK:

REMAINING' Ajj.11:

l8/5/97

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l ABWR DESIGN CERTIFICATION i 8/6/92 ACRS REVIEW l O EXAMPLES OF ABWR SYSTEM ITAAC  !

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ABWR DESIGN CERTIFICATIQH 8/5/92 ACRS 10BCOMMITTEE REVIEW O

STANDBY LIQUID CONTROL  :

SUMMARY

OF TIER 1 SYSTEM (SLCS) ENTRIES DESIGN DESCRIPTION ENTRIES:

- INJECT NEUTRON ABSORBING POISON INTO REACTOR AUTOMATIC INITIATION'ON ATWS-SIGRAL

- - KEY EQUIPMENT PERFORMANCE AND ASME CODE REQUIREMENTS-GIVEN COMPONENTS REQUIRED FOR INJECTION ARE SEISMIC CATEGORY I

[) - SIMFLIFIED SYSTEM DIAGRAM ITAAC ENTRIES:

BASIC SYSTEM CONFIGURATION POISON REQUIREMENTS

- . PUMP DESIGN LIMITS

- IN-SERVICE FUNCTIONAL TESTS ELECTRICAL POWER REQUIREMENTS SEISMIC REQUIREMINTS.

O:

AJJ 8/5/92

.. = _ _

4 l ABWR Design Document

! 2.2.4 Standby Liquid Control System O The Staader tisuid Centrei (SLC) Srsiem is desiseed ie inject neetren absorbing poison using a boron solution into the reactor and thus provide back.

up reactor shutdown capability independent of the normal reacthity control

system based on inserdon of control rods into the core. The SLC System is

capable of operation over a wide range of reat. tor pressure conditions up to and including the elevated pressures associated uith an anticipated plant transient 4 coupled with a failure to scram (ATWS).

The SLC System is designed to bring the reactor, at any time in a cycle, and at all conditions, from full power to a subcritical condiden, with the reactor in the most reactive xenon-free state, without control rod movement. The system will  ;

inject the minimum required boron soludon in 61 minutes, j The SLC System (Figure 2.2A) consists of a boron solution storage tank, two positive displacement pumps, two motoreperated injecdon valves which are provided in parallel for redundancy, and associated piping and valves used to transfer borated water from the storage tank to the reactor pressure vessel -

(RPV). The borated solution is discharged through the 'B' high pressure core Gooder (HPCF) subsystem sparger, Key equipment performance requirements are:

O (1> rumg liew (mintmum> 100,pm with s><h pumps ruuning (2) Maximurn reactor pressure 1250 psig (for injection)  ;

(3) Pumpable volume in storage 6100 U.S. gal '

tank (minimum) .

The required volume of r.olution contained in the storage tank is dependent upon the solution concentration, and this concentration can vary during reactor operatiors. A required boron solution volume /concentradon relationship is -

used tc - acceptable SLC System storage tank conditions'during plant operatin .

The SLC. System is automatically initiated during an ATWS. An AMS condition :

exists when either of the following occursi (1) , High RPV pressure (1125 psig) and Average Power Range Monitor (APRM) not down scale for 3 minutes, or-(2) Low RPV level (Level 2) and APRM not down scale for 3 minutes.

- 2.2.4 - 1, 6/1/92-h*%,F w- w et eey*twe-t g tw w wp'th- --r e t -y w --g q 74 e -p v ?* .gw.yvi gn e a g $f y $ gg- g Q P+ c ue e 'T N

I A3WR Design Docuraent When the SLC System is automatically initiated to inject a liquid neutron g absorber into the reactor, the following devices are actuated:

L)

(1) The two injection valves are opened.

(2) The two stonge tank dixharge valves nre opened.

(3) '1he two injection pumps are naned.

- (4) The reactor water cleanup isolation v:.lves are e.losed.

1 The SLC System can also be manually initiated from the main control room.

When it is manuallyinitiated to inject a liquid neutron absorber into the reactor, the following devices are actuated by each switch:

(1) One of the two irjection utves is opened.

1) One of the two storage tank discharge valves is opened.

p) One of the two injection pumps is started.

(4) One of the reactor water cleanup isolation vahrs is closed.

_ The SLC System provides borated water to the recctor core to compensate for V the various reactivity effects during the required conditions, These effecu include xenon decay, elimination of stea

.r. voids, changing water density due to the reduction in water temperature, Doppler effect in unmium, chcrges in nen2on leakage, and changes in control rod worth as boron affecta.eutron migration length. To meet this objecdve, i;is necessary to inject a quantity of i boron which produces a minimum concentration of 850 ppm of natural boron in the t eactor core at 70'F To allow for potential bakage and imperfect mb:ing in the rector sptem, an additional 25% (220) is added to the above requirement. The required concentration is thus acl-leved, accouating for dilution in the RPV with nonnal water level and including the volume in the P}IR shutdown cooling piping. This quantity of boron srolution is the amount which is above the pump suction shutofflevelin the tank, thus allowing for the~

portion of the tant . alume which cannot be injected.

The prmps are capable of producing discharge pressure to inject the solution into the reactor when the reactor is at high pressure conditions corresponding to the system relief vahr actuation (1560 psig), which is above peak ATWS '

1 pressure.

The SLC System includes sufIicient control room indication to allow for the necessary monitoring and control during design basis operational conditions.

(~]

Thir includes pump discharge pressure, storage tank liquid level and

< tempemture, as well as valve open/close and pump on/offindication for those 2- 6/1;92 2.2.4 f

ABWR oesiga vocument _

components shown on Figure 2.2.4 (with the exception of the simple check g vah es).

Q.)

The SLC System uses a dissolved solution of r, odium pemaborate as the neutron.

absorbing poison. This solution b held in a stomge tank which has a heater to maintain solution temperature above the saturation temperanire. The heater is capable of automatic operation and automatic shutoff to maintain an acceptable solution remperature. The SLC Sy; tern solution tank. a test water tank, the two positive displacement pumps, and associated valdng are all lccated in the secondary containment on the floor elevation below the operadng floor. This is a Seismic Category I structure, and the SLC System equipment is protected from phenomena such as earthquakes, tornados, hurricanes, and floods, as well as from internal postulated accident phenomena. In this area, thc SLC System is not subject to conditions such as missiles, pipe whip, and discharging fluids.

The pumps, heater, valves, and controls are powered from the standby power supply or normal offsite power. The pumps and valves are powered and

, controlled from separate buses and circuits so that single active failure will not prevent system operation. The power supplied to one motor. operated injection valve, storage tank discharge utive, and injection pump is powered from Division I,48 VAC. The power supply to the other motor <>perated i:jection valve, storage tank outlet valve, and injection pump is powered from Division II, l Q 480 VAC. The power supply to the tank heaters and heater controls is connectable to a standbfpower source. The standby power sowcc is Class IE from an on-site source and is independent of the off. site power.

Components of the SLC System which are required for injection of the neutron absorber into the reactor are classified Seismic Category 1. The major mechanical components are designed to meet ASME Code requirements as shown below:

ASME Design Conditions Component Code Class Pressure- Temperature Stcrage Tank 2 Static Head 1%'F Pump 2_ 1560 psig _ 150 7  ;

injection Valves 1 1560 psig 150'F Piping Inboard of 1 1250 psig 575'F Injection Valves q

'V 2.2.4 3 6/1/92

. ABWR Design occument; - .

'~

Piping and components not required for the injection of the neutron absorbe 1

-(e.g., test tank, samplinF system line, and storage tank vent) are cla>sified Non.

Nuclear Safety (NNS).

Design provisions to permit systern testing include a testi mk and associated piping and valves. The tank can be supplied with demineralind icater which can-be pumped in a closed loop tiuough either pump or injected into the~ reactor.-

The SLC. System is separated both physically and electrically from the Control -

Rod Drive System.

Inspections, Tests, Analyses and Acceptance Criterls ,

Table 2.2.4 provides a definition of the. inspections, tests, and/or analyses; together with associated acceptance criteria, which will be undertaken for the .

- SLC Fstem.

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= = = _ - - - - _

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U Table 2.2.4: Standby 1.lquid Control Systern u

inspections, Tests, Analyses amt Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Critoria Construction records, revisions and plant 1. It must be shown the SLC System can h 1. The minir.'um average poison 1.

- concentration in the reactor t.fter operation . visual examinations will be undertaken to - achieve a poison concentration of 850 ppm

[ assess as-built parameters listed below for or g.' ester, assuming a 25% dilution due to I' ' of the SLC System shall be equal to or

' compatibility with SLC System design non-uniform mixing in th y reacter and gre3rer than 850 ppm.

accounting for dilution in the RHR--

calculations. lf necessary, en as-built SLC

[- shutdown cooling systems. Thi; System analysis' will be conducted to demonstrate that the acceptance criteria concentration must be achieved under -

system design basis conditions.

are mat.

Critical Parameters: This requires that the SLC System meet the following values:

I

. a. - Storage tsnk pumpable volume L

a. Storage tank pumpeble volume range I'  : b. RPV water inventory at 70*F' 6100-6800 gal.

I- i' ,

c. U RHR shutdown cooling system water- b.' RPV water inventory $ 1.0C x 10616.

f~ Inventory at 70*F G

c. - RHR shutdown cooling system :  !

inventory s 0.287 x 100 lb . j

2. A simplified system configuration is shown. / 2. Inspections of installation records, together~ 2. The system configuration le iT sccordance l with plant we'kdowns,will be conducted to with Figure 2.2.4 in Figure 2.2.4.-

confirm that the installed equipment is in -- l compliance with the design configuration defined in Figure 2.2.4.

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'Ol O O Table 2.2.4: Standby Liquid Control System (Continued)

Inspections, Tests,' Analyses and Acceptance Criteria Certified Design Commitment ' Inspections, Tests, Analyses Acceptance Criteris '

' 3. ' The SLC System shall be capable of. 3. System preoperation tests will be 3. It must be shown that the SLC System can corducted to demonstrate acceptable automatically inject 100 gpm (both pumps :

' delivering 100 gpm c'so!ution with both pump and system performance.These ' . running) against a reactor pressure of 1250 :

. pumps operating against the elevated '

tests will involve establishing test psig with simulated ATWS conditions.it pressure conditions which con exist in the ' must also be shown that the SLC System reactor during events involving SLC conditions thrt simuiste conditions which will exist during an SLC System design - pumps can pump the entire storage ten 6

. System initiation.

basis event. To demonstrate' adequate Net pumpable volume.

- Positive Suction Head (NPSH), det{very of rated flow will be confirmed by tests I conducted at conditions of low leve! and maximum temperature in the storage tanic and the water will be injeded from the storage tank to ths RPV.-

t:

4. 4 The system is designed tobermit m.serwce 4. Field tests will be conducted after sys+em 4. Using normally installed controls, power il y>.

instellation to confirm that in-service supplies and other auxiliaries, the system ~

' functional testing of the SLC System.'

o has the capability to perform:

system testing can be performed.

a. Pump tests in s closed loop on the test ,

tenk.

i.

1 l b. RPV injection tests using domineralized water from the test tank.

5. .The pump, heater, valves and controla can - 5.1 System tests will be conducted after 5. The installed equipment een be powered L Installation to conf'rm that the electrical , from the standby AC power supply.

I- be powered from the ' standby AC powerJ supply es described in Sostion 2.2.4.

power supply configurations are in compliance with design commitmants.

6. See Generic Equipment Ouelification ' d

6. : SLC System components which are ' ~ 6. . See Generic Equipment Qualification verification ectivities (ITA). ' Acceptance Criteris (AC).'

r'equired for the injection of the neutron

~a bsorber into'the reactor are classified i L.

Seismic Category I and qualified for-3 2 L appropriate environment for locations -

8 ' where installed.

O O O N

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PRIMARY NNS CONTAINMENT STORAGE 2 f TANK TE A L

NNSl 2 L A SAMPUNG e _ y_g- L SYSTEM HEATER I I p_ .

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' INJECTION i I e VALVES . I I PtJMPS I-

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u Figure 2.2.4 Standby Liquid Control System (Standby Mode)

ABWR DESIGN CERTIFICA140N

$1,5/92 ACRS=SURCDMMITTEE' REVIEW:-

ABWR DESIGN CERTIfICATIDH

_ {) B/5/92 ACRS SUBCOMMITTEE REELEW RESIDUAL HEAT REMOVAL -: '

SUMMARY

OF TIER 1 ENTRIES-SYSTEM (RHR)

-?

l DESIGN-DESCRIPTION ENTRIES:

SAFETY-RELATED : SYSTEM (WITH SOME:'!ONSAFETY FUNCTIONS)

- SEISMIC: CATEGORY I; ASME CODE T sS 1 INSIDE PCPB,LCODE:

CLASS 2 OUTSIDE THREE COMPLETELY INDEPENDENT SUBSYSTEMS- '

MULTIPLE MODES OF OPERATION LOW PRESSURE CORE- FLOODING (AUTOKATICALLY INITIATED)'-

-SUPPRESSIDN POOL 1 COOLING (LONG-TERM' CONTAINMENT? COOLING)

-- SHUTDOWN COOLING : (DECAY ; HEAT 1 REMOVAL) =

0 --

WETWELL AND DRYWELL SPRAYS SUPPLEMENTAL 1 FUEL POOL:COOLINGf(AND: EMERGENCY MAXE-UP)

~

AC INDEPENDENT WATER! ADDITION '(FIRELWATER CROSS-TIEF SIMPLIFIED' SYSTEM DIAGRAMS (3)= ,

4 ITAAC ENTRIES:- v BASIC SYSTEM CONFIGURATIONJ

. SYSTEM 0PERATION IN ALL-MODES

- VALIDATION OF!ECCS AND CONTAINMENT-ANALYSES (E.s., PUMP-AND;

! HEAT EXCHANGER CAPABILITY)

INITIATION,J-ISOLATION AND1 INTERLOCK: LOGIC-(I JTC-it 8/5/92 I I

q 2

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a l ABWR Design Document _ L 2.4 Core Cooling O 24 ' 8esieu iwea1aemoveis vetem nesign Description .

The Residual Her.t Removal (RHR) System is comprised of tluce divisionally separate subsystems : hat pedorm a variety of functions utilizing th : following six basic modes of operation: (1) shutdown cooling, (2) :,uppression pool cooling, (3) wetwell and dr)well spray cooling, (4) low pressure cot e flooder (LPFL), (5) fuel pool cooling, and (6) AC independent water addkion. The configt ration of each locp is shown on its P&lD in Figure 2.4.1 (aligned in the standby mode).

The major functior.s of the various modes of operation include: (i) containment he:.t removal, (2) reauor decay heat removal, (3) emergency reactor vessellevel makeup and (4) augmented fuel pool cooling. In line with its given functions, j ponions of the system are a part of the ECCS network and the containment cooling system. Additionall), portions of the RHR System are considered a pan of the Reactor Coolant Pressure Boundary (RCPB).

c The entire RHR System is designed to safety-rebted standards, although it performs some non-safety functions (i.e., those that are not taken credit for when evaluating design basis accidents). The safety-related modes of operation include: (1) low pressure flooding, (2) suppression pool cooling, (3) vietwell b spray cooling and (4) shutdown cooling. Non-safety-related modes of operation include: (1)drywell spray cooling. (2} AC independent water additiot and (3) augmented fuel pcrol cooling. The RHR System also provides a backup, safety ,

related fuel pool makeup capability. Ancillary modes of operation include minimum flow bypass and full flow testing. (

The ECCS function of the RHR System is performed by the LPFL rnode, Following receipt of a LOCA signal ( low reactor water 'evel or high dr)vell pressure ), the RHR System automatically initiates and operates in the LPFL mode (in conjunction with the remainder of the ECCS network) to provide emergency makeup to the reactor vesselin order to keep the incter core cooled auch that the criteda ofl0 CFR 50A6 are met.The LPFL mode is accomplished by all thre: loops of the RHR System by transferring water from the suppression pool to the RW, via the RHR heat exchangers. Although the LPFL mode is automatically initiated , it may also be initiated manually. The system will alsc, automatically revert to the LPFL mode of operation from any other test or operating mode upon receipt of a LOCA signal. Each RHR loop's RPV injection udve requires a low reactor pressure permissive signal whether being opened manually or automatically in response to a LOCA signal.

(~) The containment heat removal function in the ABWR is performed by the V Containment Cooling System, which is comprised of the low pressure core ,

flooder (LPFL), suppression pool cooling, and wetwell and drywell spray cooling 2.at -1 C/16.W J

ABWR Design Document modes of the PJ1R System. Following a LOCA, the energy present within the p reactc : primary system is dumped either directly to the suppreuion pool via the V SRVs, or inciirectly via the dr>well and connecdng vents. Subsequendy, fksion product decay heat condnues to add energy to the pwl. The Containment Cooling System is designed to limit the long-term bulk tempenture of the suppression pel, and thus limis the long-term peak tempentures and pressures within the wetwell and devell regions of the containment to within their analyzed design limits, with only two of the three Ic, ops in operation (i.e., worse case single failure). The cooling requirements of the containment cooling function establish the necessary RHR heat exchanger heat removal capacity.

The LPFL mode,ia addition to its primary function of cooling the core, senes to cool the containment, as the heat exchanger is designed to always be in d:e loop. The dedicated suppression pool cooling mode is made available la cach of the three lyaps of the RHR System isy circulating suppression pool water through the respective RHR heat exchanger and then directly back to the suppression pool. This mode of RHR is usually initiated manually but will also initiate automatically in response to high suppression pool temperature. The wrtwell and drywell spmy medes of RHR ae each available in only two of the three subsystems (loops B and C). These functions are performed by drawing vwer from the suppression pool and delivering it to a common wetwell spray header and/or a cominon drywell spray header, both via the associated RHR t) heat excl auger (s). These containment spmy modes of the RHR System are typically initiated manually, with the exception of automatic initiation of wetwell spray coincident with automadc suppression pool cooling. However, the drywell spray inlet alves can only be opened if there exis:s high drywell presure and if the RFV injection valves are fully closed. Wetwell and drywell sprays serve as an augmented method of containment cooling. Weewell spray also serves to mitigate the consequences of steam bypassing the suppression pool.

The normal operational mode of the RHR System is in the shutdown cooling mode of opemtion, which is used to remove decay heat from the reactor core.

Tids raode provides the required safety-t elated capability needed to achieve and maintain a cold shutdow condition, including considention of the worst cav:

qstem single failure. The RUR heat exchanger heat remond capacity requirements in this mode are bounded by containment cooling requirements.

Shutdown cooling is initiated manually once the RPV has been depressurized below the system low preum e perminive. In this mode each loop taes suction from the RPV via its dedicated suction line, pumps the water through its respective heat exchanger, and returns the cr,oled water to the RPV. Two loops (B and C.) discharge water back to the RPV via dedicated spargers, while the l

third loop (A) utilizes the vessel spargers of one of the two feedwaterlines (FW-(,) A) . The heat removed in th RHR heat exchangers is uansputed to the ultimate -

heat ink via the respective division of reactor cooling nier and service water.

Each shutdown cooling suction valve is interlocked with that loop's suppression 2.4.1 2 6/15/92

ABWR oesign Document .

pool suction and discharge valves and wetwell spray valve to prevent draining of the reactor vessel to the suppression pool. Also, each shutdown cooling suedon

(,) salve is interlocked with, and automatically closes on, low reactor water level, nie augmented fuel pool cooling mode of the RHR System supplements /

replaces the normal fr.cl pool cooling system during infrequent conditiom of high neat load. This mode is accomplished manually in one nf two ways. When

> the reactor vessel head is removed, the cavity flooded and the fuel pool gates are removed, the RIIR System cools tle fuel pool in the normal shutdown cooling  ;

mode. When the fuel pool is oth erwise isolated from the reactor cavity, two loops l W and C) of the RHR System cm direcdy cool the pool by taking suction from sud dhcharging back to the normal fuel pool ec,aling system. This connection i also prosides fcr emergency fuel pool makeup capability by supplying a safety-related makeup path to the 'uel pool from a safety related source (i.e., the suppresion pool).

One loop (C) of the RER System also functions in an AC independent water addition mode. This mode provides a means of cross connecting the reactor buildmg fire protection system header to the RHR Systemjust outside the containment in the absence of the normal ECCS network and independent of the r.ormal essendal AC power distribution network. The connection is accomplished by manually op ming two incries valves on the cross <onnection p pipingjust upstream of its tie-in to the normal RHR piping. Fire protection

,V system water can be directed to either the RPV or the d: well spmy sparger by manuvl opening of the loop C RHR injection valve or the two loop C dgvell spray valves. These threc nives also have mano.d hand wheels. The Ore water is ,

supplied via the system's reactorbuilding distribution header by either the direct diesel <lriven fire pump or imm an external source udlizing 3 dedicated connectionjust outside the reactor building.

Each lonp of the RHR System clso has both a maimura flow mode and a full flow

, test mode. The minimum flow mode assures that there is pump flow suflicient to keep the pump cool by opening a minimum flow vahe that directs flow back to the s ippression pool anytime the pump is ruunine and the main discharge ulve is closed. Upon sensing that there is adequate Gowin the pump main discharge line, the minimum flow valve is automatically c!csed. In the full Dow test tnode, the system is essentially operated in the suppressioq>ool cooling mode, drawing suction from and discharging back to the suppression pool.

The RRR System is comprised of three s< parate loops or subsystems, each of-whicit includes a pump and a heat exchnoger, takes suction from either the RPV or the suppression pool, and directs water back to eitbei the IU'V or the n separation suppression pool. Two of the three loops can divert a Imion of the

!j suppression pool return flow to a common wetwell spray sparger or direct the entire flow to a common drywell spray sparger. The divisional subsystems of the f 2.4.1 2 wiw2

d ABWR oesign Document _

RHR Sptem are separated both mechanically and electrically, as well as being -

m physically located in different areas of the plant to address requirements R

- , pertaining to fire protection and other separation criteria. Each of the three subs estems is powered from a separate divisional power distribution bus that can l be supplied from either an on-site or off-site source. Cooling water to each q dhision of RHR equipment (he.it exchanger as well as pump and motor coolers) .j

! is supplied by the respective division of the reactor cooling water (RCW) Spiem.

The RHR Sprem aho includes provisions for containment isolation and RCPB ]

pressure isolation.

f The RHR System will maintain the capability to perform its intended safety--

related functions either fe,11owing a Safe Shutdown Earthquake (SSE) or during-l 2

the environmental conditions imposed by a LOCA, and in each case assuming the worst case single failure. The sptem will alsa accommodate calculated movement and thermal stresses. The sptem is designed so that the pumps will have necessary head / flow characteristics and available NPSH greater than required NPSH for operating modes. The system can be powered from either normal on'. site sources or by the emergency diesel generators. The RHR System is Scumic Category I and is housed in the Seismic Category I reactor building _to -

provide protection against tornadoes, floods, and other natural phenomena,

! The RHR pumps are motor <iriven centridigalpumps each capable of supplying at least 4200 gpm at 40 psid (dryweil to RPVE The pumps are ASME Code am lQ F 2 components with a design pressure of 500 psig and a design temperature of - ,

l 360*F. The pumps are interlocked from starting without' an open suction pathi j The RHR pumps are protected from possible pump runst conditions during

operation. The RRR heat exchangers are horizontal U-tube /shell type each

! sized to provide a minimum effective heat r:moval capacit y (K coeflicle'nt) of-195 Bru/sec'F. The primag end secondary sides of the heat exchangers are L ASME Code Class 2 and 3, respectively. The primary side design _ temperature =

l and pressure are 500 psig and 3604, respectively. The seconduy side design

( temperature and pressure are consistent with that of the RCW System. Each loop of the RHR Sptem has its ownjockey pump to act as a keep-fill system for that l' loop's pump discharge piping. Thejockey pumps are ASME Code Class 2.

~

The RHR System piping and valves are ASME Code Ciass 1 or 2 as shown on the.

P&ID (Figures 2.4.la, b, c). The cesign pressure and temperatury of piping anC y l l valves varies across the sptem. For that piping attached to the RPV, from the RPV out to and including the outboard containment isolation valw.a. the design

_ pressure and temperature are 1250 psig and 575TJrespectively. For other  ;

Ll l

piping open to thh containment atmosphere, out to and including the outboard containment isoladon valves, the design pressur: and temperature are 45 psig :

l and 219"F, respectively. For p; ping and valves outside the containment isolation -

)g'v valves, the design pressure and temperature depends on whetherit idocated on l - the suction or discha ge side of the main pump. Those portions on the suction 4 6/16/92

_2.4.1 i

._,., , . _, 2.. c,._ _ _ . _ . a-- ,. x , ,_ + a- ,

~ ABWR Design Document -

l n.

4

> side are rated at 300 psig and 360'F, while those portions on the discharge side; g' are rated at 500 psig and 360'F, respectively. The_ low pressure portions of the-shatdown cooling piping are prote:ted from full reactor pressure by autc matic :

pressure isolation valves that are interlocked with reactor pressure. Hight reliability of this interlock is assured by.udlizing four separate and divisionally

i independent pressure sensors in a 2eutef 4 logic. Additionally lin series:

inboard and outboard containment / pressure isolation valves in each loop are

[ powered from separate electrical divisions. Rchdvalves are also provided for

protection from overpressure.

[ The RHR System includes control room indication to allow for monitoring and _-

control during design . basis operational conditions, i.e., system flows, l

i temperatures and pressures, as well as valve open/close and pump on/off l indication for those instruments and components shown;on Figures 2.4.la, b : '

and c. with the exception of simple check nives and overpressure relief ulves *

(of the check valves shown only the testable check blves downstream of each

. loop's RPV injection valve has control room status indication). .

Inspections, Tests, Analyses and Acceptance Criteria  ;

This secuon provides a dermition~ of the inspections, tests and/or analyses .

[

L together with associated acceptance criteria which will be undertaken for the l_O RnR s rstem.

l-l' I

t' i

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. s c ., -. .- , _ - , _ , , + -m, , .. , , ,- , , - , -4,

. , -- ,,,m,- _ . , .. _ . . . . ~ . .c-.~ _~.

D .O o Table 2.4.1: Residual Hest Hemoval Syctem A..

Inspections, Tests, Analyses and Acceptance Criterin Acceptance Critada certified Design Commitment. Ir.epections, Tests Analyses Inspections of the as-built RHR 1. A :tual RHR System configuration, for The configuration of the BHR System is - 1.

.1.

configuration shall be pe. formed. . those compcnents shown, conforms with shown in Figures 2.4.ia, b and c, which are Figures 2.4.1a, b and *: and separation r each mechanically and electrically requirements.

separeted from each other.

2. The ECCS LOCA performance analysis for 2. RHR System actuation and operation is

' 2. The RHR System operates 'n the LPFL consistent with the ECCS performance

" mode'as part of the overalf ECCS network. masuring core cooling shall be validated by ..

RHR System functbriel testing, including ' analysis as follows: >

demonstration that the LPcL mode (of eech L a. RHR Flow (each loop)

RHR foop)is capeble of sutomaticalW 2 4200 gpm (et 40 psid)

~

!- Initieting and operating in response to a -

L .b. Time to Rated Flow (each loop)

. LOCA signet.

$ 36 coc .

i ..

3, The ps! mary containment performance - 3. ftHR System automatically actuatas !n the T T 3.1The RHR System .operates in the _.

? analysis for long-term peak pressure and . , ^ suppression poof cooling mode as; <

. suppression poci cooling mode to limit the . designed and RHR heet exchanger

, temperature shall be validated by RHR l

long iterm temperature snd pressure of the - performance is consistent with the containment under post LOCA conditions.? ' System functional testing demonstr? ting conteinment cooling system analysis as -

i the required flowrote through the heet i'

' ( exchenyrr and by inspection of vendor test ~ fotfows:-

dets demonetrating the heat exchanger's L .' effective best removal capability.L ' a. Effective heet remoal capability of esch C '-

Automet".c initiation in the suppression  : RHR Heat Exchanger (K coefficient) includes' effects of RCW, RSW and UHS:

fpool coolin0 mode will also tx* '

~I

p. 2195 Bttvsec*F.

demonstrated.'

b. Tube side flow of each RNR Heat

!' Exchanger-2 4200 gpm

4. ' RHit icops B and C cach separt.tefy are - 1

-4. I RHR System functional tests shall be j

4. A portion of the RHR System return f

ow (in .  ! capaba of pro Ading wetwell sprey flow L

? performed to demcnstrt te wetwell epray loops _ B & C) can be diverted to the wetwell J L r.onsistent with the suppression bool '.

[ i spray header.

f!ow capabl?ity.

bypass snelysis as follows: -

L

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. a. - Wetwell spray flow.(each foop individually) 2 500 gpm.

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Oi O O GENERIO LETTER 83-28, SUPPLEMENT 1 REQUIREDLACTIONS: BASED ON GENERIC ilMPLICATIONS ;OF SALEM ATWS EVENTS

PRESENTATIONLBEFORE THE-L ADVISORYJ. COMMITTEE ON REACTOR SAFEGUARDSL LBY:

lCARLLH.:BERLINGER, CHIEFj

~

(GENERIC! COMMUNICATIONS BRANCH-

, DIVISION OF OPERATIONAL EVENTS ASSESSMENT

L(301):504-28371

~

~

. ' AUGUST .6,1992 '

I b

c

GENERIC LETTER $28, SUPPLEMENT 1

• GL 83-28 REQUESTED LICENSEES TO IMPLEMENT LONG-TERM g

CORRECTIVE ACTIONS IN' RESPONSE TO SALEM ATWS EVENT.

• MAJOR ACT!ONS INCLUDED:

• AUTO-ACTUATION OF SHUNT TRIP ATTACHMENT FOR ALL

. TRIPS.-.

• PREVENTIVETMAINTENANCE AND SURVEILLANCE TESTING PROGRAMS.

• ACTIONf4.2.3-- PERFORM' LIFE TESTING

• ~ ACTION.4.2.4 - PERIODIC REPLACEMENT BASED ON

LIFE EXPECTANCY.

• LlCENSEES:DID NOT FULLY IMPLEMENT LIFE TESTING AND PERIODIC!REPLACEMENTiPROGRAMS;. CONTINUING TEST,

! MAINTENANCE AND TRENDING' PROGRAMS: CONSIDERED SUFFICIENT.

= ? REVIEW OF OPERATIONAL EXPERIENCE DATA:FROM 1986 l TilROUGHTEARLY:J1991.(NPRDS AND LERs):

• ONELRTBJFAILURECTO-OPEN.ON DEMAND,_MCGUIRE

' UNIT 2,fJULY 1987, CRACKED WELD, MANUFACTURING

PROBLEM..

• - lRECENT EVENT, PALO VERDE, UNIT 3, MARCH 1992, a

MAINTENANCE.

L O a o-L

. GENERIC LETTER 87-28, SUPPLEMENT 1 j i (continued) j r i 1

• STAFF HAS CONCLUDED REGARDING RTB HELIABILITY: j ll -

• ACTIONS' COMPLETED. HAVE 'BEEN EFFECTIVE .(RTB L  : RELIABILITY HAS'GONE FROM 0.01 - 0.001 TO j

-0.0001 - 0.00001 FAILURES PER DEMAND).

+

• CDF'FOR ATWS EVENTSLNOT EFFECTED FOR RTB- '

'FAILUREL RATES LESS;THAN 0.001 FAILURES l PER DEMAND. THEREFORE, .FURTHER ~ ACTIONS i TO1 ADDRESS END-OF-LIFE DEGRADATIONilN-

'RTB RELIABILITY:ARELNOT JUSTIFIED..

• i MODIFICATIONS' RESULTING: FROM - IMPLEMENTATION l L OF 10 CFR 50.62 FURTHER REDUCED RISK FROM .

[ RTB' FAILURES.

i a

p

• THIS-GL St!PPLEMENT PR'OPOSES TO REMOVE THE:

[  ; REQUIREMENT FOR LIFE TESTING '(4.2.3) AND:

PERIODIC REPLACEMENT (4.2.4) BASED ON- '

EXPECTED LIFETIMES.-

t j '. .

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,. . .. -.. ..~.~ - . . .. . - . . . . . . _ - - -

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g,'ygegfpe hP; Lid UMMlWii gg [g[ff[

l y}yl,jjj""63'66 gfgg v/g S09 avMsm . #Rx O W JL 29 A8:27 /

Union au1y 23,1992 Etscraic n

U.S. Nuclear Regulatory Commission _

ATTN: Chief, Rules and Directives Review Branch Washington, D.C. 20555

Dear Sir:

ULNRC-267 0 PROPOSED SUPPLEMENT l'TO GENERIC LETTER 83-2B

References:

1.-57FR29540-57FR29541 dated July 2,.1992

2. ULNRC-1678 dated November 13, 1987 Union Electric offers the following comments-on the proposed Supplement 1 to Generic Letter (GL) 83-28,_'as noticed in the Federal Register per Reference _1.

While Union Electric commends the proposed closure of items 4.2.3 and 4.2.4~of.GLL83-28,,which would close all GL 83-28' actions for Callaway,-the staff's conclusion should be expanded. The current ~ conclusion, " . . , the staff O co=c1uee .ta t 11ce= ee ectio== i= re vo==e to ite== 4 2 3 and 4.2.4 of GL 83-28 are not necessary.", does not reflect

?

the WOG-sponsored 3 4 ?e c :le testing documented in WCAP-10835. The staff's conclusion should mention that, in light of improved-RTB reliability, no further licensee actions'are required. The background section of the supplement'should acknowledge'the testing documented in WCAP-10835 and.

describe the issue behind the staff's' original reluctance.to accept this WCAP, i.c. the lack of thermal-aging prior to the life cycle testing. The_ arguments presented in Reference 2 have been accepted.somewhat by~the staff, ini view of the prope, sed supplement's reference _to: licensee's contentions ". . .. that further;11 meting of--the RTBs is not necessary because of their exterm a a quality assurance, preventive maintenance. and surveillance-resting-programe.";

however, this background is not' discussed:in the supplement.

With these clarifications addressed, Union-Electric strongly endorses issuance of-this supplement.

Union Electric-appreciates:the. opportunity to

~

comment on this subject. If you'have any. questions on_the above, please contact us.

Very truly-yours, DJ AL,&

Alan C. Passwater' Manager - Licensing and-Fuels-GGY/kea r .- e .r .%4-- -,-c e ,

. . ~. .

4

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- 67MMMo

^

O 79/9p-

g

- \

l August 2, 1992 j COMMENTS OF OHIO CITIZENS FOR RESPONSIELE ENERGY, INC. ("OCRE")  ;

ON " PROPOSED GENERIC COMMUNICATIO!1 SUPPLEMENT 1 TO GENERIC LETTER 83-28, ' REQUIRED ACTIONS-BASED OH GENERIC IMPLICATIONS OF SALEd ATWS EVENTS'", 57 FED. REG. 2954C (JULY 2.1992)

This proposed supplement to Generic Letter 83-28 would inform licensees that actions in response to. items 4.2,3 (life testing) and 4.2,4 (periodic replacement of breakers or components) are-not required.

OCRE -opposes the issuance of this_ supplement. The -operating-experience _ with reactor trip breakers outlined in; the Federal "egister- notice does not support relaxation of the_G9neric _Let-ter*r requirements,- The fact that.RTBs:have failed at all- indi-cates a continuing need for the requirements. Even if the Staff.

considers this operating. experience to be indicative'of improving performance, it does not necessarily' follow that relaxation .of-'

l v the requirements is warranted. Perhaps performance _is improving _

because of the very requirements to be deleted.

The NRC should not be so eager.to-relax regulatory. requirements without clear and convincing evidence that relaxation is neces-sary and desirable.

Respectfully submitted, Susan L. Hiatt e A

Director, OCRE 8275 Munson Road Mentor, OH 44060-2406 g "s '

(216) 255-3158' E :% gg

- &. ,a i

'~ g g

~

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(^) 9 l-l l i

, m ,.c. , , , . , . . , - .-.y,,-. , , , ,

.-m.,

F

('h V

i COMMENTS TO ACRS ON RTB DPO 06 AUG 92

1. By Generic Letter 83-26 Supplement 1 the staff intends to remove the " requirement" that PWR licensees lifetest (LT) Reactor Trip Breaker (RTB). The basis given is that the maintenance and surveillance implemented after issue of Generic Letter 83-28 has sufficiently improved RTB reliability so that the LT is no longer needed.

2. The basis for the supposed improvement is that only one or two failures to open were found in an NPRDS sort covering 21'000 tests. The staff concludes that the RTB reliability is better than .99999. Because of deficiencies in the NPRDS sort, I think that the RTB reliability may be an order of magnitude lower, i.e.

.9999, which is quite good enough.

3. But even if this point value for the RTB reliability is

['Ji correct from 1986 to 1992 will it be so indefinitely?

w

4. After July 1996, all structures, systems, and components, subject to 10 CFR 50.65, will be maintained according to procedures licenseas will determine. Will those procedures be such as to maintain the present apparent reliability?

5. If all the failure modes of RTB have already appeared in the NPRDS record and if they do not increase in frequency as the breakers age, then even allowing another order of magnitude increase in the RTB failure rate to account for the unknown variance of the RTB failure rate distribution, the RTB may still be reliable enough.

6. There are in the NPRDS record many more slow trips than failures to open. These have been discovered in monthly testing.

It is then reasonable to suppose that breaker degradation has been sufficiently slow cs to permit many failures to open to have been intercepted in test. But, as the breakers age, new failure modes may be encountered which do not announce themselves in time for the maintenance procedures then in effect to interrupt their appearance in service.

m N

l

F

(] Comments to ACRS -

2 -

(s/

7. None of these gambles with future RTB reliability need to be taken if the ACRS can persuade the staff to reconsider the life test requirement they propose to rescind. The wearout life distribution would be determined and any failure modes not yet met with would have been discovered and a replacement policy could have been implemented.

8. In discussions with the ACRS subcommittee the question was

' asked, WPat would you get from the LT that you don't already get from the current maintenance and surveillance? Current maintenance and surveillance is apparently preventing many failures to open. But for reasons given above and in my 04 August 92 comments to the ACRS Subcommittee it is impossible to tell the age of RTB on failure,to trend failures, or in any way to anticipate failures by timely part or breaker replacement; or to compare licensee's maintenance. And the connection between the effects of licensee maintenance and the only accessible record the NPRDS is uncertain and variable.

9. The construction of a reliability data base is a difficult occupation. It would be marvelous if we had lucked out in that

() the NPRDS were such. But until recently the staff has been told that was not the purpose of the NPRDS. The NPRDS is unsatisfactory as a reliability data base in that the entries are exceedingly variable; many different people maintain, repair, diagnose, and report the failures of breakers of undeterminable age, with unrecorded histories. Failures and repairs may be entered but if any breakers are replaced without a test failure there is no requirement for them to be recorded. In short the possibility of so many less than incidental but unknown  ;

characteristics can affect the NPRDS that only in the absence of a properly designed, operated, and controlled data base should it be resorted to. Future modifications to the collection of breaker failures may also affect its value to the NRC.

The statistical innocence of the staff has been remarked upon by the ACRS. It is nowhere more evident than in.the recent employment of the NPRDS to determine the RTB failure rates.

CHARLES MORRIS NRR/ DST /SELB r^s

O O O FINAL REGULATORY GUIDE 1.101 2

.i EMERGENCY PLANNING AND PREPAREDNESS FOR NUCLEAR REACTORS r

! ACRS

> - AUGUST 6, 1992 l

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I i' . APP. 1 OF NUREG 0654 PROVIDED EXAMPLE INITIATING  !

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. OF EXPERIENCE. j e t i * ' DEVELOPED BY NUMARC WITH SIGNIFICANT INPUT FROM L I .NRC. FEMA WAS ALSO' INVOLVED. 1

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THRESHOLD-VALUES lFOR;BWR AND/OR PWR~ FACILITIES.  ;

p ,  ;

t.

-= NUMARC CTRARLY- DEFINES " FISSION. PRODUCT BARRIER LOSS" I

AND " POTENTIAL." LOSS".

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i.

i- .

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. _ , ,, _ ,,,, . - .- . .,, ,. .~._______m__._ _ . _ _ _ _ - . . _ _ _ _ _ _ _ . _ _ . ._

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.O O o j!

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IMPORTANT ASPECTS OF NUMARC EAL GUIDELINES

! (CONTINUED) j

*: EAL THRESHOLDS UTILIZE MANY OF THE SAME OBSERVABLE

! AND QUANTIFIABLE PARAMETERS NOW USED IN PLANT EOP'S L.-

L i -

TEMPERATURES PRESSURES

~ VESSEL LEVELS I ,

INJECTION.. FLOW RATES I -

iTEDWATER' FLOW RATES

j. -

SUBCOOLING MARGIN

[ -

CONTAINMENT TEMPERATURE  ;

i CONTAINMENT PRESSURE CONTAINMENT RADIATION ]

ISOLATION SYSTEM STATUS i I -

ACTIVITY / RADIATION LEVELS -

I CRITICAL SAFETY FUNCTION STATUS j ..

-* . CLEARLYJ LIMITS "DEIAY TIMES" FOR THE - RECOGNITION OF F

FAILED MITIGATION EFFORTS.

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}

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= .,

O O O NUMARC EAL SCHEMS f

AN EVENT-BASED EAL CLASSIFICATION SYSTEM g INCORPORATING A FISSION PRODUCT BARRIER CHALLENGE / BREACH SCHEME.

PREFIX A _AENORMAL RADIOLOGICAL CONDITIONS PREFIX H _ HAZARDOUS CONDITIONS-r PREFIX-S_---S_YSTEM MALhW CTIONS i

!- PREFIX F FISSION PRODUCT BARRIER CHALLENGES / BREACHES t

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O O o

, f i

EXAMPLES  :

v i

I' AU -- UNUSUAL EVENT BASED ON ABNORMAL j RADIOLOGICAL CONDITIONS i

1 HU'-- UNUSUAL EVENT BASED'ON HAZARDOUS  :

CONDITIONS ,

1 SU -- UNUSUAL EVENT BASED ON SYSTEM n MALFUNCTIONS

! AA -- ALERT BASED ON ABNORMAL RADIOLOGICAL CONDITIONS  ;

l, HA -- ALERT BASED ON HAZARDOUS CONDITIONS i- FA'-- ALERT BASED ON FISSION PRODUCT BARRIER CHALLENGES / BREACHES i SS -- SITE AREA EMERGENCY' BASED ON SYSTEM MALFUNCTIONS

~

i

!- :AS -- SITE AREA EMERGENCY BASED ON ABNORMAL I RADIOLOGICAL CONDITIONS SG - GENERAL EMERGENCY BASED ON SYSTEM MALFUNCTIONS-f HG -- GENERAL EMERGENCY BASED ON HAZARDOUS .

CONDITIONS I FG -- GENERAL EMERGENCY BASED ON FISSION t PRODUCT BARRIER FAILURES t t.

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O O O l

CONCLUSION l

! TO EXPEDITE PUBLICATION OF FINAL REGULATORY GUIDE BY-SEPTEMBER 1, 1992 I

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_ _ - = . .

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